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| issue date = 09/28/2006 | | issue date = 09/28/2006 | ||
| title = Rev. 28 to Off-Site Dose Calculation Manual (OCDM) | | title = Rev. 28 to Off-Site Dose Calculation Manual (OCDM) | ||
| author name = Fiorenza T, Hutton J | | author name = Fiorenza T, Hutton J, Kurtz T, Schimmel M, Stinson G | ||
| author affiliation = Constellation Energy Group | | author affiliation = Constellation Energy Group | ||
| addressee name = | | addressee name = | ||
| Line 18: | Line 18: | ||
==SUMMARY== | ==SUMMARY== | ||
OF REVISIONS Revision 28 (EFfective September | OF REVISIONS Revision 28 (EFfective September | ||
: 29. 2006)PAGE 1 3.3-13,14 i13.3-6 1 4.0-1 II 2-10,26,33-36,66,67,75,80 ix, I 1.0-1, I 1.,02, IB 3.3-2,14.1-1 | : 29. 2006)PAGE 1 3.3-13,14 i13.3-6 1 4.0-1 II 2-10,26,33-36,66,67,75,80 ix, I 1.0-1, I 1.,02, IB 3.3-2,14.1-1 | ||
& !a, It, 11, I15ý .129,.Il 63, II 107, 11 108 I 3.3-9 1 3.3-10 1 3.3-7, 1 3.3-12, and 1 3.3-13 II 63, II 64, and 11107 II 3 and ii 4 iv, 1 1.6-1, 1-3.1-7, 13.2-3, 1 3.2-10, 13.2-,121I3.3A-1, 13.3-2,13.3-3, I 3.3-7, .3.3-8, 1 3.3w9, 1 3.3-10, 1 B 3.1-3, 1 B 3.2-5, 1 B 3.2-6, l B 3.3-1,, I B 3.3-2,1 4.1-la, 11 10,11113,1120, and 1123 DATE August 2000 November 2000 November 2000 November 2000 December 2001 December. | & !a, It, 11, I15ý .129,.Il 63, II 107, 11 108 I 3.3-9 1 3.3-10 1 3.3-7, 1 3.3-12, and 1 3.3-13 II 63, II 64, and 11107 II 3 and ii 4 iv, 1 1.6-1, 1-3.1-7, 13.2-3, 1 3.2-10, 13.2-,121I3.3A-1, 13.3-2,13.3-3, I 3.3-7, .3.3-8, 1 3.3w9, 1 3.3-10, 1 B 3.1-3, 1 B 3.2-5, 1 B 3.2-6, l B 3.3-1,, I B 3.3-2,1 4.1-la, 11 10,11113,1120, and 1123 DATE August 2000 November 2000 November 2000 November 2000 December 2001 December. | ||
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TABLE OF CONTENTS (Cont)B 3.2 B 3.2.1 B 3.2.2 B 3.2.3 B 3.2.4 B 3.2.5 B 3.2.6 B 33 B 3.3.1 B 3.3.2 Radioactive Gaseous Effluents Gaseous Effluents DoseRate Gaseous Effluents Noble Gas Dose*Gaseous Effluents Dose -Iodine-13)1, Iodine- 133, Tritium, and Radioactive Material in Particulate Form Gaseous Radwaste Treatment System Ventilation Exhaust Treatment System Venting or Purging instrumentation Radioactive Liquid Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring InStrumentation Radioactive Effluents Total Dose Radiological Environmental Moriitoring Monitoring Program Land Use Census Interlaboratory | TABLE OF CONTENTS (Cont)B 3.2 B 3.2.1 B 3.2.2 B 3.2.3 B 3.2.4 B 3.2.5 B 3.2.6 B 33 B 3.3.1 B 3.3.2 Radioactive Gaseous Effluents Gaseous Effluents DoseRate Gaseous Effluents Noble Gas Dose*Gaseous Effluents Dose -Iodine-13)1, Iodine- 133, Tritium, and Radioactive Material in Particulate Form Gaseous Radwaste Treatment System Ventilation Exhaust Treatment System Venting or Purging instrumentation Radioactive Liquid Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring InStrumentation Radioactive Effluents Total Dose Radiological Environmental Moriitoring Monitoring Program Land Use Census Interlaboratory | ||
'Comparison Prograni PAGE I B 3.2-1 I.B 3.2-1 I B 3-2-2 I B 3.2-3[ B 3 .2-5 I B 3.2-6 I. B3.2-7 I B 3.3-1 I B3.3-1 I B 3.3-2 lB 3.4-1 I B 3-5-1 1B 3.5-1, I B-3.5-2 I B 3.5-3 B3.4 B 3.5 B 3.5.1 B 3.5,2 B 3.5.3 SECTION 4.0 ADMINISTRATIVE CONTROLS 14.0-1 D 4.1 D4.1.1 D 4.2 Reporting. | 'Comparison Prograni PAGE I B 3.2-1 I.B 3.2-1 I B 3-2-2 I B 3.2-3[ B 3 .2-5 I B 3.2-6 I. B3.2-7 I B 3.3-1 I B3.3-1 I B 3.3-2 lB 3.4-1 I B 3-5-1 1B 3.5-1, I B-3.5-2 I B 3.5-3 B3.4 B 3.5 B 3.5.1 B 3.5,2 B 3.5.3 SECTION 4.0 ADMINISTRATIVE CONTROLS 14.0-1 D 4.1 D4.1.1 D 4.2 Reporting. | ||
Requh'ements Special Reports Major Changes to Liquid, Gaseous and Solid Radwaste Treatment Systems 14.1-1 1 4. 1.'4 14.2-1 Unit 2 Revision 28 September 2006 iii TABLE OF CONTENTS (Cont).IF7CTI"N ,'lJR R1?FVCTION PA GE IZUBJEXT REFSECTION PART II -CALCULATIONAL METHODOLOGIES | Requh'ements Special Reports Major Changes to Liquid, Gaseous and Solid Radwaste Treatment Systems 14.1-1 1 4. 1.'4 14.2-1 Unit 2 Revision 28 September 2006 iii TABLE OF CONTENTS (Cont).IF7CTI"N ,'lJR R1?FVCTION PA GE IZUBJEXT REFSECTION PART II -CALCULATIONAL METHODOLOGIES 1.0 LIQUID EFFLUENTS I .1 Liquid Effluent Monitor Alarm Setpoints 1 A.11 Basis 1.1.2 Setpoint Determination Methodology 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint 1 .1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculations 1.1.2..3 Service Water and Cooling Tower Bi0wdown Effluent Radiation Alarm Setpoint 3.1.1 3.3.1 II i 112 112 112 112 115 116:1.2 1.3 1.4 1.5 2.0 Liquid Effluent ConcentrationCalculaijon Liquid Effluent Dose CalculationMethodology Liquid Effluent Sampling Representativeness Liquid Radwaste System FUNCTIONALITY 3.1.1 DSR 3.1.1.2 3.1.2 DSR 3.1.2.1 Table D 3. A -1 note b 3.1.3 DSR 3.1.3.1 B 3.1.3 II 7 II 8 119 II 10 1112 11'12 11 12 1113 11 14 GiASEOUS EFFLUENTS 2.1 211,1.2.1,2 1.1.2 i 2..1.2.1 2.1.2.2 2.1.2.3 Gaseous Effluent Monitor Alarm Setpoints Basis 3.2..1 Setpoirit Deierminatiori MethodologY Discussion 3.3.2 Stack Noble Gas.Detector Alarm. Setpoint Equation Vent Noble Gas Detector Alarm Setpoint Equation Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation 2.2 2.2.1 2.2.2 2.2.3 Gaseous Effluent Dose. Rate Calculation Methodology X/Q and W, -Dispersion Parameters for Dose Rate, Table D 3-23 Whole Body Dose RateDue to Noble Gases Skin Dose Rate Due to: Noble Gases 3.2.1 DLCO3.2.! .a DSR 3.2 .i.1 DLCO 3.2.1.a DSR 3.2.1.1 1115 11 16 IIF16'1117 ilia iv Unit 2 Revision 28 September 2006 TABLE OF CONTENTS (Cont)SECTION SUBJECT REF SECTION PAGE 2.2.4 2,3 2-3.1 2.3.2 2.3.3 23.4 Organ Dose RateDue to 1-131,1-133,Tritium and DLCO 3.2.L.b Particulates with half-lives greater than 8 days DSR 3.2.1.2 Gaseous Effluent.Dose Calculation Methodology 3.2.2 3.2.3 3.2.5 W, and WV,:- Dispersion Parameters For Dose, Table D 3.-23 Gamma.Air Dose Due to Noble Gases 3,2.2 DSR 3.2.2.1 Beta Air Dose Due to Noble Gases 3.3.2 Organ Dose Due to 1- 31, 1-133, Tritium and Particulates 3.2.3 with half-lives 312.5 DSR 3.2.3.1 DSR 3.2.5.1 I-133 and I-135 Estimation Isokinetic Sampling Use of Concurrent Meteorological Data vs. Historical Data Gaseous Radwaste Treatment System Operation 3.2.4 Ventilation Exhaust.Treatment System Operation 3.2.5 2.4 2.5.2.6 2.7 2.8 3.0 3.1 3.2 3.3, 3.4 4.0, 4.1 4.2 4.31 1119 1I 20 II 20 1121, II 21 1i 21 II 22 11 22 II 22 HI 22 II 23 1124 I125 II 26 11 27 1I 27 11 30 II 30 11 30 1131 URANIUM FUEL CYCLE 3.4 Evaluation of Doses From Liquid Effluents DSR 3.1.2.1 Evaluation of Doses.From Gaseous Effluents DSR 3.2.2.1 Evaluation of Doses From Direct Radiation DSR 3.2.3.1 Doses to Members of the Public Within the Site Boundary 4.1 ENVIRONMENTAL MONITORING PROGRAM 3.5 Sampling Stations 3.5.1 DSR 3.5.1.1 Interlaboratory Comparison Program DSR 3.5.3.2 Capabilities for Thermoluminescent Dosimeters -used for Environmental Measurements V Unit 2 Revision 28 September 2006 TABLE OF CONTENTS (Cont)SECTION SUBJECT REF SECTION PAGE Appendix A Liquid Dose Factor Derivation Appendix B Appendix C Appendix D Plume Shine Dose, Factor Derivation Dose'Parameters.for Iodine 131 and 133, Particulates and Tritium Diagrams of Liquid and Gaseous Radwaste Treatment Systems. and Monitoring Systems 11.66 II 69 1173 II 83 11106 Appendix E Nine Mile Point On-Site and Off-Site Maps vi Unit 2 Revision 28 September 2006' LIST OF TABLES PART I -RADIOLOGICAL EFFL UENT CONTROLS TABLE NO TITLE PAGE.D 3.1.1-1 Radioactive Liquid Waste Sampling and Analysis 1 3.1-2 D 3.2.1-1 Radioactive Gaseous Waste Sampling and Analysis 1 3.2-2 D 3.3.1 -1 Radioactive Liquid Effluent Monitoring Instrumentation 13.3-6 D 3.3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation 1 3.3-13 D 3.5.1-1 Radiological Environmental Monitoring Program 1 3.5-6 D 3.5,. 1-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples I 3.5-10 D 3.5.1-3 Detection Capabilities for Environmental Sample Analyses, I 3.5-11 vii Unit 2 Revision 28 September 2006 LISTOF TARLES (Cont)PARTH -CALCULATIONAL METHODOLOGIES TABLE NO D 2-1 D 2-2 thru D 2-5 D3-1 D 3-2 D3-3 D 3-4 thru D 3-22 D 3-23 D 3-24 D 5.1 TITLE Liquid Effluent Detector Response Aiat Values -Liquid Effluent Dose Factor Offgas PrctMatmient Detector Response Finite Plume -Ground Level Dose Factorsfrom an Elevated Release Immersion Dose Factors Dose And Dose Rate Factors, Ri Dispersion Parameters at Controlling Locations,. | ||
EFFLUENTS I .1 Liquid Effluent Monitor Alarm Setpoints 1 A.11 Basis 1.1.2 Setpoint Determination Methodology 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint 1 .1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculations 1.1.2..3 Service Water and Cooling Tower Bi0wdown Effluent Radiation Alarm Setpoint 3.1.1 3.3.1 II i 112 112 112 112 115 116:1.2 1.3 1.4 1.5 2.0 Liquid Effluent ConcentrationCalculaijon Liquid Effluent Dose CalculationMethodology Liquid Effluent Sampling Representativeness Liquid Radwaste System FUNCTIONALITY 3.1.1 DSR 3.1.1.2 3.1.2 DSR 3.1.2.1 Table D 3. A -1 note b 3.1.3 DSR 3.1.3.1 B 3.1.3 II 7 II 8 119 II 10 1112 11'12 11 12 1113 11 14 GiASEOUS EFFLUENTS 2.1 211,1.2.1,2 1.1.2 i 2..1.2.1 2.1.2.2 2.1.2.3 Gaseous Effluent Monitor Alarm Setpoints Basis 3.2..1 Setpoirit Deierminatiori MethodologY Discussion | |||
Noble Gas.Detector Alarm. Setpoint Equation Vent Noble Gas Detector Alarm Setpoint Equation Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation 2.2 2.2.1 2.2.2 2.2.3 Gaseous Effluent Dose. Rate Calculation Methodology X/Q and W, -Dispersion Parameters for Dose Rate, Table D 3-23 Whole Body Dose RateDue to Noble Gases Skin Dose Rate Due to: Noble Gases 3.2.1 DLCO3.2.! .a DSR 3.2 .i.1 DLCO 3.2.1.a DSR 3.2.1.1 1115 11 16 IIF16'1117 ilia iv Unit 2 Revision 28 September 2006 TABLE OF CONTENTS (Cont)SECTION SUBJECT REF SECTION PAGE 2.2.4 2,3 2-3.1 2.3.2 2.3.3 23.4 Organ Dose RateDue to 1-131,1-133,Tritium and DLCO 3.2.L.b Particulates with half-lives greater than 8 days DSR 3.2.1.2 Gaseous Effluent.Dose Calculation Methodology 3.2.2 3.2.3 3.2.5 W, and WV,:- Dispersion Parameters For Dose, Table D 3.-23 Gamma.Air Dose Due to Noble Gases 3,2.2 DSR 3.2.2.1 Beta Air Dose Due to Noble Gases 3.3.2 Organ Dose Due to 1- 31, 1-133, Tritium and Particulates 3.2.3 with half-lives 312.5 DSR 3.2.3.1 DSR 3.2.5.1 I-133 and I-135 Estimation Isokinetic Sampling Use of Concurrent Meteorological Data vs. Historical Data Gaseous Radwaste Treatment System Operation | |||
Exhaust.Treatment System Operation 3.2.5 2.4 2.5.2.6 2.7 2.8 3.0 3.1 3.2 3.3, 3.4 4.0, 4.1 4.2 4.31 1119 1I 20 II 20 1121, II 21 1i 21 II 22 11 22 II 22 HI 22 II 23 1124 I125 II 26 11 27 1I 27 11 30 II 30 11 30 1131 URANIUM FUEL CYCLE 3.4 Evaluation of Doses From Liquid Effluents DSR 3.1.2.1 Evaluation of Doses.From Gaseous Effluents DSR 3.2.2.1 Evaluation of Doses From Direct Radiation DSR 3.2.3.1 Doses to Members of the Public Within the Site Boundary 4.1 ENVIRONMENTAL MONITORING PROGRAM 3.5 Sampling Stations 3.5.1 DSR 3.5.1.1 Interlaboratory Comparison Program DSR 3.5.3.2 Capabilities for Thermoluminescent Dosimeters -used for Environmental Measurements V Unit 2 Revision 28 September 2006 TABLE OF CONTENTS (Cont)SECTION SUBJECT REF SECTION PAGE Appendix A Liquid Dose Factor Derivation Appendix B Appendix C Appendix D Plume Shine Dose, Factor Derivation Dose'Parameters.for Iodine 131 and 133, Particulates and Tritium Diagrams of Liquid and Gaseous Radwaste Treatment Systems. and Monitoring Systems 11.66 II 69 1173 II 83 11106 Appendix E Nine Mile Point On-Site and Off-Site Maps vi Unit 2 Revision 28 September 2006' LIST OF TABLES PART I -RADIOLOGICAL EFFL UENT CONTROLS TABLE NO TITLE PAGE.D 3.1.1-1 Radioactive Liquid Waste Sampling and Analysis 1 3.1-2 D 3.2.1-1 Radioactive Gaseous Waste Sampling and Analysis 1 3.2-2 D 3.3.1 -1 Radioactive Liquid Effluent Monitoring Instrumentation 13.3-6 D 3.3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation 1 3.3-13 D 3.5.1-1 Radiological Environmental Monitoring Program 1 3.5-6 D 3.5,. 1-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples I 3.5-10 D 3.5.1-3 Detection Capabilities for Environmental Sample Analyses, I 3.5-11 vii Unit 2 Revision 28 September 2006 LISTOF TARLES (Cont)PARTH -CALCULATIONAL METHODOLOGIES TABLE NO D 2-1 D 2-2 thru D 2-5 D3-1 D 3-2 D3-3 D 3-4 thru D 3-22 D 3-23 D 3-24 D 5.1 TITLE Liquid Effluent Detector Response Aiat Values -Liquid Effluent Dose Factor Offgas PrctMatmient Detector Response Finite Plume -Ground Level Dose Factorsfrom an Elevated Release Immersion Dose Factors Dose And Dose Rate Factors, Ri Dispersion Parameters at Controlling Locations,. | |||
X/Q, W, and W, Values Parameters"Fof the Evaluation of Doses to Real Members of the Public From Gaseous ,And Liquid Effluents Radiological Environmental Monitoring Program Sampling Locations PAGE II 33.1 34 II 38 11.39 1140 1141 II 60 Ii161 11 62 VIII Unit 2 Revision, 28 September 2006 LIST OF FIGURES FIGURE NO D 1.0-1 D 5.1-1 D 5.1-2 D5.1i2 TITLE Site Area and Land Portion of Exclusion Area Boundaries Nine Mile Point On-Site Map Nine Mile Point Off-Site Map (page I of 2)Nine Mile Point Off-Site Map (page 2 .of 2)PAGE 1 1.0-4 11107 11108 11 109 ix Unit 2 Revision 28 September 2006 INTRODUCTION The OFFSITE DOSE CALCULATION MANUAL (ODCM) is a supporting document of the Technical Specifications Section 5.5.1. The previous Limiting Conditions for Operation thatwere contained in the Radiological Effluent Technical Specifications are now transferred to the ODCM as Radiological Effluent Controls. | X/Q, W, and W, Values Parameters"Fof the Evaluation of Doses to Real Members of the Public From Gaseous ,And Liquid Effluents Radiological Environmental Monitoring Program Sampling Locations PAGE II 33.1 34 II 38 11.39 1140 1141 II 60 Ii161 11 62 VIII Unit 2 Revision, 28 September 2006 LIST OF FIGURES FIGURE NO D 1.0-1 D 5.1-1 D 5.1-2 D5.1i2 TITLE Site Area and Land Portion of Exclusion Area Boundaries Nine Mile Point On-Site Map Nine Mile Point Off-Site Map (page I of 2)Nine Mile Point Off-Site Map (page 2 .of 2)PAGE 1 1.0-4 11107 11108 11 109 ix Unit 2 Revision 28 September 2006 INTRODUCTION The OFFSITE DOSE CALCULATION MANUAL (ODCM) is a supporting document of the Technical Specifications Section 5.5.1. The previous Limiting Conditions for Operation thatwere contained in the Radiological Effluent Technical Specifications are now transferred to the ODCM as Radiological Effluent Controls. | ||
The ODCM contains two parts: Radiological Effluent Controls, Part I; and Calculational Methodologies, Part I. Radiological Effluent Controls, Part 1,includes the following: | The ODCM contains two parts: Radiological Effluent Controls, Part I; and Calculational Methodologies, Part I. Radiological Effluent Controls, Part 1,includes the following: | ||
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H-3 Gross Alpha St-89 Proportional, Composite of grab samples Wd)Proportional Composite of-grab samples (d)Each batch(g) 31 days: Each batch (g) 92. days 2. Continuous Release&a. Service Water Effluent A b. Service Water Effluent B c. Cooling Thwer Blowdown 3. Coniinuous. | H-3 Gross Alpha St-89 Proportional, Composite of grab samples Wd)Proportional Composite of-grab samples (d)Each batch(g) 31 days: Each batch (g) 92. days 2. Continuous Release&a. Service Water Effluent A b. Service Water Effluent B c. Cooling Thwer Blowdown 3. Coniinuous. | ||
Release Auxiliaýr Boiler Pump Seal and Saniple Cooling Discharge (Service Water)I Grab Sample 31 day,;(e)Grab Sample 31 days (e)Grab Sample 31 days,(e):Sr-90 Fe-55 31 days (e) Principal Gamma'Emitters (c)31 days (e) 1-131 3 I days (e) Dissolved and Entrained Gases (gamma emitters)31 days(es) 11-3 3t days (e) Gross Alpha 92 days (e> Sr-89 92 days (e> Sr-0 92 days (e) :FeZ55 31 days (f. :PrincipaI Gamma Emitters (c)92 days (1) 1t-3 SAMPLELOWER LIMIT OF.DETECTION (LLD) (a).5, x101 ICi/ml I x 10-.Ci/ml I x to- 5 LCi/ml I x 104 VaCi/mi I x I0"7 pCi/ml ,5 x 10-' pCi/ml 5 x 10"' pCi/ml I xl 0- pCi/mI:5 X, 1-"7 11Ci/Mt I x 10`5 pCi/ml I x 10-' PCi/nml I x 1l0 'pCilml 1 x 10*8 PCi/mI x t O.7 paCi/mI 5 x I& pci/ml I x 10-7 PCi/mI I x 10-.Ci/mI Grab Sample Grab Sample Grab Sample Grab Sample Grab Sample Grab Sample 31 days (6)31I days,,(c)32 days (e)92 days (e).92 days (e)3 1. days (f)Grab Sample 92 days (f)Unit 2 Revision 28 September 2006 1.3.1-2 Liquid Effluents Concentration D 3.1.1 Table D 3.1.1-1 (Page 2 of 2)Radioactive Liquid. Waste Sampling and. Analysis (a) ITLe LED is defiied as the smallest conceiitration of radioactiveeii-naterial in a sampnle that will yielda net count, above system backgound, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation representsa'*r"nal" signsl.For a particular menesureinnt sy.stem, which may include radichemical separation:. | Release Auxiliaýr Boiler Pump Seal and Saniple Cooling Discharge (Service Water)I Grab Sample 31 day,;(e)Grab Sample 31 days (e)Grab Sample 31 days,(e):Sr-90 Fe-55 31 days (e) Principal Gamma'Emitters (c)31 days (e) 1-131 3 I days (e) Dissolved and Entrained Gases (gamma emitters)31 days(es) 11-3 3t days (e) Gross Alpha 92 days (e> Sr-89 92 days (e> Sr-0 92 days (e) :FeZ55 31 days (f. :PrincipaI Gamma Emitters (c)92 days (1) 1t-3 SAMPLELOWER LIMIT OF.DETECTION (LLD) (a).5, x101 ICi/ml I x 10-.Ci/ml I x to- 5 LCi/ml I x 104 VaCi/mi I x I0"7 pCi/ml ,5 x 10-' pCi/ml 5 x 10"' pCi/ml I xl 0- pCi/mI:5 X, 1-"7 11Ci/Mt I x 10`5 pCi/ml I x 10-' PCi/nml I x 1l0 'pCilml 1 x 10*8 PCi/mI x t O.7 paCi/mI 5 x I& pci/ml I x 10-7 PCi/mI I x 10-.Ci/mI Grab Sample Grab Sample Grab Sample Grab Sample Grab Sample Grab Sample 31 days (6)31I days,,(c)32 days (e)92 days (e).92 days (e)3 1. days (f)Grab Sample 92 days (f)Unit 2 Revision 28 September 2006 1.3.1-2 Liquid Effluents Concentration D 3.1.1 Table D 3.1.1-1 (Page 2 of 2)Radioactive Liquid. Waste Sampling and. Analysis (a) ITLe LED is defiied as the smallest conceiitration of radioactiveeii-naterial in a sampnle that will yielda net count, above system backgound, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation representsa'*r"nal" signsl.For a particular menesureinnt sy.stem, which may include radichemical separation:. | ||
LLD (4.66) (S, )(E)(QV) (222x1 06) {Y) e-where: LLD) The before-the-fact.lower limit of detection | LLD (4.66) (S, )(E)(QV) (222x1 06) {Y) e-where: LLD) The before-the-fact.lower limit of detection | ||
(ýCi per unit mass or volume), Sb= The standard deviation of the background counting rate or of t.he counting.rate of aiblank sample as appropriate (counts per miiiiute),. | (ýCi per unit mass or volume), Sb= The standard deviation of the background counting rate or of t.he counting.rate of aiblank sample as appropriate (counts per miiiiute),. | ||
E -The counting efficiency (counts per disintegration), V. The sample size (units of mass or volume), 2,22 x 06 The number of disintegrations per minute per jiCi, Y = The fractional radiochemical.yield, when applicable, radioactive decay constant forihe particularradionuclide (see), and At = The elapsed time betweenthemidpoint of siampie collec-tiofi and the'tim of counting (seconds). | E -The counting efficiency (counts per disintegration), V. The sample size (units of mass or volume), 2,22 x 06 The number of disintegrations per minute per jiCi, Y = The fractional radiochemical.yield, when applicable, radioactive decay constant forihe particularradionuclide (see), and At = The elapsed time betweenthemidpoint of siampie collec-tiofi and the'tim of counting (seconds). | ||
| Line 99: | Line 89: | ||
system shall bbe FUNCTIONAL. | system shall bbe FUNCTIONAL. | ||
APPLICABILITY: | APPLICABILITY: | ||
At. all -times.ACTIONS-------------------------------- | At. all -times.ACTIONS-------------------------------- | ||
: w. .N OTES -- -..... ..................---------- | : w. .N OTES -- -..... ..................---------- | ||
: 1. LCO 3.0.3 is not applicable.. | : 1. LCO 3.0.3 is not applicable.. | ||
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... ..--- .. .-_ .. .. .. .. --- -- .. .... ..-.- ..... ...CONDITION REQUIRED ACTION COMPLETION TIME A. The gaseous radwaste from the main condenser air ejector system is being discharged without treatment. | ... ..--- .. .-_ .. .. .. .. --- -- .. .... ..-.- ..... ...CONDITION REQUIRED ACTION COMPLETION TIME A. The gaseous radwaste from the main condenser air ejector system is being discharged without treatment. | ||
B. Required Action and associated Completion Time'not met.A. 1 :Restore treatment of gaseous radwaste effluent.7 days 30 days B. I Prepare and submit to the. NRC, pursuant to D 4, 1., a Special Report that includes the following: | B. Required Action and associated Completion Time'not met.A. 1 :Restore treatment of gaseous radwaste effluent.7 days 30 days B. I Prepare and submit to the. NRC, pursuant to D 4, 1., a Special Report that includes the following: | ||
(1) Identification of any nonfunctional equipment | (1) Identification of any nonfunctional equipment | ||
:or subsystems and the reason for, the,.nonfunctionality, (2) Action(s) taken to restore the nonfunctional cquipment to FUNCTIONAL status, and (3) Summary description of action(s) taken to prevent a recurrence. | :or subsystems and the reason for, the,.nonfunctionality, (2) Action(s) taken to restore the nonfunctional cquipment to FUNCTIONAL status, and (3) Summary description of action(s) taken to prevent a recurrence. | ||
Unit 2 Revision 28 September 2006 ,13.2-10 Gaseous Radwaste Treatment System D 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.4.1 Check the readings of the relevant instruments to 12 hours ensure that the GASEOUS RADWASTE TREATMENT SYSTEM is: functioning. | Unit 2 Revision 28 September 2006 ,13.2-10 Gaseous Radwaste Treatment System D 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.4.1 Check the readings of the relevant instruments to 12 hours ensure that the GASEOUS RADWASTE TREATMENT SYSTEM is: functioning. | ||
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Composite sample. (0) Gamma isotopic overa ofic'month analysis of each sample period (i) (g) once per month I. sample Site's dowinstreamr 6ooling (2) H-3 analysis of water intake (h) each composite sample and once. per 3 months b. Ground As required From oneof two-soturces if Grab samplef uie (t) Gamma isotopic likely to be atflcted () per 3 months analysis of each sample.(g) once per 3 months (2) H-3 analysis of each sample ohce per 3'months (continued) | Composite sample. (0) Gamma isotopic overa ofic'month analysis of each sample period (i) (g) once per month I. sample Site's dowinstreamr 6ooling (2) H-3 analysis of water intake (h) each composite sample and once. per 3 months b. Ground As required From oneof two-soturces if Grab samplef uie (t) Gamma isotopic likely to be atflcted () per 3 months analysis of each sample.(g) once per 3 months (2) H-3 analysis of each sample ohce per 3'months (continued) | ||
Unit 2 Revision 28, September 2006 13.5-6 Radiological Envirotmentai Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 2 of 4)Radiological Environmental MonitorinIg Program EXPOSURE PATHWAY SAMPLING AND AND)ORý NUMBER OF SAMPLE COLLECTION TYPE ANDFREQUENCY SAMPLE SAMPLES LOCATIONS.(a) | Unit 2 Revision 28, September 2006 13.5-6 Radiological Envirotmentai Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 2 of 4)Radiological Environmental MonitorinIg Program EXPOSURE PATHWAY SAMPLING AND AND)ORý NUMBER OF SAMPLE COLLECTION TYPE ANDFREQUENCY SAMPLE SAMPLES LOCATIONS.(a) | ||
FREQUENCY OF ANALYSIS 3. Waterbome (continued),c, Drinking .1 sanmpiel Ieach, One to three oflthe nearest. When analysis (1) 1-131 analysis on watersupplies that could be is performed, a each composite sample.affctcd bhy is di ccree (kM composite sample when the dose over aitWo week, calculated for the period (i); otherwise, consumption of the a composite sample water is greater than I monthly. mremlyr (I)*(2) Gross beta and gamma isotopic, analyses of each composite sample (g)monthly (3) l-34asialysis of each composite qample d. Sediment I sample From a downstream area with Twice per year once per 3 months from existing or potential recreational Shoreline value Gamrma isotopic analysis of each sample(g)4. Ingestion a, Milk (1) 3 samples fromn l[n3 locations.withirt | FREQUENCY OF ANALYSIS 3. Waterbome (continued),c, Drinking .1 sanmpiel Ieach, One to three oflthe nearest. When analysis (1) 1-131 analysis on watersupplies that could be is performed, a each composite sample.affctcd bhy is di ccree (kM composite sample when the dose over aitWo week, calculated for the period (i); otherwise, consumption of the a composite sample water is greater than I monthly. mremlyr (I)*(2) Gross beta and gamma isotopic, analyses of each composite sample (g)monthly (3) l-34asialysis of each composite qample d. Sediment I sample From a downstream area with Twice per year once per 3 months from existing or potential recreational Shoreline value Gamrma isotopic analysis of each sample(g)4. Ingestion a, Milk (1) 3 samples fromn l[n3 locations.withirt 3.5 mites Twice per month, (I) Gamma isotopic MILK (e) Api thruuh. (g) wid 1-131 analysis of SAMPLING Dec-ember (m) each sample twice per LOCATIONS month April through December (2),If there alr In each of3'aie-vis 3.550 mile.s (2) Gamma isotopic none;, distant (e) (g) and 1-13 Ianalysis of then I sample each sample once per from MILK month January through SAMPLING March if required LOCATIONS At a control location 9-20 miles (3) :1, sampie from a distant and in a leasi prevalent MILK wind direction (d)SAMPLING LOCATION b, Fish (1) 1 sample~each In the,vicihity of a plant Twice per year Gamma isotopic analysis of of discharge areta each sample (g) on edible 2 commereially portions wice per. year or recreationally important species (n)(2) 1 sample of the In areas not influenc.d by.wame spcc ies station dischaegc (d)(continued) | ||
Twice per month, (I) Gamma isotopic MILK (e) Api thruuh. (g) wid 1-131 analysis of SAMPLING Dec-ember (m) each sample twice per LOCATIONS month April through December (2),If there alr In each of3'aie-vis 3.550 mile.s (2) Gamma isotopic none;, distant (e) (g) and 1-13 Ianalysis of then I sample each sample once per from MILK month January through SAMPLING March if required LOCATIONS At a control location 9-20 miles (3) :1, sampie from a distant and in a leasi prevalent MILK wind direction (d)SAMPLING LOCATION b, Fish (1) 1 sample~each In the,vicihity of a plant Twice per year Gamma isotopic analysis of of discharge areta each sample (g) on edible 2 commereially portions wice per. year or recreationally important species (n)(2) 1 sample of the In areas not influenc.d by.wame spcc ies station dischaegc (d)(continued) | |||
Unit 2 Revision 28 September 2006, 1 3.5-7 Radiological Environmental Monitoring Program D 3.5.:1 Table D 3.5.1-1 (page 3 of 4)Radiological Environmental Monitoring Program EXPOSURE PATHWAY SAMPLING AND TYPE AND FREQUENCY AND/OR NUMBER OF SAMPLE COLLECTION OF ANALYSIS SAMP E, SAMPLES LOCATIONS (a): FREQUENCY 4. Ingestion (continued) | Unit 2 Revision 28 September 2006, 1 3.5-7 Radiological Environmental Monitoring Program D 3.5.:1 Table D 3.5.1-1 (page 3 of 4)Radiological Environmental Monitoring Program EXPOSURE PATHWAY SAMPLING AND TYPE AND FREQUENCY AND/OR NUMBER OF SAMPLE COLLECTION OF ANALYSIS SAMP E, SAMPLES LOCATIONS (a): FREQUENCY 4. Ingestion (continued) | ||
: c. Food (1) 1 sample of Any area that is. iirigated by Attimne ofharvest Gamma isotopic (g) and I-Products each principal water in which liquid plant (p) 131 analysis of each sample class af food wastes have been discharged (o) of edible portions pioduota (2) Samples of 3 Grown nearest to each of 2 different kinds different offsite locations (e) Once per.year during of broad leaf the harvest season-vegetation (such as vegesables) | : c. Food (1) 1 sample of Any area that is. iirigated by Attimne ofharvest Gamma isotopic (g) and I-Products each principal water in which liquid plant (p) 131 analysis of each sample class af food wastes have been discharged (o) of edible portions pioduota (2) Samples of 3 Grown nearest to each of 2 different kinds different offsite locations (e) Once per.year during of broad leaf the harvest season-vegetation (such as vegesables) | ||
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up to four reactors, it is highly unlikely that the resultant. | up to four reactors, it is highly unlikely that the resultant. | ||
dose tooa MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside:storage tanks, etc., are kept small. The Special Report will describe a course of actionthat should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within ihe 40 ýCFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC fromother uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the -same site or within a radius of 5 miles must be considered. | dose tooa MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside:storage tanks, etc., are kept small. The Special Report will describe a course of actionthat should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within ihe 40 ýCFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC fromother uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the -same site or within a radius of 5 miles must be considered. | ||
The dose calculation methodology | The dose calculation methodology | ||
:and parameters for:calculating the doses from the actual release rates' of radioactive noble in gaseous: effluents are consistent with the methodology provided in RG 1.109,"Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and RG 1. i 11, "Methods fr'Estimating Atmospheric Transport.and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1,"' July 1977. The ODCM equations'provided for determining the air doses at or beyond the.SITE BOUNDARY are based upon real-time meteorological conditions or the historical average atmospheric conditions. | :and parameters for:calculating the doses from the actual release rates' of radioactive noble in gaseous: effluents are consistent with the methodology provided in RG 1.109,"Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and RG 1. i 11, "Methods fr'Estimating Atmospheric Transport.and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1,"' July 1977. The ODCM equations'provided for determining the air doses at or beyond the.SITE BOUNDARY are based upon real-time meteorological conditions or the historical average atmospheric conditions. | ||
This applies to the release of radioactive.material in gaseous effluents from each unit at the site.Unit 2 Revision 28 I B 3.2-2 September 2006 Gaseous Effluents Dose -Iodine-i 31, Iodine-i133, Tritium, and Radioactive Material In Particulate Form B 3.2.3 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.3 Gaseous Effluents:Dose | This applies to the release of radioactive.material in gaseous effluents from each unit at the site.Unit 2 Revision 28 I B 3.2-2 September 2006 Gaseous Effluents Dose -Iodine-i 31, Iodine-i133, Tritium, and Radioactive Material In Particulate Form B 3.2.3 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.3 Gaseous Effluents:Dose | ||
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Licensees may choose to subnmit this information as part of the annual FSAR update, Licensee-initiated majorchanges to the radwaste treatment systems (liquid, gaseous, and solid): a. Shall be reported to the Commission in the Radioactive Effluent Releasereport for the period in which the evaluation was reviewed by the SORCG The discussion of each changes shall contain: 1. A summary of the-evaluation thatled to the determination that the change::could be ,made in accordance with 10 CFR:50.59. | Licensees may choose to subnmit this information as part of the annual FSAR update, Licensee-initiated majorchanges to the radwaste treatment systems (liquid, gaseous, and solid): a. Shall be reported to the Commission in the Radioactive Effluent Releasereport for the period in which the evaluation was reviewed by the SORCG The discussion of each changes shall contain: 1. A summary of the-evaluation thatled to the determination that the change::could be ,made in accordance with 10 CFR:50.59. | ||
: 2. Sufficient detailed information to totally support the reason for thechangc without benefit of additional or supplemental informiation; | : 2. Sufficient detailed information to totally support the reason for thechangc without benefit of additional or supplemental informiation; | ||
: 3. A detailed description of the equipment, components, andprocesses involved alnd the interfaces'with other plant systems;4. An evaluation of the change, which shows the predicted | : 3. A detailed description of the equipment, components, andprocesses involved alnd the interfaces'with other plant systems;4. An evaluation of the change, which shows the predicted | ||
:releases of radioactive materials: | :releases of radioactive materials: | ||
in liquid :and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;5'. An ev'aluation of the change, which shows the expected maximum exposures -to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimatedin the license application and amendments thereto;:6. A comparison of the predicted releases of radioactive materials, in liquid and: gaseous, effluents and in. solid waste, to the actual releases for theperiod that precedes the time when the: change is to be made;7. An estimate of the to plant operating personnel as a resultof the change; and (Continued) | in liquid :and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;5'. An ev'aluation of the change, which shows the expected maximum exposures -to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimatedin the license application and amendments thereto;:6. A comparison of the predicted releases of radioactive materials, in liquid and: gaseous, effluents and in. solid waste, to the actual releases for theperiod that precedes the time when the: change is to be made;7. An estimate of the to plant operating personnel as a resultof the change; and (Continued) | ||
Unit 2 Revision 28 1 4.2-1 September 2006 Major Changes to Liquid, Gaseous, and Solid, Radwaste Treatment System D 4.2 D 4.2 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEM (continued) 8'. Documentation of the fact that the change was reviewed and found acceptable by the SORC.b. Shall become effective upon review and acceptance by the SORC.Unit 2 Revision 28 September 2006 i 4.2ý2 PART .I- CALCULATLONAL METHODOLOGIES Unit 2 Revision 28 September | Unit 2 Revision 28 1 4.2-1 September 2006 Major Changes to Liquid, Gaseous, and Solid, Radwaste Treatment System D 4.2 D 4.2 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEM (continued) 8'. Documentation of the fact that the change was reviewed and found acceptable by the SORC.b. Shall become effective upon review and acceptance by the SORC.Unit 2 Revision 28 September 2006 i 4.2ý2 PART .I- CALCULATLONAL METHODOLOGIES Unit 2 Revision 28 September | ||
:2006 I.1, | :2006 I.1, 1.0 LIQUID EFFLUENTS Service Water A and B, Cooling Tower Blowdown and the Liquid Radioactive Waste Discharges comprise the Radioactive Liquid Effluents at Unit 2. Presently there are no temporary outdoor tanks containing radioactive water capable of affecting the nearest known or future water supply in an unrestricted area. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.1.1 Liquid Effluent Monitor Alarm Setpoints 1.1.1 Basis The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) 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 nobles gases, the concentration shall be limited to 2E-04 uCi/ml total activity.1.1.2 Setpoint Determination Methodology 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint The Liquid Radioactive Waste System Tanks are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. At the end of the discharge tunnel in Lake Ontario, a diffuser structure has been installed. | ||
EFFLUENTS Service Water A and B, Cooling Tower Blowdown and the Liquid Radioactive Waste Discharges comprise the Radioactive Liquid Effluents at Unit 2. Presently there are no temporary outdoor tanks containing radioactive water capable of affecting the nearest known or future water supply in an unrestricted area. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.1.1 Liquid Effluent Monitor Alarm Setpoints 1.1.1 Basis The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) 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 nobles gases, the concentration shall be limited to 2E-04 uCi/ml total activity.1.1.2 Setpoint Determination Methodology 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint The Liquid Radioactive Waste System Tanks are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. At the end of the discharge tunnel in Lake Ontario, a diffuser structure has been installed. | |||
Its purpose is to maintain surface water temperatures low enough to meet thermal pollution limits. However, it also assists in the near field dilution of any activity released. | Its purpose is to maintain surface water temperatures low enough to meet thermal pollution limits. However, it also assists in the near field dilution of any activity released. | ||
Service Water and the Cooling Tower Blowdown are also pumped to the discharge tunnel and will provide dilution. | Service Water and the Cooling Tower Blowdown are also pumped to the discharge tunnel and will provide dilution. | ||
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Service Water and Cooling Tower Blowdown Alarm Setpoint Equation: Alarm Setpoint< 0.8 1/CF Yi C+/-/ [Yi(Ci/MECJ)] | Service Water and Cooling Tower Blowdown Alarm Setpoint Equation: Alarm Setpoint< 0.8 1/CF Yi C+/-/ [Yi(Ci/MECJ)] | ||
+/- Background. | +/- Background. | ||
Where: Alarm Setpoint 0.8 Ci CFi MECi Background Yi (Ci/CFi)Yi (Ci/MECi)(1/CF) YiCi CF-The Radiation Detector Alarm Setpoint, cpm= Safety Factor, unitless= Concentration of isotope i in potential contaminated stream, jiCi/ml= Detector response for isotope i, net pCi/ml/cpm See Table 2-1 for a list of nominal values= Maximum Effluent Concentration, ten times the effluent concentration limit for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, pCi/ml= Detector response when sample chamber is filled with nonradioactive water, cpm= The total detector response when exposed to the concentration of nuclides in the potential contaminant, cpm The total fraction of ten times the 1OCFR20, Appendix B, Table 2, Column 2 limit that is in the potential contaminated stream, unitless.= An approximation to Yj (Ci/CFi), determined at each calibration of the effluent monitor= Monitor Conversion Factor, p[Ci/ml/cpm | Where: Alarm Setpoint 0.8 Ci CFi MECi Background Yi (Ci/CFi)Yi (Ci/MECi)(1/CF) YiCi CF-The Radiation Detector Alarm Setpoint, cpm= Safety Factor, unitless= Concentration of isotope i in potential contaminated stream, jiCi/ml= Detector response for isotope i, net pCi/ml/cpm See Table 2-1 for a list of nominal values= Maximum Effluent Concentration, ten times the effluent concentration limit for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, pCi/ml= Detector response when sample chamber is filled with nonradioactive water, cpm= The total detector response when exposed to the concentration of nuclides in the potential contaminant, cpm The total fraction of ten times the 1OCFR20, Appendix B, Table 2, Column 2 limit that is in the potential contaminated stream, unitless.= An approximation to Yj (Ci/CFi), determined at each calibration of the effluent monitor= Monitor Conversion Factor, p[Ci/ml/cpm 1.2 Liquid Effluent Concentration Calculation This calculation documents compliance with Section D 3.1.1 of Part 1: Unit 2 Revision 28 II 7 September 2006 As required by Technical Specification 5.5.4, "Radioactive Effluent Controls Program," the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) 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 microcurie/ml total activity.The concentration of radioactivity from Liquid Radwaste, Service Water A and B and the Cooling Tower Blowdown are included in the calculation. | ||
Effluent Concentration Calculation This calculation documents compliance with Section D 3.1.1 of Part 1: Unit 2 Revision 28 II 7 September 2006 As required by Technical Specification 5.5.4, "Radioactive Effluent Controls Program," the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) 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 microcurie/ml total activity.The concentration of radioactivity from Liquid Radwaste, Service Water A and B and the Cooling Tower Blowdown are included in the calculation. | |||
The calculation is performed for a specific period of time. No credit is taken for averaging. | The calculation is performed for a specific period of time. No credit is taken for averaging. | ||
The limiting concentration is calculated as follows: FMEC S Ys [Fs/y, (F.) Yi (Cis+MECi) | The limiting concentration is calculated as follows: FMEC S Ys [Fs/y, (F.) Yi (Cis+MECi) | ||
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This alarm will give indication of incomplete mixing with adequate margin before exceeding ten times the effluent concentration. | This alarm will give indication of incomplete mixing with adequate margin before exceeding ten times the effluent concentration. | ||
Service Water A and B and the Cooling Tower Blowdown are sampled from the radiation monitor on each respective stream. These monitors continuously withdraw a sample and pump it back to the effluent stream. The length of tubing between the continuously flowing sample and the sample spigot contains less than 200 ml which is adequately purged by requiring a purge of at least 1 liter when grabbing a sample.1.5 Liquid Radwaste System FUNCTIONALITY The Liquid Radwaste Treatment System shall be FUNCTIONAL and used when projected doses due to liquid radwaste effluents would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period. Cumulative doses will be determined at least once per 31 days (as indicated in Section 1.3) and doses will also be projected if the radwaste treatment systems are not being fully utilized.The system collection tanks are processed as follows: 1) Low Conductivity (Waste Collector): | Service Water A and B and the Cooling Tower Blowdown are sampled from the radiation monitor on each respective stream. These monitors continuously withdraw a sample and pump it back to the effluent stream. The length of tubing between the continuously flowing sample and the sample spigot contains less than 200 ml which is adequately purged by requiring a purge of at least 1 liter when grabbing a sample.1.5 Liquid Radwaste System FUNCTIONALITY The Liquid Radwaste Treatment System shall be FUNCTIONAL and used when projected doses due to liquid radwaste effluents would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period. Cumulative doses will be determined at least once per 31 days (as indicated in Section 1.3) and doses will also be projected if the radwaste treatment systems are not being fully utilized.The system collection tanks are processed as follows: 1) Low Conductivity (Waste Collector): | ||
Radwaste Filter and Radwaste Demineralizer or the Thermex System.2) High Conductivity (Floor Drains): Regenerant Evaporator or the Thermex System.Unit 2 Revision 28 1110 September 2006 | Radwaste Filter and Radwaste Demineralizer or the Thermex System.2) High Conductivity (Floor Drains): Regenerant Evaporator or the Thermex System.Unit 2 Revision 28 1110 September 2006 | ||
: 3) Regenerant Waste: If resin regeneration is used at NMP-2; the waste will be processed through the regenerant evaporator or Thermex System.The dose projection indicated above will be performed in accordance with the methodology of Section 1.3.Unit 2 Revision 28 II 11 September 2006 | : 3) Regenerant Waste: If resin regeneration is used at NMP-2; the waste will be processed through the regenerant evaporator or Thermex System.The dose projection indicated above will be performed in accordance with the methodology of Section 1.3.Unit 2 Revision 28 II 11 September 2006 2.0 GASEOUS EFFLUENTS The gaseous effluent release points are the stack and the combined Radwaste/Reactor Building vent. The stack effluent point includes Turbine Building ventilation, main condenser offgas (after charcoal bed holdup), and Standby Gas Treatment System exhaust. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.2.1 Gaseous Effluent Monitor Alarm Setpoints 2.1.1 Basis The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following in accordance with Technical Specification 5.5.4.g: a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mremlyr to the skin, and b. For iodine- 131, for iodine- 133, for tritium, and for all radionuclides with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.The radioactivity rate of noble gases measured downstream of the recombiner shall be limited to less than or equal to 350,000 microcuries/second during offgas system operation in accordance with Technical Specification 3.7.4.2.1.2 Setpoint Determination Methodology Discussion Nine Mile Point Unit 1 and the James A FitzPatrick nuclear plants occupy the same site as Nine Mile Point Unit 2. Because of the independence of these plants' safety systems, control rooms and operating staffs it is assumed that simultaneous accidents are not likely to occur at the different units. However, there are two release points at Unit 2. It is assumed that if an accident were to occur at Unit 2 that both release points could be involved.The alarm setpoint for Gaseous Effluent Noble Gas Monitors are based on a dose rate limit of 500 mRem/yr to the Whole Body. Since there are two release points at Unit 2, the dose rate limit of 500 mRemlyr is divided equally for each release point, but may be apportioned otherwise, if required. | ||
EFFLUENTS The gaseous effluent release points are the stack and the combined Radwaste/Reactor Building vent. The stack effluent point includes Turbine Building ventilation, main condenser offgas (after charcoal bed holdup), and Standby Gas Treatment System exhaust. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.2.1 Gaseous Effluent Monitor Alarm Setpoints 2.1.1 Basis The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following in accordance with Technical Specification 5.5.4.g: a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mremlyr to the skin, and b. For iodine- 131, for iodine- 133, for tritium, and for all radionuclides with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.The radioactivity rate of noble gases measured downstream of the recombiner shall be limited to less than or equal to 350,000 microcuries/second during offgas system operation in accordance with Technical Specification 3.7.4.2.1.2 Setpoint Determination Methodology Discussion Nine Mile Point Unit 1 and the James A FitzPatrick nuclear plants occupy the same site as Nine Mile Point Unit 2. Because of the independence of these plants' safety systems, control rooms and operating staffs it is assumed that simultaneous accidents are not likely to occur at the different units. However, there are two release points at Unit 2. It is assumed that if an accident were to occur at Unit 2 that both release points could be involved.The alarm setpoint for Gaseous Effluent Noble Gas Monitors are based on a dose rate limit of 500 mRem/yr to the Whole Body. Since there are two release points at Unit 2, the dose rate limit of 500 mRemlyr is divided equally for each release point, but may be apportioned otherwise, if required. | |||
These monitors are sensitive to only noble gases.Because of this it is considered impractical to base their alarm setpoints on organ dose rates due to iodines or particulates. | These monitors are sensitive to only noble gases.Because of this it is considered impractical to base their alarm setpoints on organ dose rates due to iodines or particulates. | ||
Additionally skin dose rate is never significantly greater than the whole body dose rate. Thus the factor R which is the basis for the alarm setpoint calculation is nominally taken as equal to 250 mRem/yr. If there are significant releases from any gaseous release point on the site (>25 mRem/yr) for an extended period of time then the setpoint will be recalculated with an appropriately smaller value for R.The high alarm setpoint for the Offgas Noble Gas monitor is based on a limit of 350,000 uCi/sec. This is the release rate for which a FSAR accident analysis was completed. | Additionally skin dose rate is never significantly greater than the whole body dose rate. Thus the factor R which is the basis for the alarm setpoint calculation is nominally taken as equal to 250 mRem/yr. If there are significant releases from any gaseous release point on the site (>25 mRem/yr) for an extended period of time then the setpoint will be recalculated with an appropriately smaller value for R.The high alarm setpoint for the Offgas Noble Gas monitor is based on a limit of 350,000 uCi/sec. This is the release rate for which a FSAR accident analysis was completed. | ||
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However, the alarm setpoint may be recalculated using an updated nuclide distribution based on actual plant process conditions. | However, the alarm setpoint may be recalculated using an updated nuclide distribution based on actual plant process conditions. | ||
The monitor nominal response values will be confirmed during periodic calibration using a Transfer Standard source traceable to the primary calibration performed by the vendor.Particulates and lodines are not included in this calculation because this is a noble gas monitor.To provide an alarm in the event of failure of the offgas system flow instrumentation, the low flow alarm setpoint will beset at or above 10 scfm, (well below normal system flow) and the high flow alarm setpoint will be set at or below 110 scfm, which is well above expected steady-state flow rates with a tight condenser. | The monitor nominal response values will be confirmed during periodic calibration using a Transfer Standard source traceable to the primary calibration performed by the vendor.Particulates and lodines are not included in this calculation because this is a noble gas monitor.To provide an alarm in the event of failure of the offgas system flow instrumentation, the low flow alarm setpoint will beset at or above 10 scfm, (well below normal system flow) and the high flow alarm setpoint will be set at or below 110 scfm, which is well above expected steady-state flow rates with a tight condenser. | ||
To provide an alarm for changing conditions, the alert alarm will normally be set at 1.5 times nominal full power background to ensure that the Specific Activity Action required by ITS SR 3.7.4.1, are implemented in a timely fashion.(3.50E+05) | To provide an alarm for changing conditions, the alert alarm will normally be set at 1.5 times nominal full power background to ensure that the Specific Activity Action required by ITS SR 3.7.4.1, are implemented in a timely fashion.(3.50E+05) | ||
(2.12 E-03) j(__C/CFi) | (2.12 E-03) j(__C/CFi) | ||
+ Background Alarm Setpoint, cpm < 0.8 F i(CO)Where: Alarm Setpoint = The alarm setpoint for the offgas pretreatment Noble Gas Detector, cpm 0.8 -Safety Factor, unitless Unit 2 Revision 28 1115 September 2006 350,000 2.12E-03 Ci CFi F Background Ei (Ci/CFi)E (Ci)= The Technical Specification Limit for Offgas Pretreatment,ýtCi/sec= Unit conversion Factor, 60 sec/min / 28317 cc/CF= The concentration of nuclide, i, in the Offgas, [tCi/cc= The Detector response to nuclide i, pXi/cc/cpm; See Table D 3-1 for a list of nominal values-- The Offgas System Flow rate, CFM= The detector response to non-fission gases and general area dose rates, cpm-- The summation of the nuclide concentration divided by the corresponding detector response, net cpm= The summation of the concentration of nuclides in offgas,ýtCi/cc 2.2 2.2.1 Gaseous Effluents Dose Rate Calculation Dose rates will be calculated monthly at a minimum to demonstrate that the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the dose rate limits specified in 10CFR20. These limits are as follows: The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited per Technical Specification 5.5.4.g to the following: | + Background Alarm Setpoint, cpm < 0.8 F i(CO)Where: Alarm Setpoint = The alarm setpoint for the offgas pretreatment Noble Gas Detector, cpm 0.8 -Safety Factor, unitless Unit 2 Revision 28 1115 September 2006 350,000 2.12E-03 Ci CFi F Background Ei (Ci/CFi)E (Ci)= The Technical Specification Limit for Offgas Pretreatment,ýtCi/sec= Unit conversion Factor, 60 sec/min / 28317 cc/CF= The concentration of nuclide, i, in the Offgas, [tCi/cc= The Detector response to nuclide i, pXi/cc/cpm; See Table D 3-1 for a list of nominal values-- The Offgas System Flow rate, CFM= The detector response to non-fission gases and general area dose rates, cpm-- The summation of the nuclide concentration divided by the corresponding detector response, net cpm= The summation of the concentration of nuclides in offgas,ýtCi/cc 2.2 2.2.1 Gaseous Effluents Dose Rate Calculation Dose rates will be calculated monthly at a minimum to demonstrate that the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the dose rate limits specified in 10CFR20. These limits are as follows: The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited per Technical Specification 5.5.4.g to the following: | ||
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It should be noted that the most conservative pathways do not all exist at the same location. | It should be noted that the most conservative pathways do not all exist at the same location. | ||
It is conservative to assume that a single individual would actually be at each of the receptor locations. | It is conservative to assume that a single individual would actually be at each of the receptor locations. | ||
2.2.2 Whole Body Dose Rate Due to Noble Gases The ground level gamma radiation dose from a noble gas stack release (elevated), referred to as plume shine, is calculated using the dose factors from Appendix B of this document. | |||
Body Dose Rate Due to Noble Gases The ground level gamma radiation dose from a noble gas stack release (elevated), referred to as plume shine, is calculated using the dose factors from Appendix B of this document. | |||
The ground level gamma radiation dose from a noble gas vent release accounts for the exposure from immersion in the semi-infinite cloud. The dispersion of the cloud from the point of release to the receptor at the east site boundary is factored into the plume shine dose factors for stack releases and through the use of X/Q in the equation for the immersion ground level dose rates for vent releases. | The ground level gamma radiation dose from a noble gas vent release accounts for the exposure from immersion in the semi-infinite cloud. The dispersion of the cloud from the point of release to the receptor at the east site boundary is factored into the plume shine dose factors for stack releases and through the use of X/Q in the equation for the immersion ground level dose rates for vent releases. | ||
The release rate is averaged over the period of concern. The factors are discussed in Appendix B.Whole body dose rate (DR)y due to noble gases: (DR)y = 3.17E-08 Yi [ViQis + Ki (X/Q)vQiv] | The release rate is averaged over the period of concern. The factors are discussed in Appendix B.Whole body dose rate (DR)y due to noble gases: (DR)y = 3.17E-08 Yi [ViQis + Ki (X/Q)vQiv] | ||
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ground X/Q values are used for the vent releases (v=vent). (sec/mi 3)3.17E-8 Conversion Factor; the inverse of the number of seconds in a year;(yr/sec)Qiv,Qis The release rate of each noble gas nuclide i, from the stack(s) or vent (v) averaged over the time period of concern, pCi/sec.2.2.4 Organ Dose Rate Due to 1-131, 1-133, Tritium, and Particulates with Half-lives greater than 8 days.The organ dose rate is calculated using the dose factors (Ri) from Appendix C. The factor Ri takes into account the dose rate received from the ground plane, inhalation and ingestion pathways. | ground X/Q values are used for the vent releases (v=vent). (sec/mi 3)3.17E-8 Conversion Factor; the inverse of the number of seconds in a year;(yr/sec)Qiv,Qis The release rate of each noble gas nuclide i, from the stack(s) or vent (v) averaged over the time period of concern, pCi/sec.2.2.4 Organ Dose Rate Due to 1-131, 1-133, Tritium, and Particulates with Half-lives greater than 8 days.The organ dose rate is calculated using the dose factors (Ri) from Appendix C. The factor Ri takes into account the dose rate received from the ground plane, inhalation and ingestion pathways. | ||
W, and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. | W, and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. | ||
The release rate is averaged over the period of concern.Organ dose rates (DR)at due to iodine- 131, iodine- 133, tritium and all radionuclides in particulate form with half-lives greater than 8 days: (DR) at 3.17E-8 Yj[ZiRijat | The release rate is averaged over the period of concern.Organ dose rates (DR)at due to iodine- 131, iodine- 133, tritium and all radionuclides in particulate form with half-lives greater than 8 days: (DR) at 3.17E-8 Yj[ZiRijat | ||
[WsQis + WvQiv] I Where: (DR)at Organ dose rate (mrem/sec) | [WsQis + WvQiv] I Where: (DR)at Organ dose rate (mrem/sec) | ||
Rijat = The factor that takes into account the dose from nuclide i through pathway j to an age group a, and individual organ t. Units for inhalation pathway, mrem/yr per pCi/m 3.Units for ground and ingestion pathways, m 2-mrem/yr per uCi/sec. (See Tables D 3-4 through D 3-22).Ws, Wv Dispersion parameter either X/Q (sec/mi 3) or D/Q (1/M 2)depending on pathway and receptor location. | Rijat = The factor that takes into account the dose from nuclide i through pathway j to an age group a, and individual organ t. Units for inhalation pathway, mrem/yr per pCi/m 3.Units for ground and ingestion pathways, m 2-mrem/yr per uCi/sec. (See Tables D 3-4 through D 3-22).Ws, Wv Dispersion parameter either X/Q (sec/mi 3) or D/Q (1/M 2)depending on pathway and receptor location. | ||
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: a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and, b. During any calendar year: Less than or equal to 15 mrem to any organ.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 from iodine and particulate releases, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.2.3.1 W, and W, -Dispersion Parameters for Dose, Table D 3-23 The dispersion parameters for dose calculations were obtained chiefly from the Nine Mile Point Unit 2 Environmental Report Appendix 7B. These were calculated using the methodology of Regulatory Guide 1.111 and NUREG 0324. The stack was modeled as an elevated release point because height is more than 2.5 times the height of any adjacent building. | : a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and, b. During any calendar year: Less than or equal to 15 mrem to any organ.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 from iodine and particulate releases, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.2.3.1 W, and W, -Dispersion Parameters for Dose, Table D 3-23 The dispersion parameters for dose calculations were obtained chiefly from the Nine Mile Point Unit 2 Environmental Report Appendix 7B. These were calculated using the methodology of Regulatory Guide 1.111 and NUREG 0324. The stack was modeled as an elevated release point because height is more than 2.5 times the height of any adjacent building. | ||
The vent was modeled as a combined elevated/ground level release because the vent's height is not more than 2.5 times the height of any adjacent building. | The vent was modeled as a combined elevated/ground level release because the vent's height is not more than 2.5 times the height of any adjacent building. | ||
Average meteorology over the appropriate time period was used. Dispersion parameters not available from the ER were obtained from C.T. Main Data report dated November, 1985, or the FES.Unit 2 Revision 28 II 20 September 2006 | Average meteorology over the appropriate time period was used. Dispersion parameters not available from the ER were obtained from C.T. Main Data report dated November, 1985, or the FES.Unit 2 Revision 28 II 20 September 2006 2.3.2 Gamma Air Dose Due to Noble Gases Gamma air dose from the stack or vent noble gas releases is calculated monthly. The gamma air dose equation is similar to the gamma dose rate equation except the receptor is air instead of the whole body or skin of whole body. Therefore, the stack noble gas releases use the finite plume air dose factors, and the vent noble gas releases use semi-infinite cloud immersion dose factors. The factor X/Q takes into account the dispersion of vent releases to the most conservative location. | ||
Air Dose Due to Noble Gases Gamma air dose from the stack or vent noble gas releases is calculated monthly. The gamma air dose equation is similar to the gamma dose rate equation except the receptor is air instead of the whole body or skin of whole body. Therefore, the stack noble gas releases use the finite plume air dose factors, and the vent noble gas releases use semi-infinite cloud immersion dose factors. The factor X/Q takes into account the dispersion of vent releases to the most conservative location. | |||
The release activity is totaled over the period of concern. The finite plume factor is discussed in Appendix B.Gamma air dose due to noble gases: DY = 3.17E-8 Yi[Mi(X/Q)v Qiv + Bi Qjis x t D= The gamma air dose for the period of concern, mrad t= The duration of the dose period of concern, sec Where all other parameters have been previously defined.2.3.3 Beta Air Dose Due to Noble Gases The beta air dose from the stack or vent noble gas releases is calculated using the semi-infinite cloud immersion dose factor in beta radiation. | The release activity is totaled over the period of concern. The finite plume factor is discussed in Appendix B.Gamma air dose due to noble gases: DY = 3.17E-8 Yi[Mi(X/Q)v Qiv + Bi Qjis x t D= The gamma air dose for the period of concern, mrad t= The duration of the dose period of concern, sec Where all other parameters have been previously defined.2.3.3 Beta Air Dose Due to Noble Gases The beta air dose from the stack or vent noble gas releases is calculated using the semi-infinite cloud immersion dose factor in beta radiation. | ||
The factor X/Q takes into account the dispersion of releases to the most conservative location.Beta air dose due to noble gases: DP -3.17E-8 YiNi[(X/Q)v Qiv + (X/Q)s Qis] X t Dp = Beta air dose (mrad) for the period of concern Ni The constant accounting for the beta air dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed on Table D 3-3, mrad/yr per uCi/m 3.(From Reg. Guide 1.109).t = The duration of the dose period of concern, sec Where all other parameters have been previously defined.2.3.4 Organ Dose Due to 1-131, 1-133, Tritium and Particulates with half-lives greater than 8 days.The organ dose is based on the same equation as the dose rate equation except the dose is compared to the 1 OCFR50 dose limits. The factor Ri takes into account the dose received from the ground plane, inhalation, food (cow milk, cow meat and vegetation) pathways.W, and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. | The factor X/Q takes into account the dispersion of releases to the most conservative location.Beta air dose due to noble gases: DP -3.17E-8 YiNi[(X/Q)v Qiv + (X/Q)s Qis] X t Dp = Beta air dose (mrad) for the period of concern Ni The constant accounting for the beta air dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed on Table D 3-3, mrad/yr per uCi/m 3.(From Reg. Guide 1.109).t = The duration of the dose period of concern, sec Where all other parameters have been previously defined.2.3.4 Organ Dose Due to 1-131, 1-133, Tritium and Particulates with half-lives greater than 8 days.The organ dose is based on the same equation as the dose rate equation except the dose is compared to the 1 OCFR50 dose limits. The factor Ri takes into account the dose received from the ground plane, inhalation, food (cow milk, cow meat and vegetation) pathways.W, and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. | ||
The release is totaled over the period of concern. The Ri factors are discussed in Appendix C.Organ dose Dat due to iodine- 131, iodine- 133, tritium and radionuclides in particulate Unit 2 Revision 28 1121 September 2006 form with half-lives greater than 8 days.Dat = 3.17E-8 yj [ Ei Rijat [Ws Qis + Wv Qiv] I X t Where: Dat = Dose to the critical organ t, for age group a, mrem t = The duration of the dose period of concern, sec Where all other parameters have been previously defined in Section 2.2.4.2.4 1-133 and 1-135 Estimation Stack and vent effluent iodine cartridges are analyzed to a sensitivity of at least 1 E- 12 uCi/cc. If detected in excess of the LLD, the 1-131 and 1-133 analysis results will be reported directly from each cartridge analyzed. | The release is totaled over the period of concern. The Ri factors are discussed in Appendix C.Organ dose Dat due to iodine- 131, iodine- 133, tritium and radionuclides in particulate Unit 2 Revision 28 1121 September 2006 form with half-lives greater than 8 days.Dat = 3.17E-8 yj [ Ei Rijat [Ws Qis + Wv Qiv] I X t Where: Dat = Dose to the critical organ t, for age group a, mrem t = The duration of the dose period of concern, sec Where all other parameters have been previously defined in Section 2.2.4.2.4 1-133 and 1-135 Estimation Stack and vent effluent iodine cartridges are analyzed to a sensitivity of at least 1 E- 12 uCi/cc. If detected in excess of the LLD, the 1-131 and 1-133 analysis results will be reported directly from each cartridge analyzed. | ||
Periodically, (usually quarterly but on a monthly frequency if effluent iodines are routinely detected) a short-duration (12 to 24 hour) effluent sample is collected and analyzed to establish an 1-135/1-131 ratio and an I-133/1-131 ratio, if each activity exceeds LLD. The short-duration ratio is used to confirm the routinely measured 1-133 values. The short-duration 1-135/I- 131 ratio (if determined) is used with the 1-131 release to estimate the 1-135 release. The short-duration 1-133/1-131 ratio may be used with the 1-131 release to estimate the 1-133 release if the directly measured 1-133 release appears non-conservative. | Periodically, (usually quarterly but on a monthly frequency if effluent iodines are routinely detected) a short-duration (12 to 24 hour) effluent sample is collected and analyzed to establish an 1-135/1-131 ratio and an I-133/1-131 ratio, if each activity exceeds LLD. The short-duration ratio is used to confirm the routinely measured 1-133 values. The short-duration 1-135/I- 131 ratio (if determined) is used with the 1-131 release to estimate the 1-135 release. The short-duration 1-133/1-131 ratio may be used with the 1-131 release to estimate the 1-133 release if the directly measured 1-133 release appears non-conservative. | ||
2.5 Isokinetic Sampling Sampling systems for the stack and vent effluent releases are designed to maintain isokinetic sample flow at normal ventilation flow rates. During periods of reduced ventilation flow, sample flow may be maintained at a minimum flow rate (above the calculated isokinetic rate) in order to minimize sample line losses due to particulate deposition at low velocity.2.6 Use of Concurrent Meteorological Data vs. Historical Data It is the intent to use dispersion parameters based on historical meteorological data to set alarm points and to determine or predict dose and dose rates in the environment due to gaseous effluents. | |||
Sampling Sampling systems for the stack and vent effluent releases are designed to maintain isokinetic sample flow at normal ventilation flow rates. During periods of reduced ventilation flow, sample flow may be maintained at a minimum flow rate (above the calculated isokinetic rate) in order to minimize sample line losses due to particulate deposition at low velocity.2.6 Use of Concurrent Meteorological Data vs. Historical Data It is the intent to use dispersion parameters based on historical meteorological data to set alarm points and to determine or predict dose and dose rates in the environment due to gaseous effluents. | |||
If effluent levels approach limiting values, meteorological conditions concurrent with the time of release may be used to determine gaseous pathway doses.2.7 Gaseous Radwaste Treatment System Operation Part I, Section D 3.2.4 requires the GASEOUS RADWASTE TREATMENT SYSTEM to be in operation whenever the main condenser air ejector system is in operation. | If effluent levels approach limiting values, meteorological conditions concurrent with the time of release may be used to determine gaseous pathway doses.2.7 Gaseous Radwaste Treatment System Operation Part I, Section D 3.2.4 requires the GASEOUS RADWASTE TREATMENT SYSTEM to be in operation whenever the main condenser air ejector system is in operation. | ||
The system may be operated for short periods with the charcoal beds bypassed to facilitate transients. | The system may be operated for short periods with the charcoal beds bypassed to facilitate transients. | ||
The components of the system which normally should operate to treat offgas Unit 2 Revision 28 1122 September 2006 2.8 are the Preheater, Recombiner, Condenser, Dryer, Charcoal Adsorbers, HEPA Filter, and Vacuum Pump. (See Appendix D, Offgas System).Ventilation Exhaust Treatment System Operation Part I, Section D 3.2.5 requires the VENTILATION EXHAUST TREATMENT SYSTEM to be FUNCTIONAL when projected doses in 31 days due to iodine and particulate releases would exceed 0.3 mrem to any organ of a member of the public. The appropriate components, which affect iodine or particulate release, to be FUNCTIONAL are: 1)2)3)HEPA Filter -Radwaste Decon Area HEPA Filter -Radwaste Equipment Area HEPA Filter -Radwaste General Area Whenever one of these filters is not FUNCTIONAL, iodine and particulate dose projections will be made for 31-day intervals starting with filter nonfunctionality, and continuing as long as the filter remains nonfunctional, in accordance with DSR 3.2.5.1.Predicted release rates will be used, along with the methodology of Section 2.3.4. (See Appendix D, Gaseous Radiation Monitoring.) | The components of the system which normally should operate to treat offgas Unit 2 Revision 28 1122 September 2006 2.8 are the Preheater, Recombiner, Condenser, Dryer, Charcoal Adsorbers, HEPA Filter, and Vacuum Pump. (See Appendix D, Offgas System).Ventilation Exhaust Treatment System Operation Part I, Section D 3.2.5 requires the VENTILATION EXHAUST TREATMENT SYSTEM to be FUNCTIONAL when projected doses in 31 days due to iodine and particulate releases would exceed 0.3 mrem to any organ of a member of the public. The appropriate components, which affect iodine or particulate release, to be FUNCTIONAL are: 1)2)3)HEPA Filter -Radwaste Decon Area HEPA Filter -Radwaste Equipment Area HEPA Filter -Radwaste General Area Whenever one of these filters is not FUNCTIONAL, iodine and particulate dose projections will be made for 31-day intervals starting with filter nonfunctionality, and continuing as long as the filter remains nonfunctional, in accordance with DSR 3.2.5.1.Predicted release rates will be used, along with the methodology of Section 2.3.4. (See Appendix D, Gaseous Radiation Monitoring.) | ||
Unit 2 Revision 28 1123 September 2006 | Unit 2 Revision 28 1123 September 2006 3.0 URANIUM FUEL CYCLE The "Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b) as follows: "Uranium fuel cycle means the operations of milling of uranium ore chemical conversion of uranium, isotopic enrichment 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, to the 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 recovered non-uranium special nuclear and by-product materials from the cycle." Sections D 3.1.2, D 3.2.2, and D 3.2.3 of Part I requires that when the 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 fuel 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 result in 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 reduce subsequent releases.The report to the NRC shall contain: 1) Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site, that contribute to the annual dose of the maximum exposed member of the public.2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from all existing pathways and sources of radioactive effluents and direct radiation. | ||
FUEL CYCLE The "Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b) as follows: "Uranium fuel cycle means the operations of milling of uranium ore chemical conversion of uranium, isotopic enrichment 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, to the 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 recovered non-uranium special nuclear and by-product materials from the cycle." Sections D 3.1.2, D 3.2.2, and D 3.2.3 of Part I requires that when the 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 fuel 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 result in 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 reduce subsequent releases.The report to the NRC shall contain: 1) Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site, that contribute to the annual dose of the maximum exposed member of the public.2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from all existing pathways and sources of radioactive effluents and direct radiation. | |||
The total body and organ doses resulting from radioactive material in liquid effluents from Nine Mile Point Unit 2 will be summed with the doses resulting from the releases of noble gases, radioiodines, and particulates. | The total body and organ doses resulting from radioactive material in liquid effluents from Nine Mile Point Unit 2 will be summed with the doses resulting from the releases of noble gases, radioiodines, and particulates. | ||
The direct dose components will also be determined by either calculation or actual measurement. | The direct dose components will also be determined by either calculation or actual measurement. | ||
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However, any updated field data may be utilized that concerns locations of real individuals, real time meteorological data, location of critical receptors, etc. 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. | However, any updated field data may be utilized that concerns locations of real individuals, real time meteorological data, location of critical receptors, etc. 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 effluent calculational data.Doses to members of the public from the pathways considered in section 2 as a result of gaseous effluents will be calculated using the methodology of Regulatory Guide 1.109 or the methodology of the ODCM, as applicable. | Environmental sample media data such as TLD, air sample, milk sample and vegetable (food crop) sample data may be utilized in lieu of effluent calculational data.Doses to members of the public from the pathways considered in section 2 as a result of gaseous effluents will be calculated using the methodology of Regulatory Guide 1.109 or the methodology of the ODCM, as applicable. | ||
Doses calculated from environmental sample media will be based on methodologies found in Regulatory Guide 1.109.Unit 2 Revision 28 II 26 September 2006 | Doses calculated from environmental sample media will be based on methodologies found in Regulatory Guide 1.109.Unit 2 Revision 28 II 26 September 2006 3.3 Evaluation of Doses From Direct Radiation The dose contribution as a result of direct radiation shall be considered when evaluating whether the dose limitations of 40 CFR 190 have been exceeded. | ||
of Doses From Direct Radiation The dose contribution as a result of direct radiation shall be considered when 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. | 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. | 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. | ||
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In addition, pathways associated with water related recreational activities, other than fishing, are not applicable here. These include swimming, boating and wading which are prohibited at the facility.Unit 2 Revision 28 1127 September 2006 The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium 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. | In addition, pathways associated with water related recreational activities, other than fishing, are not applicable here. These include swimming, boating and wading which are prohibited at the facility.Unit 2 Revision 28 1127 September 2006 The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium 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 the following equation adapted from Regulatory Guide 1.109. Table D 3-24 presents the reference for the parameters used in the following equation.NOTE: The following equation is adapted from equations C-3 and C-4 of Regulatory Guide 1.109. Since many of the factors are in units of pCi/m 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 to the Regulatory Guide equations. | Thus, the inhalation pathway is evaluated using the following equation adapted from Regulatory Guide 1.109. Table D 3-24 presents the reference for the parameters used in the following equation.NOTE: The following equation is adapted from equations C-3 and C-4 of Regulatory Guide 1.109. Since many of the factors are in units of pCi/m 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 to the Regulatory Guide equations. | ||
Dja Ei [ (Ci) F (X/Q) (DFA) ija (BR) at]Where: Dja The maximum dose from all nuclides to the organ j and age group (a) in mrem/yr; ex. if calculating to the adult lung, then Dja = DL and DFAija = DFAiL C i The average concentration in the stack or vent release of nuclide i for the period in pCi/rn 3.F Unit 2 average stack or vent flowrate in m 3/sec.X/Q = The plume dispersion parameter for a location approximately 0.50 miles west of NMP-2 (The plume dispersion parameters are 9.6E-07 (stack) and 2.8E-06 (vent) and were obtained from the C.T. Main five year average annual X/Q tables. The vent X/Q (ground level) is ten times the listed 0.50 mile X/Q because the vent is approximately | Dja Ei [ (Ci) F (X/Q) (DFA) ija (BR) at]Where: Dja The maximum dose from all nuclides to the organ j and age group (a) in mrem/yr; ex. if calculating to the adult lung, then Dja = DL and DFAija = DFAiL C i The average concentration in the stack or vent release of nuclide i for the period in pCi/rn 3.F Unit 2 average stack or vent flowrate in m 3/sec.X/Q = The plume dispersion parameter for a location approximately 0.50 miles west of NMP-2 (The plume dispersion parameters are 9.6E-07 (stack) and 2.8E-06 (vent) and were obtained from the C.T. Main five year average annual X/Q tables. The vent X/Q (ground level) is ten times the listed 0.50 mile X/Q because the vent is approximately 0.3 miles from the receptor location. | ||
from the receptor location. | |||
The stack (elevated) | The stack (elevated) | ||
X/Q is conservative when based on 0.50 miles because of the close proximity of the stack and the receptor location.(DFA)ija the dose factor for nuclide i, organ j, and age group a in mrem per pCi (Reg. Guide 1.109, Table E-7); ex. if calculating to the adult lung the DFAija = DFAiL (BR)a annual air intake for individuals in age group a in M3 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 2 Revision 28 1128 September 2006 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 Regulatory Guide 1.109 as presented in Section 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. | X/Q is conservative when based on 0.50 miles because of the close proximity of the stack and the receptor location.(DFA)ija the dose factor for nuclide i, organ j, and age group a in mrem per pCi (Reg. Guide 1.109, Table E-7); ex. if calculating to the adult lung the DFAija = DFAiL (BR)a annual air intake for individuals in age group a in M3 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 2 Revision 28 1128 September 2006 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 Regulatory Guide 1.109 as presented in Section 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. | ||
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At least two environmental TLDs will be used at one location in the approximate area where fishing occurs. The TLDs will be placed in the field on approximately the beginning of each calendar quarter and removed approximately at the end of each calendar quarter (quarter 2, 3, and 4).The average TLD readings will be adjusted by the average control TLD readings. | At least two environmental TLDs will be used at one location in the approximate area where fishing occurs. The TLDs will be placed in the field on approximately the beginning of each calendar quarter and removed approximately at the end of each calendar quarter (quarter 2, 3, and 4).The average TLD readings will be 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 used after adjusting for the appropriate time period (as applicable). | 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 used after adjusting for the appropriate time period (as applicable). | ||
In the event of loss or theft of the TLDs, results from a TLD or TLDs in a nearby area may be utilized.Unit 2 Revision 28 1129 September 2006 | In the event of loss or theft of the TLDs, results from a TLD or TLDs in a nearby area may be utilized.Unit 2 Revision 28 1129 September 2006 4.0 ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Stations The current sampling locations are specified in Table D 5-1 and Figures D 5.1-1 and D 5.1-2. The meteorological tower location is shown on Figure D 5.1-1 and is located where TLD location #17 is identified. | ||
MONITORING PROGRAM 4.1 Sampling Stations The current sampling locations are specified in Table D 5-1 and Figures D 5.1-1 and D 5.1-2. The meteorological tower location is shown on Figure D 5.1-1 and is located where TLD location #17 is identified. | |||
The Environmental Monitoring Program is a joint effort between the owners and operators of the Nine Mile Point Units 1 and 2 and the James A. FitzPatrick Nuclear Power Plants. 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 D 5-1 are based on the NMP-2 reactor centerline. | The Environmental Monitoring Program is a joint effort between the owners and operators of the Nine Mile Point Units 1 and 2 and the James A. FitzPatrick Nuclear Power Plants. 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 D 5-1 are based on the NMP-2 reactor centerline. | ||
The average dispersion and deposition parameters for the three units have been calculated for a 5 year period, 1978 through 1982. Average dispersion or deposition parameters for the site are calculated using the 1978 through 1982 data and are used to compare the results of the annual land use census. If it is determined that sample locations required by Control D 3.5.1 are unavailable or new locations are identified that yield a significantly higher (i.e., 50%) calculated D/Q value, actions will be taken as required by Controls D 3.5.1 and D 3.5.2 and the Radiological Environmental Monitoring Program updated accordingly. | The average dispersion and deposition parameters for the three units have been calculated for a 5 year period, 1978 through 1982. Average dispersion or deposition parameters for the site are calculated using the 1978 through 1982 data and are used to compare the results of the annual land use census. If it is determined that sample locations required by Control D 3.5.1 are unavailable or new locations are identified that yield a significantly higher (i.e., 50%) calculated D/Q value, actions will be taken as required by Controls D 3.5.1 and D 3.5.2 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 Interlaboratory 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 which cross check samples are available. | |||
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 Interlaboratory 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 which cross check samples are available. | |||
An attempt will be made to obtain a QC sample to program sample ratio of 5% or better. The Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission staff may evaluate the results.Specific sample media for which EPA Cross Check Program samples are available include the following: | An attempt will be made to obtain a QC sample to program sample ratio of 5% or better. The Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission staff may evaluate the results.Specific sample media for which EPA 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 0 tritium in water* 1- 13 1 in water Unit 2 Revision 28 1130 September 2006 | * gross beta in air particulate filters* gamma emitters in air particulate filters* gamma emitters in milk* gamma emitters in water 0 tritium in water* 1- 13 1 in water Unit 2 Revision 28 1130 September 2006 4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements required by the Technical Specifications are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use. In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabilities may be determined by the vendor who supplies the TLDs. Required detection capabilities are as follows.4.3.1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less 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 10 uR/hr during the field cycle. The average of the relative 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 interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constant. | |||
for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements required by the Technical Specifications are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use. In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabilities may be determined by the vendor who supplies the TLDs. Required detection capabilities are as follows.4.3.1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less than 7.5%. A total of at least 5 TLDs shall be evaluated. | |||
shall be determined by giving TLDs repeated exposures equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The average of the relative standard deviations of the responses shall be less than 3.0%. A total of at least 4 TLDs shall be evaluated. | |||
of exposure interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constant. | |||
This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures. | This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures. | ||
For these tests, the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85. At least 6 TLDs shall be evaluated. | For these tests, the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85. At least 6 TLDs shall be evaluated. | ||
4.3.4 Energy dependence shall be evaluated by the response of TLDs to photons for several energies between approximately 30 keV and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25% for photons with energies greater than 80 keV 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. | |||
dependence shall be evaluated by the response of TLDs to photons for several energies between approximately 30 keV and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25% for photons with energies greater than 80 keV 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. | 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, the TLD 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. | To accomplish this, the TLD 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 be 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 not 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 2 Revision 28 1131 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. | |||
dependence shall be 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 not 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. | The TLDs shall be exposed under two conditions: | ||
Unit 2 Revision 28 1131 September 2006 | |||
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 phosphor, as appropriate, shall be dried before readout. The response of the TLD exposed in the plastic bag containing water shall not differ from that exposed in the regular plastic bag by more than 10%. A total of at least 4 TLDs shall be evaluated for each condition. | (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 phosphor, as appropriate, shall be dried before readout. The response of the TLD exposed in the plastic bag containing water shall not differ from that exposed in the regular plastic bag by more than 10%. A total of at least 4 TLDs shall be evaluated for each condition. | ||
4.3.8 Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is less than 10 uRlhr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3). The average exposure inferred from the responses of the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least 3 TLDs shall be evaluated. | 4.3.8 Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is less than 10 uRlhr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3). The average exposure inferred from the responses of the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least 3 TLDs shall be evaluated. | ||
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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.In addition, not all dose factors are used for the dose calculations. | 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.In addition, not all dose factors are used for the dose calculations. | ||
A maximum individual is used, which is a composite of the maximum dose factor of each age group for each organ as reflected in the applicable chemistry procedures. | A maximum individual is used, which is a composite of the maximum dose factor of each age group for each organ as reflected in the applicable chemistry procedures. | ||
Unit 2 Revision 28 II 68 September 2006 APPENDIX B PLUME SHINE DOSE FACTOR DERIVATION Unit 2 Revision 28 September 2006 1169 Appendix B For elevated releases the plume shine dose factors for gamma air (Bi) and whole body (Vi), are calculated using the 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 gamma 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 follows: Gamma Air Bi = Y I. Where: K' conversion factor (see S RE) v, below for actual value).Ila mass absorption coefficient (cm 2/g; air for Bi, tissue for Vi)E Energy of gamma ray per disintegration (Mev)Vs average wind speed for each stability class (s), r/s R = downwind distance (site boundary, m)E = sector width (radians)s = subscript for stability class is I function = I + k1 2 for each stability class. (unitless, see Regulatory Guide 1.109)k2 Fraction of the attenuated energy that is actually absorbed in air (see Regulatory Guide 1.109, see below for equation)Whole Body-ýLatd Vi = 1.1lSFBie Where: td tissue depth (g/cm 2)SF -shielding factor from structures (unitless) 1.11 = Ratio of mass absorption coefficients between tissue and air.Where all other parameters are defined above.Unit 2 Revision 28 II 70 September 2006 Appendix B (Cont'd)1K = conversion factor 2k = ýL -a I-La 3.7ElOdis 1.6E-6erz Ci-sec Mev =1293 g 100 erg Im3 g-rad.46 Where: t = mass attenuation coefficient (cm 2/g; air for Bi, tissue for Vi)I-La-defined above There are seven stability classes, A thru F. The percentage of the year that each stability class 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 Unit 2 FSAR. 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 | Unit 2 Revision 28 II 68 September 2006 APPENDIX B PLUME SHINE DOSE FACTOR DERIVATION Unit 2 Revision 28 September 2006 1169 Appendix B For elevated releases the plume shine dose factors for gamma air (Bi) and whole body (Vi), are calculated using the 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 gamma 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 follows: Gamma Air Bi = Y I. Where: K' conversion factor (see S RE) v, below for actual value).Ila mass absorption coefficient (cm 2/g; air for Bi, tissue for Vi)E Energy of gamma ray per disintegration (Mev)Vs average wind speed for each stability class (s), r/s R = downwind distance (site boundary, m)E = sector width (radians)s = subscript for stability class is I function = I + k1 2 for each stability class. (unitless, see Regulatory Guide 1.109)k2 Fraction of the attenuated energy that is actually absorbed in air (see Regulatory Guide 1.109, see below for equation)Whole Body-ýLatd Vi = 1.1lSFBie Where: td tissue depth (g/cm 2)SF -shielding factor from structures (unitless) 1.11 = Ratio of mass absorption coefficients between tissue and air.Where all other parameters are defined above.Unit 2 Revision 28 II 70 September 2006 Appendix B (Cont'd)1K = conversion factor 2k = ýL -a I-La 3.7ElOdis 1.6E-6erz Ci-sec Mev =1293 g 100 erg Im3 g-rad.46 Where: t = mass attenuation coefficient (cm 2/g; air for Bi, tissue for Vi)I-La-defined above There are seven stability classes, A thru F. The percentage of the year that each stability class 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 Unit 2 FSAR. 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 | ||
([ia) and attenuation (i) coefficients were calculated by multiplying the mass absorption (FLa/P) and mass attenuation (i'/p) coefficients given in the Radiation Health Handbook by the air density equal to 1.293 E-3 g/cc or the tissue density of 1 g/cc where applicable. | ([ia) and attenuation (i) coefficients were calculated by multiplying the mass absorption (FLa/P) and mass attenuation (i'/p) coefficients given in the Radiation Health Handbook by the air density equal to 1.293 E-3 g/cc or the tissue density of 1 g/cc where applicable. | ||
The tissue depth is 5g/cm 2 for the whole body.The downwind distance is the site boundary.SAMPLE CALCULATION Ex. Kr-89-DATA E 4a =Rz F STABILITY CLASS ONLY- Gamma Air 2.22MeV k = FLLa .871 K = .46 2.943 E-3m-' 4a VF = 5.55 m/sec 5.5064E-3m-' | The tissue depth is 5g/cm 2 for the whole body.The downwind distance is the site boundary.SAMPLE CALCULATION Ex. Kr-89-DATA E 4a =Rz F STABILITY CLASS ONLY- Gamma Air 2.22MeV k = FLLa .871 K = .46 2.943 E-3m-' 4a VF = 5.55 m/sec 5.5064E-3m-' | ||
R = 1600m.39 19m vertical plume spread taken from "Introduction to Nuclear Engineering", John R. LaMarsh Unit 2 Revision 28 September 2006 1171 Appendix B (Cont'd)-I Function UOz I1 12 I-- .11= .3= .4 D I1 + k1 2= .3 + (.871) (.4) = .65 I dis. 1 0.46 [Ci-sec) (Mev/ergsi (2.943E-3m-') | R = 1600m.39 19m vertical plume spread taken from "Introduction to Nuclear Engineering", John R. LaMarsh Unit 2 Revision 28 September 2006 1171 Appendix B (Cont'd)-I Function UOz I1 12 I-- .11= .3= .4 D I1 + k1 2= .3 + (.871) (.4) = .65 I dis. 1 0.46 [Ci-sec) (Mev/ergsi (2.943E-3m-') | ||
(2.22Mev) | (2.22Mev) | ||
(.65)(iW (g/m 3) (ergs) (5.55 m/s) (.39) (1600m)(g-rad)= 3.18(-7) rad/s (3600 s/hr) (24 h/d) (365 d/y) (iE3mrad/rad) | (.65)(iW (g/m 3) (ergs) (5.55 m/s) (.39) (1600m)(g-rad)= 3.18(-7) rad/s (3600 s/hr) (24 h/d) (365 d/y) (iE3mrad/rad) | ||
Ci/s (lE6uCi)Ci 1.00(-2) mrad/yr uCi/sec Vi 1.11 (.7) L(E-2)mrad/yr] | Ci/s (lE6uCi)Ci 1.00(-2) mrad/yr uCi/sec Vi 1.11 (.7) L(E-2)mrad/yr] | ||
[e[iCi/sec-(.0253 cm 2/g) (5g/cm 2)I 6.85(-3) mrad/yr pCi/sec Note: The above calculation is for the F stability class only. For Table D 3-2 and procedure values, a weighted fraction of each stability class was used to determine the Bi and Vi values.Unit 2 Revision 28 II 72 September 2006 APPENDIX C DOSE PARAMETERS FOR IODINE 131 and 133, PARTICULATES AND TRITIUM Unit 2 Revision 28 September 2006 1173 Appendix C DOSE PARAMETERS FOR IODINE -131 AND -133, PARTICULATES AND TRITIUM This appendix contains the methodology which was used to calculate the organ dose factors for 1-131, 1-133, particulates, and tritium. The dose factor, Ri, was calculated using the methodology outlined in NUREG-0 133. The radioiodine and particulate DLCO 3.2.1 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. | [e[iCi/sec-(.0253 cm 2/g) (5g/cm 2)I 6.85(-3) mrad/yr pCi/sec Note: The above calculation is for the F stability class only. For Table D 3-2 and procedure values, a weighted fraction of each stability class was used to determine the Bi and Vi values.Unit 2 Revision 28 II 72 September 2006 APPENDIX C DOSE PARAMETERS FOR IODINE 131 and 133, PARTICULATES AND TRITIUM Unit 2 Revision 28 September 2006 1173 Appendix C DOSE PARAMETERS FOR IODINE -131 AND -133, PARTICULATES AND TRITIUM This appendix contains the methodology which was used to calculate the organ dose factors for 1-131, 1-133, particulates, and tritium. The dose factor, Ri, was calculated using the methodology outlined in NUREG-0 133. The radioiodine and particulate DLCO 3.2.1 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. | Washout was calculated and determined to be negligible. | ||
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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 2 Revision 28 1177 September 2006 Appendix C (Cont'd)C.4 Grass-Cow-Meat PathwayK'QfUapFf(r)(DFL)i,(+ | 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 2 Revision 28 1177 September 2006 Appendix C (Cont'd)C.4 Grass-Cow-Meat PathwayK'QfUapFf(r)(DFL)i,(+ | ||
fPfy +(1-fPf,)e-;{th X ) ~ ~ ~ (A1 A,,) I Ye , -Ri(M) = Dose factor for the meat ingestion pathway for radionuclide i for any organ of interest, (units = m -mrem/yr per ýtCi/sec)Ff = The stable element transfer coefficients, (units = pCi/kg per pCi/day)Uap = The receptor's meat consumption rate for age group a, (units = kg/year)th = The transport time from harvest, to cow, to receptor, (units = sec)tf = The transport time from pasture, to cow, to receptor, (units = sec)All other terms remain the same as defined for the milk pathway. Table C-2 contains the values which were used in calculating Ri(M).The concentration of tritium in meat is based on airborne concentration rather than deposition. | fPfy +(1-fPf,)e-;{th X ) ~ ~ ~ (A1 A,,) I Ye , -Ri(M) = Dose factor for the meat ingestion pathway for radionuclide i for any organ of interest, (units = m -mrem/yr per ýtCi/sec)Ff = The stable element transfer coefficients, (units = pCi/kg per pCi/day)Uap = The receptor's meat consumption rate for age group a, (units = kg/year)th = The transport time from harvest, to cow, to receptor, (units = sec)tf = The transport time from pasture, to cow, to receptor, (units = sec)All other terms remain the same as defined for the milk pathway. Table C-2 contains the values which were used in calculating Ri(M).The concentration of tritium in meat is based on airborne concentration rather than deposition. | ||
Therefore, the RT(M) is based on X/Q.RT(M) = K'K"'FfQfUap(DFL)iat | Therefore, the RT(M) is based on X/Q.RT(M) = K'K"'FfQfUap(DFL)iat | ||
[0.75(0.5/H)] | [0.75(0.5/H)] | ||
Where: All C.5 RT(M) = Dose factor for the meat ingestion pathway for tritium for any organ of interest, (units = mrem/yr per ýtCi/m 3)other terms are defined above.Vegetation Pathway The integrated concentration in vegetation consumed by man follows the expression developed for milk. Man is considered to consume two types of vegetation (fresh and stored)that differ only in the time period between harvest and consumption, therefore: | Where: All C.5 RT(M) = Dose factor for the meat ingestion pathway for tritium for any organ of interest, (units = mrem/yr per ýtCi/m 3)other terms are defined above.Vegetation Pathway The integrated concentration in vegetation consumed by man follows the expression developed for milk. Man is considered to consume two types of vegetation (fresh and stored)that differ only in the time period between harvest and consumption, therefore: | ||
Revision as of 00:43, 13 July 2019
| ML071280692 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/28/2006 |
| From: | Fiorenza T, Hutton J, Kurtz T, Schimmel M, Stinson G Constellation Energy Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML071280692 (206) | |
Text
Constellation Energy" Nine Mile Pnint Nuiclear Station NINE MILE POINT :NUCLEAR STATION NINE MILE POINT UNIT 2 OFF-SITE DOSE CALCULATION MANUAL (ODCM.)DATE APPROVALS SIGNATURES REVISION 28 Prep2red by: Reviewed by: Concurred by: T M.Kurtz Health Physicist G. R. Stinson H Phys, Is 17Gener e General Supervisor Design Engnr J. A. Hutton Pla- ene al. ,anager M. A. Schirmmel-J ' f9 Manager Engineering.S vices
SUMMARY
OF REVISIONS Revision 28 (EFfective September
- 29. 2006)PAGE 1 3.3-13,14 i13.3-6 1 4.0-1 II 2-10,26,33-36,66,67,75,80 ix, I 1.0-1, I 1.,02, IB 3.3-2,14.1-1
& !a, It, 11, I15ý .129,.Il 63, II 107, 11 108 I 3.3-9 1 3.3-10 1 3.3-7, 1 3.3-12, and 1 3.3-13 II 63, II 64, and 11107 II 3 and ii 4 iv, 1 1.6-1, 1-3.1-7, 13.2-3, 1 3.2-10, 13.2-,121I3.3A-1, 13.3-2,13.3-3, I 3.3-7, .3.3-8, 1 3.3w9, 1 3.3-10, 1 B 3.1-3, 1 B 3.2-5, 1 B 3.2-6, l B 3.3-1,, I B 3.3-2,1 4.1-la, 11 10,11113,1120, and 1123 DATE August 2000 November 2000 November 2000 November 2000 December 2001 December.
2002 March 2003 January 2004 December 2005 May 2006 September 2006 i Unit 2 Revision 28 September,2006 TABLE OF CONTENTS PAGE List of Tables vii List of Figures ix Introduction x PART I -RADIOLOGICAL EFFLUENT CONTROLS I SECTION 1.0 DEFINITIONS 1.1.0-0 SECTION 2.0 Not Used SECTION 3.0 APPLICABILITY 13.0-0 D 3.1 Radioactive Liquid Effluents I 3.-I-D 3.1.1 Liquid Effluents Concentration I13.1-1 D 3.1.12 Liquid Effluents Dose. 13.1-4 D 3;1:3 Liquid Radwaste Treatment System 1:3.1-7 D 3.2 Radioactivc Gaseous Effluents 13.2-1 D 3.2.1 Gaseous Effluents Dose Rate I3.2-i D 3.2.2 Gaseous Effluents Noble Gas Dose 1 3-2-4 D 3.2.3 Gaseous Effluents Dose- iodine-1,3 1, Iodine-1 33, Tritium, and Radioactive Material in Particulate Form 13.2-7 D 3.2.4 Gaseous Radwaste Treatment System 13.2-10 D 3.2.5 Ventilation Exhaust Treatment System I 3.2-12 D 3.2.6 Venting or Purging I 3:2-14 D 3.3 Instrumentation.
I3.3-i D 3.3.,1 Radioactive Liquid Effluent Monitoring Instrumentation i 3.3-I D 3.3:2 Radioactive Gaseous Effluent Monitoring Instrumentation 1 3.3-7 D 3.4 Radioactive Effluents Total Dose I'3.4-i D 3.5 Radiological Environmental Monitoring 1,3.5-1* 3.5.1 Monitoring Program 1 35-1, D 3.5:2 Land Use Census 13.5-13* 3.5.3 Inlerlaboratory Comparison Program 13.5-16 BASES I B 3.1-0 B 3.1 Radioactive Liquid Effluents IB 3.1-1 B 3.1.1 Liquid Effluents Concentration I B 3.1-1 B 3.1.2 Liquid Effluents Dose I B311-2 B 3.1.3 Liquid Radwaste Treatment System ILB 3 1-3 Hi Unit 2 Revision 28 September 2006.
TABLE OF CONTENTS (Cont)B 3.2 B 3.2.1 B 3.2.2 B 3.2.3 B 3.2.4 B 3.2.5 B 3.2.6 B 33 B 3.3.1 B 3.3.2 Radioactive Gaseous Effluents Gaseous Effluents DoseRate Gaseous Effluents Noble Gas Dose*Gaseous Effluents Dose -Iodine-13)1, Iodine- 133, Tritium, and Radioactive Material in Particulate Form Gaseous Radwaste Treatment System Ventilation Exhaust Treatment System Venting or Purging instrumentation Radioactive Liquid Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring InStrumentation Radioactive Effluents Total Dose Radiological Environmental Moriitoring Monitoring Program Land Use Census Interlaboratory
'Comparison Prograni PAGE I B 3.2-1 I.B 3.2-1 I B 3-2-2 I B 3.2-3[ B 3 .2-5 I B 3.2-6 I. B3.2-7 I B 3.3-1 I B3.3-1 I B 3.3-2 lB 3.4-1 I B 3-5-1 1B 3.5-1, I B-3.5-2 I B 3.5-3 B3.4 B 3.5 B 3.5.1 B 3.5,2 B 3.5.3 SECTION 4.0 ADMINISTRATIVE CONTROLS 14.0-1 D 4.1 D4.1.1 D 4.2 Reporting.
Requh'ements Special Reports Major Changes to Liquid, Gaseous and Solid Radwaste Treatment Systems 14.1-1 1 4. 1.'4 14.2-1 Unit 2 Revision 28 September 2006 iii TABLE OF CONTENTS (Cont).IF7CTI"N ,'lJR R1?FVCTION PA GE IZUBJEXT REFSECTION PART II -CALCULATIONAL METHODOLOGIES 1.0 LIQUID EFFLUENTS I .1 Liquid Effluent Monitor Alarm Setpoints 1 A.11 Basis 1.1.2 Setpoint Determination Methodology 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint 1 .1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculations 1.1.2..3 Service Water and Cooling Tower Bi0wdown Effluent Radiation Alarm Setpoint 3.1.1 3.3.1 II i 112 112 112 112 115 116:1.2 1.3 1.4 1.5 2.0 Liquid Effluent ConcentrationCalculaijon Liquid Effluent Dose CalculationMethodology Liquid Effluent Sampling Representativeness Liquid Radwaste System FUNCTIONALITY 3.1.1 DSR 3.1.1.2 3.1.2 DSR 3.1.2.1 Table D 3. A -1 note b 3.1.3 DSR 3.1.3.1 B 3.1.3 II 7 II 8 119 II 10 1112 11'12 11 12 1113 11 14 GiASEOUS EFFLUENTS 2.1 211,1.2.1,2 1.1.2 i 2..1.2.1 2.1.2.2 2.1.2.3 Gaseous Effluent Monitor Alarm Setpoints Basis 3.2..1 Setpoirit Deierminatiori MethodologY Discussion 3.3.2 Stack Noble Gas.Detector Alarm. Setpoint Equation Vent Noble Gas Detector Alarm Setpoint Equation Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation 2.2 2.2.1 2.2.2 2.2.3 Gaseous Effluent Dose. Rate Calculation Methodology X/Q and W, -Dispersion Parameters for Dose Rate, Table D 3-23 Whole Body Dose RateDue to Noble Gases Skin Dose Rate Due to: Noble Gases 3.2.1 DLCO3.2.! .a DSR 3.2 .i.1 DLCO 3.2.1.a DSR 3.2.1.1 1115 11 16 IIF16'1117 ilia iv Unit 2 Revision 28 September 2006 TABLE OF CONTENTS (Cont)SECTION SUBJECT REF SECTION PAGE 2.2.4 2,3 2-3.1 2.3.2 2.3.3 23.4 Organ Dose RateDue to 1-131,1-133,Tritium and DLCO 3.2.L.b Particulates with half-lives greater than 8 days DSR 3.2.1.2 Gaseous Effluent.Dose Calculation Methodology 3.2.2 3.2.3 3.2.5 W, and WV,:- Dispersion Parameters For Dose, Table D 3.-23 Gamma.Air Dose Due to Noble Gases 3,2.2 DSR 3.2.2.1 Beta Air Dose Due to Noble Gases 3.3.2 Organ Dose Due to 1- 31, 1-133, Tritium and Particulates 3.2.3 with half-lives 312.5 DSR 3.2.3.1 DSR 3.2.5.1 I-133 and I-135 Estimation Isokinetic Sampling Use of Concurrent Meteorological Data vs. Historical Data Gaseous Radwaste Treatment System Operation 3.2.4 Ventilation Exhaust.Treatment System Operation 3.2.5 2.4 2.5.2.6 2.7 2.8 3.0 3.1 3.2 3.3, 3.4 4.0, 4.1 4.2 4.31 1119 1I 20 II 20 1121, II 21 1i 21 II 22 11 22 II 22 HI 22 II 23 1124 I125 II 26 11 27 1I 27 11 30 II 30 11 30 1131 URANIUM FUEL CYCLE 3.4 Evaluation of Doses From Liquid Effluents DSR 3.1.2.1 Evaluation of Doses.From Gaseous Effluents DSR 3.2.2.1 Evaluation of Doses From Direct Radiation DSR 3.2.3.1 Doses to Members of the Public Within the Site Boundary 4.1 ENVIRONMENTAL MONITORING PROGRAM 3.5 Sampling Stations 3.5.1 DSR 3.5.1.1 Interlaboratory Comparison Program DSR 3.5.3.2 Capabilities for Thermoluminescent Dosimeters -used for Environmental Measurements V Unit 2 Revision 28 September 2006 TABLE OF CONTENTS (Cont)SECTION SUBJECT REF SECTION PAGE Appendix A Liquid Dose Factor Derivation Appendix B Appendix C Appendix D Plume Shine Dose, Factor Derivation Dose'Parameters.for Iodine 131 and 133, Particulates and Tritium Diagrams of Liquid and Gaseous Radwaste Treatment Systems. and Monitoring Systems 11.66 II 69 1173 II 83 11106 Appendix E Nine Mile Point On-Site and Off-Site Maps vi Unit 2 Revision 28 September 2006' LIST OF TABLES PART I -RADIOLOGICAL EFFL UENT CONTROLS TABLE NO TITLE PAGE.D 3.1.1-1 Radioactive Liquid Waste Sampling and Analysis 1 3.1-2 D 3.2.1-1 Radioactive Gaseous Waste Sampling and Analysis 1 3.2-2 D 3.3.1 -1 Radioactive Liquid Effluent Monitoring Instrumentation 13.3-6 D 3.3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation 1 3.3-13 D 3.5.1-1 Radiological Environmental Monitoring Program 1 3.5-6 D 3.5,. 1-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples I 3.5-10 D 3.5.1-3 Detection Capabilities for Environmental Sample Analyses, I 3.5-11 vii Unit 2 Revision 28 September 2006 LISTOF TARLES (Cont)PARTH -CALCULATIONAL METHODOLOGIES TABLE NO D 2-1 D 2-2 thru D 2-5 D3-1 D 3-2 D3-3 D 3-4 thru D 3-22 D 3-23 D 3-24 D 5.1 TITLE Liquid Effluent Detector Response Aiat Values -Liquid Effluent Dose Factor Offgas PrctMatmient Detector Response Finite Plume -Ground Level Dose Factorsfrom an Elevated Release Immersion Dose Factors Dose And Dose Rate Factors, Ri Dispersion Parameters at Controlling Locations,.
X/Q, W, and W, Values Parameters"Fof the Evaluation of Doses to Real Members of the Public From Gaseous ,And Liquid Effluents Radiological Environmental Monitoring Program Sampling Locations PAGE II 33.1 34 II 38 11.39 1140 1141 II 60 Ii161 11 62 VIII Unit 2 Revision, 28 September 2006 LIST OF FIGURES FIGURE NO D 1.0-1 D 5.1-1 D 5.1-2 D5.1i2 TITLE Site Area and Land Portion of Exclusion Area Boundaries Nine Mile Point On-Site Map Nine Mile Point Off-Site Map (page I of 2)Nine Mile Point Off-Site Map (page 2 .of 2)PAGE 1 1.0-4 11107 11108 11 109 ix Unit 2 Revision 28 September 2006 INTRODUCTION The OFFSITE DOSE CALCULATION MANUAL (ODCM) is a supporting document of the Technical Specifications Section 5.5.1. The previous Limiting Conditions for Operation thatwere contained in the Radiological Effluent Technical Specifications are now transferred to the ODCM as Radiological Effluent Controls.
The ODCM contains two parts: Radiological Effluent Controls, Part I; and Calculational Methodologies, Part I. Radiological Effluent Controls, Part 1,includes the following:
(1) The Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification 5.5.1 and (2) descriptions of the information that:should be included in the: Annual Radiological Environmental Operating and Radioactive-Effluent Release Reports required by Technical Specifications 5.6.2 and 5.6.3. Calculational.
Methodologies, Part 1i, describes the methodology and parameters to be used in the calculation of liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints and the calculation of offsitesdoses due to radioactive liquid and gaseous effluents.
The ODCM also contains, a listand graphical description of the specific sample locations for the radiologiCal environmental monitoring program, and liquid and gaseous radwaste treatment system configurations.
The ODCM follows the methodology and models suggestcd by NUREG-O133 and Regulatory Guide 1. 109, Revision 1. Simplifying assumptions have been applied in this manual where applicable to provide a more workable documentfor implementing the Radiological Effluent Control requirements; this, simplified approach will result in a more conservative dose evaluation for determining compliance with regulatorý requirements.
The ODCM will be maintained for use as a reference and training document of accepted methodologies and calculations.
Changes to the calculation.methods or parameters will be incorporated into the ODCM to assure that the ODCM representsthe present, methodology in all applicable areas. Any changes to the ODCM will be implemented in accordance with Section 5.5.1 of theTechnical SpecificationsK.
x Unit 2 Revision 28 September 2006 PART I -RADIOLOGICAL EFFLUENT CONTROLS Unit 2 Revision 28 September 2006 I Definitions 1.0 PART I -RADIOLOGICAL EFFLUENT CONTROLS SECTION 1.0 DEFINITIONS Unit 2 Revision 28 Scptemrbcr 2006 1 1.0-0 Definitions 1.0 1.0 DEFINITIONS
NOTE -----------------------------------------------
Technical Specifications defined terms and the following additional defined terms appear in capitalized type and are applicable throughout these specifications and bases.TERM DEFINITION FUNCTIONAL (FUNCTIONALITY)
GASEOUS RADWASTE TREATMENT SYSTEM MEMBER(S)OF THE. PUBLIC MILK SAMPLING LOCATION FUNCTIONALITY is an attribute of Structures, Systems, or Components (SSCs) that is not controlled by Technical Specifications.
An SSC shall be functional or havw functionality when it is capabIc of pcrforming its specified function~as set forth in the Current Licensing Basis (CLB).FUNCTIONALITY does not apply to specified safety functions, but does apply to tlhe ability of non-Technical Specifications SSCs to perform specified support functions.
A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting offgases from the main: condenser evacuation system 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~shall include all persons who are not occupationally associated with the Nine Mile Point Nuclear Station and James A. FitzPatrick Nuclear. Power Plant. This category does not include employees of owners and operators of the Nine Mile Point Nuclear:Station and James A, Fitzpatrick Nuclear Power Plant, their contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries..
This, category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with Nine Mile Point Nuclear Station and James A. FitzPatrick Nuclear Power Plant.A MILK SAMPLING LOCATION is a location where. 10 or more head of milk animals are available for collection of milk samples.(continued)
Unit 2 Revision:
28-September 2006 I 1.0-1 Definitions 1.0 1.0 DEFINITIONS (continued)
TERM DEFINITION OFFSITE DOSE CALCULATION MANUAL PURGE -PURIGIN REPORTABLE EVENT SITE BOUNDARY SOURCE CHECK UNRESTRICTED AREA The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the current methodology and parameters used in the calculation of offsite doses that result from radioactive gaseous and liquid effluents, in the calculation of gaseous and, liquid effluent monitoring Alarm/Trip Sctpoints, and in the conduct of the environmental radiological monitoring program. The ODCM shall also contain: (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Program required by Specification 5.5.1 of Technical Specifications and, (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 5.6.2 and 5,6.3.PURGE and PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
A REPORTABLE EVENT shall be any of those conditions specified in 10 CFR 50.73.The SITE BOUNDARY shall be that line around the Nine Mile Point Nuclear Station beyond whichthe land is not owned, leased or otherwise controlled by the owners and operators of NineMile Point Nuclear Station and James A, Fitzpatrick Nuclear Power Plant. See Figure D 1.0-1.A SOURCE CHECK shall be the qualitative assessment of channel response when the c'hanncl sensor is exposed to a source of increased radioactivity.
An UNRESTRICTED AREA shall bel 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 or for industrial, commercial, institutional, and/or recreational purposes.(continued)
Unit 2 Revision 28 September 2006 11.0-2 Definitions 1.0 1.0 DEFINITIONS (continued)
TERM DEFINITION VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM shall be 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 ads.orbers and/or 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 consideredto have any effect on noble gas efflucnts).
Engineered safety features (ESF) atmospheric cleanup systems arebnot :considered to be VENTILATION EXHAUST TREATMENT SYSTEM conpo'nents.
VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, concentration, or other operating condition, in such a manner that replacementair or gas is' not provided or required during VENTING. Vent, used in system names, does-not imply a VENTING process.Unit 2 Revision 28 September 2006 1 1.0-3 NMP'2 41DISCHARGE Definitions 1.0 RIO T A L A K E Ainer Rood Lycoming-SITE AREA AND LAND PORTION OF EXCLUSION AREA BOUNDARIES SCALE-IhL ES Niagara Mohawk Power. Corporation retains ownership in, switchyard facilities within the exclusion area boundary.by Nine .ile Point NuCI6or Station, LLC by agreement.
cer-tain trurismission line and Access and usage ore controlled Figure D 1.0-1 (Page 1 of 1)Site Area and Land Portion of Exclusion Area Boundaries Unit 2 Revision 28 September 2006 11.0-4 PART I -RADIOLOGICAL EFFLUENT CONTROLS SECTION 3.0 APPLICABILITY Unit.2 Revision 28 September 2006 I 3.0-0 Applicability 3.0 3.0 APPLICABILITY The Offsite Dose Calculation Manual (ODCM) Specifications are contained in Section 3.0 of Part I. They contain operational requirements, Surveillance Requirements, and reporting requirements.
Additionally, the Required Actions and associated Completion Times for degraded Conditions are specified.
The format is consistent with the Technical Specifications (Appendix A to the NMP2 Operating License).The rulcs of usagc for the ODCM Specification are the same as those for the Technical Specifications.
These rules are found in Technical.Specifications Sections 1.2, "Logical Connectors," 1.3, "Completion Times," and 1.4, "Frequency." The ODCM Specifications are subject to Technical Specifications Section 3.0, "Limiting Condition for Operation (LCO) Applicability and Surveillance Requirement (SR) Applicability," with the following exceptions:, 1. LCO 3.06, regarding support/supported system ACTIONS is not applicable to ODCM Specifications.
- 2. LCO 3.0.7, regarding allowances, to change specified Technical Specifications is not applicable to ODCM Specifications:
- 3. Section 3.0 requirements are not required when so stated in notes within individual.
specifications.
Unit 2 Revision 28 September 2006 13.0-1 Liquid Effluents Concentration D 3.1.1 D 3.1 RADIOACTIVE LIQUID EFFLUENTS* 3.1.1 Liquid Effluents Concentration DLCO 3.1. 1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (Figure D 1.0-1) shall be limited to;a. Ten times the concentration specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gaises; and b. 2 x 1.0-4 ýtCi/ml total activity concentration for dissolved or entrained noble gases.APPLICABILITY:
At all times.ACTIONS CONDITION REQUIRED ACTION, COMPLETION TIME A. Concentration of A. .Initiate action to restore Immediately radioactive.
material concentration to within limits.released in liquid effluents to' UNRESTRICTED AREAS exceeds limits.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.1.1 Perform radioactive liquid waste sampling and In accordance with activity analysis.
Table D 3.1.1.-1 DSR 3.1.1.2 Verify the results of the DSR 3.1. .1 analyses to In accordance with assure thatthe concentrations at the point of release Table D 3. 1.1-1 are maintained within the limits of DLCO 3.1. 1.Unit 2 Revision 28 September 2006 13.1-I Liquid Effluents Concentration D 3.1.1 Table D 3.1.1-1 (Page 1 of 2)Radioactive Liquid Waste Sampling and Analysis LIQUI D RELEASE TYPE 1, Batch Waste Release Tanks (b)a. 2LWSJ,.K4A
- b. 2LWS-TK4B C. 2LWS-TK5A d. 2LWS-TK5B SAMPLE SAMPLE TYPE FREQUENCY Grab Sample Each Batch (g)Grab"Sample One batch/31 days (g)ANALYSIS FREQUENCY Each Batch (g)31 days SAMPLE ANALYSIS Principal Gamma Emitters(e) 1-131 Dissolved and*ntrained Ga..es (gamma tirlitters)
H-3 Gross Alpha St-89 Proportional, Composite of grab samples Wd)Proportional Composite of-grab samples (d)Each batch(g) 31 days: Each batch (g) 92. days 2. Continuous Release&a. Service Water Effluent A b. Service Water Effluent B c. Cooling Thwer Blowdown 3. Coniinuous.
Release Auxiliaýr Boiler Pump Seal and Saniple Cooling Discharge (Service Water)I Grab Sample 31 day,;(e)Grab Sample 31 days (e)Grab Sample 31 days,(e):Sr-90 Fe-55 31 days (e) Principal Gamma'Emitters (c)31 days (e) 1-131 3 I days (e) Dissolved and Entrained Gases (gamma emitters)31 days(es) 11-3 3t days (e) Gross Alpha 92 days (e> Sr-89 92 days (e> Sr-0 92 days (e) :FeZ55 31 days (f. :PrincipaI Gamma Emitters (c)92 days (1) 1t-3 SAMPLELOWER LIMIT OF.DETECTION (LLD) (a).5, x101 ICi/ml I x 10-.Ci/ml I x to- 5 LCi/ml I x 104 VaCi/mi I x I0"7 pCi/ml ,5 x 10-' pCi/ml 5 x 10"' pCi/ml I xl 0- pCi/mI:5 X, 1-"7 11Ci/Mt I x 10`5 pCi/ml I x 10-' PCi/nml I x 1l0 'pCilml 1 x 10*8 PCi/mI x t O.7 paCi/mI 5 x I& pci/ml I x 10-7 PCi/mI I x 10-.Ci/mI Grab Sample Grab Sample Grab Sample Grab Sample Grab Sample Grab Sample 31 days (6)31I days,,(c)32 days (e)92 days (e).92 days (e)3 1. days (f)Grab Sample 92 days (f)Unit 2 Revision 28 September 2006 1.3.1-2 Liquid Effluents Concentration D 3.1.1 Table D 3.1.1-1 (Page 2 of 2)Radioactive Liquid. Waste Sampling and. Analysis (a) ITLe LED is defiied as the smallest conceiitration of radioactiveeii-naterial in a sampnle that will yielda net count, above system backgound, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation representsa'*r"nal" signsl.For a particular menesureinnt sy.stem, which may include radichemical separation:.
LLD (4.66) (S, )(E)(QV) (222x1 06) {Y) e-where: LLD) The before-the-fact.lower limit of detection
(ýCi per unit mass or volume), Sb= The standard deviation of the background counting rate or of t.he counting.rate of aiblank sample as appropriate (counts per miiiiute),.
E -The counting efficiency (counts per disintegration), V. The sample size (units of mass or volume), 2,22 x 06 The number of disintegrations per minute per jiCi, Y = The fractional radiochemical.yield, when applicable, radioactive decay constant forihe particularradionuclide (see), and At = The elapsed time betweenthemidpoint of siampie collec-tiofi and the'tim of counting (seconds).
Typical valuesof E; V. Y, and At should be used in the calculation, It'iShould be recognized thaet he LELD is definted.i a before.-thtfact limit iepresentin g the capability of a measuremenrt sysiemand:
not as an after-the-fact limitfor a particular measurement.(b) A batch release is the discharge of liquid wastesof volume:, Prior to sampling for analyses, each bitch shall be isolated, and then thoroughly mixed by the method described in Part 11, Section 1.4 to assure representative sampling.(c) The principal gamma emitters for which the. LLDIapplies include the following radionuecides:
Mn-541 Fe-59,Co-58,Co-60, Zn-65,<Mo-99 .Cs 13, C 137, and Ce14 Ce-144 shall also bemeasured.butw ith an LLD of 5 x 106pCi'rml.
Thishlistdoes'notnmean 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 thie Radioactive Effluent Release Report pursuant to Technical Specification 5.63 in the format outlined in RG 1.21, Appendix B,. Revision I, June 1974.(d) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste dischargedand.in which the method of sampling employed results in a specimen that is representative of the liquids releised..(e) If the alarm setpoint of the effluent monitor is exceeded', the frequency of sampling shall be increased to daily until the condition no longerexists.
Frequency of analysis shallbeincreasdto daily forprincipal gamma emitters and an incident composite for. H13, griss alpha, Sri89, Si-90, and Fc-55.If the alarm setpoint of Service Water Effluent Monitor.A and/or'. Bis.exceeded, the Ifrequeney of sampling shall be increased tow daily until the condition no longer exists. Frequency of analysis shall be increased ti daily for pnncipalgamma:emitters and an incident composite ,br H-3. gross aspha, Sr-89, Sr.-90, and Fe55. " (g) Complete prior to each release.Unit 2 Revision 28 113.1-3 September 2006 Liquid Effluents Dose D 3.1.2 D 3.1 RADIOACTIVE LIQUID EFFLUENTS D 3. 1.2 Liquid Effluents Dose DLCO 3.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials released in liquid, effluents from each unit to UNRESTRICTED AREAS (FigureD 1 0-ý1) shall be limited to: a. ,:S 1.5 mremn to the whole body and _< 5 mrem to any organ during any calendar quarter; and b. 3 mrem to the whole body and ! 0 mrem to any organ during any calendar year.APPLICABILITY:
At all times.ACTIONS........ ------- -------- NOTES,--------------------------
- 1. LCO 3.0.3 is not applicable.
- 2. LCO 3.0.4 is not applicable.
CONDITION A. Calculated dose toa MEMBER OF THE PUBLIC from the release ofradioactive materials in liquid effluents to UNRESTRICTED AREAS exceedslimits.
REQUIRED ACTION COMPLETION TIME .A.I Prepare and submit to the NRC, pursuant.to D 4.1.. 1,.a Special Report that (1) Identifies the cause(s) for exceeding the limit(s)and (2) 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 DLCO 3.1.2.30 days.(continued)
Unit 2 Revision 28 September 2006 I 3.1-4 Liquid Effluents Dose*D 3.1.2 ACTIONS(continuO.CONDITION REQUIRED ACTION COMPLETION TIME 4 B. Calculated dose to a MEMBER OF THE PUBLIC from the release, of radioactive materials in liquid effluents exceeds 2 times the limits.B.1 Calculate the annual dose to a.MEMBER OF THE PUBLIC which includes contributions from direct radiation from the units (including outside storage tanks, etc.).Immediately Inmediately AND B.2 Verify that the limits ofDLCO 3.4 have not been exceeded.C. Required Action B.2 and Associated Completion time not met.C. I Prepare and submitto the NRC, pursuant to D 4. 1. 1, a Special Report, as defined iný10 CFR 20.2203 (a)(4), of Required Action A. 1 shall also include the following:
(1) The corrcctivc action(s) to be taken to prevent recurrence of exceeding the limits of DLCO 3.4 and the schedule for -achieving conformance, (2) An analysis that estimates the 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), and (3) Describes the levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations.
30 days Unit 2 Revision 28 September 2006.I 3.1-5 Liquid Effluents Dose D 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.2.1 Determine cumulative dose contributions from liquid 3,1 days effluents for the current calendar quartet and the current calendar year.Unit 2 Revision 28 September 2006.1 3.1-6 Liquid Radwaste Treatment System D 3.1.3 D 3.1 RADIOACTIVE LIQUID EFFLUENTS D 3.1.3 Liquid Radwaste Treatment System DLCO 3.1.3 The liquid radwaste treatment.
system shall bbe FUNCTIONAL.
APPLICABILITY:
At. all -times.ACTIONS--------------------------------
- w. .N OTES -- -..... ..................----------
- 1. LCO 3.0.3 is not applicable..
- 2. LCO 3.0.4 is not applicable.
CONDITION A. Radioactive liquid waste.being discharged without treatment.
AND Projected doses due.to the liquid effluent, from the unit, to UNRESTRICTED AREAS would exceed 0.06 mrem to the whole bodyof 0.2 mrem to any organ in a 3:1 day period.AND Any portion of the liquid radwaste treatment system not in operation.
REQUIRED ACTION A. I Prepare and submit to the NRC, pursuant to D4.1.I, a Special Repolt that includes: (1) An explanation of why liquid radwaste was being discharged without treatment, identification of any nonfunctional equipment or subsystems, and the reason for the nonfunctionality, (2) Action(s) taken to restore the nonfunctional equipment to FUNCTIONAL status, and (3) Summary description of action(s) taken to prevent a recurrence.
30 days COMPLETION TIME Unit 2 Revision 28 September 2006 1 3.1-7 Liquid Radwaste Treatment System D 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.3.1 -- -----------
NOTE ---------------
Only required to be met.when liquid radwaste treatment systems are. not being fully utilized.Project the doses'due to liquid effluents from each. 31 days unit to UNRESTRICTED AREAS.Unit 2 Revision 28 September 2006, I 3:1-8 Gaseous Effluents Dose-Rate D 3.2.1 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.1 Gaseous Effluents Dose Rate DLCO 3.2.1 The dose rate, from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (Figure D 1.04) shall be limited to: a. For noble gases, < 500. mremlyr to the whole body and<30,00 mremlyr to the skin and b. For 1-131, 1-133, H-3 and all radionuclides in particulate form with half-lives
> 8 days, < 1500 mrem/yr to any organ.APPLICABILITY:
At all times.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. The dose rate(s)atpor A.1 Restore the release:rate to Immvdiately beyond the SITE within the limit.BOUND ARY due to .... .. ... .radioactive gaseous effluents exceeds limits.SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.1.1 Thedosc rate from noble gases in gaseous effluents In accordance with shall be determined to be within the limits of DLCO Table D 3.2.1,-i 3.2.1.a.DSR 3.2.1.2 The dose rate from 1-131, 1-133, H-3 and all In accordance with radionuclides in particulate form with half-lives TableD.3.2>1-1
> 8 days in gaseous effluents shall be determined to be within the limits of DLCO 3.2.1l.b.Unit 2 Revision 28 September 2006 1 3.2-1 Gaseous: Effluents Dose Rate D 3.2. 1 Table D 32.1-1 (Page 1 of 2)Radioactive Gaseous Waste Sampling and Analysis GASEOUS RELEASE TYPE Containmncnt (b)SAMPLE TYPE Grab Sample SAMPLE FREQUENCY Each Purg¢A FR~ANALYSIS SAMPLE'EQUENCY ANALYSIS (h) Principal Gamma Emitters'(c)
Each Purge H-3 (oxide)Each Purge Principal Gzimma Emitters, (c)31 days.(d) nncpal Gamma Emittere (c).'31 days (e) H-3 (oxide)7 days (g) 1-13I Main Stack.Radwaste/Reactor Building Vent Grab Sample 3. days (d)SAMPLFI P 1OWFI LIMIT OF DETECTION I x 10-6 pCi/ml I x. 1O 1 0- Ci/in lx, 1 0 A ECi/mI Ix~l p.2 i rilm I xpCi/mi.I x 10.1 pCi./m~I I'x 1011 iCiMIm Grab Sample, Charcoal Sample Particulate Sample Composite Particulate Sample 331 days (e)Continuous (4,° _Continuous (01 7 days (g) Principal Gamma Emitters (c)Gross Alpha ,Continuous (f) 92 days Sr-89.Sr-90 See the notes on the next page.Unit 2 Revision 28 September 2006 1 3.2-2 Gaseous Effluents Dose Rate D 3.2.1 Table D 3.2.1-1 (Page 2 of.2).Radioactive Gaseous Waste Sampling and Analysis" (a) The (LLD is defined as the~smalest 1 concentration of radioactive material in a sample that will yield.a net count, above system background, thatiw ll be detecte -with 95% probability withonly 5% probability of fally concluding that a blank observation represents a "real" signal.For a particular measurement systern, which may include radiochcmieal separation:
LI)D (4.66)(S,)
.(E) (V) (2.22x 010) (Y) e-'l..l) The before-the-fact lower limit of detection (fiCi per unit mass or volume), he standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (Counts per minute), E The counting efficiency (counts per:diintegration);
V = The sample size (units of mass or volume),.2.,22x. I 0ý -The number of disintegrations per minute per aCi, Y The Fractional radiochemical yield;wben applicable,-- The radioactive decay constant for the particular radionuclide (sc) and At= Th clapsed. time-bctwecn the midpoint of sample collection arid the time 6f counting (seconds):
Typical values ofE, 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 asan afterthe-fet limitfor a particular measurement.(b) Sample and analysis betore PURCE is used to deitcminc permissible PURGE rates, Sample and analysis during actual PURGE is used for offsitse dose calculiitions..(c) The principal gamma emitters for which the LLD applies include the following radionuclides:
Kr 87, Kr-88,.Xe-133, Xe-133m, Xe-135, and Xe138 in noble gas releasoes and Mn-54 Fe-59, Co-58, Co 60,-Zn-65 Mo-99,1-131.
Cs- 134, Cs- 37, Ce and Ce-I144 in iodine and particulate releases.
This list does not mean that only these ,riaclidesý are to be considered.
Othergamma peaks that are idcntiiable, iogether with: os oftsi of h e:above nticlides, shall also be analyzied and reported in the RadioactiveEffluent ReleaseRepoit pursuant to Technical Specification 5.6.3 in the format outlined in R.G 1 ..Appendis 1, Revision 1. June 1974.(d) Ift-he main stack or reactor/Tadwaste building isotopic monitor is not FUNCTIONAL.
sampling and analysis shall also be performoJ following shutdown, startup, or when there is an alarm on the offgaspretreatnientniobit0i.(e) 11-3 grab samples shall be taken once every 7 days from the reactor/radwaste ventilation system when fuel is ofiloadcd until stable H-3 release levels can be demonstrated.(f) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the tine period covered by each dose or dose rate'calculaTion made in accordarnce with DLCO 3.2. .b and DICO 3.2.3.(g) When theurelease rate of the main stack or reactoriradwaste building vent exceeds its alarm setpoint, the iodine and.particulate device shall bie- ed to determinethe chanoes in todino 'nd partirl-ste release ratis. The ariayirshill bediin onc per 24 hoburi itil the release n6 longer exceeds the alarii setpoint.
When samplesdoilted for 24 ours are analyzed; the corresponding LLDs may be increased by a factor of 10.(h) Complete prior to eachrelease:.
Unit 2 Revision 28 13.2-3 September 2006 Gaseous Effluents Noble Gas Dose D 3.2.2 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.2 Gaseous Effluents Noble Gas Dose DLCO 3.2.2 The air dose from noble gases released in gaseous effluents from eachunit to areas at or beyond the SITE BOUNDARY (Figure D 1.0-1) shall be ,limited to: a. During any calendar quarter: <, 5 mrad for gammaradiation and_ 10 mrad for beta~radiation and b. During any calendar year: _s 10 mrad for gamma radiation and<20 mrad for beta radiation.
APPLICABILITY:
Atall times.ACTIONS--------------------.-
--- ------------------
-.--------
--NOTI ES ---- -----------------------
1. LCO 3.0.3 is not applicable.
- 2. LCO3.0.4 is not applicable.
CONDITION A. Thepair dos e at or beyond the SITE BOUNDARY due to noble gases released in exceeds limits.REQUIRED ACTION A. 1 Prepare and submit to the NRC, pursuant to D 4.1-1.ia Special Report that (1) Identifies the cause(s) for exceeding the limit(s) and (2) 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 DLCO 3.2.2.COMPLETION TIME 30 days (continued)
Unit 2..Revision 28 September,2006 1 3.2-4 Gaseous Effluents Noble Gas Dose D 3.2.2 ACTIONS (continued)-
CONDITION REQUIRED ACTION i B., Calculated doselo a MEMBER OF THE,.PUBLIC from the release of radioactive materials in gaseous effluents due to noble gases exceeds 2 times the limits.B. I Calculate the annual dose-to a MEMBER OF THE PUBLIC which includes contributions from direct radiation from the units (including outside storage tanks, etc.).COMPLETION, TIME Imnediately Immediately AND B.2 Verity.that the. limits of DLCO 3.4 have not been exceeded.C. Required Action B.2 and Associated Completion time not met.C.I Special Report, asdefined in 10 CFR 20.2203-(a)(4), of Required Action A. 1 shall also include the following:
(1) The corrective action(s) to be taken to prevent, recurrence of exceeding thelimits of DLCO 3.4 and the .schedule for achieving conformance, (2) An analysis that estimates thc dose to a. MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and directfradiation, for the, calendar year that includes the release(s), and (3) Describes the:levels of radiation and concentrations of radioactive material involved and the cause of.the exposure.
levels, or concentrations.
30 days Unit 2 Revision 28 September,2006 1 3.2-5 Gaseous Effluents Noble Gas Dose D 3.2.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.2.1 Determine cumulative dose .contributions for the 331 days current calendar quarter and, current calendar year.Unit 2 Revision 28 September 2006 13.2-6 Gaseous Effluents Dose -:-1431, 1433, H-3 and Radioactive Material in ParticulateForm D 3.2.3 D3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.3 Gaseous EffluentsDose 131, 1-133, H-3 and Radioactive Materialin Particulate Form DLCO 3.2.3 The dose to a MEMBER OF THE PUBLIC from M1131,,-133,'-H-3.
and all radioactive material in particulate form with half-lives
> 8 days in gaseous effluents released, from each unit, to areas :at or beyond the SITE BOUNDARY (Figure D 1.0-1) shall be limited to: a. Duiring any calendar quarter- < 7.5imrem to any organ and bK During any calendar year: _< 15 mrem to any organ.,APPLICABILITY:
At all times.ACTIONS 1. LCO 3.0.3 is not applicable.
- 2. LCO 3.0.4 is not applicable.
NOTES -,.--------.--
.............---------------------
COMPLETION CONDITION A. The dose from I-131, 1-133, H-3 and radioactive material in particulate'form with half-lives > 8 days released in gaseous effluents at or, ,beyond the SITE BOUNDARY exceeds limits.REQUIRED ACTION COMPLETION TIME 30 days A.1 Prepare and submit to the N RC, pursuant to D 4.1.1, a Spccial Report that (1) Identifies the cause(s) for exceeding the limit(s) and (2) Defines the corrective actions that have:been takento reduce the releases and the proposed correctiVe:'actions to be taken to assure that subsequent releases will be in compliance with DLCO 3.2.3.(continued)
Unit 2 Revision 28 September 2006 13.2-7 Gaseous Effluents Dose -l-1.11, 1-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME-4-B. Calculated dose to a MEMBER OF THE PUBLIC from the-release of ,radioactive, materials in gaseous effluents exceeds 2 times the limits.B.I Calculate the annual dose to a MEMBER OF THE PUBLIC which includes contributions from direct radiation from the units (including outside storage tanks, etc.).AND'B.2 Verify that the limits of DLCO 3.4 have not been exceeded.Immediately Immediately 30 days C. Required Action B.2 and Associated Completion time not met.C.] Special Report, as defined in 10 CFR 20.2203 (a)(4), of Required Action A. 1 shall also include the following:
.(l)The corrective action(s) to be ,taken to prevent recurrence of exceeding the liirits of DLCO 3.4 and the schedule:
for achieving c0nformance, (2)An analysis that:estimates the 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), and (3)Describes the levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations.
'Unit 2 Revision 28 September 2006 13.2-8 Gaseous Effluents Dose 13 1, 1-133, H-3 and Radioactive Material in Particulate.
Form D 3.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.3.1 Determine cumulative dose contributions for the 31 days current calendar quarter and current calendar year for 1- 131, i-133, 1-13 and radioactive material in particulate form with half-lives
>8 days,.Unit 2 Revision 28 September 2006 13.2-9 Ouseous Radwaste Treatment System D 3.2.4 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.4 Gaseous Radwaste Treatment System DLCO 3.2.4 The GASEOUS RADWASTE TREATMENT, SYSTEM shall be in operation.
APPLICABILITY:
Whenever the main condenser air ejector system is in operation.
ACTIONS---.-.3------ ----- not-applicab.
..----- NOTE ---------------
--LCO 3.0.3 s not applicable.
._,.,....,------,--.-..,--,-.-_
-. .--..---.-...
... ..--- .. .-_ .. .. .. .. --- -- .. .... ..-.- ..... ...CONDITION REQUIRED ACTION COMPLETION TIME A. The gaseous radwaste from the main condenser air ejector system is being discharged without treatment.
B. Required Action and associated Completion Time'not met.A. 1 :Restore treatment of gaseous radwaste effluent.7 days 30 days B. I Prepare and submit to the. NRC, pursuant to D 4, 1., a Special Report that includes the following:
(1) Identification of any nonfunctional equipment
- or subsystems and the reason for, the,.nonfunctionality, (2) Action(s) taken to restore the nonfunctional cquipment to FUNCTIONAL status, and (3) Summary description of action(s) taken to prevent a recurrence.
Unit 2 Revision 28 September 2006 ,13.2-10 Gaseous Radwaste Treatment System D 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.4.1 Check the readings of the relevant instruments to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensure that the GASEOUS RADWASTE TREATMENT SYSTEM is: functioning.
Unit 2 Revision 28 September 2006 13.2-11 Ventilation Exhaust Treatment System D 3.2.5 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D*3.2.5 Ventilation Exhaust Treatment System DLCO 3.2.5 The VENTILATION EXHAUST TREATMENT SYSTEM shall be FUNCTIONAL.
APPLICABILITY:
Atall times, ACTIONS---------
N O TES ------------------------------------------------
- 1. LCO 3.0.3 is not applicable.
- 2. LCO 3.0.4 is not applicable..
CONDITION REQUIRED A.CTIOP COMPLETION TIME A. The radioactive gaseous waste is being discharged without treatment.
AND Projected doses in :31 days from iodine and particulate releases, from each unit, to areas at or, beyond the SITE BOUNDARY (see Figure D 1.0-1) would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.A.A Prepare and submit to the NRC, pursuant to D.4..J1, a Special Report-that includes the following:
(1) Identification of any nonfunctional equipment or subsystems and the reason for the nonfunctionality, (2) Action(s) taken to restore the nonfunctional equipment to FUNCTIONAL status, and (3) Summary description of action(s) taken to prevent a recurrence.
30 days Unit 2 Revision 28 September 2006 13.2-12 Ventilation Exhaust Treatment System D 3.2.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.5.1-NOTE ------------------------
Only required to be met when the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.Project the doses from iodine and particulate releases from each unit. toareas at or beyond the SITE BOUNDARY.31 days: Unit 2 Revision 28 September 2006 13.2-13 Venting or Purging D 3.2.6 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2,6 Venting or Purging DLCO 3.216 VENTING or PURGING of the drywell and/lor suppression chamber shall be through the standby gas treatment system.APPLICABILITY:
MODES 1, 2, and 3., ACTIONS.......--------------------
NOTES ---------..........
- 1. LCO 3.0.3 'is not applicable.
- 2. LCO 3.0.4 is not applivable.
CONDITION A. VENTING or PURGING of the drywelH and/or suppression chamber not through the standby gas treatment system.REQUIRED ACTION COMPLETION TIME Immediately
+A.I Suspend all VENTING and PURGING ofthe drywell and/or suppression chamber.Unit 2 Revision 28 September 200.6 1.3.2 Venting or Purging D 32.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.6.1 The drywell and/or suppression chamber shall be determined to be aligned for VENTING or PURGING through the standby gas treatment system.Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before start of VENTING or PURGING AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter during VENTING or PURGING Unit 2 Revision 28 September 2006 13.2-15 Radioactive Liquid Effluenl Monitoring Instrumentation D3.3.1 D13.3 INSTRUMENTATION D 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation DLCO3.3.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table D3.3.1-1 shall be FUNCTIONAL with: a. The minimum FUNCTIONAL channel(s) in service.b. The alarm./trip setpoints set to ensure *that. the limits of DLCO. 3. 1.1 are not exceeded.APPLICABILITY:
According to TableD* 3.3.1-1.ACTIONS------------..----------------------------------------
N OTES 1. LCO 3.0.3 is not applicable.
- 2. Separate condition entry is allowed for each channel.CONDITION A. Liquid effluent monitoring instrumentation channel.alarm/trip setpoint less conservative than required.REQUIRED ACTION COMPLETION TIME 1-A,I Suspend the release of radioactive liquid effluents monitored by the affected channel.rImmediately OR A.2 Declare the channel nonfunctional.
ORR A.3 Change the setpoint so it is acceptably conservative.
Immediately Immediately (continued)
Unit 2 Revision 28 September 2006 1 3.3-1 ACTIONS_(continued)
COD.ITI.O.N.....
.........CONDITION Radioactive Liquid Effluent Monitoring Instrumentation D 33.1 REQUIREDACTION COMPLETION TIME B. On"e: or more required channelh nonfunctional.
B. 1 Enter the. Corndition referenced in Table D 3.3.1-1 for the channel.Immediately AND B.2 Restore nonfunctional channel(s) to FUNCTIONAL status..30 days C. As required by Required Action B.1 and referenced in Table D 3.3.1-1.C.1 Analyze at least Z independent samples in accordance with Table D 3.1 .1- 1.AND C.2--------NOTE---------
Verification Action will be performed by at least 2 separate technically qualified members of the facilitystaff.
Prior to-initiating aýrelease Independently verify the release rate calculations and discharge line valving.Prior to initiating a release D. As required by'Required Action B. I and referenced in Table D 3.3.1-1.D.1 Collect and. analyze grab samples for radioactivity at a limit of detection of at least 5 x 1 0 Ci/ml.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter (continued)
Unit 2 Revision 28 September 2006 I .3.3-2 Radioactive Liquid Effluent MOnitoring Instrumentation D 3.3.1.ACTIONS (continued)
_ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _CONDITION E. As required by Required Action P.1 and referenced in Table D 3.3.1-1.REQUIRED ACTION COMPLETION TIME El 1 ...........
NOTE --........-
Pump performance curves generated.in place may be used to estimate flow.Estimate:
the flow rate during actuaI releases.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> AND Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter F. As requiredby Required Action B.1 and referenced in Table D 3.3. 1-1.G. Required Action B.2 and associated Completion Time not met.F. 1 Estimate tank liquid level.Immediately AND-During liquid additions to the tank In accordance with Radioactive Effluent Release Report.G. I. Explain in the next Radioactive .Effluent Release Report why the nonfunctionality was not corrected in. a-timely manner.H. Required Action and H. I Suspend liquid effluent Inmiediately associated Completion releases monitored by the Time for Condition C, D, nonfunctional channel(s).
or E not met.I. Required Action and 1.1 Suspend liquid additions to Immediately associated Completion Time the tank.monitored by the for Condition F not met. nonfunctional channel(s).
Unit 2 Revision 28 September 2006 13.3-3 Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 SURVEILLANCE REQUIREMENTS
-.--------------
NOTE--r .........------
I -------------
Refer to Table D 3.3,. i-I to determine which DSRs apply for each function.SURVEILLANCE.
IRFRQUENCY 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DSR 3.3, 1. 1 Perform CHANNEL CHECK.-f DSR 3.3.1.2 Perform CHANNEL CHECK by verifying indication of flow during periods of release.Perform SOURCE CHECK.DSR 3.3.1.3 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on any day on which continuous, periodic,*
or batch releases are made Prior to release 31 days 31 days DSR 3.3.1.4 Perform SOURCE CHECK.DSR 3.3.1.5 Perform CHANNEL FUNCTIONAL TEST. The CHANNEL FUNCTIONAL, TEST:shall.
also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the alrrni/trip setpoint; and control room alaxm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, or instrument controls not set in operate mode.DSR 3.3.1.6 Perform CHANNEL FUNCTIONAL TEST.92 days (continued)
Unit 2 Revision 28 September 2006 1 3.3-4 Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY 184 days DSR 3.3.1.7 DSR 3.3.1.8 Perform CHANNEL FUNCTIONAL TEST. The CHANNEL FUNCTIONAL TEST shall also demonstrate control.room alarm annunciation occurs fdr instrument indication levels measured.
above the alarm setpoint, circuit failure, instrument indicating a downscale failure, or instrument controls not set in operate mode.Perform CHANNEL CALIBRATION.
Theinitial 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 NIST standards, or using actual samples ofliquid effluents that have, been.analyzed on'a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its inLended range of energy and ineasurement.
For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.4 18 months 18%months DSR 313.1.9 Pertorm CHANNEL CALIBRATION.
Unit 2 Revision 218 September 2006 I 3.3-5 Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 Table D 3.3-1-1 (page 1 of 1)Radioactive Liquid Effluent Monitoring Instrumentation APPLICABILITY REQUIRED CONDITIONS OR OTHER CHANNELS REFERENCED SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTION B. I REQUIREMENTS Radioactivity Monitors Providing Alarm and Automatic.Termination of Release Liquid Radwaste Effluent (a) I C DSR3:3.t.1 Line DSR DSR 3.3,1.5 DSR.3.3. 18 2. Radioactivity Monitors Providing Alarm but not Providing Automatic Termination ofRelease a. Service Water Effluert Line A b. Service Water Effluent Line B (a)(a)I D D b DSR 3.3. t.I DSR 3.31.4 DSR 3.3.1.7 DSR 3.3. 1.8 DSR 3.3,1.1 DSR 3.3.1.4 DSR 3.3.1.7 DSR.3.3.1,8 DSR 3 '3. I.1 DSR'3.3,1.4 DSR 3.3,1,7 DSR 33.1.8 c. Cooling oIower Blowdown Line 3. Flow Rate Measurement Devices a. Liquid Radwaste Effluent Line b. Service Water. Effluent I. ine A c. Service Water Effluent Line B d. Cooling Tower Blowdovn Line (a)(a)(a)(a)1 1 E E DSR 3.3.1.2 DSR 3.3. .6 DSR 3.3.1.9 DSR:3.3.1.2 DSýR3.3.-1.6 DISR 3,3.1.9 DSR 3.3. 1-2 DSR 3.3.1.6 DSR3.3.11.9 DSR 3.1.2 DSR'3.3. 1.6 DSR,<3'3.1.9 E 4- Tank Level Indicating (b) I F DSR 3.3. I, 1 Dev ices (c) DSR 313.1.6 DSR 3.3.1.9 (a)(b)(c)During releases via this pathway, During liquid addition to (he associated tank.Tanks included in this DLCO are:thoseioutdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have jaisk overtflows and siarrcunding area drains connected to the liquid:radwaste treatment system, such as temnporar, tanks.Unit 2 Revision 28 September 2006 13.3-6 Radioactive Gaseous Effluent Monitoring.
Instrumentation D 33.2 D 3.3 INSTRUMENTATION D 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation DLCO 3.3.2 The radioactive, gaseous effluent monitoring instrumentation channels shown in Table D 3.3.2-1 shall be FNCTIONAL with-a. The minimum FUNCTIONAL channel(s);in service.b. The alarm/trip setpoints of Offgas Noble Gas Activity Monitor set to ensure that the limit of Technical Specification LCO 3.7.4 is not exceeded.c. The alarnVtrip setpoints:
of Radwaste/Reactor Building Vent Effluent Noble Gas .Activity Monitor and Main Stack Effluent Noble Gas Activity Monitor set to ensure that the. limits of DLCO 3,2.1 are not exceeded.APPLICABILITY:
According to Table D 3.3.2-1.ACTIONS, S ---------------
NOTES -----------------------------------
L. LCO 3.0.3 is not applicable.
- 2. Separate'condition entry is allowed for each channel.CONDITION REQUIRED ACTION COMPLETION TIME A. Gaseous effluent A. Suspend the release of Immediately monitoring instrumentation radioactive gaseous effluents channel alarmitrip; setpoint monitored by the affected less conservativethan channel.required.OR A.2 Declare. the channel Immediately nonfunctional.
Immediately A.3 Change the setpoint so it is acceptably conservative.
I I ,(continued)
Unit 2 Revision 28 September 2006 1 3.3-7 Radioactive Gaseous Effluent M0nitoring Instrumentation D 3.3.2 ACTIONS (continued)
CONDITION REQUIRED.
ACTION COMPLETION TIME B. One or more channels nonfunctional.
B. I Enter the Condition referenced in Table D 3.3.2-1 for the channel.AND B,2 Restore nonfunctional channel(s) to FUNCTIONAL status.Immediately 30 days C. As 'required by ReqUired Action B. I and referenced in Table D 3.3.2-1.C. I Place the nonfunctional channel in the tripped condition.
OR C.2.1 Take.grab samples.AND C.2,2 Analyze.samples for gross activity.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 12 hours AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time: of sampling completion (coWt i-nu ed)Unit 2 Revision 28 September 2006 13.3-8 Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D. 1 Estimate the flow rate for the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Action 8.1 and referenced nonfunctional channel(s).
_in Table D 3,3.2-1. 'AND Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter E. As required by Required Action B.1 and referenced in TableD D3.3.2-,.E.1 Continuously collect samples using auxiliary sampling equipment as required in Table D 3.2.1-I.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 12 hours F. As required by Required, Action B.1 and referenced in TableD 3.3.2-I.F. 1.1 Take grab sample.AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of sampling completion AND F. 1.2 AND Analyze samples. for gross activity with.a radioactivity limit of detection of at least I x 10-4 Ciml. .Restore the nonfunctional channel(s) to FUNCTIONAL status.F,2.1 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> I OR F.2.2 Through a CR, determine: (i) The cause(s).of the nonfunctional.
(2) The actions to be taken and the schedule for restoring thesystem' to FUNCTIONAL status.14 days (continucd)
Unit 2 Revision 28 Septernber 2006 13.3-9 Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (continued)
CONDITION REQUIRED ACTION G Required Action B.2 and associated Completion Time not met.I1 Required Action and associated Completion Time for Condition C, D, E orTF.l not.met.G. 1 Explain in the next Radioactive Effluent Release Report why the nonfunctionality was not corrected in a timely manner.11.1 Suspend gaseous effluent releases monitored by the nonfunctional channel(s).
COMPLETION TIME In accordance with Radioactive Effluent Release Report frequency Immediately Unit 2 Revision 28 September 2006 1 3.3-10 Radioactive ,jaseous, Effluent Monitoring Instrumentation D3..3.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.23 Perform CHANNEL CHECK. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DSR 3.3.2.2 Perform CHANNEL CHECK. 7 days DSR 3.1.2.3 Perform SOURCE CHECK. 31 days DSR 3.3.2.4 Perform CHANNEL FUNCTIONAL TEST. The 31 days CHANNEL FUNCTIONAL TEST-shall.
also demonstrate the automatic isolation capability of this pathway .and that control room alarm annunciation occurs if the instrument indicates measured levels above the alarm/trip setpoint (each channel will be tested independently so as to not initiate isolation during operation);
and control room alarm annunciation occurs for instrument indication-levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, and instrument controls not set in operate mode.DSR 3.312.5 Perform CHANNEL FUNCTIONAL TEST. 92 days DSR 3.32.6 'Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate control room alarm annunciation occurs.for instrument indication levels measured above the alarm setpoint, circuit failure,.
instrument indicating a downscale failure, and instrument controls not set in operate mode.(continued)
Unit 2 Revision 28 September 2006 13.3-11 Radioactive Gaseous. Effluent Monitoring Instrumentation D 3.3.2 SURVEILLANCE REQUIREMENTS continued)_
SURVEILLANCE FREQUENCY DSR 33.2.7 Perform CHANNEL CALIBRATION, The initial 24 months CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the ld l 11 A I-L. ; Lk I allIU aU l %A 1 -.k (NIST) or using standards that have, been obtained from suppliers that participate in measurement assurance activities with NIST, or using actual samples of gaseous effluents that have been analyzed on a system thathas been calibrated with NIST traceable sources. These standards shall permit calibrating, the system over its intended range of energy and measurement.
For subsequent CHANNEL CALIBRATION, sources, thathave been related to the initial calibration may be used..The CHANNEL CALIBRATION shall also demonstrate that automatic isolation of this.pathway occurs when the instrument channelsintdicate measured levels above the Trip Setpoint.ALIBRATION.
18 months DSR 3.3.2.8 DSR 3.3.2.9 Perform CHANNEL Perform CHANNEL CALIBRATION.
The initial CHANNEL. CALIBRATION shall be performed using one or more of the reference standards certified by, the National Institute of Standards and Technology (NIST)or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST, or using kactual samnples of gaseous. effluents-that have .been analyzed on a system that has been calibrated withNIST traceable sources. These standards shall permit calibrating the system over its intended range of energy and .measurement.
For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.18 months DSR 3.3.2.10 Perform CHANNEL CALIBRATION.
24 months Unit 2 Revision 28 September 2006 I 33-12 Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 TableD 3.3:2-1 (page I of 2)Radioactive Gaseous Effluent Monitoring Instrumentation REQUIRED CONDITIONS APPLICABILITYOR CHANNELS REFERENCED OTHER SPECIFIED PER FROM RIEQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTION B. I REQUIREMENTS Offgas System a "Nole Gas (a) *2: C DSR3.3.2,1 Activity Mlonitor DSR 33,2.4-Providing DSR 3.3.2.7 Alarm and Automatic Terminat ion of Redease.b. System Flow- (a) ID DSR 33,:2.1 Rate. Measuring DSR 3.3.2.5 Device. DSR 3.32.10 2a). D DR 33.2.1 C. Sample Flow- DSR 3.3125 Rate Measuring DSR33:32.10 Device 2. Radwaste/Reactor Building Vent Effluent System a Noble Gas Cb) F DSR 3.3.2.1 Activity Mionitor DSR 3.3.2.3 (e) DSR3'3.2326 DSR 3,3.2.9 b. lodine Samplerý (b). .E DSR 3.3:2.2 C, Particulate (b) 1 E DSR 3,32.2, Sampler d. Flow-Rate., (b) D DSR 3.3.2.1 Monitor USK 3.3.2,5 DSR 33.2.8.e. Sample Flo%-' (b) D DSR 3.3.2.1 Rate Monitor DSR 3.3.2,5 DSR 3.3.2.8.(continued).(a) During olfgas systemroperation.(b) At all times.(c) Includes high range noble gas monitoring capability.
Unit 2 Revision 28 September 2006 13.3-13 Radioactive Gaseous Effluent Monitoring Instrumentation D3.3.2 Table .D 3.3.2-1 (page 2 of 2)Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY OR.OTHER SPECIFIED CO.NDITIONS REQUIRED CHANNELS PER lNSTrR.UJMENTr CONDITIONS REFERENCED FROM REQUIRED A.CrON R.l SURVEILLANCE REQUIRF.MENTS INSTRUMENT
- 3. Main Stack E'lluent a. Noble Gas Activity Monitor (c)b Iodine Sampi¢c. Particulate Sampler d. Flow-Rate Monitor e,. Sample Flow-Rale Monitor (b>(b).(b)(b)F E D D MSR 3.3,21 DSR 3.3.2.3 DSR 3.3.2.6 DSR 3.3.29 DSR 3.3.2.2 DSR 3.3.2-2 DSR 3.312.1 DSR 3.3,2.5 DSR 3.3 2 9 DSR 3.3.:2.1 DSR 3.3.2.5 DSR 3.3.2.8 (b) At all fiijnts.(c) ,lndludes high range noble gas monitoring capability.
Unit 2 Revision 28 September 2006 13.3-14 Radioactive Effluents T1tal, Dose D 3.4 D 3.4 D 3.4 RADIOACTIVE.
EFFLUENTS TOTAL DOSE Radioactive Effluents Total Dose: DLCO 3.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium tijelcycle sources shall be limited~to 25 mrem to the Whole body or any organ, except the thyroid, which-shall be limited to 75: mrcm.APPLICABILITY:
At all times.ACTIONS-------- ------- --- ----........ ----- -IN- -1. LCO 3,0.3 is not applicable.,-----------------------------------------------2. LCO 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Estimated dose:.or dose commitment due to direct radiation and the release of radioactive materials in liquid or-gaseous effluents exceeds.the limits'A. 1 Verify the condition resulting in Immediately doses exceeding:these limits has been corrected.
B. Required Action anid associated Completion Time not met.B. 1 -----------
NOTE -----------
This is the Special Report required by D 3.1.2, D 3.2.2, or D 3.2.3 supplemented with the following,----- ---- ----- ---- -- --- --Submit a Special Report, pursuant to D 4.1. L, including a request fora variance in accordance with the provisions of 40 CFR 190. This submission is considered a timely request, and a~variance is granted until staff action on the request is complete.30 days Unit 2 Revision 28 September 2006!13.4-1 Radiological Environmental Monitoring Program D 3.5.1 D'3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING D 3.5.1 Monitoring Program DLCO 3.5J The Radiological Environmental Monitoring Program shall be conducted as specified in Table D 3.5.1 -1.APPLICABILITY:
At all times.ACTIONS---------
... ..--------.....
.. ....-------------------------
N O TES ------------------------------------------------
J. LCO3.0.3 is not applicable.
- 2. LCO13.0.4 isnot applicable.
CONDITION A. Radiological Environmental A., I Monitoring Program not i conducted as specified in I Table D 3.5.1-1. I REQUIRED ACTION Prepare and submit to the NRC n the Annual Radiological Environmnental Operating Report, a description of the easons for not conducting the programn as requited ,and the plans for preventing a ecurrence.
COMPLETION TIME In accordance with the Annual Radiological Entvironmuental Operating Report frequency r B. Level of radioactivity in an environmental sampling medium at a specified locatiohi exceeds the reporting levels of Table D 3.5.1-2 when averaged over any calendar quarter.OR B. I -----------
NOTES.--"..------
I. Only applicahle if-the radioactivity/racdionuclides are the result of plant effluents.
- 2. For radionu; clides other than those in Table D 3.5.1-2, this rcport shall indicate the methodology and parameters used to estimate the potential annufal dose to a MEMBER OF. THE PUBLIC.(continued)
Unit 2 Revision 28 September 2006 1 3.5-1 Radiological Environmental Monitoring Program D3.5.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION I TIME+/-ýMore than one of the radionuclides in Table D 3.5.1-2 are detected in the environmental sampling mediunmand Concentration I +reporting level 1 concentration 2 + ... >_ 1.0.reporting level .2 OR Radionuclides other than those in Table D 3.5,1-2 are detected in an environmental sampling medium at a specified location which are the result of plant effluents and the potential annual dose.to a MEMBER OF THE PUBLIC from all radionuclides is. the calendar year limits of D 3.1.2, D 3.22 or D3.2.3.Prepae: and submit to theNRC, pursuant to D 4. 1..1, a Special Report that (1) Identifies the cause(s) for.exceedingthe limit(,s)and (2) Defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OFITHE PUBLIC is less than the calendar year limits of D 3.1.2, D 3.2.2, or D 3.2.3.OR B.2 .--- NOTES--------
I .Only applicable if the radioactivity/radionuclides are not the result of plant effluents.
2.For radionuclides&
other than those in Table D 3.5.1-2, this report shall indicate the methodology and parameters used to .estimate the potential annual dose to a MEMBER OF THE PUBLIC.Report and describe the condition in the Annual Radiological Environmental Operating.Report.
30 days In accordancc with the Annual Radiological Environmental Operating Report frequency (continued)
Unit 2 Revision 28 September 2006 i 3.5-2 Radiological Environmental Monitoring Program D 3.5.1 ACTIONS (continued)
CONDITION-1-REQUIRED ACTION COMPLETION TIME C. Milk or fresh leafy vegetation, samples unavailable from one or more of ihe sample locations required by Table D 3.5.1 -1.D. Environmental samples required inTable D 3.5.1-1 are. unobtainable:
due to sampling equipment malfunctions.
C. 1 Identify specifiC locations-for obtaining replacement samples and add them to the Radiological Environmental Monitoring Program.AND C.2 AND C.3 Delete the specific locations from which samples Were unavailable
'from the Radiological Environmental Monitoring Prograrn.Pursuant.
to Technical Specification 5.6.3, submit in the next Radioactive Effluent Release Report documentation for a change in the ODCM reflecting the new location(s) with Supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location(s) for obtaining samples.30 days ,30 days 1n accordance with the Radioactive Effluent Release Report D. I Ensure all efforts are made to complete.
corrective action(s).
AND D.2 Report all.deviations.from the sampling schedule in the Annual Radiological Environmental Operating Report.Prior to the end of the next sampling period In accordance with the Annual Radiological Environmental Operating Report (continued)
Uniit 2 Revision 28 September 2006 I 15-3 Radiological Environmental Monitoring Program D 3.5.ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME i E. Samples required by Table D 3.5.A.1 not obtained in themedia of choice, at the most desired location, or at the most, desired time.E. I Choose suitable alternative media and locations for the pathway in question.AND E.2 Make appropriate substitutions in the Radiological Environmental Monitoring Program.AND E.3 Submit in the next Radioactive Effluent Release Report, documentation for a change in.the ODCM reflecting the newlocation(s) with supporting information identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location(s) for obtaining samples, 30 days 30 days In accordance with the Radioactive Effluent Release Report Unit 2 Revision 28 4September 2006, 1 3.5-4 Radiological Environmental Monitoring Program D 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.1 Collect and analyze radiological environmental In accordance with monitoring samples pursuant to the requirements of Table D 3,5.1-1 Table D 3.5. I -1 and the detection capabilities:
required by TableD 3.5.1-3.Unit 2 Revision 28 September 2006 13.5-5 Radiological Environmental Monitoring Program D 1..5.1 TFable D 3.5.1-1 (page 1 of 4)Radiological Environmental Monitoring Program.EXPOSURE NUMBER OF SAMPLING AND PATHWAY SAMPLES COLLECTION TYPE AND FREQUENCY AND/OR STATIONS SAMPLE FREQUENCY OF ANALYSIS SAMPLE LOCATIONS (a)I. Direct 32 routine (1) Aninner ring of stations, Once per 3 months Gamma dose: once per 3 Radiation monitoring one in each months stations (b) mcetorological sector in the general area of the SITE BOUNDARY (2) An out'r ring Of stations, one in each]and base Imeteorological sector in the 4:to 5 mile (c) range from,the site (3). The balance&ofihe stations sliould be placcd in ,specialfinterest areas such as population centers.nearby residences, schools,,and in one or two areas to serve as control stations(d)
- 2. Airborne *5locations (I) 1samples fromnoffsite Continiuoussampler Radioiodine cariister:ý,Radioiodine locations close to the site operation with Analyze weekly for I-131 and boundary (within I mile) sample 'collc-tion Particulates in different sectors (e) weekly or more Pariculate sampler: (2) s amipl]efrom the. vicinity frequently if (I) Analyze for gross beta of an established year- required by dust radioactivity
_ 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> round community (e) loading. following fiter change (3). 1 sumplo.ioa ac.ntrol (2) Perform gamma isotopic (3). l *i atpleaftomae mileslanalysis on each sample loctantaion, atleast mil(g) in which gross beta distantaand in a least etivity i3 > 10 times the Prevalent wind direction previous yearly mean of (d) control sampie5.(3) Gamma isotopic analysis of composite smxiple (g)(by once per 3 months 3- Waterbome a. Suftace I sampl!e Upstream.(d).(h)
Composite sample. (0) Gamma isotopic overa ofic'month analysis of each sample period (i) (g) once per month I. sample Site's dowinstreamr 6ooling (2) H-3 analysis of water intake (h) each composite sample and once. per 3 months b. Ground As required From oneof two-soturces if Grab samplef uie (t) Gamma isotopic likely to be atflcted () per 3 months analysis of each sample.(g) once per 3 months (2) H-3 analysis of each sample ohce per 3'months (continued)
Unit 2 Revision 28, September 2006 13.5-6 Radiological Envirotmentai Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 2 of 4)Radiological Environmental MonitorinIg Program EXPOSURE PATHWAY SAMPLING AND AND)ORý NUMBER OF SAMPLE COLLECTION TYPE ANDFREQUENCY SAMPLE SAMPLES LOCATIONS.(a)
FREQUENCY OF ANALYSIS 3. Waterbome (continued),c, Drinking .1 sanmpiel Ieach, One to three oflthe nearest. When analysis (1) 1-131 analysis on watersupplies that could be is performed, a each composite sample.affctcd bhy is di ccree (kM composite sample when the dose over aitWo week, calculated for the period (i); otherwise, consumption of the a composite sample water is greater than I monthly. mremlyr (I)*(2) Gross beta and gamma isotopic, analyses of each composite sample (g)monthly (3) l-34asialysis of each composite qample d. Sediment I sample From a downstream area with Twice per year once per 3 months from existing or potential recreational Shoreline value Gamrma isotopic analysis of each sample(g)4. Ingestion a, Milk (1) 3 samples fromn l[n3 locations.withirt 3.5 mites Twice per month, (I) Gamma isotopic MILK (e) Api thruuh. (g) wid 1-131 analysis of SAMPLING Dec-ember (m) each sample twice per LOCATIONS month April through December (2),If there alr In each of3'aie-vis 3.550 mile.s (2) Gamma isotopic none;, distant (e) (g) and 1-13 Ianalysis of then I sample each sample once per from MILK month January through SAMPLING March if required LOCATIONS At a control location 9-20 miles (3) :1, sampie from a distant and in a leasi prevalent MILK wind direction (d)SAMPLING LOCATION b, Fish (1) 1 sample~each In the,vicihity of a plant Twice per year Gamma isotopic analysis of of discharge areta each sample (g) on edible 2 commereially portions wice per. year or recreationally important species (n)(2) 1 sample of the In areas not influenc.d by.wame spcc ies station dischaegc (d)(continued)
Unit 2 Revision 28 September 2006, 1 3.5-7 Radiological Environmental Monitoring Program D 3.5.:1 Table D 3.5.1-1 (page 3 of 4)Radiological Environmental Monitoring Program EXPOSURE PATHWAY SAMPLING AND TYPE AND FREQUENCY AND/OR NUMBER OF SAMPLE COLLECTION OF ANALYSIS SAMP E, SAMPLES LOCATIONS (a): FREQUENCY 4. Ingestion (continued)
- c. Food (1) 1 sample of Any area that is. iirigated by Attimne ofharvest Gamma isotopic (g) and I-Products each principal water in which liquid plant (p) 131 analysis of each sample class af food wastes have been discharged (o) of edible portions pioduota (2) Samples of 3 Grown nearest to each of 2 different kinds different offsite locations (e) Once per.year during of broad leaf the harvest season-vegetation (such as vegesables)
Grown at lea§t 9.3 miles'distant (3) I. sample of in a least prevalent wind Once per year during each of the direction the harvest season similar broad.leaf vegetation, Unit.2 Revision 28 September 2006 13.5-8 Radiological Environmental Monitoring Program D3.5.1 Table D .35. .1-1 (page 4 of 4)Radi0logical EnVironmental Monitoring Program (a) Specific pararetcrs of distance and direction sector from the centerline of one reactor, and.additional deseriptions wh&ýpertinent, shall be provided for each and every sample location in Table 1D3.5.1-1 Referio" NUREG-.0133,."PreSaration of Radiologinal Effluent.Technical Specifications for Nuclear Power Plants, October 1978 and to Radiological Assessment Branch .Technical Position on rnvironmemal Monitoring, Revision I\November 1979. Deviations arc permitted from the requifed sampling schedule ifsspecimenris are unobtainable because 9f such circumstances as hazardous conditions, seasonal..unavmailability (whicuh theft uand uricoutpctiivc rtidni).ui t5ialfuiLti "uti uturrotic stimplifig equipinilltL (b) One or more instruments, %uch as a pressurized ion chamber, for me-asuring and recording dose rate continuously may be used in place of, or in addition to integiating dosimeters.
Each of the 32.routine monitori-ng stations shall be equipped with 2 or more dosimeters or with I instriment for measuring and recording dose rate continuously.
For the purpose of this table, a thermioluminescent'dosimeter (I.LD) is considered to be one phosphor; 2 or more phosphors in a packet are considered as 2 or more dosimeters.
Film badges shall not be used asadosimeters for measuring direct -radiation.(e) At this distance,8 windrose sectors (W, WNW, NW, NNW, N, NNE' NE\ and ENE) are.over Lake Ontario.(d) The purpose of these.samples is to obtain background information.
Iti i is not practical t1 Establish control I6eations in accordance with the distance and wind tirection criteria, other sites, which provide valid background data; may be substituted..(e) Having the.highest calculated annual site average ground-kevel D/Q based on all site licensed reactors.(:1) Airborne particulate sample filters shall be analyzed for gross'beta activity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more aflt& sampling to allow for radon and thoron daughter decay.(g) Gamma isotopic analysis means the identification and quantification of gamma -emitting radionuclides that may be attributable to the'effluents from thefacility.(b) The upstream sample shall be taken at a distance beyond significant influence ofthe discharge:
The downstream sample shall be taken in an area beyond but near the mixing zone.(i) In this program, repregentative composite sample aliquots shall be collected ati imieintervals:
that are viey shon (e.g., hourly)relative to the compositing period (e.g.,:monthly) in order to assure obtaining airepresentative sample.(j) Groundwater samples shall be taken when this source is tapped for drinking or irrigption purposes in. areas where the hydraulic gradient or recharge properties are suitable for contamination.(k) Drinking water samples shall-be.taken only when drinking water is a dose pathway.(1) Analysis for 1-I131 may be accoiniplished by Ge-.Ii analysit:provided that the'bowerl imit of detection (LLD) for 1-131 in water samples foundlon.Table D 3;5;1-2 can &e met. Doses shall be the maximum organ rnd age group.(T) Sampleswill be collected January through March if 1-131 is detected in November and December of the preceding year..(n) In the event 2 commercially or recreationally importantspecies are not available.
after 3 attempts of collection, then 2 samples or onespecies oir other species not necessarily comnercially or recreationaill important may.be'utilized.
(6) Applicable only to.major irrigation projects within 9 miles of the site in tile general downeurrent dirdetion.
.(p) If harvest occurs more than once/year, sampling'shallibe performed during eachdiscrete harvest. If harvest occurs continuously, sampling shallbe taken monthly. Attentionishould be paid to including samples-oftuberous and root f6od products;Unit 2 Revision .28 I3.5-9 September 2006 Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-2 (page 1 of I)Reporting Levels for Radioactivity in Environmental Samples AIRBORNE FOOD RADIONUCLIDE PARTIUCLATEOR FISH MILK PRODUCTS ANALYVSIS WATER (pCiIL) GASES'(pCIlt,) (pCi/kg, wet) (pCi/I.) (pCi/kg, wet)H1-3 Mn-54 Feý-59 Co-60 Z11-65 Zr-95 Nb-05 1-131 Cs-134 Cs-137 L) a-. 1 40.20,000 (a)1,000 400 1,6000 300 300 400 400 2 (b)30 ,50 30,000 10,000 30,000 10.000 20,000 0.9 10:20 I N0 1,000 2,000.60, 70 1,000 2,000 200 200 300 300 (6) Fordrinking samples. This, is a'40 CFR 141 v.lue. If/no diihking watr pathway exists; a value of 30,000 pCi/L may be used.(6) Inodrinkngvaerpathw i ae of 20 pC/I. maybe used;Unit2 Revision 28 1 3.5-10 Scptember 2006 Radiological Environmental Monitoring Program D 3.5. 1 Table D 3.5.1-3 (page I of 2)Detection Capabilities for Environmental Sample Analysis (a) (b)LOWER LIMIT OF DETECTION (1-.AIRBORNE PARTIUCLATE OR FOOD RADIONUCLIDE WATER GASES (pCi/m 3) , FISm MILK PRODUCTS SEDIMENT.ANALYSIS.i (pCi/L) (pCi/kg, wet) (pCi/L) (pCi/kg, wet) (pCi/kg, dry)Gross Beta H-3 Mn-54 Fe-59 Co-58 Co-60 Zn-65 zr-95, Nb-95 1:13.1 C.,-,34 Cs- 137 4 2,000 !d)1:5 30 15 15 30 15 1 5 115 Is 0.0O 130 260 130 130 260 0,07.0,05 0.06 I 130 150 15 18 60 60 80 150 Ba-140 i5 15 See the notes on the next page IS 15 Unit2 Revision 2.8 September 2006 13.5-11 Radiological Environmental Monitoring Program D 3.5.1 Table 3.5.1-3 (page 2 of 2)Detection Capabilities for Environmental Sample Analysis )(a) This list does not mean that only. these nuclides are to be considered.
Other peaks that are identifiable, together with those of Whe above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report.(b) Required detection capabilities for thermoluminescent, dosimeters used for environmental measurements are given in ANSI N-545 Section 4 3 197D. Allowable cxceptions to ANSI 4545, Section 4.3 arc contained in the ODCM (c) The.LLD is defined as the smallest coneentrationof radioactive material in a sample that will -yield anfet count, above. system background, that will be detected with 95% probability with only 5% probability of falsely concluding,that a blank observation represents a 4real" signal.For a particular measurement system, which may include radiochemical separation:
LLD (4-66)(Sb (E) (V) (2.22) (Y) e where: LLD The before-the-fact lower limit ofdetection (pCi per unit mass or Volume), Sb, The standard deviation of the background counting rate or of the ctunting rate of ablank sample as appropriate (counts per minute), E The counting efficiency (counts per disintegration), V The sample size (6nits of mass~or vol ume), 2.22 The number of disintegrations per minute, per pCi, y -The fiactional radiochcmical yield, when applicablc,= The radioactive decay constant for the particular radionuclide (sed'lt and At The elapsed time between environmental collection or end of the sample collection period, and the time ot counting (seconds)Typical values of V, Y, and At should*be used inthe calculation.
it should be recognized that the LLD is defined as-a beforothe'fact limit~representing the capability of a measurement system and not as an after-the-fact limit fora particular measurement.
Analyses shall be performed in such amanner that the stated LLDs will be achieved under routine conditions, Occasionally background fluctuations, unavoidable small sampksizes, the presence of interfering nuclides, or othier uncontrollable eir-cumstancesmay render-these LLDs unachievable.
In such cases, the contributing factors'shall be identified andldescribed in the Annual Radiological Environmental Operating Report.(d) If no drinking;water.pathway exists, a value of 3,000 pCi/L maybe used.(e) If no drinking wateirpathwhy.exists, a value of 15 pCiJL may be used, Unit 2 Revision 28 13.5-12 September 2006 Land Use Census D.3.5.2 D 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING D 315.2 Land Use Census DLCO 3.5.2 A land usecensus shall: a. Be conducted, b. Identifywithin a distance of 5 miles the location in each of the 16 meteorological sectors of the nearest milk animal and the nearest residence, and the nearest garden (broad leaf vegetation sampling controlled by Table D 3.5.1 -1 , part 5.c may be performed in lieu of the garden census) of> 500 ft 2 producing broad leaf vegetation, and c. For elevated releases, identify within a distance of 3 miles the locations in each of the 16 meteorological sectors of all milk animals and all gardens (broad leaf vegetation sampling controlled.by Table D 3.5. 1, part.5,c may be performed in lieu of the garden: census) > 500 ft producing broad leaf vegetation..
APPLICABILITY:
At all times.ACTIONS 1. LCO 3.0-3 is-not applicable.
- 2. LCO 3.0.4 is not applicable.
NOTES ------. ..---- -------.--------
-.: -------------------------------------
I -----------------------------------------------------------------
CONDITION A. Land use census identifies location(s) that yields a calculated dose, dose commitment, or D/Q value>.than the values currently being calculated in DSR 3.2.3.1.REQUIRED ACTION COMPLETION TIME A. 1 Identify the new location(s) in the next Radioactive Effluent Release Report.In accordance with the Radioactive Effluent Release Report (continued)
Unit 2 Revisionr28 Scptcmber 2006 13.5-13 Land Use Census D 3.5.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Land use census identifies location(s) that.yields a calculated dose, dose, commitment, or D/Q value (via the same exposure pathway) 50% > than at: a location from which samples are currently being obtained in accordance with Table. D 3.5.1-1.B. I Add the new location(s) to the Radiological Environmental Monitoring Program.AND B.2 Delete the sampling location(s), excluding the control stationlocation, having the lowest calculated dose, 'dose commitment(s) or D/Q value, via the same exposure.
pathway,, fiom the Radiological Environmerntal Monitoring Program.AND B.3 Submit in the next Radioactive Effluent Release Report documentation for 'a change in the ODCM including revised figure(s)and table(s) for the ODCM reflecting the newlocation(s) with information supporting the change in sampling locations.
30 days After October 31 of the year in which the land use census was conducted In accordance with the Radioacti've Effluent Release Report___________________________________
4 4-Unit.2 Revision 28 September 2006 1.3.5-14 Land Use: Census D 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE.
FREQUENCY----- i DSR 3.512.1 Conduct the land use census during the growing season using that information that will provide the best results,:
such as by a door-to-door survey, aerial survey, or byconsulting local agriculture authorities.
366 days DSR 3.5.2.2 Report the results of the land use census in the Annual Radiological Environmental Operating Report.In accordance with the Annual Radiological Environmental Operating Report Unit .2 Revision.28 Scptcmbcr.2006, I 3.5-15 Interlaboratory Comparison Program D 3.5.3 D 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING
.3.35.3 Interlaboratory Comparison Program DLCO 3.5.3 The Interlaboratory Comparison Program shall be described, in the ODCM., AND Analyses shall be performed on all radioactive-materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the NRC, that correspond to samples requited by TableO D 3.5'. 1- 1.Participation in this program shall include media for which environmental samples are-routinely collected and for which intercomparison samples are available.
APPLICABILITY:
At all times.ACTIONS--------------------, ....--- ---............----............------
N O T E S-........---...
....... ..1. LCO 3.0.3 is not applicable.
- 2. LCO 3.0.4zis not applicable.
CONDITION REQUIRED ACTION COMPILETION
_TIME A. Analyses not performed as A.1 Report the corrective actions Inaccordance with required.
taken to prevent a recurrence the Annual to the NRC in the Annual Radiological Radiological Environmental Environmental Operating Report' Operating Report Unit 2 Revision 28 September 2006 1 3.5-16 Interlaboratory Comparison Program D 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.3.1 Report a summary of the results obtained as part of the In accordance with Interlaboratory Comparison Program in the Annual the Annual Radiological Environmental Operating Report. Radiological Environmental Operating Report Unit 2 Revision, 28 Scptcmber 2006 I _.5-17 PART I -RADIOLOGICAL EFFLUENT CONTROLS BASES Unit 2 Revision 28 1 B 3.11-0 September 2006 Liquid Effluents Concentration B.3.11 B 3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1 1,1 Liquid Effluents Concentration BASES.This. 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 10 CFR 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 resultin exposures Within: (1) the Section II.A design objectives of Appendix I to 10 CFR 50, to a MEMBER OF THE PUBLIC and (2) the levels required by 10 CFR 20.1301(e) to the population.
The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xc- 13.5 is the controlling radioisotope and'its effluent concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in Intemational Commission on.Radiological Protection (ICRP) Publication,2.
This applies to the release of radioactive materials in liquid effluents.
from all units at. the site.The required detection capabilities for radioactive materials iii liquidwaste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of~theLLD, and other detection limits can befoundin L. .A. Currie, "Lower Limit of Detection:
Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREGICR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).
Unit 2 Revision 28 I B 31-1 September 2006 Liquid Effluents Dose B 3.1.2 B 3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.A.2 Liquid Effluents Dose, BASES This is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I to 10 CFR 50. This implements the guides set forth in Section II.A of AppendixI.l 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 liquid. effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable.
Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance.that the operation of the facility will not result in radionuclidc concentrations in the potable drinking waterthat are. in excess of the requirements of 40 CFR 141. For ýsites containing up to four reactors, it ishighly unlikely that the resultant.
dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside storage tanks, etc., are kept small. The Special Report will describe a course-of action that should result. in the limitation of the Annual dose to a MEMBER OF THE. PUBLIC to within the.40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitmentto the MEMBERS OF THE PUBLIC from otheruraniumfuel cycle sources .is negligible, with the exception that dose contributions from other nuclear, fuel. cycle facilities at the same site or within a radius of 5 miles must be considered, The dose calculation methodology and parameters implement the requirements in Section, IIl.A of Appendix I that conformance with. the guides of Appendix I be shown by Calculational procedures based on models and data, so that the: actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.
The equations specified for calculating the doses that.result from actual release rates of radioactive material in liquid effluents are consistent with the methodology provided in RO 1.109, "Calculation of Anmual Doses To Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part.5 0, Appendix I," Revision 1, October 1977 and R.G. 1.113, "Estimating Aquatic Dispersion of EffliUnts from Accidental and Routine Reactor Releases.for the Puipose of Implementing Appendix I," April 1977. This applies tothe release of radioactive materials in liquid effluents from each unit at the site. For units with shared radwaste treatment systems. the liquid effluents from the shared.system are to be proportioned among the units sharing that system.Unit 2 Revision 28 1 B 3.1-2 September 2006 Liquid Radwaste Treatment System B 3.1.3 B3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.3 Liquid Radwaste Treatment System BASES The installed liquid radwaste treatment system shall be :considered FUNCTIONAL by meeting DLCO 3,. 1.1 and DLCO 3.1.2. The FUNCTIONALITY of the liquid radwaste treatment system ensures that this system will be available for use-whenever liquid effluents require treatment before release to the i environment.
The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept as low asis reasonably achievable.
This implements.the-requirements of 10 CFR 50.36a, GDC 60.of Appendix A to lOCFR 50 and the design objective given in Section II.D of Appendix.
Ito 10 CFR 50. The~specified limits governing, the use of appropriate portions of the. liquid. radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I to 10 CFR 50 for liquid effluents.
This applies to the release of radioactive materials in liquid effluents from each unit at. the site. For units with shared radwaste'treatment systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.Unit 2 Revision 28 September 2006 I B 3.1-3 Gaseous Effluents Dose Rate B 3.2.1 B 3.2 RADIOACTIVE GASEOUS. EFFLUENTS B 312.1 Gaseous Effluents Dose Rate BASES: This is provided to ensure that the: dose rate- at any time at and beyond the SITE BOUN DARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR 20 to UNRESTRICTED AREAS.The annual dose limits. are the doses associated with the. concentration.
of 10 CFR 20, Appendix B, Table2, Column I. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of aMEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR 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.
Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in Part II. The specified release rate limits restrict, atall 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 mremryear to the whole body or to. less than or equal to 3000 mrem/yearto the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a childvia the inhalation pathway to~less than or equal to 1500 mrem/year.
This applies to the release of radioactive materials in gaseous effluents from all units at the site.The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of thelower limits of detection (LLDs). Detailed discussion oftheLLD, and other detection limits can be found in..L. A. Currie, "Lower LimitofDetection:
Definition and Elaboration of a Proposed Position for Radiological Effluent and Environments Measurements," NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).
Unit .2 Revision 28 I B3.2-1 September 2006 Gaseous Effluents Noble. Gas Dose B 3.2.2 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.2 Gaseous Effluents Noble Gas Dose BASES:*This is provided to implement the requirements of Section 1l.B, 1Il.A, and IV.A of Appendix I to 10 CFR 50. The DLCO implements the guides set forth in Section II.B of Appendixi.
The REQUIRED ACTIONS 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 Requirements implement the requirements in Section IIILA of Appendix I that conformance with the guidelines of Appendix I be shown by calculational procedures-based on models and data so that the actual cxposure of a MEMBER. OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.
For sitescontaining.
up to four reactors, it is highly unlikely that the resultant.
dose tooa MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside:storage tanks, etc., are kept small. The Special Report will describe a course of actionthat should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within ihe 40 ýCFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC fromother uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the -same site or within a radius of 5 miles must be considered.
The dose calculation methodology
- and parameters for:calculating the doses from the actual release rates' of radioactive noble in gaseous: effluents are consistent with the methodology provided in RG 1.109,"Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and RG 1. i 11, "Methods fr'Estimating Atmospheric Transport.and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1,"' July 1977. The ODCM equations'provided for determining the air doses at or beyond the.SITE BOUNDARY are based upon real-time meteorological conditions or the historical average atmospheric conditions.
This applies to the release of radioactive.material in gaseous effluents from each unit at the site.Unit 2 Revision 28 I B 3.2-2 September 2006 Gaseous Effluents Dose -Iodine-i 31, Iodine-i133, Tritium, and Radioactive Material In Particulate Form B 3.2.3 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.3 Gaseous Effluents:Dose
-todine-131, Iodine-133, Tritium., and Radioactive Material In Particulate Form BASES This is provided to implement the requirements of Sections.
II.C, III.A, and IV.A of Appendix I to 10 CFR 50. The DLCO implements the guides set forth in Section II.C of Appendix I. The REQUIRED ACTIONS provide the required operating flexibility and at the same time implementthe guides set -forth inSection IV.A of Appendix Ito assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable.
The calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on: models and data, so that the actual exposureof a MEMBER OF THE PUBLIC thtough.appropriate pathways is unlikely to be substantially underestinmated.
.For sites containing up to four reactors,.
it is highly unlikely:that the resultant dose to a MEMBER OF THE PUBLIC. will exceed the dose limits of 40 CFR 190 if the individual reactors.remain within twice the dose design objectives of Appendix I, and if direct radiation0 doses from theunits including outside storage tanks, etc., are keptsmall.
The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report,.it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC.fiom other: uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle, facilities at the same site or within a radius of 5 miles must be considered.
The calculational methodology and parameters for calculating the doses from the actual release rates of the subject materials.
are consistent with'the methodology provided in RG 1.109, "Calculationof Annual Doses to Man from Routine Releases of Reactor Effluents for. the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, " Revision 1, October 1977, :and RG 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 DLCO. for iodine- 131, iodine- 133, tritium, and radioactive material in particulate form with half-lives greater than 8 days.,are dependent upon the e.isting radionuclide pathways to man, in the areas at or beyond the SITE BOUNDARY.
The pathways that. were. examined inthe development of thesecalculations were: (1) individual inhalation of airborne radioactive material, (2) deposition of radioactive material onto green leafy vegetation Unit 2 Revision 28 I B 3.2-3 September 2006 Gaseous Effluents Dose -Iodine-13.1, Iodine-133, Tritium, and Radioactive Material In Particulate Form B'3.2.3 B 3.23 Gaseous Effluents Dose -.lodine-131,:Iodine-133, Tritium, and Radioactive Material In Particulate Form (continued) with subsequent consumption by man, (3) deposition onto grassy areas where milk-producing animals and meat-producing animals graze (human consumption of the milk and meat is assumed), and (4) deposition on the ground with subsequent exposure to man. This applies to therelease of radioactive materials in gaseous effluents from each unit at the site. For units with shared :radwaste treatment.
systems, the. gaseous effluents:
from the shared system are proportioned among the units sharing that system.Unit 2 Revision 28 September 2006 I B 3.2-4 Gaseous Radwaste-Treatment System B 3.24 B 3.2 RADIOACTIVE GASEOUS FFI7IENTS B 3.2.4 Gaseous Radwaste Treatment System BASES The FUNCTIONALITY of the GASEOUS RADWASTE TREATMENT SYSTEM ensures that..the system will be available for use whenever gaseous effluents require treatment before release to the environment.
The requirement that the appropriate portions of this system be used, when specified; provides reasonable assurance that the releases of radioactive materials in gaseous effluents will bckept as low as is reasonably achievable.
This implements the requirements of 10 CFR 50.3.6a, GDC 60 of Appendix A to 10 CFR 50, and the design objectives given in.,Section II.D of Appendix I to 10 CFR 50. Limits governing the use of appropriate portions of the system were specified as a. suitable fraction of the dose design objectives set forth in:Sections ILB and IIC of Appendix Ito 10 CFR.50, for gasous effltents, This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the sharedsystem are proportional among the units sharing that system.Unit .2 Revision 28 September 2006 IB 3.2-5 Ventilation Exhaust Treatment System B 3.2.5 B 32 RADIOACTIVE GASEOUtS EFFLUENTS B 3.2.5 Ventilation Exhaust Treatmeht System BASES The FUNCTIONALITY of the VENTILATION EXHAUSr IREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment before release to the environment.
The requirement that-the appropriate portions of this system be used, when specified, provides reasonable; assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable.
This implements the requirements of 10 CFR 50,36a, GDC!60 of Appendix A to 10 CFR 50, and the. design objectives given in Section ILD of Appendix Ito 10CFR50. Limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives.
set forth in Sections ILB and II.C of Appendix Ito 10 CFR 50, for gaseous effluents.
This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment.systems, the gaseous effluents from theshared system arc proportional among the. units sharing that- system.The appropriate components, which affect iodine or particulate release, to be-FUNCTIONAL are: 1) HEPA Filter -Radwaste Decon Area 2) HE.PA Filter -Radwaste Equipment Area 3) HEPA Filter -Radwaste General Area Whenever one of these filters is not FUNCTIONAL, iodine and particulate dose projections will be made for 31-day intervals starting with filter nonfinctionality, andcontinuing as long asthe filter remains nonfunctional, in accordance with DSR 3.2.5.1.Unit 2 Revision 28 I B 3.2-6 September 2006 Venting or Purging B 3.2.6 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.6 Venting orPurging BASES This: provides reasonable assurance that releases from drywell and/or suppression chamber purging. operations will not exceed the arnual dose limits of 10 CFR 20 for unrestricted areas.Unit 2 Revision 28 September 2006 I B 3.2-7 Radioactive Liquid Effluent Monitoring Instrumentation B 3.3.1 B 33 B 3.3.1 INSTRUMENTATION Radioactive Liquid Effluent*Monitoring Instrumentation BASES The radioactive liquid effluent instrumentation is provided to monitor and control,:as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents.
The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Part .11. to ensure that the alarm/trip will occur bcforc cxceeding tn times the. limits of 10 CFR 20. The FUNCTIONALITY and use: of this instrumentation is consistent with the requirements of GDC 60, 63, and 64 of Appendix Ato 10 CFR 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially riesult in the transport of radioactivc matcrials to UNRESTRICTED AREAS.Tanks included are those outdoor tanks that. are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liqui&radwaste treatment system, such as temporary tanks.Unit 2 Revision 28 September 2006 I B 3.3-1 Radioactive Gaseous Effluent Monitoring Instrumentation.B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Radioactive Gaseous Effluent Monitoringlnstrumentation BASES The radioactive gaseous effluent instrumentation isprovided to monitor and.control, as, applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.
The alarm/trip setpointsfor these instruments shall be calculated and. adjusted in accordance with the methodology and parameters in; PartlII to ensure that the alarm/trip will occur before cxcceding the limits of 10 CFR 20. Although the. Offgas System Noble Gas Activity Monitor is listed in Table D. 3.3.2-1, "Radioactive Gaseous Effluent Monitoring Instriumentation", these monitors are actually located upstream of the Main Stack.noble gas activity moniitor and are. not effluent monitors, They were included in Table D 3.3.2-1 in accordance with'NUREG-0473.
As such, Offgas System Noble Gas Activity.
Monitoralarm and trip setpoints:
are not based on I OCFR20. The offgas system noble gasmonitor alert setpoint is set at 1.5 times nominal full power background to assure compliance .with ITS SR 3.7.4.1 which requires offgas sampling be performed within four hours of a 50% increase in offgas monitoring readings, and to support MSLRM trip removal. The offgas system noble gas monitor trip setpoint is based on the IOCFR100 limits for the-limiting design basis gaseous waste system accident whichjis the offgas system rupture. The range of the noble gas channels of the main stack and radwaste/reactor building vent effluent monitors is sufficiently large to envelope both normal and accident levelsof noblelgas activity.
The capabilities of these instruments are consistentwith therecommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following.
an Accident," December 1980 and NUREG-0737, "Clarification oftheTMI Action Plan Requirements," November 1980. This instrumentation also includes provisions*
for monitoring and controlling the concentrations of potentially explosive gas mixtures in the offgas system. The.FUNCTIONALITY and, use ofthis instrumentation is consistent ,with the requirements of GDC Q 60, 63, and. 64 of Appendix Ato 10 CFR 50..Unit 2 Revision 28 1133.3-2 September 2006 Radioactive Effluents Total .Dose B 3.4 B 3.4 RADIOACTIVE EFFLUENTS TOTAL DOSE BASES This is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46 FR 18525. This requires the preparation and, submittal of a Special Report whenever the calculated doses from releases of radioactivity and from radiation from uranium fuel cycle sources exceed 25 mremrnto the whole body or anyorgan, except the thyroid (which shall be limited to less than or equal to 75 mrem). If the dose-to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violationof 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to .be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed.
The variance only relates to the limits of 40. CFR 190, and does not apply in any way to:the other requirements for dose limitation of 10 CFR 20, as addressed in.3.1.1 and 3.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which the individual is engaged in carrying out any operation that is part of the nuclear fuel cycle.Unit 2 Revision 28 September 2006.I B 34-1 Monitoring Program B 3.5.1 B 3.5 RADIOLOGICAL ENVIRONMENTAL'MONITORING B 3.5.1 Monitoring:
Program BASES The Radiological Environmental Monitoring Program provides representative measurements of radiation and of radioactive materials.
in-those.
exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the plant operation.
This monitoring program implementsSection IV.B.2 of Appendixj Ito 10 CFR 50 and thereby supplemeritsthe Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the. basisof the effluent measurements and the modeling of the environmental exposure pathways.
Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. Program changes may be initiated based on operational experience.
The required detection capabilities for environmental sample analyses are tabulated in terms of thelower limits of detection (LLDs). The LLDs required.by Table D 3,5.1-3 are considered optimum for routine environmental measurements in industrial laboratories.
It should be recognized that theI LLD is~defined as a before-the-fact limit representing the capability of a measurement system and not as anafter-the-fact limit for a particular measurement.
Detailed discussion of the LLD, and other detection limits, can be found in L.A. Currie, "Lower Limit of Detection:
Definition and Elaboration of a-Proposed Position for Radiological Effluent and Environmental Measurements;" NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASI.-300 (revised annually).
Unit 2 Revision 2&I B 3.5-1 September 2006 Land Use Census B 3.5.2 B 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 3.5.2 Land Use Census BASES This is provided to ensure that changes in.the use of areas.at or beyond the S ITE BO UNDARY are identified and that modifications to the Radiological Environmental Monitoring Program are made if required by the results of this census. The best intbrmation,, such as from a door-to-door survey, from an aerial survey, or'from consulting With local agricultural authorities, shall be used.This census satisfies therequirements of Section IV.B.3 of Appendix I to 10 CFR 50.Restricting the census tojgardens of greater than 500 square feet. provides assurance that significant exposure pathways via leafy vcgetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26.kg/y ear) of leafy vegetables assumed in RG 1 .109 for consumption by a child. To determine this minimum garden size,. the following assumptions were made: (!) 20% of thc garden was used for growing broad leaf vegetation (i.e.; similar to lettuce and cabbage) and (2) the vegetation yield was 2 kg/m 2.A MILK SAMPLING LOCATION, as defined in Section 1.0, requires that at least 10 milking cows.are present at a designated milk samplelocation..
It has beenfound from past experience, and, as a result ofconferring with local farmers, that a minimum~of 10 milking cows is necessary to guarantee an adequate supply of milk twice- a month for analytical purposes.
Locations with fewer than 1.0 milking cows are usually utilized for breeding.
purposes, eliminating azStable.supply of milk for samples as a result of suckling calves and periods when the adult animals are dry. Elevated releases are defined in RG 1. 111, Revision 1, July 1977.Unit 2 Revision 28 I B 3.5-2 September 2006 Interlaboratory Comparison Program B 3.5.3 B 3;5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 3.5.-3 Interlaboratory Comparison Program BASES The requirement for paiticipation in an approved Interlaboratory Comparison Program is provided to ensure that: independent checks on the precision and accuracy-of the measurements of radioactive, materials in environmental sample matrices are performed as part of the quality assurance program for environmentalmonitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR 50.Unit 2 Revision 28 September 2006 I B 3.5-3 PART I -RADIOLOGICAL EFFLUENT CONTROLS SECTION 4.0 ADMINISTRATIVE CONTROLS Unit 2 Revision 28 September 2006 1 4.0-0 Administrative Controls 4.0 4.0 ADMINISTRATIVE CONTROLS The ODCM Specifications are subject to Techikical Specifications Section 5.5.4, "Radioactive Effluent Controls Program," Section 5.6.2, "Annual Radiological Environmental Operating Report" Section 5.6.3, "Radioactive Effluent-Release Report," and ýSejction 5.5.1, "Offsite DOse Calculation Manual." Unit 2 Revision 28 September 2006 1 4.0-1 Special Reports D4.1.1 D4.1.2 D 4.1.3 D 4.1 REPORTING REQUIREMENTS D 41. .lSpecial Reports Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.D 4.1 .2Annual Radiological Environmental Operating Reports In addition to theý requirements of Technical Specification 5.6.2 the report shall also includethe following:
A summary description of the Radiological Environmental Monitoring Program; at least two legible maps, one shall cover stations near the SITE BOUNDARY and the. second shall include the more distant stations, coveting all sample locations keyed to a tablegiving distances and directions from thecenterline of one. reactor; the results of license participation in the Interlaboratory Comparison Program, required by Control D 3.5.3; discussion of all deviations from the Sampling Schedule of Table P 3.5.1-1; and.discussion of all analysis in which the LLD required by Table D 35.1-3 was not achievabic.
D 4.1 .3Radioacti ve Effluent Release Report The Radiological Effluent Release Report described in Technical Specification section 5.6.3 shall include: " An annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hourtlisting, on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation(if measured), or in the form of joint, frequency distribution of wind speed, wind direction, and atmospheric stability.
In lieu of submission with the RadiologicalEffluent Release Report, the licensee has tlhe.option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request." An assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during the previous year.(Continued)
Unit 2 Revision.
28 1. 421-1 September 2006 Special Reports D 4.1.3 D 4.1.3 Radioactive Effluent'Release Report (continued)
As assessment of radiation doses from the radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC from their activities inside the SITE BOUNDARY during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location shall be included in these reports. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in Part II.* As assessment of doses to the likelymost exposed MEMBER OF THE PUBLIC from reactor releasesand other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiatin, for the previous calendar year to show conformance with 40 CFR .190, "Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given inPart II.* A list of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquideffluents made during the reporting period..Any changes made during the reporting period to the PROCESS CONTROL PROGRAM and to the OFFSITE DOSE CALCULATION MANUAL (ODCM).* Any major changes to liquid, gaseous, or solid radwaste treatment systems pursuant to D 4.2.e A listing of new locations for dose calculations andlor environmental monitoring identified by the land use census pursuant to. Control D 3.5.2.* An explanation of why the nonfunctionality of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified.in Controls D33.1 and D 3.3.2.* Description of events leading to liquid bholdup tanks exceeding the limits, of TRM 3,7.7.Unit 2 Revision 28 1 4. i-1 a September 2006 Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment System D 4.2 D 4.2 MAJOR CHANGES TOLIQI JiD, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEM--- --------------------------
.-.------NOTE ---------------------
.---------
Licensees may choose to subnmit this information as part of the annual FSAR update, Licensee-initiated majorchanges to the radwaste treatment systems (liquid, gaseous, and solid): a. Shall be reported to the Commission in the Radioactive Effluent Releasereport for the period in which the evaluation was reviewed by the SORCG The discussion of each changes shall contain: 1. A summary of the-evaluation thatled to the determination that the change::could be ,made in accordance with 10 CFR:50.59.
- 2. Sufficient detailed information to totally support the reason for thechangc without benefit of additional or supplemental informiation;
- 3. A detailed description of the equipment, components, andprocesses involved alnd the interfaces'with other plant systems;4. An evaluation of the change, which shows the predicted
- releases of radioactive materials:
in liquid :and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;5'. An ev'aluation of the change, which shows the expected maximum exposures -to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimatedin the license application and amendments thereto;:6. A comparison of the predicted releases of radioactive materials, in liquid and: gaseous, effluents and in. solid waste, to the actual releases for theperiod that precedes the time when the: change is to be made;7. An estimate of the to plant operating personnel as a resultof the change; and (Continued)
Unit 2 Revision 28 1 4.2-1 September 2006 Major Changes to Liquid, Gaseous, and Solid, Radwaste Treatment System D 4.2 D 4.2 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEM (continued) 8'. Documentation of the fact that the change was reviewed and found acceptable by the SORC.b. Shall become effective upon review and acceptance by the SORC.Unit 2 Revision 28 September 2006 i 4.2ý2 PART .I- CALCULATLONAL METHODOLOGIES Unit 2 Revision 28 September
- 2006 I.1, 1.0 LIQUID EFFLUENTS Service Water A and B, Cooling Tower Blowdown and the Liquid Radioactive Waste Discharges comprise the Radioactive Liquid Effluents at Unit 2. Presently there are no temporary outdoor tanks containing radioactive water capable of affecting the nearest known or future water supply in an unrestricted area. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.1.1 Liquid Effluent Monitor Alarm Setpoints 1.1.1 Basis The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) 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 nobles gases, the concentration shall be limited to 2E-04 uCi/ml total activity.1.1.2 Setpoint Determination Methodology 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint The Liquid Radioactive Waste System Tanks are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. At the end of the discharge tunnel in Lake Ontario, a diffuser structure has been installed.
Its purpose is to maintain surface water temperatures low enough to meet thermal pollution limits. However, it also assists in the near field dilution of any activity released.
Service Water and the Cooling Tower Blowdown are also pumped to the discharge tunnel and will provide dilution.
If the Service Water or the Cooling Tower Blowdown is found to be contaminated, then its activity will be accounted for when calculating the permissible radwaste effluent flow for a Liquid Radwaste discharge.
The Liquid Radwaste System Monitor provides alarm and automatic termination of release if radiation levels above its alarm setpoint are detected.The radiation detector is a sodium iodide crystal. It is a scintillation device. The crystal is sensitive to gamma and beta radiation.
However, because of the metal walls of the sample chamber and the absorption characteristics of water, the monitor is not particularly sensitive to beta radiation.
Actual detector response Ei (CGi/CFi), cpm, has been evaluated by placing a sample of typical radioactive waste into the monitor and recording the gross count rate, cpm. A calibration ratio was developed by dividing the noted detector response, Yj (cGc/CFi) cpm, by total concentration of activity Yj (CGi)uci/cc. The quantification of the gamma activity was completed with gamma spectrometry equipment whose calibration is traceable to NIST. This calibration ratio verified the manufacturer's prototype calibration, and any subsequent transfer calibrations performed.
The current calibration factor (expressed as the reciprocal conversion factor, uCi/ml/cpm), will be used for subsequent setpoint calculations in the determination of detector response: Ei(CGi/CFi)
= Yi(CGi)/CF Unit 2 Revision 28 112 September 2006 Where the factors are as defined above.The calculations of the required dilution factors (RDF) are performed as follows: RDF7 = MEC gamma fraction = (CG, /MEC,)RDFTOTAL = -MEC total fraction = Z(C 1/MEC 1)i i RDFY is used to calculate the liquid radwaste effluent radiation monitor setpoint.
This monitor is a gamma detector and has little or no response to non gamma emitters.
Use of RDF, rather than RDFTOTAL, to determine the monitor setpoint prevents the condition where a tank with gamma concentrations near their LLD cannot be discharged due to spurious alarms received because the setpoint is close to the monitor background.
RDFTOTAL is used to determine the minimum dilution factor required to discharge the tank contents based on all activity, both gamma and non gamma, in the tank. This ensures that the concentrations of all radioactive materials released in liquid effluents will meet DLCO 3.1.1. Non gamma emitting nuclide activity, except tritium was initially estimated based on the expected ratios to quantified nuclides as listed in the FSAR Table 11.2.5. Fe-55, Sr-89 and Sr-90 are 2.5, 0.25 and 0.02 times, respectively, the concentration of Co-60. Currently, non gamma activity except tritium is estimated using the results from the latest analysis of composite samples.Tritium concentration is assumed to equal the latest concentration detected in the monthly tritium analysis of liquid radioactive waste tanks discharged.
Nominal flow rates of the Liquid Radioactive Waste System Tanks discharged is < 165 gpm while dilution flow from the Service Water Pumps, and Cooling Tower Blowdown cumulatively is typically over 10,200 gpm. Because of the large amount of dilution the alarm setpoint could be substantially greater than that which would correspond to the concentration actually in the tank. Potentially a discharge could continue even if the distribution of nuclides in the tank were substantially different from the grab sample obtained prior to discharge which was used to establish the detector alarm point. To avoid this possibility of "Non representative Sampling" resulting in erroneous assumptions about the discharge of a tank, the tank is recirculated for a minimum of 2.5 tank volumes prior to sampling.This monitor's setpoint takes into account the dilution of Radwaste Effluents provided by the Service Water and Cooling Tower Blowdown flows. Detector response for the nuclides to be discharged (cpm) is multiplied by the Actual Dilution Factor (dilution flow/waste stream flow) and divided by the Required Dilution Factor (total fraction of the effluent concentration in the waste stream). A safety factor is used to ensure that the limit is never exceeded.
Service Water and Cooling Tower Blowdown are normally non-radioactive.
If they are found to be contaminated prior to a Liquid Radwaste discharge then an alternative equation is used to take into account the contamination.
If they become contaminated during a Radwaste discharge, then the discharge will be immediately terminated and the situation fully assessed.Unit 2 Revision 28 II 3 September 2006 Normal Radwaste Effluent Alarm Setpoint Calculation:
Alarm Setpoint < 0.8
- TDF/PEF
- TGC/CF
- 1/RDF7 + Background.
Where: Alarm Setpoint = The Radiation Detector Alarm Setpoint, cpm 0.8 -Safety Factor, unitless TDF = Nonradioactive dilution flow rate, gpm. Service Water Flow (ranges from 30,000 to 58,000 gpm) +Blowdown flow (typically 10,200 gpm) -Tempering Ci = Concentration of isotope i in Radwaste tank prior to dilution, jtCi/ml (gamma + non-gamma emitters)CFi = Detector response for isotope i, net ýtCi/ml/cpm See Table D 2-1 for a list of nominal values PEF -The permissible Radwaste Effluent Flow rate, gpm, 165 gpm is the maximum value used in this equation MECi -Maximum Effluent Concentration, ten times the limiting effluent concentration for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, ýtCi/ml Background
-Detector response when sample chamber is filled with nonradioactive water, cpm CF -Monitor Conversion Factor, VtCi/ml/cpm, determined at each calibration of the effluent monitor CGi = Concentration of gamma emitting nuclide in Radwaste tank prior to dilution, ýtCi/ml TGC = YCGi = Summation of all gamma emitting nuclides (which monitor will respond to)(CGj/CFj)
= The total detector response when exposed to the concentration of nuclides in the Radwaste tank, cpm RDFy = Ei (CGi/MECi)
-The total fraction often times the 10 CFR 20, Appendix B, Table 2, Column 2 limit that is in the Radwaste tank, unitless.
This is also known as the Required Dilution Factor Gamma (RDFy).TGC/CF = An approximation to E (CGj/CFi) using CF determined at each calibration of the effluent monitor TDF/PEF = An approximation to (TDF + PEF)/PEF, the Actual Dilution Factor in effect during a discharge.(Cj/MECj)
= The total fraction of ten times the 10 CFR 20, Appendix B, Table 2, Column 2 limit that is in the Radwaste tank, unitless.This is also known as the Required Dilution Factor-Total and includes both the gamma and non-gamma emitters.Tempering A diversion of some fraction of discharge flow to the intake canal for the purpose of temperature control, gpm.Unit 2 Revision 28 114 September 2006 Permissible effluent flow, PEF, shall be calculated to determine that the maximum effluent concentration will not be exceeded in the discharge canal.PEF = TDF (RDFTotaI) 1.5 If Actual Dilution Factor is set equal to the Required Dilution Factor, then the alarm points required by the above equations correspond to a concentration of 80% of the Radwaste Tank concentration.
No discharge could occur, since the monitor would be in alarm as soon as the discharge commenced.
To avoid this situation, maximum allowable radwaste discharge flow is calculated using a multiple (usually 1.5 to 2) of the Required Dilution Factor, resulting in discharge canal concentration of 2/3 to 1/2 of the maximum effluent concentration prior to alarm and termination of release. If no gamma emitters are detected in the Radwaste Tank samples, then the radiation monitor setpoint will be based on assuming gamma activity at the LLD of the most limiting nuclide from recent discharges.
In performing the alarm calculation, the smaller of 165 gpm (the maximum possible flow) and PEF will be used.To ensure the alarm setpoint is not exceeded, an alert alarm is provided.
The alert alarm will be set in accordance with the equation above using a safety factor of 0.5 (or lower)instead of 0.8.1.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculation:
The allowable discharge flow rate for a Radwaste tank, when one of the normal dilution streams (Service Water A, Service Water B, or Cooling Tower Blowdown) is contaminated, will be calculated by an iterative process. Using Radwaste tank concentrations with a total liquid effluent flow rate, the resulting fraction of the maximum effluent concentration in the discharge canal will be calculated.
FMEC = Ys[Fs/ys(F.)
i(Cis MECi)]Then the permissible radwaste effluent flow rate is given by: PEF = Total Radwaste Effluent Flow FMEC The corresponding Alarm Setpoint will then be calculated using the following equation, with PEF limited as above.TGC/CF Alarm Setpoint < 0.8 + Background FMEC Unit 2 Revision 28 II 5 September 2006 Where: Alarm Setpoint 0.8 F, Ci Cis CF MECQ PEF Background TGC/CF =Z+/- (CGi/CF)Y- CFCi.]Y,; [F.)= The Radiation Detector Alarm Setpoint, cpm= Safety Factor, Unitless= An Effluent flow rate for stream s, gpm= Concentration of isotope i in Radwaste tank prior to dilution, ýtCi/ml Concentration of isotope i in Effluent stream s including the Radwaste Effluent tank undiluted, ýtCi/ml= Average detector response for all isotopes in the waste stream, net jiCi/ml/cpm
= Maximum Effluent Concentration, ten times the effluent concentration limit for isotope i from 1 OCFR20 Appendix B, Table 2, Column 2, ýtCi/ml= The permissible Radwaste Effluent Flow rate, gpm= Detector response when sample chamber is filled with nonradioactive water, cpm The total detector response when exposed to the concentration of nuclides in the Radwaste tank, cpm The total activity of nuclide i in all Effluent streams, ýtCi-gpm/ml The total Liquid Effluent Flow rate, gpm (Service Water & CT Blowdown & Radwaste)1.1.2.3 Service Water and Cooling Tower Blowdown Effluent Alarm Setpoint These monitor setpoints do not take any credit for dilution of each respective effluent stream. Detector response for the distribution of nuclides potentially discharged is divided by the total MEC fraction of the radionuclides potentially in the respective stream. A safety factor is used to ensure that the limit is never exceeded.Service Water and Cooling Tower Blowdown are normally non-radioactive.
If they are found to be contaminated by statistically significant increase in detector response then grab samples will be obtained and analysis meeting the LLD requirements of Table D 3.1.1 -1 completed so that an estimate of offsite dose can be made and the situation fully assessed.Service Water A and B and the Cooling Tower Blowdown are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. Normal flow rates for each Service Water Pump is 10,000 gpm while that for the Cooling Tower Blowdown may be as much as 10,200 gpm. Credit is not taken for any dilution of these individual effluent streams.The radiation detector is a sodium iodide crystal. It is a scintillation device. The crystal is sensitive to gamma and beta radiation.
However, because of the metal walls in its sample chamber and the absorption characteristics of water, the monitor is not particularly sensitive to beta radiation.
Unit 2 Revision 28 II 6 September 2006 Detector response i (ci/cFi) has been evaluated by placing a diluted sample of Reactor Coolant (after a two hour decay) in a representative monitor and noting its gross count rate. Reactor Coolant was chosen because it represents the most likely contaminant of Station Waters.A two hour decay was chosen by judgement of the staff of Nine Mile Point. Reactor Coolant with no decay contains a considerable amount of very energetic nuclides which would bias the detector response term high. However assuming a longer than 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> decay is not realistic as the most likely release mechanism is a leak through the Residual Heat Removal Heat Exchangers which would contain Reactor Coolant during shutdowns.
Service Water and Cooling Tower Blowdown Alarm Setpoint Equation: Alarm Setpoint< 0.8 1/CF Yi C+/-/ [Yi(Ci/MECJ)]
+/- Background.
Where: Alarm Setpoint 0.8 Ci CFi MECi Background Yi (Ci/CFi)Yi (Ci/MECi)(1/CF) YiCi CF-The Radiation Detector Alarm Setpoint, cpm= Safety Factor, unitless= Concentration of isotope i in potential contaminated stream, jiCi/ml= Detector response for isotope i, net pCi/ml/cpm See Table 2-1 for a list of nominal values= Maximum Effluent Concentration, ten times the effluent concentration limit for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, pCi/ml= Detector response when sample chamber is filled with nonradioactive water, cpm= The total detector response when exposed to the concentration of nuclides in the potential contaminant, cpm The total fraction of ten times the 1OCFR20, Appendix B, Table 2, Column 2 limit that is in the potential contaminated stream, unitless.= An approximation to Yj (Ci/CFi), determined at each calibration of the effluent monitor= Monitor Conversion Factor, p[Ci/ml/cpm 1.2 Liquid Effluent Concentration Calculation This calculation documents compliance with Section D 3.1.1 of Part 1: Unit 2 Revision 28 II 7 September 2006 As required by Technical Specification 5.5.4, "Radioactive Effluent Controls Program," the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) 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 microcurie/ml total activity.The concentration of radioactivity from Liquid Radwaste, Service Water A and B and the Cooling Tower Blowdown are included in the calculation.
The calculation is performed for a specific period of time. No credit is taken for averaging.
The limiting concentration is calculated as follows: FMEC S Ys [Fs/y, (F.) Yi (Cis+MECi)
I Where: FMEC Cis F, MECi Yj (Cis/MECD)
Ys (Fs)= The Fraction of Maximum Effluent Concentration, the ratio at the point of discharge of the actual concentration to ten times the limiting concentration of 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases, unitless= The concentration of nuclide i in a particular effluent stream s, itCi/ml= The flow rate of a particular effluent stream s, gpm= Maximum Effluent Concentration, ten times the limiting Effluent Concentration of a specific nuclide i from 1 OCFR20, Appendix B, Table 2, Column 2 (for noble gases, the concentration shall be limited to 2E-4 microcurie/ml), ýtCi/ml= The Maximum Effluent Concentration fraction of stream s prior to dilution by other streams= The total flow rate of all effluent streams s, gpm 1.3 A value of less than one for the MEC fraction is required for compliance.
Liquid Effluent Dose Calculation Methodology The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure D 1.0-1) shall be limited: a. During any calendar quarter to less than or equal to 1.5 mrem to the whole body and to less than or equal to 5 mrem to any organ, and b. During any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ.Unit 2 Revision 28 118 September 2006 Doses due to Liquid Effluents are calculated monthly for the fish and drinking water ingestion pathways and the external sediment exposure pathways from all detected nuclides in liquid effluents released to the unrestricted areas using the following expression from NUREG 0133, Section 4.3.D, = Yi [Ait ZL (ATLCiLFL)
]Where: Dt The cumulative dose commitment to the total body or any organ, t from the liquid effluents for the total time period Z, (ATT) , mrem ATL = The length of the L th time period over which CiL and FL are averaged for all liquid releases, hours CiL = The average concentration of radionuclide, i, in undiluted liquid effluents during time period ATL from any liquid release, ýtCi/ml Ait= The site related ingestion dose commitment factor for the maximum individual to the total body or any or gan t for each identified principal gamma or beta emitter, mrem/hr per pCi/ml. Table D 2-2.FL The near field average dilution factor for Cij during any liquid effluent release. Defined as the ratio of the maximum undiluted liquid waste flow during release to the product of the average flow from the site discharge structure to unrestricted receiving waters times 5.9. (5.9 is the site specific applicable factor for the mixing effect of the discharge structure.)
See the Nine Mile Point Unit 2 Environmental Report -Operating License Stage, Table 5.4-2 footnote 1.These factors can be related to batch release parameters as follows: FL = PEF / (TDF x 5.9) (Terms defined in Section 1.1.2.1 and above)ATLFL = [PEF (gpm) x ATL (min) x 1.67E-2 (hr/min)]
/ [TDF (gpm) x 5.9]= [TV x 2.83E-3 (hours)] / TDF For each batch, PEF (gpm) x ATL (min) = Tank Volume. For each batch, a dose calculation common constant (ATLFL) is calculated to be used with the concentration of each nuclide and dose factor, Ai, to calculate the dose to a receptor.
Normally, the highest dose factor for any age group (adult, teen, child, infant) will be used for.calculation, but specific age-group calculations to demonstrate compliance may be performed if required.1.4 Liquid Effluent Sampling Representativeness There are four tanks in the radwaste system designed to be discharged to the discharge canal. These tanks are labeled 4A, 4B, 5A, and 5B.Liquid Radwaste Tank 5A and 5B at Nine Mile Point Unit 2 contain a sparger spray ring which assists the mixing of the tank contents while it is being recirculated prior to Unit 2 Revision 28 119 September 2006 sampling.
This sparger effectively mixes the tank four times faster than simple recirculation.
Liquid Radwaste Tank 4A and 4B contain a mixing ring but no sparger. No credit is taken for the mixing effects of the ring. Normal recirculation flow is 150 gpm for tank 5A and 5B, 110 gpm for tank 4A and 4B while each tank contains up to 25,000 gallons although the entire contents are not discharged.
To assure that the tanks are adequately mixed prior to sampling, it is a plant requirement that the tank be recirculated for the time required to pass 2.5 times the volume of the tank: Recirculation Time = 2.5T/RM Where: Recirculation Time= Is the minimum time to recirculate the Tank, min 2.5 = Is the plant requirement, unitless T Is the tank volume, gal R -Is the recirculation flow rate, gpm.M -Is the factor that takes into account the mixing of the sparger, unitless, four for tank 5A and B, one for tank 4A and B.Additionally, the Alert Alarm setpoint of the Liquid Radwaste Effluent monitor is set at approximately 60% of the High alarm setpoint.
This alarm will give indication of incomplete mixing with adequate margin before exceeding ten times the effluent concentration.
Service Water A and B and the Cooling Tower Blowdown are sampled from the radiation monitor on each respective stream. These monitors continuously withdraw a sample and pump it back to the effluent stream. The length of tubing between the continuously flowing sample and the sample spigot contains less than 200 ml which is adequately purged by requiring a purge of at least 1 liter when grabbing a sample.1.5 Liquid Radwaste System FUNCTIONALITY The Liquid Radwaste Treatment System shall be FUNCTIONAL and used when projected doses due to liquid radwaste effluents would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period. Cumulative doses will be determined at least once per 31 days (as indicated in Section 1.3) and doses will also be projected if the radwaste treatment systems are not being fully utilized.The system collection tanks are processed as follows: 1) Low Conductivity (Waste Collector):
Radwaste Filter and Radwaste Demineralizer or the Thermex System.2) High Conductivity (Floor Drains): Regenerant Evaporator or the Thermex System.Unit 2 Revision 28 1110 September 2006
- 3) Regenerant Waste: If resin regeneration is used at NMP-2; the waste will be processed through the regenerant evaporator or Thermex System.The dose projection indicated above will be performed in accordance with the methodology of Section 1.3.Unit 2 Revision 28 II 11 September 2006 2.0 GASEOUS EFFLUENTS The gaseous effluent release points are the stack and the combined Radwaste/Reactor Building vent. The stack effluent point includes Turbine Building ventilation, main condenser offgas (after charcoal bed holdup), and Standby Gas Treatment System exhaust. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.2.1 Gaseous Effluent Monitor Alarm Setpoints 2.1.1 Basis The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following in accordance with Technical Specification 5.5.4.g: a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mremlyr to the skin, and b. For iodine- 131, for iodine- 133, for tritium, and for all radionuclides with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.The radioactivity rate of noble gases measured downstream of the recombiner shall be limited to less than or equal to 350,000 microcuries/second during offgas system operation in accordance with Technical Specification 3.7.4.2.1.2 Setpoint Determination Methodology Discussion Nine Mile Point Unit 1 and the James A FitzPatrick nuclear plants occupy the same site as Nine Mile Point Unit 2. Because of the independence of these plants' safety systems, control rooms and operating staffs it is assumed that simultaneous accidents are not likely to occur at the different units. However, there are two release points at Unit 2. It is assumed that if an accident were to occur at Unit 2 that both release points could be involved.The alarm setpoint for Gaseous Effluent Noble Gas Monitors are based on a dose rate limit of 500 mRem/yr to the Whole Body. Since there are two release points at Unit 2, the dose rate limit of 500 mRemlyr is divided equally for each release point, but may be apportioned otherwise, if required.
These monitors are sensitive to only noble gases.Because of this it is considered impractical to base their alarm setpoints on organ dose rates due to iodines or particulates.
Additionally skin dose rate is never significantly greater than the whole body dose rate. Thus the factor R which is the basis for the alarm setpoint calculation is nominally taken as equal to 250 mRem/yr. If there are significant releases from any gaseous release point on the site (>25 mRem/yr) for an extended period of time then the setpoint will be recalculated with an appropriately smaller value for R.The high alarm setpoint for the Offgas Noble Gas monitor is based on a limit of 350,000 uCi/sec. This is the release rate for which a FSAR accident analysis was completed.
At Unit 2 Revision 28 1112 September 2006 this rate the Offgas System charcoal beds will not contain enough activity so that their failure and subsequent release of activity will present a significant offsite dose assuming accident meteorology.
Initially, in accordance with Part I, Section D 3.3.2, the Germanium multichannel analysis systems of the stack and vent will be calibrated with gas standards (traceable to NIST) in accordance with DSR 3.3.2.9. Subsequent calibrations may be performed with gas standards, or with related solid sources. The quarterly Channel Functional Test will include FUNCTIONALITY of the 30cc chamber and the dilution stages to confirm monitor high range capability. (Appendix D, Gaseous Effluent Monitoring System).2.1.2.1 Stack Noble Gas Detector Alarm Setpoint Equation: The stack at Nine Mile Point Unit 2 receives the Offgas after charcoal bed delay, Turbine Building Ventilation and the Standby Gas Treatment system exhaust. The Standby Gas Treatment System Exhausts the primary containment during normal shutdowns and maintains a negative pressure on the Reactor Building to maintain secondary containment integrity.
The Standby Gas Treatment will isolate on high radiation detected (by the SGTS monitor) during primary containment purges.The stack noble gas detector is made of germanium.
It is sensitive to only gamma radiation.
However, because it is a computer based multichannel analysis system it is able to accurately quantify the activity released in terms of uCi of specific nuclides.Only pure alpha and beta emitters are not detectable, of which there are no common noble gases. A distribution of Noble Gases corresponding to offgas is chosen for the nominal alarm setpoint calculation.
Offgas is chosen because it represents the most significant contaminant of gaseous activity in the plant. The release rate Qj, corresponds to offgas concentration expected with the plant design limit for fuel failure. The alarm setpoint may be recalculated if a significant release is encountered.
In that case the actual distribution of noble gases will be used in the calculation.
The following calculation will be used for the initial Alarm Setpoint.O. 8R Y-i (Qi)Alarm Setpoint, [tCi/sec < Y-i(QiVi)0.8 = Safety Factor, unitless R -Allocation Factor. Normally, 250 mrem/yr; the value must be 500 mrem/yr or less depending upon the dose rate from other release points within the site such that the total dose rate corresponds to< 500 mrem/yr Qi= The release rate of nuclide i, ptCi/sec Vi -The constant for each identified noble gas nuclide accounting for the whole body dose from the elevated finite plume listed on Table D 3-2, mrem/yr per ptCi/sec Yi (Qi) The total release rate of noble gas nuclides in the stack effluent,ýtCi/sec Unit 2 Revision 28 1113 September 2006 Ei (QiVi) The total of the product of each isotope release rate times its respective whole body plume constant, mrem/yr The alert alarm is normally set at less than 10% of the high alarm.2.1.2.2 Vent Noble Gas Detector Alarm Setpoint Equation: The vent contains the Reactor Building ventilation above and below the refuel floor and the Radwaste Building ventilation effluents.
The Reactor Building Ventilation will isolate when radiation monitors detect high levels of radiation (these are separate monitors, not otherwise discussed in the ODCM). Nominal flow rate for the vent is 2.37E5 CFM.This detector is made of germanium.
It is sensitive to only gamma radiation.
However, because it is a computer based multichannel analysis system it is able to accurately quantify the activity released in terms of pCi of specific nuclides.
Only pure alpha and beta emitters are not detectable, of which there are no common noble gases. A distribution of Noble Gases corresponding to that expected with the design limit for fuel failure offgas is chosen for the nominal alarm setpoint calculation.
Offgas is chosen because it represents the most significant contaminant of gaseous activity in the plant.The alarm setpoint may be recalculated if a significant release is encountered.
In that case the actual distribution of noble gases will be used in the calculation.
0.-8R Fi (Qi)Alarm Setpoint, uCi/sec < (X/Q)v 2I(QiKi)Where: 0.8 -Safety Factor, unitless R = Allocation Factor. Normally, 250 mremlyr; the value must be 500 mrem/yr or less depending upon the dose rate from other release points within the site such that the total rate corresponds to < 500 mrem/yr Qi = The release rate of nuclide i, [tCi/sec (X/Q)v = The highest annual average atmospheric dispersion coefficient at the site boundary as listed in the Final Environmental Statement, NUREG 1085, Table D-2, 2.OE-6 sec/mi 3 K i The constant for each identified noble gas nuclide accounting for the whole body dose from the semi-infinite cloud, listed on Table D 3-3, mrem/yr per ýtCi/m 3 Unit 2 Revision 28 1114 September 2006 Yi (Qi) -The total release rate of noble gas nuclides in the vent effluent, uCi/sec Ei (QiKi) -The total of the product of the each isotope release rate times its respective whole body immersion constant, mrem/yr per sec/mr The alert alarm is normally set at less than 10% of the high alarm.2.1.2.3 Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation: The Offgas system has a radiation detector downstream of the recombiners and before the charcoal decay beds. The offgas, after decay, is exhausted to the main stack. The system will automatically isolate if its pretreatment radiation monitor detects levels of radiation above the high alarm setpoint.The Radiation Detector contains a plastic scintillator disc. It is a beta scintillation detector.
Detector response yj (ci/cFi) has been evaluated from isotopic analysis of offgas analyzed on a multichannel analyzer, traceable to NIST. A distribution of offgas corresponding to that expected with the design limit for fuel failure was used to establish the initial setpoint.
However, the alarm setpoint may be recalculated using an updated nuclide distribution based on actual plant process conditions.
The monitor nominal response values will be confirmed during periodic calibration using a Transfer Standard source traceable to the primary calibration performed by the vendor.Particulates and lodines are not included in this calculation because this is a noble gas monitor.To provide an alarm in the event of failure of the offgas system flow instrumentation, the low flow alarm setpoint will beset at or above 10 scfm, (well below normal system flow) and the high flow alarm setpoint will be set at or below 110 scfm, which is well above expected steady-state flow rates with a tight condenser.
To provide an alarm for changing conditions, the alert alarm will normally be set at 1.5 times nominal full power background to ensure that the Specific Activity Action required by ITS SR 3.7.4.1, are implemented in a timely fashion.(3.50E+05)
(2.12 E-03) j(__C/CFi)
+ Background Alarm Setpoint, cpm < 0.8 F i(CO)Where: Alarm Setpoint = The alarm setpoint for the offgas pretreatment Noble Gas Detector, cpm 0.8 -Safety Factor, unitless Unit 2 Revision 28 1115 September 2006 350,000 2.12E-03 Ci CFi F Background Ei (Ci/CFi)E (Ci)= The Technical Specification Limit for Offgas Pretreatment,ýtCi/sec= Unit conversion Factor, 60 sec/min / 28317 cc/CF= The concentration of nuclide, i, in the Offgas, [tCi/cc= The Detector response to nuclide i, pXi/cc/cpm; See Table D 3-1 for a list of nominal values-- The Offgas System Flow rate, CFM= The detector response to non-fission gases and general area dose rates, cpm-- The summation of the nuclide concentration divided by the corresponding detector response, net cpm= The summation of the concentration of nuclides in offgas,ýtCi/cc 2.2 2.2.1 Gaseous Effluents Dose Rate Calculation Dose rates will be calculated monthly at a minimum to demonstrate that the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the dose rate limits specified in 10CFR20. These limits are as follows: The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited per Technical Specification 5.5.4.g to the following:
- a. For noble gases: Less than or equal to 500 mremlyr to the whole body and less than or equal to 3000 mrem/yr to the skin, and b. For iodine-13 1, iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ: X/Q and W, -Dispersion Parameters for Dose Rate, Table D 3-23 The dispersion parameters for the whole body and skin dose rate calculation correspond to the highest annual average dispersion parameters at or beyond the unrestricted area boundary.
This is at the east site boundary.
These values were obtained from the Nine Mile Point Unit 2 Final Environmental Statement, NUREG 1085 Table D-2 for the vent and stack. These were calculated using the methodology of Regulatory Guide 1. 111, Rev. 1. The stack was modeled as an elevated release point because its height is more than 2.5 times any adjacent building height. The vent was modeled as a ground level release because even though it is higher than any adjacent building it is not more than 2.5 times the height.The NRC Final Environmental Statement values for the site boundary X/Q and D/Q terms were selected for use in calculating Effluent Monitor Alarm Points and compliance with Site Boundary Dose Rate specifications because they are conservative when compared with the corresponding Nine Mile Point Environmental Report values. In Unit 2 Revision 28 1116 September 2006 addition, the stack "intermittent release" X/Q was selected in lieu of the "continuous" value, since it is slightly larger, and also would allow not making a distinction between long term and short term releases.The dispersion parameters for the organ dose calculations were obtained from the Environmental Report, Figures 7B-4 (stack) and 7B-8 (vent) by locating values corresponding to currently existing (1985) pathways.
It should be noted that the most conservative pathways do not all exist at the same location.
It is conservative to assume that a single individual would actually be at each of the receptor locations.
2.2.2 Whole Body Dose Rate Due to Noble Gases The ground level gamma radiation dose from a noble gas stack release (elevated), referred to as plume shine, is calculated using the dose factors from Appendix B of this document.
The ground level gamma radiation dose from a noble gas vent release accounts for the exposure from immersion in the semi-infinite cloud. The dispersion of the cloud from the point of release to the receptor at the east site boundary is factored into the plume shine dose factors for stack releases and through the use of X/Q in the equation for the immersion ground level dose rates for vent releases.
The release rate is averaged over the period of concern. The factors are discussed in Appendix B.Whole body dose rate (DR)y due to noble gases: (DR)y = 3.17E-08 Yi [ViQis + Ki (X/Q)vQiv]
Where: DRy Whole body dose rate (mrem/sec)
Vi The constant accounting for the gamma whole body dose rate from the finite plume from the elevated stack releases for each identified noble gas nuclide, i. Listed on Table D 3-2, mrem/yr per pCi/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, i. Listed in Table D 3-3, mremlyr per uCi/m 3 (From Reg.Guide 1.109)X/Qv The relative plume concentration at or beyond the X/Qs land sector site boundary.
Average meteorological data is used.Elevated X/Q values are used for the stack releases (s=stack);
ground X/Q values are used for the vent releases (v-vent).
Listed on Table D 3-23 (sec/mi 3)Qis,QiV The release rate of each noble gas nuclide i, from the stack (s) or vent ,(v). Averaged over the time period of concern. (4tCi/sec)
Unit 2 Revision 28 1117 September 2006 3.17E-08 Conversion Factor; the inverse of the number of seconds in one year.(yr/sec)2.2.3 Skin Dose Rate Due to Noble Gases There are two types of radiation from noble gas releases that contribute to the skin dose rate: beta and gamma.For stack releases this calculation takes into account the dose from beta radiation in a semi infinite cloud by using an immersion dose factor. Additionally, the dispersion of the released activity from the stack to the receptor is taken into account by use of the factor (X/Q). The gamma radiation dose from the elevated stack release is taken into account by the dose factors in Appendix B.For vent releases the calculations also take into account the dose from the beta ([3) and gamma (y) radiation of the semi infinite cloud by using an immersion dose factor.Dispersion is taken into account by use of the factor (X/Q).The release rate is averaged over the period of concern.Skin dose rate (DR)+ due to noble gases: (DR) 3.17E-8 Yi [ (Li (X/Q)sý+I.IIBi)
Qir+ (Li+I.II1Mi) (X/Q) vQiv]Where: (DR) + = Skin dose rate (mrem/sec)
Li The constant to account for the gamma and beta skin dose rates for each noble gas nuclide, i, from immersion in the semi-infinite cloud, mrem/yr per ptCi/m 3 , listed on Table D 3-3 (from R.G. 1.109)Mi The constant to account for the air gamma dose rate for each noble gas nuclide, i, from immersion in the semi-infinite cloud, mrad/yr per p.Ci/m 3 , listed on Table D 3-3 (from R.G. 1.109)1.11 = Unit conversion constant, mrem/mrad.7 = Structural shielding factor, unitless Bi The constant accounting for the air gamma dose rate from exposure to the overhead plume of elevated releases of each identified noble gas nuclide, i. Listed on Table D 3-2, mrad/yr per ýtCi/sec.Unit 2 Revision 28 1118 September 2006 (X/Q)S The relative plume concentration at or beyond the land (X/Q)v sector site boundary.
Average meteorological data is used. Elevated X/Q values are used for the stack releases (s=stack);
ground X/Q values are used for the vent releases (v=vent). (sec/mi 3)3.17E-8 Conversion Factor; the inverse of the number of seconds in a year;(yr/sec)Qiv,Qis The release rate of each noble gas nuclide i, from the stack(s) or vent (v) averaged over the time period of concern, pCi/sec.2.2.4 Organ Dose Rate Due to 1-131, 1-133, Tritium, and Particulates with Half-lives greater than 8 days.The organ dose rate is calculated using the dose factors (Ri) from Appendix C. The factor Ri takes into account the dose rate received from the ground plane, inhalation and ingestion pathways.
W, and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways.
The release rate is averaged over the period of concern.Organ dose rates (DR)at due to iodine- 131, iodine- 133, tritium and all radionuclides in particulate form with half-lives greater than 8 days: (DR) at 3.17E-8 Yj[ZiRijat
[WsQis + WvQiv] I Where: (DR)at Organ dose rate (mrem/sec)
Rijat = The factor that takes into account the dose from nuclide i through pathway j to an age group a, and individual organ t. Units for inhalation pathway, mrem/yr per pCi/m 3.Units for ground and ingestion pathways, m 2-mrem/yr per uCi/sec. (See Tables D 3-4 through D 3-22).Ws, Wv Dispersion parameter either X/Q (sec/mi 3) or D/Q (1/M 2)depending on pathway and receptor location.
Average meteorological data is used (Table D 3-23). Elevated Ws values are used for stack releases (s=stack);
ground Wv values are used for vent releases (v-vent).Qi, Qiv The release rates for nuclide i, from the stack (s)and vent (v) respectively, pCi/sec.When the release rate exceeds 0.75 uCi/sec from the stack or vent, the dose rate assessment shall, also, include JAF and NMP 1 dose contributions.
The use of the 0.75 pCi/sec release rate threshold is conservative because it is based on the dose conversion Unit 2 Revision 28 1119 September 2006 factor (Ri) for the Sr-90 child bone which is significantly higher than the dose factors for the other isotopes present in the stack or vent release.2.3 Gaseous Effluent Dose Calculation Methodology Doses will be calculated monthly at a minimum to demonstrate that doses resulting from the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the limits specified in 10 CFR 50. These limits are as follows: The air dose from noble gases released in gaseous effluents, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following.
- a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radioactive material in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and, b. During any calendar year: Less than or equal to 15 mrem to any organ.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 from iodine and particulate releases, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.2.3.1 W, and W, -Dispersion Parameters for Dose, Table D 3-23 The dispersion parameters for dose calculations were obtained chiefly from the Nine Mile Point Unit 2 Environmental Report Appendix 7B. These were calculated using the methodology of Regulatory Guide 1.111 and NUREG 0324. The stack was modeled as an elevated release point because height is more than 2.5 times the height of any adjacent building.
The vent was modeled as a combined elevated/ground level release because the vent's height is not more than 2.5 times the height of any adjacent building.
Average meteorology over the appropriate time period was used. Dispersion parameters not available from the ER were obtained from C.T. Main Data report dated November, 1985, or the FES.Unit 2 Revision 28 II 20 September 2006 2.3.2 Gamma Air Dose Due to Noble Gases Gamma air dose from the stack or vent noble gas releases is calculated monthly. The gamma air dose equation is similar to the gamma dose rate equation except the receptor is air instead of the whole body or skin of whole body. Therefore, the stack noble gas releases use the finite plume air dose factors, and the vent noble gas releases use semi-infinite cloud immersion dose factors. The factor X/Q takes into account the dispersion of vent releases to the most conservative location.
The release activity is totaled over the period of concern. The finite plume factor is discussed in Appendix B.Gamma air dose due to noble gases: DY = 3.17E-8 Yi[Mi(X/Q)v Qiv + Bi Qjis x t D= The gamma air dose for the period of concern, mrad t= The duration of the dose period of concern, sec Where all other parameters have been previously defined.2.3.3 Beta Air Dose Due to Noble Gases The beta air dose from the stack or vent noble gas releases is calculated using the semi-infinite cloud immersion dose factor in beta radiation.
The factor X/Q takes into account the dispersion of releases to the most conservative location.Beta air dose due to noble gases: DP -3.17E-8 YiNi[(X/Q)v Qiv + (X/Q)s Qis] X t Dp = Beta air dose (mrad) for the period of concern Ni The constant accounting for the beta air dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed on Table D 3-3, mrad/yr per uCi/m 3.(From Reg. Guide 1.109).t = The duration of the dose period of concern, sec Where all other parameters have been previously defined.2.3.4 Organ Dose Due to 1-131, 1-133, Tritium and Particulates with half-lives greater than 8 days.The organ dose is based on the same equation as the dose rate equation except the dose is compared to the 1 OCFR50 dose limits. The factor Ri takes into account the dose received from the ground plane, inhalation, food (cow milk, cow meat and vegetation) pathways.W, and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways.
The release is totaled over the period of concern. The Ri factors are discussed in Appendix C.Organ dose Dat due to iodine- 131, iodine- 133, tritium and radionuclides in particulate Unit 2 Revision 28 1121 September 2006 form with half-lives greater than 8 days.Dat = 3.17E-8 yj [ Ei Rijat [Ws Qis + Wv Qiv] I X t Where: Dat = Dose to the critical organ t, for age group a, mrem t = The duration of the dose period of concern, sec Where all other parameters have been previously defined in Section 2.2.4.2.4 1-133 and 1-135 Estimation Stack and vent effluent iodine cartridges are analyzed to a sensitivity of at least 1 E- 12 uCi/cc. If detected in excess of the LLD, the 1-131 and 1-133 analysis results will be reported directly from each cartridge analyzed.
Periodically, (usually quarterly but on a monthly frequency if effluent iodines are routinely detected) a short-duration (12 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) effluent sample is collected and analyzed to establish an 1-135/1-131 ratio and an I-133/1-131 ratio, if each activity exceeds LLD. The short-duration ratio is used to confirm the routinely measured 1-133 values. The short-duration 1-135/I- 131 ratio (if determined) is used with the 1-131 release to estimate the 1-135 release. The short-duration 1-133/1-131 ratio may be used with the 1-131 release to estimate the 1-133 release if the directly measured 1-133 release appears non-conservative.
2.5 Isokinetic Sampling Sampling systems for the stack and vent effluent releases are designed to maintain isokinetic sample flow at normal ventilation flow rates. During periods of reduced ventilation flow, sample flow may be maintained at a minimum flow rate (above the calculated isokinetic rate) in order to minimize sample line losses due to particulate deposition at low velocity.2.6 Use of Concurrent Meteorological Data vs. Historical Data It is the intent to use dispersion parameters based on historical meteorological data to set alarm points and to determine or predict dose and dose rates in the environment due to gaseous effluents.
If effluent levels approach limiting values, meteorological conditions concurrent with the time of release may be used to determine gaseous pathway doses.2.7 Gaseous Radwaste Treatment System Operation Part I, Section D 3.2.4 requires the GASEOUS RADWASTE TREATMENT SYSTEM to be in operation whenever the main condenser air ejector system is in operation.
The system may be operated for short periods with the charcoal beds bypassed to facilitate transients.
The components of the system which normally should operate to treat offgas Unit 2 Revision 28 1122 September 2006 2.8 are the Preheater, Recombiner, Condenser, Dryer, Charcoal Adsorbers, HEPA Filter, and Vacuum Pump. (See Appendix D, Offgas System).Ventilation Exhaust Treatment System Operation Part I, Section D 3.2.5 requires the VENTILATION EXHAUST TREATMENT SYSTEM to be FUNCTIONAL when projected doses in 31 days due to iodine and particulate releases would exceed 0.3 mrem to any organ of a member of the public. The appropriate components, which affect iodine or particulate release, to be FUNCTIONAL are: 1)2)3)HEPA Filter -Radwaste Decon Area HEPA Filter -Radwaste Equipment Area HEPA Filter -Radwaste General Area Whenever one of these filters is not FUNCTIONAL, iodine and particulate dose projections will be made for 31-day intervals starting with filter nonfunctionality, and continuing as long as the filter remains nonfunctional, in accordance with DSR 3.2.5.1.Predicted release rates will be used, along with the methodology of Section 2.3.4. (See Appendix D, Gaseous Radiation Monitoring.)
Unit 2 Revision 28 1123 September 2006 3.0 URANIUM FUEL CYCLE The "Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b) as follows: "Uranium fuel cycle means the operations of milling of uranium ore chemical conversion of uranium, isotopic enrichment 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, to the 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 recovered non-uranium special nuclear and by-product materials from the cycle." Sections D 3.1.2, D 3.2.2, and D 3.2.3 of Part I requires that when the 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 fuel 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 result in 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 reduce subsequent releases.The report to the NRC shall contain: 1) Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site, that contribute to the annual dose of the maximum exposed member of the public.2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from all existing pathways and sources of radioactive effluents and direct radiation.
The total body and organ doses resulting from radioactive material in liquid effluents from Nine Mile Point Unit 2 will be summed with the doses resulting from the releases of noble gases, radioiodines, and particulates.
The direct dose components will also 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 by Technical Specification 5.6.3. The doses from Nine Mile Point Unit 2 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.Unit 2 Revision 28 1124 September 2006 For the purpose of calculating doses, the results of the Environmental Monitoring Program may be included to provide more refined estimates of doses to a real maximum exposed individual.
Estimated doses, as calculated from station effluents, may be replaced by doses calculated from actual environmental sample results.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 effluent data and Regulatory Guide 1.109 methodology or by calculating a dose to man based on actual fish sample 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 adult. 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 sediment sample analysis data.Equations used to evaluate fish and shoreline sediment samples are based on Regulatory Guide 1.109 methodology.
Because of the sample medium type and the half-lives of the radionuclides historically observed, the decay corrected portions of the equations are deleted. This does not reduce the conservatism of the calculated doses but increases the simplicity from an evaluation point of view. Table D 3-24 presents the parameters used for calculating doses from liquid effluents.
The dose from fish sample media is calculated as: Rapj = Yi [Cif (U) (Daipj) f] (IE+3)Where: Rapj The total annual dose to organ j, of an individual of age group a, from nuclide i, via fish pathway p, in mrem per year; ex. if calculating to the adult whole body, then Rapj = Rwb and Daipj = DiWB Cif = The concentration of radionuclide i in fish samples in pCi/gram U = The consumption rate of fish 1 E+3 = Grams per kilogram (Daipj) The ingestion dose factor for age group a, nuclide i, fish pathway p, and organ j, (Reg. Guide 1.109, Table E- 11) (mrem/pCi).
ex. when calculating to the adult whole body Daipj = DiWB f The fractional portion of the year over which the dose is applicable Unit 2 Revision 28 1125 September 2006 The dose from shoreline sediment sample media is calculated as: Rapj = Yj [Cis (U) (4E+4) (0.3) (Daipj) f]Where: Ra'j .The total annual dose to organ j, of an individual of age group a, from nuclide i, via the sediment pathway p, in mrem per year; ex. if calculating to the adult whole body, then Rapj = RWB and Daipj = DiWB Ci= The concentration of radionuclide i in shoreline sediment in pCi/gram 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 0.3 = The shore width factor for a lake Daipj = The dose factor for age group a, nuclide i, sediment pathway s, and organ j. (Reg. Guide 1.109, Table E-6) (mrem/hr per pCi/m 2); ex.when calculating to the adult whole body Daipj = DiWB f = The fractional portion of the year over which the dose is applicable NOTE: Because of the nature of the receptor location and the extensive fishing in the area, the critical individual may be a teenager or an adult.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 2 of the calculational methodologies section will be considered and include ground 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, etc. 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 effluent calculational data.Doses to members of the public from the pathways considered in section 2 as a result of gaseous effluents will be calculated using the methodology of Regulatory Guide 1.109 or the methodology of the ODCM, as applicable.
Doses calculated from environmental sample media will be based on methodologies found in Regulatory Guide 1.109.Unit 2 Revision 28 II 26 September 2006 3.3 Evaluation of Doses From Direct Radiation The dose contribution as a result of direct radiation shall be considered when 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 location and 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 due to their activities inside the site boundary as defined by Figure D 1.0-1. A member of the public, would be represented by an individual who visits the sites' Energy Center for the purpose of observing the educational displays or for picnicking and associated activities.
Fishing is a major recreational activity in the area and on the Site as a result of the salmon 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 boundary.
Based on the nature of the fishermen and undocumented observations, it is conservatively assumed that the maximum exposed individual spends an average of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per week fishing from the shoreline at a location between the Energy Center and the Unit 1 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 with the resultant lung dose, the ground dose pathway with the resultant whole body and skin dose and the direct radiation dose pathway with the associated total 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, possible direct radiation dose from the facility and a ground plane dose (deposition).
Because the location is in close proximity to the site, any beta plume submersion dose is felt to be insignificant.
Other pathways, such as the ingestion pathway, are not applicable.
In addition, pathways associated with water related recreational activities, other than fishing, are not applicable here. These include swimming, boating and wading which are prohibited at the facility.Unit 2 Revision 28 1127 September 2006 The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium 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 the following equation adapted from Regulatory Guide 1.109. Table D 3-24 presents the reference for the parameters used in the following equation.NOTE: The following equation is adapted from equations C-3 and C-4 of Regulatory Guide 1.109. Since many of the factors are in units of pCi/m 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 to the Regulatory Guide equations.
Dja Ei [ (Ci) F (X/Q) (DFA) ija (BR) at]Where: Dja The maximum dose from all nuclides to the organ j and age group (a) in mrem/yr; ex. if calculating to the adult lung, then Dja = DL and DFAija = DFAiL C i The average concentration in the stack or vent release of nuclide i for the period in pCi/rn 3.F Unit 2 average stack or vent flowrate in m 3/sec.X/Q = The plume dispersion parameter for a location approximately 0.50 miles west of NMP-2 (The plume dispersion parameters are 9.6E-07 (stack) and 2.8E-06 (vent) and were obtained from the C.T. Main five year average annual X/Q tables. The vent X/Q (ground level) is ten times the listed 0.50 mile X/Q because the vent is approximately 0.3 miles from the receptor location.
The stack (elevated)
X/Q is conservative when based on 0.50 miles because of the close proximity of the stack and the receptor location.(DFA)ija the dose factor for nuclide i, organ j, and age group a in mrem per pCi (Reg. Guide 1.109, Table E-7); ex. if calculating to the adult lung the DFAija = DFAiL (BR)a annual air intake for individuals in age group a in M3 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 2 Revision 28 1128 September 2006 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 Regulatory Guide 1.109 as presented in Section 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 fishing is not performed from the shoreline but is instead performed in the water (i.e., the use of waders), then the ground dose pathway (deposition) will not be evaluated.
The direct radiation gamma dose pathway includes any gamma doses from an overhead plume, submersion in the plume, possible 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 used at one location in the approximate area where fishing occurs. The TLDs will be placed in the field on approximately the beginning of each calendar quarter and removed approximately at the end of each calendar quarter (quarter 2, 3, and 4).The average TLD readings will be 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 used after adjusting for the appropriate time period (as applicable).
In the event of loss or theft of the TLDs, results from a TLD or TLDs in a nearby area may be utilized.Unit 2 Revision 28 1129 September 2006 4.0 ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Stations The current sampling locations are specified in Table D 5-1 and Figures D 5.1-1 and D 5.1-2. The meteorological tower location is shown on Figure D 5.1-1 and is located where TLD location #17 is identified.
The Environmental Monitoring Program is a joint effort between the owners and operators of the Nine Mile Point Units 1 and 2 and the James A. FitzPatrick Nuclear Power Plants. 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 D 5-1 are based on the NMP-2 reactor centerline.
The average dispersion and deposition parameters for the three units have been calculated for a 5 year period, 1978 through 1982. Average dispersion or deposition parameters for the site are calculated using the 1978 through 1982 data and are used to compare the results of the annual land use census. If it is determined that sample locations required by Control D 3.5.1 are unavailable or new locations are identified that yield a significantly higher (i.e., 50%) calculated D/Q value, actions will be taken as required by Controls D 3.5.1 and D 3.5.2 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 Interlaboratory 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 which cross check samples are available.
An attempt will be made to obtain a QC sample to program sample ratio of 5% or better. The Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission staff may evaluate the results.Specific sample media for which EPA 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 0 tritium in water* 1- 13 1 in water Unit 2 Revision 28 1130 September 2006 4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements required by the Technical Specifications are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use. In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabilities may be determined by the vendor who supplies the TLDs. Required detection capabilities are as follows.4.3.1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less 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 10 uR/hr during the field cycle. The average of the relative 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 interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constant.
This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures.
For these tests, the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85. At least 6 TLDs shall be evaluated.
4.3.4 Energy dependence shall be evaluated by the response of TLDs to photons for several energies between approximately 30 keV and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25% for photons with energies greater than 80 keV 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, the TLD 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 be 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 not 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 2 Revision 28 1131 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 phosphor, as appropriate, shall be dried before readout. The response of the TLD exposed in the plastic bag containing water shall not differ from that exposed in the regular plastic bag by more than 10%. A total of at least 4 TLDs shall be evaluated for each condition.
4.3.8 Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is less than 10 uRlhr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3). The average exposure inferred from the responses of the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least 3 TLDs shall be evaluated.
Unit 2 Revision 28 1132 September 2006 TABLE D 2-1 LIQUID EFFLUENT DETECTORS RESPONSES*
NUCLIDE (CPM/iiCi/ml x 101)Sr 89 Sr 91 Sr 92 Y91 Y 92 Zr 95 Nb 95 Mo 99 Tc 99m Te 132 Ba 140 Ce 144 Br 84 1131 1132 1133 1134 1135 Cs 134 Cs 136 Cs 137 Cs 138 Mn 54 Mn 56 Fe 59 Co 58 Co 60 0.78E-04 1.22 0.817 2.47 0.205 0.835 0.85 0.232 0.232 1.12 0.499 0.103 1.12 1.01 2.63 0.967 2.32 1.17 1.97 2.89 0.732 1.45 0.842 1.2 0.863 1.14 1.65* Values from SWEC purchase specification NMP2-P28 IF.Unit 2 Revision 28 1133 September 2006 TABLE D 2-2 Aiat VALUES -LIQUID'ADULT mrem -ml hr- uCi NUCLIDE T BODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H3 Cr 51 Cu 64 Mn 54 Fe 55 Fe 59 Co 57 Co 58 Co 60 Zn 65 Sr 89 Sr 90 Sr 92 Zr 95 Mn 56 Mo 99 Na 24 1131 1132 1 133 1 135 Ni 65 Cs 134 Cs 136 Cs 137 Ba 140 Ce 141 Nb 95m Nb 95 La 140 Ce 144 Tc 99m Np 239 Te 132 Zr 97 W 187 Ag 110m Sb 124 Zn 69m Au 199 As 76 3.67E-1 1.26 1.28 8.38E2 1.07E2 9.28E2 5.43E1 2.01E2 6.36E2 3.32E4 6.38E2 1.36E5 1.44E-2 7.59E-1 3.07E-2 1.60E1 1.34E2 1.16E2 4.34E-3 1.22E1 1.32E0 1.14E-2 5.79E5 8.42E4 3.42E5 1.37E1 3.79E-2 1.51El 1.31E2 1.62E-2 3.03E-1 2.05E-2 1.8E-3 1.1 8E3 5.08E-4 4.31E1 1.09El 4.72E1 5.40E1 3.95 5.94 3.67E-1 3.13E2 2.33E2 1.34E4 2.62E2 8.06E3 5.36E2 1.81E3 4.93E3 4.63E4 3.57E3 1.60E4 6.61 2.83E2 5.52 1.95E2 1.34E2 5.36EI 2.33E-3 3.59E1 3.79E0 6.35E-1 1.24E4 1.33E4 1.01E4 4.30E2 8.81E1 1.44E6 1.48E6 3.72E3 6.15E2 9.54E-01 4.47E2 5.97E4 3.39E2 4.04E4 3.94E2 3.36E2 3.60E4 7.33E2 1.24E4 1.18E-2 3.98 6.62E2 1.03E3 1.07 6.47E1 2.31 E4 2.22E4 5.55E5 3.34E-1 9.77E-1 1.97E-3 1.34E2 1.42E2 4.64E-3 2.30E1 1.28E0 1.93E-1 2.98E5 2.96E4 3.82E5 2.09E2 6.93E-2 3.53E1 4.38E2 1.03E-1 2.02 5.71E-4 2.28E-2 1.95E3 5.44E-3 1.48E2 1.14E1 1.07E3 2.46E2 1.26E-1 1.60E-1 3.67E-1 1.1 8E-2 2.73 4.38E3 4.57E2 2.42E3 2.11E1 9.04E 1 3.24E2 7.3 5E4 6.18E-5 7.88E-1 1.73E-1 8.42E1 1.34E2 2.03E2 1.24E-2 3.99E1 3.36E0 2.50E-2 7.08E5 1.17E5 5.22E5 3.04E-1 5.83E-2 2.74E1 2.44E2 5.36E-2 9.66E-1 1.6 1E-3 2.78E-3 1.26E3 1.1OE-3 1.23E2 1.13E1 4.33E1 5.90E2 4.67 6.19 3.67E-1 2.86E-1 6.89 1.31 E3 7.53E-1 1.07 6.47E1 4.92E4 6.18E-5 8.39E-1 2.20E-1 1.91E2 1.34E2 3.48E2 1.98E-2 6.97E1 5.39E0 2.29E5 6.51 E4 1.77E5 1.31E-1 4.60E-2 2.70E1 2.41 E2 2.83E-3 6.57E-1 2.45E-2 7.40E-3 1.22E4 1.66E-3 4.43E-5 1.22E1 4.31EI 3.57E2 1.79E1 1.16El 3.67E-1 7.56E-1 3.98 7.53E-1 1.07 6.47E1 2.21 6.18E-5 6.99E- I 1.97E-3 1.34E2 6.65E4 4.34E- 1 5.87E3 2.22E2 2.04EI 3.28E-1 3.10EI 4.17E-2 3.53E-2 3.56E-1 2.83E-3 2.06E-1 5.95E-4 1.39E3 7.11E-6 4.43E-5 1.04El 4.31E1 6.90E-2 1.26E-1 1.60E-1 3.67E-1 1.66 3.98 2.55E2 6.76E2 1.07 6.47E1 2.21 6.18E-5 6.99E- I 1.97E-3 1.34E2 2.77E-2 7.61 E4 8.92E3 5.89E4 1.92E-1 3.53 E-2 3.56E-1 2.83E-3 2.06E- 1 7.90E-4 5.95E-4 2.66E-3 7.11E-6 4.43E-5 1.04El 5.12E1 6.90E-2 1.26E-1 1.60E-1 Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.Unit 2 Revision 28 1134 September 2006 TABLE D 2-3 Aiat VALUES -LIQUID'TEEN NUCLIDE H3 Cr 51 Cu 64 Mn 54 Fe 55 Fe 59 Co 57 Co 58 Co 60 Zn 65 Sr 89 Sr 90 Sr 92 Zr 95 Mn 56 Mo 99 Na 24 1131 1132 1133 1135 Ni 65 Cs 134 Cs 136 Cs 137 Ba 140 Ce 141 Nb 95m Nb 95 La 140 Ce 144 Tc 99m Np 239 Te 132 Zr 97 W 187 Ag 110m Sb 124 Zn 69m Au 199 As 76 T BODY 2.73E-1 1.35 1.35 8.75E2 1.15E2 9.59E2 1.44E2 2.1 0E2 9.44E2 3.40E4 6.92E2 1.14E5 1.54E-2 3.96 3.22E-2 1.71EI 1.38E2 1. 14E2 4.56E-3 1.28E1 1.76E0 1.21E-2 3.33E5 7.87E4 1.90E5 1.44E1 2.OOE-1 1.69E1 1.17E2 2.97E-2 1.25 2.11E-2 4.63E-3 1.23E3 5.68E-4 4.55E1 5.85E1 2.45E2 5.76E1 4.85 7.18 GI-TRACT 2.73E-1 2.16E2 2.23E2 8.84E3 2.13E2 5.85E3 4.08E2 1.23E3 3.73E3 3.08E4 2.88E3 1.30E4 9.19E1 2.1 0E2 1.19E 1 1.60E2 1.38E2 4.21El 5.54E-3 3.17E1 3.84E0 1.44 9.05E3 9.44E3 7.91E3 3.40E2 6.85E1 1.14E6 1.05E6 3.01E3 4.83 E2 1.07 3.78E2 4.13E4 3.11 E2 3.52E4 3.17E2 4.53E2 3.43E4 5.78E2 1.06E4 BONE 6.56E-2 2.22E1 6.93E2 1.06E3 5.98 3.61 E2 2.10E4 2.42E4 4.62E5 3.61 E- 1 4.19 1.1OE-2 1.38E2 1.52E2 4.86E-3 2.47E1 1.34E0 2.08E-1 3.05E5 2.98E4 4.09E5 2.21 E2 2.33E-1 3.87E1 4.43E2 1.22E-1 3.07 5.84E-4 2.82E-2 2.06E3 5.84E-3 1.59E2 5.89E1 2.51 E2 2.65E2 7.04E-1 8.92E-1 mrem -ml hr -uCi LIVER 2.73E-1 6.56E-2 2.87 4.32E3 4.91 E2 2.48E3 2.19E1 9.47E1 6.20E2 7.28E4 3.45E-4 3.99 1.81E-I 8.95E1 1.38E2 2.12E2 1.27E-2 4.19E1 3.46E0 2.66E-2 7.18E5 1.17E5 5.44E5 5.03E-1 2.21E-1 2.99E1 2.47E2 6.82E-2 1.94 1.63E-3 5.67E-3 1.30E3 1. 19E-3 1.30E2 5.88E1 2.41 E2 6.24E2 5.60 7.40 KIDNEY 2.73E-1 3.47E-1 7.27 1.31E3 4.20 5.98 3.61 E2 4.66E4 3.45E-4 4.03 2.29E-1 2.05E2 1.38E2 3.66E2 2.OOE-2 7.35E1 5.47E0 2.28E5 6.38E4 1.85E5 3.25E-1 2.08E-1 2.96E1 2.39E2 1.58E-2 1.62 2.43E-2 1.07E-2 1.25E4 1.78E-3 2.47E-4 5.97EI 2.41 E2 3.79E2 2.01EI 1.33E1 THYROID 2.73E- 1 7.79E- 1 2.22E1 4.20 5.98 3.61E2 1.24E1 3.45E-4 3.90 1.10E-2 1.38E2 6.19E4 4.29E- 1 5.85E3 2.23E2 1.14E2 1.83 1.73E2 2.33E-1 1.97E-1 1.99 1.58E-2 1.15 3.32E-3 1.37E3 3.97E-5 2.47E-4 5.79E 1 2.41 E2 3.85E-1 7.04E- 1 8.92E- 1 LUNG 2.73E-1 1.90 2.22E1 3.11 E2 7.84E2 5.98 3.61 E2 1.24E1 3.45E-4 3.90 1.1OE-2 1.38E2 1.55E-1 1.02E-4 8.72E4 1.01E4 7.21 E4 4.15E-1 1.97E-1 1.99 1.58E-2 1.15 9.04E-4 3.32E-3 1.48E-2 3.97E-5 2.47E-4 5.79E1 2.50E2 3.85E-1 7.04E-1 8.92E-1'Calculated in accordance with NUREG 0133, Section 4.3. 1; and Regulatory Guide 1. 109, Regulatory position C, Section 1.Unit 2 Revision 28 1135 September 2006 TABLE D 2-4 Ajat VALUES -LIQUID'CHILD NUCLIDE H3 Cr 51 Cu 64 Mn 54 Fe 55 Fe 59 Co 57 Co 58 Co 60 Zn 65 Sr 89 Sr 90 Sr 92 Zr 95 Mn 56 Mo 99 Na 24 1131 1132 1133 1135 Ni 65 Cs 134 Cs 136 Cs 137 Ba 140 Ce 141 Nb 95m Nb 95 La 140 Ce 144 Tc 99m Np 239 Te 132 Zr 97 W 187 Ag 1 10m Sb 124 Zn 69m Au 199 As 76 T BODY 3.34E-1 1.39 1.60 9.02E2 1.50E2 1.04E3 6.24EI 2.21 E2 7.03E2 3.56E4 9.13E2 1.06E5 1.85E-2 8.95E-1 3.73E-2 2.22E1 1.51E2 1. 14E2 5.08E-3 1.51El 1.53E0 1.46E-2 1.27E5 6.26E4 7.28E4 1.87E1 4.61 E-2 2.14E1 1.45E2 1.93E-2 4.31E-1 2.29E-2 2.40E-3 1.38E3 6.99E-4 5.37E1 1.29E1 5.69E I 6.80E1 5.58 8.31 GI-TRACT 3.34E-1 7.29E1 1.25E2 2.83E3 8.99EI 2.18E3 1.62E2 4.20E2 1.25E3 1.01E4 1.24E3 5.62E3 8.73 9.36E1 2.39E1 7.42E1 1.51E2 1.80E1 1.30E-2 1.60E1 2.30E0 3.07 3.28E3 3.40E3 3.12E3 1.62E2 4.14E1 5.28E5 3.75E5 1.33E3 2.92E2 7.87E-1 1.79E2 1.1 5E4 1.77E2 1.68E4 1.24E2 1.68E2 1.87E4 2.75E2 5.47E3 BONE 1.37E-2 4.65 9.15E2 1.29E3 1.25 7.55E1 2.15E4 3.20E4 4.17E5 4.61E-I 1.22 2.30E-3 1.51 E2 2.00E2 6.01E-3 3.22E1 1.68E0 2.66E-1 3.68E5 3.52E4 5.15E5 3.19E2 1.08E-1 4.99E1 5.21 E2 1.39E-1 3.81 7.05E-4 3.44E-2 2.57E3 8.11E-3 2.02E2 1.35E1 6.92E1 3.37E2 1.47E-1 1.86E-1 mrem -ml hr -uCi LIVER 3.34E-1 1.37E-2 2.65 3.37E3 4.85E2 2.09E3 2.OOE1 7.30E1 2.88E2 5.73E4 9.04E- 1 1.65E-1 8.98E1 1.51E2 2.01E2 1.1OE-2 3.98E1 3.02E0 2.5 1E-2 6.04E5 9.67E4 4.93E5 3.28E-1 7.43E-2 2.92E1 2.03E2 5.09E-2 1.36 1.38E-3 3.12E-3 1.14E3 1.18E-3 1.20E2 1.30E1 5.06E1 5.75E2 5.02 6.58 KIDNEY 3.34E-1 2.22E- 1 6.41 9.49E2 8.78E-1 1.25 7.55E1 3.61 E4 9.43E-1 2.OOE-I 1.92E2 1.51 E2 3.31 E2 1.69E-2 6.64E1 4.63E0 1.87E5 5.15E4 1.61E5 1.40E-1 5.57E-2 2.68E1 1.91E2 3.30E-3 8.61E-1 2.01E-2 7.70E-3 1.06E4 1.69E-3 5.16E-5 1.39E1 5.03E1 3.34E2 1.80El 1.15EI THYROID 3.34E- 1 7.76E- 1 4.65 8.78E- I 1.25 7.55E1 2.58 8.15E-1 2.30E-3 1.51E2 6.66E4 5.13E-1 7.40E3 2.67E2 2.38E1 3.82E-1 3.62E I 4.87E-2 4.12E-2 4.16E-1 3.30E-3 2.40E- 1 6.94E-4 1.66E3 8.29E-6 5.16E-5 1.21El 5.04E1 8.05E-2 1.47E-1 1.86E-1 LUNG 3.34E-1 1.41 4.65 2.74E2 6.08E2 1.25 7.55EI1 2.58 8.15E-1 2.30E-3 1.51E2 3.23E-2 6.72E4 7.68E3 5.78E4 2.15E-1 4.12E-2 4.16E-1 3.30E-3 2.40E-1 7.02E-4 6.94E-4 3.1OE-3 8.29E-6 5.16E-5 1.21El 6.08E1 8.05E-2 1.47E-1 1.86E-1 1 Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.Unit 2 Revision 28 1136 September 2006 TABLE D 2-5 Ajat VALUES -LIQUID'INFANT mrem -ml hr- uCi NUCLIDE T BODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H3 Cr 51 Cu 64 Mn 54 Fe 55 Fe 59 Co 57 Co 58 Co 60 Zn 65 Sr 89 Sr 90 Sr 92 Zr 95 Mn 56 Mo 99 Na 24 1131 1132 1133 1135 Ni 65 Cs 134 Cs 136 Cs 137 Ba 140 Ce 141 Nb 95m Nb95 La 140 Ce 144 Tc 99m Np 239 Te 132 Zr 97 W 187 Ag 110m Sb 124 Zn 69m Au 199 As 76 1.87E-1 8.2 1E-3 1.96E-2 2.73 1.45 1.25E1 1. 13E0 5.36 1.55E1 1.76E1 4.27E1 2.86E3 1.56E-5 2.12E-2 1.81E-6 2.65 9.61E-1 9.78 3.43E-6 8.26E-1 2.3 8E2 2.96E-6 4.30E1 2.81E1 2.63E1 4.88 3.3 1E-3 1.02E3 5.87E-3 6.52E-4 1.01E-1 3.17E-4 2.08E-4 4.08 1.38E-4 4.13E-2 2.91E-1 1.87E-1 2.39E-1 8.70E-1 4.42 6.91E-1 1.52E1 2.37E0 5.36 1.56E1 3.22E1 3.06E1 1.40E2 4.54E-3 1.49E1 9.56E-4 4.48 9.61E-1 7.94E-1 7.80E-6 4.77E-1 2.36E2 4.96E-4 1.16 1.14 1.16 2.33E1 1.45E1 1.20El 8.57 2.98E1 1.03E2 7.14E-3 1.06E1 1.62E1 1.92E1 7.02 2.28E1 3.93E1 3.50 5.38 2.85E1 8.42 1.82E1 1.11El 1.49E3 1.12E4 4.2 1E-4 1.23E-1 9.61E-1 1.89E1 4.75E-6 1.94 3.29E2 5.75E-5 2.28E2 2.56E1 3.17E2 9.48E1 4.6 1E-2 2.39E3 2.47E-2 6.43E-3 1.80 1.19E-5 4.12E-3 8.83 1.76E-3 1.72E-1 6.02E-1 1.27E1 1.24E-1 1.87E-1 4.24E-2 1.20E1 5.44 3.18E1 6.95E1 2.15 6.55 3.81E1 2.99E-2 1.05E-5 1.36E1 9.61E-1 2.22E1 9.63E-6 2.82 6.54E2 6.5 1E-6 4.26E2 7.53E1 3.71 E2 9.48E-2 2.81 E-2 1.73E3 1.02E-2 2.53E-3 7.37E-1 2.46E-5 3.68E-4 4.37 3.02E-4 1.19E-1 4.39E-1 1.87E-1 2.52E-1 2.48E-1 8.46E-2 1.87E-1 1.1 7E-3 7.17E-2 2.67 1.85E1 3.23E-2 9.05E-6 2.03E1 9.61E-1 2.60E1 1.07E-5 3.31 7.28E2 1.1 0E2 3.OOE1 9.95E1 2.25E-2 8.67E-3 1.1 0E3 7.28E-3 2.98E-1 2.64E-4 7.34E-4 2.74E1 3.04E-4 6.28E-1 1.02E-1 6.26E-1 1.03E-1 1.87E-1 5.36E-3 9.61E-1 7.31 E3 4.52E-4 5.13E2 5.86E0 6.46 3.38E-2 1.87E-1 1.04E-2 2.66 9.41 9.61E-1 4.50E1 6.13 4.03E1 5.82E-2 1.28E-5 7.98 3.95 2.30E-2 2.23E-1 8.67E-2'Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.Unit 2 Revision 28 1137 September 2006 TABLE D 3-1 OFFGAS PRETREATMENT*
DETECTOR RESPONSE NUCLIDE Kr 83m Kr85 Kr 85m Kr87 Kr88 Kr89 Xe 131m Xe 133 Xe 133m Xe 135 Xe 135m Xe 137 Xe 138 NET CPMh/Ci/cc 4.28E+03 3.85E+03 6.68E+03 3.97E+03 6.48E+03 1.69E+03 4.91 E+03 6.89E+03 5.5 1E+03* Values from calculation H21 C-070 Unit 2 Revision 28 September 2006 1138 TABLE D 3-2 PLUME SHINE PARAMETERS 1 NUCLIDE Kr 83m Kr85 Kr 85m Kr 87 Kr88 Kr89 Kr 90 Xe 131m Xe 133 Xe 133m Xe 135 Xe 135m Xe 137 Xe 138 Xe- 127 Ar 41 Bi mrad/yr uCi/sec 9.01E-7 6.92E-7 5.09E-4 2.72E-3 7.23E-3 1.15E-2 6.57E-3 7.76E-6 7.46E-5 4.79E-5 7.82E-4 1.45E-3 6.25E-4 4.46E-3 1.96E-3 5.OOE-3 Vi mrem/yr uCi/sec 4.91E-4 2.57E-3 7.04E-3 1.13E-2 4.49E-3 6.42E-5 3.95E-5 7.44E-4 1.37E-3 5.98E-4 4.26E-3 1.31E-3 4.79E-3 Bi and Vi are calculated for critical site boundary location; 1.6km in the easterly direction.
See Appendix B. Those values that show a dotted line were negligible because of high energy absorption coefficients.
Unit 2 Revision 28 1139 September 2006 TABLE D 3-3 IMMERSION DOSE FACTORS 1 Nuclide Kr 83m Kr 85m Kr85 Kr87 Kr88 Kr89 Kr 90 Xe 131m Xe 133m Xe 133 Xe 135m Xe 135 Xe 137 Xe 138 Ar 41 K_ (y-Body)2 7.56E-02 1.17E3 1.61El 5.92E3 1.47E4 1.66E4 1.56E4 9.15E1 2.51 E2 2.94E2 3.12E3 1.81E3 1.42E3 8.83E3 8.84E3 Li (3-Skin) 2 1.46E3 1.34E3 9.73E3 2.37E3 1.01E4 7.29E3 4.76E2 9.94E2 3.06E2 7.11E2 1.86E3 1.22E4 4.13E3 2.69E3 Mi(y-Air) 3 1.93E1 1.23E3 1.72E1 6.17E3 1.52E4 1.73E4 1.63E4 1.56E2 3.27E2 3.5 3E2 3.36E3 1.92E3 1.51E3 9.2 1E3 9.30E3 1 (1Air) 3 2.88E2 1.97E3 1.95E3 1.03E4 2.93E3 1.06E4 7.83E3 1.11 E3 1.48E3 1.05E3 7.39E2 2.46E3 1.27E4 4.75E3 3.28E3'From, Table B-i .Regulatory Guide 1.109 Rev. 1 2 mrem/yr per uCi/m 3.3 mrad/yr per uCi/m 3.Unit 2 Revision 28 September 2006 1140 TABLE D 3-4 DOSE AND DOSE RATE Ri VALUES -INHALATION
-INFANT'm rem/yr uCi/m 3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY 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 1 133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag I 10m 2.65E4 1.97E4 1.36E4 1.93E4 3.98E5 4.09E7 1. 15E5 1.57E4 3.79E4 1.32E4 3.96E5 5.49E5 5.60E4 5.05E2 2.77E4 3.19E6 7.94E3 9.99E3 6.47E2 5.3 1E3 2.53E4 1.17E4 2.35E4 1.22E3 8.02E3 6.26E4 2.79E4 6.43E3 1.65E2 4.44E4 1.92E4 7.03E5 6.12E5 5.60E I 2.00E2 1.67E4 1.21E6 8.13E3 7.22E3 6.47E2 5.3 1E3 8.95E1 4.98E3 3.33E3 9.48E3 1.82E3 1.18E4 3.11 E4 1.14E4 2.59E6 2.03E4 3.78E3 3.23E1 1.96E4 5.60E3 7.45E4 4.55E4 2.90E3 5.15E1 1.99E3 1.76E5 5.00E2 5.00E3 6.47E2 5.3 1E3 5.75E1 1.48E7 3.56E6 6.47E2 5.3 1E3 1.32E1 4.98E3 3.25E4 3.11 E4 4.72E3 2.65E2 5.18E4 2.24E4 1.90E5 1.72E5 1.34E1 5.25E3 5.38E5 3.15E3 1.09E4 6.47E2 5.31 E3 1.28E4 1.00E6 8.69E4 1.02E6 7.77E5 4.5 1E6 6.47E5 2.03E6 1.12E7 1.75E6 4.79E5 1.35E5 7.97E4 7.13E4 1.60E6 1.68E5 5.17E5 9.84E6 3.22E5 3.67E6 6.47E2 5.3 1E3 3.57E2 7.06E3 1.09E3 2.48E4 1.11 E4 3.19E4 5.14E4 6.40E4 1.3 1E5 2.17E4 1.27E4 4.87E4 1.06E3 2.16E3 1.33E3 1.33E3 3.84E4 8.48E4 2.16E4 1.48E5 3.12E4 3.30E4'This and following 1K Tables Calculated in accordance with NUREG 0133, Section 5.3.1, except C 14 values in accordance with Regulatory Guide 1.109 Equation C-8.Unit 2 Revision 28 September 2006 1141 TABLE D 3-5 DOSE AND DOSE RATE Ri VALUES -INHALATION
-CHILD mrem/yr uCi/m 3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY 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 1131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 110m 3.59E4 4.74E4 2.07E4 4.26E4 5.99E5 1.01E8 1.90E5 2.3 5E4 4.81 E4 1.66E4 6.5 1E5 9.07E5 7.40E4 6.44E2 3.92E4 6.77E6 1.08E4 1.69E4 1.12E3 6.73E3 4.29E4 2.52E4 3.34E4 1.77E3 1.31E4 1.13E5 4.18E4 9.18E3 1.72E2 4.81 E4 2.03E4 1.0 1E6 8.25E5 6.48E1 2.25E2 1.95E4 2.12E6 8.73E3 1.14E4 1.12E3 6.73E3 1.54E2 9.5 1E3 7.77E3 1.67E4 3.16E3 2.26E4 7.03E4 1.72E4 6.44E6 3.70E4 6.55E3 4.26E1 2.73E4 7.70E3 2.25E5 1.28E5 4.33E3 7.55EI 2.90E3 3.61 E5 6.81 E2 9.14E3 1. 12E3 6.73E3 8.55E1 1.62E7 3.85E6 1.12E3 6.73E3 2.43E1 1.00E4 7.14E4 5.96E4 8.62E3 3.92E2 7.88E4 3.38E4 3.30E5 2.82E5 2.1 lEl 8.55E3 1.1 7E6 4.81 E3 2.12E4 1.12E3 6.73E3 1.70E4 1.58E6 1.1 IE5 1.27E6 1.11E6 7.07E6 9.95E5 2.16E6 1.48E7 2.23E6 6.14E5 1.35E5 1.21E5 1.04E5 1.74E6 1.83E5 5.44E5 1.20E7 3.28E5 5.48E6 1.12E3 6.73E3 1.08E3 2.29E4 2.87E3 7.07E4 3.44E4 9.62E4 1.63E4 1.67E5 3.43E5 6.11E4 3.70E4 1.27E5 2.84E3 5.48E3 3.85E3 3.62E3 1.02E5 2.26E5 5.66E4 3.89E5 8.2 1E4 1.00E5 Unit 2 Revision 28 September 2006 II 42 TABLE D 3-6 DOSE AND DOSE RATE Ri VALUES -INHALATION
-TEEN mrem/yr uCi/m 3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY 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 1 133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 1 10m 2.60E4 3.34E4 1.59E4 3.86E4 4.34E5 1.08E8 1.46E5 1.86E4 3.54E4 1.22E4 5.02E5 6.70E5 5.47E4 4.79E2 2.84E4 4.89E6 7.86E3 1.38E4 1.27E3 4.87E3 5.11 E4 2.38E4 3.70E4 2.07E3 1.51E4 1.34E5 4.58E4 1.03E4 1.69E2 4.91 E4 2.05E4 1.13E6 8.48E5 6.70E1 2.36E2 1.90E4 2.02E6 8.56E3 1.31 E4 1.27E3 4.87E3 1.35E2 8.40E3 5.54E3 1.43E4 2.78E3 1.98E4 6.24E4 1.25E4 6.68E6 3.15E4 5.66E3 3.22E1 2.64E4 6.22E3 5.49E5 3.11 E5 3.52E3 6.26E1 2.17E3 2.62E5 5.13E2 7.99E3 1.27E3 4.87E3 7.50E1 1.46E7 2.92E6 1.27E3 4.87E3 3.07EI 1.27E4 8.64E4 6.74E4 1.00E4 4.11E2 8.40E4 3.59E4 3.75E5 3.04E5 2.28E1 8.88E3 1.2 1E6 5.02E3 2.50E4 1.27E3 4.87E3 2.10E4 1.98E6 1.24E5 1.53E6 1.34E6 8.72E6 1.24E6 2.42E6 1.65E7 2.69E6 7.51E5 1.54E5 1.46E5 1.21E5 2.03E6 2.14E5 6.14E5 1.34E7 3.72E5 6.75E6 1.27E3 4.87E3 3.00E3 6.68E4 6.39E3 1.78E5 9.52E4 2.59E5 4.66E4 3.71 E5 7.65E5 1.49E5 9.68E4 2.69E5 6.49E3 1.03E4 9.76E3 8.48E3 2.29E5 4.87E5 1.26E5 8.64E5 1.82E5 2.73E5 Unit 2 Revision 28 September 2006 1143 TABLE D 3-7 DOSE AND DOSE RATE Ri VALUES -INHALATION
-ADULT m rem/yr uCi/m 3 T. BODY THYROID KID NUCLIDE BONE LIVER NEY 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 1 133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag I 10m 1.82E4 2.46E4 1.18E4 3.24E4 3.04E5 9.92E7 1.07E5 1.41E4 2.52E4 8.64E3 3.73E5 4.78E5 3.90E4 3.44E2 1.99E4 3.43E6 5.27E3 1.08E4 1.26E3 3.41 E3 3.96E4 1.70E4 2.78E4 1.58E3 1.15E4 1.03E5 3.44E4 7.82E3 1.21E2 3.58E4 1.48E4 8.48E5 6.2 1E5 4.90E1 1.74E2 1.35E4 1.43E6 6.10E3 1.00E4 1.26E3 3.4 1E3 1.00E2 6.30E3 3.94E3 1.06E4 2.07E3 1.48E4 4.66E4 8.72E3 6.1 0E6 2.33E4 4.21 E3 2.30E1 2.05E4 4.52E3 7.28E5 4.28E5 2.57E3 4.58E1 1.53E3 1.84E5 3.65E2 5.94E3 1.26E3 3.41 E3 5.95E1 1.19E7 2.15E6 1.26E3 3.41E3 2.28E1 9.84E3 6.90E4 5.42E4 7.74E3 2.91 E2 6.13E4 2.58E4 2.87E5 2.22E5 1.67E1 6.26E3 8.48E5 3.56E3 1.97E4 LUNG 1.26E3 3.41E3 1.44E4 1.40E6 7.2 1E4 1.02E6 9.28E5 5.97E6 8.64E5 1.40E6 9.60E6 1.77E6 5.05E5 9.12E4 9.76E4 7.52E4 1.27E6 1.36E5 3.62E5 7.78E6 2.21E5 4.63E6 GI-LLI 1.26E3 3.41 E3 3.32E3 7.74E4 6.03E3 1.88E5 1.06E5 2.85E5 5.34E4 3.5OE5 7.22E5 1.50E5 1.04E5 2.48E5 6.28E3 8.88E3 1.04E4 8.40E3 2.18E5 4.58E5 1.20E5 8.16E5 1.73E5 3.02E5 Unit 2 Revision 28 September 2006 II 44 TABLE D 3-8 DOSE AND DOSE RATE Ri VALUES -GROUND PLANE ALL AGE GROUPS M 2_mrem/yr uCilsec 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 1 133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 110m TOTAL BODY SKIN 4.65E6 1.40E9 2.73E8 3.80E8 2.15E10 7.46E8 2.16E4 2.45E8 1.36E8 3.99E6 1.72E7 2.39E6 6.83E9 1.03E10 2.05E7 1.92E7 1.37E7 6.96E7 8.46E6 3.44E9 5.50E6 1.64E9 3.20E8 4.45E8 2.53E10 8.57E8 2.5 1E4 2.85E8 1.61E8 4.63E6 2.09E7 2.91 E6 7.97E9 1.20El0 2.35E7 2.18E7 1.54E7 8.07E7 1.01E7 4.0 1E9 Unit 2 Revision 28 September 2006 1145 TABLE D 3-9 DOSE AND DOSE RATE Ri VALUES -COW MILK -INFANT m 2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 3 *-- 2.38E3 2.38E3 2.3 8E3 2.3 8E3 2.38E3 2.38E3 C 14" 3.23E6 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 Cr 51 ..-- 8.35E4 5.45E4 1.19E4 1.06E5 2.43E6 Mn 54 -- 2.51E7 5.68E6 -- 5.56E6 -- 9.21E6 Fe 55 8.43E7 5.44E7 1.45E7 .... 2.66E7 6.91 E6 Fe 59 1.22E8 2.13E8 8.38E7 .... 6.29E7 1.02E8 Co 58 -- 1.39E7 3.46E7 ...... 3.46E7 Co 60 -- 5.90E7 1.39E8 ...... 1.40E8 Zn 65 3.53E9 1.21E10 5.58E9 -- 5.87E9 -- 1.02E10 Sr 89 6.93E9 -- 1.99E8 ...... 1.42E8 Sr 90 8.19E 10 -- 2.09E 10 ...... 1.02E9 Zr 95 3.85E3 9.39E2 6.66E2 -- 1.01E3 -- 4.68E5 Nb 95 4.21 E5 1.64E5 1.17E5 -- 1.54E5 -- 3.03E8 Mo 99 -- 1.04E8 2.03E7 -- 1.55E8 -- 3.43E7 1131 6.81E8 8.02E8 3.53E8 2.64E11 9.37E8 -- 2.86E7 1133 8.52E6 1.24E7 3.63E6 2.26E9 1.46E7 -- 2.10E6 Cs 134 2.41E10 4.49E10 4.54E9 -- 1.16E10 4.74E9 1.22E8 Cs 137 3.47E10 4.06E10 2.88E9 -- 1.09E10 4.41 E9 1.27E8 Ba 140 1.21E8 1.21 E5 6.22E6 -- 2.87E4 7.42E4 2.97E7 La 140 2.03E1 7.99 2.06 ...... 9.39E4 Ce 141 2.28E4 1.39E4 1.64E3 -- 4.28E3 -- 7.18E6 Ce 144 1.49E6 6.1OE5 8.34E4 -- 2.46E5 -- 8.54E7 Nd 147 4.43E2 4.55E2 2.79E1 -- 1.76E2 -- 2.89E5 Ag 110m 2.46E8 1.79E8 1.19E8 -- 2.56E8 -- 9.29E9 mrnrem/yr per ýtCi/m 3.Unit 2 Revision 28 1146 September 2006 TABLE D 3-10 DOSE AND DOSE RATE Ri VALUES -COW MILK -CHILD m -mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY 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 1 133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 110m 1.65E6 6.97E7 6.52E7 2.63E9 3.64E9 7.53E10 2.17E3 1.86E5 3.26E8 4.04E6 1.50E10 2.17E10 5.87E7 9.70 1.15E4 1.04E6 2.24E2 1.33E8 1.57E3 3.29E5 1.35E7 3.07E7 1.06E8 6.94E6 2.89E7 7.00E9 4.77E2 1.03E4 4.07E7 3.28E8 4.99E6 2.45E10 2.08E10 5.14E4 3.39 5.73E3 3.26E5 1.81E2 8.97E7 1.57E3 3.29E5 5.27E4 3.59E6 1.15E7 5.26E7 2.13E7 8.52E7 4.35E9 1.04E8 1.91El0 4.25E2 5.69E4 1.01E7 1.86E8 1.89E6 5.18E9 3.07E9 3.43E6 1.14 8.51 E2 5.55E4 1.40El 7.17E7 1.57E3 3.29E5 2.93E4 1.08El I 9.27E8 1.57E3 3.29E5 7.99E3 3.78E6 4.41 E9 6.83E2 1.00E5 8.69E7 5.39E8 8.32E6 7.6 1E9 6.78E9 1.67E4 2.5 1E3 1.80E5 9.94E1 1.67E8 1.57E3 3.29E5 5.34E4 2.09E7 3.06E7 2.73E9 2.44E9 3.07E4 1.57E3 3.29E5 2.80E6 1.13E7 6.85E6 1.10E8 4.05E7 1.60E8 1.23E9 1.41E8 1.01E9 4.98E5 4.42E8 3.37E7 2.92E7 2.01E6 1.32E8 1.30E8 2.97E7 9.45E4 7.15E6 8.49E7 2.87E5 1.07E10*rnrem/yr per PtCi/m 3.Unit 2 Revision 28 September 2006 II 47 TABLE D 3-11 DOSE AND DOSE RATE Ri VALUES -COW MILK -TEEN m 2_mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 3 -- 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 C 14" 6.70E5 1.34E5 1.34E5 1.34E5 1.34E5 1.35E5 1.34E5 Cr 51 .... 2.58E4 1.44E4 5.66E3 3.69E4 4.34E6 Mn 54 -- 9.01E6 1.79E6 -- 2.69E6 -- 1.85E7 Fe 55 2.78E7 1.97E7 4.59E6 .... 1.25E7 8.52E6 Fe 59 2.81E7 6.57E7 2.54E7 .... 2.07E7 1.55E8 Co 58 -- 4.55E6 1.05E7 ...... 6.27E7 Co 60 -- 1.86E7 4.19E7 ...... 2.42E8 Zn 65 1.34E9 4.65E9 2.17E9 -- 2.97E9 -- 1.97E9 Sr 89 1.47E9 -- 4.21E7 ...... 1.75E8 Sr 90 4.45E10 -- 1.1E 1..--... 1.25E9 Zr 95 9.34E2 2.95E2 2.03E2 -- 4.33E2 -- 6.80E5 Nb 95 1.86E5 1.03E5 5.69E4 -- 1.00E5 -- 4.42E8 Mo 99 -- 2.24E7 4.27E6 -- 5.12E7 -- 4.01 E7 1131 1.34E8 1.88E8 1.01E8 5.49E10 3.24E8 -- 3.72E7 1133 1.66E6 2.82E6 8.59E5 3.93E8 4.94E6 -- 2.13E6 Cs 134 6.49E9 1.53E10 7.08E9 -- 4.85E9 1.85E9 1.90E8 Cs 137 9.02E9 1.20E10 4.18E9 -- 4.08E9 1.59E9 1.71E8 Ba 140 2.43E7 2.98E4 1.57E6 -- 1.01 E4 2.00E4 3.75E7 La 140 4.05 1.99 5.30E-1 ...--. 1.14E5 Ce 141 4.67E3 3.12E3 3.58E2 -- 1.47E3 -- 8.91E6 Ce 144 4.22E5 1.74E5 2.27E4 -- 1.04E5 -- 1.06E8 Nd 147 9.12E1 9.91E1 5.94E0 -- 5.82E1 -- 3.58E5 Ag lr0m 6.13E7 5.80E7 3.53E7 -- 1.11E8 -- 1.63E10 mrem/yr per ýiCi/m 3.Unit 2 Revision 28 II148 September 2006 TABLE D 3-12 DOSE AND DOSE RATE Ri VALUES -COW MILK -ADULT m 2-mrem/yr uCi/sec T. BODY THYROID KI NUCLIDE BONE LIVER DNEY LUNG GI-LLI H3**C 14 Cr51 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 Ag 110m 3.63E5 1.57E7 1.61E7 8.7 1E8 7.99E8 3.15E10 5.34E2 1.09E5 7.4 1E7 9.09E5 3.74E9 4.97E9 1.35E7 2.26 2.54E3 2.29E5 4.74E1 3.71E7 7.63E2 7.26E4 5.41E6 1.08E7 3.79E7 2.70E6 1.10E7 2.77E9 1.71E2 6.07E4 1.24E7 1.06E8 1.58E6 8.89E9 6.80E9 1.69E4 1.14 1.72E3 9.58E4 5.48E1 3.43E7 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-1 1.95E2 1.23E4 3.28E0 2.04E7 7.63E2 7.26E4 8.85E3 3.47E10 2.32E8 7.63E2 7.26E4 3.26E3 1.61E6 1.85E9 2.69E2 6.00E4 2.8 1E7 1.82E8 2.76E6 2.88E9 2.31 E9 5.75E3 7.99E2 5.68E4 3.20E1 6.74E7 7.63E2 7.26E4 1.96E4 6.04E6 1.06E7 9.55E8 7.68E8 9.69E3 7.63E2 7.26E4 3.72E6 1.66E7 6.21 E6 1.26E8 5.47E7 2.06E8 1.75E9 1.28E8 9.11E8 5.43E5 3.69E8 2.87E7 2.80E7 1.42E6 1.56E8 1.32E8 2.77E7 8.3 5E4 6.58E6 7.74E7 2.63E5 1.40E10*mrem/yr per tCi/m 3.Unit 2 Revision 28 September 2006 II 49 TABLE D 3-13 DOSE AND DOSE RATE Ri VALUES -GOAT MILK -INFANT m 2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY 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 Ag 110m 3.23E6 1.1 0E6 1.59E6 4.24E8 1.48E10 1.72E11 4.66E2 9.42E4 8.17E8 1.02E7 7.23E10 1.04El 1 1.45E7 2.430 2.74E3 1.79E5 5.32E1 2.95E7 6.3 3E3 6.89E5 3.01E6 7.08E5 2.78E6 1.67E6 7.08E6 1.45E9 1.13E2 3.88E4 1.27E7 9.63E8 1.49E7 1.35Ell 1.22E 1I 1.45E4 9.59E-1 1.67E3 7.32E4 5.47E1 2.15E7 6.33E3 6.89E5 1.00E4 6.82E5 1.89E5 1.09E6 4.16E6 1.67E7 6.70E8 4.24E8 4.38E110 8.04E1 2.24E4 2.47E6 4.23E8 4.36E6 1.36E10 8.63E9 7.48E5 2.47E-1 1.96E2 1.00E4 3.35E0 1.43E7 6.33E3 6.89E5 6.56E3 3.16E11 2.71 E9 6.33E3 6.89E5 1.43E3 6.67E5 7.04E8 1.22E2 2.78E4 1.89E7 1.12E9 1.75E7 3.47E10 3.27E10 3.44E3 5.14E2 2.96E4 2.11El 3.07E7 6.33E3 6.89E5 1.28E4 3.46E5 8.21 E5 1.42E 10 1.32E10 8.91 E3 6.33E3 6.89E5 2.93E5 1.11E6 8.98E4 1.33E6 4.16E6 1.68E7 1.23E9 3.04E8 2.15E9 5.65E4 3.27E7 4.17E6 3.44E7 2.52E6 3.66E8 3.81 E8 3.56E6 1.13E4 8.62E5 1.03E7 3.46E4 1.1 1E9 mrem/yr per pCi/m 3.Unit 2 Revision 28 September 2006 II 50 TABLE D 3-14 DOSE AND DOSE RATE Ri VALUES -GOAT MILK -CHILD m 2-mrem/yr uCi/sec T. BODY THYROID KI]NUCLIDE BONE LIVER D)NEY 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 1131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 110m 1.65E6 9.06E5 8.52E5 3.15E8 7.77E9 1.58E11 2.62E2 5.05E4 3.91E8 4.84E6 4.49E10 6.52E10 7.05E6 1.16 1.38E3 1.25E5 2.68E1 1.60E7 4.17E3 3.29E5 1.62E6 4.8 1E5 1.38E6 8.35E5 3.47E6 8.40E8 5.76E 1 1.96E4 4.95E6 3.94E8 5.99E6 7.37E10 6.24E10 6.18E3 4.07E-1 6.88E2 3.91 E4 2.17E1 1.08E7 4.17E3 3.29E5 6.34E3 4.31 E5 1.49E5 6.86E5 2.56E6 1.02E7 5.23E8 2.22E8 4.01El0 5.13E1 1.40E4 1.22E6 2.24E8 2.27E6 1.55E10 9.21 E9 4.12E5 1.37E-1 1.02E2 6.66E3 1.68E0 8.60E6 4.17E3 3.29E5 3.52E3 1.30E 1I 1.11E9 4.17E3 3.29E5 9.62E2 4.54E5 5.29E8 8.25E1 1.85E4 1.06E7 6.46E8 9.98E6 2.28E10 2.03E10 2.01E3 3.02E2 2.16E4 1.19E 1 2.00E7 LUNG 4.17E3 3.29E5 6.43E3 2.72E5 3.99E5 8.19E9 7.32E9 3.68E3 GI-LLI 4.17E3 3.29E5 3.36E5 1.36E6 8.91 E4 1.43E6 4.87E6 1.92E7 1.48E8 3.01E8 2.13E9 6.01E4 3.63E7 4.09E6 3.50E7 2.41 E6 3.97E8 3.91E8 3.57E6 1.13E4 8.59E5 1.02E7 3.44E4 1.28E9 mrem/yr per ý.Ci/m 3.Unit 2 Revision 28 September 2006 1151 TABLE D 3-15 DOSE AND DOSE RATE Ri VALUES -GOAT MILK -TEEN m 2-mrem/yr uCi/sec 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 1131 1133 Cs 134 6.70E5 3.61 E5 3.67E5 1.61E8 3.14E9 9.36E10 1.13E2 2.23E4 1.61E8 1.99E6 1.95E10 2.71E10 2.92E6 4.86E-1 5.60E2 5.06E4 1.09El 7.36E6 2.64E3 1.34E5 1.08E6 2.56E5 8.57E5 5.46E5 2.23E6 5.58E8 3.56E1 1.24E4 2.72E6 2.26E8 3.38E6 4.58E10 3.60E10 3.58E3 2.39E-1 3.74E2 2.09E4 1.19E 1 6.96E6 2.64E3 1.34E5 3.11 E3 2.15E5 5.97E4 3.31 E5 1.26E6 5.03E6 2.60E8 8.99E7 2.3 1E10 2.45E1 6.82E3 5.19E5 1.2 1E8 1.03E6 2.13E10 1.25E10 1.88E5 6.36E-2 4.30E1 2.72E3 7.13E-1 4.24E6 2.64E3 1.34E5 1.73E3 6.59E10 4.72E8 2.64E3 1.34E5 6.82E2 3.23E5 3.57E8 5.23E1 1.20E4 6.23E6 3.89E8 5.93E6 1.46E10 1.23E10 1.21E3 1.76E2 1.25E4 6.99E0 1.33E7 2.64E3 1.35E5 4.44E3 1.62E5 2.70E5 5.56E9 4.76E9 2.4 1E3 2.64E3 1.34E5 5.23E5 2.22E6 1.11 E5 2.03E6 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 1.07E6 1.27E7 4.29E4 1.96E9 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 110m.mrem/yr per PtCi/m 3.Unit 2 Revision 28 September 2006 II 52 TABLE D 3-16 DOSE AND DOSE RATE Ri VALUES -GOAT MILK -ADULT m 2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 3 -- 2.03E3 2.03E3 2.03E3 2.03E3 2.03E3 2.03E3 C 14* 3.63E5 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 Cr 51 .... 1.78E3 1.06E3 3.92E2 2.36E3 4.48E5 Mn 54 -- 6.50E5 1.24E5 -- 1.93E5 -- 1.99E6 Fe 55 2.04E5 1.41 E5 3.28E4 .... 7.85E4 8.07E4 Fe 59 2.10E5 4.95E5 1.90E5 .... 1.38E5 1.65E6 Co 58 -- 3.25E5 7.27E5 ...... 6.58E6 Co60 -- 1.32E6 2.91 E6 ...... 2.48E7 Zn 65 1.05E8 3.33E8 1.51E8 -- 2.23E8 -- 2.10E8 Sr 89 1.70E9 -- 4.89E7 ...... 2.73E8 Sr 90 6.62E10 -- 1.63E10 ...... 1.91E9 Zr 95 6.45E1 2.07E1 1.40El -- 3.25E1 -- 6.56E4 Nb 95 1.31E4 7.29E3 3.92E3 -- 7.21 E3 -- 4.42E7 Mo 99 -- 1.51E6 2.87E5 -- 3.41 E6 -- 3.49E6 1131 8.89E7 1.27E8 7.29E7 4.17E10 2.18E8 -- 3.36E7 1133 1.09E6 1.90E6 5.79E5 2.79E8 3.31 E6 -- 1.71E6 Cs 134 1.12E10 2.67E10 2.18E10 -- 8.63E9 2.86E9 4.67E8 Cs 137 1.49E10 2.04E10 1.34E10 -- 6.93E9 2.30E9 3.95E8 Ba 140 1.62E6 2.03E3 1.06E5 -- 6.91 E2 1.16E3 3.33E6 La 140 2.71E-1 1.36E-1 3.61E-2 .-- -- 1.00E4 Ce 141 3.06E2 2.07E2 2.34E1 -- 9.60E1 -- 7.90E5 Ce 144 2.75E4 1.15E4 1.48E3 -- 6.82E3 -- 9.30E6 Nd 147 5.69E0 6.57E0 3.93E-1 -- 3.84E0 -- 3.15E4 Ag 110m 4.45E6 4.12E6 2.45E6 -- 8.09E6 -- 1.68E9 mrem/yr per C i/m 3.Unit 2 Revision 28 II153 September 2006 TABLE D 3-17 DOSE AND DOSE RATE Ri VALUES -COW MEAT -CHILD m 2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY 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 1 133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 110m 5.29E5 2.89E8 2.04E8 2.38E8 2.65E8 7.01E9 1.51E6 4.10E6 4.15E6 9.38E-2 6.09E8 8.99E8 2.20E7 2.80E-2 1.17E4 1.48E6 5.93E3 5.62E6 2.34E2 1.06E5 5.15E6 1.53E8 3.30E8 9.41 E6 4.64E7 6.35E8 3.32E5 1.59E6 5.42E4 4.18E6 1.16E-1 1.00E9 8.60E8 1.93E4 9.78E-3 5.82E3 4.65E5 4.80E3 3.79E6 2.34E2 1.06E5 4.55E3 1.37E6 4.74E7 1.65E8 2.88E7 1.37E8 3.95E8 7.57E6 1.78E9 2.95E5 1.14E6 1.34E4 2.37E6 4.39E-2 2.11E8 1.27E8 1.28E6 3.30E-3 8.64E2 7.91 E4 3.72E2 3.03E6 2.34E2 1.06E5 2.52E3 1.38E9 2.15E1 2.34E2 1.06E5 6.90E2 1.44E6 4.00E8 4.75E5 1.50E6 1.16E5 6.86E6 1.93E-1 3.1 0E8 2.80E8 6.27E3 2.55E3 2.57E5 2.64E3 7.05E6 2.34E2 1.06E5 4.61 E3 8.66E7 9.58E7 1.11E8 1.01E8 1.15E4 2.34E2 1.06E5 2.41 E5 4.32E6 2.84E7 3.44E8 5.49E7 2.57E8 1.12E8 1.03E7 9.44E7 3.46E8 2.95E9 4.48E4 3.72E5 4.67E-2 5.39E6 5.39E6 1.11 E7 2.73E2 7.26E6 1.21E8 7.61 E6 4.52E8*mrem/yr per jtCi/m 3.Unit 2 Revision 28 September 2006 1154 TABLE D 3-18 DOSE AND DOSE RATE Ri VALUES -COW MEAT -TEEN m 2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY 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 1131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 1 10m 2.8 1E5 1.50E8 1.15E8 1.59E8 1.40E8 5.42E9 8.50E5 2.37E6 2.24E6 5.05E-2 3.46E8 4.88E8 1.19E7 1.53E-2 6.19E3 7.87E5 3.16E3 3.39E6 1.94E2 5.62E4 4.50E6 1.07E8 2.69E8 8.05E6 3.90E7 5.52E8 2.68E5 1.32E6 3.90E4 3.13E6 8.57E-2 8.13E8 6.49E8 1.46E4 7.5 1E-3 4.14E3 3.26E5 3.44E3 3.20E6 1.94E2 5.62E4 2.93E3 8.93E5 2.49E7 1.04E8 1.86E7 8.80E7 2.57E8 4.01E6 1.34E9 1.84E5 7.24E5 7.43E3 1.68E6 2.6 1E-2 3.77E8 2.26E8 7.68E5 2.OOE-3 4.75E2 4.23E4 2.06E2 1.95E7 1.94E2 5.62E4 1.62E3 9.15E8 1.20El 1.94E2 5.62E4 6.39E2 1.34E6 3.53E8 3.94E5 1.28E6 8.92E4 5.40E6 1.50E-1 2.58E8 2.21E8 4.95E3 1.95E3 1.94E5 2.02E3 6.13E6 1.94E2 5.62E4 4.16E3 6.77E7 8.47E7 9.87E7 8.58E7 9.81E3 1.94E2 5.62E4 4.90E5 9.24E6 4.62E7 6.36E8 1.11E8 5.09E8 2.34E8 1.67E7 1.52E8 6.19E8 5.63E9 6.98E4 6.20E5 6.48E-2 1.01E7 9.24E6 1.84E7 4.31 E2 1.1 8E7 1.98E8 1.24E7 9.01E8.mrem/yr per ý,Ci/m 3.Unit 2 Revision 28 September 2006 1155 TABLE D 3-19 DOSE AND DOSE RATE Ri VALUES -COW MEAT -ADULT m 2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 3 -- 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 C 14* 3.33E5 6.66E4 6.66E4 6.66E4 6.66E4 6.66E4 6.66E4 Cr 51 .... 3.65E3 2.18E3 8.03E2 4.84E3 9.17E5 Mn 54 -- 5.90E6 1.13E6 -- 1.76E6 -- 1.81E7 Fe 55 1.85E8 1.28E8 2.98E7 .... 7.14E7 7.34E7 Fe 59 1.44E8 3.39E8 1.30E8 .... 9.46E7 1.13E9 Co 58 -- 1.04E7 2.34E7 ...... 2.12E8 Co60 -- 5.03E7 1.11 E8 ...... 9.45E8 Zn 65 2.26E8 7.19E8 3.25E8 -- 4.81E8 -- 4.53E8 Sr 89 1.66E8 -- 4.76E6 ...... 2.66E7 Sr 90 8.38E9 -- 2.06E9 ...... 2.42E8 Zr 95 1.06E6 3.40E5 2.30E5 -- 5.34E5 -- 1.08E9 Nb 95 3.04E6 1.69E6 9.08E5 -- 1.67E6 -- 1.03E10 Mo 99 -- 4.71 E4 8.97E3 -- 1.07E5 -- 1.09E5 1131 2.69E6 3.85E6 2.21 E6 1.26E9 6.61 E6 -- 1.02E6 1133 6.04E-2 1.05E-1 3.20E-2 1.54E1 1.83E-1 -- 9.44E-2 Cs 134 4.35E8 1.03E9 8.45E8 -- 3.35E8 1.11E8 1.81E7 Cs 137 5.88E8 8.04E8 5.26E8 -- 2.73E8 9.07E7 1.56E7 Ba 140 1.44E7 1.81E4 9.44E5 -- 6.15E3 1.04E4 2.97E7 La 140 1.86E-2 9.37E-3 2.48E-3 ...... 6.88E2 Ce 141 7.38E3 4.99E3 5.66E2 -- 2.32E3 -- 1.91 E7 Ce 144 9.33E5 3.90E5 5.01E4 -- 2.31E5 -- 3.16E8 Nd 147 3.59E3 4.15E3 2.48E2 -- 2.42E3 -- 1.99E7 Ag 110m 4.48E6 4.14E6 2.46E6 -- 8.13E6 -- 1.69E9 mrem/yr per ptCi/m 3.Unit 2 Revision 28 1156 September 2006 TABLE D 3-20 DOSE AND DOSE RATE Ri VALUES -VEGETATION
-CHILD m -mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 3* -- 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 C 14* 3.50E6 7.01E5 7.01E5 7.01E5 7.01E5 7.01E5 7.01E5 Cr 51 .... 1.17E5 6.49E4 1.77E4 1.18E5 6.20E6 Mn 54 -- 6.65E8 1.77E8 -- 1.86E8 -- 5.58E8 Fe 55 7.63E8 4.05E8 1.25E8 .... 2.29E8 7.50E7 Fe 59 3.97E8 6.42E8 3.20E8 .... 1.86E8 6.69E8 Co 58 -- 6.45E7 1.97E8 ...... 3.76E8 Co 60 -- 3.78E8 1.12E9 ...... 2.10E9 Zn 65 8.12E8 2.16E9 1.35E9 -- 1.36E9 -- 3.80E8 Sr 89 3.59E10 -- 1.03E9 ...... 1.39E9 Sr 90 1.24E12 -- 3.15E11 ...... 1.67E10 Zr 95 3.86E6 8.50E5 7.56E5 -- 1.22E6 -- 8.86E8 Nb 95 1.02E6 3.99E5 2.85E5 -- 3.75E5 -- 7.37E8 Mo 99 -- 7.70E6 1.91E6 -- 1.65E7 -- 6.37E6 1 131 7.16E7 7.20E7 4.09E7 2.38E10 1.18E8 -- 6.41E6 1 133 1.69E6 2.09E6 7.92E5 3.89E8 3.49E6 -- 8.44E5 Cs 134 1.60E10 2.63E10 5.55E9 -- 8.15E9 2.93E9 1.42E8 Cs 137 2.39E10 2.29E10 3.38E9 -- 7.46E9 2.68E9 1.43E8 Ba 140 2.77E8 2.43E5 1.62E7 -- 7.90E4 1.45E5 1.40E8 La 140 3.25E3 1.13E3 3.83E2 ...... 3.16E7 Ce 141 6.56E5 3.27E5 4.85E4 -- 1.43E5 -- 4.08E8 Ce 144 1.27E8 3.98E7 6.78E6 -- 2.21E7 -- 1.04E10 Nd 147 7.23E4 5.86E4 4.54E3 -- 3.22E4 -- 9.28E7 Ag 110m 3.21E7 2.17E7 1.73E7 -- 4.04E7 -- 2.58E9 mrem/yr per gCi/m 3.Unit 2 Revision 28 1157 September 2006 TABLE D 3-21 DOSE AND DOSE RATE Ri VALUES -VEGETATION
-TEEN M 2_mrem/vr uCi/sec 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 1131 1133 Cs 134 Cs 137 Ba 140 La 140 Ce 141 Ce 144 Nd 147 Ag 110m BONE 1.45E6 3.1 0E8 1.79E8 4.24E8 1.51El0 7.51El 1 1.72E6 4.80E5 3.85E7 9.29E5 7.10E9 1.01El0 1.38E8 1.81E3 2.83E5 5.27E7 3.66E4 1.51E7 LIVER 2.59E3 2.91 E5 4.54E8 2.20E8 4.18E8 4.37E7 2.49E8 1.47E9 5*44E5 2.66E5 5.64E6 5.39E7 1.58E6 1.67E10 1.35E10 1.69E5 8.88E2 1.89E5 2.18E7 3.98E4 1.43E7 T. BODY 2.59E3 2.91 E5 6.16E4 9.01E7 5.13E7 1.61E8 1.01E8 5.60E8 6.86E8 4.33E8 1.85E1 1 3.74E5 1.46E5 1.08E6 2.89E7 4.80E5 7.75E9 4.69E9 8.91 E6 2.36E2 2.17E4 2.83E6 2.3863 8.72E6 THYROID KIDNEY LUNG 2.59E3 2.91E5 3.42E4 1.57E10 2.20E8 2.59E3 2.91 E5 1.35E4 1.36E8 9.41E8 7.99E5 2.58E5 1.29E7 9.28E7 2.76E6 5.3 1E9 4.59E9 5.74E4 8.89E4 1.30E7 2.34E4 2.74E7 2.59E3 2.91E5 8.79E4 1.40E8 1.32E8 2.03E9 1.78E9 1. 14E5 GI-LLI 2.59E3 2.91 E5 1.03E7 9.32E8 9.53E7 9.89E8 6.02E8 3.24E9 6.23E8 1.80E9 2.11 E10 1.26E9 1.1 4E9 1.01E7 1.07E7 1.19E6 2.08E8 1.92E8 2.13E8 5.1 0E7 5.40E8 1.33E10 1.44E8 4.03E9.mrem/yr per gCi/m 3 Unit 2 Revision 28 September 2006 1158 TABLE D 3-22 DOSE AND DOSE RATE Ri VALUES -VEGETATION
-ADULT m 2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 3* -- 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 C 14" 8.97E5 1.79E5 1.79E5 1.79E5 1.79E5 1.79E5 1.79E5 Cr 51 .... 4.64E4 2.77E4 1.02E4 6.15E4 1.17E7 Mn 54 -- 3.13E8 5.97E7 -- 9.31E7 -- 9.58E8 Fe 55 2.00E8 1.38E8 3.22E7 ..-- 7.69E7 7.91E7 Fe 59 1.26E8 2.96E8 1.13E8 .... 8.27E7 1.02E9 Co 58 -- 3.08E7 6.90E7 ...... 6.24E8 Co 60 -- 1.67E8 3.69E8 ...... 3.14E9 Zn 65 3.17E8 1.01E9 4.56E8 -- 6.75E8 -- 6.36E8 Sr 89 9.96E9 -- 2.86E8 ...... 1.60E9 Sr 90 6.05E11 -- 1.48E11 .. ... 1.75E10 Zr 95 1.18E6 3.77E5 2.55E5 -- 5.92E5 -- 1.20E9 Nb 95 3.55E5 1.98E5 1.06E5 -- 1.95E5 -- 1.20E9 Mo 99 -- 6.14E6 1.17E6 -- 1.39E7 -- 1.42E7 1131 4.04E7 5.78E7 3.31E7 1.90E10 9.91E7 -- 1.53E7 1133 1.00E6 1.74E6 5.30E5 2.56E8 3.03E6 -- 1.56E6 Cs 134 4.67E9 1.1IEl10 9.08E9 -- 3.59E9 1.19E9 1.94E8 Cs 137 6.36E9 8.70E9 5.70E9 -- 2.95E9 9.81E8 1.68E8 Ba 140 1.29E8 1.61E5 8.42E6 -- 5.49E4 9.25E4 2.65E8 La 140 1.98E3 9.97E2 2.63E2 ...... 7.32E7 Ce 141 1.97E5 1.33E5 1.51E4 -- 6.19E4 -- 5.09E8 Ce 144 3.29E7 1.38E7 1.77E6 -- 8.16E6 -- MO1El1 Nd 147 3.36E4 3.88E4 2.32E3 -- 2.27E4 -- 1.86E8 Ag 110m 1.05E7 9.75E6 5.79E6 -- 1.92E7 -- 3.98E9*mrem/yr per ýiCi/m 3 Unit 2 Revision 28 II159 September 2006 TABLE D 3-23 DISPERSION PARAMETERS AT CONTROLLING LOCATIONS' X/Q,W, and W, VALUES VENT Site Boundary 2 Inhalation and Ground Plane Cow Milk Goat Milk 3 Meat Animal Vegetation STACK Site Boundary2 Inhalation and Ground Plane Cow Milk Goat Milk 3 Meat Animal Vegetation NOTE: DIRECTION DISTANCE (m)E E (104)ESE (1300)SE (1400)E(114)E (96)1,600 1,800 4,300 4,800 2,600 2,900 X/O (sec/m 3)2.00 E-6 1.42E-7 4.1 IE-8 3.56E-08 1.17E-7 1.04E-7 DIO Wm)2. 1OE-9 2.90E-9 4.73E-10 5.32E-10 1.86E-9 1 .50E-9 E 1,600 1,700 4.50E-8 8,48E-9 6.OOE-9 1 .34E-9 E (109°)ESE (1350) 4,200 SE (140°) 4,800 E (114) 2,500 E (96) 2,800 Inhalation and Ground Plane are annual average values.1.05E-8 3.64E-10 2.90E-08 5.71E-10 1.13E-8 1.15E-9 1.38E-8 9.42E-10 Others are grazing season only.1 X/Q and D/Q values from NMP-2 ER-OLS.2 X/Q and D/Q from NMP-2 FES, NUREG-1085, May 1985, Table D-2.3 X/Q and D/Q from C.T. Main Data Report dated November 1985.Unit 2 Revision 28 September 2006 II 60 TABLE D 3-24 PARAMETERS FOR THE EVALUATION OF DOSES TO REAL MEMBERS OF THE PUBLIC FROM GASEOUS AND LIQUID EFFLUENTS Pathway Fish Parameter U (kg/yr) -adult Value 21 Reference Reg. Guide 1.109 Table E-5 Reg. Guide 1.109 Table E-1 1 Fish Daipj (mrem/pCi)
Each Radionuclide Shoreline U (hr/yr)-adult-teen 67 67 Reg. Guide 1.109 Assumed to be Same as Adult Shoreline Daipj (mrem/hr per pCi/m 2)Each Radionuclide Reg. Guide 1.109 Table E-6 Reg. Guide 1.109 Table E-7 Inhalation DFAija Each Radionuclide Unit 2 Revision 28 September 2006 1161 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)Direct Radiation (TLD)* Map = See Figure TABLE D 5.1 NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS* Map Collection Site Location (Env. Pro2ram No.)I Nine Mile Point Road North (R-l)2 County Route 29 & Lake Road (R-2)3 County Route 29 (R-3)4 Village of Lycoming, NY (R-4)5 Montario Point Road (R-5)6 North Shoreline Area (75)7 North Shoreline Area (76)8 North Shoreline Area (77)9 North Shoreline Area (23)10 JAF East Boundary (78)11 Route 29 (79)12 Route 29 (80)13 Miner Road (81)14 Miner Road (82)15 Lakeview Road (83)16 Lakeview Road (84)17 Site Meteorological Tower (7)18 Energy Information Center (18)19 North Shoreline (85)s D 5.1-1 and D 5.1-2.Location 1.8 mi @ 88°E 1.1 mi@ 1040 ESE 1.5 mi @ 132' SE 1.8 mi @ 143° SE 16.4 mi @ 42* NE 0.1 mi@ 5°N 0.1 mi @ 25* NNE 0.2 mi @ 45° NE 0.8 mi @ 70* ENE 1.0 mi @ 90 E 1.1 mi @ 115* SE 1.4 mi @ 133' SE 1.6 mi @ 159 SSE 1.6 mi @ 181 S 1.2 mi @ 200° SSW 1.1 mi @ 225' SW 0.7 mi @ 250 WSW 0.4 mi @ 265 W 0.2 mi @ 294' WNW Unit 2 Revision 28 September 2006 1162 Type of Sample 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)Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)Direct Radiation (TLD)TABLE D 5.1 (Cont'd)NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS* Map Collection Site Location (Env. Program No.)20 North Shoreline (86)21 North Shoreline (87)22 Hickory Grove (88)23 Leavitt Road (89)24 Route 104 (90)25 Route 51A (91)26 Maiden Lane Road (92)27 County Route 53 (93)28 County Route 1 (94)29 Lake Shoreline (95)30 Phoenix, NY Control (49)31 S. W. Oswego, Control (14)32 Scriba, NY (96)33 Alcan Aluminum, Route IA (58)34 Lycoming, NY (97)35 New Haven, NY (56)36 W. Boundary, Bible Camp (15)37 Lake Road (98)Location 0.1 mi @ 315 NW 0.1 mi @ 341 NNW 4.5 mi @ 97o E 4.1 mi@ 111 ESE 4.2 mi @ 135" SE 4.8 mi @ 156" SSE 4.4 mi @ 183 S 4.4 mi @ 205* SSW 4.7 mi @ 2230 SW 4.1 mi @ 237 WSW 19.8 mi @ 163" S 12.6 mi @ 226* SW 3.6 mi @ 199* SSW 3.1 mi @ 220* SW 1.8 mi @ 143' SE 5.3 mi @ 123' ESE 0.9 mi @ 237* WSW 1.2 mi @ 101 E* Map= See Figures D 5.1-1 and D 5.1-2.Unit 2 Revision 28 September 2006 II 63 Type of Sample Surface Water Surface Water Shoreline Sediment Fish Fish Fish Milk Milk Milk Milk (CR)Food Product Food Product Food Product Food Product Food Product TABLE D 5.1 (Cont'd)NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS* Map Collection Site Location (Env. Pro2ram No.)38 OSS Inlet Canal (NA)39 JAFNPP Inlet Canal (NA)40 Sunset Bay Shoreline (NA)41 NMP Site Discharge Area (NA)42 NMP Site Discharge Area (NA)43 Oswego Harbor Area (NA)76 Milk Location #76 64 Milk Location #55 66 Milk Location #4 77 Milk Location (Summerville) 48 Produce Location #6**(Bergenstock) (NA)49 Produce Location #I**(Culeton) (NA)50 Produce Location #2**(Vitullo) (NA)51 Produce Location #5**(C.S. Parkhurst) (NA)52 Produce Location #3**(C. Narewski) (NA)Location 7.6 mi @ 235* SW 0.5 mi @ 70' ENE 1.5mi @ 80°E 0.3 mi @ 315 NW (and/or)0.6 mi @ 55* NE 6.2 mi @ 235' SW 6.3 mi @ 120* ESE 9.0 mi @ 95° E 7.8 mi @ 113* ESE 13.9 mi @ 191 SSW 1.9mi @ 141 SE 1.7 mi @ 96* E 1.9 mi @ 101* E 1.5 mi @ 114'ESE 1.6 mi @ 84* E* Map (NA)CR See Figures D 5.1-1 and D 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 D/Q.Not applicable.
Control Result (location).
Unit 2 Revision 28 II 64 September 2006 TABLE D 5.1 (Cont'd)NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS* Map Type of Sample Location Food Product 53 Food Product (CR)Food Product (CR)Food Product Food Product Food Product Food Product Food Product (CR)Food Product Food Product Food Product Food Product Collection Site (Env. Program No.)Produce Location #4**(P. Parkhurst) (NA)Produce Location #7**(Mc Millen) (NA)Produce Location #8**(Denman) (NA)Produce Location #9*(O'Connor) (NA)Produce Location #10"*(C. Lawton) (NA)Produce Location # 11 *(C. R. Parkhurst) (NA)Produce Location #12**(Barton) (NA)Produce Location #13**(Flack) (NA)Produce Location #14**(Koeneke) (NA)Produce Location # 15*(Whaley) (NA)Produce Location #16**(Murray) (NA)Produce Location #17**(Battles) (NA)Product Location # 18*(Kronenbitter) 2.1 mi@ 110° ESE 15.0 mi @ 2230 SW 12.6 mi @ 2250 SW 1.6 mi@ 171 S 2.2 mi @ 1230 ESE 2.0 mi @ 1120 ESE 1.9 mi @ 1150 ESE 15.6 mi @ 2250 W 1.9 mi @ 95' E 1.7 mi @ 1360 SE 1.2 mi @ 2070 SSW 1.76 mi @ 970 E 1.52 mi @ 850E Location Food Product*Map = SeeFigures D 5.1-1 andD 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 D/Q.(NA) = Not applicable.
CR = Control Result (location).
Unit 2 Revision 28 1165 September 2006 APPENDIX A LIQUID DOSE FACTOR DERIVATION Unit 2 Revision 28 September 2006 II 66 Appendix A Liquid Effluent Dose Factor Derivation, Aiat Aiat (mrem/hr per uCi/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 E- 11 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 Aiat of each nuclide i age group a, and organ t, hence Al at. It should be noted that the fish ingestion pathway is the most significant pathway for dose from liquid effluents.
The water consumption pathway is included for consistency with NUREG 0133.The equation for calculating dose contributions given in section 1.3 requires the use of the composite dose factor Air for each nuclide, i. The dose factor equation for a fresh water site is: U + -Uf BFi e -+ U, We-""(I -e-: 'b )DFS1 Aiat K 0 U .I DF, 0 + 9.Where: Ajat Is the dose factor for nuclide i, age group a, total body or organ t, for all appropriate pathways, (mrem/hr per uCi/ml)Ko Is the unit conversion factor, 1. 14E5=1E6pCi/uCi x 1E3 ml/liter 8760 hr/yr Uw Water consumption (liters/yr);
from Table E-5 of Reg. Guide 1.109 Uf = Fish consumption (kg/yr); from Table E-5 of Reg. Guide 1.109 us = Sediment Shoreline Usage (hr/yr); from Table E-5 of Reg. Guide 1.109 BFi Bioaccumulation factor for nuclide, i, in fish, (pCi/kg per pCi/liter), from Table A- I of Reg. Guide 1.109 DFLiat Dose conversion factor for age, nuclide, i, group a, total body or organ t, (mrem/pCi);
from Table E-11 of Reg. Guide 1.109 DFSi = Dose conversion factor for nuclide i and total body, from standing on contaminated ground (mrem/hr per pCi/mi); from Table E-6 of Reg.Guide 1.109 Unit 2 Revision 28 II 67 September 2006 Appendix A (Cont'd)Dw Dilution factor from the near field 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 of the City of Oswego. (Unitless)
DS Dilution factor from the near field area within one quarter mile of the release point to the shoreline deposit (taken at the same point where we take environmental samples 1.5 miles; unitless)69.3 conversion factor .693 x 100, 100 = K, (liters/kg-hr)*40 kg/rn 2*24 hr/day/.693 in liters/m 2-d, and K, = transfer coefficient from water to sediment in liters/kg per hour.tpw, tpf, Average transit time required for each nuclide to reach the tPS point of exposure for internal dose, it is the total time elapsed from release of the nuclides to 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 of facility operating life), (hrs).decay constant for nuclide i (hf1)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/mi 2 (DFL)iat = 1.95E-3 mrem/pCi tPW = 40 hrs. (w = water)BFj = 15 pCi/kg per pCi/liter tpf = 24 hrs. (f = fish)Uf = 21 kg/yr tb = 1.314E5 hr (5.48E3 days)Dw = 62 unitless Uw = 730 liters/yr DS = 17.8 unitless Ko = 1.14E5 (pCi/uCi)(ml/kg) us = 12 hr/yr (hr/yr)W = 0.3 ki = 3.61E-3hr' tps = 7.3 hrs (s=Shoreline Sediment)These values will yield an Aiat Factor of 6.65E4 mrem-ml per uCi-hr as listed in Table D 2-2. It should be noted that only a limited number of nuclides are listed on Tables D 2-2 to D 2-5. These are the most common nuclides encountered in effluents.
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.In addition, not all dose factors are used for the dose calculations.
A maximum individual is used, which is a composite of the maximum dose factor of each age group for each organ as reflected in the applicable chemistry procedures.
Unit 2 Revision 28 II 68 September 2006 APPENDIX B PLUME SHINE DOSE FACTOR DERIVATION Unit 2 Revision 28 September 2006 1169 Appendix B For elevated releases the plume shine dose factors for gamma air (Bi) and whole body (Vi), are calculated using the 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 gamma 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 follows: Gamma Air Bi = Y I. Where: K' conversion factor (see S RE) v, below for actual value).Ila mass absorption coefficient (cm 2/g; air for Bi, tissue for Vi)E Energy of gamma ray per disintegration (Mev)Vs average wind speed for each stability class (s), r/s R = downwind distance (site boundary, m)E = sector width (radians)s = subscript for stability class is I function = I + k1 2 for each stability class. (unitless, see Regulatory Guide 1.109)k2 Fraction of the attenuated energy that is actually absorbed in air (see Regulatory Guide 1.109, see below for equation)Whole Body-ýLatd Vi = 1.1lSFBie Where: td tissue depth (g/cm 2)SF -shielding factor from structures (unitless) 1.11 = Ratio of mass absorption coefficients between tissue and air.Where all other parameters are defined above.Unit 2 Revision 28 II 70 September 2006 Appendix B (Cont'd)1K = conversion factor 2k = ýL -a I-La 3.7ElOdis 1.6E-6erz Ci-sec Mev =1293 g 100 erg Im3 g-rad.46 Where: t = mass attenuation coefficient (cm 2/g; air for Bi, tissue for Vi)I-La-defined above There are seven stability classes, A thru F. The percentage of the year that each stability class 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 Unit 2 FSAR. 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
([ia) and attenuation (i) coefficients were calculated by multiplying the mass absorption (FLa/P) and mass attenuation (i'/p) coefficients given in the Radiation Health Handbook by the air density equal to 1.293 E-3 g/cc or the tissue density of 1 g/cc where applicable.
The tissue depth is 5g/cm 2 for the whole body.The downwind distance is the site boundary.SAMPLE CALCULATION Ex. Kr-89-DATA E 4a =Rz F STABILITY CLASS ONLY- Gamma Air 2.22MeV k = FLLa .871 K = .46 2.943 E-3m-' 4a VF = 5.55 m/sec 5.5064E-3m-'
R = 1600m.39 19m vertical plume spread taken from "Introduction to Nuclear Engineering", John R. LaMarsh Unit 2 Revision 28 September 2006 1171 Appendix B (Cont'd)-I Function UOz I1 12 I-- .11= .3= .4 D I1 + k1 2= .3 + (.871) (.4) = .65 I dis. 1 0.46 [Ci-sec) (Mev/ergsi (2.943E-3m-')
(2.22Mev)
(.65)(iW (g/m 3) (ergs) (5.55 m/s) (.39) (1600m)(g-rad)= 3.18(-7) rad/s (3600 s/hr) (24 h/d) (365 d/y) (iE3mrad/rad)
Ci/s (lE6uCi)Ci 1.00(-2) mrad/yr uCi/sec Vi 1.11 (.7) L(E-2)mrad/yr]
[e[iCi/sec-(.0253 cm 2/g) (5g/cm 2)I 6.85(-3) mrad/yr pCi/sec Note: The above calculation is for the F stability class only. For Table D 3-2 and procedure values, a weighted fraction of each stability class was used to determine the Bi and Vi values.Unit 2 Revision 28 II 72 September 2006 APPENDIX C DOSE PARAMETERS FOR IODINE 131 and 133, PARTICULATES AND TRITIUM Unit 2 Revision 28 September 2006 1173 Appendix C DOSE PARAMETERS FOR IODINE -131 AND -133, PARTICULATES AND TRITIUM This appendix contains the methodology which was used to calculate the organ dose factors for 1-131, 1-133, particulates, and tritium. The dose factor, Ri, was calculated using the methodology outlined in NUREG-0 133. The radioiodine and particulate DLCO 3.2.1 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 all pathways.
However, for dose compliance calculations, 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.1 Inhalation Pathway Ri(I) -K'(BR)a(DFA)ija where: Ri(I) dose factor for each identified radionuclide i of the organ of interest (units = mrem/yr per uCi/m 3);K' = a constant of unit conversion, 1 E6 pCi/ýtCi (BR)a -Breathing rate of the receptor of age group a, (units = m 3/yr);(DFA)ija The inhalation dose factor for nuclide i, organ j and age group a, and organ t (units = mrem/pCi).
The breathing rates (BR)a for the various age groups, as given in Table E-5 of Regulatory Guide 1.109 Revision 1, are tabulated below.Age Group (a) Breathing Rate (m 3/yr)Infant 1400 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DFA)ija for the various age groups are given in Tables E-7 through E- 10 of Regulatory Guide 1.109 Revision 1.Unit 2 Revision 28 1174 September 2006 Appendix C (Cont'd)C.2 Ground Plane Pathway-xit Ri(G) = K'K' (SF) (DFG)i (1-e xi Where: Ri(G) = Dose factor for the ground plane pathway for each identified radionuclide i for the organ of interest (units = m-mrem/yr per uCi/sec)K' = A constant of unit conversion, 1 E6 pCi/uCi K" = A constant of unit conversion, 8760 hr/year ki= The radiological decay constant for radionuclide i, (units = secl)t = The exposure time, sec, 4.73E8 sec (15 years)(DFG) = The ground plane dose conversion factor for radionuclide i; (units mrem/hr per pCi/mr 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 DFGi values is presented in Table E-6 of Regulatory Guide 1.109 Revision 1.Unit 2 Revision 28 II 75 September 2006 Appendix C (Cont'd)C.3 Grass-(Cow or Goat)-Milk Pathway K'QfUapFm(r)(DFL)ia, (fIf (1-fpf)e-Ah]
Ai Rk( -(Ai Aý) f[ f + fe1 je-" Where: Ri(C) Dose factor for the cow milk or goat milk pathway, for each identified radionuclide i for the organ of interest, (units = m2-mrem/yr per uCi/sec)K' 1 A constant of unit conversion, 1E6 pCi/ýtCi Qf = The cow's or goat's feed consumption rate, (units = kg/day-wet weight)Uap = The receptor's milk consumption rate for age group a, (units = liters/yr)
Yp = 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 = Fraction of deposited activity retained on cow's feed grass (DFL)iat = The ingestion dose factor for nuclide i, age group a, and total body or organ t (units = mrem/pCi)?i The radiological decay constant for radionuclide i, (units=sec
-1)-w The decay constant for removal of activity on leaf and plant surfaces by weathering equal to 5.73E-7 sec -I (corresponding to a 14 day half-life) tf = The transport time from pasture to cow or goat, to milk, to receptor, (units = sec)th = The transport time from pasture, to harvest, to cow or goat, to milk, to receptor (units sec)fP Fraction of the year that the cow or goat is on pasture (dimensionless) fs Fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless)
Unit 2 Revision 28 II 76 September 2006 Appendix C (Cont'd)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 1, the value of fs 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 fp was obtained from the environmental group.Table C-1 contains the appropriate values and their source in Regulatory Guide 1.109 Revision 1.The concentration of tritium in milk is based on the airborne concentration rather than the deposition.
Therefore, the RT(C) is based on X/Q: RT(C) = K'K"' FmQfUap(DFL)iat 0.75(0.5/H)
Where: RT(C) = Dose factor for the cow or goat milk pathway for tritium for the organ of interest, (units = mrem/yr per ptCi/mi 3)K"' = A constant of unit conversion, 1E3 g/kg H = Absolute humidity of the atmosphere, (units = g/m 3)0.75 = The fraction of total feed that is 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 2 Revision 28 1177 September 2006 Appendix C (Cont'd)C.4 Grass-Cow-Meat PathwayK'QfUapFf(r)(DFL)i,(+
fPfy +(1-fPf,)e-;{th X ) ~ ~ ~ (A1 A,,) I Ye , -Ri(M) = Dose factor for the meat ingestion pathway for radionuclide i for any organ of interest, (units = m -mrem/yr per ýtCi/sec)Ff = The stable element transfer coefficients, (units = pCi/kg per pCi/day)Uap = The receptor's meat consumption rate for age group a, (units = kg/year)th = The transport time from harvest, to cow, to receptor, (units = sec)tf = The transport time from pasture, to cow, to receptor, (units = sec)All other terms remain the same as defined for the milk pathway. Table C-2 contains the values which were used in calculating Ri(M).The concentration of tritium in meat is based on airborne concentration rather than deposition.
Therefore, the RT(M) is based on X/Q.RT(M) = K'K"'FfQfUap(DFL)iat
[0.75(0.5/H)]
Where: All C.5 RT(M) = Dose factor for the meat ingestion pathway for tritium for any organ of interest, (units = mrem/yr per ýtCi/m 3)other terms are defined above.Vegetation Pathway The integrated concentration in vegetation consumed by man follows the expression developed for milk. Man is considered to consume two types of vegetation (fresh and stored)that differ only in the time period between harvest and consumption, therefore:
V) = K' r (DFL) iat[ ULaFLe + USaFgeXith Yv(Xi + X.)Unit 2 Revision 28 September 2006 1178 Appendix C (Cont'd)Where: Ri(V) = Dose factor for vegetable pathway for radionuclide i for the organ of interest, (units = m 2-mrem/yr per pLCi/sec)K' A constant of unit conversion, 1 E6 pCi/gCi ULa = The consumption rate of fresh leafy vegetation by the receptor in age group a, (units = kg/yr)Usa -The consumption 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 Fg = The fraction of the annual intake of stored vegetation grown locally tL -The average time between harvest of leafy vegetation and its consumption, (units = sec)th = The average time between harvest of stored vegetation and its consumption, (units = sec)Yv The vegetation areal P density, (units = kg/m 2)All other factors have been defined previously.
Table C-3 presents the appropriate parameter values and their source in Regulatory Guide 1.109 Revision 1.In lieu of site-specific data, values for FL and Fg of, 1.0 and 0.76, respectively, were used in the calculation.
These values were obtained from Table E-15 of Regulatory Guide 1.109 Revision 1.The concentration of tritium in vegetation is based on the airborne concentration rather than the deposition.
Therefore, the RT(V) is based on X/Q: RT(V) = K'K'" [ULa fL + USa fg](DFL)iat 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 jtCi/m 3).All other terms are defined in preceeding sections.Unit 2 Revision 28 1179 September 2006 Parameters for Parameter Qf (kg/day)r (DFL)ija (mrem/pCi)
Fm (pCi/liter per pCi/day)Y, (kg/mr 2)Yp (kg/mr)th (seconds)tf (seconds)Uap (liters/yr)
TABLE C-1 Grass -(Cow or Goat) -Milk Pathways Reference Value (Reg. Guide 1.109 Rev. 1)50 (cow) Table E-3 6 (goat) Table E-3 1.0 (radioiodines)
Table E-15 0.2 (particulates)
Table E-15 Each radionuclide Tables E-1 1 to E-14 Each stable element Table E-1 (cow)Table E-2 (goat)2.0 Table E-15 0.7 Table E-15 7.78 x 106 (90 days) Table E-15 1.73 x 10 5 (2 days) Table E-15 330 infant Table E-5 330 child Table E-5 400 teen Table E-5 310 adult Table E-5 Unit 2 Revision 28 September 2006 1180 TABLE C-2 Parameters for the Grass-Cow-Meat Pathway Parameter r Ff (pCi/kg per pCi/day)Uap (kg/yr)(DFL)ija (mrem/pCi)
Yp (kg/mr 2)Y, (kg/mr 2)th (seconds)tf (seconds)Qf (kg/day)Value 1.0 (radioiodines) 0.2 (particulates)
Each stable element 0 infant 41 child 65 teen 110 adult Each radionuclide 0.7 2.0 7.78E6 (90 days)1.73E6 (20 days)50 Reference (Reg. Guide 1.109 Rev. 1)Table E- 15 Table E- 15 Table E-1 Table E-5 Table E-5 Table E-5 Table E-5 Tables E- II to E- 14 Table E-15 Table E- 15 Table E-15 Table E- 15 Table E-3 Unit 2 Revision 28 September 2006 1181 TABLE C-3 Parameters for the Vegetable Pathway Reference (Reg. Guide 1.109 Rev. 1)Parameter Value r (dimensionless)(DFL)ija (mrem/pCi)
UL)a (kg/yr) -infant-child-teen-adult US)a (kg/yr) -infant-child-teen-adult 1.0 (radioiodines) 0.2 (particulates)
Each radionuclide Table E- 1 Table E-1 Tables E- 11 to E-14 0 26 42 64 0 520 630 520 Table E-5 Table E-5 Table E-5 Table E-5 Table E-5 Table E-5 Table E-5 Table E-5 Table E- 15 Table E- 15 tL (seconds)th (seconds)Yv (kg/mr 2)8.6E4 (1 day)5.18E6 (60 days)2.0 Table E- 15 Unit 2 Revision 28 September 2006 II 82 APPENDIX D DIAGRAMS OF LIQUID AND GASEOUS TREATMENT SYSTEMS AND MONITORING SYSTEMS Unit 2 Revision 28 September 2006 1183 Liquid Radwaste Treatment System Diagrams Unit 2 Revision 28 September 2006 1184 SPENT FUEL -POOL COOLING THERMEX SYSTEM RADWASTE, DEMINERALIZER REACTOR WATER CLEANUP SYSTEM REGENERANT.
EVAPORATOR'I CONDENSATE4 DEMINERALIZERS' REACTOR BUILDING EQUIPMENT DRAINS PHASE SEPARATOR RECOVERY, SAMPLE TURBINE BLDG EQUIPMENT DRAINS'RESIDUAL HEAT REMOVAL SYSTEM'RECOVERY, SAMPLE SYSTEM RADWASTE, FILTERS FLOOR DRAIN FILTER RADWASTE FILTERS WASTE COLLECTION II 85 SERVCt AIR WASTE COLLECTOR TREATMENT SYSTEM II 86 FV 330 EVAP WATER HG F IGE AOIv I RANGEHIH E33 142 ~FV 331 F DRAIN OR TANKS LOW FE 331*RANGE RECOVERY SAMPLE SYSTEM and WASTE DISCHARGE SAMPLE SYSTEM 1187 REGEN CST RX BLDG THERMEX BLDG)RAINS SYSTEM DRAINS SPENT RESIN TANKS.AUX.BOILER BLDG SUMP RW BLDG DRAINS I -TYPICAL OF 2------------OTHER FLOOR DRN SUCT LINE RW DEMINS w WASTE FLOOR COLLECTOR COLLECTOR SURGE TK 17 TANK OTHER FLOOR O DRAIN COLLECTOR TANK FLOOR DRAIN COLLECTION SYSTEM 1188 WASTE COLLECTOR--CI-SURGE TANK[.1 AOV230 1 FLOOR DRAIN COLLECTOR SURGE TANK-INSTRUMENT AIR ICOND. MAKEUP AND DRAW OFF FLOOR DRAIN COLLECTOR TANKS REGEN. WASTE TANKS I WASTE COLLECTOR D-TANKS SLUDGE TANK AOV214REGEN. WASTE TANK AOV127 WASTE DISCH.SAMPLE TANK AOV123 14., FLOOR DRAIN COLLECTOR-TANK AOV126_ WASTE COLLECTOR:
TANK FLOOR DRAIN FILTER SYSTEM'II 89 THERMEX SYSTEM RW TYPICAL OF 2------------------------
4 "AV3 BACK OTHER F WASTE REGENERANT TANK REGEN EVAP 61-THERMEX SYSTEM AoV93 I---- --------------------I REGENERANT WASTE SYSTEM 1190 i npwi~r cm EVAP AND STORAE AOV218 COLL. TANKS BOTTOMS.. PUMPS FLOOR DRAIN 2AOV87 COLL. TANKS WASTE DISCH AOV147 SAMPLE TANKS REGENERANT EVAPORATOR SYSTEM 1191 Gaseous Treatment System Diagrams Unit 2 Revision 28 September 2006 II 92 II 93 Title: CONDENSER AIR REMOVAL SYSTEM fO DRY~fIS 1194 Title: OFFGAS RECOMBINERS t0 CHANCOAL AesSOwml famcw~1eCLCW ,ALL LINES AND EQUIPMENT
.i;LOCATED INSIDE THIS BOUNDARY iTO BE ABANDONED IN PLACE...................Title: OFFGAS DRYERS 1195 FROM -I OFFr GAS DRYERS F E AOv 103 To Main Stack 1196 Title: OFFGAS SYSTEM CHARCOAL ABSORBERS F"O RlEACYOR M~kDWON VENHIL PEN FR" PYHJARY CONTAPOANT F I OUTSVE'Aln II 97 Title-STANDBY GAS TREATMENT SYSTEM Liquid Radiation Monitoring Diagrams Unit 2 Revision 28 September 2006 1198 FROM MRS SYSTEM RHR SERVICE WATER (A)RHR SERVICE WATER (8)QA CAT I FROM RHR HEATEx. EIA FROM CCF HEATerX. EIA.t .,C OA C$t El r. WE\146A t I I FROM RHR HEATEX. EIB9 FROM CCS HEATEX. EIA.IB,IC FROM SJAe PRECOOLERS eA.2B111 TA CAT r'RE\F 408lýDISCHARGE RAY 6 J g COOLING TOWER BLOWDOWN OA CAT El ALL LIOUID RADWASTE SFC SYSTEM LIQUID RADIATION MONITORING SHEET 2 OF 2 SHEET 2 OF 2 II 99 NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT FLOW PURGE/TEST I ; SAMPLE CONNECTION T T INLET I) GLOBE VALVE, ALL OTHER MANUALLY OPERATED VALVES ARE BALL VALVES CALIBRATION/DRAIN CONNECTION OFFLINE LIQUID MONITOR 2CCP-CABII5 2CCS-CA8152 2CWS-CABI57 LEGEND ( PRESSURE INDICATOR O FLOW INDICATING SW.7S SOLENOID OPERATED SW.H NORMALLY CLOSED VALVE M NORMALLY OPEN VALVE OFF-LINE LIQUID MONITOR NIAGARA MOH-AIWIK POWER CORPORATION NINE MILE POINT-UNIT 2 UPDATED SAFETY ANALYSIS REPORT 11100 95ALX XLiVJ.UD 3w OCOE 1,;,.UH '"K 39 91 FLOW LE PURGE/TEST SAMPLE ET CONNECTION INLET LOW FLOW DATA ACOUISITION UNIT t : --- ---(DAU)* I DETECTOR ji I CF HECK SOURCE I I I -PURGE CONTROL GRAB SAMPLER I I_ PUMP CONTROL I II) PUMP PURGE OUTLET SAMPLER CALIBRATION/DRAIN OFFLINE LIOUID MONITOR CALIBRATION TEST/ CONNECTION VENT CONNECTION 2LWS-CAB206 LEGEND NOTES.(1) GLOBE VALVE. ALL OTHER MANUALLY OPERATED VALVES ARE BALL VALVES PRESSURE INDICATOR O FLOW INDICATING SW.W SOLENOID OPERATED SW.H NORMALLY CLOSED VALVE H NORMALLY OPEN VALVE OFF-LINE LIQUID MONITOR NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 11101 UPDATED SAFETY ANALYSIS REPORT I USAR REVISION 0 APRIL 1989 USAR REVISION 0 APRIL 1989 LOW FLOW/NORMAL FLOW I FS I d-- FLOW SAMPLE OUTLET DATA ACOUISITION UNIT (DAU)I I113 SAMPLE I INLET I-~ I SAMPLE SPURGE ol XI ui Ul I I I.I I I II----2 I (I) P.......NOTES a) GLOBE VALVE. ALL OTHER MANUALLY OPERATED VALVES ARE BALL VALVES ZCALIBRATION TEST/VENT CONNECTION CALIBRATION/DRAIN CONNECTION OFFLINE LIOUID MONITOR>SWPCAB23A 2SWP-CAS238 2SWP-CABI46A 2SWPOCABI468 LEGEND PRESSURE INDICATOR OFLOW SW.FS] SOLENOID OPERATED SW.H NORMALLY CLOSED VALVE NORMALLY OPEN VALVE OFF-LINE LIQUID MONITOR 11102 NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 UPDATED SAFETY A1NALYSIS REPORT Gaseous Effluent Monitoring System Diagrams Unit 2 Revision 28 September 2006 11103 I I ~ ~ ~ ~ ~~c I I
- II
- I. I LINER FILLING HOOD EXH. 0 HEPA FILTER CFm CAM RDAT81E0 9 CM RADWASTE BUILDING 211 RADWASTE GENERAL AREA l TICM DECONTH.A HL ER A FILTER AUX.__EXH. AREA lI SERVICE BLDG.GENERAL EXH. AREA VENT H300 CFM RADWASTE 149101 CFM 1711 EOUIP. EX-, I FOA CAT 11 RADWASTE M --BUILDING 283 F GEN. AREA VENTILATION I265C EXH. HEPA FILTER 231 I OA CAT I-'66500-69000 CFM RECIRCULATION MODE -NO-, 00nn f.Fu rNTAINlMTNT
.. P1 CAM ACF: AUTOMATIC CONTROL FUNCTION PAM: POST ACCIDENT MONITOR-ý PARTICULATE
& IODINE SAMPLING CAPABILITY CAM: CONTINUOUS AIRBORNE MONITOR I : SAFETY-RELATED MONITOR NOTES, I. MODIFICATION 95-O11 HAS BEEN INSTALLED TO ALLOW CONCURRENT OPERATION OF ALL 3 EXHAUST FANS. WHEN ALL 3 FANS ARE RUNNING THERE WILL BE AN ADDITIONAL EXHAUST OF z 17.SOD CFM.I-GASEOUS RADIATION MONITORING NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 YUPDATED SAFET ANALYSIS REPORT U IDT SAN EIISI IAFEOVEIBER,998 I I i I " I USAR REVISION*
10 NOVEMBER 19 8 STACK SpV P V P V V p p FRESH AIR Fp PUMP CONTROLDETECTOR DE" O I NOT F FLOW SENSORS AM LFI R 7AMPLIFIER SO R E VALVES, ETC, OPERATIONAL
~'& ADC 1, ------ rAMPLIFIER
& ADC .0 dD Ly ULTICANNI CONTROLLER HOST COMPUTER DUAL DISK DRIVE BLOCK DIAGRAM F-nTYPICAL GASEOUS EFFLUENT MONITORING SYSTEM NIAGARA MOHAWK POWER CORPORATION 11105 NINE MILE POINT-UNI1`2 UPDATED SAFETY ANALYSIS REPORT I APPENDIX E NINE MILE POINT ON-SITE AND OFF-SITE MAPS Unit 2 Revision 28 September 2006 11106 FIGURE 5.1 -1 5 W Sw -A". (o.U -,,= A,.g*..tS*I 'l N= l-.a I I I S-11107 odcmfig5l2.dgn SCALE OF MILES' .FIGURE D5.1-2 5 LEGEND NINE MILE POINT InteA r tate ..............................
U.S* & Slate Highways ....Count Road& ........................
to n ...........
.............
...Tow Line.. .Count ..........................
(12 25)City L Village Lines ...............
... .. ..Railroad...............
ENVIRONMENTAL SAMPLE ...........
LOCATION Lstitude 43*28'N.Longitude 7630'W.at Oswego County Bldg. Oswego, N.Y.Land Area 968 Square rnines Rainbow Shores 68 49 52 40 67 6 Sekltar 553 LAKE_ ONTARIO 'A..00 25 Mexico Poin i e i Fleasn ik Point 16~u umlstaBdd"\AeSaal....... I n
-I- .1. i. -..-im M -.IL kv I! 'L .AM lI7 tkA Y I J CTo N.. 370 COUNTY W Ce t-31\"- .oNONDAGA COUNTY To Sgoaoo. ____STc Sposcooe Three Ij Comers FIGURE D5.1-2 20 M 79