ML19336A407
| ML19336A407 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 09/30/1980 |
| From: | Croney S, Mandler J, Motes B ALLIED CHEMICAL CORP., EG&G, INC. |
| To: | |
| References | |
| CON-FIN-A-6075 NUREG-CR-1629, NUDOCS 8010230017 | |
| Download: ML19336A407 (600) | |
Text
{{#Wiki_filter:- I NUREG/CR-1629 RR In-Plant Source Term Measurements at Turkey Point Station - Units 3 and 4 Manuscript Completed: July 1980 Date Published: September 1980 Prepared by
'J. W. Mandler, S. T. Croney, N. C. Dyer, C. V. Mcisaac, A. C. Stalker
**B. G. Motes, J. H. Keller, T. E. Cox, D. W. Akers, J. W. Tkachyk, S. W. Duce
'EG6G Idaho, Inc.
Idaho National Engineering Laboratory Idaho Falls, ID 83401
" Allied Chemical Corporation Idaho National Engineering Laboratory Idaho Falls, ID 83401 Prepared for Division of Safeguards, Fuel Cycle and Environmental Research Offica of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Wcshington, D.C. 20555 NRC FIN No. A6075 l
i l A3 b
ABSTRACT This report presents data obtained at Turkey Point Units #3 and #4 as a part of the in-plant source term measurement program in operating pressurized water reactors (PWR*s). The work was conducted for the Office of Nuclear Regulatory Research in support of the Effluent Treatment Systems Branch of the Office of Nuclear Reactor Regt:ation. The primary objective of this program is to provide the Nuclear Regulatory Commission 3 (NRC) with operational data that can be used in evaluation of plant l designs for liquid and gaseous waste treatment systems. Data presented were obtained at the Turkey Point Power Station operated by Florida Power and Light, located south of Miami, Florida. In-plant measurements were conducted during the time period from November, 1977 through May, 1978. This plant is the third in a planned series of six operating PWR's to be studied, two from each of the taajor PWR vendors. Data from all plants will be combined and interpreted to provide a data base for radioisotope inventory in plant systems, radioactive waste treatment system performance, and source terms for both liquid and gaseous systems. One of the primary objectives in performing measurements at Turkey Point was to study primary-to-secondary leaks if they occurred and to determine partition factors in steam generators. The opportunity to study primary-to-secondary leaks occurred twice during the in-plant measurement period. Results of these studies together with measurements performed on the liquid and gaseous systems at Turkey Point are presented. i i I l iii i
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TABLE OF CONTENTS Page Abstract . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii Acknowl edgements . . . . . . . . . . . . . . . . . . . . . . . . . . xxvi i , 1. Introduccion ............. . . . . . . . . . . . . I 1.1 Objectives of the In-Plant Measurement Program . . . . . . I 1.2 Turkey Point . . . . . . . . . . . . . . . . . . . . . . . 2
- 1.2.1 In-Plant Measurements at Turkey Point . . . . . . . 2 1
1.2.2 Description of Turkey Point . . . . . . . . . . . . 2 1.2.3 Plant Data .................... 3
- 2. Summary and Conclusions . . . . . . . . . . . . . . . . . . . 7 2.1 General Plant Operations during i i-Plant iteasurements . . 7 2.2 L i q u i d Sy s t em s . . . . . . . . . . . . . . . . . . . . . . 7 2.2.1 Description of Liquid Systems . . . . . . . . . . . 7 4
2.2.2 Reactor Coolant . . . . . . . . . . . . . . . . . . 7 2.2.3 Chemical and Volume Control System . . . . . . . . 13 2.2.4 Boric Acid Recovery and Radwaste Systems . . . . . 14 2.2.5 Secondary System ................. 16 2.2.6 Spent Fuel Pit .................. 19 } 2.2.7 General Conclusions - Demineralizers and Evaporators . . . . . . . . . . . . . . . . . . . . 19 l 2.3 Gaseous Systems ..................... 25
- 3. Reactor Coolant e.s Letdown System . . . . . . . . . . . . . . 38 3.1 System Description and Sample Points . . . . . . . . . . . 38 3.1.1 Reactor Coolant System . . . . . . . . . . . . . . 38 3.1.2 Letdown System ................. . 38 y
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i l l l TABLE OF CONTENTS (cont'd) i Page 1 3.1.2.1 System Description . . . . . . . . . . . . 38 3.1.2.2 Sample Points .............. 44 3.1. 2. 3 Sample Collection ............ 44 3.2 Discussion of Measurement Data - Reactor Coolant . .... 44 l 3.2.1 Radionuclide Concentrations in Reactor Coolant .. 44 J 3.2.2 Predicted Radionuclide Concentrations in Coolant Waters .................. 56 3.2.3 Spiking Studies . . . . . . . . . . . . . . . . . . 60 3.3 Discussion of Measurement Data - Letdown . Demineralizers . . . . . . . . . . . . . . . . . . . . . . 69 3.3.1 Unit #3 . . . . . . . . . . . . . . . . . . . . . . 69 , 3.3.2 Unit #4 . . . . . . . . . . . . . . . . . . . . . . 79 3.3.3 Conclusions . . . . . . . . . . . . . . . . . . . . 83 i i 4. Boric Acid Recovery and Liquid Radwaste Systems . . . . . . . . 95 4.1 System Gescription and Sample Points . . . . . . . . . . . 95 I 4.1.1 Boric Acid Recovery System ............ 95 i l 4.1.2 Radwaste System . . . . . . . . . . . . . . . . . . 98 2 4 4.2 Discussion of Measurements Data - Boric Acid Recovery System ....................102 4.2.1 Measurements ..................102 4.2.2 Base Cation Demineralizers . . . . . . . . . . . 102 4.2.3 Boric Acid Evaporator . . . . . . . . . . . . . . 105 108 4.2.4 Condensate Demineralizer and Filter . . . . . . . 4.2.5 Summary of Results . . . . . . . . . . . . . . . 10 8 vi
TABLE OF CONTENTS (cont'd) . Page 4.3 Discussion of Measurement Data - Liquid Radwaste Sy s t em . . . . . . . . . . . . . . . . . . . . . . . . . 13 3 4.3.1 Introduction ..................133 4.3.2 Radwaste Evaporator . . . . . . . . . . . . . . . 133 4.3.3 Condensate Demineralizer . . . . . . . . . . . . 149
. 4.3.4 Test Demineralizer . . . . . . . . . . . . . . . 149 i
4.4 Radionuclide Concentrations in Tanks . . . . . . . . . . 153 4.5 Conclusions ......................153
- 5. Spent Fuel Pit . . . . . . . . . . . . . . . . . . . . . . . 160 5.1 Sy s t em De sc ri pt i o n . . . . . . . . . . . . . . . . . . . 160 5.2 Mea sure .ient s . . . . . . . . . . . . . . . . . . . . . . 160 5.3 Resul ts and Di scussion . . . . . . . . . . . . . . . . . 162 5.3.1 Unit #3 Fuel Pit Area Extrapolated Annual Gaseous Releases for 1311, 3H, and 14C . . . . . 162 5.3.2 Unit #3 Spent Fuel Pit and Associated Water Tritium Mass Balance . . . . . . . . . . . . . . 164 5.3.3 Unit #3 Spent Fuel Pit Demineralizer Decontamination Factors . . . . . . . . . . . . . 168
- 6. Secondary System . . . . . . . . . . . . . . . . . . . . . . 174 6.1 Introduction . . . . . . . . . . . . . . . . . . . . . . 174 6.2 System Description . . . . . . . . . . . . . . . . . . . 174 6.2.1 Steam Generator . . . . . . . . . . . . . . . . . 174 6.2.2 Turbine Train . . . . . . . . . . . . . . . . . . 174 i
6.2.3 Gland Seal . . . . . . . . . . . . . . . . . . . 177 6.2.4 Moisture Separator Reheater . . . . . . . . . . . 177 6.2,5 Steam det Air Ejectors .............177 vii ;
l 1 I l I l l l TABLE OF CONTENTS (cont'd) 1 Page 6.2.6 Main Condenser .................177 6.2.7 Condensate Storage Tank . . . . . . . . . . . . . 178 6.2.8 Condensate Pumps . . . . . . . . . . . . . . . . 178 i 6.2.9 Feedwater Reheaters . . . . . . . . . . . . . . . 178 6.2.10 Chemical Injection . . . . . . . . . . . . . . . 178 6.2.11 Feedwater Pumps . . . . . . . . . . . . . . . . . 178 6.2.12 Circul ating Water Pumps . . . . . . . . . . . . . 179 6.2.13 R ated Fl ows . . . . . . . . . . . . . . . . . . . .,179 6.3 Sample Points .... .................179 6.4 Sample Types and Procedures . . . . . . . . . . . . . . 1 81 6.5 Results ...... ..................181
- 6. 5.1 Pl a nt Ch emi st ry . . . . . . . . . . . . . . . . . 181 6.5.1.1 Reasons for Chemistry Control . . . . . 181 6.5.1.2 Reactor Coolant System . . . . . . . . . 182 6.5.1.3 Secondary Chemistry . . . . . . . . . . 18 2 6.5.2 Primary-to-Secondary Leak Rates . . . . . . . . . 198 C.5.3 Relative Iodine Isotopic Age . . . . . . . . . . 19 9 203 I 6.5.4 Blowdown Flash Tank . . . . . . . . . . . . . . .
6.5.5 Steam Generator Decontamination Factors . . . . . 208 6.5.6 High Pressure Drains . . . . . . . . . . . . . . 211 l 5 6.5.7 Main Steam Air Ejector . . . . . . . . . . . . . 213 6.5.8 Turbine Gland Seal Exhaust Vent . . . . . . . . . 213 l l i 6.6 Conclusions .......... . . . . . . . . . . . . 216 l viii l
1 TABLE OF CONTENTS (cont'd) Page
- 7. Waste Gas Processing and Containment Building Systems . . . . 217 7.1 Waste Gas Processing System . . . . . . . . . . . . . . 217 7.1.1 System Description . . . . . . . . . . . . . . . 217 7.1.2 Measurement Data and Methods . . . . . . . . . . 217 7.1.3 Results and Discussion . . . . . . . . . . . . . 219 7.2 Containment Building System . . . . . . . . . . . . . . 228 7.2.1 System Description . . . . . . . . . . . . . . . 228 7.2.2 Measurement Methods . . . . . . . . . . . . . . . 228 7.2.3 Results and Discussion . . . . . . . . . . . . . 230 7.2.3.1 Reactor Coolant Effective Radionuclide Inventory Leakage Rates, Effective Partition Factors, and Iodine Species . 230 7.2.3.2 Containment Purge Frequency . . . . . . 237 7.2.3.3 Extrapolated Annual Radionuclide Release ................243
- 8. Auxiliary Building Ventilation System . . . . . . . . . . . . 252 8.1 System Description and Sampling Methods . . . . . . . . 252 8.2 Measurement Data . . . . . . . . . . . . . . . . . . . . 256 8.3 Results and Discussion . . . . . . . . . . . . . . . . . 256 8.3.1 Normalized 1311 and 3H Release Rates . . . . . . 258 8.3.2 Iodine-131 Sources and Annual Release Rates . . . 262
- 8.3.3 Particulate Source Terms and Annual Releases i for Gaseous Effluents . . . . . . . . . . . . . . 266 8.3.4 Stack Gaseous 3H and 14C Release Rates . . . . . 266 8.3.5 Decontamination Factor for HEPA Filters . . . . . 270
- 8.3.6 Effective Reactor Coolant Leakage Rates, Partition Factors, and 131 I Species . . . . . . . 270 ix
TABLE OF CONTENTS (cont'd) i Page 8.3.7 Stack Release Rates for Selected Beta-Emitting Nuclides . . . . . . . . . . . . . . . . . . . . 275 8.3.8 Stack Release Rates for 133Xe . . . . . . . . . . 275 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . 279 l Appendix A . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 A.1 Sample Handling . . . . . . . . . . . . . . . . . . . . . A-1 l A.l.1 Sample Volumes and General Sampling Procedures . . A-1 A.l.2 Sampl e Val idation . . . . . . . . . . . . . . . . . A-1 A.l.2.1 Liquid Sampl es . . . . . . . . . . . . . . A-1 A.l.2.2 Gaseous Samples . . . . . . . . . . . . . A-2 A.l.3 Validation of Sample Analyses . . . . . . . . . . . A-2 A.2 Data Handling . . . . . . . . . . . . . . . . . . . . . . A-3 Appendix B . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 l l l l l x 1
LIST OF TABLES 4 Page 2.1 "Best Value" Decontamination Factors for CVCS 4 Mixed-Bed Demineralizers . . . . . . . . . . . . . . . . . 15 2.2 "Best Value" Decontamination Factors for Boric Acid Recovery System ..................... 17 ] 2.3 "Best Va!ue" Decontamination Factors for Radwaste System ......................... 17 2.4 "Best Value" Decontamination Factors for Spent Fuel Pit Demineralizer . . . . . . . . . . . . . . . 20 2.5 Average Release Rates During Combined Refueling and Non-Refueling Periods .................. 27 2.6 Auxiliary Building 131I Sources ............. 29 2.7 Average Release Rates During Non-Refueling Interval ... 30 2.8 Average Release Rates During Refueling Interval ..... 31 2.9 Main Stack 3 H and 131 I Normalized Release Rates and Effective Partition Coefficients . . . . . . . . . . . . . 32 4 2.10 Average 131I Distribution at Turkey Point ........ 34 2.11 Average Leakage Rates of Reactor Coolant into the Containment and Auxiliary Building . . . . . . . . . . . . 35 3.1 Principal CVCS Component Data Summary .......... 42 4 3.2 Average Radionuclide Concentrations in Reactor Coolar.t, Reactor Power Operations (Non-Spiking) - Unit #3 . . . . . 46 3.3 Iodine Ratios in Unit #3 Reactor Coolant . . . . . . . . . 49 3.3A Theoretical Iodine Ratios Due to Recoil, Diffusion, and Equilibrium Fuel Fail Mechanisms . . . . . . . . . . . . . 49 3.4 Average Radionuclide Concentrations in Reactor Coolant, j Reactor Power Operations (Non-Spiking) - Unit #4 . . . . . 51 3.5' Iodine Ratios in Unit #4 Reactor Coolant . . . . .... 56
- 3.6 Predicted Radionuclide Concentrations in Reactor Coolant and Secondary Coolant .................. 57 i
xi
LIST OF TABLES (cont'd)
- Page 3.7 Parameter Values Used to Modify N237 Predicted Radionuclide Concentrations for Turkey Point Unit #3 l or #4 .......................... 59 Comparison of Predicted and Measured Radionuclide
~ 3.8 Concentrations in Reactor Coolant ............ 61 3.9 131I Concentrations in Reactor Coolant for Shutdown Spike Unit #3, 5/19-20/78 ................ 68 3.10 Unit #3 Letdown Sample Information . . . . . . . . . . . . 70 3.11 DF's for Unit #3 CVCS Mixed Bed B Demineralizer ..... 72 3.12 Means and Ranges for Radionuclide Concentrations and "Best Value" DF's for Unit #3 CVCS Mixed Bed B Demineralizer ...................... 80 3.13 Unit #4 Letdown Sample Information . . . . . . . . . . . . 81
- 3. ~14 DF's for Unit #4 CVCS Mixed-Bed A Demineralizer ..... 84 j
3.15 Means and Ranges for Radionuclide Concentrations and "Best Value" DF's for Unit #4 CVCS Mixed Bed A Demineralizer . . 92 4.1 Principal Components in Radwaste and Boric Acid
> Rec ov e ry Sy s t ems . . . . . . . . . . . . . . . . . . . . . 97 4.2 Boric Acid Recovery System Feed Contents and Sources . . 103 4.3 DF's for Base Cation Demineralizers . . . . . . . . . . 110 4.4 DF's for Boric Acid Evaporator . . . . . . . . . . . . . 112 4.5 DF's for BAE Condensate Demineralizer . , . . . . . . . 114 4.6 DF's for BAE Condensate Demineralizer Filter . . . . . . 116
- 4.7 Boric Acid Recovery System with Base Cation Demin. C Mean hadionuclide Concentrations and "Best Value" DF's . 118
- 4.8 Boric Acid Recovery System with Base Cation Demin. A Mean Radionuclide Concentrations and "Best Value" DF's . 122 4.9 Ri dwaste Evaporator Operati.ig Parameters . . . . . . . . 134 4.10 Means and Ranges for Radionuclide Concentrations in Radwaste Evaporator Feed, Distillate, and Bottoms . . . 136 xii
i i j LIST OF TABLES (cont'd) 4 Page i 4.11 Decontamination Factors for Radwaste Evaporator . . . . 137 4.12 "Best Value" DF's for Radwaste Evaporator . . . . . . . 148 i 4.13 DF's for Radwaste Evaporator Condensate Demineralizer . 150 a 4.14 Means and Ranges for Radionuclide Concentrations and "Best Value" DF's for Radwaste Evaporator Condensate Demineralizer . . . . . . . . . . . . . . . . . . . . . 151 1 4.15 Decontamination Factors for Test Demineralizer Train . . 152 4.16 Radionuclide Concentrations in Holdup and Waste i Holdup Tanks . . . . . . . . . . . . . . . . . . . . . . 154 4.17 Radionuclide Concentrations in Radwaste Building Monitor Tanks . . . . . . . . . . . . . . . . . . . . . 156 4.18 Radionuclide Concentrations in Auxiliary Building Monitor Tanks . . . . . . . . . . . . . . . . . . . . . 157 i
- 4.19 Extrapolated Annual Releases from Monitor Tarks . . . . 158
, 5.1 Extrapolated Annual Releases of Gaseous Tritium, 1311, and I"C from the Unit #3 Fuel Pit Area . . . . . . . . . 163 5.2 Average Release Rates of Gaseous 3H,14C, and 1311 from Unit #3 Fuel Pit Area . . . . . . . . . . . . . . . 165 5.3 Unit #3 Fuel Pit Area Duct Gaseous 3H and 14C Species . . . . . . . . . . . . . . . . . . . . . . 166 5.4 Unit #3 Tritium Mass Balance . . . . . . . . . . . . . . 167 5.5 Sample Information for Unit #3 SFP Demineralizer Tests . 169 l , 5.6 Measured DF's for #3 SFP Demineralizer . . . . . . . . . 170 l
l 5.7 Unit #3 SFP Demineralizer DF's for Beta-0nly-Emitting 172 Radionuclides . . . . . . . . . . . . . . . . . . . . . 5.8 Means and Ranges for Radionuclide Concentrations and "Best Value" DF's for #3 SFP Demineralizer . . . . . . . . . . 173 6.1 Rated Flows for Secondary System . . . . . . . . . . . . 180 6.2 Rated Flows for Primary System . . . . . . . . . . . . . 180 l l xiii l i
l i l 4 1 i i LIST OF TABLES Page ! i Average Primary Chemistry - Unit #3 I 6.3 Period: 11/9-21/77 ..................183 l 6.4 Average Primary Chemistry - Unit #4 Period: 1/18-26/78 . . . . . . . . . . . . . . . . . . 184 6.5 n, cage Primary Chemistry - Unit #4 Period: 2/3-9/78 ...................185 6.6 Average Secondary Chemistry - Unit #3 j Period: 11/9/77 . . . . . . . . . . . . . . . . . . . . 188 l 6.7 Average 5es.adary Chemistry - Unit #3 Period: 11/14-21/77 . . . . . . . . . . . . . . . . . . 190 6.8 Average Secondary Chemistry - Unit #4 Period: 1/18-24/78 . . . . . . . . . . . . . . . . . . 192 6.9 Average Secondary Chemistry - Unit #4 Period: 1/26/78 . . . . . . . . . . . . . . . . . . . . 194 6.10 Average Secondary Chemistry - Unit #4 Period: 2/3-9/78 ...................196 4 6.11 Primary-to-Secondary Leak Rate . . . . . . . . . . . . . 200 6.12 Isotopic Age Calculations ...............202
- 6.13 Blowdown Flash Tank Iodine Activity . . . . . . . . . . 204 6.14 Blowdown Flash Tank Inlet . . . . . . . . . . . . . . . 205 6.15 Flow Through Blowdown Flash Tank . . . . . . . . . . . . 206 6.16 Blowdown Flash Tank Vent . . . . . . . . . . . . . . . . 207 6.17 Average Steam Generator DF's . . . . . . . . . . . . . . 212 6.18 Fraction of Feed Activity Coming from HP Drains l
for Unit #4 . . . . . . . . . . . . . . . . . . . . . . 214 6.19 Air Ejector Iodine Species Sample . . . . . . . . . . . 215 7.1 131I Analyses of Waste Gas Decay Tanks . . . . . . . . . 220 7.2 Noble Gas Analysis Data of Waste Gas Decay Tanks . . . . 222 . 7.3 Turkey Point Watta Gas Tank Releases During In-Plant Study Period . . . . . . . . . . . . . . . . . . . . . . 224 xiv
LIST OF TABLES (cont'd) Page 7.4 Extrapolated Annual Radionuclide Releases from WGDT (Combined Refueling and Non-Refueling) . . . . . . . . . 226
; 7.5 Extrapolated Annual Radionuclide Releases from WGDT's (Refueling and Non-Refueling) . . . . . . . . . . . . . 227 7.6 Waste Gas Decay Tanks Iodine Species . . . . . . . . . . 229 7.7 Radionuclide Average Reactor Effective Coolant Leak
- Rates to the Containment Building . . . . . . . . . . . 233 i
7.8 Radionuclide Effective Leakage Rate to Containment Building, Unit #3 . . . . . . . . . . . . . . . . . . . 234 7.9 Radionuclide Effective Leakage Rate to Containment Building, Unit #4 . . . . . . . . . . . . . . . . . . . 235 7.10 Containment Effective Partition Factors . . . . . . . . 236 7.11 Unit #3 Containment Atmosphere Iodine Species Averages . 238 7.12 Unit'#4 Containment Atmosphere Iodine Species Averages . 239 7.13 Turkey Point Unit #4 Containment Purges and Vents . . . 241 7.14 Turkey Point Unit #4 Containment Purges and Vents . . . 242 i /.15 Extrapolated Annual Containment Building Radionuclide Releases . . . . . . . . . . . . . . . . . 244 7.16 Data for Unit #3 Containment Annual 131 1 Releases . . . 247 7.17 Data for Unit #4 Containment Annual 1311 Releases . . . 248 7.18 Data for Annual 3H Releases from Containment Purges . . 249 7.19 lC and Particulate Radionuclide Averages . . . . . . . 251 ! 3.1 Auxiliary Building Sampling Station Feeds with Design I D uc t F' is . . . . . . . . . . . . . . . . . . . . . . . 2 54 I 8.2 Ventilation Duct Flow Measurements . . . . . . . . . . . 257 8.3 Average 131I Reactor Coolant Concentrations . . . . . . 259 8.4 Main Stack 3H and 131I Normalized Release Rates and l Effective Partition Coefficients . . . . . . . . . . . . 260 l 1 xV i I. .. . ., ,. . , - _. . - _ , -- - - - .
LIST OF TABLES (cont'd) Pa9e 8.5 Normalized 1311 Release Rates . . . . . . . . . . . . . 261 8.6 Station #4 131I Airborne Activity Release Rates . . . . 263 8.7 Extrapolated Annual Stack and Auxiliary Building Particulate Releases for Gaseous Effluents . . . . . . . 267 8.8 Average Particulate Release Mass Balance . . . . . . . . 268 8.9 Extrapolated Annual Stack Releases of Gaseous 3H and 14C . 269 8.10 Containment PL~ge and WGDT Contributions to Annual Stack Release at 14C and 3H . . . . . . . . . . . . . . 269 4 8.11 Stack Gaseous 3H and 14C Species . . . . . . . . . . . . .,271 8.12 Average Fractional Percentage for 131I Species . . . . . 272 8.13 Main Stack Effective Reactor Coolant Leak Rates . . . . 274 l 8.14 Stack Release Rates for Selected Beta-Emitting Nuclides . 276 8.15 Stack 133Xe Releases (4/29-6/7/78) . . . . . . . . . . . 277 A.1 Comparisons of Results from Replicate Samples of Reactor Coolant Analyzed by FPL and INEL . . . . . . . . A-4 A.2 Results of Split and Replicate Samples Analyzed by d 00E-RESL and INEL ................... A-5 B.1 Radionuclide Concentrations in Reactor Coolant - Unit #3
; Power Operations Prior to Refueling .......... B-2 I B.2 Radionuclide Concentrations in Reactor Coolant - Unit #3 During Refueling . . . . . . . . . . . . . . . . . . . . B-4 ;
1 B.3 Radionuclide Concentrations in Reactor Coolant - Unit #3 Power Ope ations Af ter Refueling . . . . . . . . . . . . B-6 B.4 Radionuclide Concentrations in Reactor Coolant - Unit #4 Power Operations . . . . . . . . . . . . . . . . . . . . B-14 i B.5 Beta-Only-Emitting Radionuclide Concentrations in Reactor Coolant. Unit #3 - During Refueling .......... B-26 B.6 Beta-0nly-Emitting Radionuclide Concentrations in Reactor l Coolant, Unit r3 - Power Operations After Refueling .. B-27 1 xV1 l l
. .~,-
i i LIST OF TABLES (cont'd) , Page 8.7 Beta-Only Emitting Radionuclide Concentrations in Reactor Coolant, Unit #4 - Power Operations .......... B-28 B.8 3H and 131I Concentrations in Reactor Coolant, Unit #3 FPL Measurements ................... B-29 B.9 3H and 1311 Concentrations in Reactor Coolant, Unit #4
! FPL Measurements .................... B-30 B.10 Radionuclide Concentrations in Reactor Coolant 1/24-26/78 Shutdcwn of Unit #4 ............. B-31 B.11 Radionuclide Concentrations in Reactor Coolant 5/19-22/78 Shutdown, Startup of Unit #3 . . . . . . . . . B-35 B.12 Radionuclide Concentrations in Unit #3 CVCS and Demineralizer DF's ................... B-53 B.13 Radionuclide Concentrations in Unit #4 CVCS and Deminerali7er DF's ........... . . . . . . . B-65 B.14 Radionuclide Concentrations in Base Caticn Demi neral izer C Inl et . . . . . . . . . . . . . . . . . . B-85 B.15 Radionuclide Concentrations in Base Cation Demi neral i zer A Inl et . . . . . . . . . . . . . . . . . B-86 B.16 Radionuclide Concentrations in BAE Feed . . . . . . . . . B-87 B.17 Radionuclide Concentrations in BAE Distillate . . . . . . B-89 B.18 Radionuclide Concentrations in BAE Bottoms . . . . . . . B-91 B.19 Radionuclide Concentrations in BAE Condensate Demineralizer Effluent ................. B-93 8.20 Radionuclide Concentrations in BAE Condensate i Demineralizer Filter Effluent . . . . . . . . . . . . . . B-95 B.21 Radionuclide Cu wentrations in Radwaste Evaporator feed . . . . . . . . . . . . . . . . . . . . . B-97 B.22 Radionuclide Concentrations in Radwaste Evaporator Distillate .......................
s B-100 8.23 Radionuclide Concentrations in Radwaste Evaporator
. Bottoms . . . . . . . . . . . . . . . . . . . . . . . . . B-103 4
i xvii
i i
^
l l LIST OF TABLES (cont'd) : Page B.24 Radionu:lide Concentrations in Radwaste Evaporator Distillate and Bottoms, Feed Shut Off and Bottoms Concentrating During Sample Period . . . . . . . . . . . B-106 L.25 Radionuclide Concentrations in Inlet and Outlet for l Radwaste Condensate Demineralizer . . . . . . . . . . . . B-107 B.26 Radionuclide Concentrations in Waste Holdup Tank No. 2 . . . . . . . . . . . . . . . . . . . . . . . B-109 B ?7 Radionuclide Concentrations in Radwaste Building Monitor Tank A . . . . . . . . . . . . . . . . . . . . . B-lll
! B.28 Radionuclide Concentrations in Radwaste Building Monitor Tanks B and C . . . . . . . . . . . . . . . . . . B-ll2 B.29 Radionuclide Concentrations in Auxiliary Building l Monitor Tanks . . . . . . . . . . . . . . . . . . . . . . B-ll3
, B.30 Concentrations of Beta Only Emitting Radionuclide; in Waste Holdup and Monitor Tanks . . . . . . . . . . . . . B-ll4
! B.31 Unit #3 Spent Fuel Pit Demineralizer Inlet and Outlet Radionuclide Concentrations . . . . . . . . . . . . . . . B-ll6 )
8.32 Beta-Only-Emitting Radionuclides for Spent Fuel Pit and Associated Waters . . . . . . . . . . . . . . . . . . B-ll8 I t B.33 Unit #3 Spent Ft 1 Pit and Associated Water Samples . . . B-120 B.34 Radionuclide Concentrations in Secondary Waters l Unit #3, 11/8-9/77 . . . . . . . . . . . . . . . . . . . B-125 f B.35 Radionuclide Concentrations in Secondary Waters Unit #3, 11/14/77 . . . . . . . . . . . . . . . . . . . . B-130 B.36 Radionuclide Concentrations in Secondary Waters Unit #3, 11/16/77 . . . . . . . . . . . . . . . . . . . . B-134 , 1 B.37 Radionuclide Concentrations in Secondary Waters Unit #3, 11/17-18/77 . . . . . . . . . . . . . . . . . . B-135 B.38 Radionuclide Concentrations in Secondary Waters Unit #3, 11/21/77 . . . . . . . . . . . . . . . . . . . . B-141 B.39 Radionuclide Concentrations in Secondary Waters Unit #3, 11/22/77 . . . . . . . . . . . . . . . . . . . . B-145 xviii 1
xix l 502-B . . . . . . . . . . . . . . . . . . . . . snoitartnecnoC ; edilcunoidaR enrobriA knaT yaceD sag etsaW 65.B 402-B . . . . selpmaS rotcejE riA fo sisylanA H3-C41 fo stluseR 55.B 202-B . . . . . . selpmaS yradnoceS fo sisylanA ateB fo stluseR 45.B i 002-B . . . . . . . . . . . . . . . . . . . . 87/42/1 ,4# tinU selpmaS rotcejE riA ni seicepS enidoI 35.B 891-B . . . . . . . . . . 87/02/1 ,4# tinU ,selpmaS knaT hsalF nwodwolB dna rotcejE riA ni snoitartnecnoC edilcunoidaR 25.B 691-B . . . . . . . . . . . . . . . . . . . . 87/81/1 ,4# tinU selpmaS rotcejE riA ni snoitartnecnoC edilcunoidaR 15.B 491-B . . . . . . . . . . . . . . . . . . 87/9-7/2 ,4# tinU sretaW yradnoceS ni snoitartnecnoC edilcunoidaR 05.B 091-B . . . . . . . . . . . . . . . . . . . . 87/3/2 ,4# t inU sretaW yradnoceS ni snoitartnecnoC edilcunoidaR 94.B 281-B . . . . . . . . . . . . . . . . . . . . ?7/62/1 ,4# tinU sretaW yradnocul ni snoitartnecnoC edilcunoidaR 84 8 671-B . . . . . . . . . .. .......... 87/42/1 ,4# tinU sretaW yradnoceS ni snoitartnecnoC edilcunoidaR 74.B 271-B . . . . . . . . . . . . . . . . . . . . 87/32/1 ,4# tinU sretaW yradnoceS ni snoitartnecnoC edilcunoidaR 64.B 861-B . . . . . . . . . . . . . . . . . . . . 87/22/1 ,4# tinU sretaW yradnoceS ni snoitartnecnoC edilcunoidaR 54.B 461-B . . . . . . . . . . . . . . . . . . . . 87/02/1 ,4# tinU sretaW yradnoceS ni snoitartnecnoC edilcunoidaR 44 851-B . . . . . . . . . . . . . . . . . . . . 87/91/1 ,4# tinU sretaW yradnoceS ni snoitartnecnoC edilcunoidaR 34.B 251-B . . . . . . . . . . . . . . . . . . . . 87/81/1 ,4# tinU sretaW yradnoceS ni snoitartnecnoC edilcunoidaR 24.B l 841-B . . . . . . . . . . . . . . . . . . . 77/41/11 ,3# t inU selpmaS rotcejE riA ni snoitartnecnoC edilcunoidaR 14.B o 641-B . . . . . . . . . . . . . . . . . . . . 77/9/11 ,3# tinU l selpmaS rotcejE riA ni snoitartnecnoC edilcunoidaR 04.B egaP
)d'tnoc( SELBAT FO TSIL
l LIST OF TABLES (cont'd) Page ; B.57 Iodine Species - Measurements of Waste Gas Processing f System . . . . . . . . . . . . . . . . . . . . . . . . . B-209 B.58 Waste Gas Decay Tank 14C and 3H Concentrations . . . . . B-210 B.59 linit #3 Containment Airborne Radionuclide Lancentrations Before Refueling . . . . . . . . . . . . . B-211 B.60 Unit #3 Containment Airborne Radionuclide Concentrations After Refueling . . . . . . . . . . . . . B-213 l B.61 Iodine Species Measurements of Unit #3 Containment Atmosphere . . . . . . . . . . . . . . . . . B-215 B.62 Unit #3 Containment Airborne 14C and 3H Concentrations . B-216 B.63 Unit #4 Containment Radionuclide Concentrations . . . . . B-217 B.64 Iodine Species Measurements of Unit #4 Containment Atmosphere . . . . . . . . . . . . . . . . . B-219 B.65 Unit #4 Containment Airborne 14C and 3H Concentrations . B-220 B.66 Ventilation Airborne Radionuclide Release Rates Sampl e Station #1 . . . . . . . . . . . . . . . . . . . . B-221 B.67 Ventilation Airborne Radionuclide Release Rates Sampl e Stati on #2 . . . . . . . . . . . . . . . . . . . . B-226 i B.68 Ventilation Airborne Radionuclide Release Rates Sampl e Stati on #3 . . . . . . . . . . . . . . . . . . . . B-231
; B.69 Ventilation Airborne Radionuclide Release Rates Sampl e Stati on #4 . . . . . . . . . . . . . . . . . . . . B-236 B.70 Ventilation Airborne Radionuclide Release Rates i
Sample Station #4A . . . . . . . . . . . . . . . . . . . B-241 1 B.71 Ventilation Airborne Radionuclide Release Rates Sampl e St ation #5 . . . . . . . . . . . . . . . . . . . . B-245 'l B.72 Ventilation Airborne Radionuclide Release Rates Unit #3 Fuel Pit Area Duct . . . . . . . . . . . . . . . B-250 B.73 Ventilation Airborne Radionuclide Release Rates Uni t #3 Sampl e Room . . . . . . . . . . . . . . . . . . . B-255 B.74 Ventilation Airborne Radionuclide Release Rates Gas Stripper Room . . . . . . . . . . . . . . . . . . . . B-256 XX
LIST OF TABLES (cont'd) Page B.75 Ventilation Airborne Radionuclide Release Rates Main Stack . . . . . . . . . . . . . . . . . . . . . . . B-257 B.76 1311 Species Data, Sample Station #1 . . . . . . . . . . B-262 B.77 131 1 Species Data, Sample Station #2 . . . . . . . . . . B-263 B.78 131 I Species Data, Sample Station #3 . . . . . . . . . . B-264 B.79 181 I Species Data, Sam 91e Station #4 . . . . . . . . . . B-265 B.80 131 1 Species Data, Sample Station #4A . . . . . . . . . . B-266 B.81 131 I Species Data, Sample Scotion #5 . . . . . . . . . . B-267 B.82 131 I Species Data, Unit #3 Fuel Pit Area Duct . . . . . . B-268 B.83 131 I Species Data, Unit #3 Sample Room . . . . . . . . B-269 B.84 131 I Species Data, Gas Stripper Room . . . . . . - . . . B-269 B.85 131 I Species Data, Main Stack . . . . . . . . . . . . . . B-270 B.86 Auxiliary Building 131 I Sources . . . . . . . . . . . . . B-271 XXi
LIST OF FIGURES Page 1.1 Diagram of Liquid and Solid Systems . . . . . . . . . . . 4 1.2 utagram of Gaseous Waste System . . . . . . . . . . . . . 5 2.1 Unit #3 Power Level and 131I Concentrations in Reactor Coolant . . . . . . . . ............. 8 2.2 Unit #4 Power Level and 1311 Concentrations in Reactor Coolant . . . . . . . . . . . . . . . . . . . . . 9 2.3 Simplified Block Diagram of Reactor Coolant, Letdown, and Fuel Pit Cleanup Systems .............. 10 2.4 Simplified Block Diagram of Boric Acid Recovery and Radwaste Systems .................... 11 2.5 60Co Decontamination Factors for Mixed-Bed Demineralizers ..................... 21 2.6 131I Decontamination Factors for Evaporators ...... 22 2.7 137 Cs Decontamination Factors for Evaporators . . . . . . 23 2.8 60 Co Decontamination Factors for Evaporators 24 2.9 Auxiliary Building and Spent Fuel Pit #3 Ventilation System ......................... 26 3.1 Reactor Coolant System, Units #3 and #4 . . . . . . . . . 39 3.2 Chemical and Volume Control System, Turkey Point Plant Units #3 and #4 . . . . . . . . . . . . . . . . . . 41 3.3 Shutdown Spike - Unit #4, 1/25/78 . . . . . . . . . . . . 64 3.4 Shutdown Spike - Unit #3, 5/19-20/78 .......... 66 3.5 Startup Spike - Unit #3, 5/21-22/78 . . . . . . . . . . . 67 3.6 Unit #3 Letdown Operational Information . . . . . . . . . 71 3.7 131 I and 133I Inlet Concentrations and DF's for Unit #3 CVCS Mixed Bed B Demineralizer ............. 75 3.8 134Cs and 137Cs Inlet Concentrations and DF's for Unit #3 CVCS Mixed Bed B Demineralizer ............. 76 i xxiii
l LIST OF FIGURES (cont'd) Page 3.9 seCo and soCo Inlet Concentrations and DF's for Unit #3 CVCS Mixed-Bed B Demineralizer ............. 77
- 3.10 95Zr and 124Sb Inlet Concentrations and DF's for Unit #3 CVCS Mixed-Bed B Demineralizer ..... ....... 78 3.11 Unit #4 Letdown Operational Information . . . . . . . . . 82 1 1
3.12 1311 Inlet Concentrations and DF's for Unit #4 CVCS I
- Mixed-Bed A Demineralizer . . . . . . . . . . . . . . . . 88 I
3.13 134Cs and 137Cs Inlet Concentrations and DF's for Unit #4 CVCS Mixed-Bed A Demineralizer ............. 89 3.14 60Co Inlet Concentrations and DF's for Unit #4 CVCS Mixed-Bed A Demineralizer ............. 90 3.15 54Mn and 59Fe Inlet Concentrations and DF's for Unit #4 CVCS Mixed-Bed A Demineralizer ............. 91 4.1 Diagram of Boric Acid Recovery System . . . . . . . . . . 96 4.2 Diagram of Liquid Radwaste System . . . . . . . . . . . . 99 4.3 Radionuclide Concentrations.in Inlet to Boric Acid Recovery System and Demineralizer DF's . . . . . . . . 104 131 1 Concentrations in BAE Feed, Distillate, and 4.4 Bottoms and BAE DF's . . . . . . . . . . . . . . . . . 1 06 i !~ 4.5 seCo Concentrations in BAE Feed, Distillate, and Bottoms and BAE DF's . . . ... . . . . . . . . . . . . 107 i . 4.6 1311 Corcentration in Inlet to BAE Condensate Demineralizer and Demineralizer DF . . . . . . . . . . 109 4.7 Boric Acid Recovery System Average Concentrations and "Best Value" DF's for Operation with Base Cation Demineralizer C in Service . . . . . . . . . . . . . . 126 4.8 Boric Acid Recovery System. Average Concentrations and "Best Value" DF's for Operation with Base Cation Demineralizer A in Service . . . . . . . . . . . . . . 127 4.9 Boric Acid Evaporator Average Concentrations and Ratios ! of Bottoms to Distillate Activities . . . . . . . . . . 128 l 1 I xxiv
i i , LIST OF FIGURES (cont'd) i Page i 4.10 Correlation Between 131I DF and Inlet Concentration i for Base Cation Demineralizer A . . . . . . . . . . . . . 130 i 4.11 Correlation Between 58Co and 60Co DF and Feed i Concentration for Boric' Acid Evaporator . . . . . . . . . 1 31 4.12 Correlation Between 131I Concentration in BAE Distillate ) and Bottoms . . . . . . . . . . . . . . . . . . . . . . . 132 i 4.13 Radwaste Evaporator Operating Parameters ........ 135 i i 4.14 1311 Concentrations and DF's for Radwaste Evaporator . . 141 4.15 54 2 Mn Concentrations and DF's for Radwaste Evaporator . . 142 4 4.16 s0 Co Concentrations and DF's for Radwaste Evaporator . . 143 4.17 Correlation Between 60Co Concentration in Distillate and Bottoms, Radwaste Evaporator ............ 144 4.18 Correlation Between 131I Concentration in Distillate and Bottoms, Radwaste Evaporator ............ 145 i
! 4.19 Correlation Between 54Mn Concentration in Distillate j and Bottoms, Radwaste Evaporator ............ 146
- 4.20 Correlation Between 60Co Concentration in Distillate and Bottoms, Radwaste Evaporator ............ 147 5.1 ochematic Diagram of Unit #3 Spent Fuel Pit System . . . 161 6.1 Diagram of Secondary System . . . . . . . . . . . . . . . 175 6.2 Schematic Diagram of Secondary System . . . . . . . . . . 176 6.3 Ma i n Steam Sampl e P rob . . . . . . . . . . . . . . . . . . 210 l 7.1 Schematic Diagram of Waste Gas Processing System . . . . 21 8 7.2 131 I Fractional Percent Organic lodines as a Function of Time After Last Purge ................ 240 i 8.1 . Auxiliary Building Ventilation System . . . . . . . . . 253 l 8.2 Auxiliary Building 131 1 Release Rates . . . . . . . . . 265 4
- xxv
., ,n- ,, ,. y ....,-v---.-- . -
, , - - . -n _- ,,__.,,.,,-n.- . ~ - -y . - , ,,-y , ,
I LIST OF FIGURES (cont'd) Page B.1 Reactor Cool ant System P&ID . . . . . . . . . . . . . . . B-272 B.2 Chemical and Volume Control System P&ID . . . . . . . . . B-273 B.3 Chemical and Volume Control System P&ID - Boric Acid { l and Demineralizer Systems . . . . . . . . . . . . . . . . B-274 B.4 Chemical and Volume Control System P&ID - Boron Recycle System . . . . . . . . . . . . . . . . . . . . .B-275 B.5 Main Steam System P&ID . . . . . . . . . . . . . . . . .B-276 I B.6 Extraction and Auxiliary Steam Systems P&ID . . . . . . . B-277 B.7 Condensate System P&ID . . . . . . . . . . . . . . . . .B-278 B.8 Feedwater System P&ID . . . . . . . . . . . . . . . . . . B-279 B.9 Condensate and Feedwater Auxiliary Systems P&ID . . . . . B-280 B.10 Radwaste Facility Waste Disposal - Liquid System P&ID . . B-281 B.ll Radwaste Facility Waste Disposal - Liquid System P&ID . . B-282 B.12 Waste Disposal Systems P&ID - Liquids . . . . . . . . . . B-283 B.13 Waste Disposal System P&ID - Gas . . . . . . . . . . . .B-284 B.14 Sampling System and Spent Fuel Pit Cooling System P&ID .B-285 B.15 Containment Ventilation System P&ID . . . . . . . . . . .B-286 B.16 Auxiliary Building Ventilation System P&ID . . . . . . .B-287 B.17 Levels for Holdup Tanks A, B, and C . . . . . . . . . .B-288 .l B.18 Levels for Waste Holdup Tank #1 and Chemistry I Drain Tank . . . . . . . . . . . . . . . . . . . . . . .B-289 B.19 Levels for Laundry Drain Tanks A and B . . . . . . . . .B-290 B.20 Levels for Waste Condensate Tanks A and B . . . . . . . . B-291 B.21 Levels for Monitor Tanks A and B in Auxiliary Building . . . . . . . . . . . . . . . . . . . . . . . .B-292 l xxvi I l i
ACKNOWLEDGEMENTS The valuable assistance of Florida Power and Light persennel made the measurement program at Turkey Point possible. H. Yaeger and his staff, especially J. Wade, L. Cushen, J. Strong, J. Puckett, and E. LaPierre contributed greatly to the success of the measurements. Also, from the Florida Power and Light corporate offices, R. Uhrig and G. Ledicotte provided valuable assistance. The financial support and direction of the U. S. Nuclear Regulatory Commission provided by C. Bartlett and D. Solberg of RES and J. Collins, R. Bangart and J. Lee of ETSB made the measurement program at Turkey Point possible. The assistance of C. Pelletier of Science Applications, Inc., in data interpretatf or, and report preparation was of great value. The contributions to the program made by H. Cadwell and A. Marley in sample and data collection and W. Killian and R. Kynaston in software and hardware support contributed to the success of the program. The typing and editing services provided by J. Robinson made preparation of this report possible. f I xxvii
- 1. INTRODUCTION 1.1 Objectives of the In-Plant Measurement Program The primary objective of the in-plan; =arce tem measurement study at operating pressurized water reactors (PWR's) is to provide the Nuclear Regulatory Commission (NRC) with operational data that can be used in evaluation of plant designs for liquid and gaseous w3ste treatment systems. This evaluation requires a knowledge of the sources and quantities of radioactive waste materials generated at a nuclear power reactor during nomal operation including anticipated operational occurrences, how these sources vary with plant design, the radioisotope inventory in plant systems, how radioactive materials move through plant systems, and radioactive waste treatment system performance.
Specific objectives of the in-plant measurement study are:
- 1. Obtain data on the inventory of radioisotopes present (i.e. , locations, concentrations, etc.) in operating reactbr plant systens during nomal operation and anticipated operational occurrences.
- 2. Study radioactive waste treatment system performance and detemine decontamination factors (DF's) for demineralizers, evaporators, filters, and gaseous cleanup systems.
- 3. Detect and measure primary-to-secondary leaks and determine isotopic partition factors for steam generators and main condenser.
- 4. Obtain data on radioisotope concentrations in fuel pool waters and perform a tritium balance during refueling.
- 5. Detemine the releases of radioactive materials in the gaseous and liquid effluents.
- 6. Estimate annual release of airborne activity from the auxiliary building ventilation system, process gas system, and containment buildings.
- 7. Provide additional source term information so that the parameters used in calculational models (1) can be updated as necessary.
Measurements are to be made during the three stages of plant operation (i.e., power generation prior to refueling, during refueling operations, and power generation following refueling) so that the data can be used to estimate equipment performance and radioactivity releases over the lifetime of a nuclear power plant.
1 The In-Plant Source Term Measurement Program is being carried out by the Idaho National Engineering Laboratory (INEL) and is a joint effort involving EG&G Idaho, Inc., and Allied Chemical Corp. In order to provide a data base for currently operating PWR's, a total of 6
- PWR's will be studied, 2 from each of the major vendors (Westinghouse, Combustion Engineering, and Babcock & Wilcox). In-plant measurements were initiated during the summer of 1976. During 1976 and 1977 measurements were made at the Fort Calhoun Station, Blair, Nebraska (operated by Omaha Public Power District) and at the Zion Station, Zion, Illinois (operated by Commonwealth Edison Co.). Results of these measurements are reported in references 2 and 3. This is a report on the results of measurements at Units #3 and #4 of the Turkey Point Power Station.
1.2 Turkey Point 1.2.1 In-Plant Measurements at Turkey Point The measurement program at Turkey Point was initiated in November, 1977. First, sample points and locations in the liquid and gaseous process streams were selected. This was accomplished by examining the piping and instrument diagrams (P&ID's) to determine where samples should be taken, discussing the proposed sample points with plant personnel, inspecting the actual systems to verify the efficacy of the sample points and locations. Results were used to generate a measurement plan for the specific studies to be made at Turkey Point. The NRC Mobile Laboratory was then moved to Turkey Point on 11/1/77. Actual in-plant measurements began on 11/8/77. In-plant measurements at Turkey Point spanned the period 11/8/77 to 6/1/78. Samples from both liquid and gaseous process streams were collected and analyzed using the procedures described in reference 4.
~
During this 7-month period, Unit #3 was down for refueling from 11/24/77 to 2/17/78, and Unit #4 was down for steam generator repairs from 2/14/78 to 3/9/78. t One of the main objectives at Turkey Point was to study primary-to-secondary leaks if they occurred and to detennine partition factors in steam generators and attempt an iodine balance around the secondary i system. The opportunity to study primary-to-secondary leaks occurred during the in-plant measurement study at Turkey Point. During November, 1977 steam generator C on Unit #3 had a primary-to-secondary leak and steam generator A on Unit #4 developed a leak during January-February, 1978. 1.2.2 Description of Turkey Point The Turkey Point Generating Station, operated by Florida Power and Light (FPL), is located on the western shore of Biscayne Bay, about 25 miles south of Miami, Florida. Two gas- and oil-fired generating plants, Turkey Point Units #1 and #2, and two nuclear power plants, Turkey Point Units #3 and #4, share the site. Units #3 and #4 are identical 2
pressurized light water reactors supplied by Westinghouse. Each unit has a generating capacity of 2200 MWt and a gross electrical power output of 760 MWe. Unit #3 reached initial criticality in 10/72 and began commercial operations in 12/72 while Unit #4 reached criticality in 6/73 and began commercial operations in 9/73.
.Each nuclear unit consists of a pressurized water reactor, reactor coolant system, secondary system, spent fuel storage pool, and associated auxiliary fluid systems. The reactor coolant system has three coolant loops, each with a vertical U-tube steam generator with integral moisture separaters. The auxiliary systems are used to charge the reactor coolant system, add makeup water, purify reactor coolant water, provide chemicals for corrosion inhibition and reactivity control, cool system components and the spent fuel storage pool, remove residual heat when the reactor is shut down, and provide for emergency coolant injection.
Some components of the auxiliary and waste treatment systems are shared by Units #3 and #4. These shared systems and components include the holdup tanks (used for reactor letdown solution), boric acid tanks, primary water storage tanks, refueling water storage tanks, base-cation demineralizers (FPL nomenclature for cation demineralizers used to remove pH control chemicals upstream of the boric acid evaporator), boric acid evaporator and condensate demineralizer, waste holdup tank, radwaste evaporator and condensate demineralizer, evaporator concentrates holding tank, and monitor tanks. Figures 1.1 and 1.2 show simplified schematic diagrams of the liquid and solid systems and the gaseous waste treatment system, respectively, at Turkey Point. The Piping and Instrument Diagrams (P&ID), which contain more details for each system, can be found in Appendix B, Figures B.1-B.16. For purpost.s of measurement, Turkey Point was divided into two major systems - liquid and gas. The liquid system was subdivided into i
, six basic subsystems - reactor coolant, secondary, letdown or chemical and volume control system (CVCS), boric acid recovery, liquid radwaste, and spent fuel pit cleanup. The gas system was subdivided into three basic subsystems - auxiliary building ventilation, process gas, and containment. Each of these subsystems and the data obtained are discussed in detail in the following sections of this report. Sample and data handling procedures are discussed in Appendix A. Appendix B contains the measured data.
1.2.3 Plant Data Wherever possible, plant data were collected to supplement i the data obtained during the measurements and to help interpret the measurements. Plant operational data used to characterize samples included the control room legs (each unit), the auxiliary building operator log, the radwaste building operator log, the daily water reports (both primary and secondary), plus information obtained in discussions with plant personnel. 4 3
Primary Primary I ( 1 } Steam (
' loop
["a it #3 g U"at e g#4 ( loop
) Steam (
generator - j condensers
)
Condensers - -- generator JL if JL if JL Jk Blowdown CVCS #4 Blowdown CVCS #3 if V M Makeup water d Blowdown and chemicals Blowdov n Vent to flash 4 flash ->-Vent atmosphere to atmospher+e tank tank Reactor Reactor e coolant coolant drain tank drain tank m I f Chemical laboratory 4 l1r Containment sumps
. Waste Boric acid
- Spent fuel m Holdup holdup & Floor drains pits -
tanks tanks tank Laundry JL 3 Shower 7 Evaporatol Evaporator Radwaste Boric acid Concentrates 2 evaporator r co holding tank y tem bottoms
' bottoms system niak Intake canal f Solid waste drumming facility lf Shipment II off-site
' Discharge
' tanks canal Discharge canal gg Figure 1.1 Diagram of Liquid and Solid Systems
Overall gaseous waste system Unit #3 Unit #4 4 i x i \ Vent 4 & 4 & Turbines Turbines If Il rs en rs e> Condenser Condenser Primary- - coolant- {- Primary-} coolant - { Exhaust j gEnhaust j j Exhaust A Exhaust Blowdown Steam Jet Blowdown Steam jet flash airejector Mash airejector tank g and gland tank and gland seal exhaust v seal exhaust 35,000 35.000 D crm cfm D ' 35.000 cfm Roughing filter Unit #3 Unit #4 - containment containment
- I l Roughing
- To CVCS l f'Her _
35 000 cfm holdup tanks for reuse ; 525 cu ft ) A f Gas decay CVCS. Waste gas tank (6) holdup _ compressors inleakage HEPA tanks
- 13.500 cfm
_ filter
-> ~
g , - Auxiliary g, _ building -- Prefilters air - 13.5. cfm ventilation system - Roughing 40.000 cfm Roughing filters Roughing
-HEPA x aust filter 1000 hIterS HEPA fjners cfm 1i HEPA
~g, iter 1000 cfm
~""
Unit #3 - ~9,ner _ Unit #4
'" ' a
a.
- + +
a:a 20 c,m - a aaa Prefitter cfm 20.0 _ _ C7.-* 20.0 2000 _ Prefilter cfm 4 cfm 7500 cfm I inieakage inisakage New rad waste building 7500 cfm
'CVCS - chemical and volume control system
,%gt...,a 73a l
Figure 1.2 Diagram of Gaseous Radwaste System 1 5 l
o Information obtained from the control room log includes power level, reactor coolant flow, steam generator flow and levei, reactor coolant temperature and pressure, and rod position. This infonnation l was tabulated either hourly or every 4 hours throughout the day. Plots
; of power level for Units #3 and #4 during the period 11/1/77 to 6/1/78 l are shown in Figures 2.1 and 2.2, respectively.
Information obtained from the auxiliary building operator log and the radwaste building operator log included tank levels, pressure drop
- across the letdown demineralizers, spent fuel pit level and water i temperatu:e, and waste gas decay tank pressure. This information was i tabulated avery 4 hours throughout the day. Tank level information pertinent to source tenn measurements is shown in Appendix B, Figures
. B.17 to B.21. i Also included in the operators' logs were times when monitor tanks l 1 were filling or were on recirculation prior to discharge, times when a l j reactor coolant drain tank was transferred to a containment sump, timos l , when evaporator bottoms were transferred and where they were transferred, j times when liquids were transferred to holdup tanks, operational status of the evaporators and spent fuel pit demineralizers, and miscellaneous j status information concerning systems and components. The daily water reports for the primary system provided information such as letdown flow rate, pH, conjuctivity, and boron and lithium i concentrations for both Units #3 and #4. The daily water reports for i the secondary system contained t'tal blowdown rates, pH, and conductivity for steam generator water together aith pH, conductivity, and chromate and phosphate concentrations in ti.e makeup water. ! Results of radiochemical analyses of reactor coolant samples perfortned by plant personnel were also obtained. Tables B.8 and B.9 in Appendix B
- list 1311 and 3H concentrations in Unit #3 and #4 reactor coolant as measured by FPL. Since the results of replicate samples indicated good agreement between FPL and INEL (see Appendix A), the 1311 concentrations i were used to supplement INEL data. The 1311 concentrations shown in Figures 2.1 and 2.2 include both INEL and FPL data.
i l I 6 4
I i f i i i 2.
SUMMARY
AND CONCLUSIONS i 2.1 General Plant Operation During Ir.-Plant Measurements Measurements were conducted at Turkey Point from 11/1/77 to 6/1/78. During this 7-month period, Unit #3 was down for refaling from 11/24/77 to 2/17/78. In addition, outages of shorter duration occurred during 3/21-25/78 and on 4/20/78 and 5/11/78, and during 5/19-22/78. At the 4 beginning of the measurement period, Unit #4 was down for steam generator l repairs. Power operations resumed on 11/12/77. This unit was down again for steam generator repairs from 2/14-3/9/78. Shorter outages occurred on 12/9/77, 12/17/77, 12/26-27/77, and 1/25/78. Figures 2.1 and 2.2 show the power levels for Units #3 and #4, respectively, during the measurement period. l 2.2 Liquid Systems i 2.2.1 Description of Liquid Systems I Figures 2.3 and 2.4 show simplified block diagrams of the l liquid systems at Turkey Point. For measurement purposes, the liquid l syste;;. has been divided into six basic subsystems: ! 1. reactor coolant
- 2. secondary
,' 3. letdown or chemical and volune control system (CVCS)
- 4. boric acid recovery
- 5. liquid radwaste
- 6. spent fuel pit cleanup.
The systems are, however, interrelated. Reactor coolant is cleaned up
- by the letdown demineralizers and is then returned to the core. Reactor coolant can also be diverted to the reactor coolant drain tank and to the holdup tanks and the boric acid recovery system (boric acid evaporator and condensate demineralizer). Radwaste water is drained into the waste holdup tank and from there to the radwaste evaporator and condensate
! demineralizer. Effluent from both condensate demineralizers is routed to ! the monitor tanks where it is sampled prior to release from the plant. Spent fuel pit water is treated by the spent fuel pit demineralizer. 2.2.2 Reactor Coolant Reactor coolant samples were obtained from Unit #3 during the period 11/9/77 to 6/1/78 and from Unit #4 during 12/2/77 to 5/23/78. For Unit #3, this encompassed the three stages of operation - power generation prior to refueling, refueling operations, and power generation following refueling. Unit #4 was in power generation during the in-plant measure-ment period. l Figures 2.1 and 2.2 show the power level and 131I concentration in reactor coolant for Units #3 and #4, respectively, for the period 11/1/77 to 6/1/78. Data obtained by FPL are included in these plots. It should 7
Fioure 2.1 Unit #3 Power Level and 1311 Concentrations in Peactor Coolant l= 3 i N i i i i i :
#3 reactor coolant 131; -
_ e INEL measurement _ _ a FPL measurement _ 10-1 y
,s f
il a A
=
E _ , g I 1 g as b ' 's ei f, g f 'g i !
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> t : i
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1 g _ 1 i _- o \ 8 10-3 _ t 's -
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=
'r ^
F l ll E E y _ _ g _ _ 10 N 4/1 5/15 6/1 11/1 11/15 12/1 12/15 2/15 3/11 3/15 4/15 5/1 Year (1977-78) INEL-A-10 986
Finure 2.3 Simplified Block Diagram of Reactor Coolant, Let bwn, and Fuel Pit Cleanup Systems Steam gen O _ C ' t Non-regen. G 1 U g$erator ; _ heat exchanger e i NI l Reactor } I B core g L L Volume A A B A B Steam generator 1I ss ss A control V To 1F A tank Ls holdup '
. J L JL
" tanks _ Deborating Cation Mixed bed demins demin demins Heat _ exchanger Refueling water 4 storage tank j m _- _ _ _3 , _ - , 1f _ Spent I' fuel pit 1f Jk INEL-A-10 972 Y Demineralizer
Firiure 2.2 Unit #4 Power Level and 131I Concentrations in Peactor Coolant l=- 1 i i i i i i i i i i I:
~
#4 reactor coolant 131i =
e INEL measurment :
- 8 4 FPL measurement -
a'.As a c 10-1 g-E s \ se 18
- g. -
=
8 : i 1 a t s 8 i,
- e
,! \,
g
$s g\ s
,/ i. 5 k g
^ #
.b 10-2 =- k \a gg '. =M' \sv f V L 3 E i-
- g '4,,,/ . ,s -a--a---E t : :
g _ .. _ C - o o 0 10-3 g. - 10-4 ' ' ' ' ' ' ' ' ' ' ' ' ' ' 100 , , , ,, , , , , , , , , so Q-10 8 I I l ' ' ' ' ' ' ' ' ' ' 11/1 11/15 12/1 12/15 1/1 1/15 2/1 2/15 3/1 3/15 4/1 4/15 5/1 5/15 6/1 Year (1977-78) INEL-A-10 978
Fiaure 2.4 Simplified Block Dianran of Boric Acid Recoverv Ed Radwaste Systems Reactor coolant letdowns Flcac:or coolant drain tanks Spent fuel pits Concentrates holding tank u i fN _ Bonc acid _ A- B C A B C A a
. demineralizer M II II -
To
)[ d a d L 5 monitor Holdup tanks tanks Base-cation Filter demineralizers a
Laundry & hot shower tanks Chemical drain tank Containment sumps ][ X Reactor coolant drain tanks CVCS demineralizers Steam gen blowdown tanks Fuel transfer canals Floor drains Equipment drains I _ Radwaste _ g , (Waste (holdup tank)
] evaporator To L = = + monitor Condensate -
tanks demineralizer Filter INEL- A-10 973
4 be noted that the dashed lines are not intended to indicate the 131I i concentration between measurement points. Their purpose is only to aid I in detennining trends. The data in these figures indicate that in both units the 181I concentration spiked (i.e., increased dramatically) not i only upon reactor shutdown and startup, but each time the power was i altered. A small spike was even observed after the power was reduced in 3 Unit #3 from 100% to 90% on 11/19/77. This sensitivity to power change
; was not observed during measurements at Zion, where the reactor exhibited i definite -iodine spikes upon shutdown and startup, but only small changes ;
i in iodine concentration when only a power change occurred (3). i I In order to provide more information on the spiking phenomenon, reactor coolant samples were obtained on approximately an hourly basis during two hot shutdowns. On 1/25/78, Unit #4 underwent a hot shutdown that lasted almost 12 hours. Iodine, cesium, and barium and the crud-l associated radionuclides (e.g., Co, Zn) exhibited spiking during this l shutdown. The longer-lived radionuclides reached their maximum concentrations about 5 hours after poser reduction began (about 3 hours after 0% power was reached). Iodine-131,137Cs, and seCo peaked at concentrations 38, 4.5, and 31 times the pre-spike levels, respectively. Unit #3 underwent a hot shutdown on 5/19/78 that lasted until 5/22/78. The iodine, cesium, and cobalt radionuclides exhibited spiking upon shutdown. Concentration of the longer-lived f odines reached a maximum about 6 hours after power reduction began (about 2-1/2 hours after 0% power was reached). Iodine-131 peaked at a concentration 7.5 ! times its pre-spike level, while 137Cs and seCc peaked later and showed 4 smaller increases. Of the gaseous fission products, only 133Xe exhibited an increase in concentration after shutdown. Immediately after startup on 5/22/78 the cesiums and the longer-lived iodines exhibited spiking, but the magnitudes of the spikes were not as great as during shutdown (1311 reached a concentration 3 times its pre-spike level). The crud- ! associated radionuclides showed only slight indications of spiking. During the spiking study, it was found that after shutdown, the 1
- 132I concentration initially spiked, quickly returned to pre-spike levels, and then followed the 132Te half life. This indicates that the source
! of the 132I after the spike was 132Te. This same phenomenon was observed in studies made at another power reactor (5). Because 1:2Te was detected in the reactor coolant at only very low levels (insufficient to account for the 1321), the telluriu:n must be attached to the surfaces of the reactor internal structures. This can explain why tellurium is not normally detected in reactor coolant. I
- In addition to spikes due to power changes, a higher than nonnal iodine level of long duration occurred in Unit #4 reactor coolant during 4/9-14/78 (see Figure 2.2). This elevated level was caused by the
! letdown demineralizer for Unit #4 being bypassed. After the demineralizer 4 was put back into operation again, the iodine, cesium, and crud-associated radionuclide concentrations dropped, but never decreased to previous l 12
~ .-_ . _- - --- . _
i I levels. In addition, the iodine isotope ratios (ratios of 133, 135, 132, 134y with 1311 ) measured after the demineralizer had been put back into service showed a change (an increase of about 20%) from pre-bypass values. It is not known why these changes occurred. Mere bypassing of the letdown demineralizer would not be expected to cause this. Average radionuclide concentrations in Unit #3 and #4 reactor coolant were obtained for non-spiking periods of rower operation. These periods were 11/1-22/77 and 2/21-6/1/78 for Unit /3 (i.e., before and after refueling) and 11/12/77-3/28/78, 4/9-14/78, and 4/17-5/31/78 for Unit #4 (i.e,, before, during, and after the elevation due to bypass of the letdown demineralizer). These average concentrations are presented in Tables 3.2 and 3.4. General observations of these concentrations are as follows. The levels of the gaseous fission products were higher in Unit #3 reactor coolant than in Unit #4 reactor coolant. The 1311 concentration was higher in Unit #3 prior to refueling but after refueling it dropped to about the level measured in Unit #4. Concentrations of , the shorter-lived iodines were much higher (about an order of magnitude) in Unit #3 than in Unit #4. Unit #4, on the other hand, exhibited l higher concentrations for all crud-associated radionuclides except spCo. Except for 132I and 134 1, the radionuclide concentrations were lower in both units than the concentrations suggested by the ANSI N237-1976 standard. Iodina isotope ratios in reactor coolant indicated that in Unit #3 the release mechanism for fission product release to the reactor coolant was dominated by recoil. The majority of the fission products would seem to originate from tramp uranium. This tramp uranium was probably released during an earlier fuel failure and plated out on the fuel rods and on the reactor core internals. The iodine isotope ratios in Unit #4 reactor coolant indicated that the fission product release was dominated
, by diffusion. This means that most of the fission products in Unit #4 l reactor coolant ware released from small defects in the fuel cladding.
2.2.3 Chemical and Volume Control System Samples uf inlet water to and outlet water from the demineralizers in the chemical and volume control systems for both Units #3 and #4 were obtained in order to study the operation of the demineralizers and determine demntamination factors (DF's). Unit #3 CVCS was studied during the period 2/21-5/25/78 and Unit #4 CVCS during 11/30/77-5/23/78. l Demineralizer B in CVCS #3 contained relatively new resin during the measurement period (it had been in service for 25 days and had processed 7.5(3) bed volumes when measurements began).* Demineralizer A in CVCS
#4 was older, having been in service 180 days and having processed 6.3(4) bed volumes when measurements began. ,
*In this report, the number in parentheses following a number presents the power of ten multiplier, i.e., 3.97(4) = 3.97 x 104 13
The measurements carried out on the CVCS demineralizers indicate that the DF's for er sin radionuclides tend to be directly related to inlet concentratici 0F's for cesium and iodine were found to be highly correlated with int . concentration while DF's for certain other radionuclides showed a lesser degree d correlation. This was especially evident during spikes in the reactor coolant concentration. As will be shown below (sectio- '.7), the same phencmenon was observed for other demin-eralizers at .vKey Point as well as at Zion and Fort Calhoun. The data base is not yet sufficient to determine the exact dependence of DF on inlet concentration; therefore, DF values that best represent the DF's over the range of observed inlet concentrations were determined. These "best value" DF's were obtained for each demineralizer by detennining the average inlet and outlet concentrations and then using these average concen-trations to obtain the DF's. This tectinique allows the estimation of DF's for those radionuclides not detected with any regularity in either the inlet or outlet streams. When the data base is sufficient to quantify the functional dependence between DF and inlet concentration, this function will be used in place of "best value" DF's because it is more representative of actual plant performance. Table 2.1 lists the "best value" DF's for the radionuclides most frequently seen in CVCS demineralizer inlet and outlet samples. More detailed tabulations are contained in Tables 3.12 and 3.15. 2.2.4 Boric Acid Recovery and Radwaste Systems At Turkey PJnt, the boric acid recovery system and liquid radwaste system are similar (see Figure 2.4). Both are fed by holdup tanks - 3 holdup tanks feed the boric acid recovery system while a single waste holdup tank feeds the radwaste t-tem In the boric acid recovery system, liquid flows from the holdup tanks, through the base-cation demineralizer, evaporator, condensate demineralizer, and filter and finally to the monitor tanks. In the radwaste system, liquid flows from the waste holdup tank directly to the evaporator, then through the condensate demineralizer and filter to the monitor tanks. Except for the base-cation demineralizers, the two systems are identical. The boric acid recovery system normally processes reactor coolant quality water. This water is held for about 5-6 days which allows short-lived radionuclides to decay before processing. This water contains relatively long-lived radioisotopes and high baron concentrations. The radwaste system, on the other hand, has a shorter holdup time because the system has only one tank which can accept feed at the same time processing is occurring. Since, on occasion, water from the primary system obtained through the reactor coolant drain tank is processed, the radwaste system can contain more short-lived radionuclides and variable boron concentrations. The chemistry of this water is variable and sometimes contains high chromate concentrations. At times the radwaste system was used to process reactor coolant type water. In these cases, water from the reactor coolant drain tank was transferred to the containment sump and from there to the waste holdup tank. Also, during the early part of May, the boric acid recovery system was used to process water from spent fuel pit #3. For these operations, the cation resin in the base-cation demineralizer was replaced by a mixed-bed resin. 14
l TABLE 2.1 1
"BEST VALUE"tDECONTAtiINATION FACTORS FOR CVCS MIXED-BED DEMINERALIZERS "Best Value" Decontamination Factor CVCS #3 CVCS #4 Nuclide n
,em'n. B Demin. A 131I 2600 3600 133I 1900 1100 lasI 2200 650
[ 134Cs 1.0 1.0 137Cs 1.0 1.0 24N6 2500 550
- 51Cr 23 15 54Mn 110 16 l 59Fe 43 17 l
seCo 69 13 60Co 95 20 99Mo 520 200 124Sb 5.6 3.4 i t nBest Value o DF = Average Inlet Concentration Average Outlet Concentration i 1 I l l 15 l
l i When the base-cation demineralizers were loaded with a mixed-bed . resin, the DF's for 1811 and the cesiums were found to be highly correlated to the inlet concentration. This is the same observation I that was made for the mixed-bed letdown demineralizers (see Section 2.2.3). l The measurements carried out on the two evaporators also showed that the DF (as measured by the ratio of the feed to distillate activities) was dependent spon the inlet concentration. For the radwaste evaporator, the DF's for lHI, the cesiums, and the cobalts were found to be highly correlated with the feed concentrations for these isotopes. As with the mixed-bed demineralizers, DF tended to increase as the inlet concentration increased. For the boric acid evaporator the DF's for the cobalts showed a high degree of correlation with feed activities, but neither i 1311 nor the cesiums showed this effect. When these measurements were made, however, the concentrations of iodine an:i cesium in the feed were very low. Distillate cov,entration for this type of evaporator was found to be . related to bottoms concentration rather than tne feed concentration. In . particular, it was found that when the concentration in the feed is drastically reduced the distillate and bottoms concentrations remained essentially constant. Hence the feed-to-distillate DF actually went to a value less than cra whereas the bottoms-to-distillate ratio remained very large. i l Since the data base is not yet sufficient to quantify the relationship between DF and feed concentration for evaporators, "best value" DF's were obtained by dividing the average inlet (or feed) concentration by the average outlet (or distillate) concentration for each radionuclide. Tables 2.2 and 2.3 list "best value" DF's for selected radionuclides in the various components of the boric acid recovery system and the radwaste system, respectively. More detailed tabulations of "best value" DF's for the boric acid recovery system and radwaste system can be found in Tables 4.7, 4.8, 4.12, and 4.14. For the boric acid recovery system, the DF's for the base-cation demineralizer and the evaporatc,* were obtained from mc surements made only while the demineralizer contained a l i cation resin (i.te. data obtained while the base-cation duineralizer ] was loaded with a mixed-bed resin were not included). The DF's for the , condensate demine-alizer and filter, however, were obtained using all i
- the data obtained during the in-plant measurements. )
2.2.5 Secondary System i One of the primary objectives in carrying out in-plant j measurements at Turkey Point was to study steam generator leaks if they occurred. During the measurement period, leaks occurred in steam generator i C of Unit #3 (November 1977) and in steam generator A of Unit #4 (knuary- l February, 1978) . Detailed studies of these steam generator leaks were i made and lead to the following conclusions and observations. ' Based on letdown demineralizer DF's for both plants, the iodine in the reactor coolant system is >99% ionic. The reducing conditions of the reactor coolant (i.e., hydrogen overcover together with reduction 16 l
TABLE 2.2 l "BEST VALUE"tDECONTAMINATION FACTORS FOR BORIC ACID RECOVERY SYSTEM "Best Value" Decontaminatior. Factor for System Component t i Base-Cation Condensate j Nuclide Demin. Evaporator Demin. Filter 1311 1.6 50 11 1.I 134Cs > 900 --- 1 1.4 137Cs > 280 --- I l.9 seCo 24- 780 l I.8 60C0 19 31 0 2.6 1.8 l TABLE 2.3 "BEST VALUE'* DECONTAMINATION FACTORS FOR RADWASTE SYSTEM ,
"Best Value" DF for System Component
- Condensate Nuclide Evaporator Demin.
131I 24 4.6 133I 410 _____ 134Cs 910 0.2 136Cs 37 ----- j 137Cs 1100 0.9 seCo 510 170 60Co 440 18 Averace Inlet (or Feed) Concentration t "Best Value" DF = Average C utlet (or Distillate) Concentration i 17 l
due to thennal and radiation conditions) favors I as the principal i"ine form (5,11). For this reason, iodine entering a steam generator om the reactor coolant is principally I . Measurements showed that
.11 the activity entering the blowdown flash tank from steam generator blowdown leaves the flash tank in the liquid stream. Iodine activity released via the blowdown flash tank vent was below the detection limits of the measurements. Therefore, little, if any, iodine was vaporized and released tnrough the flash tank vent. This indicates that iodine entered the secondary system in a nonvolatile form and remained nonvolatile in the steam generators and blowdown flash tank.
Measurements indicated that the average steam generator iodine { partition factor for Unit #4 was 9 + 4(-4). This very low partition factor indicates that there was very little volatile iodine in the secondary system. Because of insufficient data, the iodine partition factor for Unit #3 steam generators could ' lot be determined. Another result that indicates that there was very little volatile iodine in the secondary system is the high fraction of iodine activity that entered the main feed via the high pressure (HP) drains. Measurements indicated that an average of 75% of the iodine activity in the feed came via the HP drains even though only 28% of the flow follous that path. This result is consistent with the conclusion that most of the iodine activity gets into the main steam entrained in moisture droplets because most of the moisture droplets should be removed by the moisture separator and this removed moisture is routed to the HP drain system. Comparisons of 1331/131 I ratios in the reactor coolant and the generator bottoms indicated that the age of the iodine in Unit #3 secondary was about 20 hours older than reactor coolant. In Unit #4, the age of the secondary water was about 8.5 hours older than reactor coolant. Measurements indicated that the steam leaving the steam generators contained about the expected amount of moisture. The moisture entrain-ment fraction (i.e., moisture carryover) was measured to be 0.3%, which is approximately the steam generator design value (design carryover is 0.25%). The primary to secondary leakage in the two units was different. Unit #3 had a steady-state leak which v;ried between 3 and 8 gulars per hour during the leak period (mid-August to late November 1977). The leak in Unit #4 increased constantly throughout the period of measurement, starting at 0.6 gallons per hour on 1/15/78 and increasing to about 20 gallons per hour on 2/14/78 before the, unit was shut down for repairs. The steam jet air ejector (SJAh; is the release point for iodine leaving the main condenser to the environment. Average daily releases measured during the period of primary-to-secondary leakage were 18
l e 4 l 4.7(-4) pCi/sec for 131 1 and 3.8(-4) pCi/sec for 133 I in Unit #3, and 1.8(-5) pCi/sec and 2.3(-5) pCi/sec for 1311 and 133I , respectively, I in Unit #4. Based on these release rates, the average percent of main steam iodine activity released via the air ejector is 1.3(-1)% for
. Unit #3 and 1.0(-1)% for Unit #4. This small fraction of iodine leaving
] the secondary system via the air ejector is consistent with the observa-tion that there was very little volatile iodine in the secondary system. As measured by iodine species samplers, the bulk (approximately 90%) of the iodine which was discharged from the air ejectors was in the organic I form. The gland seal discharge which exhausts to atmosphere was not l measured due to lack of a sampling point. 2.2.6 Spent ael Pit The spent fuel pit (SFP) cooling system consists of a pump,
- heat exchanger, filter and a mixed-bed demineralizer. The pump draws water from the SFP, circulates it through the heat exchanger, and returns it to the SFP. Nominal flow through the cooling loop is 2000 gallons per minute. To maintain clarity and to remove any radioactive contaminants i approximately 5 percent of the cooling loop flow is diverted through the filter and the mixed-bed demineralizer.
During February and April five sets of measurements were made on the Unit #3 spent fuel pit system to determine the mixed-bed demineralizer DF's. The mixed-bed demineralizer resin was replaced in January,1978. A summary of the "best value" DF's for the most frequently observed radionuclides is presented in Table 2.4. Table 5.7 contains a more complete listing of "best value" DF's. Samples were taken of the SFP, reactor coolant, and the refueling water storage tank (RWST) vaters before and after transfer of fuel from Unit #3 core to the SFP. The data indicate that the total number of curies of 3 H in all the previously mentioned waters decreased from approximately 18 prior to fuel transfer to approximately 11 curies after fuel transfer. The balance is due to release via the monitor tanks. 2.2.7 General Conclusions - Demineralizers and Evaporators The measurements carried out on the denineralizers and j evaporators indicate a high degree of correlation between DF and inlet (or feed) concentration. Figures 2.5-2.8 show plots of DF vs. inlet concentration for selected radionuclides in demineralizers and evaporators. Included are data from Turk:y Point and from the other two PWR's studied in the in-plant measurement program (2,3). In addition, all mixed-bed demineralizers (letdown, spent fuel pit, evaporator condensate demineralizers, and the base cation demineralizers at Turkey Point when loaded with mixed-bed resin) and all evaporators (boric acid and radwaste evaporators from all vendors studied) are included. I I 19
l t ) TABLE 2.4 "BEST VALUE'* DECONTAMINATION FACTORS FOR SPENT FUEL PIT DEMINERALIZER "Best Value" Nuclide DF 131I 2.5(0) 134Cs 1.0(0) 137Cs 1.0(0) 54Mn 8.6 55Fe 2.7 seCo 8.3 60Co 1,2 63N i 3.4 90Sr 1.0 91Y 1.4 I 95Nb 2.7 124Sb 1.5 125Sb 1.2(0 l l 0 t "Best Value" DF = Average Inlet Concentration Average Outlet Concentration ] l l 1 4 a 20
l i r Firiure 2.5 60 Co Decontamination Factors for Mixed-Bed Deafneralizers 104 , , i i . ,,ig i i i iiiiii i i i s inisi i i i siinal i ' ' ' ' ' ' ' 103 ,_, e - S e
- g e ,
E 102 _ e o g S i * ** e a
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- e
$ 101 _
, e
- 8 e e* e o e Turkey Point demineralizers O .p
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, a 8 e e a e s = *
- Zion demineralizers
~
e I 8 ~ e Fort Calho. '1 demineralizer
-1 e i t ' ItIII ' i 8 8 'I ' ' ' I'i ' ' ' ' ' ' ' ' ' ' ' '
10-6 10-5 10-4 10-3 10-2 10-1 Inlet concentratici ( p Ci/ml ) INEL-A-11040 i
Figure 2.6 131I Decontamination Factors for Evaporators 104 i i i i isig; i i iiising i i i iising i i i iniin i , i,iii, i ,,,iiig
- Turkey Point evaporators 103 _
-
- Zion evaporators 2
$ e Fort Calhoun evaporator
- 102 - m
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i .
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o
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' ' 'I ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '' ' ' ' ' ' ' '
10-1 10-2 10-1 10 4 10-5 10-4 10-3 10-7 IN EL. A-11038 Feed concentration ( Ci/ml)
k Figure 2.7 137 Cs Decontamination Factors for Evaporators 104 . . , isig , , ,
,,,ig i i i . iiig i i i si f
~
e e ~ 103 =__ e a e
, =
a : 2 - e ; g _ * - e _
.9 N
ca .E 102
= o =
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e a e Turkey Point evaporators .- o - e a Q Zion evaporators - 101 ~ a e Fort Calhoun evaporator 1
' ' ' ' '! ' ' ' ' 'I * ' ' ' 'I ' ' ' ' '
10_-6 10-5 30-4 10-3 10-2 Feed concentration (y Ci/ml) INEL A-11039
I Figure 2.8 60Co Decontamination Factors for Eve 9 orators 105 , , , , , , , , , , , , , , i i siis [
- Turkey Point evaporators e
104 -
- Zion evaporators e
e Fort Calhoun evaporator 3 U - J!! 103 -
.9 e
% a
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e g
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,/ 4 A e i i eiiil i i i i eieil I i i i a a il i i e a i e ii i
3 10-5 10-4 10-3 10-2 10-6 Feed concentration (p C /rni ) INEL-A-11041
l l Although the scatter of data points in Figures 2.5-2.8 is large I (probably due to the variety of operating conditions, demineralizer bed lives, evaporator bottoms concentrations, etc.), the data indicate that the DF's for these radionuclides tend to increase with inlet (or feed) concentration. If the data in these figures are fit (via the nethod of ; j least squares) with a straight line, i.e., an equation of the form B D = AC where I C - inlet concentration i A,8 -- constants D - decontaminat. ion fdctor, ! the approximate slopes (i.e., constant B) obtained are 0.6 for 1311 and 0.5
- for 60Co in mixed-bed demineralizers, 0.6 for 1311, 0.5 for 137Cs, and 0.8 for 60Co in evaporators. The straight lines in the figures show the' fits of the above equation to the data. It must be emphasized that thu e
! results are based on a limited amount of data obtained under a variety of operating conditions. More data and more detailed analyses (including effects of operating conditions, demir.eralizer bed ii~v'es, evaporator bottoms concentrations, etc.) is required to verify that this relationship is generally applicable to all demineralizers and evaporators and to quantify the functional relationship between DF and inlet concentration. When the data base is sufficient to quantify the functional dependence between DF and inle.t (or feed) concentration, the function will be used in place of "best value" DF because it is more representative of actual plant performance. 2.3 Gaseous Systems At Turkey Point there are five major areas or systems on the primary side which are potential sources of gaseous radioactivity. Potential gaseous source tenns associated with the secondary system are discussed l separately in section 2.2.5 above. The five major areas or systems i associated with the primary side are the
- 1. auxiliary building,
, 2. containment buildings (Unit #3 and #4),
- 3. waste gas processing system,
- 4. the spent fuel pit areas (Unit #3 and #4), and
- 5. new radwaste building.
With the exception of the Unit #3 spent fuel pit area, all the potential sources feed into the plant's main exhaust stack. Unit #3 spent fuel pit area is independent and has its own exhaust stack. Figure 2.9 shows block diagrams of the auxiliary building and spent fuel pit #3 ventilation systems with sampler locations noted. For reference, Figure 1.2 presents the overall gaseous waste system at Turkey Point. Table 2.5 summarizes 25 l
i Figure 2.9 , 4 i Auxilf ary Building and Spent Fuel Pit 6 Ventilation Systems
+
1 Particulate.
\ Particulate, iodine,14C \ iodine,14C j and 3H samplers Stack k and 3H samplers Stackk t
- l Fdters l A i Unit #4 Electrical Machine penetration room shop room
' *
- 0'
N Unit #3 spent Fitters di di 16 h fuel pit area Unit #4 Seal water Unit #4 Waste 00 Containment ; *- gas decay and residual heat tan to tank rooms exch room spray pump room 2 h "3 Unit #4 Holdup
- Res&al heat ;
Unit #4 ;
- Pipeways Sample room tank rooms exchanger room Urut #3 Unit #3 Unit #4 Unit #4 Charging - Electrical e- Non-regen heat i Non-regen heat -
exchanger room exchanger room pump room yetration room Boric acid evap Unit #3 Unit #3 ,,Old., waste *- Seal heat Charging - r : package rooms evaporator room (units #3 and #4) exchanger room pump room U" "3 Solid waste Unit #3 e Containment Monitor ~ = ' = Pipeways tank room drumming room spray pump room Concentrate
" ' nit #3 exchange room -
Safety injection pump room h- " Sample room *- holding tank (units #3 and #4) - Deborating and son Gas Ressdual heat Residual heat *-- stnpper rooms .- removal pump - exchangerooP1 exchanger rooms (units #3 and #4) rooms (units #3 and #4) - Sampfer notes:
- 1. Circled numbers indicate longterm sampling stations with i s i samplers for particulate and iodine.
- 2. Circled letters show short term sampler locations with ,
r 4 particulate and iodme samplers.
- 3. Longterm sampfers are indicated at both stack sample focations. ?
I i 1 _ _ . _ . _ _ . _ _ _ _ ___________ _ __ _ _ m_ . - - , - - . -
TABLE 2.5 AVERAGE RELEASE RATES [8] DURING COMBINED REFUELING AND NON-REFUELING PERIODS - (pC1/sec) i Nuclides System 1311 134Cs 137Cs seco 60Co 54Mn 3H 14C Stack 2.5(-2) 2.6(-4) 4.2(-4) 3.4(-3) 1.4(-3) 2.0(-4) 4.3(-1) 2.2(-1) Auxiliary Building 2.3(-2) 2.5(-7) 4.4(-7) 3.5(-7) 1.9(-7) 2.0(-8) 5. S 2) 1.5(-1) (<0.1)[7] (0.1)[7] (0.1)[7] (0.1)[7] (0.1)[7] (11.9)[2.4] (G8.1)[2.4] Containments (Units1.9(-4) #3 & #4)E93 (92.5)[1] 2.4(-6) 6.0(-6) 1.0(-7) 9.5(-7) [3] 2.9(-1) 4.6(-3) (0.7) (<0.1) (1.5) (<0.1) (0.1) 37.4) (2.1) i Waste Gas Processing System 3.8(-5) 1.2(-11) 2.7(-11) 2.7(-10) 9.2(-11) [3] 7.6(-4) 5.2(-2) 1 (0.1) (<0.1)[7] (<0.1)[7] (<0.1)[7] (<0.1)[7] (<1.0) (23.6) Unit #4 Spent Fuel Area and 1.8(-3) 2.6(-4) 4.1(-4) 3.4(-3) 1.4(-3) [3] to New Radwaste Building 8.8(-2) 1.2(-2) (.72)[2.4] (99+)[2.4] (97.6)[2.4] (99+)[2.4] (99+)[2.4] (20.4)[4.5] (5.5)[4.5] Unit #3 Spent Fuel Area [6] 4.8(-4) [3] 5.0(-7) [3] [3] [3] 8.8(-2) 1.2(-2) [1] Ntanbers in parentheses below release rate values are percent of stack release rate. [2] Calculated from differences of other sources which feed stack. [3] Radionuclide not detected or insufficient data to calculate values. [4] Areas not sampled for radionuclide. [5] Value assumed to be the same value as Unit #3 spent fuel area (see section 5.3.1). [6] The Unit #3 spent fuel area does not contribute to the stack releases; therefore, to obtain the total plant release rate the stack and Unit #3 spent fuel area release rates must be added. To obtain an extrapolated annual release (pC1) for any of the radionuclides the average release rate can be multiplied by the number of seconds per year. 3.15(7). [7] Reit.ase rates have a DF of 100 applied for exhaust HEPA filters. [8] All release rates are downstream of exhaust filter banks. [9] Release rates from containment are for non-refueling interval only, no containment samples were taken of containment during refueling.
the average release rates for both Units #3 and #4 of measured activities during the in-plant measurement period at Turkey Point. From inspection of Table 2.5 it is apparent that the auxiliary building was the predominant source of 131 I during the in-plant measurement period. It contributed 92.5% of the total. Table 2.6 shows a breakdown cf the data taken from sampling stations in the auxiliary building. Note that sampling station
- 4 was consistently the major contributor of 131 1 from the. auxiliary building . Sampling station #4 included innong other things (Figure 2.9),
the radwaste and boric acid evaporator (BAE) rooms, their associated gas stripper rooms, and the Unit #3 sample room. As is shown in section 8, the Unit #3 BAE room was the major source of 131 1 in the auxiliary building. It should be noted here that the differences between the auxiliary building and stack releases are due to waste gas decay tank and containment purge releases as well as a 20% uncertainty in the measured values (see section 8.3.2). Tables 2.7 and 2.8 present the combined average release rates during the non-refueling and refueling intervals. During both intervals the auxiliary building was the predominant source of 131 1 In fact, the only time an area other than the auxiliary building was a significant contributor to the stack 131 1 release rate was during the refueling interval. Although samples were not taken of the containment atmospheres during refueling, the data indicate that the containments account for approximately 30% of the stack 131I releases. This is supported by the facts that release rates from the waste gas processing system are insignifcant during both the refueling and non-refueling intervals and that the 131 1 release from the Unit #4 spent fuel area and new radwaste building should be similar to the Unit #3 and spent fuel area. The Unit #3 spent fuel area 13'I release rate is insignificant compared to the stack 1311 release rate. During non-refueling periods, the major sources of I"C and 3 H were the auxiliary and cor.tainment building, -espectively. During refueling, a distribution similar to the non-refueling distribution for the two radionuclides is believed to exist, however, data are not available to detemine an exact distribution. In all cases, the Unit #4 spent fuel and new radwaste building areas were the predominant sources of particulates (radiocesiums, radiocobalts and 5"Mn). The data indicate the solid waste solidification operation is the source of the particulates (see section 8.3.3). l Table 2.9 shows the 1311 and 3H main stack release rates nonnalized to their respective average reactor coolant concentrations during the i sampling intervals. The normalized release rates are the measured release I rates divided by the avegage reactor coolant concentration. It should be pointed out that the I nomalized rates in Table 2.7 were calculated by including 131 1 concentrations due to radioiodine spiking associated with reactor power transients. For comparative 131I normalized release rates excluding 131 I reactor coolant spikes refer to Table 8.4. The l average 131 1 normalized release rate including spiking concentrations is 2.0 (uCi/sec)/(uCi/gm). If spiking is not included the average 131I normalized rate is 2.9 (pCi/sec)/(pCi/gm). The 3 H average normalized release rate (including Unit #3 spent fuel area) is 2.95 (uCi/sec)/(uCi/gm). i The normalized 131I and 3 H averages quoted represent total plant releases. l l 1 28 l i
TABLE 2.6 AUXILIARY BUILDING 131I SOURCES Auxiliary Percent of Totd Auxiliary Building 131I Releases Stack Building Total (total 1311) Station Station Station Station Station Station Sample Pe:fod (uCf/sec) (uci/sec) #1 #2 #3 #4
#4A #5 11/10-11/21/77 4.00 1 0.05(-2) 4.58 0.03(-2) 0.5 1.3 3.6 94.5 N.S. 0.1 11/21-12/4/77[4] 1.04 1 0.02(-2) 4.14 1 0.02(-3) 1.8 7.5 7.2 83.3 N.S. 0.2 12/4-12/14/77[4] 1.74 i 0.02(-2) 2.6110.04(-3) 2.9 23.3 2.6 64.8 N.S. 6.a 12/14-12/28/7{4] 6.77 1 0.04(-2) 3.79 0.01(-2) 0.4 15.4 3.6 80.7 N.S. 0.06 12/28-1/11/78[4]4.39i0.03(-2) 3.61 1 0.01(-2) 0.3 3.0 3.2 93.4 N.S. 0.05 1/11-1/25/78[4] 3.3910.09(-3) 1.79 i 0.03(- 3) 1.2 8.4 6.5 83.7 48.2[1] 0.2 1/25-2/8/78[4] 7.66 1 0.04(-2) 7.16
- 0.02(-2) 0.6 1.5 0.9 96.8 34.9 0.2 g 2/8-2/22/78[4] 5.08 1 0.04(-2) 4.41 1 0.01(-2) 0.4 3.3 2.3 93.6 20.3 0.4 2/22-3/9/78 1.9310.02(-2) 1.20 1 0.01(-2) 0.9 6.2 5.6 85.7 31.3 1.6 3/9-3/21/78 1.43 0.02(-2) 1.8910.01(-2) 0.3 5.0 3.2 91.4 17.0 0.1 3/21-4/3/78 2.52 1 0.01(-3) 1.10
- 0.02(-3) 0.8 4.6 8.6 85.6 72.0 0.4 4/3-4/20/78 4.6910.09(-3) 3.30 1 0.04(-3) 0.5 9.4 3.7 86.1 19.7 0.2 4/20-5/4/78 2.56 1 0.02(-2) 2.80
- 0.01 (-2) 0.2 0.2 2.3 97.2 34.3 0.05 5/4-5/18/78 1.13 i C.05(-3) 1.09 0.02(-4)[2] 1.1 2.2 1.5 26.7[2,3] 95.0 0.2 5/18-6/1/78 1.8310.06(-3) 3.910.1(-4)[2] 0.8 2.3 1.3 75.0[2,3] 90.6 4.9
[1] Short sangle period. [2] Auxiliary building Iodine-131 total based on sample station #4A due to sar.tpler malfunction at station #4 and consequently, the sum is low. [3] The nunbers represent the percent of the total stack releases. [4] Samples associated with the refueling interval. 1 _ _ _ _ rA_
. - _ _ . .- _ -_ - - - _ = __ _ . _ - -. ._ . _.
TABLE 2.7 AVERAGE RELEASE RATES DURING NON-REFUELING INTERVAL [5] (uti/sec) System 181I 134Cs 137Cs seco 60Co 54Mn 3H 14C Stack 1.4(-2) 2.7(-5) 4.6(-5) 5.5(-3) 1.3(-3) 2.3(-4) 5.7(-1) 1.8(-1) Auxiliary Building 1.8(-2) 1.7(-8)[1] 5.8(-8)[1] 2.8(-7)[1] i.1(-7 1.7(-8)[1] 1-6(-1)[4] 1.4(-1)[4] (100)[7] (0.2) (1.3) (0.0) (0.0))[1] (13.0) (75.8) Containments (Units #3 & #4) 1.9(-4) 2.4(-6) 6.0(-6) 1.0(-7) 9.5(-7) [2] 2.9(-1) 4.6(-3) (0.0) (8.8) (13.0) (0.0) (0.0) (50.8) (2.5) Wasta Gas Processing System 5.7(-6) 1.0(-11)[1] 1.3(-11)[1] 4.4(-11)[1] 8.6(-11)[1] [2] 2.5(-4) 2.5(-2)
- (0.0) (0.0) (0.0) (0.0) (0.0) (0.0) (13.9) g Unit M Spent Fuel Area and [2.4] 2.5(-5)[4] 4.0(-5)[4] 5.5(-3)[4] 1.3(-3)[4] [2] 1.2(-1)[3] 1.4(-2)[3]
New Radweste Building (91.0) (86.8) (100) (100) (21.2) (7.8) Unit #3 Spent Fuel Area [6] 2.6(-4) [2] 3.5(-7) [2] [2] [2] 1.2(-1) 1.4(-2) [1] Release rates have a DF of 100 applied for exhaust HEPA filters. [2] Radionuclide not detected or insufficient data to calculate values. [3] Value asstmed to be the same value as Unit #3 spent fuel area (see section 5.3.1). [4] Areas not sampled for radionuclide, values calculated from differences of other sources feeding the stack. [5] All release rates are downstream of exhaust filter banks. 4
~6] The Unit #3 spent fuel area does not contribute to the stack releases; to obtain the total plant release rate the i Unit #3 spent fuel area and the stack release rates must be added.
[7] Numbers in parentheses below the release rate values are percent of stack release rate.
TABLE 2.8 AVERAGE RELEASE RATES DURING REFUELING INTERVALE23 (pC1/sec) System 131I 134Cs 137Cs seco 60Co 54Mn 3H 14C Stack 3.9(-2) 5.0(-4) 8.8(-4) 1.3(-3) 1.5(-3) 8.9(-5) 3.2(-1) 3.0(-1) Auxiliary Building 2.8(-2) t.8(-7 [5] (/1.8)[4] (0.1) )[1] (0.1) 7.6(-7)[1](0.0)4.0(-7)[1] 2.5(-7)[1] (0.0) (0.0)2.4(-8)[1] [5] Containments (Units #3 & #4) [5] [5] [5] [5] [5] [5] [5] [5] Waste Gas Processing System 3.1(-5) 2.9(-12) 1.4(-11) 2.2(-10) 1.2(-11) [6] 5.8(-4) 2.7(-2) (0.0) (0.0)[1] (0.0)[1] (0.0)(1] (0.0)[1] (0.2) (9.0) Unit #4 Spent Fuel Area and [5] [5] [5] [5] [5] [5] 2.9(-2) 9.8(-3) 3 Nw Radwaste Building (9.1)[5.7] (3.3)[5.7] Unit #3 Spent Fuel Area 7.5(-4) [6] 6.0(-7) [6] [6] [6] 2.9(-2) 9.8(-3) [1] Release rates have a DF of 100 applied for exhaust HEPA filters. [2] All rc:aase rates are downstream of the exhaust HEPA filters. [3] The Unit #3 5;:ent fuel area does not contribute to the stack releases; to obtain the total plant release rates the Unit #3 spent fuel area and the stack release rates must be added. [4] Nw6ers in parentlasis below the release rate values are precent of stack release rate. [5] Area not sampled for radionuclide. [6] Radionuclide not detected. [7] Value assumed to be the same value as Unit #3 spent fuel area.
TABLE 2.9 . MAIN STACK 3H AND 131I NORMALIZED RELEASE RATES [10] AND EFFECTIVE PARTITION COEFFICIENTS Normalized 3H Normalized Reactor Coolant Release Rate Reactor Coolant 1311 Release [8] Average 3H Average 1311 vC1/sec
) f5,7]
uC1/sec)[6,7] Date (pC1/ml)f91 pC1/qm (vCi/ml)f31 Rate (iiC1/amEPF 11/10-11/21 7.1(- 0.44 1.3 - 3.08 7.0 l 11/21-12/4 1.9 - 0.83 1.4 - 0.74 0.89 12/4-12/14 1.1 - 4.4 1. 3 - '11 1.34 0.30 1 12/14-12/28 1.5 - 4.5 1.2 - 'l ' 5.64 1.25 l 12/28-1/11 1.6 - O.94 8.3 - 'l ' 5.29 5.62 1/11-1/25 2.3 - 'l' O.83 6.5 - ' 1 '. 0.52 0.62 1/25-2/8 1.6 - 'lf
. 1.5 1.4 - '1] 5.47 3.65 1 2/8-2/22 8.0 - 3.0 [4] [4]
2/22-3/9 1. 3 - [2] 3.5 1.1 - [2] 1.75 0.50 3/9-3/21 1.4 - 1.6 1.2 - 1.19 0.74 I 3/21-4/3 1.5 - 1.8 1.2 - 0.21 0.12 l 4/3-4/20 2.2 - 7.7 3.2 - 0.15 1.9(-2) 4/20-5/4 2.3 - 1.4 1.2 - 2.13 1.52 5/4-5/18 1.85(- ) 0.11 1.0 - 0.11 1.0 5/18-6/1 2.0(-1 5.0 8.6 - 0.21 4.2(-2) [1] Unit #4 shutdown; Unit #3 rei.ctor coolant concentrations used. [2] Unit #3 shutdown; Unit #4 reactor coolant concentrations used. [3] Reactor coolant average 131I concentration includes spike concentrations. ) Also includes plant analyses. [4] Insufficient reactor coolant analyses to calculate average reactor coolant l concentration of release rate. l [5] Effective Partition Factor (EPF) 'is the ratio of normalized 1311 release rate to the nomalized tritium release rate. [6] The 3H releases from the Unit f3 spent fuel pit (SFP) are not included in the above nomalized 3H release rates. Including the Unit #3 SFP area would increase the total plant nomalized rates by approximately 17 percent. Therefore, the corrected nomalized total plant release rate is 2.95 (pCf/sec)/(pCi/gm). [7] One ml of reactor coolant sample weighs one gram, i.e., gram and ml are essentially interchangeable. [8] The 131I releases from the Unit #3 spent fuel area are not included in the above non.alized 1311 release rates. However since 1311 releases from the Unit #3 are less than 2 percent of the total plant I31I releases the main stack 1311 normalized release rates represent total plant normalized rates. [9] Average Unit f3 and #4 3H reactor coolant concentrations except as noted. [10] Normalized release rates are for both Units #3 and #4.
Table 2.10 presents a summary of the average 181 1 species distributions at Turkey Point. As indicated, with the exception of Unit #3 spent fuel pit area, where the species is H0I, the predominant iodine chemical species is organic iodine. This includes station #4 of the auxiliary building. The existence of organic iodine is normally associated with older fodines, i.e., farther removed in time from the primary coolant. Because of the ! close proximity of sample station #4 to the major 1311 source (i.e., #3 BAE), the predominant iodine species would not be expected to be organic. As #3 BAE processes primary coolant type water, one would expect a significant fraction of the iodine to be in the elemental fom. However, as indicated the species is predominantly organic with very little elemental. This was also observed in the radwaste evaporator area at i another plant (7). The controlling chemistry for this phenomenon is not known; however, one can postulate that a percentage (10%) of the iodine collected on the cleanup demineralizers is converted to an organic species and subsequently transferred to the BAE. Alternately, one could postulate that the organic iodine is fomed in the bottoms of the BAE's. The latter postulation is supported (as discussed in section 8.3.2) by nomalization of the 131 1 releases to the BAE bottoms. Table 2.11 presents the effective reactor coolant leakage rates into the containment and auxiliary buildings from both Units #3 and #4. The leakage rates are temed effective reactor coolant leakage rates as the leak rates are based on airborne radionuclide concentrations. For a given amount of reactor coolant leakage into the auxiliary or containment buildings only a fraction of the reactor coolant liquid vaporizes, the balance going to the waste drain tanks in the buildings. The amount of liquid that goes to the waste drain tanks is not included in the effective reactor coolant leak rates presented. As a result, the leak rates presented are lower than the actual reactor coolant leak rates. , However, the leak rates calculated based on airborne concentrations represent more appropriate leak rates for predicting the gaseous radio-nuclide inventory in containment. For example, for a given radionuclide reactor coolant concentration and an effective leak rate for the same radionuclide, the actual amount of the radionuclide that leaks into i the containment or auxiliary buildings and is vaporized can be predicted. This is significant in that it is the airborne radionuclide inventories which are available for release to the environment. The approach taken in this study was to detemine effective leak rates into the containment buildings based on 3H and radioiodine and to calculate effective leak rates into the auxiliary building based on 3H only. The auxiliary building effective leakage rates were based on 3H only since effective partition factors (EPF's) greater than 1.0 were observed in the auxiliary building for the radiofodines. This approach was taken as EPF's should not exceed 1.0 (cf sections 8.3.6 and 8.3.2). The average effective leakage rates into the auxiliary building in Table 2.9 are based on the average stack value (475 lbs/ day, Table 8.12) and the percent 3 H coming from the auxiliary building (11.9%, Table 2.5). 1 The effective leak rates from the Unit #3 and #4 containment buildings l are an average of Unit #3 and #4 containments and are based on measure-ments of the containment building atmospheres. 33
TABLE 2.10 AVERAGE 131I DISTRIBUTION AT TURKEY POINT (percent) location Parti'.dlate Fil ter I 2 HOI Organic Stack 1.3 5.9 27.9 64.9 Auxiliary Building 2.5 9.2 24.3 64.0
- 3 Spent Fuel Area 1.8 11.8 57.9 28.5 Waste Gas System <1.0 <l.0 5.5 93.7 Containment Buildings <1.0 3.6 11.1 84.2 Station #4 1.2 4.7 19.7 74.5 l
1 I 34 (
TABLE 2.11 _ AVERAGE EFFECTIVE LEAKAGE RATES OF REACTOR COOLANT INTO THE CONTAINMENT AND AUXILIARY BUILDING (During Refueling and Non-refueling Combined) Based on Based on 3H 131I Auxiliary Building 56 --- (ibs/ day) Containment Buildings 2.7(-2) 8.7(-4) (percent / day) l 35
. = _ _ .. - - -._ --
The following are conclusions based on the measurements made at ( Turkey Point: 3
- 1. The average, total plant H and 131 I normalized release rates are 2.9 and 2.9 (uCi/sec)/(pCi/gm), respectively. The normal-i ized 1311 released rates do not include spikes in the 1311 reactor coolant concentrations. The 1311 release rate is nomimally a factor of 40 higher than the average total release rate observed in a study by Science Applications, 3 Inc. (SAI) at three other PWR's (7), while the normalized H release rate observed in this study was nominally a factor of 1.5 lower than the total release rate measured by SAI. j i
- 2. The auxiliary building is the major source of gaseous 1311, ranging between nominally 70 and 90 percent of the total 1311 releases during both refueling and non-refueling.
- 3. Within the auxiliary building the #3 boric acid evaporator i room is the major source of 131I activity. It consistently l accounts for more than 80% of the total auxiliary building releases.
, 4. Of the spent fuel and new radwaste areas, the waste gas processing system, and the containment buildings, only the containment buildings contribute significant quantities of J 131I compard to the auxiliary building. The containment buildings (.onVibute approximately 30% of the plant 131I l releases dtring the refueling interval, the only time the
! containment buildings' contribution is significant.
- 5. The average iodine species distribution released to the
; environment via the plant stack is 64.9% organic, 27.9%
H0I, 5.9% elemental, and 1.3% particulate. In the auxiliary ) building the average percent distribution of the sum of the sampling locations is 64.0 organic, 24.3 HOI, 9.2 elemental, and 2.5 particulate. The distribution in the major 131 1 source (#3 boric acid evaporator room) of the auxiliary building is j 74.5% organic,19.7% percent H0I, 4.7% elemental, and 1.2% particulate.
- 6. The new radwaste building contributes greater tiia 95% of the extrapolated annual particulate releases (134Cs, '37Cs, 58C0, soCo, and 54Mn). It is believed to be due to the :,olid waste solidification operation.
- 7. The spent fuel areas combined with the containment buildings are the major sources of 3H. The contributions from the i containment range between 50 and 67 percent and the spent fuel areas range between 10 and 20 percent. Tritium from i the auxiliary building was most significant during non-refueli g, when it was 28 percent.
i l I 36
- 8. The 14C releases are 70,1, 24, and 5 percent, respectively, for the auxiliary building, containment buildings, waste gas processing system, and the spent fuel pit areas during the refueling and non-refueling combined. A similar distri-bution was observed during non-refueling.
l
- 9. The 131 1 effective partition factors in some cases exceed a value of 1.0. This is indicative of a source of 131 I higher in concentration than the reactor coolant.
- 10. The average effective leak rate of reactor coolant (based on tritium) into the axuiliary building is approximately 56 pounds per day.
- 11. The best estimate of the effective reactor coolant leakage rate into the containment buildings is 5.4(-4) % per day for 3 H and 8.7(-4) % per day for 1311.
l 37
l l l
- 3. REACTOR COOLANT AND LETDOWN SYSTEM l 3.1 System Description and Sample Points 3.1.1 Reactor Coolant System The reactor coolant system is used to circulate the heated, high-pressure reactor coolant from the reactor to the steam generators.
For Turkey Point Units #3 and #4, this system consists of three loops, one for each of .the steam generators. Figure 3.1 shows a simplified ! diagram of the reactor coolant system and applies to both units. l As shown in Figure 3.1, the available sample locations in the reactor coolant system are located on hotlegs A and B. The valving allows individual samples to be taken from each hotleg or a mixed sample from both hotlegs. Liquid flows from the hotleg(s) through a cooler and then to a reactor coolant sample sink for each unit. Normal sampling procedure was to open the two valves to obtain a mixed sample from hotlegs A and B and allow liquid to recirculate for a minimum of 15 minutes. The sample line to the sink was then purged for a minimum of 10 minutes prior to collection of a sample. Samples were taken by two methods. Most samples were collected in 4 50 ml bottles as described in the Source Term Procedures (4). Since an indeterminate amount of the noble gases may be lost from these samples, when noble gas concentrations were desired, a more specialized sampling technique was snployed. This technique consisted of plumbing a 35 ml liquid bomb sampler i'n series with the sample line, purging the sampler for a minimum of 2 minutes, then closing the stopcocks to collect a sample that had retained its dissolved noble gases. Noble gas concentrations are reported only for reactor coolant samples obtained using the 35 ml liquid bomb sampler. Reactor coolant samples were obtained from both units approximately
; weekly between 11/9/77 and 6/1/78. When the letdown system was being studied and spiking due to shutdown or startup was being investigated, more frequent campling was utilized. Samples for analysis of beta-only-emitting raf +1 ides were obtained on a less frequent basis.
3.1.2 g oo a System 3.1.2.1 System Description The chemical and volume control system (CVCS) provides a means of purifying and degassing reactor coolant. The system is used to control reactor coolant boron concentration and provides for j the addition of corrosion inhibiting chemicals and makeup water to the j i reactor coolant. 1 I l 38 f-
Fiqure ?.1 Reactor Coolant System, linits 13 and #^ Steam generator C Reactor p coolant pump C O o g To RHR ____ JL [ AA Jf Steam generator B w To p letdown Reactor j( Reacter A coolant core U pump B To Steam pressurizer d __. generator A , Reactor I l coolant I I pump A n
~
11 i a_ 1o v, - sample INEL-A-10 974 system
Figure 3.2 is a diagram of the CVCS purification system and applies to both Unit #3 and Unit #4. Table 3.1 is a summary of CVCS , I principal component information which also specifies which components are shared by Unit #3 and Unit #4. : Reactor coolant leaves the reactor coolant system by way of a letdown line~ located on the discharge side of reactor coolant pump B and enters the chemical and volume control system through the shell side of the
- regenerative heat exchanger. Coolant leaving the heat exchanger then passes through one of three letdown orifices, two of which are sized to allow 60 gpm flow and the third 45 gpm flow. The coolant then leaves containment and in the auxiliary building undergoes a second temperature reduction in the tube side of the non-regenerative heat exchanger. Component cooling water flows through the shell side of the non-regenerative heat exchanger, and the letdown stream outlet temperature is controlled automatically by a temperature control valve in the component cooling water outlet stream. The coolant is next further reduced in pressure by the low pressure letdown valve and is routed to one of two mixed-bed purification demineralizers.
The demineralized reactor coolant flows through the reactor coolant filter and enters the volume control tank through a spray nozzle. The filter, which houses three disposable synthetic filter elements, is l designed for 98% retention of particulates larger than 25 microns. Hydrogen i is supplied to the vapor space in the volume control tank for the purpose of removing oxygen from the reactor coolant. Periodically the tank is vented to the waste gas decay tanks to remove fission gases from the reactor coolant. Excessive rise of the volume control tank water level is prevented by automatic actuation of a three-way diversion valve, which routes the reactor coolant letdown flow to the holdup tanks. The charging pumps return the coolant from the volume control tank to the reactor coolant system through the tube side of the regenerative heat exchanger. Also, the pumps supply high pressure water to the reactor coolant pump seals. Part of the flow enters the reactor coolant system through a labyrinth seal on the pump shaft and the remainder cools the lower radial bearing, passes through the seals, is cooled in the seal water heat exchanger, filtered, and returned to the volume control tank. Seal water inleakage to the reactor coolant system is compensated by continuous letdown of the reactor coolant system. The makeup system supplies water to the reactor coolant system to compensate for normal reactor coolant system leakage, to adjust reactor coolant boron concentration, and to control reactor coolant chemistry. Makeup water to the reactor coolant system comes from the following sources: (1) the primary water storage tank for dilution; (2) boric acid tanks for boration; (3) refueling water storage tank for emergency makeup; (4) and the chemical mixing tank for hydrazine or pH control chemical addition. Small quantities of boric acid solution are metered from the discharge of a boric acid transfer pump for blending with primary water as makeup for normal leakage or for increasing the boron concentration. 40 I ' - - - - . . _ _ . _
5 F..ure 3.2 Chemical and Volume Control System. Turkey Point Plant Units #3 and #4 i I [' ; I Non-regen HX
, , ,- ei Dets atmg Cation g i" Muod bed
, H )Il e s ,qh, demar-shr -
dommersluer commershmers
-----g 5 '
( ) * * * "I g-- -- 1 L J re To GA 0 From h
' o.
s- To WOS Gas
} s' n
6 RHR HX outlet _- g a (K (-
- 1 i g n J i
I $ k i i Regenerateve g heat enchanger
) Seal water HX ,g Letdowet from loop B - _
t g cad kg _e, t g g 1 ' 1 Cation A B
,, , Seat water I p,-_g ,1 a
To loop A Cold leg O return i \/ e filters 1 I g To PRZ aux 3"ay g D$- g 2 l g g To ioop C het seg as-D(- # l
! 2 2*--
p.,c,o, i i --*1 1 I I I g 0 coolant foter To lg cold legs e-, [, To H T. l4 4 -- L - -- J lo - J -----J
& B A E. ,,
Seal water I ' ' ' RCP sea, ,etur, L miect,- ' a- nor'c l . -- - % , , - , . 2 - T ~ 't+r* i L----wa' s 'c"8""*P' fi y l
-e---_--_-----
- - " l" 3'"*" a ' H $, L - an- *v - f-sto,.ge tan.
a , , TT: l l I--l ll ~~=:~ TI A si ; ] _ l y r'*------- 1 a From RWST l Chemicas and Volume Control System To RCP C - 4
*~~~~ Units 3 and 4 3
lB l l r el ~ emergency rnakeup l To RCP B - C Le - - - - - - - e-------------* 1 To RCP A :W lC l: ! r*! * --- From chem pot unit 4 g, , Charg.ng pumps L t t
TABLE 3.1 PRINCIPA!. CVCS COMPONENT NATA
SUMMARY
Heat Letdown Letdown Design Design
- y Transfer Flow C Pressure Tc=perature Quantity S tu/hr lb /hr -F ps12,shell/ tube F, shell/ tube Heat Exchangers 6
Regenerative 1 8.65 x 10 6 29,826 265 2485/2735 650/650 Non regenerative 1 14.8 x 10 6 29,826 163 150/600 250/400 Seal water 1 2.17 x 10 126,756 17 150/150 250/250 l Excess letdown 1 4.75 x 10 6 12,400 360 150/2485 250/650 Capacity Design Design 1 Each Pressure Te=perature a Quantity M gpm Head psig y Pumps Charging 3 Pos. displ. 77 2385 psi 3000 250 Boric acid transfer 4* Canned 60 235 ft 150 250 3 Holdup tank recirculation 1* Centrifugal 500 100 ft 150 200 I 2*
?.onitor tank Centrifugal 100 150 ft 150 200 Concentrates holding tank transfer 2* Canned 20 150 f t 75 250 Gas stripper feed .3* Canned 25 185 ft 150 200 7
- Cas stripper bottom 2 Canned 12.5 93 ft 75 300
, Design Design j y Pressure Temperature
; quantity Type Volume, Each psig 7 a
Taaks 3 Volu=e control 1 Vert. 300 ft 3 75 Int /15 Ext 250 Charging pump accum. 3 Vert. 100 in 3000 250 Boric acid 3* Vert. 7500 gal Atmos. 250 Chemical edxing i Vert. 6.0 gal 150 250 Batching 1* Jacket Bta. 800 gal 3 Atmos. 250 Holdup 3* Horizontal 13,000 ft 15 200 1
TA3LE 3.1 (Cont'd) Design Design
, Pressure Temperature
^
Quantity Type Volume psig F Tanks (continued) Concentrates holding 1* Vertical 925 gal Atmos. 250 Monitor 2* Diaphragm 10,000 gal Atmos. 150 Resin Design Design y Flow Pressure Tenperature Vo}uma Quantity Type ft E psig F Demineralizer Vessels Mixed bed 2 Flushable 30 109 200 250 Cation bed 1 Flushable 20 60 200 250 3ase and cation ion exchangers 3* Flushable 30 25 150 250 a Evaporator condensate 2* Fixed 30 25 200 250 Deborating 2 Fixed 43 120 200 250 1 Quantity per unit unless othetvise specified. Shared or capable of being shared by Unit 3 and Unit 4. m
i f ! Two flushable mixed-bed demineralizers serve the CVCS to remove fission and corrosion produc'.s from the reactor coolant. They are placed in parallel in the leidown stream at the head of a demineralizer train that also includes one :ation demineralizer and two anion deborating demi neralizers. The cation dimineralizer is charged with resin in the hydrogen form and is used inte mittently to control the concentrations , of lithium and cesium. Input t o the cation demineralizer is the effluent i from one of the mixed-bed demir.tralizers. When a deborating demineralizer is in service, which would normall,' be near the end of a core life, it receives i the effluent of a mixed-bed demineralizer and passes deboratcd recctor i coolant to the volume control tink via the reactor coolant filter. The purpose of the reactor coolant iilter is to protect against resin fines. Each mixed-bed demineraliztr holds 30 ft3 of Li-0H resin, and each l bed has a surface area of 5.4 ft2 and a depth of 5.5 ft. Turkey Point I purchases resins from three vendors and sometimes mixes resins from different vendors when charging the mixed-bed demineralizers. Mixed Bed 'B' of Unit #3 was charged on 6/8/77, and put in service 1/20/78. Mixed l Bed ' A' of Unit #4 was charged on a date unknown and put in ser" ice 5/10/77. 1 These mixed beds remained in service throughout the sampling period. 1
- l 3.1.2.2 Sampla 'oints '
4 . The available valid sample locations in the CVCS purification systems of Units #3 & #4 were (1) downstream from the non-regenerative heat exchanger; upstream from the three-way diversion ) valve which is used to divert letdown to the holdup tanks; upstream from the purification demineralizers train and (2) downstream from the purification demineralizer train; upstream from the reactor coolant filter. No valid sample paint was available downstream from the reactor coolant filter of i either unit. Liquid samples of the letdown streams of Units #3 & #4 were taken at their respective reactor coolant sample sinks. 3.1.2.3 Sample Collection i l Sample lines were purged for a minimum of 10 ninutes prior to collection of mixed-bed demineralizer inlet and outlet samples. Usually less than two minutes elapsed between inlet and outlet collaction times. Inlet samples were collected in 50 ml volumes on all sample dates and during December, 1977, 800 m1 samples were taken concurrently for the purpose of resin concentration. In addition, 450 ml inlet samples were takea concurrently during April, 1978. Volumes of outlet samples were 450 ml with the exception of those samples taken in December, 1977, and February,1978, which had volumes of several liters and were taken at the same time for the purpose of resin cor: centration. 3.2 Discussion of t'casurement Data - Reactor Coolant 3.2.1 Radionuclide Concentrations in Reactor Coolant Reactor coolant samples were obtained from Unit #3 during the period from 11/9/77 to 6/1/78. This period encompassed the three 44 t
~ ,.r.__.,.m _ . - - ._ _. ~ -_
i stages of Unit #3 operation - power generation prior to refueling, refueling operations, and power generation following refueling. During
- power operations, samples were obtained by the method described in i Sectic,,3.1.1. During refueling, however, samples were obtained either through the RHR system or directly from the reactor cavity (i.e., a dip sample) Results of analyses of these samples are contained in Appendix Ta. les B.1, B.2, B.3, B.5, B.6, B.8, and B.11.
Figure 2.1 shows plots of the power level and 1311 concentration in reactor coolant for Unit #3. Data obtained by FPL are included. It should be noted that the dashed lines are not intended to indicate the 131I concentration between measurement points (e.g., the indicated shapes of most spikes are inaccu 7te due to the limited number of data points). Their purpose is to aid in determining trends. Examination of 4 Figure 2.1 indicates that the 1311 concentration in Unit #3 reactor i coolant spiked not only upon reactor shutdown and startup but each time the power was altered. A small spike was even observed after a power reduction from 100% to 90% on 11/19/77. If spikes are ignored, the 4 average 1311 level before refueling was about 1.5(-2) pCi/ml and about 7{-3) pCi/ml after refueling. No attempt was made to obtain an average 131! concentration that included the contribution due to spikes. In order to obtain such a time-weighted average, all spikes must be well characterized as to intensity and duration. This information, however, was obtained only for two spikes ('.ee Section 3.2.3). Table 3.2 lists average (i.e., arithmetic mean) radionuclide concen-trations in Unit #3 reactor coolant measured during non-spiking periods i prior to refueling and after refueling. The methods used to obtain these averages and others in this report are discussed in Appendix A. No attempt was made to obtain averagt radionuclide concentration for the refueling period because of the large variations in concentration. These variations are due to several factors including (1) radioactive I decay, (2) the addition of 4 gal. of peroxide to the reactor coolant at shutdown to dissolve crud, and (3) the mixing of reactor coolant with water from the fuel pool and refueling water storage tank for Unit #3. An indication of the large reduction in concentration of the short-lived radionuclides due to reactor shutdown and refueling can be obtained by a comparison of the concentrations obtained during refueling (Table B.2) with the corresponding data obtained before and after refueling
; (Tables B.1 and B.3).
Inspection of Table 3.2 indicates that as a result of refueling the
- iodine concentration was reduced by about a factor of two, rubidium decreased by about 50%, and cesium decreased by approximately a factor r of two. No trend in the concentrations of crud-associated radionuclides
- is discernible. A reduction in iodine, rubidium, and cesium levels after refueling is exincted because the new fuel should contain fewer leaks than the older fuel that it replaced. Examination of the iodine l ratios before and after refueling (see Tables 3.3 and 3.3A), indicates l that in both cases the release mechanism for fission products appears to i
l t 45 l
TABLE 3.2 AVERAGE RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT REACTOR POWER OPERATIONS (NON-SPIKING) - UNIT #3 Measured Prior to Refuelin[ Measured After Refueling " 11/1/77 - 11/22/77 2/21/78-6/1/78 Mean Range Mean Range Nuclide (pCi/ml) (uti/ml) (uCi/ml) (uCi/ml) amKr ** ** 1.8(- ) 1.5-2.3(-2) asKr <4( 5 t rt l 87Kr 3.4 - ) 2.6-4.1(-2) , seKr 4.7 - ) 3.9-6.3(-2) ) 131mXe
<2( 3) ttt 133mXe 4.3 - 1.2-8.7(-3) 133Xe 2.0 - 0.5-3.1(-1) ;
135mXe 1.1 - 0.5-3(-1) ' 13sXe ** ** 1.5 - j ** ** 1.2-1.9(-1) 137Xe 3(-2) 2-4(-2) l 13sXe 1.1(-1) 0.9-1.3(-1) 64Br ** ** 1.6(-2) 1.4-2.2(-2) 131I 1.5(-2) 1.1-1.8(-2) 6.9(-3) 4.3-8.8(-3) 132I 1.9(-1) 1.7-2.2(-1) 9.5(-2) 0.4-1.15(-1) i 133I 1.2(-1) 1.1-1.3(-1) 5.6(-2) 5.3-6.1(-2) 1 134I 3.4(-1) 3.3-3.5(-1) 1.8(-1) 1.63-1.97 -1
- lasI 2.1(-1) 2.0-2.2(-1) 1.0(-1) 0.9-1.09((-1))
88Rb 1.0 - 0.8-1.3 -1 7.0 - 5.8-8.0(-2 89Rb 1.0 - 0.9-1.1 - 6.1 - 4.8-7.6(-2 134Cs 1. 2 - 0. 3-1.6 - 6. 2 - 0.05-1.85( ) 136Cs 1.5 - 1.1-2.2 - 1.1 - 0.35-5.6(- 137Cs 2.1 - 0.5-2.7 - 8.5 - 0.06-2.6 -3 l 13eCs 3.0-1) 2.8-3.3 - 1.8 -1 1.6-2.48 -?
; 139Cs 2.8 -1) 2.2-3.3 - 1.5 -1 0.3-2.6(1) 3H 1.3(-1) 0.57-1.9(-1 ) 1.7(-1) 0.41-3.2(-1) 41Ar ** **
5(-3) <0.05-2.3(-2)
, 14C 4.6(-5 2.5-6.7(-5) 5.8(-3) 3.0-7.0(-3) 7.4(-3 24Na 0.24-1.26(-2) i 32P ** ** 3.7 - 3.0-4.37(-3) 4 51Cr 5(-5) 0.2-1.7(-4) 1.5 - 0.6-5.3(-4)
- 54Mn 1.4(-5) 1.0-1.9(-5) 2.9 - 0.02-1.08(-4) l ssFe ** ** 1. 5 - 0.97-2.1 -5)
- 59Fe 1.0(-5) 0.9-1.1(-5) 9 -6) 0.25-2.6-5) 5700 9(-6) 0.6-1(- ) 2 -6) <0.7-5.3 -6) seCo 4.7(-4) 0.1-1.5 -3) 4-4) 0.01-1.16(-3) 60C0 1.2(-4) 0.6-3.3**
-4) 2 -4) 0.04-6.8(-4) 63Ni 5.3(-6) 1.9-9.1(-6) 65Zn 1.3(-5) 1.3-2.0(-5) 4(-6) <1-9(-6) 893p ** ** ],9(.5) 1.76-2.11(-5) 90Sr ** **
2.1(-7) 0.17-4.1(-7) 46
, n- -, e,--- - -.- ,y- ----
m-a--- , - - - -- ---r e -w-
k TABLE 3.2 (cont'd) AVERAGE RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT REACTOR POWER OPERATIONS (NON-SPIKING) - UNIT #3 Measured Prior to Refueling t Measure'd After Refueling ii 11/1/77 - 11/22/77 2/21/78-6/1/78 Mean Range Mean Range luclide (uCi/ml) (uCi/ml) (pCi/ml) (pCi/ml) 91Sr 8(-4) 0. 3-1,3(- 3) 3 2(-4) *** 91Y ** ** 5.85(-7) 5.8-5.9(-7) 9 3Y 2. 3 - 2-5(-3) 2(-3) 0.9-5(-3) 95Zr 2.5 - 1.4-6.5 - 3.9 - 0.09-1.07 - 95Nb 3.5 - 2.4-6.9 - 5.0 - 0.05-1.28 - 99Mo 1.2 - 0.2-1.6 - 9.3 - 0.13-1.52 - 103Ru 1.6 - 1.6-2.4(-5) 1.6(-5 0.4-5.7(-5) lo6Rh *** 6 2(-5) 2.3(-5 0.05-1.0(-4) 110 mag <2(-4) ttt 6(-6) 0.18 ' .1(- 5) 124Sb 7(-6) 0.6-2.0(-5) 9(-6) 0.19 .d(-5) 12sSb 1.51.4(-5) *** 9(-6) 0.4-2.l(-5) 129mTe 5(-6) *** 5 5 3(-5) *** 129Te <5(-3) ttt 2.1 0.2(-2) ' 131mTe 3 2(-4) <2(-4) ttt 131Te ** ** *** 7 1(-3) 132Te 2(-5) 0. 3-4(-5) 3(-5) 0.16-1.3(-4) 139Ba 1.l(-2) 0.6-1.7-2) 1.5(-2) 0.3-1.9(-2) 140Ba 9(-5) 0.4-1.9 -4) 2.l(-4) 0.16-6.0(-4) 14cLa 2(-4) 0.2-7(- ) 3(-4) 0.01-2.3(-3) 141Ce 1.4(-5) 1.4-2(-5) 5(-6) 0.2-3(-5) 143Ce <1 - ttt <2(-5) ttt 144Ce <9 - ttt 1(-5) <0.4-3( -5) 152Eu <2 - ttt <4 (- 6 ttt 154Eu <5 -6) ttt <2(-6 ttt issEu ** ** *** 2 1 -6) 187W <7(-5) ttt 1.5(-4) 0.8-3.0(-4) 2 39Np 7(-5) 0.6-1.l(-4) 5(-5) <0.2-1.8(-4)
- - Analysis not performed for radionuclide
- - Radionuclide detected in only one sample, therefore, a range of measured concentrations was not obtained.
- - Analysis not performed for radionuclide
- Data obtained from the following sample.s (see Appendix Table B.1) 1037,11/14/77 1735, 11/18/77 1623,11/16/77 1710, 11/21/77 1158, 11/17/77 131I data includes the above samples and the following FPL samples (seeAppendixTableB.8) 0200,11/1/77 0936,11/22/77 0105,11/15/77 3H data obtained from FPL samples obtained during period 11/3-21/77 (see Appendix Table B.8) 47
1 TABLE 3.2 (cont'd) AVERAGE RADIONUCLIDE CONCENThATIONS IN REACTOR COOLANT REACTOR POWER OPERATIONS (NON-SPIKING) - UNIT #3 tt - Data obtained from the following samples (see Appendix Table B.3) 1316, 2/21/78 1006-1009,4/25/78 1133, 2/23/78 0918,5/1/78 0805-0806,4/9/78 1013, 5/9/78 1824, 4/12/78 0951, 5/16/78 1126, 4/13/78 0923,5/25/78 1123,4/15/78 0938,6/1/78 1000-1004,4/17/78 1 131I data includes results from the above samples and the fol10 wing FPL samples (see Appendix Table B.8) i 2/ 21 / 78 0835,5/2/78 ; 0800, 3/21/78 5/9/78 2120,3/21/78 0849,5/16/78 0815,4/4/78 1840,5/23/78 4/11/78 0235,5/24/78 l 0835, 4/18/78 0855, 5/30/78 Beta-only-emitting radionuclide data obtained from samples (see Appendix Table B.6) 1010,4/25/78 0940,6/1/78 3H data includes FPL sam 2/20-5/29/78 l (see Appendix Table B.8)ples obtained during period ttt - Radionuclide was not detected (only a detection limit was obtained). l A range of measured concentrations, therefore, was not obtained. i l l l l l 48
'I be dominated by recoil. The majority of the fission products are, therefore, coming from tramp uranium. Examination of the iodine levels i in the reactor coolant before and after refueling indicates that refueling ! had reduced the magnitude of the source of fission products by about a factor of two. Hence, much of the tramp uranium remained in the core l after refueling. Reactor coolant samples were obtained from Unit #4 from 12/2/77 to 5/23/78. During this period, the reactor was down for steam generator repairs for about 3 weeks (2/14 - 3/9/78). Results of analyses of samples obtained from Unit #4 can be found in Appendix Tables B.4, B.7, B.9, and 8.10. Alpha analysis of a composite of samples obtained between 12/12/77 and 1/25/78 yielded a 23ePu concentration of 5.4 + 1.6(-9) pCi/mi i and a 239,240Pu concentration of 4.3 + 1.4(-9) pCi/ml. Figure 2.2 shows
, plots of the power level and 131I concentration in reactor coc,lant for Unit #4. As in the corresponding plots for Unit #3 (figure 2.1), FPL data has been included.
i
! TABLE 3.3 IODINE RATIOS IN UNIT #3 REACTOR COOLANT Ratio Ratio Before Refueling After Refueling 1331/131I 8.0 8.1 j
135171311 14,o 14.6 132g71311 12.7 13.8 134 If131I 22.7 25.8 TABLE 3.3A i THEORETICAL IODINE RATIOS DUE TO REC 0ll, DIFFUSION, > AND EQUILIBRIUM FUEL FAILURE MECHANISMS Theoretical Ratio
- for Ratio Recoil Diffusion Equilibrium 133 16 5.1 1.7 1f1311 135yflail 29 5.4 1.0 13217131I 26 2.9 0.31 134 If131I 54 3.7 0.25 l _ _ _ _
- Ratios are based on models in reference 8.
i l l l l I 49 l
t
\
l Examination of Figure 2.2 indicates that the 131I concentration in Unit #4 spiked whenever the reactor power level was altered. Units #3 and #4, therefore, exhibit similar spiking characteristics. In addition to spikes due to power changes, a higher than nonnal iodine level of long duration was observed in samples obtained during the period 4/9-14/78. This elevation was caused by the letdown demineralizer for Ur.it j #4 being bypassed for change out of the letdown filter. Although it is not known exactly when bypass of the demineralizer began, the shape of ; the elevation indicates that it probably began on 4/7 or 4/8/78. Because of the letdowa demineralizer bypass, average radionuclide ' concentrations in Unit #4 reactor coolant were obtained for three i periods - before the bypass of the letdown demineralizer, during the i i 4/9-14/78 period of bypass, and after the demineralizer had been put back into service. These average concentrations (see Table 3.4) do not incluoe data obtained during spikes due to changes in reactor power. i Results from the sample obtained at 09:27 on 12/12/77, however, are ] included because this sample was obtained during a non-spiking period. This sample contained much higher levels of crud-associated radionuclides l j than other samples obtained from Unit #4 although iodine, cesium, rubidium, and sodium concentrations were normal. Although these anomalously high j levels of crud-associated radionuclides certainly influence the averages i j and ranges given in Table 3.4, they were inclu N d because the sample was valid. In addition, FPL personnel indicated " t they have observed similar " crud bursts" in reactor coolont at Turkey Point. Examination of the average concentrations in Table 3.4 measured before and after the 4/9-14/78 bypass of the letdown demineralizer indicate increases in the levels of iodine cesitsn, and most crud-associated
- radionuclides. . For example, the average 1511 concentration exhibited a 23% increase. Concentrations of shorter-lived iodines increased by about 50% and 134Cs and 137Cs levels increased by about a factor of two.
ftost crud-associated radionuclides exhibited increased concentrations also. As expected, fission gas levels were essentially the same before and after the bypass of the demineralizer. It is not known why the 1 average iodine, cesit"n, and crud-associated radionuclide levels showed increases after the period of demineralizer bypass. It would be expected that the levels of these radioisotopes would return to pre-bypass values l soon after the demineralizer had been put back into operation. , An indication of fuel quality can be obtained from the iodine ratios. Table 3.5 lists the iodine ratios before and after the 4/9-14/78 demineralizer bypass. Examination of these ratios indicates that the release mechanism for fission products appears to be dominated by diffusion both befcre and after the demineralizer bypass (see Table 3.3A). It is not known why the iodine ratios were higher after the i demineralizer bypass than before it. Mere bypassing of the letdown demineralizer would not be expected to cause this. A comparison of the reactor coolant iodine activities for the two l reactor units indicates that after refueling the release rate for 1311 l 1 l l 50 !
TABLE 3.4 AVERAGE RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT REACTOR POWER OPERATIONS (NON-SPIKING) - UNIT #4 Measured During Period t Measured During Period it Measured During Periodttt 11/12/77-3/28/78 4/9-14/78 Mean Range 4/17-5/31/78 Mean Range Mean Range Nuclide (uCi/ml) (uCi/ml) (uci/ml) (uCi/ml) (uCi/ml) (uCi/ml) asmKr 1.7(-2) ** esKr 1.6-1.9(-2) ** 1.7(-2) 1.5-1.91(-2)
<2(-5) tttt ** **
81 e 2.3(-2) ** <8(-4) tttt 8 0.. 2.1-2.4(-2) ** 2.5(-2) 2.1-2.8(-2) 3.4(-2) 3.2-3.8(-2) ** ** 3.8(-2) 131 axe 4 1(-4) *** ** ** 3.2-4.43(-2) 133mXe 6.7(-3) ** 2(-4) 0.4-4(- ) 5.3-7.9(-3) ** 6.6(-3) 5.7-6.0 -3) 133Xe 2.1 - 1.3-2.7(-1 ** ** 1.9 - 2.2(-1 1.9-3.0 -1 3 135mXe 13sXe 1.2 - 1.7-2(-2) ) ** 2.3(-2 <2-3.2( 2))
- 1.02-1.49(-1) 1.2 -1 1.1-1.4 -
137Xe ** ** 138Xe 1.2 -2 0.6-1.9 - 3.3(-2) 2.3-4.4(-2) ** ** 4.0 -2) 2.9-4.7 - e4Br ** ** 131I 1.9(-3) 1.9-2.0(-3) 3.3(-3) 2.0-5(-3) 6.5(-3) 0.36-1.0(-2) 8.9(-2) 0.66-1.01(-1) 8.0(-3) 132I 1.3(-2) 0.99-1.46(-2) 0.63-1.03(-2) 133I 2.2(-2) 1.89-2.45(-2) 1.9(-2) 1.83-2.06(-2) 1.2(-2 0.99-1.66-2) 4.7 - 3.3-5.2(-2) 1.7(-2) 134I 1.4(-2 1.09-1.72 -2) 2.1 - 2.01-2,18(-2 1.68-1.8(-2) 135I 1.2(-2 2.2(-2) 1.96-2.43(-2) 1.09-1.44 -2) 2.7 - 2.0-3.05(-2)) 1.9(-2) 1.80-1.98(-2) serb 5.5(-2) 4.6-7.4(-2) 6.2(-2) 5.0-8(-2) 5.8(-2) 5.1-6.6(-2) 89Rb 9.5(-3) 0.68-1.4(-2) 1.0(-2) 0.8-1.2(-2) 1.4(-2) 134Cs 9.4(-4) 0.44-1.49(-3) 1.14-1.63(-2) 1.7(-3) 1.73-1.8(-3) 2.C(-3 1.61-3.l(-3 136Cs 3.9(-5) 1.5-9(-5) 2.1(-4) 137Cs 1.8(-3) 1.2-2.9(-3) 3.2(-3) 1.45-2.7(-4) 3.8(-5 3.5 -3 2.4-5.2(-5)) 13eCs 3.06-3.44(-3) 3.0-4.5(-3) 3.9(-2) 3.6-4.4(-2) 5.5(-2) 4.7-7(-2) 5.8 -2) 139Cs 1.3(-2) 0.01-5(-2) 5.0-7.8(-2) 1.8(-2) 1.5-2.3(-2) 1.3 -2) 0.2-4(-2) l 6.*
TABLE 3.4 (cont'd) AVERAGE RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT REACTOR POWER OPERATIONS (NON-SPIKING) - UNIT iF4 ttt Measured During Periodi Heasured During Period tt Measured During Period 11/12/77-3/28/78 4/9-14/78 4/17-5/31/78 Mean Range Mean Range Mean Range Nuclide (uCi/ml) (uCi/ml) (uCi/ml) (uCi/ml) (pC1/ml) (vCi/ml) 3H 1.3(-1 0.15-2.5(-1) 2.1(-1) *** 2.3(-1) 1.6-3.2(-1) 41Ar 6(-4) ) 0.5-1.05(-3) ** <2(-4) tttt 14C 1.9 -6) 1.22-3.2(- 5) ** ** 9.5 i 1.0(-6) 24Na 5.6 -3) 4.2-7.6(-3) 2.2(-2) 1.64-2.43(-2) 9.3(-3) 0.84-1.16(-2)
** ** 8.2 0.2(-3) ***
32P 9.7 -5) 0.36-1.58 - *** slCr 4. -4) 0.004-4.0 - 1.7 1 0.6(-4) 6.3(-4) 0.06-1.8(-3) 54Mn 7(- ) 0.019-6.0 - 1.3(-5) 0.53-2.l(-5) 1.7(-4) 0.15-5.0(-4) ssFe 2.0 - 0.38-5.3(-5) ** ** 9.0 1 0.1(-4)
$ 59Fe 4.5 - 0.016-5.3 -4) 8(-6) 0.2-1.6(-5) 9(- 5) 0.06-3.2 -4) 57C0 1.1 - 0.13-8.0( 5) <3(-6) 7(-6) 0.13-1.9 -5) seCo 1.3 - 0.007-1.2 -2 7(-4) 0.26-1.14(-3) 2.9(-3) 0.26-8.5 -3) 60Co 1.2 - 0.003-1.0 -3 3.2(-5) 1.4-4.5(-5) 2.6(-4) 0.47-4.6 -4) 63Ni 5.9 - 0.125-1.49 ** ** 1.76 i 0.06(-4)
; 65Zn 8(-6) 0.16-4(-5)(- ) 3(-6) 2.9-3(-6) 1.9(-5) <0.2-2.5(-5) 89Sr 1.2(-5) 0.048-3.2(-5) ** 5.9 0.2(-6) 90Sr 2.4(-7) 0.2-6.3(-7) ** ** 4.3 0.5(-7) 91Sr 2 2(-4) *** <2(-4) 2 2(-4) ***
** ** 1.35 1 0.09(-6) 91Y 1.9(-7) 0.5-4(-7) ***
93Y 4(-4) 0.03-2.2(-3) 9 4(-4) <3(-4) tttt 95Zr 4.5(-5) 0.013-4.6(-4) 2.5(-5) 1.0-5(-5) 8(-5) 0.17-1.6(-4 9sNb 4.2(-5) 1.5(-5) 0.83-2.3(-5) 1.l(-4) 0.11-2.2(-4 i 99Mo 1.7(-4) 0.02-4.4(-4)) 0.001-1.08(-3 1.3(-4) 1.2-1.5(-4) 1.3(-4) 0.71-1.7(-4 103Ru 0.10-3.5(-5) 4(-6) 3.5-9(-6) 2.2(-5) 0.6-6(-5) 106Rh 8(-6))
<4(-6 tttt <2(-5) <2(-5) tttt 110 mag 3.4 0.8(-5) *** 5(-6) 2.7-7(-6) 1.8(-5) 0.5-3(-5)
o TABLE 3.4 (cont'd) AVERAGE RADIONtiCLIDE CONCENTRATIONS IN REACTOR COOLANT REACTOR POWER OPERATIONS (NON-SPIKING) - UNIT #4 Measured During Period i Measured During Period tt ifeasured During Period *II 11/12/77-3/28/78 4/9-14/78 4/17-5/31/78 Mean Range Mean Rarge !!ean Range Nuclide (uC1/ml) (uC1/ml) __ (pCi/ml) (pCi/ml) (pCi/ml) (uCi/ml) 124Sb 1.9(-5) 0.008 - 1.5(-4) 1.3(-5) 0.43-2.6(-5) 3.9(-5) 0.4-8.3(-5) 12sSb 8(-6) 0.4-3.0(-5) <8(-6) tttt <6-6) tttt 129mTe <3(-6) tttt <6(-5) tttt <3 -5) tttt 12sTe <6(-3) tttt <4(-3) tttt <1 -2 tttt
!31mie <4(-5) ' tttt 1.6 0.8(-3) ***
131Te ** ** <2(- tttt
<9(-4) tttt <8(- tttt 132Te 3(-5) 0.4-5(-5) 1.9 0.9(-5) ***
3(-5 <0.07-1.l(-4) 139Ba 4.5(-3) 2.5-7.6(-3) 6(-3) 3.8-9(-3) 6.0(-3) 4-8(-3)
$ 140Ba 1.l(-4) 0.15-4.0(-4) <2(-5) 140La tttt 2.3(-5) 0.04-7.6(-3) 5(-5) 0.05-2(-4) <1(-5) tttt 9(a) <0.03-3.4(-3) 141Ce 6(-6) 0. 3-2. 5 (- 5) <6(-6) tttt 1 1(-5) ***
143Ce 4 1 4(-5) ***
<4(-5) tttt <6(-5 tttt 144Ce 7(-6) 0.15-3(-5) <3- tttt <2 - tttt 152Eu <4(-6) tttt <3 - tttt <2-
*** tttt 154Eu 5 3(-6) <2 - tttt <2 - tttt 155Eu <9 - tttt .<6(-6 tttt 187W 1(-4) <0.2-5(-4) <2- tttt <2(-4 tttt 239Np 4(-5) 0.05-1.6(-4) <5- tttt 6 4(-5) ***
* - Radionuclide not detected
** - Analysis not performed for radionuclide'
*** - Radionuclide detected in only one sample, therefore, a range of measured concentrations was not obtained.
_ _ . _ _. . . - _ _ _ _ _ _ _ _ . . . _ , _ _ _ . . _ __ = _ _ . ._ TABLE 3.4 (cont'd) AVERAGE RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT REACTOR POWER OPERATIONS (NON-SPIKING) - UNIT #4 t - Data obtained from the following samples (see Appendix Table B.4) 0945,12/8/77 1127, 1/19/78 0943, 2/3/78 1052, 12/11/77 1305,1/20/78 0950, 2/7/78 0927, 12/12/77 1116,1/22/78 0955, 2/8/78 1618,1/7/78 1114,1/23/78 1104, 2/9/78 2030,1/11/78 1031,1/24/78 1036,3/17/78 1127,1/18/78 0414,1/25/78 1916-1935,3/25/78 1311 data includes the above samples and the following FPL samples (see Appendix Table B.7)
- 11/15/77 1/17/78 3/21/78 0835,11/29/77 0825,1/31/78 0900, 3/28/78 E 0025, 12/13/77 0832, 3/14/78 Beta-only snitting radionuclide data obtained from samples (see Appendix Table B.7) 1610,11/30/77 0929,12/2/77 0933,12/12/77 3H data includes FPL samples obtained during period 11/7/77-4/3/78 (see Appendix Table B.9) tt - Data obtained from the following samples (see Appendix Table B.4) 0903-0904,4/9/78 1138, 4/13/78 1355,4/14/78 i 1753,4/12/78 131I data includes the above sampia and the following FPL sample (see Appendix Table B.9) i 0919,4/11/78 3H data obtained from the following FPL sample (see Appendix Table B.9) 4/10/78
TABLE 3.4 (cont'd) AVERAGE RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT REACTOR POWER OPERATIONS (NON-SPIKING) - UNIT #4 ttt - Data obtained from the following samples (see Appendix Table B.4) 1045-1046,4/17/78 1012,5/16/78 0909, 5/1/78 1030,5/23/78 0921,5/9/78 1311 data includes the above samples and the following FPL samples (see Appendix Table B.9) 4/18/78 0833,5/23/78 0920, 5/9/78 0405, 5/31/78 5/16/78 g Beta-only-emitting radionuclide data obtained from the 1025,4/24/78 sample (see Appendix Table B.7) 3H data includes FPL samples obtained during period 4/17-5/29/78 (see Appendix Table B.9) tttt - Radionuclide was not detected (only a detection limit was obtained). A range of measured concentrations, therefore, was not obtained.
TABLE 3.5 I0 DINE RATIOS IN UNIT #4 REACTOR COOLANT Ratio Before Ratio After 4/9-14/78 4/9-14/78 1 Demineralizer Bypass Demineralizer Bypass l l 1331 1311 1.8 2.1 ] 135 1 131I 1.8 2.4 1321 131I 2.0 2.4 134g f131I 2.1 2.7 I in Unit #3 was about the same as that in Unit #4. However, because the l release mechanism was predominantly recoil in Unit #3 and diffusion in Unit #4, even after refueling the release rates for the other iodines l were much higher in Unit #3 than in Unit #4. l A comparison of the average radionuclide concentrations in the two units indicates the following. In general, the levels of the gaseous fission products were slightly higher in Unit #3 reactor coolant. The levels of 133Xe, bowever, were about equal in the two units. The 1311 concentration was higher in Unit #3 prior to refueling but after refueling it dropped to about the level measured in Unit #4. Concentrations of < the shorter-lived iodine radionuclides were much higher (about an order of magnitude) in Unit #3 than in Unit #4. Unit #4 reactor coolant exhibited hi her lgels of crug-associated radionuclides (e.g., 51Cr, 5" Mn , s s pe , Fe, g9 -60Co, 110 Ag, 124Sb) than did Unit #3. The only major crud radionuclide that exhibited higher concentrations in Unit #3 than in Unit #4 was 60Co. The reason for this apparent anomalous behavior J of 60Co is not known. In general, the rubidium and the shorter-lived cesium concentrations were higher in Unit #3 reactor coolant. l i 3.2.2 Predicted Radionuclide Concentrations in Coolant Waters The American Nuclear Society, Standards Committee Working Group ANS-18.1 has prepared a set of typical radionuclide concentrations for use in estimating the radioactivity in the principal fluid streams of a light water reactor over its lifetime (9). Expected radionuclide activity levels in the primary and secondary coolants for Turkey Point I i can be derived from the ANSI N237-1976 values by adjusting the parameters l of the reference PWR to those of Turkey Point. Table 3.6 presents these l expected activity levels and Table 3.7 lists the parameters used for I adjustment of the reference PWR to Turkey Point. These parameters are for U-tube steam generators and assumes all volatile treatment (AVT) of l the secondary coolant. Although the techniques used to collect the data in the in-plant measurement study are capable of detecting all gamma-emitting radionuclides 56 i
TABLE 3.6 PREDICTED RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT AND SECONDARY COOLANTt i Coo a Water kam Nuclide (ut /gm) (pCi/rm), (pCi/gm)
* ":l; 1::: "11 :l(::
5:!: "11 Ml:
.; 1:!: "il Bl:
l'N: 133Xe M: 1.6 + n11 Nil 0: 7.1 - 135mXe 1.2 - Nil 5.0 - l'2" i:2: "il M: 138Xe 3.9 - Nil 1.7-8)
* :' ; U:81 H:y3 i:!l:p *ar 2:7- !:1:1) H:
l'!! B: i:::5 !::: 1:4 lasI 8: 1.6 - O:' 2.8 - M: 2.8 -
- l3:
134Cs 6:*: 2.0 - B: 1.9 - M: 1.9 -
)
136Cs 1.0 - 9.6 - 9.6 - 137Cs 1.4 - 1.5 - 1.5 .
- 1sN 4.0(+1) 1.7(-6) 1.7(-7) 3.4 1.0(0) 1.0(-3) 1.0(-3) 1 l4" ssFe 1:":
- 1. 3 -
U: 1.3 - U:'))
- 1. 3 -
59Fe 7.8 - 9.4 - 9.4 - ) seCo 1.3 - 1.3 - 1.3 - 60Co 1,6 ],7 1,7 9 89Sr 2.7 - 3.7 - 3.7 - 90Sr 7.8 - 9.4 - 9.4 -
*:W 91Y U:
5.0(-5 B: 5.6 - U: 5.6 - 57 l
TABLE 3.6 (cont'd) PREDICTED RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT AND SECONDARY COOLANT o a Water team Nuclide (uC1/gm)_ hCi/gm) (uC1/gm) l 95Zr i:i: 4.7 - i:*: 5.6 - l 1:": 5.6 - I l ssNb 3.9 - 5.6 - 5.6 - 99Mo 6.7 - 5.6 - 5.6 - 99mTc 4.1 - 5.3 - 5.3 - 10 3Ru 3.5-5) 3.7 - 3.7 -
- 106Ru 7.8 -ET 9.4 - 9.4 -
- 10 3mRh 4.0 - 3.2 - 3.2 - i lo6Rh 8.9 - 7.7 - 7.7 -
- 12smTe 2.3 - l .7 - 1.7 - i
- 127mJe 2.2 - I .7 - l.7 - l
- 127Te 7.l - 3.6 - 3.6 -
129mTe 1.1 - 1.1 - l.1 -9) 129Te 1.4 - 9.7 - 9.7 - 131mie 2.0 - 9.3 - 9.3 - 131Te 9.8 -4) 7.9 - 7.9 - 132Te 2.1 -2) 1.5 - 1.5(-8)
- 137mBa 1.4 -2) 1.2 - 1.2(-8) 140Ba I .7 -4) 1.7 - 1.7-10) 140La 1.2 -4) 1.5 - 1.5 -10) 141Ce 5.5-5) 5.6 - 5.6-11) 143Ce 3.2 -5) 1.7 - 1.7 -11) 1
- l"e i:M:ll 1:?: i:?: l
- 144Pr 2.9(-5) 3.1 - 3.1 -
239Np 9.6(-4) 5.6 - 5.6 - t - Prediction based on ANSI N237-1976 Standard (9)
* - Radionuclide listed in ANSI N237-1976 but not directly 1 measured at Turkey Point ! ** - Calculation assumes all volatile treatment chemistry ,
for secondary water ' 58
TABLE 3.7 PARAMETER VALUES USED TO MODIFY N237 PREDICTED RADIONUCLIDE CONCENTRATIONS FOR TURKEY POINT UNIT #3 OR #4 Turkey Point Value Parameter _ Symbol Unit Unit #3 or #4 Thennal Power P MWt 2200* ! Steam flow rate (all generators) FS lbs/hr 9.6(6)* Weight of water in reactor coolant system WP lbs 4.0(5)** Weight of water in all steam generators WS lbs 2.6(5)*** Reactor coolant letdown flow (purification) FD lbs/hr 3.0(4)* Reactor coolant letdown flow (yearly average FB lbs/hr 3.0(2)* for boron control) Steam generator blowdown flow (total) FBD lbs/hr 3.2(4)* Fraction of radioactivity in blowdown stream NBD ------ 1.0* which is not returned to the secondary coolant system Flow through the purification system cation FA lbs/hr 3.0(3)* demin'eralizer Ratio of condensate demineralizer flow rate NC ------ 0.0* to the total steam flow rate Ratio of the total amount of noble gases Y ------ 0.0* routed to gaseous radwaste frnm the purification system to the total amount routed from the primary coolant system to the purification system (not including the boron recovery systen) o Based on information obtained during measurement program 0* Information from FSAR (10) o** Information from NRC l l 59 l
l present in any sample, some radionuclides treated in ANSI N237-1976 were not observed at Turkey Point (cf the radionuclides denoted with an asterisk in Table 3.6). These radionuclides were not observed because they were either not present in detectable quantities, they have very short half lives, or they emit only very low energy gamma rays. For example, esRb and 1301 are present only in very small amounts because i of their very low fission yields (about 2(-3)% for each isotope). In addition, their gamma rays are masked by gamma rays of about the same
-energy emitted by more abundant radionuclides. The radionuclides 16 N, esBr, and 89K r have very short half-lives which precluded their detection by the measarement tech }" Rues utilized in the in-plant studies. The radionuclides 83Br and Q 'Kr could not be detected due tg tng very low gmga-raz06energ g Other radionuclides such as 90Y, 91 Y, Tc, Rh, Ru, Ba, 143Pr, and 144 Pr could not be detected due to very low gamma-ray energies or interferences from other radionuclides but they are parents or daughters of radionuclides that were measured. l The concentrations of these radionuclides can be estimated by assuming I that they are in equilibrium with their parent or daughter. This assumption is valid when the reactor is producing power under steady state conditions.
Although attempts were made to measure the concentrations of the tellurium isotopes, they have not been observed with any consistency at Turkey Point. Teg tum-129m was detected in only 2 samples,129Te in I sample, and e in 1 sample. Tellurium-132 was detected more frequently, but its concentration was about 3 orders of magnitude lower than predicted by ANSI N237-1976. There is evidence (see Section 3.2.3) that indicates that tellurium is fixed on the internal reactor core components. In Table 3.8 the radionuclide concentrations predicted for Turkey Point reactor coolant are compared with the average concentrations actually measured for Units #3 and #4. Except for 132I and 1341, the measured radionuclide levels are generally lower than the predicted levels. i I 3.2.3 Spiking Studies As noted in section 3.2.1, the data indicate that the iodine concentrations in the reactor coolant of both units spike whenever the reactor power is changed by about 10% or more. No data is available to indicate whether spiking occurs for power changes of less than 10%. In order to study the spiking phenomenon in more detail, reactor coolant samples were obtained on approximately an hourly basis during two hot shutdowns. i l On 1/25/78 Unit #4 underwent a hot shutdown that lasted almost 12 hours. Power reduction began at 04:00 and reached 0% just after 06:00. , Startup began at about 17:30 and 100% was reached at 19:00. Figure 3.3
. shows plots of the concentrations of 1311, 137Cs, and 58 Co during the .
shutdown. Data obtained during the spike can be found in Appendix Table 1 1 B.10. I l 60 l i l
. . -. . - - - = . ._ . .._ __ _--- . - . -- . . _ - - _ _ .
TABLE 3.8 COMPARIS0N OF PREDICTED AND MEASURED RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT Predicted Measured Concentration in Unit #3 Measured Concentration in Unit #4 Concentration Before Refueling After Refuelin9 After Bypass ***
- Nuclide (pC1/ml) (uC1/ml) (uCi/ml) Before (pCi/ml Byp) ass *** (uCi/ml) esmKr 9.8(-2) **
1.8(-2) 1.7(-2) 1.7(-2) esKr 1.6(-1) <a( 5) <2(-5) 8(-4) 87Kr 5.3(-2) ** 3.4 -2) 2.3(-2) 2.5(-2) eeKr 1.8(-1) 4.7 -2) 3.4(-2) 3.8(-2) 131mXe 1.0(-1) **
<2( 3) 4(-4)* 2(-4) 133mXe 2.0(-1) 4.3(-3) 6.7(-3) 6.6(-3) 133Xe 1.6 +1) **
2.0(-1) 2.1(-1) 2.2 - 135mXe 1.2 -2) 1.l(-1) 1.9(-2) 2.3 - 13sXe 3.1 -1) 1.5(-1 1.2(-1) 1.2 - 9 137Xe 8.0 -3) ** 3(-2)) t 1. 2 - 13 axe 3.9(-2) 1.l(-1) 3.3(-2) 4.0(-2 84Br ** 2.3(-3) 1.6(-2) ** 3.3(-3) 131I 2.1(-1) 1.5(-2) 6.9(-3) 6.5(-3) 8.0(-3) 132I 8.7(-2) 1.9(-1) 9.5-2) 1.3(-2) 1.9(-2) 133I 3.l(-1) 1.2(-1) 5.6 -2) 1.2(-2) 1.7(-2) 134I 4.2 -2 3.4 -1) 1.8 -1) 1.4(-2) 2.2(-2) 135I 1.6 -1 2.1 -1) 1.0 -1) 1.2(-2) 1.9(-2) serb 1,8(-1) 1.0 - 7.0 -2) 5.5 - 5.8 - 89Rb tt 1.0 - 6.1 - 9.5 - 1.4 - 134Cs 2.0 - 1.2 - 6.2 - 9.4 - 2.0 - 136Cs 1.0 - 1.5 - 1.1 - 3.9 - 3.8 - 137Cs 1.4 - 2.1 - 8.5 - 1.8 - 3.5 - 13eCs tt 3.0 - 1.8 - 3.9 - 5.8(-2 139Cs tt 2.8(-l 1.5(-l) 1.3(-2) 1.3(-2) 3H 1.0(0) 1.3(-1) 1.7(-1) 1.3(-1 ) 2.3(-1) 41Ar tt 5(-3) 6(-4) <2(-4) 1 l
TABLE 3.8 (cont'd) COMPARIS0N OF PREDICTED AND MEASURED RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT Predicted Measured Concentration in Unit #3 Measured Concentration in Unit #4 Concentration Before Refueling After Refueling Before Bypass *** After Bypass *** Nuclide (pCi/ml) (uCi/ml) (t.C i/ml) (uC1/ml) (uCi/ml) 14C ** tt 4.6 -5) 1.9 -6) 9.5 -
- 2i.Na tt 5.8(-3) 7.4 -3) 5.6 -3) 9.3 -
32P ** SICr tt 3.7 -3) 9.7-5) 8.2 -
- 1.5(-3 5(-5) 1.5(-4) 4.5 -4) 6.3 -
54Mn 2.4(-4 1.4 (- 5) 2.9(-5) 7(-5) 1.7(-4) 55Fe 1.3(-3 ** 1.5(-5) 2.0(-5 9.8(-4)* . 39Fe 7.8(-4 1.0(-5) 9 -6) 4.5 - 9(-5) 57Co 99 g(_6) 2 6) ],] . 7(_g) r seCo 1.3(-2) 4.7(-4) 4 -4) 1.3 - 2.9(-3) !
$ 60Co 1.6(-3) 1.2(-4) 2(-4) 1.2(-4) 2.6(-4) 63Ni **
tt 5.3(-6) 5.9(-6) 1.76(-4)* 6sZn tt 1.3(-5) 4(-6) 8(-6) 1.9(-5) , 89Sr 2.7-4) ** 1.9(-5) 1.2(-5) 5.9(-6)* 90Sr 7.8 -6) 2.1(-7) 2.d(-7) 4.3(-7)* 91Sr 5.5 -4) 8(-4) ** 3(-4)* 2(-4)* 2(-4)* 91Y 5.0(-5 5.8(-7) 1.9(-7) 1.35(-6)* 2.8(-5 i
- 93Y 2.3(-3) 2(-3) 4(-4) <3(A) i 95Zr 4.7(-5 2.5 -5) 3.9(-5) 4.5(-5) 8(-5) 95Nb 3.9 (-5 3.5 -5) 5.0(-5) 4.2(-5) 1.1(-4 99Mo 6.7(-2) 1.2 -3) 9.3 - 1.7(-4) 1.3(-
103Ru 3.5(-5) 1.6(-5) 1.6 - 8(-6) 2.2(- 106Rh 8.9(-6) 6(-5)t 2.3 - <4(-6) <2(-5 110 mag tt <2(-4) 6(-6) 3.4(-5)* 1.8(-5) 124Sb tt 7(-6) 9(-6 1.9(-5) 3.9(5) 12sSb tt 1.5(-5)* 9(-6 8(-6) <6 - l 129mTe 1.1-3) 5(-6)* 5(-5 * <3(-6) <3- ' 129Te 1.4 -3) <5{-3) 2.1(-2)* <6(-3) <1 - 131mTe 2.0 -3) 3(-4)*
<2(-4) <4(-5) <2 - ,
131Te 9.8 -4) 7(-3)* **
<8(-4 132Te 2.1(-2) 2(-5) 3(-5) 3(-5) <3(-5) 1
t 1 TABLE 3.8 (cont'd) ' COMPARISON OF PREDICTED AND MEASURED RADIONUCLIDE CONCENTRATIONS IN REACTOR C0OLANT Predicted Measured Concentration in Unit #3 Measured Concentration in Unit #4 Concentration Before Refueling After Refueling Before Bypass *** After Bypass *** Nuclide (uCi/ml) (uCi/ml) (uC1/ml) (vC1/ml) (uCi/ml)
'139Ba tt 1.1(-2 1.5(-2) 4.5(-3) 6.0(-3) 140Ba 1.7(-4 9(-5) ) 2.l(-4) 1.1(-4) 2.3(-5) 14cLa 1.2 -4 3(-4) 5(-5) 9(-4) 141Ce 5.5 -5 2(-4) )
1.4(-5 5(-6) 6(-6) 1(-5)* r 143Ce 3.2 -5) <1 - <2(-5) 4(-5)* <6(-5) 144Ce 2.6 -5) <9 - 1(-5) 7(-6) <2 -
<2 -
152Eu tt <4(-6) <2 -
<4(-6) i 154Eu tt <5 - <2(-6) 5(-6)* <2 -
155Eu ** ** w tt 2(-6)* <6 - 3 187W tt <7(-5) 1.5(-4 1(-4) <2-239Pu 9.6(-4) <7(-5) 5(-5) ) 4(-5) 6(5)* i i t - Radionuclide not detected I tt - Radion.tclide not listed in ANSI N237-1976 +
* - Radionuclide detected in one sample only
** - Measurement not made for radionuclide
*** - Measurements made during period before letdown demineralizer was bypassed ar.d period 4
after the letdown demineralizer was put back into operation. j l l
Figure 3.3 Shutdown Spike - Unit #4,1/25/78 1
- i i i i i -
' ~
, l
, . 131 \
..... - \
k i 10~ \ -
.I
\
.... \
C
- \
wmu* 1 E s ~ i U 58Co e / l ~s % g 10'2 - _ 137 N T l
/a a.-a. *Cs ~~ ,_% %, _ ,
f'
- - \
./
8 2 / 2 0 - ' %,g,' % ' ~~ - a .I j , N . , ,,, ' - . _
.f N - .
10-3 _ ._ j J 10'4 I I ' I I IM - i i i - i
; _ l
~
- l u
y _ _ o
- a. _ _
10 h I ' ' ' 0000 0600 1200 1800 0000 0600 1200 1/25/78 Time 1/26/78 INEL A-10 980 64
i i i
! As indicated in Figure 3.3 and by the data in Appendix Table B.10, j spiking for most longer-lived radionuclides reached a maximum at about
- 09
- 17 (short-lived radionuclides reached a maximum earlier), about 5 i hours after power reduction began and about 3 hours after 0% power was reached. Iodine-131, 137Cs, and seCo peaked at concentrations of 38, 4.5, and 31 times the pre-spike levels, respectively. Most detected fission products and crud-associated radionuclides showed spiking except i for the rubidiums. As expected, shorter-lived radionuclides exhibited spikes of smaller magnitudes (e.g., radioiodine inventory in the plenum is
, inversely proportional to half life) and soluble activation products (e.g., 2*Na) did not spike.
Unit #3 underwent a hot shutdown on 5/19/78 that lasted until 5/22/78. Power reduction began at 20:00 on 5/19/78 and 0% was reached at about 23:30. Startup began at about 04:30 on 5/22/78 and full power i was reached at 06:00. Figures 3.4 and 3.5 show plots of the concentrations of 131 1, 137Cs, and 58Co during the shutdown and startup, respectively. Data obtained during this hot shutdown can be found in Appendix Table B.11. As indicated in Figure 3.4 and Appendix Table B.11, the concentrations of the longer-lived iodines reached a maximum at about 01:46, 5/20/78, almost 6 hours after power reduction began and about 2-1/2 hours after 0% power was reached. Iodine-131 peaked at a concentration 7.5 times its pre-spike level. Cesium-137 and seCo peaked later than did 1311 - and showed smaller increases over pre-spike levels (about 4 for 137Cs and 1.8 for seCo). Of the gaseous fission products, only 133Xe exhibited an increase after shutdown. Most fission products and crud-associated radionuclides exhibited spiking upon shutdown. Sodium-24, however, did not spike. Immediately after startup on 5/22/78 the longer-lived iodine isotopes and the cesiums exhibited spiking, but the magnitudes of the spikes
' were not as great as during shutdown. Iodine-131 reached a concentration ; 3 times its pre-spike level and 137Cs reached a concentration about 2.6 times its pre-spike level. The crud-associated radionuclides gave only 24 Na did not spike at all.
slight indications of spiking and 1 i One additional observation that can be made about the reactor coolant radionuclide concentrations during the hot shutdown of Unit #3 is that 1321 decayed with a half-life much longer than 2.3 hours. Table 3.9 lists the measured 1321 concentrations as a function of time after ether with decay corrected concentrations obtained using the ! shutdown 132I and 13togTe half lives. The data indicate that after recovering from the spike, the decay of the 132I concentration followed the 132Te half I f fe. This indicates that the source of the 1321 after the spike was 132Te. Moreover, because 132Te was detected in the reactor coolant at levels orders of magnitude below predicted (see Tables 3.6 and B.11), j the tellurium must be attached to the surfaces of the reactor internal ( structures. This explains why tellurium is not normally detected in reactor coolant. i 65 I
Figure 3.4 ) Shutdown Spike - Unit #3, 5/19-20/78 10-1 _ N , , , , _ l
/ -
* ***-*. g %-g -
% ,'*-e l31 I e -
l l I e 10-2 -
~
E s
- l i
~
O _ _ ! E - 3 ,,,,,,,,,_ m""N _ , 58Co - I 8 -
/ a _
*_ a-A' w"'a ,137 Cs -
8 C
/
o 0 10-3 -- A.N a_ l
' ' ' l 10-4 % '
l 100 _ N i i i i _ I g _ _ O , a _ _ l 10 1000 N l' 0000 0600 1200 1800 0000 , 5/16a8 5/20n8 Time INEL-A-10 989 66 l I
Finure 3.5 Startup Spike - linit #3, 5/21-22/78 10-1 _ i i i -
- *-es -
e _
,e.,
1 131 1 E *** N
- e I O - 6
%e 3 a C
.9 10-2 e 58Co C
/"\_
~
5 ,e " y - :-p
~_
e -
/s'%Ra E
o - e O
- &# ~
137Cs 4 I l I I 10-3 100 - 1 s i 1 -
- f _
~
?_
e - 3 e - - 10 0000 0600 1200 1800
- 5/22/78 Time INEL-A-10 988 l
l l l l 67 l i
i 1 TABLE 3.9
- 132I CONCENTRATIONS IN REACTOR COOLANT FOR SHUTDOWN SPIKE UNIT #3 5/19-20/78 i
Concentration Corrected
- Concentration Corrected
- Time Date Concentration Using 132I Half life Using 132Te Half life 2141 5/19/78 1.0310.03(-1 1.03 0.03 -1 1.03 0.03 -
2251 5/19/78 1.11 2 0.02 - 1.58 1 0.03 - 1.12 0.02 - 2352 5/19/78 1.04 1 0.03 - 2.0110.06- 1.06 0.03 - 0051 5/20/78 1.30 0.02 - 3.38 1 0.05 - 1.34 1 0.02 - 0146 5/20/78 1.41 1 0.05 - 4.82 0.17 - 1.46 0.05 - 0248 5/20/78 1.29
- 0.04 - 6.04 1 0.19 - 1.35 0.04 -
0340 5/20/78 1.19 0.02 - 7.21 0.12 1.25 0.02 - 044' 1 5/20/78 1.13 1 0.02 - 9.2610.16 1.20 0.02 - 0501 5/20/78 1.09 1 0.02 - 1.2110.02 1.17 i 0.02 - 064) 5/20/78 1.05 1 0.01 - 1.57 1 0.02 1.1410.01-074) 5/20/78 1.00 1 0.01 2.02 1 0.02 1.09 i 0.01 - 0840 5/20/78 9.610.1(-2) 2.63 0.03 1.06 0.01 - 0940 5/20/78 9.6 0.l(-2) 3.55 1 0.04 1.07 0.01 - 1045 5/20/78 1. I l 0.02(1) 5.19 1 0.10(0) 1.13 0.02 - 1140 5/20/78 9.1 0.1(-2 6.15 i 0.07(0) 1.03 0.01 - 1247 5/20/78 9.2 0.2 - 8.79 0.19 1.05 0.02 - t 1345 5/20/78 9.1 0.2 - 1.15 1 0.03 1.05 1 0.02 - 1440 5/20/78 8.8 0.2 - 1.47 0.03 1.02 0.02(-1 < 2005 5/21/78 7.7 0.1 - 9.67 0.13 1.16 0.02(-1
- Decay corrected to 2141, 5/19/78 i
l 4 l i 68
t I 4 f 3.3 Discussion of Measurement Data - Letdown Demines alizers 3.3.1 Unitf_33 Table 3.10 lists dates and times of sampling of the Unit #3 CVCS mixed-bed demineralizer B inlet and outlet for the determination of decontaminationfactors(DF's). Also listed are the power level, letdown flow rate, number of bed volumes that passed through the demin-eralizer, and reactor coolant chemistry infomation for each sample date. Figure 3.6 presents these parameters graphically. The increase in reactor coolant boron concentration on 5/21/78 was the result of boration during zero-power operation. j The sampling period for the Unit #3 letdown started on 2/21/78 and ended 5/25/78. During this 94-day period, twelve sample sets were j collected. Following refueling, Unit #3 was brought up to 100% power on
! 2/20/78. The Unit #3 reactor experienced two trips and also underwert five power reductions from April through May, 1978. Power reduction and escalation account for the higher than average inlet concentrations measured on 4/27-29/78 and on 5/21/78.
Inlet and outlet concentrations and decontamination factors of all ! measured radionuclides for the Unit #3 CVCS mixed-bed demineralizer B
- are presented in Appendix B, Table B.12. Table 3.11 lists DF's for all i'
measurements on Unit #3 CVCS demineralizer and Figures 3.7 to 3.10 are graphs of inlet concentrations and decontamination factors of selected nuclides. On 2/21/78, 4 days after startup following refueling, decon-tamination factors for 58C0, 60Co, 95Zr,134Cs, and 137Cs were considerably 2 higher than their respective average values while the DF's of 124Sb and i 125Sb were considerably lower than their average values. On 2/21/78 and j 2/23/78 the inlet concentrations measured for 24Sb and 12sSb were considerably higher than the concentrations measured in the reactor coolant. The reason for this is not known. Inlet concentrations of the above nuclides decreased dramatically from 2/21/78 to 2/23/78 while , outlet concentrations increased slightly, resulting in DF's on 2/23/78 which were lower roughly in proportion to the percentage decreases in
- inlet concentrations. The increases in outlet concentrations from
! 2/21/78 to 2/23/78 imply a decrease in the number of free ion-exchange sites and/or increases in the contributions to outlet concentrations from ion-exchange processes in which the radionuclides of interest were replaced by other (non-radioactive) ions.
The reduction to zero power on 5/20-21/78 caused moderate spikes in , reactor coolant radionuclide concentrations when measured on 5/21/78. ! These spikes in inlet concentrations were accompanied by increases in i DF's for most radionuclides. Reactor coolant radionuclide concentrations
- decreased on 5/25/78 to near their 5/19/78 values and concurrently the l DF's of 95 Zr, 1311, 124 Sb, 134Cs, and 137Cs decreased.
( The fluctuations of DF from one sample date to the next are such that it is difficult to assess the logical diminishing of bed ion-removal I 69
. _ . ~ .. .- - _ - _ _ __-
TABLE 3.10 UNIT #3 LETDOWN SAMPLE INF0PFATION Power Letdown Days Demin Bed Volumes pH Sample Sample Level Flow Rate in Thru Demin Baron @ Cond. Date Time (%) (opm) Service _ (x 104) M 25'C (umhos) 2/21/78 13:52 99 45 25 0.75 867 6.13 5.7 2/23/78 12:21 100 45 27 0.80 848 6.04 6.7 4/12/78 18:21 100 55 75 2.43 712 6.88 16.5 4/13/78 11:21 100 53 76 2.45 732 6.98 14.0 4/15/78 11:17 100 53 78 2.53 703 6.98 13.7 4/27/78 13:54 100 55 90 2.99 673 6.58 11.8 4/29/78 09:45 100 55 92 3.05 681 6.61 11.2 g 5/9/78 10:15 100 53 102 3.40 678 6.73 12.5 5/16/78 09:55 100 54 109 3.66 605 6.85 10.1 5/19/78 21:20 97% 21:00 52 112 3.78 599 6.68 9.85 l 27% 22:00 5/21/78 20:00 0 55 114 3.85 728 6.6 10.1 5/25/78 09:10 100 50 118 3.97 589 6.55 8.30 l '
Figure 3.6 Unit #3 Letdown Operational Information i i i i i i i i i i i
- e ~
m \.-*~.~,_O
~ ~
- i. .- *
*~e-o ,\e -
Boron (ppm) _ 589_ O-0.-O 'i 3.97x104 o#o'O"
/
o-0"O Bed volumes thru demineralizer - 1 .\ e-.N i
} Conductivity"N t
D E
=
e'e\.-.- \ - a _ j o
- gg$p H
.- p O-Os a_a-a-a - o-o_8.30 a ~
a-*- 6.55 -
,_ ! Letdown flow rate (gpm)a .._. *s ,_
~
50 ng M t I I i l i I I i t i f b$$$$$$$$$6$ n a w n n n n c c c c 1978 INEL A-14 163 t I i { 71 l
TABLE 3.11 DF's FOR UNIT #3 CVCS MIXED BED B DEMINERALIZER 13:52 12:21 18:21 11:21 Nuclide 2/21/78 2/23/78 4/12/78 4/13/78 1311 3.110.3(3) 2.8 0.7(3) 2.7 0.4 6.2 i 1.1(2) 132I 2.310.2(4) >4.3(2) 4.0 1 1.5 >8.0(2) 133I 6.3 1 0.4 2.5 0.3(3) 2.6 0.5 8.3 0.5(2) 134I 4.5 0.3 >8.1(2) >1.9(3) >1.4(3) 1351 3.2 i 0.5 5.4 1.7(3) >7.8(2) 1.4 1 0.5(3) esRb 2.7 2.6 0.3(0) 1.13 0.02(0) 3.7 i 0.2 0) 89Rb 2.1 0.3(0)1 0.3(1 1.9 0.3(1) >4.7(1) 3.1 0.4 1) 134Cs 1.34 0.04(1) 2.5510.07(0) 1.1 0.05(0i 9.7 0.7 -1) 136Cs 9.4 4.5(2) 6.221.2(0) 4.9i0.60) 137Cs 1.27 0.03(1) 3.6 2.85 i 3.4(2) 0.07(0 ) 1.04 i 0.05(0) 1.0 0.1 0) 138Cs 1.5 0.11) 8.7 0.4(0) 4.2 0.l(0) 5.3 0.30) ' 139Cs >6.3(1) >4.2(1) l l 24Na >6.8(4) >1.6(4) 1.5i0.6(3) 1.0 0.3(3) 51Cr 3.0 i 1.2 >1.6(2) >1.4(1) >1.3 ' 54Hn 3.4 0.5 3.3 0.8(2) 8.1 0.4(2) >1.7 ') 59Fe 6.7 1.2 >8.7(1) >5.7(1) >2.0 ) 4.5 1.6 *
- 57Co 6.4 1 2.7(0) 58C0 2.4 1 0.1 2) 5.510.2(1) 1.6 0.2(2) 4.0 0.3(1) l 60Co 4.4 0.2 2) 8.6 0.4(1) 2.3 0.07(2) 6.3 0.5(1) ;
- 65Zn >1.2(2)
>3.l(1) 1
' 84Br >1.1 >1.2 91Sr >3.5(4) >5.3(3) >1.4 >1.3 ' 91mY ** ** >3.7 >6.6 ) 93Y >7.2(3) >1.7(3) >1.0 >5.2(0 ) 95Zr 6.2 1 0.7(1) 4.3 0.8(1) >1.6 >9.5(0 1 95Nb 2.1 0.8 4.3 i 0.9 8.6 3.9(0)
- I 99Mo 2.8 0.2 2.0 i 0.3 >3.5(2) >2.5(2) lo3Ru 8.2 1 2.6 4.3 1 0.8 >8.4(0) >5.2(0) lo6Ru * *
<1.3(1) >2.9(2) 1.7 0.4 3) *
- 110 mag >7.6(0) 124Sb 9.4 i 0.5 -1) 1.9 0.1 (-1 ) 3.6 0.8(0) 2.4 1.0(0) 12sSb 9.4 0.2 -1) 1.56 0.06(-1) >1.7(0) I 123Te >3.2(0) l 129mre * * * * .
131Te 37.9(0) l 132Te 5.821.0(1) 7.6 2.3(0) >1.6(1) >1.1(1) 139Ba 2.4 0.l(0) ? ? i 0.1(0) 1.4 0.1(0) 2.1 0.1(0) 140Ba >6.9 . A .1 >6.1(2) >3.1(2) 140La >4.8 >1.6 >1.7(1) >2.2(1) 141Ce >4.0 >1.8 * *
<1.1
- 1 144Ce >7.1(1) >l.7(1) 187W >1.1 >4.8(2) >3.8(1) >S.8(1) 3.8 1.2 *
- j 239Np 1.7(2) 0.5(2) i
- Radionuclide not detected
/ ** Radionuclide not measured 72 /
TABLE 3.11 (cont'd) DF's FOR UNIT #3 CVCS MIXED BED B DEMINERALIZER 11:17 13:54 09:45 10:15 Nuclide 4/15/78 4/27/78 4/29/78 __ 5/9/78 , 1311 3.8 i 1.9(3) 3.5 0.7 >8.9(3) 4.5 2.3(3) 1321 >1.3(3) 2.7 0.9 >l.l(3)
- 133I 5.7 i 1.5(3) 5.8 1.7 4.1 1.4(3) >3.8 134I >l .4 >4.7 >l.6 >1.3 1351 >7.5 >1.1 >4.6 >1.4 serb 2.310.10) 1.57 0.05(0) 1.64 0.05(0) 3.0 i 0.1(-1) 89Rb 2.8 1 0.3 1) 6.9 0.7(1) 3.4 0.4(1) >3.2(1) 134Cs 9.7 i 0.1 -1) 1.05 0.02(0 9.1 0.2(-1) 9.4 0.2(-1) 13sCs 5.3 i 0.2 0) 2.4 0.7(0) ) <3.4(0) 3.5 0.5(0) 137Cs 1.06 0.02(0) 1.1610.02(0) 9.1 1 0.2 ) 1.02 1 0.02(0) 13eCs 9.4 1 2.3(0) 1.02 1 0.02(1) 5.9 0.1 4.23 0.07(0) 139Cs >2.7(1) >3.0(1) 1.6 0.8 >5.9(0) 24Na >1.9(3) >1.4(3) >2.6(3) >3.1(3) 51Cr <2.8(-1) 54Mn 1.5 0.5(2) 7.3 2 1.3(1) >l.3 >l.6 59Fe >1.8(1) >l.l(1) >5.2 >1.7. j 50C0 4.9 0.6(1) 6.8 0.3(0) 1.0 0.2(2) 1.6 0.3(2) 60Co 1.1*1 0.3(2) 1.23 1 0.05(1) 7.5 0.6(1) 9.9* i 1.0(2) 65Zn *
>l.8(l)
- 84Br >1.9(1) >2.8(l) >l.2(1) 91Sr >5.6(1) >9.l(1) >2.5(1) >4.l(1) 9ImY >l.5(1) >3.6(1) >2.2(1) 93Y >1.0(1) >1.0(2) >3.3 95Zr >1.5(1) 4.9 0.6 9.3 4.9(0) >3.8 95Nb 8.6 2.7(0) 3.0 1 0.6 6.8 2.3(0) >2.0 99Mo >2.5(2) 5.9 0.8 >3.4(2) >6.1 10lRu >2.2(0) 3.2 i 1.2 >8.7(0)
- lo6Ru *
>5.l(1) 3.412.9(0)
IMAg >1.2(1) 124Sb 3.4 i s.4(0) 4.2 0.9(0) 3.8 2 1.1(0) 4.7 2.0(0) 12sSb >1.7(0) >1.6(I ** 129Te
>l.3(1) >1.8(0 129mTe
>l.1(2)
>3.4(2
>8.5(-6) 131Te 132Te >4.1(0) >l.2(1) >3.9(0) >4.5(0) 139Ba 3.2 0.1(0) 3.7 0.5(0) 1.7 0.1(0) 2.00 0.05(0) 140Ba >2.2(2) 9.8 1 0.3(0) 3.5 0.2(1) 1.9 1 0.9(2) 140La >1.4(2)
<7.6(l)
<3.4(2 141Ce >1.4(0 144Ce
>9.l(0) >2.1(1 187W >2.l(1)
>3.8(0) 2 39Np
- Radionuclide not detected 0* Radionuclide not measured 73
TABLE 3.11 (cont'd) DF's FOR UNIT #3 CVCS MIXED BED B DEMINERAllZER l 09:55 21:20 20:00 09:10 I Nuclide 5/16/78 5/19/78 5/21/78 5/25/78 I 131I 1.1 1 0.1 2.3 0.1(3) 5.4 1 0.3(3) 4.8 1.8(3) 132I 2.6 i 1.0 >8.1(4) >4.8(4) >7.7 133I 2.9 1 0.4 2.5 1 0.8(4) >6.0(2) >3.2 134I >1.4 3) >1.4 5 * >1.9 lasI >2.3 3) >7.8 4 >3.8(2) >5.6 2) esRb 3.0 1 0.6(3) ** 3.52 0.09(0) 89Rb >1.2(0) >7.4(3) 2.9 0.3(1) 134Cs 1.04 1 0.01f0) 9.7i0.2(-1) 3.65 1 0.06(0) 1.20 0.02 l 136Cs 6.6 0.3(0) 8.5 0.4(0) 6.0 0.1(1) 1.21 0.07 137Cs 1.07 0.01f0) 1.05 C .01 (0 3.67 0.04(0) 1.34 0.09 13BCs 4.5210.07(0) 2.4 0.3(4)) ** 5.81 0.2(0) 139Cs >1.4(3) >1.9(1) . 244a 1.7 0.6(3) >2.2(4) >7.8(2) >6.8(2) , 51Cr >1.1(1) >1-1(1) >7.8(1) 9.0 3.4(1) i j 54Mn 1.8 1 0.4(2) 3.2*0.3(1) 4.4 0.8(2) 2.30 i 0.08(1) ! 59Fe >3.9(1) >2.3(1) >1.5(2) >2.l(1) 57C0 <l.4(0) >3.3(0) >2.0(1) <8(-1) seCo 1.7 i 0.l(2) 4.29 i 0.05(1) 2.00 0.02(2) 4.3 1.7(2) ) 60Co 1.6 0.4(2) 1.1 0.4(2) 1.5 0.1(2) >1.6(1) 65Zn * *
>5.2(1) >8.0(0) 94Br 91Sr >1.2(2) >2.1(3) >1.2(2 >1.5 91mY >2.6(1) >4.7(2) >2.8(3 >2.1 i 9 3Y >1.8(3 >8.0 95Zr *
>2.9(1) 8.113.5(1) 5 2(0) 95Nb >3.1 1 8.1 1.8(0) 3.2 0.3(1) 4.1 0.5(0) 99Mo >4.7 2 >5.1 >3.1 3) >3.6
' 103 >7.9 0 >6.0 >9.1 1) >1.1 110 mag * *
>5.5(0) >3.4(1) 124Sb 6.3 1.6(0) 6.2 1.1(0) 1.24 0.07(1) 4.0 0.4(0) 125Sb * *
<3.0(1) >2.6(0) 129Te >3.0(0) 129mTe <6.8(0) <3.6 131Te * *
>6.4
>1.6(0) 132Te >5.2(0) >1.2(1) >4.5(1) >1.2 3.3 **
139Ba 0.2(0) >2.3(2) 4.6 t 0.3(0) 140Ba 2.5 0.5(2) 6.1 0.6(1) 3.7i0.3(2) 2.9 0.6(2) 140La >1.3(3) 3.5*1 0.1(2) 141Ce
<1.2(0) 144Ce
- 187W * * *
- 239Np >l.5(1)
- Radionuclide not detected
** Radionuclide not measured 74
Figure 3.7 131I and 133I Inlet Concentrations and DF's for Unit #3 CVCS Mixed-Bed B Demineralizer _ i i i i i , i i i i i i _
= 4.912.3(3) 131; D.F. 2 i
gN o - 4 %, k%..o_ ._
~~ -
l E
.L o1.8810.03(2) -
o-o 1311 pCi/ml
~
O o ig 0-ONo
/ 0 6.02 0.02(-3) o/
o y --
~
g 2.510.8(4)
- ~
, D.F. = Decontamination factor _
i
.=
E 5 -
~
E 5
., < o .7 1.5(3).
0 133l D.F. -
%e\
.. e _
- n --
- E g
- I ~ e a :
o-O"U'O' No O' - 4.9710.03(-2) 1331 pCi/mf - l es=
=
6 i Y o - t I I t i I I I I f I i 1 O
~ ~ , , , , , e e e e l 1978 IN EL- A-14 160 l
75 i
Figure 3.8 134Cs and 137Cs Inlet Concentrations and DF's for Unit #3 CVCS Mixed-Bed B Demineralizer i i i i i i i I I i i I I I e 1.34
- 0.04(1) 01.3820.01(-3)
; 134Cs pCi/ml
~
_ -O [ 1 _ No o - N %o . 2.8910.02(-4) _. O e 1.05
- 0.02(0) e
'No-o *N,~e#**e 134Cs D.F. 3 o - o -
.E
( - e 3.6710.04(0)~ e 2.8510.07(0) E 4 - 0 137Cs -
; , D.F.
e -sN e e~e __
) [ ONO 2 O*~O O
~
0-- / 137CspCilmf~ 0 i 4.1610.02(-4) , O f -. D.F. = Decontamination factor --
~
_ O [ I I i i i I i i I I I I I 5 b N O N N b g 3 3 5 N
$ 6 4 4 4 4 4 . h h h 1978 INEL-A 14 165 l
76
Figure 3.9 seCo and 60Co Inlet Concentrations and DF's for Unit #3 CVCS Hixed-Bed B Demineralizer T _ I I I I I I I I I I i _ r - _ Unit 3 CVCS Mixed Bed B 9.911.0(2) _
- ~
60Co Decontamination factors j _ _
~ ~~
;7.510.6(1) . I
~ ~"
4.7 1 0.1 (-4) l _ 2.0610.04(-4) _ l 60Co Inlet Concentrations 2 _ 4.311.7(2) ) _ 58Co Decontamination _ factors ,, k 1.610.2 (2) ) -
\ ..
_- 5
) 8.08 2 0.07T l
_ (-4) _ l _ _ l 58 Co inlet concentrations 1.8610.02(-4) l l I I I I I I I I I l 2-21 2-23 4-12 4-13 4-15 4-27 4-29 5-9 5-16 5-19 5-21 5-25 19/8 i 6810.3(0) INEL A-10 762 i 77 i
Figure 3.10 MZr ano 12',Sb Inlet Concentrations and DF's for Unit #3 CVCS Mixed-Bed B Demineralizer i _ i i l i i i I iil i 3 _
" 8.1 3.5(1) : i
_ D.F. = Decontamination factor _ 0 _ l _ l i l l .. j a _ . o 95Z r D.F. _ _ i . _ 4
- o -
l .. . _ o -5) ,, _ _ O _
$C 95Zr pCilmf 6 1.24 0.07(1) m - _
M o d - ~ I II
~
{..g/ y/ N o
'2;p : ~
~
\.o[ .. __
. 2.4 0.1(-5)
~
124Sb pCi/mf
- Q T l -
l
. .~./ _
I..s./
- e j 1.910.1(-1) 1.920.7(-6)
- I i 1 I l t i i i i t ! I h'
~ ~, , , , , e 7
e ho e l 1978 IN EL.A-14 159 i l i 78 i I
__ _ .- _=. efficiency with bed usage. 'nhen the average inlet concentration of i 131I is divided by its average outlet concentration for each of the February, April, and May, the average DF's are respectively months},2.22(3{b 2.89(3 and 2.97(3). The average DF's for each of the above months are for Co: 2.45(2), 3.82(1), and 1.02(2); and for 124Sb: 5.06(-1), 3.60(0), and 7.62(0). The measurements carried cut on the demineralizers at Turkey Point indicate that DF's for many radionuclides tend to be correlated with inlet concentration. There is, therefore, no single DF value that exactly represents the DF for all values of inlet concentration for i the radionuclide. It is of interest to have a single DF value that reflects the average performance of a demineralizer over the range of observed inlet concentrations for a radionuclide. We obtained these representative or "best value" DF's by first determining average inlet and outlet concentrations and then using the average concentrations to obtain DF's. Table 3.12 contains these "best value" DF's together with mean inlet and outlet concentrations and ranges for che inlet and outlet concentrations for demineralizer B. The "best value" DF's for 134Cs and 137Cs are equal to 1.0 which is the value of the average DF's of 134Cs
- and 137 Cs during the period 4/12 to 5/19/78. In detenining the "best value" DF's for cesium, the data taken during February while the resin in demineralizer B was very fresh were not included. While the resin was fresh, cesium DF's were much higher than nonna1, rangirg up to approximately 13.
3.3.2 Unit #4 Table 3.13 lists dates and times of sampling of the Unit #4 CVCS mixed-bed demineralizer A inlet and outlet for the determination of decontamination factors (DF's). Also listed are the power level, letdown flow rate, number of bed volumes that passed through the demin-eralizer, and reactor coolant chemistry infonnation for each sample date. Figure 3.11 presents these parameters graphically. The higher than normal reactor coolant conductivity observed on 4/12-15/78 was due to bypass of the CVCS demineralizer. The elevated conductivity on 12/6-8/78 has no apparent explanation. The sampling period for the Unit #4 letdown started on 11/30/77 and ended 5/23/78. During this 175-day period, twenty sample sets were collected. The Unit #4 reactor experienced one trip (12/9/78 at 00:54) and also underwent three load reductions from 11/30/77 to 5/23/78. The percentage power level minimums on tte following dates are: 12/1/77, 50%; 12/9/77, 44%; 2/14/78 to 3/9/78, 0%; and 4/24/78, 32%. The February-March power outage was for the purpo!,e of steam generator tube inspection and repairs. Also of note is the fact that the Unit #4 CVCS mixed-bed demineralizer was bypassed on sample dates 4/12-13/78. The above-mentioned ! operational occurrences account for the significantly higher than average inlet concentrations measured on 12/9/77 and on 4/12-13/78, and 12/14/78. ' l l 79 l
. . _ . - . ~ __- _ _ - - ___ ..
i TABLE 3.12 __ MEANS AND RANGES FDR ADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's i FOR UNIT #3 CVCS MIXED BED B DEMINEPALIZER I Inlet Concentration (pCi/ml) Outlet Concentration (pC1/ml) "Best Value" , Nuclide Mea Range Mean Range DF 131 8. 5 0.' 37 8 - 3 0.11-1.2(- 2.6 132I 8.8 - 0.345-1.05 - 5.5 - 0.015-2.6(- 1.6 3 ' las! 4.7 - 0.72-6.0(-2) 2.4 - 0.023-1.0(- 1.9 3) , 1341 1.6 - 1.27-1.92(-1) 4.8 - <0.012-<1.6 4) 6(3
- 135I 8.9 - 0.0034-1.0 (-1) 4.1 - 4.009-<2.3 -4) 2.2
- serb 2.1 -
0.59-4.8(- 1.3 - 0.00037-4.14(-1) 1.6 , 89Rb 4.9 - 2.41-6.3(- 3.3 - <0.00057- <4. 7 1.5 ' 134Cs 4.7 - 0.050-1.38 3) 3.7 - 0.143-8.5(-4)(-2)1.3 13sCs 1.0 - 0.35-5.65( ) 1.1 - 0.0096-2.7(-5) 9.4 137Cs 6.6 - 0.074-1.90 3) 4.9 - 0.0188-1.13 - ) 1.3 l 13eCs 1.9 - 0.90-2.44( ) 6.6 - 0.00007-4.3 - ) 2.9 ' 139Cs 1.0 - 0.54-1.6(- 5.9 - <0.046-9.9( 3 1.8
% *:*: Mid'2P l N: '8:8B:B:!l 54Mn 59Fe 8.4 -
1.5 - 0.58-1.74(-4) 0.77-4.0(-5) 7.8 - 3.5 -
<0.052-3.2 -6
<1.9-<8.7( 7) 2:1 1.1 A.3 ,
l 1ll s7Co 1.3 - <0.6-4.5(-6) 3.2 - 0.99-9.9(- ) 3.9 seCo 3.4 - 0.112-1.45(-3) 5.0 - 0.05-2.65(-5) 6.9 )> 60C0 2.2 - 0.88-8.7(-4) 2. 3 - 0.10-9.0(-6) 9.5 ') ssZn 4.1 - <0.13-1.3( ) <4.0 - >1. 0 i 91Sr 3.2 - 0.36-6.0 - <4.7 - >6.9
; 91mY 2.2 - 0.82-5.6 - <5.4 >4.1 i 93Y 1.5 - 0.11-6.2 - <2.3 >6 . 4 95Zr 1.4 - <0.35-3.1 (- 5) 7.5 - <0.30-3.5 - 1.8( , i 95Nb 2.3 - 0.0108-1.4( ) 9.8 - <0.31-2.6 - 2 . 31 99Mo 1.1 - 0.42-4.59(- 2.2 - 0.015-1.5 - 5.2f '
10 3Ru 9.2 - 0.24-3.12(- 4.7 - <0.23-1.'/(-6 1.9I
! 106Ru l.1 - 0.29-4.1(-5) 1.3 - <0. 57-<6. 8(-6) 110 mag 3.7 -5 <0.0095-3.3I;-4) 2.7 - <0.014-<5.4(-5) 8.4I(
1.3 124Sb* 6.2 - 0. 30-5.0(-5 1.1(- ) 0.61-1.9(-6) 5.6 1 12sSb* 2.8 - 0.18-<6.3(-) 1.0 - <0.17-2.1(-6) 2.8
! 13984 1.6 - P42-3.3(-2 5.2 - <0.049-8.2(-3) 3.0 14cBa 1.2 - 0.59-3.11(- 3 1.5 - <0.0085-9.2(-5) 8.2 14cLa 9.9 - 0.028-6.38(- ) 5.9 - <0.0084-1.37(-5) 1.7 l 141Ce 1.9 - <0.32-3.2( 6 5.1 - <0.08-<2.4(-6) 3 . 81 ,
1 14*Ce 3.8 - <0.003-1.7 ) 1.2 - <0.10-<5.9 (-6) 3.01, 197W 4.9 - <0.17-1.1( <4.5(-5 >1.1(1) 139 Np 3.4 - <0.1.1 -1. 4 (- 5.8(-6)) 0.026-<6.0(-5) 5.9(0)
- Data from 2/21/78 and 2/23/78 not included because ? questionable inlet 4
concentrations. Inlet concentrations were much higher than concentrations in reactor coolant. i 80
TABLE 3.13 UNIT #4 LETDOWN SAMPLE INFORMATION Power Letdown Days Domin Bed Volines pH Sample M Level Flow Rate in Thru Demin Baron 9 Cond. Date , , (%) (t) Service (x 105) ggni 25'C (umhos) 11/30/77 1 100 60 180 0.63 651 6.80 18.0 12/6/77 14. ., 100 60 186 0.65 630 6.82 16.2 12/8/77 12:42 100 60 188 0.66 637 6.73 16.0 12/9/77 09:38 100 62 189 0.67 638 6.50 11.5 12/10/77 09:18 100 65 190 0.67 644 6.68 11.1 12/11/77 10:54 100 60 191 0.67 621 6.68 11.05 12/12/77 09:31 100 53 192 0.68 616 6.68 11.5 c3 12/13/77 14:14 100 62 193 0.68 620 6.70 11.8 2/10/78 11:41 100 65 252 0.92 448 6.88 11.2 2/10/78 16:03 100 65 252 0.92 44 8 6.88 11.2 2/14/78 13:25 95 62 256 0.94 428 6.73 11.3 4/12/78 17:55 100 65 299 1.10 337 7.22 16.1 4/13/78 11:33 100 65 300 1.11 336 7.35 17.5 4/14/78 13:53 100 65 301 1.11 320 7.28 17.2 4/15/78 09:57 100 60 302 1.11 323 7.28 17.6 4/27/78 14:04 100 45 31 3 1.15 288 7.05 13.2 4/29/78 09:53 100 45 315 1.16 288 7.01 12.9 5/9/78 09:27 100 55 325 1.19 250 7.15 12.0 5/16/78 10:05 100 55 332 1.21 233 7.30 11.0 5/23/78 10:23 100 47 339 1.23 21 3 7.23 11.8 a i
Figure 3.11 Unit #4 Letdown Operational Inforination i i i i i i i i i i i i i i i i i i i i i i e .-. .-.. . ._, ~ Boron (ppm) e-e s e e-e s,
\e-o\*s e 213 _
oo Conductivity o- Bed'o volumes
~ \ thru demineralizer s
e a~o s 1.23x105 fn o so_g o oso_o_o ,,,,_,_, a-e Ng 11.8
.--. =/
g _ _ T3 ~ 1 4 pH
- o -o-o 7.23 ~
!.,_S S"2"~U'o'o-o-o-o-
'*'a-a-aNa o~
_ a -=Q-*S,w2-04 a-
- N a _
N'/ Letdown flow rate a-.
\'47 _ l (gpm) a-a i
~ _
l l I i 1 i t I t i i i 1 i i t i t i l I 1 i h$$$SEEE$$$EN$$$$NO e - e e yyyyn n n n n n n n h c 1 INELAM 162 1977 1978 l 82
Appendix B Table B.13 presents inlet and outlet radionuclide concentrations and DF's for the Unit 4 CVCS mixed-bed demineralizer A. A summary of DF's for all measurements on Unit #4 CVCS mixed-bed A demineralizer (except while the demineralizer was being bypassed) is presented in Table 3.14 and Figures 3.12 to 3.15 show plots of inlet concentrations and decontamination factors for selected nuclides. When the data for any given radionuclide are examined as a whole there appears to be poor correlation between DF and inlet concentration. However, if we examine the data in the vicinity of 12/9/77 reactor coolant spike we see that the change in DF is roughly proportional to the change in inlet concentration for a number of nuclides, especially 134Cs, 137Cs, and 59Fe. In order to assess the change in DF with resin bed usage, average DF's were computed for selected nuclides for the following periods: 11/30-12/13/77, 2/10-14/78, 4/14-29/78, and 5/9-23/78. For the above periods the average DF's for 1311 are 3.94(3),1.20(4), 2.26(3), and 4.30(3); for 60Co they are 3.46(1),2.40(1),2.60(0),1.10(1); and for 12'Sb they are 6.41(0), 3.29(0), 1.22(0), and 2.19(0). The dependence of DF on the number of bed volumes that passed through the demineralizer is not made clear using these average DF's possibly because that dependence is obscured by the effects due to the large variations in inlet concen-trations. The demineralizer bypass of 4/12-13/78 caused an increase in reactor coolant radionuclide concentrations. The DF's of s4Mn, 59Fe, 58Co, 60Co, 124Sb, exhibit minimums on 4/14/78 but show large increases on each of the next three sample dates and by 4/29/78 those DF's had increased to near their February values. The "best value" DF's for CVCS #4 demineralizer A together with mean inlet and outlet radionuclide concentrations and ranges for the inlet and outlet concentrations are presented in Table 3.15. 3.3.3 Conclusions The Unit #3 letdown data of 5/19,21,25/78 and the Unit #4 letdown data of 12/8-13/77 indicate that reactor coolant spikes of several radionuclides are effectively stopped by the mixed-bed demineralizers and that spikes of most radionuclides are diminished by tie demineralizers. Demineralizer outlet concentrations of 134Cs and 137Cs decreased during spiking and no delayed outlet spike was measured either during the four days following the reactor coolant spike of Unit #4 or on the one sample date which followed by four days the spike of Unit #3 reactor coolant. Demineralizer outlet concentrations of 131I did increase during both spikes but in each case the percentage increase in the outlet concentrations was much smaller than the percentace increase in j the inlet concentration. Hence, percentage removal of IIII improved I dramatically with a large increase in 131I inlet concentration. The spikes in inlet concentrations on 12/9/77 and 5/21/78 were accompanied by increases in decontamination factors for most of the other measured l radionuclides. l l l l 83 l 1 i
TABLE 3.14 DF's FOR UNIT #4 CVCS MIXED-BED A DEMINERALIZER 16:09 14:55 12:42 09:38 09:18 Nuclitt 11/30/77 12/6/77 12/8/77 12/9/77 12/10/77 1311 >3.7 >9.1 ll 6.2 i 1.7(2) 4.2i0.1(3) 1.27 0.03(4) 132I >6.8 >9.8 f >2.0(2) >4.3(2) >1.3(2) 1331 >1. 4 >1.5 ll 8.9 3.2(1) 5.02 1 0.08(3) 2.2*1 0.3(4) 1 34I >4. 91 >8.2 I >2.1(2) >1.3(2) 135I >6.7( >6.2 !l >3.3(4) 6.9 i 0.5(3) >1.0(4) serb 1.8 1 0.4(0) 3.9 i 0.6(0) 4.510.4(0ll 3.6 1 0.6(0
- 89Rb >9.2(0) >2.0(1) 3.8 1 0.9(1 ) 9.8 i 2.1(0 134Cs 1.96 i 0.03 'l 9.0 i 0.2 - ) 1.05 i 0.02[0) 5.86 i 0.09 0) 1.68 0.03 136Cs 1.62 2 0.10 1 1.6 i 0.2 0 1.2 0.1(0) 7.4 0.4(1) 1.45 1 0.05 9.7 i 0.2 - ) 1.05i0.01(0) 5.07 i 0.09(0) 1.61 1 0.03 137Cs 2.31 i 0.03 / 1.25 i 0.08 1) 1.57i0.08(1) 1.6*1 0.1(1)
- 13aCs 1.00 1 0.10 J
- 139CS *
- 32,4(1) 3H 9.8 i 0.4(-1) 14C 4.1 i 0.6(-2) 2 i.Na >2.7(2) >3.0(3) 2.3 1 0.1 ll 6.4 1 0.6 7.5i1.9(2) 51Cr >1.1(2) 1.6 i 0.3 g 5.0 1 1.7 s' Mn 7.6 1 0.4(1) ,5.4(0) 7.7 1.4 j 5.8 0.9 8.6 1.67 ii 1.9(0) 0.08(1 )
55Fe 3.7 2 0.4(0) 59Fe *
- 3.0 i 1.6(1) 1.7* i 0.1(2) 1.4 i 0.2(1) 57Co >1.7(1)
>1.9(0) >5.0(0) 7.4t0.4(0) 4.0 1 0.1(1) 8.1 1 0.2(0) I seco 1.45 1 0.02(2) 6.9 i 0.7(0) 60Co 1.24 1 0.03(2) 1.0 0.2(1) 7.5**i 1.9(0) 1.3 1 0.1(1) 5.0**t 0.2(0) 6 3N1 5.3 1 0.2(0) 65Zn * *
- 2.2 1.0(2) >2.6(1)
)
89Sr 2.3 i 0.5(2) ** " ** ** 1 I 90Sr 9.0* ' 4.0(1) l 91$p *
>3.1(1) >1.3(1) >3.9(0) 91Y 1.7*1 0.8(0) 93Y >3.1 >4.1(3) >2.4(4) >6.3(1) 95Zr >2.2(1) >4.6 8.711.7 1.3 0.3(1 2.9 0.3 95Nb 2.6 i 1.0(1) <6.3 1.7 i 0.6 1.4 1 0.3(1 1.9 i 0.5 99Mo >1.81) >7.9 1.310.1 5.7* i 0.3(2 6.8 0.1 to 3Ru <1.0 1)
- 1.3 i 0.6 >3.2(0) 11emAg >4.4 -1) >6.0(-1) >1.3(0) 124Sb 6.6 i 0.5(1) <1.1(0) 3.0 i 1.1(0) 6.6 i 1.5(0) 1.3 i 0.2(0) 12sSb 4.9 2.2(0) *
<3.3(0) <1.4(1) <1.4(0) 129mTe >2.1(2) <3.0(1) 132Te * * *
- 8.3 i 3.6(0) 139Ba 1.3010.04(0) 1.39 i 0.03(0) 1.820.1 1.310.1(0) <3.4(-1) 140Ba >2.5(1) >1.1 (1 *- 1.9 i 0.7 >l.2(2) >7.2(1) 140La *
- 1.9 i 0.5 4.3 1 0.6(2) 3.310.6(1) 141Ce *
* * * <1.5(0) 143Ce *
* * * <1.5(2) 144Ce 187W >4.5(1) >1.2(2) >7.7(1) >5.5(1) >1.6(1) 239Np *
- 2.8i0.5(0) 1.0 i 0.6(1) 9.3 i 6.4(0)
- Radionuclide not detected
** Radionuclide not measured 84
l TABLE 3.14 (cont'd) DF's FOR UNIT #4 CVCS MIXED-BED A DEMINERALIZER 10:54 09:31 14:14 11:41 16:03
'f uelide 12/11/77 12/12/77 12/13/77 2/10/78 2/10/78 131I 8.9 0.1(3) 2.0 .1(4) 7.1 2 0.2(3) >4.5(3) >2.3(3 132I >1.1(2 >4.1 >5.0(2) >3.5(2) >1.9(2 133I >2.1 >7.9 3.3 i 0.5(4) 2.6 0.6(2) >2.0 134I >9.9 >l.1 >5.0 2 >2.4 2) >1.4 las! >9.1 i :l.1 >3.4 4 >1.1 3) > 1.1 serb 3.7i1.2(0) 3.3* i 1.7(-1) 6.7 i 0.5(0) 4.8 i 0.8 0) 4.1 0.8(0) 89Rb >9.8(0) 7.4 i 1.6(1) 2.0 0.4 1) >4.0(1) 134Cs 1.04 i 0.03(0) 1.01 0.02(0) 9.710.3(-1) 9.7 0.1 - ) 9.4 0.1 - )
136Cs 2.9 i 0.3(0) 2.4 i 0.3(0) 1.4 0.4(0) 2.2i0.30 2.1 1 0.3 0 137Cs 1.09 0.03(0) 1.08 t 0.02(0) 1.03 1 0.03(0) 9.87 i 0.09 -1) 9.9 0.1 - ) ' 13eCs 2.1* i 0.3(1) 9.9*1 1.5(-1) 2.5 i 0.1(1) 1.6 i 0.2(1) 1.6i0.21 139Cs >8.7(1) >3.1(1) 311 ** ** ** ** **
; 14C ** ** ** ** **
24Na 1.5 i 0.4 3 1.4 1 0.2 3 1.6 1 0.2 >2.8(3) 3.321.0(3) 51Cr 2.0 i 1.2 1 1.8 0.31 2.3 1 0.6 >6.3(1) >2.8(1) 54 6.8 1 0.7 1 3 4.9 3 0 81 6.2 0.4 8.5 j 0.6(0) 4.3 i 0.5(1) 59Fe 2.410.6(1) 7.3 i 4.0(1) 2.2 i 0.5(1) *
>1.6(0) 57C0 >2.3(0) >3.1(0) >2.3(0) >1.6(0) 58Cn 3.2 1 0.1 1 9.8 i 0.3 2.8 i 0.1 1 1.11 1 0.07(2) 1.17 i 0.03 i so; 2.3 1 0.1 1 8.5 j 0.3 1.3i0.11 6.6 1.9(1) 1.33 0.09 65Z: * *
>3.1(0) >8.3(0) >6.9(0) 895g 90Sr ** ** ** ** **
r >1.2(1) >1.5(1) >1.4(1) >2.9(1) >2.1(1) 93Y >3.l(2) >2.6(3) >3.5(3) >2.6 >1.3(1) 95Zr 5.4 0.6 1.9 1 0.3 7.6 0.7 >7.5 4.0i1.1(0) 9sNb 4.2 i 0.5 1.7 0.3 5.8 0.6 >9.1 3.5 0.6(0) 99Mo 1.3 i 0.1 1.5 0.4 1.2
- 0.2 >2.1 >1.0(1) lo3Ru * *
>1.7(0) >1.2(0) >1.9(0) 110'% 9 >1.2(0) 124Sb 3.6i0.5(0) 2.6 0.3(0) 3.410.3(0) 3.1 0.9(0) 3.4* i 1.0(0) 12sSb >1.1(0) >2.1(0) >3.4(0) 129mTe <1.0(-1) 132Te * *
>2.9(0) >4.2(0) >1.1(0) 1398a 1.1310.06(0) 5.9 i 0.5(-1) 2.4 i 0.1(0) 1.80 1 0.04(0) 1.40 0.04(-1) 140Ba >6.0 >1.3(2) >9.5(1) >4.7(1) >2.1(1) twoLa >1.6 1.010.8(3) 2.510.3(2)
>l.3(1) 141Ce >1.6 143Ce >6.2-1)
<5.0(-1) 146Ce
- 187W >3.5(1) >3.1(1) >9.1(1) >3.5(1) >2.5(1) 339Np' * *
! <1.5(0) 2.2- . 1.7(0) <8.7(1)
- Radionuclide not detected
** Radionuclide not measured 85 ;
\
i 4 TABLE 3.14 (cont'd) DF's FOR lNIT #4 CVCS MIXED-BED A DEMINERALIZER 13:25 13 53 09:57 14:04 Nuclide 2/14/78 4/14/78 4/15/78 4/27/78 131I 5.8 i 4.1(4) 1.0 i 0.2(4) 2.5 0.8(3) 3.8 t 0.1(2) 1321 >4.7 >1.9 >3. >1.1(3) I 133I >4.5 >3.3 >9.4 7.8 3.4(2) l 134I >3.6 >2.3 >2.6 >4.4(2) 135I :1.4 >4.5 >8.6 , >2.9(2) 1 1.0 2 , serb ..I 0.4(0 3.4 3.31 0.06(0) 5.3 1 0.2(0) 1 89Rb 2.2 1 0.5(1 1.6 i 0.4 1 2.7 0.7 1 >4.4(1) 134Cs 1.04 0.02 0) 8.2 0.2-) 9.2 0.1 - ) 9.8 1 0.1(- ) 136Cs 2.1 i 0.2 0 2.9 0.2 0 4.4 i 0.6(0 137Cs 9.8 1 0.1 - ) 1:7 i 0.2 9.0 0.4-11)) 9.5 i 0.1 - ) 1.0320.020) 13sCs 1.3 1 0.1 1 8i.82 1 0.39(0) 6.7 0.1 0 1.41 0.031) ) 139Cs >1.2(1) > 3.5(0) i 3H ** ** ** ** I l li.C ** ** ** ** 2i.Na 1.9 i 0.3(3) 7.1 1 3.2 3 <+ 1.7 0.2 2 4.9i1.G(-2) 4
>5. b0(0) '. 0 .5 0.05(1) 59Fe 8.112.0(1) 1.5 i 0.2(0) 4.3i1.0(0) 1.1 0.2(1) 57Co <l.8(0) >8.1(-1) seCo 2.43 0.07(1 1.10 i 0.04 2.46 i 0.02(0) 6.8 i G.1 60 2.7 + 0.4(1) ) 1.3310.04 3.1 0.1(0) 3.9 i 0.2 65Zn * *
<4.0(1) >1.9(0) )
893p 903p ** ** ** ** l 91Sr * * *
>5.0(2) 91y ** ** ** **
9W * * *
- 95Zr 1.5 1 0.3 1.11 i 0.08(0) 2.5 0.4(0) 2.6 1.0(0) 95Nb 5.6 1.2 8.7 0.6(-1) 2.1 1 0.2(0) 4.8 i 1.0(0) 99Mo 1.5 1.2 >2.0(1) >2.1 1 >1.0(2)
* >1.4 103Ru >3.0(0) ) >2.3(0) 110mA9 >1.6(0) 1.9 i 0.7(0) >1.4 ?
)
124Sb 3.3 i 1.0(0) 9.5*1 0.7(-1) 1.1 0.1(0) 1.8 0.2(0) ! 1255b * *
<1.4(1) *
- 129mie *
>8.1(0) 132Te <4.3(1) >1.7(0) 1398a 1.00 i 0.05(0) 1.1 0.1(0) 1.1 1 0.1(0) 4.3 2.7(0) 140Ba >3.4(2) >3.5(0) >3.0(0) 5.0 1.6(0) thola
- 6.9 i 2.4(1) >1.3(2)
>1.l(1) 141Ce < 3.8(1 ) <1.5(-1) 143Ce * * *
- 144Ce <4.9(1)
- 187W >8.7(1) >1.7(1) >8.5(0)
- 239Np * *
>3.0(0)
- Radionuclide not detected
** Radionuclide not measured 86
TABLE 3.14 (cont'd) DF's FOR UNIT #4 CVCS MIXED-BED A DEMINEPALIZER 09:53 09:27 10:05 10:23 Nuclide 4/29/78 5/9/78 5/16/78 5/23/78 131I 6.2 i 1.8(3) >5.1(3) >3.4 >1.0 132I >2.7 * >7.5 >4.1 133I >1,1 7.9 i 2.9(2) >5.4 >1.6 1 34I >1. 9 >2.5(2) >4.3 >3.2(2 lasI >1.0 >5.9(2) >7.2 >1.1(2 88Rb 1.18 i 0.03(0) 9.8 i 0.2(-1)
- 3.12 i 0.09(0 89Rb >l.7(l) >2.2(l)
- 5.2i1.71))
134Cs 9.3 1 0.1 - ) 9.4 0.2(-1 1.04 1 0.02(0) 9.1 i 0.2 - ) 136Cs 4.8 i 0.6 0 4.4 1 0.4(0)) 5.4 0.7(0) 5.4 2 1.0 0 137Cs 9.8 0.2 - ) 1.00 i 0.02(0) 1.09 0.02(0) 9.8 i 0.1 - ) 1380s 5.2 1 0.2 0 6.4*1 0.2(0) >6.1(0) 7.3 0.1 0 139Cs * *
- 311 ** ** ** **
14C ** ** ** ** 24Na 2.7 i 0.5(2) 3.1i0.2(2) 3.5* i 0.2(2) 3.2 i 1.0(2) SICr >1.4(2) >2.8(1) >4.5(0) 54Mn sspe 4.7 1 0.6(1) 2.4 1 0.3(1) 5.1 1 0.9(1) 9.2 1.1(0) 59Fe 2.2* i 'J.5(1) 9.2*1 2.9(0) >1.2(1) >4.8(0) 57Co
- 38,1(.])
58Co 1.91 0.04(1) 3.9 0.1(0) 1.82 0.04(1) 5.0 1 0.2(0) 60C0 1.3.**t 0.03(1) 7.6**i 0.3(0) 1.4**i 0.2(1) >7.4(0) , 6 3Ni ** ' ssZn * * *
- 8937 ** ** ** **
90Sr ** ** ** ** 91Sr *
<2.5(0) *
- 91y ** ** ** **
93Y * * *
- 95Zr 5.7 i 1.2(0) 5.5 i 1.2(-1) >7.6 6.7 1.0 0.1(-1) 95Nb 0.4(1) 6.5 1 0.6(-1) >1.9 9.6 i 0.2(-2) 99Mo >2.9(2)
>7.9(1) >6.4 >1.7(1) 103Ru *
- 110 mag " *
- 2.2*2 0.8(-1) 124Sb 3.0*1 0.3(0) 1.5* i 0.2(0) 3.710.6(0) 2.5* t 0.3(0) 12sSb
- 129mTe >l.l(l)
- 1 132Te >1.3(0)
- 1 139Ba 1.4i0.1(0) 2.0 0.1(0) 8.320.7(-1) 2.0 0.1(0)
- 140Ba 3.5 i 0.6(0) >1.4(1) >1.7(1) l 140La >7.9 1) >3.5(2) >6.1(2;
- l 141Ce <1.6 -1 * *
<6.4(-2) l 143Ce >1.21))
144Ce
<7.5-2) 187W .l.1(2)
>5.5(1) >1.1(2) >7.00) 239Np 1.7 i 1.4(-1) *
>9.4-1)
Radionuclide not detected
** Radionuclide not measured
~
l 87 i
Figure 3.12 1311 Inlet Concentrations and DF's for Unit #4 CVCS Mixed-Bed A Demineralizer i i i i i i i i ~~ i i i i i i i i i _ _1 e
- R 5.814.1(4) _
- R: Outlet concentration -
determined by resin
- R R ,,
R ,g ; _ R _ o R 4.2 i 0.1 (3) _ {
- <i -
I _ l
~
I 3.8 0.1 (2) l f 2.0010.04(-2)
! i I 1 l
_ j - 1 .
~
6.4910.04(-3) _ i .. 1311 Decontamination factors
}
~
- 131 e 1 Inlet concentrations
~
Unit 4 CVCS Mixed Bed A I I I I I I I I I I I I I I i I I I I I 8999SE#D SS3NC3$S$*?$$
&&4444 4 44 i s 3SSS3$g3 1978 INEL-A-10 771 1977 88
l l ' Figure 3.13 134Cs and 137Cs Inlet Concentrations and DF's for Unit #4 CVCS Mixed-Bed A Demineralizer i
~4 I I I I I I I I I I I I I I I I I I I i 1~
_ D.F. = Decontamination factor _ 5.86 0.09(0) 4.82 t 0.05(-3) _ O e~e-e#*\ 134Cs pCi/mf 7
\
O'O-O-O-o
/ ~ _e_.3
_o' o --
% . 8-o \ g,O"O-
[/ O/ ! 134Cs 8.220.2(-1) D.F. - O 5.07 0.09(0) - g _ 6 _ g _ b O g _ O 137Cs
$ D.F.
7 OA ~O 'O ~ O ~O - gao O'O ~V O T l 7.7110.09(-3) : l _
-. 137C s pCi/mf -
2 ._, . .-. =~. . _
, \ - . ...-._.
I I I I I l t I i I I I I I i ! ! I I i i I h82$$5EE$$$E95$,$h$$h
- - -m m m m a am e e e e e o o 1978 INEL.A 14 161 1977 89
Figure 3.1A 60 C0 Inlet Concentrations and DF's for Unit #4 CVCS Mixed-Bed A Demineralizer MC I i l _ I I i i l i I --I I I I I I I I I 6.611.9(1) : _f 60Co Decontamination factors _ g 2.4510.04 (-3) 1.3310.04(0) 3.910.2(-4) ; 60Co Intet concentrations - 1 l _ j l l l 2.810.4(-5)
- Unit 4 CVCS Mixed Bed A 1 I I i 1 I I I I I I I I I I I I I I I t
o 99 9o e N m o o v N m 4 m yg?e g aSSS&AAA AAA4 4 44 4 4*44 1977 1978 INEL-A-10 760 ) l i ! 90 i
*w ,
Fiqure 3.15 54Mn and 59Fe Inlet Concentrations and DF's for Unit #4 CVCS Mixed-Bed A Demineralizer 7 i i i i i iiiiii i i i I i i i i I-2 *-% o 54Mn! ~ Y* e f 4.710.6(1)* / e D.F. [ _ 3.3(-4) _ o e
- e T O *
- 2 2 s
- l _
0 O~o-o'o O o -0 - 54Mn pCilmf - I f y - g 2.1(-4) e - E 1.96 0.07(0)
$ ~
e 8.1 2.0(1) 2 _- 4 - 59pe -
~ ' D.F. -
- a h, l } '.,$ e-o , ,PJ
-- ..J u . ,v ! .e ss _
o ~ 59Fe pCilmf e _ _ D.F. = Decontamination factor _ e 1.5 0.2(0) 1 1 1 i l i t ii1 1 I i i l I i l 1 I I I
$$$$$hEES$EES$$$$$$O e ee nnunununnnnnnn oc ma. A.14 164 1977 1978 l
91
TABLE 3.15 MEANS AND RANGES FOR RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's FOR UNIT #4 CVCS MIXED-BED A DEMINEPALIZER ' Inlet Concentration (pC1/ml) Outlet Concentration (pCi/ml) "Best Value" Nuclide Mean Rance Mean Range DF 131I 1.9(-2 0.59-9.8( 2 5.4(-6) 0.014-3.3(-5) 3.6(3) 1321 1.5(-2 0.348-8,5 - <6.1(-5) >2.5(2) 133I 2.0 - 0.70-3.23 - 1.9(-5) <0.001-1.1(-4) 1.1(3) 134I 1.5 - 0.36-2.41 - <1.2(-4) >1.2(2) 1351 1.6 - 0.48-5.2( 2) 2.5(-5) <0.0029-<1.9(-4) 6.5(2) 88Rb 1.0 - 0.44-3.34 - ) 5.5 - 0.0039-3.4(-1) 1.8 89Rb 8.3 - 0.42-1.5(2 9.6 - 0.0098-<2.2(-2) 8.7 134Cs 1. 3 - 0.70-4.82 - ) 1.0 - 0.577-1.98(-3) 1.2 136Cs 2.0 - 0.016-2.65(-3) 1.8 - 0.37-4.7(-5) 1.1 8 137Cs 2.4 - 1.40-7.71(-3) 1.9 - 0.95-3.54(-3) 1.2(0 13eCs 4.9 - 1.3-7.9(-2) 5.7 - 0.147-1.34(-2) 8.6(0 139Cs 4.0 - <0.0013-<5.6(-1) <4.5(- ) >8.9(0)
*** 1.63 ***
3H 1.60 1 0.05(-1) 0.05(-1) 1.0(0) 3.2 *** *** 14C 0.3(-6) 7.9 i 0.8(-5) 4(-1) 24Na 7.0 - 0.326-1.56(-2) 1.3 - 0.14-5.4(- ) 5.5 51Cr 2.4 - <0.0059-2.1(-3) 1.8 - <0.062-9.4 -5) 1.5 54Mn 7.3 - 0.40-3.3(-4) 4.4 - 0.071 -2.6( 5) 1.6 ssFe 5.320.5(-5) *** 1.45 0.04(-5) *** 3.7 59Fe 2.5(-5 0.085-2.1(-4) 1.5(-6 0.01 3-1.1(-5) 1.7 57C0 2.6(-6 0.062-2.2(-5) 5.9(-7 <0.032-2.0(-6) 4.4 seco 1.5(-3 0.021-1.49(-2) 1.1(-4 0.026-9.4(-4) 1.3 l 60Co 2.0(-4 0.028-2.45(-3) 9.9(-6 0.045-4.6(-5) . 2.0 63Ni 1.49 *** 2.8 0.1(-6) *** ' 0.04(-5) 65Zn 5.1(-6) <0.068-5.6(-5) 2.9(-7) 0.042-<1.5(-6) 5.3(0{ 1.7(1,
TABLE 3.15 (cont'd) 1 ; EANS AND RANGES FOR RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's FOR UNIT #4 CVCS MIXED-BED A DEMINERALIZER s Nuclide Inlet Concentration (uCi/ml) Outlet Concentration (pCi/ml) "Best Value" Mean Range Mean __ Range DF 89Sr 3.2 0.l(-5) *** 1.4 0.3 *** 90Sr 6.3 0.6(-7) *** 2.3(2) 91Sr 1.5(-4 7 1 3(-9)(-7) *** 91y <0.52-7.9(-4) 1.9(-5) <0.0024-<l.1(-4) 9(1)) 8.0(0 ' 4 g 2( 7) *** 2.4 1 0.2(-7) *** 93Y 9.4 - 1.7(0)
<0.015-5.1(-3 <9.3( ) >1.0( )
95Zr 1.7 - 0.036-1.3(-4))- 8. 3 - 95Nb 1.9 - 0.042-8.5(-5) 2.1 0.049-1.3(-4) 9.5 - <0.049-8.9 - 2.0 99Mo 3.1 - 0.046-2.43(-3) 1.6 - 0.052-<8.0 - 2.0 lo3Ru 2.8 - 0.083-<1.l(-5) 1.6 - 0.066-1.34 - 1.8 110 mag 1.3 - 7.0 - 0.013-2.9(-4) <0.0042-<3.1 4) -1.6 e w 124Sb 1.5 - 0.017-1.06(-4) 4.2 - 0.056-2.1(-5) 3.4 12sSb 3.8 - <0.023-3.1(-5) 1.8 - 0. 31 -6.3(-6) 2.2 139Ba 6.0 - 4.6 - 0.057-1.07(-2) 0.217-1.2(-2) 1.3 140Ba 6.3 - 0.33-1.06(- ) 4.5 - 1.4 140La 7.9 - <0.016-2.6-5) 0.096-<7.4j-6) 0.067-3.8(- ) 5.7 - . 1.4 187W .
- 9.8 - 0.37-2.3(-3 <3.6(- ) >2.7(1)
- 239Np 1. 5 - <0.18-5.9(-5) 7. 3(-6
' <0.014-5.9(-5) 2.l(0) l *** One measurement, only, for this nuclide. , 4 I L .
l l The behavior of the Cs decontamination factors, high instantaneous values during spiking and values which may sometimes be less than one during normal operation, suggests that a measured instantaneous DF is determined by the processes of ion exchange of Cs onto and off of the resin. Rapid increases in Cs inlet concentrations during spiking yield high instantaneous Cs DF's because t5e duration of the spike is short compared to the Cs ion residence tine in the demineralizer. That is, the quantity of Cs loaded onto the bed during a spike is small compared to the total Cs (or competing ion) inventory of the bed and it is the latter that is the dominant determinant of the concentration of Cs in the outlet via Cs ion exchange onto the resin and ion exchange off of the resin and migration dowa through the bed. The "best value" DF's listed in Tables 3.12 and 3.15 for mixed-bed demineralizers A and B when compared show that with the exceptions of 1311, lasCs, 57Co,110*Ag, and 12sSb, the DF's of mixed bed B are larger than the corresponding DF's of mixed bed A. This may be due to the longer bed life of the mixed bed A as compared with that of mixed bed B. O l l l l ! i I 94 l 1
- 4. BORIC ACID REC 0VERY AND LIQUID RADWASTE SYSTEMS 4.1 System Description and Sample Points 4.1.1 Boric Acid Recovery System The boric acid recovery system is used to recover boron from i the plant letdown flow and from the fuel pool or other boron containing
- streams. In addition to the recovery of boron, it also reduces the radioactivity concentration of the stream prior to discharging it to the
{ monitor tanks. The system is shown schematically in Figure 4.1 and systems components are listed in Table 4.1. The boric acid recovery system functions in the following manner. One of three 97,000 gallon tanks, holdup tank (HUT) A, B, or C is l processed through the systm.1 whenever it is filled. Normally one tank j is filling while another is being processed with the third tank being i full or empty as required. These tanks are filled from several sources ! which include: I
- 1. The letdown flow from both Units #3 and #4. These streams can either be sent directly to a HUT or they can be processed j through the letdown demineralizers.
! 2. Reactor Coolant Drain Tank. Water in this tank can come from any of several sumps or directly from the reactor coolant cold legs. During a cold shutdown the water from the reactor coolant loop would be processed through this tank if removal of the water is required for any reason. This tank also handles the excess letdown flow through the excess letdown coolers. The tank pumps automatically to a HUT whenever it is full. l
- 3. The spent fuel pit. Any time a spent fuel pit is to be drained, the water will be processed through the boric acid i recovery system by routing it through the holdup tanks.
l Figure B.17 shows the HUT levels during the seven-month measurement period. Feed to the boric acid evaporator (BAE) is passed through a base cation resin (30 ft3 volume, 5.4 ft2 cross sectional area) and a filter before being introduced into the evaporator itself. The base cation demineralizer is used to reduce th~e levels of soluble cationic species in the feed water. It also acts as a filter for some particulate species. The evaporator is an AMF type (identical to the radwaste evaporator) which has an upper evaporator section and a lower concentrate (bottoms) holding section. The feed is introduced into the evaporator through a gas stripper and into the lower concentrate holding section. A pump moves the bottoms to the upper section where it flows down heated trays 95
- - - - - . - g* * - -
t .e *. , - .,.m- g- q y -, ---y - -emp---a++m y
Holdup tanks Base cation demineralizers l AMF evaporator Sample 1 A (boric acid concentrator) l c x A B C Condenserf ) Concentrator i Gas Filter stripper J L Hea ter g y Yh +
+
Distillate
- B holding Feed tank tank
% M l r %
Samp;e 2 9 I $ Sample 3 Demineralizer C From
,l, radwaste evaporator y Sample 4 N l C X ]4 J
A Concentrates B hold tank Sample 5 -&C Filters
'W JL JL JL JL Reactc. int letdown Reactor coola it drain tanks Mannitgr on tgr Spent fuel pits 10 k gal. _ 10 k gal Concentrates holding tank INEL- A-11 200 Figure 4.1 Diag am of b)ric Acid hcovery Syster
TABLE 4.1 PRINCIPAL COMPONENTS IN RADWASTE AND BORIC ACID RECOVERY SYSTEMS Capacity Design Design j Each Pressure Temp Quantity Type opm Head psig *F Pumps Boric acid transfer 4* Canned 60 235 ft 150 250 Monitor tank 2* Centrifugal 100 150 ft 150 200 Ccncentrates holding 2* Canned 20 150 ft 75 250 tank transfer Reactor Coolant 1 per Horiz cent, 150 175 ft 100 267 Drain (A) unit canned Reactor Coolant 1 per Horiz co't, 50 175 ft 100 267 Drain (b) unit canned Chemical Drain 1* Horiz cent 20 100 ft 150 180 Volume Design each Design Temp Quantity Type tank Pre rure 'F Tanks Reactor Coolant Drain 1 per unit Horiz 350 gal 25 psig 267 Laundry & Hot Shower 2* Vert 600 gal Atm 180 Chemical Drain 1* Vert 600 gal Atm 180 Waste Holdup fl 1* Horiz 3242 ft3 Atn, 150 Spent Resin Storage 1* Vert 300 ft3 100 psig 150 Waste Condensate 2* Vert 1000 gal Atm 180 Concentrates holding 1* Vertical 925 gal Atmos. 250 Monitor 2* Diaphragm 10,000 gal Atmos. 150 l Monitor 3* 5000 gal Atmos l I Waste Holdupf2 1* Horizontal 24300 gal Atmos ! Btric Acid 3* Vert. 7500 gal Atmos. 250 Holdup 3* Horizontal 13,000 ft3 15 200 Design Design j Volume Flow Pressure Temp Quantity Type, ft3 opm asig "F Demineralizer Vessels Eysporator condensate 2* Fixed 30 25 200 250 Base - cation ion 3* Flushable 30 25 150 250 exchangers IQuantity per unit unless otherwise specified.
- Shared or capable of being shared by Unitg and Unit #4
3 under a partial vacuum. The solution temperature is normally about 150 F with a vacuum of 15 to 25 inches of Hg. Boiling occurs in the upper >section of the evaporator. The two sections are connected with a standpipe so that the upper section does not overfill. The bottoms, with
; fresh feed, are pumped to the upper section where boiling occurs and the overflow of concentrated solution can flow back to the lower tank. This ; operation continues until the feed is stopped. The evaporation continues to reduce the water content and then the bottoms are pumped to the concentrate holding tank (CHT).
The distillate is passed through a demisting screen and then allowed to condense into the condensate holding tank. Fromtherei$ispumped through a cooler, a demineralizer filled with a mixed-bed H -0H resin (30 ft3 volume, 5.4 ft2 cross sectional area), and a filter en route to ! one of two monitor tanks. Both of the filters in the stream behind the demineralizers are there to protect against resin fines and are not primarily designed for the collection of particulate radioactive material.
- A set of samples for the boric acid recovery system consisted of j the following
i 1. Inlet to the base cation demineralizer. This sample is l essentially a sample of the HUT. e
- 2. Base cation demineralizer effluent. This sample was also used as the feed to the evaporator. It must be noted, however, that there is a filter between this sample point and the evaporator itself.
- 3. The evaporator distillate which is also the inlet to the condensate demineralizer.
- 4. The evaporator bottum or concentrate.
- 5. The condensate demineralizer effluent.
- 6. The condensate demineralizer filter effluent. This filter was shared with the radwaste system and the effluent occasionally gave a mixed sample.
4.1.2 Radwaste System I An evaporator and a mixed-bed demineralizer are used to clean up waste water from drains, sumps, collecting tanks and, on occasion, water from the primary system through the reactor coolant drain tank. Figure 4.2 shows a diagram of this system and Table 4.1 lists system components. All liquid wastes processed through the evaporator feed from one 7 23,000 gal waste holdup tank (WHT #1). A second waste holdup tank 1 (10,000 gal WHT #2) is used as an overflow tank for WHT #1 but the water l must be sent back to WHT #1 to be processed through the evaporatur. Because the only feed path is from the WHT #1, water from any tanks to be processed will be mixed with the liquid previously pumped to the holdup tank. 98
h Finure 4.2 Dianram of Linuid Rarvaste System 1 Lsquid Rad %aste Treatment Systern Ecuipment drains a reactor coolant drain tank spent resin storage tank
~-
floor drains i 1 1 1 sumps reactor containment AMF (boric acid evaporator) l( now used for rad waste
- Condensor g
. Chem Laundry Laundry SS1
)'P hhhh i
i drain hot shower hot shower , C_oncentrator e, Ec tank tank tank Waste holdup p ggj di - tank - Heater 4 6m"-
- l 24 3 k gal Mv -
Feed tank ainer F7er SS2 ~~
){ SS3 -
_ , "1
$. , )[ SS8
\
Domineralizer
\ d {
334
' ' I 7 Waste holdup l D tank s2 '
M SSS From
, Waste W,,ste l Falter l _ boric condensate condensate l l a tank tank if l
Deminerafizer l l ll l o pwats Fdters 5 I l SS6 l W Return I '" I [ [
- to rad waste Demineralizer "
j holdup tank tn 5 k gal A ll 8 ll C f rnons 5 ~ SS9 1P Monitor Monitor
. .J db tank A tank B Discharge to k gal 10 k gal -
Discharge INEL-8-10 763 SS7 I l
m i I, The reactor coolant drain tank from both units can be drained to the containment sump which pumps to the WHT #1. All auxiliary building l and reactor building drains and sumps feed to the WHT #1 but do not have sample locations. Most of these drains and sumps pump down automatically, l making it difficult to detemine where water to be processed originated. l The two laundry and hot shower drain tanks and chemical drain tanks share one pump and can be pumped to the WHT #1 or to the two waste condensate tanks which can then be discharged or sent back to WHT #1 to be processed by the evaporator, depending on the activity level of the l liquid. There is a demineralizer and filter through which the waste condensate tanks can be recirculated. i A recirculation line on the discharge of the pump for WHT #1 '1 continuously circulates liquid in the tank through a 20 micron filter i even if the flow is shut off to the evaporator. The sample point for the tank is between the filter and the recirculation line so that this i
; is also a sample point for the feed to the evaporator. It is not possible to get a sample of the tank contents before it has passed through the filter.
The evaporator is an AMF vacuum distillation evaporator (identical to the boric acid evaporator) which consists of two 1000 gal. horizontal tanks, one above the other. The liquid in the lower tank (the feed j tank) is pumped to the top tank (the concentrator) where the distillation l occurs at a pressure below atmospheric pressure. There is a continual I circulation of water from the concentrator to the feed tank so the 4 concentrations in each tank are similar. f I The feed and distillation continues until the liquid in the feed tank reaches a boron concentration of 22,000 ppm, at which time the "7 liquid flow to the evaporator is shut off and the liquid in the evaporator (bottoms) is allowed to concentrate as distillation continues until the volume of liquid in the feed tank is reduced to 250 gal. At this time, the total bottoms volume is only reduced by 40% to 1250 gal. (i.e., the feed tank contains 250 gal. while the concentrator tank still contains I 1000 gal .) . The contents of the feed tank (20% of the total bcttoms volume) are then dumped. For this reason the bottoms activity is never j reduced by more than 20% when the bottoms are dumped. The bottoms that are dumped are drummed in concrete to be sent off site. j The liquid in the concentrator is boiled by a heating bundle and the vapor is condensed on cooling coils. The condensate liquid drains I into a 60 gal. helding tank which is pumped to the condensate demineralizer l
) and associated filter designed to collect resin fines. There are actually' two filters in parallel (see Figure 2.4), and condensate from both the boric acid and waste evaporators pass through both filters. From the filters the condensate goes to one of two monitor tanks where it is sampled before being released to the environment. If the activity level i
is not low enough or if the high activity monitor alam is tripped
]
during release, the liquid is sent back to WHT #1 for reprocessing. 1 100 I
. _ ~ . _-
The sample points for the evaporator package can be found on Figure 4.2, and were pulled at the same time. Sample Sample Point
- 1. Feed (Wilt #1) SS 2
- 2. Bottoms SS 3
- 3. Distillate SS 4
- 4. Distillate Demineralizer out SS 5 When an evaporator sample series was taken, the plant operating parameters and liquid chemistry data were obtained. Selected series taken included a sample of the outlet of the distillate demineralizer.
A new demineralizer package was installed for testing in early 107; which would take liquid from WHT #2 and pass it through a tiRr and then a preshielded removable demineralizer vessel. The demineralizer bed is replaced with a fresh one when necessary, and the loaded bed is shipped off site. Sample points for the test demineralizer system are shown on Figure 4.2. Sample Sample Point
- 1. Feed to Demineralizer (WllT #2) SS 8
- 2. The monitor tank the demin- SS 9 eralizer was feeding.
The monitor tanks in the new radwaste building were sampled as being representative of the output from the test demineralizer system. Samples of the chemical drain tanks and the laundry and hot shower drain tanks were not taken because they were all recirculated in common and would not have provided infonnation as to how much each contributed. Samples of sumps and drains could not Le obtained as they go directly to the WHT #1. The sample point on WitT #1 is after the pump and filter but before the recirculation lines (Fig. 4.2). It was not possible to take an unfiltered sample of the tank. Even though the sample point for WiiT #2 was not after a filter the liquid had been 41tered to some extent by the recirculation when it was in WitT #1 before being transferred to WitT
#2. The samples of liquid from WiiT #2 were obtained downstream of the pump that is used to send the liquid to the test demineralizer.
The bottoms sample from the evaporator was taken from the outlet of the recirculating pump and was diluted if the boron concentration was high enough to cause severe settling upon cooling. The distillate sample was pulled at the same time as the feed and bottoms. When samples were taken, boron concentration of the bottoms and conductivity of th' l distillate were determined. 101
1 I l l i a i In an attempt to increase sensitivity, resin concentrations were ! taken on distillate and demineralizer outlet samples in a few cases, but little added information was gained even though the increase in volume was a factor of 9. Auxiliary building and radwaste building monitor tanks were also Samples were obtained while the tanks were on recirculation. sampl ed. These tanks are representative of liquid which will be analyzed for radioactive content and released. The liquids in the auxiliary building
- monitor tanks are not indicative of the integrated radwaste evaporator j
.' distillate because boric acid evaporator disti'ilate is also added to ;
) these tanks. The radwaste building monitor tanks, however, contain the l integrated outlet from the test demineralizer system.
Levels of tanks in the radwaste system were obtained on a routine ] bas s (every 4 hours) during the in-plant measurement period. Appendix
. L) Tigures B.18-21 show plots of these tank levels.
4.2 Discussion of Measurement Data - Boric Acid Recovery System
.4.2.1 Measurements Two intensive measurement periods for the boric acid recovery system occurred. They covered the time periods 2/16-22/78 and S/2-
! 17/78. During the first period Unit #3 was at power and Unit #4 had j just been shut down (2/14/78) for steam generator repairs. During the a second measurement period both units were at power, but the boric acid
; recovery system was also processing water Trom spent fuel pit #3.
Radionuclide concentrations measured in samples from the boric acid recovery system during the two intensive measurement periods can be found in Appendix B, Tables B.14-B.20. Table 4.2 shows the feeds to the holdup tanks during these measurement periods. These data were taken from the operations logs and represent 4 the best estimate of the tank contents, although they may not include all of the sources of water in the holdup tanks.
. 4.2.2 Base Cation Demineralizers 1
The feed to the boron recovery system varied during the two l i time periods studied. Figure 4.3 shows the concentrations of 1311, 137Cs, and 58 Co that were fed to the base-cation demineralizer. The figure also shows the DF's of the demineralizer used to treat the feed to the evaporator. During the first two runs in February the contents of holdup tank A were being processed. Unit #4 was shut down on 2/14/78 and the resulting spiking increased the reactor coolant activity level. At this time the letdown flow was being directed to baldup tant. C. The i C HUT was lined up to the evaporator train and processing began on 2/20/78. The second two runs occurred while this water was being processed through the BAE train. During the February period and for the first two days of the May period, the base-cation demineralizer was loaded with a ! l 1 102 i
- - - . - - - . . - . - - . . - - - . - - - .-- -- - -.--.\
) i TABLE 4.2 BORIC ACID RECOVERY SYSTEM FEID CONTENTS AND SOURCES i i Measurement CVCS-j Date Holdup Tank Feeds to Holduo Tank 2/16/78 A ) 43 letdown 10:00,2/7/78 i 2/17/78 A .[ #3 RCS thmugh RCDT 08:30 on 1/31/78 ll to 15:00 on 8/31/78 2/20/78 C #3 and " 'etdown from 10:30 on 2/!4/78 i to C . , 2/16/78 i 2/22/78 C
~
) E/2/78 A t
#3 RWST from 07:45 on 4/29/78 to 12:55, 4/29/78 l ) #3 SFP 19:10 on 5/1/78 to 06:30, 5/2/78 #3 .
letdown from 08:00, 4/26/78 to 06:30, 5/2/78 5/3/78 A s I 5/4/78 C #3 letdown 12:30 on 5/2/78 to 12:00 5/3/78;
#3 SFP 12:30 to 13:11 on 5/2/78 and l 06:10 to 12:00 on 5/3/78 5/9/78 A #3 letdown 03:05 on 5/8/78 and #3 SFP 02:45 5/A/78; 4
#3 letdown A #3 SFP 12:00 to 17:30 on 5/3/78 a 5/11/78 C #4 letdown 03:05 on 5/8/78; #3 & #4 letdown :
i 16:30 on 5/6/78; SFP #3 04:10 to 08:45 on
, 5/8/78; #3 SFP 19:40 on 5/6/78; #4 letdown 12:30
- 5/2/78 to 05
- 50 5/4/78 1
! 5/15/78 B
#3 letdown and #3 SFP 06:30 to 12:30 on 5/2/78 l
CHT on 5/6/78 0 19:40
- #3 letdown 05:50 5/4/78 to 16:30 5/6/78 i
5/17/78 A #3 & #4 letdown 16:45 5/11/78 to 06:00 5/14/78
#3 letdown 06:00 5/14/78 to 07:45 on 5/17/78 ;
I l l ! i i l l < I l
- 103
( l
, . . . . _ . . - - - - , , _ . _ _ , , . . - _.-,..._.....,,._.,_,_._..,__._.__....,.._..c_. . . . - . _ . . _ . . . . _ . . . _ . . - -
E i Finure 4.3 Radionuclide Concentrations in Inlet to Boric Acid Recovery System and Demineralizer DFR i i 104 4 _, , , , , , , , % , , , , ,, , , , , , ,, , ,, ,_ 10
- Boric acid evaporator unit :
_ 1 Base cation demineralizer - -
~
fi e 131: o op
/\
~
a 137Cs o DF
- / \ e 58Co o DF -
/
10 / - 103
- j Demin C ---- Demin A :
~ ~
l - -
-h / -
1
-y / -
7
; - \
\ f K g 9
] 10-3 -
k /
\ -p- 10 2
8 : / I p _ 4 4
/ l k\ -
I o
-\
l / i / ;
\
\ / sr i c _\ /
/ I y o
8 g l \ \ c /
$ 10-4 - k' g
Y - 10 1 , E \l \E ' a 2 l l e- _ l 4 A
\l &-
N l
- s g
<y / T I 10-5 _p. g __ 1 5 7 E i
10-6 I i t i I I I g iiiiiI i1 1 1 I II I I I I j o-1 4 16 17 18 19 20 21 22 1 3 5 7 9 11 13 15 17 February 1978 May 1978 INEL A-10 773 i 1 ! l l i' 104 1 I
. - ..._,,.m, _ . - __ _ --, _.
M. l l cation resin only. This is the nonnal resin used for this demineralizer,
- and its purpose is Cs removal. The cation resin had a reasonable DF for the crud-associated-radionuclides, removing over 90% of the seCo.
During the May sampling period, demineralizer C was in service for the first two samples taken. The feed for these samples was a mixture of letdown water from the two units and water from the #3 spent fuel pie.
- which was being drained for repairs. On 5/5/78 demineralizer C was
- taken out of service and demineralizer A was put into service. This new
- demineralizer contained a mixed-bed resin (H-OH form) for the removal of
- both cations and anions. The DF for 131I increased as a result of this resin change and the DF for 137Cs decreased. It is also interesting to note that the DF for seCo also changed with the mixed bed resin in service. It dropped from a value of about 10 to a value between 2 and j 3.5.
4.2.3 Boric Acid Evaporator The evaporator takes its feed from the outlet of the base , cation demineralizer cokmn. During the two sampling periods the evaporator saw a variety of feed activities. The spike activity from the 2/14/78 shutdown was fed to the evaporator on 2/20/78 and 2/22/78. Large quantities i of fuel pit water, which had very low levels of the shorter-lived radio-
~
nuclides including 131I, was processed through the system during the May sampling period. Figures 4.4 and 4.5 show the concentrations of 131I and 58C0 in the feed, bottoms and distillate along with the DF's* for both sampling periods. These two radionuclides have the highest activity levels of all of the radionuclides detected in the evaporator feed. The iodine is of interest because the evaporator is the source of iodine in the auxiliary building ventilation system (see Section 8). Data for SeCo are plotted to show the behavior of the crud-associated radionuclides. Cesium is not shown since its concentrations were very low in the feed to the evaporator. I The plot of the iodine behavior shows how a DF of less than one can i occur for a short period of time with a well functioning evaporator. In j May the concentration of 131I in the feed was dropping very rapidly due to the processing of fuel pit water by the evaporator (Figure 4.4), but the drop in activity in the bottoms tended to lag the drop in the feed. As a result, the DF was less than one for two runs until the bottoms i concentration was sufficiently reduced to give a distillate with a lower concentration than the feed. Note that the concentration of the bottoms and the distillate follow a very similar pattern. Cobalt-58 did not i exhibit this behavior. One sample of BAE bottoms, taken on 5/17/78, was analyzed for radionuclides. alpha-emittingPuand2+1(-8)for uCi/ml for 23 _ ResultinguPu.concentrationswere2+1(-8) 239, -~
*For an evaporator the DF is defined as the ratio of feed to distillate concentrations.
i 105 l
Figure 4.4 131I Concentration; in BAE Feed, Distillate, and Bottoms and BAE DF's
- I I i 10 3 i i i i i i i i i i i i i i i i i i i l- -
1 j R Boric acid evaporator unit : )
~
131 1 Concentration - 1
/ \
_ / e Feed _
/ a Bottoms
/ \ e Distillate 10 \ 102 DF
\ 0
/ U Bottoms transferred 2
/ g :
p _ j [10-3 __d gl j- -- 10 1h U I \ /\: E a _ - n _ l \- %
\
8
/ I
[ g - l
/
\-- u.
\ O i / b 8
E E '} / B
, o 10~4 --
l g -- 1 g 2 { / E $ Z / 2 _ l/ _ _ y - 10-5 __ _ 10-1 x10 - i < l l-j " 10-6 I I I I I I I yiI I II I I I I I I I I I I 2 16 17 18 19 20 21 22 1 3 5 7 9 11 13 15 l February 1978 May 1978 inet.A-10 774 i 106
a. Figure 4.5 j 5800 Concentrations in BAE Feed, Distillate, and Bottoms and BAE DF's 1 10-2 O \ 4 1 1 I i i iiiI I iiI I I I I i i l- 10 1/ J 2 / I T :
- / \ n _
_q / \ J -
\ #
~
Y 1 l' ~ l \ I I 10-3 _ l __ 10 3 j Z l i
- I i
\
_ \ \ -
\ g p 10-4 _ l '
k 102g
- s
- g fl + - d _: 5 0 -
\- g
- L / \-
I
/
B o g Boric acid evaporator ! / 2 g - 58Co Concentration M ~ 05 i N -
- o e Feed lii 10-5 _ & Bottoms 10 1 2 e Distillate E $
o DF T - -s.
- 10-6 _ ._. 1
' ' ' 8 'I 8 II '''I 10-7 16 17 18 19 21 22 I
'% 1 3 5 7 9 11 13 15 17 I I 10'l l February 1978 May 1978 INEL-A-10 772 1
107
4.2.4 Condensate Demineralizer and Filter After leaving the evaporator the distillate is condenrad and then goes through the mixed-bed condensate demineralizer for furtht:r 4 cleanup. Decontamination factors measured across the condensate i demineralizer were, in general, much lower than those measured for the letdown demineralizers. The inlet radionuclide concentrations were . also much lower than the corresponding input to the letdown demineralizers. Figure 4.6 shows a plot of 131I concentrations in the inlet to the BAE condensate demineralizer and demineralizer DF's. After leaving the condensate demineralizer the liquid passes through a filter then goes to one of the monitor tanks. During nonnal operation, the valve on the line connecting the inlets to the filters in the boric acid recovery system and the radwaste system is open. Hence, either i stream sees two sets of filters hooked up in parallel (see Figure 2.4). < In order to obtain DF's across the filter in the boric acid recovery ! system, this valve was closed temporarily. The measured DF's for this I filter exhibited considerable variation for the crud-associated radionuclides, ranging from about 0.5 when crud-associated radionuclides were washed off the filter up to about 10. 4.2.5 Summary of Results All of the individual DF's for each system component are
- presented in Tables 4.3 through 4.6.
i ' To determine "best value" DF's for the boric acid recovery system, j the analysis was split into two sections. The first was the time period 4 when the base cation demineralizer C, which contained a cation resin, was in service. Cs isotopes and low As expected,131I.this DF's for The demineralizer second period gave washigh whenDF's the base for the 4 cation demineralizer A was in service with a mixed-bed ion exchange I resin. This demineralizer gave lower DF's for the Cs isotopes than the
! cation bed but higher DF's for 1311. Next, average values of concentrations were used to calculate "best value" DF's for the two measurement periods.
)i Tables 4.7 and 4.8 present means and ranges for radionuclide concentrations j and "best value" DF's for system components during the periods when base cation demineralizer C and A were in servicgb' Figures 13 4.7 and 4.8 present values for the radionuclides 1311, Cs and 57-soCo for the periods when the two different types of resins were used in the base cation demineralizer column. Note that for decreasing feed, the values of the evaporator DF for 1311 (Figure 4.8) approaches 1.0. However, in 4 the operation of an evaporator of this type, the distillate quality is dependent on the bottoms concentration and not directly on the feed. If we calculate the ratio of bottoms to distillate concentration, we arrive at the values shown in Figure 4.9. Examination of the data obtained for lodine and cesium when the
- base. cation demineralizer was loaded with a mixed-bed resin indicates that the DF's are a function of the inlet concentrations of the 108 i
i 3_
- --- _-< _.. m- e--,_, v,-. ~ . , . . _ . - - _ . - - - - . ~ . , - , , . - _ . ,,w , ,, _ . .-m.-
Figure 4.6 1311 Concentration in Inlet to BAE Condensate Demineralizer and Demineralizer DF 104 - i i i i i i i NI i i i i i i i 104
~
BAE condensate demineralizer 1313 e inlet concentration - a Decontamination factor _ e 104 -- -- 103 e
- _ \, _
e 10-5 _. - 102 E - e', - g 1 6 - _ 2 3 - ~ E c - a e t
- _ .c c a- a _
S O C 8
- 8 4
o a a
- U 10-6 __
/' _ 10' O 2\
2 ea 1 10-7 -- - I 10-8 1 i i I l 1 I i i i i i i i i i 10-1 l 16 17 18 19 20 21 22 1 3 5 7 9 11 13 15 17 Feb.1978 May 1978 INEL A-10 979 109 l
TABLE 4.3_ DF's FOR BASE CATION DEMINERALIZERS Base Cation Demineralizer C (Cation Resin) Nuclide 15:10, 2/16/78 13:15, 2/17/78 18:15; 2/20/78 16:05, 2/22/78 09:05, 5/2/78 09:52, 5/3/78 131I 1.07 0.03(0) 1.00 1 0.02(0) 1.92 1 0.05(0) 1.45 3.8 0.10(0) 0.l(-1) 7.6 0.3(-1) 134Cs >1.5(3) >1.4I >1.5(3) >1.5(4) 5.8 2.9(1) >l.4(1) 137Cs >4.~0(2) >1.0(2)2) >5.3(2) >4.5(3) >3.l(0) >2.8(0) 51Cr 2.0 0.1(0) 1.4 0.4( 2) 1.6 0.1(0) 3.3 0.2(0) 7.0 i 1.5(-1) 1.4 0.5(-2) 54Mn 4.3 0.1 } 1.3 0.1 1.23 0.04(2) 5.9 0.2(2) 3.8 0.4(0) 59Fe 1.710.2 1.3 0.1 1.7 0.1 5.1 0.3-2) 2.22 0.04(0) 8.8 0.4(-1 ) 3.3 1.3-2) 5700 2.5 1 0.2 4.3 1.6 1.6 0.3 1.4 0.8(2) <2.3(0) 5.1 1.3 -2
. seCo 4.4 0.1 9.6 4.0 0.1 1.61 0.05(2) 1.46 0.03(0) 3.0 0.1-2{ ,
g 60Co 5.9 0.1 1.44 0.01(1) 2.03 0.04(1) 1.30 0.02(2) 2.45 0.09(0) 1.00 0.01(-1 ) 65Zn 6.4 1.3 2.8 0.7 0) 4.5 1.0(0 3.7 0.81) *
<5.3(-2) 95Zr 2.6 0.1 1.6 0.1 0) 1.1 0.l(0 5.3 0.2 0) 9.322.2(-1) 4.6 0.5 -
9sNb 2.7 0.1 1.5 0) 1.0 0.1(0 5.0 0.3 0) 6.9 0.4(-1) 3.9 0.5 - 99Mo <1.4(1) 2.89 0.06(0) 1.93 0.05(0) 5.8 1.0(0) 3.5 i 0.6 - 103Ru 5.0 0.5(-2) 1.26 0.04(0) 2.1 1 0.9(-1) 2.7 0.5(0) <3.6(-1) 4.7 1.2 - 106Ru <2.6(-1) 9.7 0.8-1) <2.5(-1) *
<l.2(0) 1.7 0.6(-1 110 mag <5.3(1) 3.7 1.0 0) <3.6(1) <9.2(1) 8.7 1.8(-l 4.0 i 0.5(-2 124Sb 1.13 1.43 0.04(0) 9.810.2-) 9.7 0.3 - ) 1.51 0.03(-1) 9.4 0.9(-1 2.5 0.2-1) 12sSb 0.03(0) 9.8 1.2 - ) 5.1 0.5 - ) 1.39 0.06(0) 9.7 1.4(-l 4.0 1 0.3 -1)
- 140La 5.8 1 0.6(0) 2.5 0.30 2.8 0.1 0 9.1 0.2(0) 2.4 0.3(-l 1.4 0.1 0) 141Ce 1.9 0.9(1) 1.6 0.5(0) * * *
<3.6(-1) 144Ce <6(-1) *
- 1.5 1.0(-1 )
* - Radionuclide not detected.
TABLE 4.3 (cont'd) DF's FOR BASE CATION DEMINERALIZERS Base Cation Demineralizer A (Mixed-Bed Resin) Nuclide 14:34; 5/4/78 14:50; 5/9/78 11:58; 5/11/78 10:35; 5/15/78 14:46; 5/17/78 131I 6.6 0.5(1) 1.05 0.07(1) 1.46 0.03(1) 1.83 0.05(1) 3.0 1.3(0) 134Cs 1.1 0.3 2) 2.7 0.2(1) 8.5 0.2(1) 4.9 0.5(1) 2.7 0.3(2) 137Cs 6.3 0.31) 1.40 0.06(1) 4.5 0.3(1) 2.2 1.3 0.1(1) 0.1(2) 51Cr 1.4 0.3(0) 2.3 0.2(0) 1.6 0.10) 2.3 0.2 2.4 54Mn 2.02 2.7 0.9 0) 0.09(0) 0.1 3.7 0.4 0) 4.5 0.3 4.0 0.7 59Fe 1.5 0.4(0) 3.1 0.6 2.2 0.3 0) 1.6 0.4 1.4 0.7 57Co <1.3(0) 1.8 0.7 2.0 0.30) : 3.3 0.9 1.8 0.6 SECo 2.5 0.3(0) 2.81 0.05(0) 3.3 0.1 0) 3.6 0.1(0) 2.1
-; 60Co 2.2 0.1(0) 0.1(0 2 2.6 0.1(0) 3.4 0.2(0) 3.4 0.1(0) 3.3 0.4 0) ssZn <l.1(0) <2.4(0)
<2.5(0) <2.8(0) 1.1 0.5 0) 95Zr 2.0 0.3(0 2.6 0.3(0) 2.7 0.3(0) 2.0 1 0.2(0) 2.7 0.50) 95Nb 1.7 0.1(0 2.7 0.2(0) 2.2 0.1(0) 2.0 99Ho 5.5
- 0.1(0) 2.1 0.4 0) 3.2(0 * *
- 103Ru 2.3 0.9(0) 2.8 0.6(0) 1.6 0.1(0) 2.4 0.2(0 1.0 106Ru
- 1.5 0.2(0)
- 1.1(-1) <1.7(0) 3.9 0.6(0 <9.4(-1) 110 mag 2.6 0.7(0) 2.2 0.4(0) 2.8 0.3(0) 3.3 0.2(0 1.5 0.30) 124Sb 5.8 0.9(0) 7.6 1 1.1(0) 7.1 0.8(0) 6.7 0.4 0) 3.7 12sSb 1.0 6.2 1.1 0) 0.5(1) 0.7(0) 1.0 0.2(1) 9.2 0.50) 5.7 1.60) 14cLa 2.0 0.5(0) 9.4 0.7(-1) 2.7 0.3(0) 4.1 0.3 0) 2.3 0.9 -1) 141Ce <8(-1 ) >6.2(0) <2.4(0)
- 144Ce * <3.3(0) 1.1 0.4(0) <1.6(0) <1.3(0) 1.4 i 0.7(0)
* - Radionuclide not detected
TABLE 4.4 - DF's FOR BORIC ACID EVAPORATOR Nuclide 2/16/78 2/17/78 2/20/78 2/27/78 5/2/78 5/3/78 131I 1.24 1 0.04(1) 3.1 0.1(1) 6.9 0.2(2) 3.7 1 0.2(1) 1.05 0.03(1) 1.72 1 0.08(1) ,
*
- 3.0
- 1 34Cs <1.1(1) <2.3(2) 1.8(0)
- 137CS <4. 3(1 ) <2.1 (2) s1Cr >3.9(3) 1.4 0.8(3) <l.3(5) 8.5 7.3(2) >3.5(1) >4.9(2) 54Mn >2.0(2) 9.8 2.7(1) 1.8 1 0.2(3) >3.4(1) 6.6 0.8(0) 4.3 i 0.2(3) 4 59Fe 5700
>7.6(2) >6.7(2) 5.8 0.5(3) 8.7 1.1(1) >1.5(1) 6.5 3.l(2)
>6.5(1) >6.8(1) >1.0(3) >l.6(0) >9.9(1) 5dCo 60 Co 3.2 0.8(3) 1.8 0.3(3) 1.29 0.04(4) 6.910.3(1) 7.3 0.4(1) 7.8 0.8(3)
>1. 6 6.2 4.2(2) 2.48 1 0.09(3) 3.6 0.6(1) 1.6 i 0.1(1) 6.1 2.4(3)
-: ssZn >1.2 >6.6(1) >4.3(3) >5.3(0) >9.4(1) 9.9 1 2.8
~
N 95Zr >7.5 >4.8(2) 7.0 1.4 7.1 1 4.9(1) 1.8 0.6(1) 9sNb >l .1 4.9*1 3.3(2) 4.0 1.5 5.812.5(1) 5.8 1.5 (1) 2.2 i 1.4 99Mo >9.3 8.5 1 3.1 9.3 1.3(1) >3.21) 3.1 i 1.a lo3Ru >l.0 4.3 1.4(2) 5.2 1.0 >2.3(1) >2.1 1) 1.0 0.4 106Ru >6.3 >7.9(2) >2.3(4) <1.7(-1) >6(0 >4.2(1) 1104g 7.8 i 3.9(1) 6.4 1 1.8 1) 3.6 0.2(2) 1.6
- 0.2(1) >1.7(1) 1.010.2(2) 124Sb >9.6(3) 4.2 1.3 3) 3.3 0.2(3) 9.6 1.5(1) 4.9 2.5(1) 1.6 0.8(3) 12sSb 2.4 0.6(3) 2.5 0.5 3) >3.9(1) 5.9 0.4(1) >3.9(1) 3.3 0.6(2) luoLa >7.1(1) 2.4 1.1 1) >1.l(5) 4.3 0.3(2) >l.8(2) 1.5 0.5(2) 141Ce >6.5(0) >4.6(1) >l.2(1) 144Ce >2.2(2) <5.3(1) * *
- 1.9 1 1.l(1)
- * - Radionuclide not detected.
TABLE 4.4 (cont'd) DF's FOR BORIC ACID EVAPORATOR I Nuclide 5/4/78 5/9/78 5/11/78 5/15/78 5/17/78 1311 3.1 0.3(-1) 9.2 0.8(-1) 1.9 0.1(0) 9.5 2.7 2.8(0) 1.3(0) 134Cs >9.0(0) 7.9 0.9(0) 1.710.8(2) 2.0 1.5(2) 7.1 2.4(0) 137CS 8.9 1 1.2(l) 9.7 0.4(0) 5.4 2.5(1) 2.1 1 1.5(1) 5.7 4.1(2) 51Cr 5.8 3.1(1) 9.4 1.6(1) 2.0 0.4(2) 1.2 0.5(2) 7.8 2.0(1) < 54Mn 2.1 0.6(1) 6.0 0.5(1) 2.1 0.8(1) 6.8 0.6(1) 1.7 0.7(1) 59Fe >2.6(1) >5.5(0) 9.4 2.8(1) >1.4(1) >4.7(0) 57Co saco
>1.2(1) 3.6
>3.1(0) 1.3i0.8(2) >3.4(0) 3.3 1.5
' 0.6(2) 1.5 0.3(2) 4.2 1.8(2) 2.3 0.5(2) 1.2 0.3 i 60Co 1.9 0.3(2) 9.4 0.3(1) 8.3 0.4(1) 9.7 0.6(1) 7.6 i 1.0
- ssZn >1.0(1) >3.2{0) >1.4(0) >2.3(0)
>2.9(0) 95Zr 4.1 0.5 8.5 1.2(1) 6.7 0.9(1) 8.0 2.9(1) 7.6 2.3(1) 9sNb 5.1 4.6 9.0 1.1(1) 7.8 0.8(1) 1.4 0.2(2) 8.6 1.9(1) 99Mo 9.4 6.4 * * *
>6.7(0) 103Ru >7.4(0) >5.9(0) 3.6 : 0.6(1) 1.6 0.5(2) 8.1 losRu 2.5(1)
<1.l(1) 7.7 4.2(1) >3.1(0) >5.4(0) >4.8(0) 11aAAg 4.3 0.7(0) 5.1 0.9(1) 2.8 0.2 2.6 0.31) 2.8 0.5 124Sb 5.1 1.1(1) >8.1(0) 5.8 1.6 1.5 0.5 2.3 0.7 12sSb 2.0 1.0(1) 2.6 0.4(1) 1.4 0.3 8.0 1.0 1.5 0.4 140La >1.1(1) >7.4(0)
>5.6(0) 6.1 1 3.6 141Ce >1.7(1) 2.7 0.9(1) *
>1.21))
>9.2 -1 144Ce >3.3(1) 1.3 0.4(1) >3.8(0) >2.80)
* - Radionuclide not detected.
TABLE 4.5 DF's FOR BAE CONDENSATE DEMINERALIZER Nuclide 2/16/78 2/17/78 2/20/78 2/22/78 5/2/78 5/3/78 131I 6.5 0.2(0) 4.7 0.2(0) 9.1 0.3(0) 1.17 0.03(1) 1.32 0.07(1) 3.8 1.3(1) 134Cs 3.8 2.2(-1) 1.0 0.7(0) 1.9 0.6(-1) <2.5(0) >1.7(0) <1.2(0) 137Cs <2.3(-1) 3.5 2.3(-1) 3.6 1. 3(-1 ) <2.8(0) <5.9(-1) 3.1 0.8(-2) 51Cr
- 1.8 4.8 4.2(-1)
- 1.7(0) <9.2(-2) <1.4(0) 54Mn <5.5(-1) 4.7 1.6(-1) 5.2 0.6(-1) <1.9(0) >6.9(0) >2.2(0) 59Fe 2.3 0.8(0) 1.2 0.6(1) >2.6(0) 57Co * * * * *
- seCo 2.4't 0.6(-1) 4.0 0.7(-1) 1.06 0.04(-1) 8.0 0.5(-1) 4.9 0.9(0) 4.6 0.5 d 60C0
- 6.3*1 5.0(-1) 5.5 0.3(-1) 1.3 0.30) >2.2(1) 1.4 0.3 95Zr 1.1 0.4(0) 2.4 1 2.7 0) >1.3(0) 1.4 0.7
- 2.2 ssNb
>8.2(-1) >1.4(0) 1.31) >2.2(0) >5.0 0) 99Mo <1.1(0) 6.4 2.7(-1) 9.1 2.1 1) *
.>1.4 0) 103Ru 7.1 3.2(-1) 2.0 1.6(0) <4.4(0) >1.90)
*
- 0.6
- 106Ru <1.3(-1) 1.5(2) <1.0 1) 11onag 2.9* i 1.5(-1) 2.0 0.6(-1) 2.8 0.2(- ) 1.810.4(0)
- 2.9 0.5(1) 124Sb 1.2 0.5(0) 3.4 0.4(0 5.2 8.1(1) >l.3(0)
- 12sSb >4.9(0) 3.9 0.9(1) 2.9 0.4 2.0 1.4(1) *
>2.6(0) 140La *
>1.8(0)
<4.8(-1)(0
>4.2(0) 7.7 3.7(0) 141Ce 144Ce >1.6(0) >l.2(1)
* - Radionuclide not detected.
TABLE 4.5(cont'd) DF's FOR BAE CONDENSATE DEMINERALIZER Nuclide 5/4/78 5/9/78 5/11/78 5/15/78 5/17/78 131I 1.3 0.l(1) 8.0 1.7(0) 2.2 0.4(1) 3.2 2.6 1.2(1) 0.8(0) 134Cs <2.1(0) 2.4 0.6(1) 7.5 i 3.5(-2) 1.5 1.2 -1) 1.1 0.5(0) 137CS 2.9 0.4(-l) 9.7 0.7(0) 1.0 0.9(0) 2.2 2.2 0) 3.3 1 3.1(-2) 51Cr >1.l(2) 6.321.3(-1) 2.1 0.7(0) >1.7(-1) >7.8(-2) 54Mn 4.1 1.2(1) >9.2(-1) 7.8 2.9 5.5 3.3 2.9(1)
- 3.2(-1) ssFe 57C0 *
<l.2(1)
>1.4(-1) (0) *
>4.3(-2) >5.4(-2) seCo 4.5 0.7(0) 4.510.9(-1) 5.6 2.7(-1) 3.6 1 0.9(0) 1.9 d 60C0 1.3 0.2 1.0(0) 1 .71 i 0.04(0) 5.0 4.2(0) 6.5 1.2(0) 4.8 0.8(0) u' ssZr 8.2 3.4 8.5 1.4(-1) >3.0(-1) 5.8 1.3 4.3(0) >1.5(-1) 95Nb 1.5 8.4 0.5(-1) 5.5 1.7(-1) >5.5(-1) 7.1 99Mo 2.4 1.4
- 3.1(-1 1.0 1 0.6
- 1.6i1.1(0))
103Ru 106Ru
<9.2(0) >4.0(-1) *
(0) >1.9(-1) >1.4(-1)
>9.8(0) >7.9(-1) *
- 110 mag 1.0 0.21) 1.9 0.2(-1) 3.1 0.5(0) 5.9i1.7(-1) 1.5 2.4 1 0.7 0) 124Sb 0.30) <2.3(0) 2.6 1.4(0) 1.9 1.2(0) ' 2.6 1.6 0) 12sSb 1.2 0.1 0) 9.4 1.2(-1) 4.3 1.5(-1) 3.2 1.2(-1) 2.0 140La * * *
- 0.4 0)
* * >7.2(-2) 141Ce 144Ce * *
>l.l(-1) *
- 4.4 3.6(0) * *
* - Radionuclide not detected.
TABLE 4.6 DF's FOR BAE CONDENSATE DEMINERALIZER FILTER Nuclide 2/16/78 2/17/78 2/20/78 2/22/78 5/2/78 5/3/78 131I 9.6 0.3(-1) 1.07 0.03(0) 1.00 0.02(0) 1.09 2 0.03(0) 9.3 0.5(-1) 9.7 0.3(-1) 134Cs 2.7 1.3(0) 1.1 0.9(0) 9.4 1.3 -1) >3.0(-1)
- 8.2 3.0(0) 136Cs 9.2 6.8-1)
- 1.0 1 0.2 0) 137Cs 1.2 0.30) 6.5 1.6(-1) 9.1 0.3 -1) 5.6 3.3(-1) 1.0 0.3(0) 8.0 1.1(0) 51Cr
- 9.3 8.6(0) 4.0 1.4(0) >4.l(0) <1.7(0) 54Mn >3.2(0) 2.6 1.8(0) 4.7 1.0(0) >5.d(-1) <1.7(0) <1.7(0) 5.9 *
- 59Fe *
<9.3(-1) 4.1 2.3(0) 3.6(-1) *
* * <2.7(0) 57Co <6.2(0) <2.2(0) .
-: seCo 4.9 0.5(-1) 1.1 0.l(0) 4.3 0.2(0) 8.8 0.4(-1 ) 2.5 0.7(0) 7.1 1.0(-1) og 60Co <4. 6(-1 ) 1.6 1.2(0) 3.5*1 0.3(0) 1.1 0.1(0) <7.2(-2) 8.6 0.9(-1 )
6sZn 95Zr *
<6.2(0) >2.6 >5.6 -1) >9.3(-1) 95Nb *
<7.7(0) <3.7 >1.4-1) *
<4. 6 (-1 )
99Mo *
<9.8(0) >2.6 >2.1 -1) *
- 103Ru
- 1.1 0.7(1) 1.1 1.4(0) 3.1 3.9(-1)
- 106Ru *
>7.9(0) *
>2.8(-1) >l.2(0) 110qAg 1.0 0.2(0) 1.0 0.2(0) 8.3 0.5(-1 8.0 3.0(-1) 4.7 2 0.8(-1) 124Sb *
>2.0(Of 5.5 2 1.9(0)) >3.6(-1) *
<8.5(-1) 12sSb *
>1.4(-1) 2.8 0.8(0) >5.3(-1) *
<l.0(0)
* * * >3.7(-1) 14cLa >1.1(0) * *
- 141Ce 144Ce <1.4(0)
* - Radionuclide not detected.
9
TABLE 4.6 (cont'd) DF's FOR BAE CONDENSATE DEMINERALI2ER FILTER Nuclide 5/4/78 5/9/78 5/11/78 5/15/78' 5/17/78 1311 9.4 1.1(-1) 1.3 0.5(0) 1.2 0.3(0) 1.2 5.4 0.5(-1) 1.3(0) 134Cs 2.1 0.1(0) 7.8 1.9(-2) 6.9 3.5(-1) 4.4 13sCs * * *
- 1.9(-1) 1.5 0.8(0) 137Cs 2.4 0.2(0) 9.7 8.2(0) 6.7 5.7(-1) 4.3 3.1(0) 1.1 1.1(1) 51Cr *
>3.40) >2.6(0) *
- 5'+Mn >1.3(0) <1.7 0) 1.8 0.4(0) 1.1 0.6(0) 5.0 4.2(-1) ssFe * *
>1.5 -1) *
- 57Co * * * * '
_. 58C0 7.8 0.5(-1) 1.8 0.2(1) 1.8 0.7(0) 9.0 0.8(-1 6.6 1.42 3.5(-1) 0 60Co 0.05(0) 2.1 0.9(1) 6.1 5.1(-1) 2.0 0.7(0)) 2.3 1.0(0) 65Zn * * *
- ssZr >1.5(0) >8.7(-1) <1.7(1) >3.6(-2)
- 95Nb 8.9 7.3 8.1 6.5(-1) 6.4 2.1 <1.4(1) 7.4 3.0 99Mo >7.6(-1)
(-1) *
>3.7(-2) (0) *
>1.7(-1) (0) 103Ru >2.3(-1) <1.5(1)
- 106Ru * <2.4(1)
<1.6(-1) * * *
- 11onWg 7.1 1.7(-1) 6.9 1.7 8.112.0-1) 4.8 1.5(0) 1.2 0.3(0) 124Sb >3.7(0) >9.3(-1) (0) 6.0 2.5 -1) 1.1 1.0(0) 6.5 4.5(-1) 12 5Sb 1.0 0.1(0) 2.4 0.3(0) 5.3 1.9 0) 3.5 1.3(0) 1.1 0.3(0) 140La * * *
<4.8(0)
- 141Ce * *
- 144Ce * * *
>6.2(-2) *
* - Radionuclide not detected.
l
TABLE 4.7 BORIC ACID RECOVERY SYSTEM WITH BASE CATION DEMIN. C MEAN RADIONUCLIDE CONCENTPATIONS AND "BEST VALUE" DF's Base Cation Demin. C (Cation Resin) Inlet Concentration (pCi/ml) Outlet Concentration (pCi/ml) "Best Value" Nuclide Mean Range Mean Range DF 131I 3.3(-3) 0.010-1.00(-2) 2.1(-3) 0.136-5.2(-3) 1.6(0) 134Cs 9.0(-4) 0.031-2.16(-3) <1(-6) >9(2) 137Cs 1.5(-3) 0.0046-3.24(-3) <5.2(-6) >2.8(2) 51Cr 3.5(-4) 0.019-8.2(-4) 2.7(-4) 0.30-5.0(-4) 1.3(0) 54Mn 9.7(-4) 0.0058-2.57(-3) 2.7(-5) 0.039-1.13(-4) 3.6(1) 59Fe 1.05(-4) 0.0010-2.5(-4) 6.2(-5) 0.025-1.44(-4) 1.7(0) g 57Co 3.5(-5 0.04-9.1(-5) 3.1 - 0.39-7.8( 6) 1.2 seCo 2.0(-2 0.0074-4.29(-2) 8.3 - 0.125-2.5-) 2.4 soCo 4.6(-3 0.0073-1,35(-2) 2.4 - 0.476-7.3 - ) 1.9 65Zn 2.2(-5 <9(-7)-5.1(-5) 6.2 - 0.09-1.7( 5 3.5 95Zr 4.3(-5) 0.032-1.15(-4) 3.3(-5) 0.54-6.9(-5) 1.3(0) 9sNb 7.3(-5) 0.048-2.02(-4) 5.8(-5) 0.14-1.23(-4) 1.2(0) 1.7(-4) 99Mo 0.012-6.13(-4) 7.4 -5) 0.0067-2.12(-4 2.3(0 103Ru 9.2(-6) 0.09-3.45(-5) 2.2 -5) 0.21-4.18(-5) ) 4(-1)) lo6Ru 1.1(-5) 0.7-3.7(-5) 3.0 - <0.02-5.7(-5) 4(-1 110 mag 1.5(-4) 0.024-<5.8(-4) 1.7 - 0.46-6.0(-5) 9(0)) 124Sb 2.4 - 0.081-6.2(-4) 2.2 - 0.099-5.5(-4 1.1(0) 12sSb 3.1 - 0.014-1.06(-3) 2.7 - 0.154-7.39(-) 1.1(0) luoLa 1.2 - 0.026-3.77(-4) 2.9 -5 0.035-1.10(- ) 4.2(0) 141Ce 2.9 - <0.62-7.4(-6) 1.4 -6) 0.39-1.7(-6) 2.1(0) 144Ce 2(- ) *** 4.9 -6) <0.024-1.3(-5) 4(-1)
*** Radionuclide detected in one measurement only.
TABLE 4.7(cont'd) BORIC ACID RECOVERY SYSTEM WITH BASE CATION DEMIN. C MEAN RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's Boric Acid Evaporator Nuclide Feed Concentration (uci/ml) Distillate Concentration (uCi/ml) "Best Value" Mean Range Mean Rance DF 1311 2,1(.3) 0.136-5.2(-3) 4.2(-5) 0.075-1.64(-4) 5.0(1) 134Cs <1(-6) 137Cs 9.6(-8) 0.057-2.0(-7) <1(1)
<5.2(-6) 4.7(-8) 1.5-6.7(-8) <1.1(2) s1Cr 2.7(-4) 54Mn 0.30-5.0(-4) 2.2(-7) <0.037-2.8(- ) 1.2(3 2.7(-5) 0.039-1.13(-4) 1.3(-7) 0.114-5.9(-7 2.0(2 59Fe 6.2(-5) d 57Co 0.025-1.44(-4) 2.5(-7) 0.046-1.26(- ) 2.5(2
- 3.1(-6) 0.39-7.8(-6) <8.4(-8) >3.6(1)
SECo 8.3(-4) 0.125-2.5(-3) 1.1 (-6) 0.083-3.86(-6) 60Co 2.4(-4) 7.8(2) 6sZn 0.476-7.3(-4 7.9(-7) 0.12-2.9(-6) 3.1(2) ssZr 6.2(-6) 0.09-1.7(-5)) <1.1(-7) >5.5(1) 3.3(-5) 0.54-6 9(-5) 9.4(-8) 0.07-3(-7) 3.5(2 9sNb 5.8(-5) 0.14-1.23(-4) 1.15(-7) 0.21-2.6(-7) 5.1(2 99Mo 7.4 - 0.0067-2.12(-4) 3.7 -7) 103Ru 2.2 - 0.0025-2.1(-6) 2.0(2 0.21-4.1 3.9 - 0.76-6.4(-8) 5.6 106Ru 3.0 - <0.02- 5.78 - }
, 3.8 - <0.0017-1.2(-6) 8.1 110 mag 1.7 -5) 0.46-6.0( 5 2.4 - 0.42-6(-7) 7.2 124Sb 2.2 - 0.099-5.5(-4) 2.8 -
12sSb 2.7 - 0.02-1.3(-6) 7.8 0.154-7.39(-4) 3.9 - 0.069-1.4(-6) 6.9 4 140La 2.9 - 0.035-1.10(-4) 4.3 - 6.7 141Ce 1. 3 - <0.10-9.6(-8)
- 0. 39-1.7(-6) <1.1(-7) >1.2(1) 144Ce 4.9(-6 <0.024-1.3(-5) 2.0(-7) <0.064-7(-7) 2.4(1) l
_ = _ - - . _ _ - . - _ .. _ - _ - _ _ _ - i t* TABLE 4.7(cont'd) BORIC ACID RECOVERY SYSTEM WITH BASE CATION DEMIN. C MEAN RADIONUCLIDE CONCENTRATIONS A;D " PEST VALUE" DF's Condensate Demineralizer 4 Inlet Concentration (uCi/ml) Outlet Concentration (uCi/ril) "Best Value" Nuclide Mean Range Mean Range DF 1311 4.2(-5) 0.075-1.64(-4) 4.2(-6) 0.021-1.40(-5) 1.0(1) 134Cs 9.3(-8) 0.057-2.0(-7) 1.0(-7) 0.30-2.1(-7) 9(-1) 137Cs 4.7(-8) 1.5-6.7(-8) 2.2(-7) 0.50-4.9(-7) 2(-1) 51Cr 2.2(-7) <0.037-2.8(-7 2.2(-7) 0.40-6(-7) 1.0(0) 54Mn 1.3(-7) 0.114-5.9(-7)) 6.8(-8) <0.12-1.38(-7) 1.9(0) 59Fe 2.5(-7) 0.046-1.26(-6) 4.3(-8) 0.11-1.0(-7) 5.7(0
% 57C0 <8.4(-8) <5.8(-8) seCo 1.1(-6 0.083-3.86(-6) 1.2(-6 0.35-4.8(-6) 9(-
60Co 7,9( 7 p,12 2,9(.6) 2.9(-7 0.88-9.9(-7) 2.7(0 ssZr 9.4(-8 0.07-3(-7) 4.1(-8 0.5-5.5(-8) 2.3(0 95Nb 1.1 - 0.21 -2.6(-7) 2.9(-8) <0.11-<1.1(-7) 4.0 ! 99Mo 3.7 - 0.0025-2.1(-6) 2.8 -8) 0.39-4.8(-8 1.3 103Ru 3.9 - 0.76-6.4(-8) 3.6 - 0.38-9.0(-8 1.1 2 106Ru 3.8 - <0.0017-1.2(-6) 1.4 - 0.19-3.8(-7 2.7 110 mag 2.4 - 0.42-6(-7) 2.5 - 0.21-5.4(-7 1.0 124Sb 2.8 - 0.02-1.3(-6) 4.1(-8 1.7-8.6(-8) 7.0 . 12sSb 3.9 - 0.069-1.4(-6) 4.3(-8) 0.59-6.9(-8) 9.1 l 14cLa 4.3 -8) <0.10-9.6(-8) 5.7(-8) 0.021-<5.2(-7) 7.5(-1) l
) 141Ce <1.1(-7) <4.9(-8) j 144Ce 2.0(-7) <0.064-7(-7) <4.2(-8) >4.9(0)
I
i TABLE 4.7 (cont'd) BORIC ACID REC 0VERY SYSTEM WITH BASE CATION DEMIN. C
!!EAN RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's Condensate Demineralizer Filter Inlet Concentration (uCi/ml) Outlet Concentration (uCi/ml) "Best Value" Nuclide Nean Range Mean Range DF 1311 4.2(-6) 0.021-1.40(-5) 4.0(-6) 0.0216-1.28(-5) 1.0(0) 134Cs 1.0(-7) 0.30-2.1(-7) 5.8(-8) 2.2-8.7(-8) 1.8(0) 137Cs 2.2(-7) 0.50-4.9(-7) 1.6 - 0.61-2.9(-7) 1.4(0) 51Cr 2.2(-7) 0.40-6(-7) 5.7 - 0.059-<3.9(-7) 3.9(0) 54Mn 6.8(-8) <0.12-1.38(-7) 3.3 - 0.47-5.3(-8) 2.0(0) a 59Fe 4.3(-8) 0.11-1.0(-7) 5.5 - 0.027-1.7(-7) 8(-1) 3 5700 <5.8(-8) 2.2 - 0.088-3(-8) <2.6(0) 58C0 1.2 -6) 0.35-4.8(-6) 1.3 - 0.099-5.44(-6) 9(-1) 60Co 2.9 - 0.88-9.9(-7) 5.8 - 0.063-1.8(-6) 5(-1) 652r 4.1 - 0.5-5.5(-8) 2.5 - <0.25-<9.8(-8} 1.6(0 ssNb 2.9 - <0.11-<1.1(-7) 2.7 - 0.041-<1.1(-i, 1.1(0 99Mo 2.8 - 0.39-4.8(-8) <4.1(-8) >7(-1 103Ru 3.6(-8) 0.38-9.0(-8) 2.9(-8) 0.033-1.11(-7) 1.2(
106Ru 1.4(-7 0.19-3.8(-7) <7.6(-7) >2(-
- 110 mag 2.5(-7 0.21-5.4(-7) 2.6 - 0.45-5.2(-7) 1.0(
124Sb 4.1(-8 1.7-8.6(.8) 2.3 - 0.31-<8.2(-8) 1.8(0) 12sSb 4.3(-8 0.59-6.9(-8) 3.6 - 0.085-<1.2(-7) 1.2(0) 140La 5.7(-8) 0.C21-<5 2(-7) <4.0-) 41.4(0) 141Ce <4.9 - <4.7 - )
<4.2 - ***
144Ce 5.5( 8 <7(-1)
*** Radionuclide detected in one reasurement only, i
l 1
TABLE 4.8 BORIC ACID RECOVERY SYSTEM WITH BASE CATION DEMIN. A MEAN RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's Base Cation Demin. A (Mixed-Bed Resin) Inlet Concentration (uCi/ml) Outlet Concentration (pCi/ml) "Best Value" Nuclide Mean Range Mean Range DF 131I 7.0(-5) 0.21-1.85(-4) 2.9(-6) 0.70-6.9(-6) 2.4(1) 134Cs 1.2(-4) 0.297-2.26(-4) 1.1(-6) 0.85-1.8(-6) 1.0(2) 137Cs 1.8(-4) 0.42-3.59(-4) 3.4(-6) 2.3-5.7(-6) 5.4(1) 51Cr 2.2(-5 1.0-3.4(-5) 1.3-5) 0.42-2.5(-5) 1.7(0) 54Mn 1.2(-5 0.68-1.95(-5) 3.8 -6) 1.7-7.2(-6) 3.0(0) _. 59Fe 2.8(-6 0.77-3.7(-6) 1.5-6) 0.56-2.5(-6) 1.9(0) g 57Co 6.2(-7 0.25-1.0(-6) 3.a -7) 1.4-6.2(-7) 1.8(0) seCo 1.7(-4) 0.427-2.51(-4) 5.8(-5) 0.20-1.0(-4) 2.9(0) 60Co 1.1(-4) 0.50-2.03(-4) 3.7(-5) 2.28-7.7(-5) 2.9(0) 65Zn 6(-7) 5.3-7) 3.7-8.1(-7) 1.00) 95Zr 5.8(-6) 4.1-8.6(-6) 2.4 -6) 1.6-3.3(-6) 2.4 0) 95Nb 9.8 -6) 0.66-1.63(-5) 4.5 -6) 3.1-6.1 (- ) 2.2 0 99Mo 7.6-7) <0.33-2.7(-6) 1.6(-7) <0.63-4.9 -7 4.7 103Ru 1.6 -6) 0.8-2.6(-6) 7.8 -7) 0.65-1.09 -6 2.0 106Ru 3.9 -6 0.092-1.4(-5) 3.2-6) <0.61-6.1(-6 1.2 11amAg 4.3 - 1.8-6.4(-6) 1.7 - 1.2-2.27(-6) 2.6 0) 124Sb 7.6 - 0.15-1.2 -5) 1. 2 - 0.4-1.9(-6) 6.4 125Sb 1.5 - 0.46-3.3 -5) 1. 7 - 0.8-3.6(-6) 8.5 140La 2. 8 - 6 0.18-9.2 -6) 1.1 - 0.73-2.26(-6) 2.5 , 141Ce 4.4 -7 <0.36-1.0(-6) 3.3 - 0.11-1.2(-6) 1.3 144Ce 2.0 -6) <0.83-4.6(-6) 2.0 -6) <0.37-4.3(-6) 1.0 i
*** Radionuclide detected in one measurement only.
I
TABLE 4.8 (cont'd) BORIC ACID RECOVERY SYSTEM WITH BASE CATION DEMIN. A MEAN RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's Boric Acid Evaporator Nuclide u Feed Concentration (uCi/ml) Distillate Concentration (sci /ml) "Best Value" Mean Ranoe Mean Range DF 131I 2.9(-6) 0.70-6.9(-6) 2.9(- 6) 0.26-9.0(-6) 1.0(0) 134Cs 1.l(-6) 0.85-1.8(-6) 7.4(-8) 0.047-1.4(-7) 1.5(1) 137Cs 3.4(-6) 2. 3-5. 7(- 6) 1.1(-7) 0.040-3.1(-7) 3.0(1) 51Cr 1.3(-5 0.42-2.5(-5) 54Mn 3.8 - 1.5-7) 0.54-4.3(-7) 8.7(1 1.7-7.2(-6) 59Fe 1. 5 - 1.2 -7) 0.37-2.2(-7) 3.1 1 0.56-2.5(-6) 5.6 -8) 0.17-<1.7(-7) 2.71 3 57Co seCo 3.4 - 1.4-6.2(-7) 2.8 - 0.019-<1.3(-7) 1.2 60Co 5.8 - 3.7 - 0.20-1.0-) 2.4 - 1.4-3.8(-7) 2.4 2.28-7.7 ) 3.9 - 1.2-8.2(-7) 9.6 ssZn 5.3 - 3.7-8.1( <1.8(-7) >2.9 ) 95Zr 2.4 - 1.6-3.3(- 3.8(-8) 2.1-7.5(-8) 95Nb 4.5 - 6.3(1) 99Mo 3.1 -6.1 (- 6 5.7(-8) 0.33-1.0(-7) 7.9(1) 1.6(-7) <0.63-4.9(-7) 3.8(-8) 0.29-5.2(-8) 4.2(0) 103Ru 7.8(-7) 0.65-1.09(-6) losRu 2.8(-8) 0.069-<1.1(-7) 2.8(1) 3.2(-6) <0. 61 -6.l ( -6) 2.5(-7) 0.57-<8.3(-7) 11 crag 1.7(-6) 1.3(1) 1.2-2.27(-6) 2.4 -7) 0.39-3.0(-7) 7.0(0) 124Sb 1.2(-6) 0.4-1.9(-6) 2.6 - 12sSb 1.2-3.7(-8) 4.5 1.7(-6) 0.8-3.6(-6) 6.3 - 4.5-7.7(-8) 2.8 14Cla 1.1(-6 0.73-2.26(-6) 4.2 - 0.037-<1.3(-7) 2.7 141Ce 3.3(-7 0.11-1.2(-6) 4.8(-8 0.079-<1.4(-7) 6.9 144Ce 2.0(-6 <0.37-4.3(-6) 1.6(-7 1.1-<6.3(-7) 1.2
- _ _ - _ _ _ _ _ _ - - - - - - - - - - - - - - - -- .n-. ---- + - - - - - - - - -
TABLE 4.R (cont'd) BORIC ACID RECOVERY SYSTEM WITH BASE CATION DEMIN. A MEAN RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's Condensate Demineralizer Inlet Concentration (vCi/ml) Outlet Concentration (pCi/ml) "Best Value" Nuclide Mean Range Mean Range JF 131I 2.9(-6) 0.26-9.0(-6) 2.3(-7) 0.085-6.7(-7) 1.3(1) 134Cs 7.4(-8) 0.047-1.4(-7) 6.2(-8) 0.059-1.1(-7) 1.2(0) 137Cs 1.l(-7) 0.040-3.1(-7) 9.8(-8) 0.32-2.2(-7) 1.2(0) , s1Cr 1.5(-7) 0.54-4.3(-7) 1.4(-7 <0.39-<6.9(-7) 1.0 54Mn 1.2(- 0.37-2.2(-7) 3.2(-8 0.54-6.7(-8) 3.9 _,, 59Fe 5.6 - 0.17-<1.7(-7) 5.0(-8 0.14-<2.6(-7) 1.1 g 57Co 2.8 - 0.019-<1.3(-7) <5.2(-8) 0.61-8,5-7)
>5(-1 seCo 2.4 - 1.4-3.8(-7) 2.6 -7) 9(-1) 60Co 3.9 - 1.2-8.2(-7) 1.5 -7) 0.58-4.8 -7) 2.6
, 85Zr 3.8 - 2.1-7.5(-8) 3.6 - 0.43-4.6 - 1.1 95Nb 5.7 - 0.33-1.0(-7) 5.0 - 0.76-8.3 - 1.1 99Mo 3.8 - 0.29-5.2(-8) 1.7 - 0.022-<1 - 2.2(0 103Ru 2.8 - 0.069-<1.1(-7) 1.8 - <0.49-<6.9 8) 1.5(0 i lo6Ru 2.5 - 0.57-<8.3( 7) <3.3(-) >8(-1 11orAg , 2.4 - 0.39-3.0(- ) 7.7(-8 0.18-2.0 -7) 3.2(0 124Sb 2.6 -8 1.2-3.7(-8 1.7(- 0.64-4.0-8) 1.6(0 12sSb 6.3 -8 4.5-7.7(-8h 9.6(- 0.27-1.8 -7) 7(-1) ! 140La 4.2 -8 0.037-<1.3(-7) <5.3(-8) i 141Ce 4.8 -8 0.079-<1.4(-7) <9.0(-8) >8(-1)
>5.3(-1 )
144Ce 1.6 -7 1.1-<6.3(-7) 1.1(-7) <0.031-<5.9(-7) 1.4(0) i
TABLE 4.8 (cont'd) BORIC ACID RECOVERY SYSTEM WITH BASE CATION DEMIN. A MEAN RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's Condensate Demineralizer Filter Inlet Concentration (pCi/ml) Outlet Concentration (vCi/ml) "Best Value" Nuclide Mean Range Mean Range DF 131I 2.3(-7) 0.085-6.7(-7) 2.1(-7) 0.18-7.1(-7) 1.1(0) 134Cs -6.2(-8) 0.059-1.1(-7) 7.3(-8) 0.46-1.0(-7) 9(-1) 137Cs 9.8(-8) 0.32-2.2(-7) 4.1(-8) 0.33-9.3(-8) 2.4(0) 51Cr 1.4(.7) <0.039-<6.9(-7) 8.7(-8) <0.043-<4.2(-7) 1.6(0) 54Mn 0.54-6.7(-8) 3.1(-8) <0.4-7.8(-8) 1.0(0) m 59Fe 3.2(-d}- 5.0(-8 0.14-<2.6(-7) <8.2(-8) >6(-1) g 57C0 <5.2(-8) <4.9(-8) 58C0 2.6 - 0.61-8.5 - 9.3 - 0.48-1.4(-7) 2.8
- 60C0 1. 5 - 0.58-4.8 - 4.9 - 2.3-9.8(-8) 3.1 95Zr 3.6 - 0.43-4.6 - 3.0 - 0.057-<l.2(-7) 1.2 95Nb 5.0-) 0.76-8.3(-8) 2.7(-8) 0.044-1.0(-7) 1.9(0 99Mo 1.7 - ) 0.022-<1(-7) <5.7(- ) >3(-1 103Ru 1.8 - ) <0.49-<6.9(-8) 1.1(-8 0.29-<5.3(-8) 1.6(0 106Ru <3.3(-7) 1.6(-7 0.37-<6.7(-7) <2 .1 .)
11amAg 0.18-2.0(-7) 2.8-8) 1.5-4.1(-8), 2.7(0) 124Sb 7.7-8} 1,7 -8, 0.64-4.0(-8) 1.0 -8) 0.157-1,2(- ), 1.6(0) lasSb 9.6(-8) 0.27-1.8(-7) 3.8-8) 2.5-6.1(-8) 2.5(0) 140La <5.3(-8) 1.8 -8) <0.56-<7.6(-8) <2.9(0) 141Ce <9.0(-8) <6.8(-8) 144Ce 1.1(-7) <0.031-<5.9(-7) <3.0(-7) >4(-1) , d 1
h AVERAGE CONCENTRATION pCi/mi 131r 134Cs 137Cs seCo 60Co
+ sans 3.3 x 10- 3 9.0 x 10-4 1.5 x 10-3 2.0 x 10-2 4.6 x 10-3 l
L s..e 3L. Aca,s.4 v,;., A 0 c DF =,1.6 >900 >280 24 19 Y '{' l fation Resin i._, , , ,' ,, , 2.1 x 10-3 <l x 10-6 <5.2 x 10-6 8.3 x 10-4 2.4 x 10-4 l.. 7 r_,..
~ a n,,,m ..
l%# i Sco.s k.o M . DF = 50 --- --- 780 310 i_ $^~ 9 4.2 x 10-s 9.6 x 10-8 4.7 x 10-8 1.1 x 10-6 7.9 x 10-7
- v. . . - 3 _
s s~rs e A _ o 1 c ...,.. , . P us- ~4 DF = 10 0.9 0.2 0.9 2.7 - n e e..s> L.9'N + s4 mete 4.2 x 10-6 1.0 x 10-7 2.2 x.10-7 1.2 x 10-6 2.9 x 10-7 na- ,._L1_ DF = 1.0 1.8 1.4 0.9 0.5 [~~h] p] sa-m r.ir4 4.0 x 10-6 5.8 x 10-8 1.6 x 10-7 1.3 x 10-6 5.8 x 10-7 { 1 -- , -- v . I a i i i e FIGURE 4.7 ' BORIC ACID REC 0VERY SYSTEM AVERAGE CONCENTRATIONS AND "BEST VALUE" DF's FOR OPERATION WITH BASE CATION DDilMRALIZER C IN SERVICE 196-
ff! L AVERAGE CONCENTRATION pC1/ml 1311 134Cs 137Cs seCo 60C0
* - saa n e 7.0 x 10-5 1.2 x 10-4 1.8 x 1S-4 1.7 x 10-4 1.1 x 10-4 x .x x w..c.,,.., p.; .., DF = 24 100 54 2.9 2.9 A o e Mix g' y bid Rzsin , _ , , , ,
y",,as2.9 x 10-6 1.1 x 10-6 3.4 x 10-6 5.8 x 10-F 3.7 x 10-5 [ ~] r .,4.
~. i e , ,,. . .
la-# Soo.. % M CF = 1.0 15 30 240 96 p,uh 2.9 x 10-6 7.4 x 10-8 1.1 x 10 - 2.4 x 10-7 3.9 x 10-7 1..., _ s . ~ ,. i-o F cuer*d ' DF = 13 1.2 1.2 p.oi...u, 0.9 2.6 D G ste s as s_ s 6[J'.7. ; % s.orte 2.3 x 10-7 6.2 x 10-8 9.8 x ,10-e 2.6 x 10-7 1. 5 x 10-7 c..YQa .-- DF = 1.1 0.9 2.4 2.8 3.1 r{J r i r ... 2.1 x 10-7 7.3 x 10-8 4.1 x 10-e 9.3 x 10-8 4.9 x 10-e _O 7 3..m I C I. Y FIGURE 4.8 BORIC ACID REC 0VERY SYSTEM AVERAGE CONCENTRATIONS AND "BEST VALUE" DF's FOR OPERATION WITH BASE CATION DEMINERALIZER A IN SERVICE 127
l D ,
,u AVERAGE CONCENTRATION pC1/mi 1311 134Cs 137Cs seCo 60C0
--g- sano c I
LA s..,c.....,ov l d A o c Y E
-*- s . e rt s L-R~]r-- !
. 1 ,,,,,.... I
, h.. g 2.4 x 10-5 1.0 x 10-7 9 x 10-8 6.9 x 10-7 6.1 x 10-7 I l 4
%.s k.a M I
DF = 670 260 990 5070 2790 o_ +Eswwm Oet - g ,, ' OMS 4 1.6 x 10-2 2.6 x 10-5 8.9 x 10-s 3.5 x 10-3 1.7 x 10-3 A c ...,. . , . Pous .,6 D E M eaJ as.e ' be[WaE -M* S a s* P Lif (e.#. _j , l E I E'l I i r << i*' *'
. . _ , _v .
3 a , v l
?
i 1
=
4 FIGURE 4.9 BORIC ACID EVAPORATOR AVERAGE CONCENTRATIONS AND RATIOS OF BOTTOMS TO DISTILLATE ACTIVITIES l 12e l
I 1 , radionuclides. Figure 4.10 shows plots of the DF's for 1311 and 137Cs vs. inlet concentration on log-log graph paper. The functional fonn , of the expression can be approximated by b DF = a C where DF = decontamination factor i a = constant C = inlet activity concentratio;; (uCi/ml) I b = slope of line For this case the value of b is near 1 (one). Because of this apparent dependence of DF on inlet concentration, the "best value" DF's represent average DF's obtained within the observed range of inlet concentrations. They, therefore, are valid only in this range of inlet concentrations. The base cation demineralizer C when charged with a cation red n did not show this correlation, for two reasons. (1) Iodine was not taken out at all, and (2) Cs was taken out so well that al e**. all of the DF's measured were lower limit values. The evaporator data were examined in detail to see if the same type of correlation with feed and DF could be found. There were essentially two effects observed. The first was a correlation between the DF for the crud-associated radionuclides 57 soCo and the feed concentration to the evaporator. This correlation is shown in Figure 4.11. Data for the radwaste evaporator (see Section 4.3) and three other evaporator types at two other nuclear power plants (2,3) are included. The functional form of the expression is essentially the same as that found for the mixed-bed demineralizer with essentially the same slope (i.e., about 1). j Since DF is defined as feed concentration divided by distillate concen-tration, a slope of 1 implies that the distillate concentration is constant (i.e., independent of the feed concentration). Therefore,'the DF of the evaporator for particulate type activity will depend upon the feed l concentration. The second effect was a correlation between the bottoms concentration and the distillate concentrati(q of 1311. This correlation is shown in Figure 4.12 and includes data from the radwaste evapcrator (see Section 4.3). This correlation is important since it shows how iodine spiking in the coolant can be transmitted through the processing stream and radwaste system and affect the discharge to the environment. That is, the iodine concentration in the distillate will not increase until the bottoms concentration increases. Since feed concentrations are normally j orders of magnitude lower than bottoms concentrations, an iodine spike l will be attenuated and delayed in time by the evaporator. The non-l volatile radionuclides, i.e., the cesiums and the cobalts, did not show this correlation in the BAE. In the radwaste evaporator (see section 129
. . - _ . - = - . .. - .~ -. ... _ - .. .
Y j Figure A.10 i Correlation Between 131I DF and Inlet Concentration
. for Base Cation Demineralizer A l (Mixed-Bed Resin) 103 , , , ,,,,,l , , , , , ,
4
,,,,,l ,,,L 1
- _ \
~ -
BAE - Base cation demineralizer A 2 , N i
- .9 102 _ e 137Cs _
E a 131; C - l '5 a ,
~
g - - C - - O e g - _
- u ~
e ~ i
.o '
1 % - * -
! E * ! 8 .
- C i 8 101 - ' -
E : : 1 3o - 1 1 3 - _ i - - 3 a - I ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' 1 10-6 10-5 10-4 10-3 Inlet concentration (pCi/ml) INEL A-10 987 1 100
1 Figure 4.11 Correlation Between 58C0 and 60Co DF and Feed Concentration : for Boric Acid Evaporator ' i i 105- i i irisi i i i I s ili i i i i s ili i isiluj i i i i lii i [ Evaporator 60Co : j i e #3 BAE e #4 BAE ' s 104 :- o.a.o.g.a Other plants a 1
- e :
l ! _ d* _ di f 103 ---
$ 2 e a m -
g - s - e o . - g - g _ e l $o a t 2 10 e
- o e =
C * ; O
,E : I o
e e _ o e _m _ O _ ! c
- Sm 101 : :
.E : :
E - o o - 1 2 - -
. 8
- 8 - -
a 1 --
=
2 l 10-1 =- = l o : l 10-2 I I 111111! I I I 11111l l 1 111111! I I I f f!HI I I 11111 l 10-7 10-6 10 10-4 10-3 10-2 l Feed concentration (pCi/ml) INEL.A 10 976 131
i Figure 4.12
- Correlation Between 131I Concentration in BAE Distillate and Bottoms I
l 10-3
^
_ Evaporator 131; - e Boric acid evap - a Rad waste evap
- 10-4 --
= _
E 1 _ a a O ~ x a _ a g # - I 8 _
=
2 E 10-5 8
$ 3
- e 3 6 -
o -
; ~
e ~ Q _ 3 - 2 e 5 l
$10-6 _-- --
Z
- i
, - l
- j 1
10-7 I I I I f ill! I I IItill! I I IIfill! I I II1111I I I I II'll 10-5 10-4 10-3 10-2 10-1 1 131 Bottoms 1 concentration ( Ci/ml) INEL-A-10 977
)
l 132 I I
4.3.2) the reverse trend (i N concentrations of cesiums and cabalts in the distillate decreased as concentrations in the feed increased) was observed. 4.3 Discussion of Measurement Data - Liquid Radwaste System 4.3.1 Introduction During the measurement period at Turkey Point, two different systems were used to process liquid radwaste. Prior to April 1978 the radwaste evaporator was used exclusively. Beginning in April the new filter-demineralizer system underwent testing and the evaporator was used infrequently. Both radwaste systems were studied. Prime emphasis, j however, was placed on the radwaste evaporator system (including the
- condensate demineralizer) and 15 sample sets were obtained. Attempts were made to study the filter-demineralizer system being tested, but samples were obtainable only from the tank feeding this system and the monitor tanks at the outlet of the system. Therefore, only average DF's for batches could be obtained.
4.3.2 Radwaste Evaporator , Studies of the radwaste evaporator system spanned the period January-April,1978, with intensive measurements occurring in January and February. A summary of pertinent chemistry and operating , parameters for the sample sets is listed in Table 4.9 and is presented graphically in Figure 4.13. The operational parameters (i.e., temperature, vacuum) remained relatively constant for the measurements. Feed rate ranged from 8 to 14 gpm, the average being approximately 10 gpm. In general, the conductivity was in the range of 3 to 4 mho except during the period when the feed was shut off prior to dumping of 1he bottoms. Boron concentration in the bottoms ranged from about 7,>00 to 22,000 ppm, but was normally about 15,000 ppm. Boron concentiation was high on 2/1/78 and 2/5/78 because the samples were taken shortly before the bottoms were dumped. Forithe 4/28/78 sample set, the boron concentration was low because the evaporator had been in service a short time (less than 1/2 day) and was processing laundry wastes which would be expected to contain low amounts of boron. Samples of the radwaste evaporator feed, distillate, and bottoms were obtained during both normal operation and during one period when the feed had been shut off and the bottoms volume was being reduced in preparation for dumping. Measured radionuclide concentration in these samples can be found in Apnendix B, Tables B.21-B.24. The means and ranges for these data are given in Table 4.10. Examinatign of the ranges indicates that for all radionuclides detected, except 110 Ag, the range in distillate activities is much smaller (factors of 1.7 to 38) than the range in feed activities. Similarly in most cases the range in bottoms activities is also smaller (factors of 1.9 to ll) than the range in feed ' activities. The feed-to-distillate DF's (see Table 4.11) sho" these range differences, i.e., the DF's are not constant but exhibit variations of up to several orders of magnitude for some radionuclides. l ! 133
. _ . -. _ _ - . - . - - . _ _ _ - - _ = _ - - ~ _ - . _ _- _. -._ . . - .
1 T1hlt 4.9 RADWASTE EVP.PORATOR OPERATING PARAMETERS Feed Feed Waste Boron Tank Tank Concr..trator Vacuum Holdup Distillate Conc. Distillate Temp Level Temp in Evap Tank Level In Bottori.s Conductivity Date; Time 'F inches *F " Ha % inches PPM umhos gal / min 1/5/78; 13:15 160 20 180 25 45 60 10.600 3.7 10 1/6/78; 16:30 160 45 180 25 48 60 14.000 3.9 11 1/8/78; 09:46 160 45 1 80 24 48.5 60 11.300 2.3 11 1/8/78; 12:35 160 33 180 24 48.5 60 14.400 2.9 9 1/8/78; 13:20 160 32 180 24 49.5 60 14.400 3.1 9 l 1/8/78; 17:22 160 33 180 24 40 63 14.400 4.0 9 5d1/10/78; 09:13 180 33 1 80 25 32 62 14.700 4.0 10 ) FEED SHUT OFF 1/10/78; 10:05 160 30 180 25 NA 60 10.1 9 1/10/78; 10:10 160 15 180 24 NA 60 12.8 9 1/10/78; 10:15 160 10 180 24 NA 60 14.0 9 1/11/78; 12:45 160 35 180 25 35 60 14.461 8.0 9 2/1/78; 16:00 160 30 179 24 64 18.600 3.4 10 2/5/78; 10:00 32 22.400 4.8 8 23.4 25 58 14 2/18/78; 13:30 167 32 178 4/28/78; 10:44 155 33 190 20 60 7.110 3.5 12 i
, i i
Figure 4.13 Radwaste Evaporator Operating Parameters 105 i
- I 'I i is i 1 i l i l l l l
~
e/ - 104 - 2 -e e Boron concentration (ppm) a Concentrator temperature (* F) - a Feed tank temperature (*F)
- Feed tank level (in.) -
0 Vacuum (in. Hg) o Feed:<is. (gal / min) 103 T Conductivity (pmhos) in distillate T y _- _ g -
& I R-5 W8%! 3 R:~dI 102 _
e _ e
~
%-* *
- e_e 7 -o-o- " 3 o o _
s N/o - 0 U 101 -
- 0 % CD o- U U
_ b 1 i l 1 1/5 1/6 i I i in i I 1/9 1/10 i I i l l l l
)
1/11 1/12 2/1 2/5 2/18 4/28
- Year (1978) INEL-A-10 975 135 i l
TABLE 4.10 MEANS AND RANGES FOR RADIONUCLIDE CONCENTRATIONS IN RADWASTE EVAPORATOR FEED, DISTILLATE. AND BOTTOMS Feed Bottoms Distillate Mea n Range Mean Range Mean Fange Nuclide (uC1/ml) (uCi/ml) (uC1/ml) (uCi/ml) (uCi/ml) (uC1/ml) 131I 7.45(- ) 3.0-270(-5) 1.5(-2) 2.5-44(-3) 3.1(-5) 1.1-5.5(-5) 132I 1.0(-5
- 4.8-14( 6) * *
- 133I 4.9 -5 2.5-264 -6) 3.9(-4 4.9-190(-5). 1.2(-7) 8.6-240(-8) 134I 1.8-4) 5.6-550 -6) 4.4(-3 4.9-2200(-5) 1.4(-6) 4.9-29(-7) 135I 7.8 -5) 5.0-320 -6) 3.3(-4 5.3-110(-5) 2.3(-7)** **
134Cs 2.1 -4) 1.1-75(-5) 4.0(-3 1.9-12(-3) 2.3 - 9.6-61(-8) 13sCs 5.5 -6) 1.0-23(-6) 2.1 - 1.0-5.4(-4) 1.5 - 1.1-1.7(-7) 137Cs 4.0 M 2.4-160(-5) 7.5 - 3.8-24(-3) 3.5 - 6.2-97(-8)
- 13eCs 1 . 7 P ,) 5.3-820(-6) 1. 2 - 3.7-530(-5) *
- 24Na 5.1 - ) 1.7-8.6(- ) 3.2(-5) 1.2-4.9(-5) *
- g 51Cr 1.8 - 2.3-71(-5 9.9(-4) 3.4-33(-4) 2.0(-6) **
54Mn 9.6-li 1.1-47 -5 1.3(-3) 4.8-38(-4) 4.2(-7) 5.3-92(-8) 59Fe 2.5 - ) 2.8-79 -6 1.6(-4) 6-36(-5) 2.7(-7 ** ** 57Co 1.7(-5) 1.4-96 -6) 1.7 -4) 4.8-36(-5) 1.2(-7 4.3-17(-8) SBCo 6.7(-3) 4.1-400(-4) 6.9-2) 9.2-250(-3) 1.3(-5 1.0-32(-6) 60Co 1.7 -3) 1.8-93(-4) 1.8 -2) 5.9-66(-3) 3.9(-6 4.2-1 30(-7) 652n 3.0 -5 3.4-62(-6) 4.0(-4) 7.3-110(-3) 95Zr 2.9 -5 2.9-110(-6) 2.2(-4 9.4-43(-5) 2.4(-7) 4.2-49(-8) 95Nb 4.7(-5 2.2-150(-6) 1.9(-4 3.5-62(-5) 4.2(-7) 1.1-9.5(-7) 99Mo 1.5(-5 3.5-26(-6) 2.3(-4 A.7-4.2(-5) 103Ru 2.7-5) 1.7-120(-6) 1.4(-4 2.2-50(-5) 1.4(-7) 6.1-21(-8) 106RuD 5.1 -5) 6.2-170(-6) 2.4(-4 110 mag 8.7 -6 2.9-12(-6) 3.2(-7) 7.2-75(-8) 124Sb 1.7 - 1.0-92(-5) 1.9 - 2.3-77(-4) 1.7 - 4.0-26(-8) 125Sb 1.8 - 1.0-100(-5) 1.7 - 3.9-77(-4) 2.4 - 1.4-3.6(-7) 129mTe 1.6 - <0.11-4.8(-5) 1.5 - ** 3.6 - ** ** 140La 3.5(-6) 5.6-77(-7) 2.1(-5) 1.0-51 (-6) 1.1(-7)**
- Radionuclide not detected.
** Radionuclide detected in one sample only.
TABLE 4.11 i DECONTAMINATION FACTORS FOR RADWASTE EVAPORATOR i Nuclide 1/5/78; 13:15 1/6/78; 16:31 _1/9/78; 09:46 1/9/78; 12:30 1/9/78; 13:20 1/9/78; 17:15 131I 1.6 0.3 3.110.1 25 1 23 1 24 1 15 1 132I * * * *
>24
- 1331 *
>44 2.4 1.2(2) 1.5 0.6(3) >2.1(2) >1(2) 1341 431 *
- 319
- 135I >1.2(2) >65 >l50 >50 >83 134Cs 77 1 3 2.7 1.0(2) *
>350 >3.3(2)
- 136Cs 14 3 >25 >15 >20 * * '
137Cs 25 79 >445 1.5 0.9(3) 2.0 1.6(3) 1.1 13eCs * *
- 0.6(3) 1.1* 1 0.7(3)
>38
- 24Na >16 >23 >21 * >86
- e siCr >4.5(2) >3.7(2) * * >28
- g 54Mn 3.6 1 3.1(1) 4.5 0.1(1) 50 12 40 6 36 8 39 9 w 59Fe >l4 >13 * * *
- t 57C0 >27 5.912.8(1) >34 >30 *
^
seCo 5.0 0.3(1) 5.3 0.3(1) 48 2 59 3 48 11 59 2 GoCo 6.4 0.3(1) 56 4 74 6 74 6 58 t 3 61 6 65Zn >2.5(2) >2.8(2) >1.3(2) >1.2(2) >1.1(3)
>1.1(2) ssZr 3.1 1.0(1) >30 * *
>l.8
- 95Nb 5.1 1.1(1) >47 >16 >11 >3d
- 103Ru 55 24 >12 >13 *
- 103mRh
- 55 *
{ 24 >12 >13
- i * * *
- 106Ru *
- i * * * *
- 106Rh
- 11cmAg * * * * *
- 124Sb 93 19 >3.4(2) >3.0(2) >2.5(2) >2.8(2) >250 ;
12sSb 46 i* 17 >1.0(2) >260 >53 * ' { ' 129mTe >160 *
* >7 140La >11 >10 >20
- l i
1 . .I
i TABLE 4.11 (cont'd) DECONTAMINATION FACTORS FOR RADWASTE EVAPORATOR 1 Nuclide 1/10/78; 09:13 1/11/78; 12:43 2/1/78; 16:10 2/5/78; 10:40 2/18/78; 13:30 4/28/78; 10:44 131I 13 1 9.7 0.2 90 4 36 2 21.9 0.5 7.0 0.5 1321 * * * * *
>4.8(2) 133I >43 1.4 1.1(2) >1.7(2) >8.7(2) 1.2 0.1(2) 21 2 1 34I >3 *
>50 1.5 0.6(2) *
- 135I >17 * * *
- 1.5 0.1(2) 134Cs >4.4(2) 5.6 2.0(2) >3(3) >2.0(3) 1.1 0.2 1.1 t 0.1(2) 13ECs >15 >10 *
- 2.1 0.6 >11 137Cs <1(5) >700 >2.7(3)
- 5.9 0.4 2.1 1.8(3) ,
13eCs * * * **
>8 >18 3H 0.90 2 0.04 ** ** **
b$ ** ** 14C 1.3 0.2(1) ** ** ** 24Na >12 * >4 * * >78 51Cr >2.6(2) >200 > 2.4 (2) >1.8(3) 33 6 >a0
- 2.0 0.7(3) 54Mn 74 17 5.5 1 3.7(2) > 2.3( 3) 72 5 61 28 sspe 4.0 0.2(1) ** ** ** **
59Fe >1.3(2) 31 9 >14 30 *
- 57C0 9 > 3.2(2) 17 1 >28 seCo 71 3 2.3 0.3(2 4.0 2.4(4) 6.P 0.9(3) 27 1 3.7 0.4(2) 60Co 97 6 3.9 1 0.5(2 1.5**: 1.1(4) 8.6 1.1(2) 26 2 1.8 0.2(2) 63Ni ** 3.3 0.7(2 ** ** **
6sZn >123 * *
>1.7(2) >1.0(2) >1.3(2) 895r ** ** ** **
>2.8(2) 90Sr 5.6 ** ** ** **
1.0(1)
- 91Sr * >15 >6
- 91y ** 8.0 ** ** ** **
0.5(1) i 4
TABLE 4.11 (cont'd) DECONTAMINATION FACTORS FOR RADWASTE EVAPORATOR Nuclide 1/10/78; 09:13 1/11/78; 12:43 2/1/78; 16:10 2/5/78; 10:40 2/18/78; 13:30 4/28/78; 10:44 95Zr >80 >55 95Nb
>2.7(2) > 4.1 (2) 31 5 2.2 0.5(2)
>1.8(2) >105 >7.5(2) > 6.9(2) 27 4 1.4 lo3Ru >50 >28
- 0.4(2)
>600 38 8 >19 10 sinRh >50 >28 >600
- 38 8 >19 106Ru * *
* * >31 >13 losRh * *
* >31 >l 3 110 mag >l20 * *
- 1.5 124Sb 0.3(2)
>3.1(2) >170 >1.0(3)
- 1.6 0.3(2) 2.5 0.7(2) 12sSb 2.2(2)
>2.0(3)
- 1.6 0.5(2) 1.2 0.2(2) 129mTe * * * * *
, IsoBa * * * * *
* >25 8 140La .
>l5 *
>53 >41 141Ce * * *
- 7.6 4.1 *
- Radionuclide not detected in feed and/or distillate.
** Radionuclide not measured.
i , J , l 4 4 l The variations in DF's show no correlation with either operating or chemistry parameters. There is, however, a correlation between DF and feed concentration. Figures 4.14 to 4.16 show plots of DF and feed, distillate, and bottoms activities for 1311, 54Mn, and 60Co, respectively. These plots show a definite correlation between DF and i feed activity (i.e., the DF tends to increase as the feed concentration increases and to decrease as the feed concentration decreases), although there is insufficient data to uniquely define the correlation function. ' Cesium-137 also exhibits a correlation between DF and feed activity. l l Because the bottoms is the liquid being evaporated in this type of l evaporator, the feed activity will not immediately affect the bottoms j I concentrations. The feed concentration levels for 1311 and 137Cs generally i are 10 to 100 times lower than the bottoms, and if a spike in the feed 1 occurs, the effect on the bottoms does not show up immediately, so that i the distillate does not change. This tends to attenuate or stop the i
! propagation of a spike through the evaporator to the distillate. On i l 1/9/78 and 2/1/78 spikes occurred in the facd concentrations but did not , show up in the distillate. For example, from 1/6/78 to 1/9/78 1811 and i 137Cs feed Concentrations rose by factors of 7 and 1.5, but the bottoms showed no change for 1311 and a decrease in Cs concentrations. The distillate showed essentially no change. From 1/11/78 to 2/1/78 the 1311 l concentrations in the feed rose by a factor of 25, the bottoms by 4 and '
the distillate 2.5.
; A comparison of distillate to bottoms activities indicates that the distillate actually can become better as the bottoms concentration i increases. Figure 4.17 shows plots of the distillate activity as a function
{ of the bottoms concentration and of the feed concentration for 60Co. The j data indicate that the distillate concentration tends to decrease as
! the bottoms or feed concentration increases. An explanation for this l l behavior is not readily apparent. l 4
Figures 4.18 to 4.20 show plots of the ratio of the bottoms-to-distillate activity vs. bottoms activity for 131 , s4Mn, 1 and 60Co. ! These plots indicate that the bottoms-to-distillate ratio increases with bottoms activity, i.e., the distillate activity remains the same or even may decrease with increasing bottoms activity. A similar plot for 137Cs i provided inconclusive results due to the small range in bottoms activities and large uncertainties in the measured distillate activities. Table 4.12 lists "best value" DF's for the radwaste evaporator. !g These DF's compare favorably with the corresponding "best value" DF's i for the boric acid evaporator (see Tables 4.7 and 4.8). For most ! radionuclides, the radwaste evaporator DF f alls between the two values measured for the boric acid evaporator. This occurs because the feed concentrations fall between those for the boric acid evaporator. One sample of radwaste evaporator bottoms, obtained on 1/11/78, was analyzed for alpha-emitting 38 radionuclides. Resulting concentgions ,240Pu. were 3.2 + 0.2(-7) pCi/ml for Pu and 1.9 + 0.2(-7) pC1/ml for 140
--. - . . _ = . _ _
1 a Figure 4.14 131I Concentrations and DF's for Radwaste Evaporator I 106 l l l I I I I i - j Radwaste evaporator 131 1 activities [ : _ e Feed activity _- _ o Bottoms activity _ 1 l o Distilate activity I 10-1 - e Decontamination factors - 105 q 10-2 _ _ 1045 k
' a - c
= _ _ g e -
.2 o E D - -
2 2 o 8 o E; 10-3 _. __ 103g u _ es _ c g g _- s _ 4 s - _ s _ ', _ %s _
-- N f
g
~
4 s i Ke \ i
-g., ' ' ' ' - s 10-4 --
~ '
[ l- y. 10 2 i _
-l 1
( 1 I 4 _ ,s** _ I I I I I I I I 1 10-5 10 1/5 1/6 1/9 1/10 1/11 1/12 2/1 2/5 2/184/28 ! Date of sample (1978) iNEL-A-10 768 l l \ 141 l
.. _- .- . . _ ~ -. _. . . . _ _ _ _- _-
i 4 Fiqure 4.15 l 5Mn Concentrations and DF's for Radwaste Evaporator I l l l l l 8 l l - g - I Radwaste evaporator 54Mn activities l _ i O O !' o 4 10-3 _. -_ 10
> Z C i
- 3 .- r 9
b 'l - e Feed activity - T - 0 Bottoms activity i o Distillate activity . 10-4 t - 103 y g e Decontamination factors -- g I : S ' d ' ( 0
- ) > _ c
_ / g S
/ o c -
.a \ .E C / 1 E u -
/ t *
/ ' 8 E / t a a
! E 10-5 __ _ ,/ 5 102y u -
c ~ <( 8 -
.AL
-i
,/x -16 gr- -- ,tf k
4 .. 10-6 _. i-
} -
~
101
~
41 ~
~_ .. l> > -. -
' ' ' -- ' ' ' ' ' 1 10-7 1/5. 1/6 1/9 1/10 1/11 1/12 2/1 2/52/184/28 Date c1 sample (1978) INEL-A-10 767 1 142
i Firiure 4.16 60C0 Concentrations and DF's for Radwaste Evaporator i 6 10 t i i i i i i i i _ 10 j _ 1 j - l 8 O 10-2 __ __ 105 3 A ~
~
j _ _ I i _ Radwaste evaporator 60Co activities [ i e Feed activity 4 i 10-3 _ jo4 0 Bottoms activity I \j _ i E t
; o Distillate activity i I
i - a Decontamination factors -. g g l~ s 2 ' 8 - I _ _8 3 O c
- t a e 10-4 -~
i 10 3 j.@
! 5! I : E i _ m
, 2 -
i - 8
\
1 g u _w _ f
, i g
_ a c i LJ
/
/ g k I
! / lE p i l 10-5 _ _ j' _,_ 302 b*~~'
- C J' i
,b ij-
~
1 10-6 __ __ 101 i _ k l I 10-7 8 ' ' ' ' ' ' ' 1 1/5 1/6 1/9 1/10 1/11 1/12 2/1 2/5 2/184/28 Date of sample (1978) INEL- A-10 766 , I 'i 143
Figure 4.17 Correlation Between 60C0 Concentration in Distillate and Bottoms, Radwaste Evaporator l l 10-4 i i i i i ,,,,ig i i i i s i i_ i i ,,is; i 3 t [ Radwaste evaporator 60C o _ t
- ~
a Bottoms a Bottoms (Feed Shut Off) -
- Feed j _ _
e 10-5 1
- E -
, 1 -
- o .. -
l 3 - ee e & a - , g e a z - e . 2 i e $a y _ _ g a ! c o e U 10-6 - , a ei _ e _
= _ _
p _ e a _ i z - - ! .:e _ e a _ o l
' ' ' ' 'l ' ' ' i ' ' ' ' ' ' ' '
10-7 10-4 10-3 10-2 30-1 ) INEL- A-10 981 Bottoms Concentration (pCi/ml) r 3 I 144
i. i i Figure 4.18 4 I Correlation Between 131I Concentration in Distillate and Bottoms. Radwaste Evaporator l 5 I 10 4 i i : i i i i i i i_ 4 i i i iiil _ Radwaste evaporator 131 i l _ _ i l C o
.g _ _
l M
' 2;
- C 8 _ _
C O O e G l 3 .* ! 5103 u - o - e , _ 1 E ~ ~ O g _ e e _ 9 s em 1 j e _
= _
1 ,i ! e 102 I I i i ilil I i e i i i i I 10-3 10-2 10-1 ( { Bottoms concentration (pCi/ml) INEt.-A-10 984 l l i l 145 l l
- ,, , - _ . . . -., _...._ .- , ,_,. , r - ---.-,s , ,, , _, ,. ---,_._,-,,mm..,,
ll Figure 4.19 i Correlation Betwaen 54Mn Concentration in Distillate and Bottoms, Radwaste Evaporator j 4 1 I i l l i
! 105 1 I I i iI I Il I I i I i i l i i_
1 t
~
_ Radwaste evaporator 54Mn
<> I C
.e - I 3 at E
~
8 -
~
C O 4> 2m
= n M 104 -
il u - 2 - a, i. _ E o t> , 1
= -
? ~
6 -
= -
e c - Il i ' d - j '. , b j t> I I I i J tII iil 103 i i i , , 10-4 10-3 10-2 Bottoms concentration (y Ci/ml) INEL- A-10 983 i I t I 146
r 4 i Figure 4.20 Correlation Between 60Co Concentration in Distillate and Bottoms. Radwaste Evaporator i 106 , , , , , , , , , , ,_ _ , , , , ,l { _ - ) I _ Radwaste evaporator 60Co i j E los ___
=
g _ _ C - - y -
-o
. 8 - - O y _ _ t 5 E i S O e - - E < O ' t
'. =
O i 9
.9
- j10 4 --
0 ! e i
** e l
- e -
i i i e , e I I I f I I I II
- I I I I I i l i 10 3 10-3 10-2 10-1 Bottoms concentration ( Ci/ml) INEL-A-10 982 f'
- r. - . . . . - --a w - - . _ . _,._.--m. .m.m,.-. -
- _ - - .~ .- . _ - .. - -- . ._ .. . - . . - - _ . . . - - -
L 2 i 1 TABLE 4.12 l "BEST VALUE" DF's FOR RADtlASTE EVAPORATOR
]
1 i i "Best Value"
! , Nuclide DF 131I 2.4(1) i 133I 4.l(. 2) 134Cs 9.1 136Cs 3.7 137CS 1,)
14C 1.3 I 54Mn 2.3 SsFe 4.0 57CO l .4 58C0 5.1 60Co 4,4 s3Ni 3.3 90Sr 5.6 91Y 8.0 95Zp 1,2 95Nb 1.1 103Ru 1.9 110PAq' 2.7 124Sb 1.0 a 12sSb 7.5 s T i
?
i 1 a e 148 l
i 4.3.3 Condensate Demineralizer i Five sets of measurements of the DF across the radwaste evaporator condensate demineralizer were made. Measurement data can be found in Appendix Table B.25. Table 4.13 lists the resulting DF's. Since very little activity was detected in the inlet and outlet for-this i demineralizer (even using resin concentration techniques), only a limited j number of DF's were obtained, and these DF's showed considerable variation. 3 For example, the DF for 1311 ranged from 2.5 to 17 for essentially the same inlet concentration and the DF for 58C0 ranged from 18 to 1100 while the inlet concentration varied by about a factor of 23. Cesium j concentrations in the outlet were approximately constant even though the inlet concentration changed and cesium DF's were less than 1.0. This 4 indicates that the cesium concentrations in the inlet had probably been higher in the past. i Because of the sparse number of DF values, no information concerning ;
- the relationship between DF and inlet concentration can be inferred.
Table 4.14 lists means and ranges for radionuclide concentration and "best value" DF's for the radwaste evaporator condensate demineralizer.
- 4.3.4 Test Demineralizer i ,
; Although actual samples of feed to and outlet from the test 4
demineralizer train could not be obtained, an average batch DF could be
- obtained by sampling the tank (waste holdup tank #2) feeding this system and the monitor tank (monitor tank A, B, or C in the radwaste building) that the test system was feeding. This was possible because the test i demineralizer was used to process batches of liquid radwaste. That is, 1 a batch of liquid was transferred from waste holdup tank #1 to waste j holdup tank #2 (10,000 gal.). This batch of liquid was then processed i through the test demineralizer and deposited into one of the 5000 gal.
l monitor tanks (that was initially empty at the start of the run). When this tank was filled, the outlet flow from the test demineralizer system was diverted to another (empty) monitor tank. Therefore, by sampling waste holdup tank #2 while a specific monitor tank was being filled and then sampling this monitor tank after it was full, an average
- DF could be obtained. Data for these samples appear in Appendix B, Tables B.26 to B.28. i
; i Table 4.15 lists the measured batch DF's for the test demineralizer ,
system (demineralizer and filter). Only nuclides that were consistently 4 observed in the monitor tanks are listed. A comparison of the DF's l l for this test demineralizer with those for the radwaste evaporator l system (evaporator and condensate demineralizer) indicates that DF's for iodine and cesium are higher (factors of about 2 to 20) for the test demineralizer, and DF's for cobalt, manganese, antimony, and silver are higher (factors of about 5 to 200) for the evaporator system. Since
'samoling of the test demineralizer system was not ideal (i.e., only the tanks could be sampled, not the actual inlet and outlet) and no information concerning the test demineralizer resin was available, no further inter-pretation of the data could be made.
l l L 149
TABLE 4.13 DF's FOR RADWASTE EVAPORATOR CONDENSATE DEMINERALIZER Nuclide 1/6/78 1/10/78 1/11/78 2/1/78 4/28/78 131I 6.1 1 0.3(0) 3.7 0.2(0) 4.8 0.2(0) 1.7 0.1(1) 2.5 0.1(0)
- 2.5 1331 * *
>4(-1) 0.3(0) 134Cs 2 1 1(-1) <4(-1 ) 2.1 0.8(-1) <3(-1) 2.3 0.7(-1) 137Cs <2(-1) 6 6(-2) <2(-1) <2(-1) 1.1 0.9(-1) 3H ** 1.03 0.05(0) 14C ** ** 9.7 i 1.4(-1) 54Mn >3.5(0) >5(0) >2.6(-1) <2.9(1) 1.9 1.0(0) 55Fe <1.6(-2) g 57Co >5(-1) >2.1(0) o seCo 3.6 1 2.1(2) 1.1 1.1(3) >2.4(1) 1.8 1.1(1 ) 1.9 0.3(1) 60Co >4.9(0) >5.2(1) >4.9(0) 1.2 1 0.9(1) 2.3 0.5(1) 63Ni <1.4(- 1) ** **
89Sr ** ** >6(0) ** ** 90Sr *~- ** 2.3 0.5(0) **
** ** <4,4(o) **
91y
* * * * >4.2(0) 95Zr * *
- 103Ru >5(-1)
- 110 mag * * * >3.6(0)
* * *
- 3.3 1.6(0) 124Sb 125Sb * * *
- 2.8 2 0.4(0)
- Radionuclide not detected
** Radionuclide not measured
TABLE 4.14 EANS AND RANGES FOR RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's FOR RADWASTE EVAPORATOR CONDENSATE DEMINERALIZER Inlet Concentration (pCi/ml) Outlet Concentration (pCi/ml) "Best Value" Nuclide Mean Range Mean Range DF 131I 2.2(-5) 1.14-2.7(-5) 4.9(-6) 0.18-1.09(-5) 4.6(0) 134Cs 1.2(-7) 0.96-1.7(-7) 5.9(-7) 4.2-7.5(-7) 2(-1) 137Cs 9.7(-8) 0.62-<3(-7) 1.l(-6) 0.84-1.3(-6) 9(-2) 3H 3.2(-2) *** 3.1 (-2) *** 1.0(0) 14C 5.8 -7) *** 5.6(-) *** 1.0(0) S'Mn 2.7-7) 0.53-5.0-7) 7.9(- ) 0.068-1.9(-7) 3.5(0) 57Co 1.1 -7) S 58C0 0.43-1.7 -7) <8(-8 >1.3(0) 1.0 -5) 0.10-2.5-5) 5.8(-8) 0.23-<2(-7) 1.7(2) 60C0 2.8 -6 0.63-5.2 -6) 1.6(-7 0.053-<l(-6) 89Sr 6(-8) ) *** 1.8(1) 90Sr 2.5(-8) ***
<l(-8)) ***
>6(0) 91y <6(.g) *** 1.l(-8) ***
2.3(0)
] ,4(.8) *** <4,4 ssZr 4.2 - *** ***
110 mag 7.2 - *** <1(-8) >4.2
<2(-8) *** >3.6 124Sb 4.0 - ***
1.2 -E's *** 12sSb 1.4 - *** *** 3.3(0) 5.0 -d) 2.8(0) Results based on one measurement -
TABLE 4.15 DECONTAMINATION FACTORS FOR TEST DEMINERALIZER TRAIN Nuclide 4/25/78; 10:30 4/26/78; 13:30 4/29/78; 11:28 5/2/78; 14:14 5/10/78; 15:44 5/16/78: 17:55 131I 4 1 2(3) >1.4(3) >9.6(2) >7.7(3) >1.2(3) 1.3 0.2(3) 134Cs 8.3 1 3.8(2) 8.5 3.3(2) 2.8 0.6(2) >8.8(2) >1.6(3) >6.5(2) 137Cs 2.321.2(2)' 5.1 0.7(2) 3.1 0.3(2) 6.0 1.6(2) >2.8(3) >1.6(3) 54Mn 4.0 0.5(1) >6.4(1) 1.3 0.l(2) 6.2 0.8(1) 2.9 0.2(1) >3.6(2) 57Co 2.2i1.l(2) >9 1.5 1.2 0.7(2) >2.5 >1.4(1) 2.7 0.1(2)
>3.6(1) 4.4 0.3(2) 58C0 3.1 0.2(2) 1.5 0.4(3) 0.1(3) 4 1 60C0 2.0 0.2(2) 3.0 0.3(2) 1.2 0.1(2) 4.6 0.3(1) 3.7 0.4(2) 95Nb 4.1 1 0.8(1) >5.8(1) >3.92) >1.7(2 >6.6(1) >1.1(2) . 110 mag 6.2 2.3(1) >1.1 2) >6.6(1 2.4 1.2 1.1 0.2(1)
E" 124Sb >1.2(2) >4.7(1) >7.01) >7.0(1 3.6 0.9 >4.7(1) 12sSb >0.6 >3.9(1) 9.7 3.9(1) >1.5(2 2.8 1 0.3 4.8 1.6(1) Nuclide 5/20/78; 11:10 5/23/78; 13:55 _ 5/25/78; 08:50 134Cs 6.8 2.6(2) 1.0 0.4(3) >7.5(2) 137Cs 1.8 0.8(3) 2.1 1.4(3) >8.7(2) 54Mn 8.3 1 0.8 6.7 1.5(1) 9.2 1.6(1 seCo 6.3 0.8 2.8 0.5(2) 2.9 0.2(2 60Co 1.7 0.1 1.3 0.1(2) 2.6 0.7(2 124Sb >4.7(1) >5.0(1) 7.3 1.6(1) 12sSb 2.4 0.5(1) A.8 2.1(1) 7.7 1.5(1) 4
4.4 Radionuclide Concentrations in Tanks During measurements at Turkey Point, samples were obtained from the various holdup and monitor tanks. The ohiective in obtaining these samples was twofold: (1) to characterize feed streams to evaporators and demineralizers and (2) to characterize the radionuclide concentration in the tanks and obtain infonnation concerning radionuclide inventory in the p' nt. Appendix B, Tables B.14, B.15, B.21, P.26-B.30, contain the results from analysis of these samples. Note that the radionuclide concentrations in the feed to the boric acid recovery system and radwaste evaporator are characteristic of the concentrations in the holdup tanks and waste holdup tank No.1, respectively. Plots of the levels in the holdup, waste holdup tank #1, and auxiliary building monitor tanks are shown in Appendix B Figures B.17, B.18 and B.21, respectively. Information concerning the levels in waste holdup tank No. 2 and the radwaste building monitor tanks was unavailable. Means and ranges for radionuclide concentration in the holdup tanks, waste holdup tanks, and monitor tanks are listed in Tables 4.16-4.18. During the measurement period (1/11/77-6/1/78), total releases from the auxiliary building monitor tanks were 2.74(6) gallons. Of this total, 1.78(6) gallons were processed through the boric acid recovery system and 9.6(5) gallons through the radwaste evaporator system. In addition, 5.2(5) gallons of liquid processed through the test demineralizer were released from the rag;aste building monitor tanks. Using the mean radio-nuclide activities measured in the monitor tanks (see Tables 4.17 and'4.18) the extrapolated annual average releases from the monitor were obtained (Table 4.19). It must be noted that these annual averages were extrapolated from monitor tank concentrations obtained during a very short period (4/29-5/25/78) and may not be representative of the releases during the total 11/1/77-6/1/78 period. 4.5 Conclusions The following conclusions have been reached with regard to the operation of the two evaporator systems.
- 1. Spikes from the reactor coolant do propagate to the feed of
, the boric acid and radwaste evaporator systems. The holdup tanks reduce the effect by allowing the short-lived radionuclides to decay before entering the evaporator system. The base cation resin reduces the effect in the evaporator inlet by removing the cation radionuclides, such as cesium, and filtering out some of the crud-associated radionuclides, e.g., the cobalt. Iodine-131 on the other hand is not removed by the cation resin and enters the evaporator system essentially unchanged. In the radwaste evaporator only one tank collects waste and feeds the evaporator so the retention time is not as great. However, the spike levels are much diluted with the addition of " older" 153
TABLE 4.16 RADIONUCLIDE CONCENTRATIONS IN HOLDUP AND WASTE HOLDUP TANKS Holdup Tanks A, B and C i Waste Holdup Tank No.1** Waste Holdup Tank No. 2 iti . Mean Range Mean Range Mean Range Nuclide (uC1/ml) (uti/ml) (pCi/ml)_ (uCi/ml) . (pC1/ml) (pC1/ml) 131I 1.8(-3) 0.0002-1.00(-2) 7.4(-4) 0.08-2.7(-3) 1.9(-4 0.73-4.6(-4) , 132I
- 1.0 - 0.48-1.4(-5) 6.5 -6 5.2-8.l(-6) 133I
- 4.9 - 0.025-2.64(-4) 7.1 -5 0.124-1.45(-4)
- 1. 8 -
134I 0.056-5.5(-4) 9.9 -6) 0.45-1.58(-5) 1351 .* 7.8 -5) 0.050-3.2(-4) ?.1-5) 0.76-6.38(-5) 134Cs 5.4(-4) 0.030-2.16(-3) 2.1(-4) 0.11-7.5(-4) 1.3(-4) 0.88-1.98(-4) 136Cs 5.5(-6) 0.10-2.3(-5) 3.a(-6) 0.93-7.2(-6) 137Cs 9.0(-4) 0.0042-3.24(-3) 4.0(-4) 0.053-1.6(-4) 2.2(-4) 1.6-3.6(-4)
*** 0.685-2.9(-2) 4.7(-2) 3.22-6.34(-2)
{ 3H 3.5(-2) 1.0(-3) 14C 8.0(-6) *** 2.6(-6 1.5-3.69(-6) 2.l(-5 1.19-3.54(-5) 32P 3.2(-5) *** 2.2 -5 1.9(-5 1.20-2.71(-5) 24Na
- 5.1 -6 1.7-8.6(-6) 7.8 -6 0.11-1.6(-5) 51Cr 2.0 - 0.019-8.2(-4) 1.8 - 0.23-7.1(-4) 3.9 -5) 0.11-1.3(-4) 54Mn 5.3 - 0.0058-2.57(-3) 9.6 - 0.11-4.7(-4) 3.9 -5) 0.184-1.01( a) 55Fe 4.0 - *** 2.3 - 0.549-4.06(-4) 7.4(-4 5.22-9.53(-4) 59Fe 5.8 - 0.0010-2.5(-4) 2.5 - 0.28-7.9(-5) 7.8(-6 0.74-8.2(-6) 57C0 1.9 -5 0.025-9.1(-5) 1.7 - 0.14-9.6 (- 5 ) 3.6(-6 0.84-9.1(-6) seCo 1.1 -2 0.0043-4.29(-2) 6.7 - 0.041-4.0(-2) 4.0(-4 0.0024-1.35(-3) 60Co 2.6 - 0.0050-1.35(-2) 1.7(-3 0.18-9. 3(-3) 5.3 -4 63Ni 3.8 - ***
1.2(-4 0.734-1.66(-4) 1.5 -4 0.24-1.6(-3)) 0.996-2.29(-4 65Zn I.9 - 0.06-5.1(-5) 3.0(-5 0.34-6.2(-5) 5.2 -6) 0.17-1.8(-5) 893p 1,4 _ *** 7.5(-6 0.276-1.22(-5) 8.8 -6) 0.376-1.80(-5) 90Sr 8.8-8) *** 4.7(-7 3.3-6.2(-7) 1.l(-6) 0.122-2.41(-6) 91Y 3.7 -7) *** 7.6(-7 0.49-1.04(-6) 5.5(-7) 2.6-7.4(-7) 95Zr 2.6 -5) 0.032-1.15(-4) 2.9(-5 0.29-1.10(-4) 2.6(-5) 0.026-1.4(-4) 95Nb 4.4 -5) 0.048-2.02(-4) 4.7(-5 0.022-1.50(-4) 2.5(-5) 0.57-8.6(-5) 99Mo 9.3(-5) <0.0033-6.13(-4) 1.5(-5) 0.35-2.6(-5) 7.3(-6) 0.14-3.4(-5)
TABLE 4.16 (cont'd) RADIONUCLILE CONCENTRATIONS IN HOLDUP AND WASTE HOLDUP TANKS Holdup Tanks A, B, and C i Waste Holdup Tank No. 1** Waste Holdup Tank No. 2 iti Mean Range Mean Range Mean Range Nuclide (pCi/ml) (pCi/ml) _(pCi/ml) (pCi/ml) (uCi/ml) (pC1/ml) 103Ru 5.7 - 0.08-3.45(-5) 2.7 - 0.017-1.2(- ) 3.1 - 1.1-6.1(-6 lo6Ru 7.8 - 0.092-3.7(-5) 5.1 - 0.062-1.7(- ) 4.7 - 0.097-1.4()) 110 mag 8.4 - 0.018-<5.8(-4) 8.7 - 0.29-1.2(-5 1.7 - 0.17-9.6(- 124Sb 1.3 - 0.015-6.2(-4) 1.7 - 0.10-9.2(-4 1.2 - 0.47-2.8(- 125Sb 1.8 - 0.0046-1.06(-3) 1.8 - 0.010-1 .0(- 3) 2.7 - 1. 3-6.1 (-5) 129mie 1.6 - < 0.11 -4. 8 (- 5) 1.3 - 0.88-1.98(-4) 140Ba
- 1.0-5) 0.34-2.7(-5) 140La 6.7(-5) 0.0018-3.77(-4) 3.5(-6) 0.56-7.7(-6) 5.7* -6) 0.20-1.31(-5) 141Ce 1.8(-6) <0. 36-7.4 (- 6) 144Ce 2.0(-6) <0.83-4.6(-6) *
- Radionuclide not detected
*** One measurement, only, for radionuclide t Data obtained from inlet to base cation demins.
tt Data obtained from feed to radwaste evaporator, ttt Based on the following samples l 1130,4/25/78 1611,5/16/78 1434,4/26/78 1755, 5/16/78 . , 1143, 4/29/78 1110,5/20/78 1424,5/2/78 1250, 5/23/78 1044,5/10/78 0858,5/25/78
TABLE 4.17 RADIONUCLIDE CONCENTRATIONS IN PADWASTE BUILDING MONITOR TANKS Monitor Tank A i Monitor Tank B it Monitor Tank C ttt Mean Rance Mean Range Mean Range Nuclide (u".1/ml) (pC1/ml) _(pCi/ml) (uCi/ml) (pCi/ml) (uCi/ml) 131I ** 5(-8) 1(-7) **
<1(-7) 134Cs 2.6(-7) 1.1-6.3(-7) **
6.5(-8) <7(-8) 137CS 4.l(-7) 0.07-1.0(-6) 1.2(-7) <0.8-2.7(-7) 5(-8) ** 3H 4.5(-2) 4.0-4.9(-2) 3.8(-2) 3.3-4.2(-2) 3.0(-3) *** 12C 5.7 - 0.28-1.1(-6) 3.6(-6) 0.82-6.3(-6) 5.5(-7) *** 32P 2.1 - 1.1-3.1(-7) <4(-7) <9(-7) 54Mn 4.3 - 2.9-7.8(-7) 4.0(-7) <l-7.4(-7) 7.l(-7) 6.6-7.9(-7) ssFe 5.4 0.65-9.7(-6) 8.3(-6) 0.44-1.22(-5) 1.l(-5) *** m 57Co 5(-8 3.8-6.2(-8) <4(-8) <6(-8)
- g seCo 7(-7 0.34-1.1 - 6.4(-7) 5.0-8.9(-7) 1.5(-6 0.88-2.7(-6) 60C0 2.l(-6) <0.2-3.9 - 2.7(-6) 0.99-4.8(-6) 5A(-6 4.8-6.l(-6) 63Ni 2.6(-6) 0.18-5.8 - 5.6(-6) 0.085-1.03(-5) 1.9(-6 ***
89Sr <0.35-3.8(-8) *** 1.5(-8) <6(-9) 1.3(-8) *** 90Sr 2.0(-8) 0.84-3.6(-8) 3.2(-8) 2.8-3.6(-8 3.9(-8) 91Y 2(-9) <2-5(-9) 3(-9) 2.1-4(-9) ) 4(-9) *** ssNb 1(-7) <0.93-1.6(-7) <5(-8) <1(-7) 99Mo <1(-7) 5(-8) <6(-8) 106Ru 1(-6) <4(-7) 1(-6) <0.08-2.4(-6) 110 mag 1(-7) ** 1.4(-7) <0.7-3.3(-7) 3.5(-7) 0.71-8.8(-7) 124Sb 2(-7) 1.3(-7) 8(-8) 0.44-1.4(-7) 12sSb 5.1(-7) 3.1-6.8(-7) 3(-7) <2-4.6(-7) 5.1(-7) 0.78-9.0(-7) 140Ba <1(-7) <2(-7) 1.4(-7) t Samples 4/25/78; 1032 5/20/78; 1110 (gamma analysis only) 4/26/78; 1330 5/23/78; 1355 (gamna analysis only) 4/29/78; 1128 5/25/78; 1545 (gamma analysis only) tt 3amples 5/2/78; 1414 ttt 5/7/78; 1513 5/8/78; 1743 (gama analysis only) 5/10/78; 1584 (cama analysis only) 5/16/78; 2135 5/25/78; 1020 (gama analysis only)
** Radionuclide detected in only one sample *** One measurement, only, for this radionuclide.
TABLE 4.18 RADIONUCLIDE CONCENTRATIONS IN . AUXILIARY BUILDING MONITOR TANKS i t it Tank A Tank B Mean Range Mean Range Nuclide (pCf/ml) (pCi/ml) (pCi/ml)_ (di/ml) ] ! 131I 5.7(-6) <0.0097-1.70(-5) 1.7(-5) 0.015-5.0(-5) 133I 1(-7) <0.8-1.4(-7) <2(-8) i 134Cs 2.7(-8) 0.98-5.7(-8) 6(-9) <5-8.1(-8) 137Cs 1.9(-7) 1.5-2.3(-7) <4(-8) 3H 6.7(-2) 0.159-1.00(-1) 7.5(-3) 14C 2.9 - ) 0.24-4.3(-6) 5.1(-7) 32P 3.5 - ) <4-9(-7) <6(-7) l SICr <5( 7 8(-7)
! 54Mn 4(-) 0.38-8.2(-8) 1(-7) <0.7-21(-7) ssFe 5.2 - 0.21-1.11(-) 1.9 - ***
58C0 4.9 - 0.70-9.4(-7 6.2 - 0.091-1.62(-6) 60C0 3.6 - 0.88-8.9(-7 7.3 - 0.25-1.8(-6) 63Ni 6.0 -7) 3.7-7.8(-7) 9.1 - ) 89Sr 1(-8) <0.3-2.7 - <4( 9
, 9aSr 7.6 - 0.28-1.2 - 1.3 - ) ***
sly 1.2 - 0.22-2.7 - 24-) *** 95Zr 1.1 - <1-1.9(-7) 2- ** 95Nb 1.0 - <0.8-2.3(-7) 3- ** 99Mo <6 - 6- 3.0-8.7(-8) 103Ru <6 - 3- ** 106Ru <7 3- ** 110 mag 1(- . <1-1.4(-7) 6- ** j 12sSb 3(- <2-6.2(-7) 1.3(-7) <1-1.8(-7) i t Samples 4/29/78,11:24 5/16/78,14:48 , 5/18/78, 11:16 tt Samples 1 5/2/78,11:17(gammaanalysisonly) 5/8/78,18:13 (gama analysis only) 1 5/10/78,14:06 Co Radionuclide detected in only one sample. l
** One measurement, only, for this radionuclide. 1 i
i 157 _ . _ , .,. -. -_ , _ . - , m,-- _ _ _ ~ .
l TABLE 4.19 i EXTRAPOLATED ANNUAL RELEASES FROM HONITOR TANKS Extrapolated Annual Release Nuclide (pCi) 131I 1.9(5) 134Cs 1.3(3) 137Cs 2.6(3) l 3H 1.0(9) 14C 5 4) 32p 7 ! SICr 9 l 54Mn 3 ; sspe 1,0 ( 57Co 1.3 seCo 1.3 60Co l 63Ni 2(4') 2(4 89Sr 1.8(2) 90Sr 3 91Y 3 957r 3 9sNb 4(3 99Mo 1 0 3) 103Ru 5 106Ru 8 11onAq~ 2 l 124Sb 5 12sSb 2 140Ba 3 t Extrapolated values based on radionuclide l concentrations measured during 4/25-5/25/78 I period and volume of liquid released during ( 7-month period from 11/1/77 to 6/1/78. l l l 158
water from other sourc . There is a correlation of the activity levels in tb. ;vaporator bottoms with the distillate - such that a higher levc. in the bottoms is reflected as a higher concentration in the distillate.
- 2. The DF's across the base cation demineralizer when loaded with a mixed-bed resin show a high correlation with inlet concentration for ionic species such as 1311 and 134 137Cs.
- The crud-associated radionuclides do not show this correlation.
- 3. The crud-associated radionuclides show a high correlation between the feed concentration and the DF across the evaporator l (inlet / distillate). This correlation is apparently valid for all evaporators studied at three nuclear power plants. The radwaste evaporator showed a similar correlation for cesium
- and iodine radionuclides. The boric acid evaporator had insufficient levels of 134~137Cs to establish a correlation and the iodine levels dropped to such a level that an accurate determination DF's was not possible and any correlation could not verified.
- 4. Because of the effects noted in 1-3 above, the boric acid recovery system and the radwaste evapurator system tend to attenuate, if not stop, spikes that reach their inlets and prevent the spikes from propagating to the monitor tanks.
S. The actual efficiency of the evaporator is related to the bottoms concentration rather than the feed. This was shown to
! be true when the feed concentration was drastically reduced when compared to the bottoms and the DF actually dropped to a i value less than one. During this time the bottoms-to-distillate ratio remained relatively constant.
- 6. The radwaste evaporator exhibited the phenomenon of having less cobalt activity in the distillate when the bottoms or feed was more concentrated. The opposite case was observed ,
, for 1311. The boric acid evaporator exhibited a correlation between distillate and bottoms concentrations for 131I. No correlation was found for cesium and cobalt.
i 1 l l 1 159
- 5. SPENT FUEL PIT
5.1 System Description
There are two spent fuel pits (SFP) at Turkey Point, one for each reactor. The fuel pits and associated cooling, purification and ventilation systems are identical with the following exceptions. Spent fuel pit #3 has a leak recovery system as well as its own ventilation release vent. Spent fuel pit #4 does not have a leak recovery system and ventilation exhaust feeds the main plant exhaust stack. A detailed drawing of spent fuel pit cooling and purification systems for Units #3 and #4 is shown in Appendix B Figure B.14. The cooling system consists of a pump, heat exchangers, filters and a mixed-bed demineralizer. In nonnal operation, water from the SFP is circulated through the heat exchangers and returned to the SFP. A nominal flow of 2000 gallons per minute results in one SFP volume (330,000 gallons) change every 2.75 hours. The purification loop for the cooling system consists of three 25-micron flushable filters and the mixed-bed demineralizer. The mixed-bed demineralizer bed has a depth of 5.5 feet and a diameter of 31.5 inches (5.4 ft 2 cross sectional area) and contains 30 ft of resin (Rohm & Haas Amberlite IRN-150 or equivalent). Nominal flow through the purification loop is 100 gallons per .ninute. Based on this flow and a 30 ft3 bed volume, the bed residence time is 2.3 minutes. The SFP's ventilation system consists of two supply fans and an exhaust fan (see Figure 1.2). The exhaust fan is rated at 20,000 cubic feet per minute. The supply fans have a total capacity of 3000 cubic feet per minute, the balance of the exhaust flow coming from s n-leakage. The SFP ventilation exhaust is processed through a series of prefilters and HEPA filters (40 of each type filter) before going to the environment. The prefilters are two feet square by two inches deep while the HEPA filters are two feet square by 11.5 inches deep. None of the filters has ever been changed. Also, as noted in Figure B.14, each SFP has a spent fuel pit skimmer system. These skimmers were not in use during the measurement progran at Turkey Point. The leak recovery system for Unit #3 SFP was used intennittently to remove water, which had leaked between the SFP stainless steel and concrete liers. The recovered liquid was processed through the SFP purification sy em. When operating, the flow was nominally 1-2 gallons per minute. 5.2 Measurements Figure 5.1 shows a schematic diagram of the Unit #3 SFP Cooling and purification system. The indicated sample points are the sampling locations used in the study. Sample I was a dip sample taken directly out of the SFP. Samples 2-4 were collected through pennanently installed sample lipas. 160
Figure 5.1 Schematic Diagram of Unit #3 Spent Fuel Pit System 4
; L . _ _ __
SFP Heat m exc l HUT l A y RWST , r----- , o _h _ ,
- d RCDT e
k i d +, L _ _ _ _ _ .s y m A A
- v v o
H+-O H - Leak demin chase 30 f t3 pump
~~~~
e l RCS n I l hot l i 8 legs f I L_____. - Filter - INEL-A-11083
% Sampling stations 161 1------
't In the Unit #3 SFP study t 131 1, particulates, 14C, and 5H .heResults ventilation exhaust was of the measurement period,monitored for which extended from 12/8/77 to 6/1/78, are in Appendix B, Tables B.72 and B.82. Data associated with the SFP demineralizer DF measurements are in Appendix B Table B.31. Tables 0.32 and B.33 of Appendix B contain ,
l the results for samples of the spent fuel pit and associated water for 1 beta-only-emitting and gamma-emitting radionuclide analyses. Sampling methods used in the st dy are described in reference 4. 5.3 Results and Discussion After Unit #3 went down for its fourth refueling, 40 fuel assemblies were transferred from the Unit #3 reactor core to the Unit #3 SFP. The fuel assemblies were transferred between 1/4/78 and 1/25/78. The original 23sU enrichment of the transferred fuel assemblies ranged from 1.8% to 2.7%. The 235 U content of the fuel assemblies, when transferred, ranged from 0.44% to 0.50%. The transferred assemblies had a combined useage l of 72,676 megawatt days. Unit #3 spent fuel pit previously contained ' 107 fuel assemblies from prior refuelings. During the interval 3/23/78 to 4/25/78 all fuel assemblies ir Unit l
#3 SFP were transferred to Unit #4 SFP. This was done to facilitate l repair of the Unit #3 SFP liner. The water from #3 SFP was processed through the #3 BAE during May,1978 (see section 4.2).
5.3.1 Unit #3 Fuel Pit Area Extrapolated Annual Gaseous Releases for 1311, 3H, and 14C The average 3H and 14C release rates downstream of the HEPA exhaust filters from Unit #3 SFP via the vapor pathway are given in Table 5.1 together with extrapolated annual releases. The releases for both radionuclides are based on sampling interval data (Table B.72 of Appendix B) where both analysis of oxidized and unoxidized species was obtained. The annual releases presented include data for the refueling and non-refueling interval, the fuel movement from SFP #3 to SFP #4 interval, and the interval of water removal from SFP #3. By including the fuel transfer and water removal operations, uncommon practices, i one would expect the release rates to be lower than during nonnal ) operations since there were intervals with no fuel or water in the #3 SFP. I However, for these two radionuclides (3H and 14 C), the average release rates are lower if the release rates during the fuel movement and water removal operations are excluded. Tnc release rates obtained by excluding the release rates during these operations are 2.9(-2) and 9.8(-3) UCi/sec for H3 and 14 C respectively. Consequently, the reported releases (Table 5.1) are upper limits i.e., the release rates would be lower during more normal power operation. It should be emphasized that the reported 14C and 3H releases are for Unit #3 SFP only. To correct to total plant extrapolated annual releases from the SFP's, the 3H release i 162 l l
1 4 1 TABLE 5.1 EXTRAPOLATED ANNUAL RELFASES OF GASEOUS TRITIUM,131I AND 14C FROM THE UNIT #3 FUEL PIT AREA (For Refueling and Non-refueling Combined) Average Extrapolated + Release Rate Annud' Release Nuclide (uci/sec) (C1/ year) 3H 8.8(-2) 2.7 I 14C 1.2(-2) 0.4 131I 4. 8(-4) 0.02 i. 1 e I d 163
l should be multiplied by approximately 2. This is based on FPL 3H analyses of both Unit #3 and Unit #4 SFP's waters. The two analyses gave approximately the same results. Although 14 C analyses were not performed for the SFP 4 waters, the same argument could be applied to the 14 C extrapolated total
- plant annual releases from the SFP's. Extrapolated radionuclide annual l
releases in this section are obtained by multiplying the average release rate by the number of seconds per year (3.15(7)).
- Also shown in Table 5.1 is the extrapolated annual release of 1311 downstream of the exhaust HEPA filter. This result is discussed in detail in Section 8. Species data for 1311 from the Unit #3 SFP are 4 also presented and discussed in Section 8. The only gaseous particulate <
radionuclide detected in the Unit #3 SFP ventilation exhaust was 137Cs. The respective release rates, downstream of the HEPA filters, were 5.0(-7), 3.5(-7) and 6.0(-7) pCi/sec during the combined refueling-non-refueling interval, the non-refueling interval, and the refueling in%rral (Tables 2.5,2.7,2.8). For comparison purposes the 131 1, 3H, and 14C average release rate, downstream of the exhaust HEPA filter, during the refueling interval, i the non-refueling interval, the fuel transfer from Unit #3 to Unit #4 i SFP interval and the Unit #3 water removal intervals are shown in Table ! 5.2. J The measured distributions of H3 and 14C chemical species are shown in Table 5.3. The average levels of oxidized fonns of 3H and 14C are 53% and 24%, respectively. The reason for the chemical species variability is unknown. i 5.3.2 Unit #3 Spent Fuel Pit and Associated Water Tritium Mass Balance l The data in Table 5.4 represent a 3H mass-balance before and after fuel movement from the Unit #3 core. Fuel movement occurred during the interval 1/14-25/78. The total 3H in the refueling water storage tank (RWST), the reactor coolant, and the spent fuel pit waters was i observed to decrease from approximately 18 curies prior to the outage to approximately 11 curies during the outage. Tritium can be released into the ventilation air and into the monitor tanks via the boric acid evaporator. i Ventilation measurements indicate that only a small fraction of the 7-curie J balance (i.e., 0.1-0.2 curies) was released via the air pathway during the period 11/21/77-1/11/78. Although it is not possible to identify the portion of H 3 in the liquid wastes that originated from Unit #3, the total release is sufficiently large to account for the 7-curie difference 4 observed during refueling of Unit #3. For example, during December,1977
- and January,1978 a total of 75 monitor tank volumes were released. In order to account for 7 curies of 3H by the release of 75 monitor tanks (10(4) gallons each), the average 3H concentration in the monitor tanks would i
have to be 2.5(-3) pCi/ml. Although H3 measurements in monitor tank liquids i 164 i i
.~-
TABLE 5.2 AVERAGE RELEASE RATES OF GASEOUS 3H,14C, AND 131I FROM UNIT #3 FUEL PIT AREA (For refueling, non-re'ueling, fuel transfers, and water removal) Interval Refueling Non-refueling Fuel Transfers Water Removal Nuclide _(pC f/sec)_ (uCi/sec) (uCi/sec) (uCi/sec) 3H 2.9(-2) 1.2(-1) 2.0(-2) 8.6(-2) 14C 9.8(-3) 1.4(-2) 2.1(-2) 1.1 (-3) , 131I 7.5(-4) 2.6(-4) 5.6(-5) 4.7(-5) 165
l i '
- TABLE 5.3 )
1 UNIT #3 FUEL PIT AREA DUCT I GASEOUS 3H AND 1"C SPECIES i Sanele Period HTO (%)[A] CO2 (%)[B] l 12/8-12/12/77 ** ** ! 12/14-12/28/77 [C] [C] ) 12/28-1/11/78 83 33 1 1/11-1/25/78 67 9 1/25-2/8/78 47 21 2/8-2/22/78 [C] [C] 2/22-3/9/78 [C] [C] l 3/9-3/15/78 [D] [D] , 3/21-4/3/78 88 35 ! 4/3-4/20/78 24 2 4/20-5/4/78 57 58 l 5/4-5/18/78 7 27 5/18-6/1/78 53 11 [A] 0xidized fraction (%) of total 3H in sample. [B] 0xidized fraction (%) of total 14C in sample. [C] 0xidized 3H and 14C only. [D] Short samle period due to power failure.
** - No Sample I
166
TABLE 5.4 UNIT #3 TRITIUM MASS BALANCE Concentration Volume Activity Date Sample (uci/ml) (x 103 cal) (Curies) 11/21/77 EI3 SFP 2.3(-3) 296.9 2.58 RWST 4.3(-3) 328.0 5.34 Reactor Coolant 3.8(-2) 70.5 10.14 Total 695.4 18.06 12/30/77[2] SFP 3.0 - 331.2 3.76 RWST 4.4 - 270.0 4.50 Reactor Coolant 1.7 - 70.5 4.50 Total 671.7 12.76 1/3/78[3] SFP 2.8(-3) 332.0 3.52 RWST 3.7 -3 35.0 0.49 Reactor Coolant 5.1 -3 361 .5 6.98 Total 728.5 10.99 1/ll/78 E43 SFP 3.2(-3 326.5 3.95 RWST 3. 7 -3 37.0 0.52 Reactor Coolant 4.5 -3 361.5 6.16 Total 725.0 10.63 [1] Samples were collected during power operations, before outage. [2] Samples collected during outage. The transfer canal had been flooded and its contents (3H activity) are included in the SFP total. Reactor cavity had not been flooded. ' [3] Samples collected during outage. The transfer canal had been flooded and its contents (3H activity) are included in the SFP total. Also the reactor cavity had been f?toded. The reactor cavity contents (3H activity) are included in the reactor coolant total. Fuel movement from the core to the SFP had not started. [4] Samples collected during outage. After fuel movement from core to SFP had started. The transfer canal and reactor cavity 3H activities are included in the * ' and reactor coolant totals. 167
4 were not performed during the Unit #3 refueling period, subsequent measure-ments made in April and May (see section 4.4) indicate an average 3H l concentration of 4(-2) pCi/ml in monitor tanks. Therefore the 7 curie 3 H balance can be accounted for by as little as 6% of the IH in the monitor tanks originating from Unit #3 refueling waters. l 5.3.3 Unit #3 Spent Fuel Pit Demineralizer Decontamination Factors ' l During February and April, five sets of samples were taken l from the #3 SFP demineralizer inlet and outlet streams to determine l demineralizer decontamination factors. Table 5.5 lists pertinent sample ! parameters for these sample sets.
]
Table 5.6 presents the measured DF's. The mixed-bed resin was replaced January 20, 1978, and was not replaced again during the SFP demineralizer study.. I l The inlet sanples obtained during February,1978 were single 450 ml samples taken at the beginning of each measurement. The effluent samples were nominally 100 liter samples collected on ion exchange resin samplers (4) for approximately 22 hours. For the 4/01/78 measurement, three sets of inlet samples were collected. Each set consisted of duplicate 450 mi samples which were collected at the beginning, middle, and end of the 24-hour sampling period. Average radionuclide concentratives for the 6 i samples are presented in Appendix B Table B.31. The effluent sample was a 216 liter sample collected again in the 100 exchange sampler. Inlet and outlet samples obtained on 4/15-16/78 were 450 ml. As indicated in Table 5.6, the DF's measured on 4/1/78 were exceptionally high in comparison to the results obtained in February. During this l time period, fuel elements were being transferred from #3 SFP to #4 SFP. j The mode of operation during this period was to secure the #3 SFP purification l loop when fuel was being moved. If the purification loop was in service, l turbidity of the SFP waters hindered visibility. Consequently, the SFP purification system was put into service on weekends only. Samples were collected during purification system operation. The bed volumes, taken from plant information, denoted in Table 5.5 reflect this mode of operation. Measurements (see Appendix B, Table B.31) indicated that on 4/1/78 the radionuclide concentration in the inlet to the SFP demineralizer had increased (e.g., by approximately a factor of 2 for cobalt,10 for antimony and cesium, and 100 for iodine) while the radionuclide concentration in the outlet had decreased (e.g., by approximately a factor of 100 for cobalt and iodine,1000 for antimony, and 10 for cesium). The tendency for demineralizer DF's to be directly releated to inlet concentration observed for other mixed-bed demineralizers (see section 2.2.7) cay be responsible for a fraction, but not all, of this increase in DF's, The full explanation for the unusually high DF's observed on 4/1/78 has not yet been found. 168
- __ - . _ = _
. .- -__ .=. _ _ - . -
TABLE 5.5 SAMPLE INFORfMTION FOR UNIT #3 SFP DE!11NEPALIZER TESTS Parame ter 2/5/78 2/6/78 4/1/78 4/15/78 4/16/78 Bed Volumes 1.0(4) 1.l(4) 4.1(4) ' 4.3(4) 4.4(4) pH 4.76 4.83 a.93 5.00 5.01 Boron (ppm) 2100 2120- 2010 2000 1970 Conduc ivity 6.76 6.75 6.2 5.95 6.03 Temperature SFP 97 96 97 '97 95 (*F)
TABLE 5.6 MEASURED DF's FOR UNIT #3 SFP DEMINERALIZER Nuclide 2/6/78 2/7/78 4/1/78 4/15/78 4/16/78 3.8 *
- 131I 2.1 0.3(0) 2.9 1 0.9(0) 0.4(4) 134Cs 3.0 0.8(2) 1.9 0.4(2 7.5 0.4(4) 9.5 0.7(-1) 8.8 0.6(-1) 137Cs 2.9 i 0.4(2) 1.2 0.2(2 7.1 0.3(4) 8.5 0.6(-1) 7.9 0.3(-1)
SICr 1.70 0.08(1) 4.3 0.2 8.1 1 0.9(3) 54Mn 5.8 0.6(1) 5.1 0.4 1.37 1 0.05(5) 3.4 0.9(0) 5.0 1.1(0) 59Fe 6.8 1.3(0) 1.4 0.5 3.3 1 0.4(4) *
- 57Co 1.90 0.11(2) 1.4 1 0.2 4.0 0.3(4) seCo 1.62 0.04 2) 1.32 0.06(2 1.8 0.1(5) 2.7 0.2(1) 4.3 0.2(1) 60Co 5.210.1 3.6 0.1(2)) 4.0 0.2(5) 5.8 1 0.2(1) 5.8 1 0.4(1) 2! ssZn 3.0 0.9
- 95Zr 1.6 0.1 1.0 0.3 1.6 0.2(a) 9sNb' 7.0 1 0.7(0 5.8 0.5 1.48 0.08(4) 3.2* i 1.1(0) 1.2 0.d(0) 103Ru 2.0810.051) 1.7 0.2 1.2 0.2(4) 106Ru 1.9 0.3(1) 1.8 0.4
- 110 mag 6.7 0.7(1) 4.2 1.0 8.7 0.9(4 124Sb 3.2* i 0.2(0) 3.! 0.1 5.3 1 0.5(4 1.i : _0.i;0)
'(0) 1.1 1 0.1(0) 12sSb 2.7 i 0.1 2.1 1 0.1(4 1.0 1.3 0.1(0) 129tTTe 2.0 0.1(0 1.7 0.3(0
- 129Te 9.5 1 2.0(0 * *
- 141Ce 1.5 0.2 1 2.310.3(1) *
- 144Ce 2.5 0.4 1 2.0 1 0.3(1) 6.6 3.1(4)
- Radionuclide not detected
1 Inspection of Table 5.6 also indicates that on 4/15-16/78 the DF's were much lower than observed in February although the inlet concentrations were in the same range. One reason for the observed decrease in DF's may be that the resin was becoming saturated (with both radioactive and non-radioactive species) due to high inlet concentrations caused by stirring up of SFP water during fuel transfer. The DF's for beta-only-emitting radionuclides for the mixed-bed demineralizer are shown in Table 5.7. Table 5.8 lists the means and ranges for radionuclide concentrations in #3 SFP demineralizer inlet and outlet streams together with "best value" DF's. In obtaining these means, ranges, and DF's, data from the 4/1/78 measurements were not included due to the presence of high particulate concentrations. Comparison of these DF's with DF's obtained for the mixed-bed CVCS demineralizers (Tables 3.12 and 3.15) indicates that when the inlet concentrations were l in the same range, the DF's were in the same range also. In cases where CVCS inlet concentrations were higher than for the SFP demineralizer, , the CVCS demineralizer exhibited larger DF's. This gives further support to the observation that DF is related to inlet concentration. t 1 4 a i 1 I F j 171
4 1 TABLE 5.7 i UNIT #3 SFP DEMINERALIZER DF's FOR BETA-ONLY EMITTING RADIONUCLIDES Date 3H 14C 91Y 893p 90$p 55Fe 63Ni 11/21/77 0.98 0.30 32 >930 324 12.4 >1650 ~ 12/30/77 0.97 0.82 >22.5 >120 100 5.1 >6900 1/25/78 1.04 0.96 13.1 >830 58 1.4 127 h i i 4 h 172
TABLE 5. 8 KANS AND RANGES FOR RADIONUCLIDE CONCENTRATIONS AND "BEST VALUE" DF's FOR UNIT #3 SFP DEMINERALIZER Inlet Concentration (vCi/ml) Outlet Concentration (uci/ml) "Best Value" Nuclide Mean Rance Mean Range DF 131I 5.3(-8) 5.3-5.3(-8) 2.1(-8) 1.8-2.5(-8) 2.5(0) 134Cs 6.0(-6) 0.11-1.0(-5) 5.9(-6) 0.00037-1.22(-5) 1.0(0) 137Cs 7.8(-6) 0.22-1.34(-5) 8.1(-6) 0.00086-1.65(-5) 1.0(0) 14C 3.5(-8) 1.6-5.3(-8) 5.1(-8) A.4-5.4(-8) 7(-1) 51Cr 2.3(-5) 1.29-3.3(-5) 7.6(-7) 7.6-7.7(-7) 3.0(1) 54Mn 1.9(-6) 1.5-2.1(-6) 2.2 -7) 0.36-4.4(-7) 8.6(0) ssFe 4.1(-5) 2.96-5.20(-5) 1.1 -5) 0.42-2.18(-5) 3.7(0) 59Fe 3.6 - 1.7-5.5(-7) 3.1 -8) 2.5-3.8(-8) 1.2(1 57Co 2.0 - 1.8-2.15 -6) 1. 3 -8) 1.13-1.5(-8) 1.5(2 O seCo 4.0 - 2.49-5.8 - 4.8-6) 3.14-5.8(-6) 8.3(1 60Co 4.6 - 3.55-5.7 3.8 -6) 1.05-6.6(-6) 1.2 2 6 3Ni 6.8 - 2.76-9.9 - 2(-6) <0.04-6(-6) 3.4 2 ssZn *** *** 3.6 - 1.2(-8 3.02) 95Zr 1.0 - 0.82-1.2(-7) 7.8(-8 7.8-7.9(-8) 1.3 0) 9sNb 1.1 - 0.80-1.9(-6) 4.2(-7 1.33-8.3(-7) 2.7(0) 89Sr 9.4 - 0.12-1.86(-5 <1(-8) >9.4(2) 9aSr 2.1 - 1.02-2.92(-6 2(-8) 0.9 A.0(-8) 1.0(2) 91Y 1.0 - 0.18-2.62(-6 7.1 - <0.09-2.0(-7) 1.4(1 103Ru 1.0 - 0.95-1.08(-6 5.3 - 5.2-5.5(-8) 1.9(1 106RL 1.2 - 1.2-1.3(-6) 6.8 - 6.4-7.3f-8) 1.8(1 110 mag 6.4(-7 6.1-6.7(-7) 1.3(-8) 0.91-1.6(-8) 4.9(1) 124Sb 6.6(-6 5.3-8.0(-6) 4.5(-6) 1.64-7.5(-6) 1.5(0) 125Sb 7.7-6) 3.16-1.1(-5) 6.3(-6) 1.19-8.9(-6) 1.2(0) 129Cre 3.3 -6) 3.3-3.4(-6) 1.8(-6) 1.66-1.98(-6) 1.8(0) 141Ce 5.5 -7) 5.2-5.8(-7) 3.0(-8) 1.8(1) 144Ce 2.8 -6) 1.9-3.7(-6) 1.2(-7) 2.5-3.5(-8)) 0.93-1.47(-7 2.3(1)
*** Radionuclide detected in one measurement set only.
i i l i
- 6. SECONDARY SYSTEM i.
6,1 Introduction ! This section presents the results of studies of primary-to-secondary
- leakage at the Turkey Point Plant. Unit #3 operated with a primary-to-secondary leak in the C steam generator from late August 1977 until t
late November 1977 when the unit was shut down for scheduled refueling. During this peri;,d, Unit #3 also experienced main condenser tube leakage i resulting in high chloride levels in the secondary system water due to '
- sea water in-leakage. Primary-to-secondary leakage studies on Unit #3 were performed during the time period 11/9-21/77.
1 Primary-to-secondary leakage occurred in the 4A steam generator in mid-January 1978 and continued until mid-February 1978 when Unit r4 was shut down to repair the leaking steam generator. Studies on Unit #4 were conducted from 1/8/78 to 2/9/78. l Figure 6.1 is a diagram of the Turkey Point secondary system which i represents both Units #3 and #4. Rated flows for the various secondary i system flow paths are included on Figure 6.1. Figure 6.2 is a flow diagram for both Units #3 and #4 secondary systems. This figure shows the analysis points where samples were taken. Results from these samples were used to evaluate the performance of the secondary system components during a primary-to-secondary leak. 1 The following system description applies to both Unit #3 and Unit #4.
6.2 System Description
6.2.1 Steam Generators q There are three Westinghouse U-tube steam generators per i nuclear unit. Each generator is rated at 3.2(6) lbs/hr steam flow at j 730 psig, 510 F and 0.25% moisture content. Normal operating volume is i 25,000 gallons. Swirl Vane assemblies and chevron mist extractors
! remove moisture from the steam as it exits the top of the steam generator.
6.2.2 Turbine Train The turbine train was fabricated by Westinghouse and consists of one high pressure (HP) and two low pressure (LP) turbine units. I 1. HP Turbine - double flow type with one impulse stage and seven rows of reaction blades at each end of the turbine.
- 2. LP Turbine - double flow type with nine rows of reaction blades at each end of the turbine.
l 174
~ . , - - - - - , -
- , . , _ _ , .e-,- .-,..,----.--w.-,rea -
f t 4 i Figure 6.1 Diagram of Secondary System , 6.876.794 lbs/hr I 155 psig 476*F s e I Generator P turbine I l l Gross Generation 728.317 KW ( 66 psig ( l A u x ilia ries- 35.210 21 psig f~. [ _- Net Generation 693.107 2.4 in hg i 24.1 in hg ' 5 595.921 lbs/ hr _ 27.5 in. hg I i l 1 ( Rehtr. drarn tank Condenser 108 7* F Main Steam from steam gland seals 334.578 lbs/hr 215.609 lbs/hr _ 1.000 lbs/hr 5,970 lbs/hr O'.371,0 l p l 6.890.214 [ l Htr n1 ) [ l Htr a; ?) { l Htr #3 SJAE. Gland seal I ' l l l Condensate condenser . pump , o o j 1.294 293 lbs/hr 119 7* F ~ 1765'F ' 210.1' F Drain to sump 175
)
l J dIh ; 620.080 lbs/hr l
- - . , . p) 5101 i
730 ps.<;
~'
e, 1000 lbs/hr 360 lbs/hr i 2 _ Moisture 7 651.884 lbs'hr ; separator - reheater y.. To SJAE To
" 167 psig 10 psig 162 psig f 167 95i9 aj L: }
Stm X HP qen turbine c L> 775.090 lbs/hr 393 psig _ 1 1 r ' 1,917.189 167 psig , i Ibs /hr. r 1 3 C"lan "3 2.692.279 lbs'hr pump
= -
M 547.882 o ' lbs/hr s3 lbs/hr* gy OE 620,080 lbs/hr
, 2531
- F 304 5* F 749.227 lbs/hr
, f F 368 9* l 9 582 493
-- O Ibs hr
) i l Htr #4 ) l Htr #5 ) = Stm Gen lHtra6] j
- b _ _ . feed pump I 436 5'F o O O
+ e }'364 7'F( s 263*F 378 9 'F INEL-B 11081 Note Numerical values shown are at 100*o of rated flow.
Flow values (lbs/hr) are for total plant flow
/
\
)
Figure 6.2 Schematic Diagram of Secondary System j j g To atmosphere j gg C O
~
Low pressure j Air L turbines (2) ejectors condenser To atmosphere IIGland J Main condenser Condensate g AL seal l' condenser
)L Reheat system l Moisture Drains High pressure -
m separator - drain tank reheater O3 Main feed g l J g i f
' Analysis points l
D Primary and letdown systems D Steam generator and main steam systems High @ HP turbine, moisture separator and HP drain pressure system turbine LP turbine and main condensate systems Steam air ejector unit Main feed system l j D Blowdown system
- Samples Taken Steam Letdown generator demin Air Blowdown ed vent it Ak II Reactor Blowdown I flash tank lf To discharge canal m n-A-n 084 176
t
- 3. Turbine Train Rating I
, Output - 728 MW Exhaust Pressure - 1.5 in. Hg Speed - 1800 RPM Speed Controt - Oil Governor . Steam Inlet Pressure - 730 psig Overspeed Trip - 1998 RPM Steam Inlet Temp. - 510'F Low Vacuum Trip - 20 io. Hg 6.2.3 Gland Seal The steam used in the gland seal comes from two sources. At less than 15% HP turbine speeds, the gland seal is provided to the HP and LP turbines from main steam through a 3- to 5-1b pressure regulator. Rated steam flow is 5,280 lbs/hr. At greater than 15% HP turbine speeds, leakage from the HP turbine supplies steam to the gland seal. , 6.2.4 Moisture Separator Reheater (MSR) There are four of these U-tube and shell type reheaters per { turbine train. The MSR's remove moisture from the steam between the HP and LP turbine units. As steam passes through knit wire demisters,10% i by volume of the entering steam leaves as moisture, the remaining 90% exits to the LP turbine as superheated steam. 6.2.5 Steam Jet Air Ejectors (SJAE) Foster Wheeler (2 types)
- 1. Hogging Ejector - used for startups and emergencies. Flow rating unknown.
1
- 2. Main jet bank - the main SJAE bank consists of two sets of two-stage steam jet air ejectors and is used when J
the main condenser vacuum is between 25 and 29 inches. i The flow rating is 15 cfm/ set. One set is used while the other remains in standby. Both types of SJAE's utilize low pressure (approximately 200 psig) steam for operation. 1 6.2.6 Main Condenser The main condenser was fabricated by Foster Wheeler and consists of 2 shells per turbine unit. Each shell serves one LP turbine. The shells are interconnected with equaliziing lines between the steam space and the condenser hotwell. Hotwell level is maintained automatically between 25,000 and 40,000 gallons. Makeup water is supplied automatically from the condensate storage tank. i b l 1 177
1 t i 4 1 I 6.2.7 Condensate Storage Tank j The condensate storage tank has a capacity of 250,000 gallons, of which,185,000 gallons (80% level) is reserved for emergency cooldown. It is used to supply makeup to the main condenser hotwell by pump (450 gpm) during startup and automatically by gravity feed when the condenser is under vacuum. Makeup is shut off if the tank level falls below 79.3% l evel . I 6.2.8 Condensate Pumps There are two 4-stage, 1170 RPM vertical condensate pumps ! per unit. These pumps are rated at 8300 GPM (60% of full condensate system flow) with a total discharge head of 980 ft of water. They are designed to operate at a maximum condensate temperature of 180'F. 6.2.9 Feedwater Reheaters i
- 1. High Pressure reheaters - there are two of these U-tube and shell reheaters. They utilize the HP turbine and moisture j separator reheater drains to reheat feedwater. The drain
- water returns to the main feed system.
- 2. Low Pressure reheaters
; a. The air ejector and gland seal condensers are used to initially reheat the main condensate. Condensed steam from the air ejectors or gland seal return to the 4
condenser hotwell.
- b. Four U-tube and shell reheaters utilize the low pressure i turbine drains to reneat the main condensate. The LP turbine drain water returns to the main condenser hotwell.
6.2.10 Chemical Injection There is one chemical tank and one chemical addition pump
; per unit. In addition there is one pump that can be shared by both ! units. The pumps are positive displacement.'
6.2.11 Feed Water Pumps i i i There are two motor driven pumps per unit, each rated at 60%
- of full feedwater system flow. These pumps are horizontal, centrifugal, multi-stage split case units built by the Byron Jackson Co. The pumps are directly connected to 3600 RPM, 7000 horsepower drip proof GE motors.
Pump speed / capacity is controlled by the steam generator level control system. i 178
- - - - - - - , _ . , - - . , , ,..e , - , , - . -,_,,,,,-,...n-, _ -
4 1 ) )
- 6.2.12 Circulating Water Pumps There are four Foster Wheeler vertical, mixed-flow circulating
+ water pumps per unit. Each pump is rated at 156,250 GPM with a dynamic water head of 23 feet. The pumps are driven at 236 RPM by 1250 horsepower Westinghouse motors. 6.2.13 Rated Flows Tables 6.1 and 6.2 contain rated flows for a single unit in both lbs/hr and gm/sec. These values apply to both Unit 63 and #4 and are the flows used for calculations in this report. 6.3 Sample Points
'During the sampling program, samples from the secondary system were taken locally and/or at a sampling sink located on the mezzanine level of the turbine deck.
! Local samples were taken directly fran the sampled component. These samples were:
- a. Blowdown flash tank liquid stream and air vent
- b. Condensate storage and recovery tanks
- c. Air ejector exhaust I
The condensate str age and recovery tanks were purged for 10 minutes at 2 liters / min prir to san.pling. t The sampling sink contained sampling points for both Units #3 and I
#4. Samples taken at this location were routed through sample coolers.
Samples available at the sink were:
- d. Main Steam Samples - plant procedures require a flow of 1.1 liters / min for a representative sample. These sample lines are about 250 feet long and require a 30-minute purge prior to sampling.
- e. . Steam Generator Blowdown Samples - plant procedures require a 1, flow of 1.2 liters / min for a representative sample. These lines are about 350 feet long and require a 45-minute purge prior to sampling.
- f. Main Condensate Samples - plant procedures require a flow of 1.2 liters / min for a representative sample. These sample lines are about 250 feet long and require a 30-minute purge prior to sampling.
- g. Main Feed Samples - plant procedures require a flow of 1.2 i liters / min for a representative sample. These samnle lines are about 250 feet long and require a 30-minute purge prior to sampling.
179
TABLE 6.1 RATED FLOWS FOR SECONDARY SYSTEM System Flow lbs/hr F_ low gm/sec j Main Steam (total) 9.6(6) 1.2(6) i Main Steam 3.2(6) 4.0(5) One (1) Steam Generator l High Pressure 2.7(6) 3.4(5) ! Drain Main Condensate 6.9(6) 8.7(5) l Main Feed 9.6(6) 1.2(6) Steam Jet Air 15 cfm 7.1(3) Ejectors Fan Exhaust Makeup Rate = 120 gal /hr per 1000 lbs/hr of blowdown High Pressure Drains (condenser bypass) = 28 percent of total flow Condensate System = 72 percent of total flow. TABLE 6.2
\
RATED FLOWS FOR P;11 MARY SYSTEM System ! Flow lbs/hr Flow om/sec Primary Coolant 1.3(7) 1.7(6) Primary Letdown 5.4(4) 6.9(3) Conversion Factor 1 lb/hr = 0.12623 gm/sec 180 l
3 i i 6.4 Sample Types and Procedures Samples obtained at Turkey Point were collected using the procedures outlined in reference 4. This included liquid, resin and gaseous samples:
- a. Grab Samples - 450 ml of the liquid sample stream were collected in a glass bottle containing 9 ml of concentrated hcl and counted.
- b. Resin Samples - a measured volume of liquid sample stream (100 ml/ min) was passed through premeasured volumes of cation and anion
! resins. Total sample volumes varied from 10 to 40 liters. The resins were transferred to plastic vials and counted.
- c. Iodine species samples - a measured volume (at 0.25 cfm) of the gaseous sample stream was passed through a species sample train.
l The sampler was disassembled and the individual species cups l placed in plastic vials and counted.
- d. Noble gas samples - 250 m1 glass bombs were purged (at 0.7 1/ min) with the gaseous sample stream for 10 bomb volumes (about 4 minutes) then sealed. The glass bombs were then counted directly.
- e. 14C 3H Samples - a measured volume (at 100 ml/ min) was passed through a scmple train. The sample train was returned to INEL for processing and counting by liquid scintillation methods (12).
- f. Air ejector and blowdow;. flash tank vent samples - portions of the sample streams were diverted to a sampler (iodine species i' sample train or 250 al glass bomb) through 1/4 inch stainless steel tubing (probe) which had a 90 bend near the sampling
*end. The probe was placed in the center of the exhaust tube ] stream. This was accomplished in the blowdown flash tank vent (which rises about 12 feet above the main steam safety deck) by hanging the probe over the lip of the open vent and moving it toward the center of the vent.
Radionuclide concentrations for all samples (liquid und gaseous) obtained from the secondary system can be found in Appendix B, Tables
; 8.34-B.55.
- 6.5 Results 6.5.1 Plant Chemistry 4
6.5.1.1 Reasons for Chemistry Control In a PWR, the separation between the primary and [ secondary system fluids is at the steam generators. If this separation is breached, then the higher pressure (approximately 2000 psig) reactor , coolant leaks into the lower pressure (approximately 700 psig) secondary 1 181
system fluid. The result is a release of a portion of the radioisotopic inventory, which has accumulated in the reactor coolant from various sources, to the secondary system fluid in the steam generator. Once released to the secondary system, it is possible for a portion of these radioisotopes to be released to the environment via 'iquid and/or gaseous discharges from the secondary system. Other objectives of chemistry control are the following. Chemicals are added to the reactor coolant system in order to reduce corrosion and to aid in reactivity control. Chemicals are added to the secondary waters in order to prevent corrosion and to reduce the fouling of heat transfer surfaces. 6.5.1.2 Reactor Coolant System At Turkey Point hydrogen is added to the reactor coolant system through the chemical and volume control system to prevent corrosion by reducing the oxygen content of the reactor coolant. Lithium hydroxide (LiOH) is a chemical agent added for pH control. Boric acid is added to the reactor coolant to aid in reactivity control of the nuclsar core by absorbing neutrons. A sidestream of reactor coolant is reduced in pressure and passed through filters and ion exchangers in order to remove ionic corrosion products, certain fission products, and insoluble corrosion products. s from fission product iodine are iodide (I-), iodate (IIthasbeendetermined(11)thatthe 2 and organic iodine (CH3 I) and that thermal and radiation reductions occurring in the core yield largely I and some 12 The s species expected in the reactor coolant, therefore, isbecause I~ principal of iodine hydrogen overpressure, alkalinity, and the occurrence of radiation reduction reactions. Decontamination factors (DF's) for the reactor coolant system at Turkey Point (see Section 3.3) indicate that > 99% of the radiciodine in the reactor coolant is iodide (I-). Earlier studies on reactor coolant cleanup systems have also indicited that iodide (I~) is the principal form present in the reactor coolant (11). Tables 6.3 to 6.5 tabulate the rer. tor coolant chemistry conditions for Units #3 and #4 during the leak studies. It should be noted that Unit #3 was near end of core life at thc time of the leak study. 6.5.1.3 Secondary Chemistry
. Previous studies (5,6,11) of radiciodine behavior in secondary system fluids during periods of primary to secondary leakage have yielded variable results. Based on these studies and the reducing chemistry conditions maintained in the secondary system the following radiofodine behavior could be expected.
182
TABLE 6.3 AVERAGE PRIMARY CHEMISTRY - UNIT #3* Period: 11/9-21/77 4 Reactor Pressurizer Limit Coolant Measurement Liquid Limit 4.2-10.5 '7.80 pH 7.90 4.2-10.5 1 Conductivity 7.41 <1-40
<1 -40 7.60 gnhos gen (02 ) <0.005 <0.10 <0.10 <0.005 <0.15 <0.05 Ch oride (Cl-) <0.05 <0.15 <0.15 <0.04 oride(F-) <0.04 <0.15 0-4000 21.62 Boon (B) 21 .76 0-40001 p
j *j Gross 6 y 15 min pCi/mi Gross 8 y 7 day i 1*8(-2) pCi/ml Demin Flow 44.4 GPM 0.22-2.22 0.57 L thium (L1) 0.58 0.22-2.22 25-35 Normal Hydrogen (H2 ) 25-35 Nonnal 19.50 cc/Kg 12.73 15-50 Off-Normal 15-50 Off-Normal Crud
-< 1.0 PM Gross a pCi/ml 1311/133I 6.8(-2) Tritium (H3 )
pCi/ml Chromate (Cr0g) ppm
- FPL analysis.
183
TABLE 6.4, AVERAGE PRIMARY CHEMISTRY - UNIT #4* Period: 1/18-26/78 Reactor Pressurizer Limit Coolant Measuremen t liquid Limit i 4.2-10.5 6.68 pH 6.71 4.2-10.5 Conductivity 13.1 <1-40
<1 -40 12.1 pmhos <0.10 <0.005 On9'" ( 2) <0.005 <0.10 9 pm <0.15 <0.05 Chloride (Cl-) <0.05 <0.15 ppm <0.15 <0.04 Fluoride (F-) <0.04 <0.15 0-4000 51 3 Bomn (B) 51 3 0-4000 p
Gmss B-y 15 min 2.5(-1) pCi/ml Gross s y 7 day 5.5(-3) pCi/ml Demin Flow j 63.4 GPM ; 0.22-2.22 1.10 L thium (Li) 1.11 0.22-2.22 25-35 Nonnal Hydrogen (H2 ) 5.5 25-35 Normal 18*03 cc/Kg 15-50 Off-Nonnal 15-50 Off-Nonnal Crud l
< 1.0 ppm I
Gross a pC1/ml 131I/1331 1.9(-1) Tritium (H3 ) pCf/ml ! Chromate (Cr03 ) l ppm
- FPL analysis 184 l
1 TABLE 6.5 , AVERAGE PRIMARY CHEMISTRY - UNIT #4* Period: 2/3-9/78 Reactor Pressurizer Limit Coolant Measurement Liquid Limit 4.2-10.5 6.67 pH 6.80 4.2-10.5
<1-40 10.8 Conductivity 10.3 <1-40
- unhos i
<0.10 O ngen (02 )
<0.005 0.006 <0.10 ppm
<0.15 <0.05 Ch oride (Cl-) <0.05 <0.15
<0.15 <0.04 Fluoride (F-) <0.04 <0.15 j p 0-4000 450 Bo on (B) 458 0-4000 pp j,9(_j)
Gross 8 y 15 min uCi/ml Gross s y 7 day 1.7(-2) pC1/ml Demin Flow l 62 GPM 0.22-2.22 0.99 Lithium (Li) 0.99 0.22-2.22 25-35 Normal flydrogen (H2 ) 25-35 Nonnal 15-50 Off-Nonnal 17*6 cc/Kg 8*0 15-50 Off-Normal Crud
< 1.0 <0.01 Gross a 4.7(-10) pC1/ml 1311/133I 1.6 (-1 )
Tritium (H3 )
- pCi/ml Chromate (Cr0g)
Ppm
- FPL analysis.
185 1 l i
i
- 1. Radiofodine in the steam generator water probably exists as I". This valence state is favored by the anerobic and alkaline conditions existing in the steam generator as a result of feedwater chemical treatment.
- 2. Radioiodine in the steam probably exists as volatile and non-volatile fractions. The volatile fraction could constitute up to 25% of the total iodine fraction coming over in the steam. Possible iodine species in the volatile fraction are I 2, HOI, 10X and CH 31. The non-volatile fraction could constitute 75% to >95% of the total iodine fraction coming over in the steam. Possible iodine species in the non-volatile ,
fr*ction is I .
- 3. Iodide (I-) passing through the turbines and main condenser may convert to other species of iodine as a result of air in-leakage and moist air. Possible iodine species are I2 and l HOI.
l 4. Iodide entering the air ejectors or gland seal exhausts could convert to other species of iodine and be discharged l to the atmosphere. Possible iodine species are 12 , H0I, and CH3 I. i 5. Iodide entering the secondary system in the steam could deposit on system surfaces as a result of iodine species conversion.
- 6. Iodine returning to the steam generators in the feedwater !
is probably I . This valence state is favored by the addition of hydrazine (N2H4) and other corrosion reducing
, chemicals to the feedwater.
Pressurized water reactors (PWR's) are subjected to a number of system changes which can result in a change of radioisotopic concentrations and/or changes in the secondary system chemistry. These changes could 1-J1uence the behavior of radiciodine in the secondary system. The
! observed changes are:
- 1. Changes of the primary to secondary leak rate. This can result in increases or decreases in the amount of radiciodine entering the secondary system.
- 2. Power level changes usually result in a temporary elevation of radioisotopic concentrations in the reactor coolant.
1 This phenomena known as spiking does result in tenporary
- increases of radiciodine entering the secondary system.
i i l l 1 86
4 i
- 3. Water loss from the system as a result of leaks and/or planned releases from the steam generator to reduce j
the total dissolved solids in the generator water (this
- operation is kncwn as steam generator blowdown) result in the loss of radiofodine from the system.
; 4. Water replacement (this operation is known as makeup) i to the system results in an apparent loss of iodine because the system volume is diluted.
- 5. The main condenser is operated under vacuum to increase turbine efficiency. Leaks in the main condenser result in raw circulating water being introduced into the j
- secondary system waters. This can result in steam generator fouling, pH depression and changes in the i reducing potential of the system.
- 6. During periods of primary to secondary leakage, boric acid (H380 3 ) used in the reactor coolant as a reactivity
, control agent can leak into the secondary system. If the l core is relatively new, then H 803 3 concentrations are high and leakage to the secondary system could result 4 in pH depression. 2 At Turkey Point, ammonium hydroxide (NH40H) is added to Unit #3 ) secondary water in order to maintain system pH between 8.5 to 9.0 in the , steam generators. Kydrazine (N2H4) is added to suppress oxygen. In i Unit #4, morpholine (tetrahydro-1,4-isoxazine) is added to maintain pH between 8.5 to 9.0 in the steam generators. Morpholine is recommended by Westinghouse for plants using sea water cooling systems. Hydrazine ! (N24 H ) is added to suppress oxygen. In addition to these measures, in both Units #3 and #4 steam generator blowdown is utilized to control total dissolved solids, ! thereby reducing the fouling of heat transfer surfaces. 4 During the in-plant measurement study, Unit #3 periodically experienced a high chloride problem due to sea water leaking into the 4 main condenser. l Tables 6.6 through 6.10 represent the average secondary chemistry ! conditions existing during each of the leak study periods. Tables ! 6.6 and 6.9 represent nontypical chemistry days for each unit. Table 6,6 shows the chemistry for Unit #3 at reduced power (50%), and Table 6.3 shows the chemistry for Unit #4 when power was increased from hot ! shutdown to 100%. i
- All secondary chemistry measurements were made by FPL personnel.
i i ( 187 1
TABLE 6.6 AVERAGE SECONDARY CHEMISTRY - UNIT #3 Period: 11/9/77 Steam Generator Main Feed Main Measurement A B C Water Condensate limits pH 7.56 7.74 7.33 8.87 8.89 *SG 8.5-9.0 Feed & Cond. 8.8-9.2 Cation 50.0 50.0 190.0 SG only Conductivity <2.0 pmhos Total 2.90 Feed & Cond. Conductivi ty only <4.0 pmhos Sodium (Na+) 1.74 2.10 3.23 SG only <0.10 ppm chloride (Cl-) 3.75 5.35 14.87 <0.05 SG - Feed - Cond.
<0.15 ppm Silica (SiO2 ) 0.15 0.22 0.25 SG only <1.0 ppm Total Hydroxide 0.22 0.19 0.15 SG only THC(OH-) ppm Annonia 0.16 0.10 0.14 0.34 SG - Feed - Cond.
(NH3 ) ppm Hydrazine 0.005 0.005 0.004 0.008 Feed - Cond. (N2Hg) ppm 0.005 ppm > 02 ppm
)
Frso Hydroxide +0.06 +0.09 +0.01 SG only ! (OH-) ppm <40.05 ppm 1 0xygtn (02 ) <0.005 0.005 SG - Feed - Cond. i Dissolved ppm
<0.005 ppm Iron (Fe+3) ppm <0.01 Fr.ed-Cond. <0.01 ppm Copper (Cu+2) ppm <0.005 Feed-Cond. <0.005 ppm Gross 8-y SG only <mDA Activity (Composite)
Tritium Activity 1.8(-4)pCi/ml SG only <mDA (Composite) 131I Activity SG only (Composite) Tech Spec 3.8 Individual 5.8(-5) 6.0(-5) 2.8(-3) SG only Gross 8-y pCi/ml pCi/ml pCi/ml NHg0H - Unit #3 SG's only Morpholine - Unit #4 188
TABLE 6.6 (cont'd) AVERAGE SECONDARY CHEMISTRY - UNIT #3 Period: 11/9/77 Chemical Treatment Volumes in Liters Unit #3 Unit #4 Hydrazine NH60H Hydrazine Morpholine 3ASG 4ASG 3BSG 4BSG 3CSG 4CSG 3FW 2.0 4FW Total 2.0 Total
- SG = Steam Generator i
1 189
TABLE 6.7 AVERAGE SECONDARY CHEMISTRY - UNIT #3 l Period: 11/14-21/77 ) l Steam Generator Main Feed Main Measurement A B ____C Water Condensate Limits pH 8.8 8.8 8.7 9.1 9.0 SG 8.5-9.0 Feed & Cond. 8.8-9.2 Cation Conductivity 1.0 1.6 1.3 SG only <2.0 pmhos Total 3.3 3.4 Feed & Cond. only Conductivity <4.0 pmhos l Sodium (Na+) 0.02 0.03 0.03 SG only <0.10 ppm Chloride (C1-) 0.12 0.10 0.08 SG Feed-Cond.
<0.15 ppm Silica (S102 ) 0.03 0.06 0.05 SG only <1.0 ppm 1 Total Hydroxide 0.33 0.30 0.29 SG only THC(OH-) ppm )
Ammonia 0.26 0.24 0.21 0.46 SG - Feed - Cond. (NH3 ) ppm Hydrazine 0.04 0.04 0.05 0.02 Feed - Cond. (N2Hg) ppm 0.005 ppm > 02 ppm Fr:e Hydroxide +0.05 +0.03 +0.05 SG only <+0.05 ppm (OH-) ppm Oxygen (02 ) <0.005 0.011 SG - Feed - Cond. Dissolved ppm <0.005 ppm Iron (Fe+3) ppm <0.01 Feed-Cond. <0.01 ppm Copper (Cu+2) ppm <0.005 Feed-Cond.<0.005ph Gross S-y SG only <mDA l Activity (Composite) I Tritium Activity 1.3(-4) pCi/ml SG only <mDA (Composite) 131I Activity SG only (Composite) Tech Spec 5.8 Individual 1.6(-4) 2.1(-4) 2.5(-3) SG only l Gross 8-y pCi/ml pCi/ml pC1/ml ' NHg0H - Unit #3 SG's only ) Morpholine - Unit #4 190 ! 1
TABLE 6.7 (cont'd) AVERAGE SECONDARY CHEMISTRY - UNIT #3 , Period: 11/14-21/77 Chemical Treatment Volumes in Liters Unit #3 Unit #4 Hydrazine NHu0H Hydrazine Morpholine 3ASG 4ASG 3BSG 4BSG 3CSG 4CSG 3FW 10.3 1.74 4FW Total 10.3 1,74 Total l i l 191
TABLE 6.8 AVERAGE SECONDARY CHEMISTRY - UNIT #4 Period: 1/18-24/78 I
, Steam Generator Main Feed Main Measurement A B C Water Condensate Limits pH 8.66 8.78 8.67 9.05 9.05 SG 8.5-9.0 Feed & Cond. 8.8-9.
Cation Conductivity 1.72 0.89 1.06 SG only <2.0 tahos Total 3.00 3.19 Feed & Cond. Conductivity only <4.0 tsnhos i Sodium (Na+) 0.02 0.02 0.02 SG only <0.10 ppm Chloride (Cl-) <0.05 <0.05 <0.05 SG - Feed - Cond.
<0.15 ppm Silica (SiO2) 0.04 0.05 0.06 SG only <1.0 ppm Total Hydroxide 0.37 0.42 0.36 SG only !
i THC(0H-) ppm Ammonia 0.19 0.27 0.20 0.42 SG - Feed - Cond. (NH3 ) ppm I Hydrazine 0.05 0.04 0.05 0.025 Feed-Cond. (N2Hg) ppm 0.005 ppm > 02 ppm Free Hydroxide +0.02 -0.05 -0.05 SG only <+0.05 ppml (OH-) ppm j 0xygen(0) 2 <0.005 <0.005 SG - Feed - Cond. Dissolved ppm <0.005 ppm I Iron (Fe+3) ppm <0.01 Feed-Cond.<0.01pj Copper (Cu+2) <0.005 Feed-Cond. <0.005 [i l Gross 8 y SG only <mDA l Activity (Composite) , Tritium Activity 9.0(-5) pCi/mi SG only <mDA
/ 1mposite) 131I Activit SG only (Composite)y Tech Spec 5.8 Individual 1.5(-4) 2.3(-5) 2.9(-5) SG only Gross 8 y pCi/ml pCi/ml pCi/mi NHg0H - Unit #3 0.80 0.91 0.87 SG's only Morpholine - Unit #4 192
TABLE 6.8 (cont'd) AVERAGE SECONDARY CHEMISTRY - UNIT #4 Period: 1/18-24/78 Chemical Treatment Volumes in Liters Unit #3 Unit #4 Hydrazine NHu0H Hydra zine Morpholine 3ASG 4ASG 3BSG 4BSG 3CSG 4CSG 3FW 4FW 0.62 Total Total 0.62 193 1
i i TABLE 6.9_ , AVERAGE SECONDARY CHEMISTRY - UNIT #4 ; l Period: 1/26/78 l Main I Steam Generator Feed l Measurement A B C Water Condensate Limits ! pH 8.06 8.21 8.07 8.75 8.76 SG 8.5-9.0 Feed & Cond. 8.8-9 Cation Conductivity 2.7 3.5 2.7 SG only <2.0 tehos Total 2.50 2.45 Feed & Cond. ) Conductivity only <4.0 pmhos i Soditsu (Na+) 0.12 0.26 0.15 SGonly<0.10 ppm l
- Chloride (C1-) 0.15 0.30 0.15 0.05 SG - Feed - Cond.
<0.15 ppm )
l 1
\
Silica (SiO2 ) 0.06 0.07 0.07 SG only <1.0 ppm ' Total Hydroxide SG only i THC(OH-) ppm Ammonia SG - Feed - Cond. (NH3 ) ppm Hydrazine (N2Hg) ppm 0.11 Feed - Cond. 0.005 ppm > 02p J Free Hydroxide SGonly<+0.05pp (OH-) ppm l Oxygen (02 ) <0.005 <0.005 SG - Feed - Cond. i
<0.005 ppm + Dissolved ppm Iron (Fe+3) ppm <0.01 Feed-Cond.<0.01q Copper (Cu+2) ppm <0.005 Feed-Cond.<.0.005i Nin Steam 3H 3.2(-4)pCi/ml SG only <m0A Composite i Tritium Activity 4.8(-4)pCi/ml SG only <mDA (Composito) 131I Activity SG only i (Composite) Tech Spec 3.8 Individual 6.5(-3) 1.4(-3) 1.1 (- 3) SG only Gross 8 y pC1/ml pCi/ml pCi/ml NHg0H - Unit #3 SG's only Morpholine - Unit #4 194
TABLE 6.9(cont'd) AVERAGE SECONDARY CHORISTRY - UNIT #4 .' Period: 1/26/78 Chemical Treatment Volumes in Liters Unit #3 Unit #4 Hydrazine NHu0H Hydrazine Morpholine 3ASG 4ASG 38SG 4BSG 3CSG 4CSG 3FW 4FW 1.0 Total Total 1.0 i l l e )
- 195
TABLE 6.1_0_ AVERAGE SECONDARY CHEMISTRY - UNIT #4 Period: 2/3-9/78 l Steam Generator Main Feed Measurement A B C Water Condensate Limits pH 8.30 8.37 8.25 8.98 9.07 SG 8.5-9.0 ; Feed & Cond. 8.8-9.2 Cation Conductivity 1.62 1.69 1.80 SG only <2.0 rmhos ' Total 3.64 3.72 Feed & Cond. , Conductivity only <4.0 pmhos l S:dium(Na+) 0.03 0.03 0.03 SG only <0.10 ppm Chloride (Cl-) 0.06 0.07 0.05 SG - Feed - Cond.
<0.15 ppm Silica (SiO2 ) 0.05 0.06 0.06 SG only <l.0 ppm Total Hydroxide 0.44 0.60 0.57 SG only I
THC(OH-) ppm Ammonia 0.15 0.22 0.17 0.36 SG - Feed - Cond. (NH3 ) ppm Hydrazine 0.043 0.04 0.04 0.024 Feed - Cond. (N2Hg) ppm 0.005 ppm > 02 ppm Free Hydroxide -0.22 -0.17 -0.15 SG only <+0.05 ppm (OH-) ppm Oxygen (0) 2 <0.005 <0.005 SG - Feed - Cond. Dissolved ppm <0.005 ppm .l Iron (Fe+3) ppm <0.01 Feed-Cond. <0.01 p3 Copper (Cu+2) ppm <0.005 Feed-Cond. <0.005 g l Gross B-y SG only <mDA Activity (Composite) Tritium Activity 6.9(-4)pCi/ml SG only <mDA (Composite) l 131I Activity SG only (Composite) Tech Spec 5.8 Individual 1.3(-3) 2.7(-4) 2.8(-4) SG only Gross 8-y pCi/mi pCi/ml pCi/mi j l NHg0H - Unit #3 2.45 2.70 2.71 SG's only i Morpholine - Unit #4 196
TABLE 6.10_(cont'd)
- AVERAGE SECONDARY CHEMISTRY - UNIT #4 Period
- 2/3-9/78 Chemical Treatment Volumes in Liters Unit #3 Unit #4 Hydrazine NH OH Hydrazine Morpholine !
3ASG 4ASG 3BSG 4BSG 3CSG 4CSG 3FW 4FW 2.5 1.58 Total Total 2.5 1.58 l-t 1 197
E I l 1 l i i 6.5.2 Primary-to-Secondary Leak Rates f
- The source of the radioisotopes in the secondary waters is i reactor coolant that leaks into the steam generator. In order to investigate the behavior of radioisotopes in the secondary waters, it
- is necessary to detemine a viable leak rate for reactor coolant coming into the steam generator water. One method of detemining a primary-to-secondary leak rate is to use tritium (3 H) concentrations, because 3H is volatile, has a long i half-life, and does not plate out on the system surfaces. This is the j method used by the plant to detemine primary-to-secondary leak rates. Another method utilizes concentrations of other radionuclides (e.g., radioactive isotopes of iodine, sodium, or cesium). These radionuclides
- can yield valid leak rates only if there is no plate out on system surfaces l and if effects of radioactive decay are considered. An indication of losses (plate out or any other) in the secondary system can be obtained by a comparison of the leak rate values detemined using 3H with corresponding values determined using the other radionuciides.
For each steam generator we can write the activity balance equation for a radionuclide : h=1Cp-bC3g-sCMS + s Cpy - AM CSG t i where A = total activity of the radionuclide in the steam generator j t = time 1 = leak rate (gm/sec)
,b = blowdown rate (gm/sec) s = steaming rate (gm/sec)
A = decay constant (sec-I) for radionuclide M = mass of water in the steam generator (gms) Cp = radionuclide concentration in reactor coolant (pCi/gm) C3g = radionuclide concentration in steam generator bottoms (uCi/gm) Cg3 = radionuclide concentration in main steam (uCi/gm) CFW = radionuclide concentration in feedwater (yCi/gm) i Atequilibriumh=0. Therefore, if we assume equilibrium the activity balance equation can be solved for the leak rate 1: 198
i l [C3g(b + AM) + s(Cg3 - Cpy)] (1) 1= C P i ~ Summing equation (1) for the three steam generators yields [IC3g(b + AM) + s(ICg3 - 3C79)] Total Leak Rate (L) = C P [IC3g(b + AM) + 3s (average Cg3 - Cpg)] (2) L= C P If we assume there are no losses in the secondary system except for the losses due to radioactive decay in the steam generator and due to steam generator blowdown (both of which are accounted for in equation (2)), then Cpy = Average Cg3 and the second tenn of equation (2) = 0. We therefore get IC3g (b + AM) (3) L= C P l Plant leak rates, blowdown rates and steaming rates are usually expressed as gal /hr and lbs/hr. The metric conversions are: 1 gal /hr = 1.05 gm/sec 1 lb/hr = 0.12623 gm/sec The data in Table 6.11 show that, except on 1/26/78 and 2/3/78, the leak rates detennined using iodine, sodium, and cesium compare well with leak rates detennined using 3H. The reason for the differences between the leak rates calculated using 3H and the other radionuclides on 1/26/78 and 2/3/78 is not apparent. In addition, except during periods when the reactor power level changed and equilibritsn cannot be ! assumed, iodine, sodium, cesium, and 3H yielded approximately the same leak rate. This indicates that there is little or no losses (such as l plate out) of iodine, sodium, or cesium in the secondary system. 6.5.3 Relative Iodine Isotopic Age The average isotopic ages relative to the reactor coolant were determined for the samples obtained from the secondary system (e.g., steam generator water, flash tank effluent, main steam, condensate l L j 199 l i . _ _ . _ _ - - -- .
TABLE 6.11 PRIMARY-TO-SECONDARY LEAK RATE (gal /hr) Steam Generator Power Plant Blowdown Level Valve Date Unit 1b/hr % 131I 133I last 24Na 134Cs 137Cs (3H) 11/9/77 3 8.3(3) 50** 3.2 1 0.3 4.6 1 0.4 t 4.4 1 0.7 6.3 0.3 7.4 1 0.9 7.7 11/14/77* 3 1.2(4) 100 6.8 t 0.4* 5.0 1 0.2* t t t t 4.6 11/17/77 3 1.2(4) 100 6.9 0.8 0.5 0.Stt 6.7 0.5 4.7 0.4 5.2 0.5 6.6 0.6 4.5 11/18/77 3 1.2(4) 100 6.7 0.7 5.6 1 0.7 5.5 0.4 4.6 0.5 6.0 0.7 5.8 1 0.6 4.9 11/21/77 3 1.1(4) 90** 5.7 2 0.4 2.1 0.1 4.4 i 0.5 t t t 3.9 1/18/78 4 1.5(4) 100 1.0 1 0.2 1.1 0.2 1.1 0.2 .9 0.2 1.1 0.3 0.7 0.2 1.3 y 1/19/78 4 1.5(4) 100 1.4 1 0.1 1.4 0.1 1.6 0.2 1.3 i 0.2 1.3 0.6 1.3 0.3 1.4 8 1/20/78 4 1.5(4) 100 1.5 i 0.2 1.4 0.2 1.2 0.2 1.3 i 0.2 1.5 0.4 1.2
- 0.3 1.7 1/22/78 4 1.5(4) 100 4.0 1 0.5 4.0 1 0.4 4.5 0.5 3.3 1 0.3 3.4
- 0.5 3.3 1 0.4 3.6 1/23/78 4 1.5(4) 100 4.4 0.5 4.5 0.4 4.5 i 0.4 4.2 i 0.5 3.3 0.6 4.4 0.5 3.0 1/24/78 4 1.5(4) 100 5.7 2 0.7 5.9 0.3 6.1 1 0.5 4.1 i 0.4 3.2 0.4 4.1 1 0.7 5.1 1/25/78 - Unit #4 In Hot Shutdown - Return to power 0530 1/26/78 1/26/78 4 1.1(4) ** 30.0 3.0 32.4 2.8 21.2 1.4 13.5 i 1.1 24.3 t 2.8 26.5 3.0 16.4 AM 0900 1/26/78 4 1.0(4) ** 25.5 2.7 26.1 2.2 18.3 1.2 13.5 1.1 20.1 i 2.2 22.3 1 2.4 15.6 PM 1300 2/3/78 4 1.4(4) ** 37.8 i 3.6 31.2 1 3.0 28.8 2.0 23.8 i 2.6 28.4 3.9 28.7 3.4 16.9
* - Plant values used for calculations.
11/14/77 131I = 4.3 - pC1/gm and 133I = 2.2 4) pCi/gm for steam generator 3C. 11/21/77 181I = 4.4 - ,133I = 1.1(-5) and i{SI = 7.5(-6) pCi/gm for steam generator 3B lasI =.3.5 - pCi/gm for steam generator 3C.
** - Transient dates when a plant power change occurred, t - No data for calculation.
tt - Data is questionable.
l and main feed) using 1311 to 133I ratios. The' method of calculation is shown in Table 6.12. Data for the calculations were taken from the l Appendix tables. The calculations were made in order to determine the i age of iodine in the secondary system and to estimate the relative residence times for iodine in the various components of the secondary system. Analysis of the age data indicated the following. In Unit #3 the age of the water in all three steam generators was about the-same (approximately 20 hours older than reactor coolant). This is to be expected because steam generator 3C had been leaking at a constant rate for almost 3 months and equilibrium conditions should exist. In Unit #4 the ege of the iodine in the 4A steam generator was younger (about 6 hours) than the non-leaking steam generators. The other secondary system components (i.e., steam, air ejectors, condensate, feed) were the same age (about 8.5 hours older than reactor coolant) as the non-leaking steam generators. This is to be expected because the 4A steam generator had been leaking at a steadily increasing rate for a few days (Unit #4 had been leaking for 2 days before the study began) and equilibrium conditions should not exist. However, in Unit #4 all the component ages became olter and the age difference between 4A steam generator and the other system components became smaller as time progressed, indicating that the system would reach an equilibrium in time. The losses of fodine in the secondary system have been attributed to losses in the steam generator as a result of radioactive decay and
, blowdown (see Section 6.5.2). If we assume the average total blowdown rate for Unit #3 is 3.3(4) lbs/hr and for Unit #4 is 4.2(4) lbs/hr (see Table 6.11), the secondary system operating volume (H2 0) is 270,000 gallons (data supplied by FPL), and the rated secondary system steam flow is 9.2(6) lbs/hr (Table 6.1), then the secondary system recirculation time can be calculated to be about 15 minutes. In Unit #3 a complete exchange of the secondary water will occur in about 68 hours and in Unit #4 a complete exchange of water will occur in about 54 hours. If iodine losses.were occurring in the secondary system due to deposition and resuspension, then the average system ages could approach or exceed these system exchange times. The measured system ages indicate that this is not the case. The average system ages of 20 hours older than reactor coolant for Unit'#3 and 8.5 hours older than reactor coolant for Unit #4 point- to steam generator blowdown as the principal source of iodine loss. This reinforces the conclusion that iodine losses in the secondary system occur as a result of radioactive decay and blowdown.
T Analysis of the age data for the blowdown flash tank indicated the following. For both Units #3 and #4 the iodine entering the blow-down flash tank was the same age as the iodine leaving the tank. This means the transit time for iodine in the blowdown flash tank was short, , indicating that the iodine enters and exits the tank in the liquid. The residence time for iodine in the blowdown flash tank would preclude losses due to deposition and/or iodine conversion in the tank. 4 4 201
,-e,. -
, ~ - , -- - , - n-- - - - - , n.- w- ~r .c---, - - - -
l l TABLE 6.12 ISOTOPIC AGE CALCULATIONS The equation used for calculations is:
)
(A131-A133)t ) R=R ge Solving for t R 1 t = in x x .x l Where A =tI" and t 1311 = 8.04 days = 192.96 hours 1/2 1/2 133I = 20.8 hours 131 = j g . ours = 3.59(-3)/ hours A A13 3 = 2 8h rs = 3.33(-2)/ hours R = 1331/1311 ratio in a liquid Rg = 1331/1311 ratio in a source (i.e., primary coolant) t = time interval between Rgand R l l 1 I i L i 202
-e ..--r yy- - -
6.5.4 Blowdown Flash Tank The blowdown from each of the three steam generators for a unit discharges to a common blowdown flash tank (BDFT). Water leaving the blowdown flash tank is routed to the discharge canal or the radwaste system depending on the activity level of the effluent. During the in-plant studies, all blowdown from both secondary units went to the discharge canal . Water vapor " flashing off" in the tank is vented directly to the atmosphere. Samples were taken from each individual steam generator blowdown and the BDFT liquid effluent on both Units #3 and #4. In addition, an iodine species and two noble gas samples were taken from the Unit #4 BDFT vent. Average inlet values were calculated from the steam generator blowdown concentrations for radionuclides entering the BDFT. The average value of the three steam generator blowdown radionuclide concentrations and the total blowdown rates for a given sample day were used 'n this calculation. Table 6.13 tabulates these data for radiciodine for Units
#3 and #4. Table 6.14 tabulates the same data for the other radionuclides detected in Units #3 and #4.
Direct measurements of the vapor flow through the BDFT vent and the liquid effluent flow could not be made. Direct detennination of the release rate (pCi/sec) of radionuclides leaving the BDFT, therefore, cannot be made. It is possible, however, to estimate the fractions of the blowdown that leave the BDFT as vapor and as liquid by assuming that all the 24Na that enters the BDFT leaves it in the liquid effluent (i.e., 24Na is not volatile so no 24 Na leaves via the BDFT vent). The rate at which 24Na enters the BDFT (i.e., 24Na concentration in blowdown times total blowdown rate) then is equal to the rate at which it leaves the BDFT (i.e., 2"Na concentration in BDFT liquid effluent times liquid effluent flow rate), and the liquid effluent flow rate can be calculated. Table 6.15 presents the calcuiated fractions of blowdown water leaving the BDFT in the liquid and vapor phases. Average fractional flows are 56 i 12% 1iquid, 44112% vapor for Unit #3 and 77 + 3% liquid, 2313% vapor for Unit #4. Results obtained from one iodine species and two noble gas samples taken on the Unit #4 BDFT vent are tabulated in Table 6.16. As indicated, no radioiodine was detected, but low concentrations of cobalt and cesium were detected in the iodine species sample and one of the noble gas samples. No radioactive noble gases were detected in the noble gas samples. An upper limit for radiofodine leaving the BDFT vent can be obtained by using the lower limits of the iodine species sample (5.8(-11) pCi/cc for 131 1 and 9.8(-10) vCi/cc for 1331 ), the average total blowdown rate for Unit #4 (4.2(4) lbs/hr), and the average fractional flow through the BDFT vent for Unit #4 (23%). Resulting upper limits are 7.l(-8) uC1/sec for 131 1 and 1.2(-6) pCi/sec for 1331. Comparing these upper limits with the inlet values obtained on the same date as the iodine species sample (Table 6.13)133 indicates that less than 7(-4)% of the 1311 and less than 9(-3)% of the I which enters in the blowdown liquid leaves the BDFT through the vent. 203 l
TABLE 6.13 ) BLOWDOWN FLASH TANK IODINE ACTIVITY Inlet Activity Rate * (uCi/sec) Blowdown Date Unit Rate 1b/hr 131I 133I 11/9/77 3 2.0(d) i.6(-1) 2.9 (-2)
*11/14/77 3 3.7(4) 1.0(-1) 4.7(-1) i 1/18/78 4 4.5(4) 7.4(-3) 1.2(-2) 1/19/78 4 4.5(4) 9.7(-3) 1.4(-2)
, 1/20/78 4 4.5(4) 1.0(-2) 1.3(-2) 1/22/78 4 4.5(4) 2.4(-2) 3.5(-2) 1/23/78 4 4.5(d) 3.1(-2) 4.6(-2) l 1/24/78 4 4.5(4) 3.5(-2) 5.3(-2) ! 1/26/78 4 3.2(4) 1.8 8.9 (-1 ) 1/26/78 4 3.0(4) 1.2(-1) 6.l(-1) 2/3/78 4 4.2(4) 2.0(-1) 3.5(-1) l l Conversion Factor 1 lb/hr = .12623 cc/sec t Calculated using the equation Inlet Activity Rate = A + B + x BD 3 Where A.B.C = Steam Generator Bottom Radionuclide Concentrations (pCi/cc) BD = Total Blowdown Rate (cc/sec) of Steam Generators
* - Plant data used for S.G. Blowdown 'C' .
204
TABLE 6.14 BLOWDOWN FLASH TANK INLET Inlet Activity Rate t (uCi/sec) Date Unit 24Na 54Mn seCo 60Co 65Zn 134Cs 136Cs 137Cs 13eCs 11/9/77 3 6.5(-3) <2.3(-4) <6.0(-4) <5.5(-5) <7.1(-4) 7.3(-3) 1.0(-3) 1.4(-2) 4.0(-2) 3 11/14/77 3 Insufficient samples to calculate these values 1/18/78 5.3(-3)
- 4 1.5(-3) 3.4(-3) 3.0(-3) <4.7(-4) 8.1(-4) 1.7( 4) 1.l(-3) 1/19/78 4 6.3(-3) <6.4(-4) <1.5(-3) <2.7(-3) <1.2(-3) 1.0(-3) <8.0(4) 2.0(-3) 9.1(-2) 1/20/78 4 5.9(-3) 1.6(-4) <4.7(-4) <3.7(-2) <5.4( 4) 2.6(-3) <5.9(-4) 2.3(-3) <2.6(-2) 1/22/78 4 1.3(-2) <2.4(-3 <1.3(-3) 3.8(-4) <7.7(-4) 2.9(-3) <8.6(-4) 5.6(-3) 1.9(-2) g 1/23/78 4 2.0(-2) <2.0(-3) <1.5(-3) <1.0(-3) <1.3(-3) 3.5(-3) <4.7(-4) 8.2(-3) 1.3(-2) 1/24/78 4 2.2(-2) <2.6(-3) <2.2(-3) <1.2(-3) <1.1(-3) 4.4(-3) <1.2(-3) 8.7(-3) 2.0(-2) 1/26/78 4 3.7(-2) <1.1(-2) 7.3(-3) 1.2(-3) <5.1(-4) 5.5(-2) 2.8(-2) 8.3(-2) 7.8(-2) 1/26/78 4 3.6(-2) <8.7(- ) <4.9(-3) 3.1(-3) <1.5(-3) 4.0(-2) 1.9(-2) 6.2(-2) 5.3(-2) 2/3/78 4 8.1(-2) <1.6(-3) 5.4(-4) <1.2'-3) <1.8(-3) 2,.3(-2) <1.1(-3) 4.5(-2) 1.1(-1)
* - Radionuclide not detected.
t - Calculated using equation in footnote to Table 6.12.
TABLE 6.15 FLOW THROUGi BLOWDOWN FLASH TANK 24Na Concentration (uCi/ml) Estimated Fractional Outlet Flows ' Date Unit Inlet Outlet Liouid (%)t y, pop (y)M 11/9/77 3 2.6 0.3(-6) 4.6 1 0.8(-6) 56 12 44 12 1/18/78 4 7.2 i 1.1(-7) 1.0 1 0.2(-6) 72 18 28 18 1/19/78 4 1.1110.09(-6) 1.8 0.3(-6) 62 11 38 11 ! 1/20/78 4 1.0 0.1(-6) <9.2(-7) 1/22/78 4 2.3 0.1 (-6) 3.4 0.2(-6) 68 5 32 5 1/23/78 4 3.4 0. 2(-6) 4.2 1 0.4(-6) 81 9 19 9 1/24/78 4 3.8 0. 2(-6), 5.2 1 0.4(-6) 73 7 27 7 l l 1/26/78 4 9.1 0.4(-6) 1.15 0.08(-5) 79 7 21 7 1/26/78 4 9.5 0.3(-6) 1.07 0.09(-5) 89 i 5 11 5 l 2/3/78 4 1.53 0.04(-5) 2.1 0.l(-5) 7314 27 4 l 24 t % Liquid = "Na Inlet Concentration x 100 Na Outlet Concentration tt % Vapor = 100 - % Liquid i 1 I i .l l 206
TABLE 6.16 BLOWDOWN FLASH TANK VEtU Radionuclide Concentrations (pC1/cc) Date Unit Sample 1311 1321 :33I 1341 1351 134Cs 1370s seCo 60Co 1/19/78 4 Noble Gas <4.9(-8) *
<3.0(-7) *
<3.5(-6) * * *
- 1/20/78 4 *
- Noble Gas <9.7(-8) <2.2(-7)
- 6.4 3.8 7.9 3.8 <6.3(-8) <1.5(-7)
(-8) (-8) 1/20/78 4 Iodine * * *
<5.8(-11) <9.8(-10) 2.2 2.1 5.8 1.8 3.9 1.1 <l.l(-9)
(-10) (-10) (-10)
- Radionuclide not detected.
O i 1
.w.
l l
.i i
l I 6.5.5 Steam Generator Decontamination Factors i l Steam generator decontamination factors (DF's) can be used i to determine information about the operation of a steam generator and i about radionuclides entering the secondary system via the main steam. The steam generator DF is defined as 36 - ' DF = Cg3 l where l I CSG - radionuclide concentration in steam generator water
- Cg3 - radionuclide concentration in main steam.
For any radionuclide the activity in the main steam is made up of j two components - the activity in the vapor and the activity entrained in moisture droplets. Since activity is the product of concentration and mass, we can write l CMS "MS = Cy M y+CgMg where j CMS, Cy , CM radionuclide concentration in main steam, vapor component, moisture droplet component i l MMS' "V' "M - mass of main steam, vapor component, moisture 1 droplet component. Solving for M g/MMS which is the entrainment fraction (moisture carryover fraction) and noting that the concentration in the moisture droplets is equal to concentration in the steam generator water (i.e., CM=CSG), we get i !- C Cy My "M MS "MS Cg %hg3 i I 24 For a non-volatile radionuclide (such as Na),Cy = 0. Hence Mg Cg3 (non-volatile radionuclide) l Entrainment Fraction = MMS "C M (non-volatile radionuclide) i. l 1 j " DF (non-volatile radionuclide) 1 I l i ! 208
4 4 l For a volatile radionuclide (such as radiciodine) Cy / 0 and the partition factor (PF) can be defined
- Cy PF = C 3g Using the Lbove equations we get PF = =
C My (Cg3 Mg3 - Cg M) g SG SG M C g3 M
=
gMS (C - MMS g} V SG ' N MS 1 1
*{E DF (volatile radionuclide) - DF (non-volatile radionuclide)
In a steam generator, the moisture carryover is small (e.g., the Turkey Point steam generators are rated at 0.25%) so My: M g3 Hence i PF : DF (volatile radionuclide) DF (non-volatile radionuclide) In order to evaluate DF's and PF's for the steam generators, the data obtained from the sets of secondary samples (see Figure 6.2 for sample
- points) were analyzed. The initial analysis indicated several inconsistencies.
When the measured radionuclide concentrations in the main steam were used to determine DF's, the results indicated that the steam generators were delivering excessive quantities of entrained moisture to the main steam. For example, the indicated entrainment factors for both units averaged about 1.2% which is greatly in excess of the manufacturer's design value of 0.25%. In addition the average concentrations of the radionuclide; observed (i.e., 24Na, I31 1, 133I, 135I, 134Cs, 137Cs) in the main steam greatly exceeded (by up to about a factor of 10) the corresponding concen-trations in the condensate and main feed. This would imply that lcrge amounts of tur radionuclides were lost somewhere in the secondary system between the main steam and'the condenser. This conclusion is inconsistent
- with the conclusion in section 6.5.2 that no such loss existed. The only conclusion, therefore, that can be drawn is that the main steam samples were biased.
l l At Turkey Point, main steam samples are taken using main steam samples probes illustrated in Figure 6.3 (6). A steam sample is taken by pulling a portion of the main steam through the probe holes and into a cooler where the sample is condensed. This sample is supposed to quantitatively represent the main st am ccmposition. The probe is I 1 designed isokinetically in order to measure the two-phase main steam flow. Samples are condensed and cooled to about 100 F in the sample cooler, j ( 209
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4 i 1 1 REFi.RLIX.L - AEME h t'.'?. P TC 4.1. - 1M 9 - ASill 512. PART 23 (196@ h j g g,, ggn l l . Figure 6,3 _ . _ _ _ S AMPLE PROSE l i MAIN STEAM SAMPLE' PROBE ! 210 I i
There has been some controversy about the quantitativeness of the main steam probe. Some authors have obtained results that indicate the probe to be positively biased by as much as 60% (6). It is believed by some scientists that the steam probe is preferential to entrained moisture simply as a result of its presence in the main steam stream (15). These scientists feel that flow perturbations created in the area of the probe result in the concentration of moisture droplets in the area of the probe. The result of such a moisture concentration would be a bias of the probe for entrained species. It appears that the main steam probes at Turkey Point are biased. Since the assumption made in section 6.5.2 that the radionuclide concentration in the feed equals the average concentration in the main steam appears to be valid, the feed concentration was used with the average steam generator concentration to determine average DF's. Table 6.17 contains the results of these determinations. Using these DF's, average entrainment fractions were found to be <0.5% for Unit #3 and 0.29 + 0.02% for Unit
#4. Average partition factors for iodine and cesium f'or Unit #4 were found to be 9 + 4(-4) and 2.0 + 0.5(-3), respectively. The very low partition factor for iodine in3fcates that there was very little volatile iodine in the secondary system.
1 6.5.6 High Pressure Drains No samples could be obtained from the high pressure (HP) drain system because no adequate sample points were available. An indirect method was therefore used to determine the amount of activity passing through this system. The main feed is made up of two components: 28% of the main feed flow comes from the HP drains and 72% comes from the main condenser. The following equation can be written for the activity passing through the HP drains into the main feed:
^HP = Fgp CHP " IF C p-F C c c where AHP - activity passing through HP drains Cgp, C p , Cc - radionuclide concentrations in HP drains, main feed, and condensate FHP, F 7 , Fc - flow rates through HP drains, main feed, and condensate The fraction of the feed activity that came via the HP drains is 0.72 C c 1, ,
Cp 211 l t
TABLE 6.17 AVERAGE STEAM GENERATOR DF's Average DF Date Uni t 24Na 131I 133I 134Cs 1370s 11/4/77
- 3 >2.6(1) > 6. *5(2 )
>3.9(2) >5.1(1)
- 2.0 0.6(2). * ,
- i 11/14/77 3 3.1 2.0(2)t 11/17/77 3 >3.1(1) >6.6(1) 2.5 2.9(1) >6.1(0) >3.9(1) 1/18/78 4 >3.3 >1.0 >2.1(1 >2.7(0) >2.8(0) 1/19/78 4 >1.7 >7.4 >4.4(1 >2.6(0) 9 8(0) 1/20/78 4 >7.4 >2.2 >4.6(1 6 3(0) >7.7(0) 1/22/78 4 5.3 0.9(2) 3.3 0.1(2) 3.5 0.2(2) 2.5 1 0.6 2) 2.5 0.5(2 1/23/78 4 2.0 0.2(2) 1.83 0.05(2) 1.81 0.05(2) 1.5 0.2 1.8 0.1(2 1/24/78 4 3.2 1 0.3(2) 3.7 0.1 3.8 0.1 2.6 i 0.5 2.5 1 0.3(2 ro 1/26/78 AM 4 >2.7 4.6 0.2 4.8 0.3 1.8 i 2.1 3.8 1 0.8(1 M 1/26/78 PM 4 >6.9 2.4 1 0.1 2.0 0.2 1.2 0.7 >1.5(2) 2/3/78 4 >4.8 2.1 0.7 2.5 0.8(2 >3.9(1) >1.0(2)
Unit #3 Average
- 2.0 3.1 *
- 0.6(2) 2.0(2)
Unit #4 Average tt 3.5 0.3(2) 2.7 0.2(2) 2.7 1 0.2(2) 1.9 1 0.3(2) 2.310.2(2)
- Insufficient data to determine value.
t Plant data tt Only real values obtained on 1/22/78, 1/23/78, 1/24/78, 1/26/78 PM, and 2/3/78 were used to determine average.
Table 6.18 lists the fraction of activity in the main feed due i to the HP drains calculated for Unit #4. The data were insufficient to perform similar calculations for Unit #3. The results in Table 6.18 indicate that an average of 75% of the activity in the feed come from the HP drains even though only 28% of the flow follows that path. This high fraction is not unexpected since the DF data (section 6.5.5) indicates that most of the activity gets into the main stream entrained in moisture droplets and the water removed by the moisture separator enters the HP drain system. The results in Table 6.18 also indicate that the relative amounts of radiosodium, radiocesium, and radiofodine in the HP drains are equal. This is consistent with the conclusion in section 6.5.5 that there is very little volatile iodine in the main steam (i.e., most of the iodine is entrained in moisture droplets). 6.5.7 Main Steam Air Ejector Iodine species samples were obtained from the air ejector vent on several occasions for each unit. Iodine activity data obtained from these samples are given in Table 6.19. The measured iodine release rates were very low (e.g.,1.6(-6) to 8.7(-4) pCi/sec for 1311). Since the main steam samples are assuned to be biased, the main steam / air ejector partition factors cannot be determined. However, the fraction of the iodine in the steam that is released via the air
- ejector can be obtained. Table 6.19 lists the estimated percent of main steam activity released through the air ejector vent. These releases are based on main steam activities determined using measured steam
; generator activities and average steam g?nerator DF's (see section 6.5.5). The average percent of main steam iodine activity released via the air ejector is 1.3 + 0.4(-1)% for Unit #3 and 1.0 + 0.1(-1)% for Unit #4. This small fra'ction of iodine leaving the secondary system via j
the air ejector is consistent with the observation (section 6.5.5) that
, there was very little volatile iodine in the secondary system.
Based on the chemistry of the secondary system (section 6.5.1), the iodine species which could be expected in the air ejector vent are I2, HOI, and CHal (2,11). The iodine species samples taken from the air
- ejector vent indicated that the principal iodine species discharged
, was. organic iodine. The amount of organic iodine discharged averaged i
about 85 percent of the total air ejector discharge for both units. It is therefore concluded that the air ejector vents are discharging less than 1 percent of the iodine activity entering the secondary system and that this iodine discharge is principally organic iodine. 6.5.8 Turbine Gland Seal Exhaust Vent t l The gland seal exhaust vent is another possible discharge i path for iodine. Unfortunately, because of plant location of these exhaust fans, measurements of the gland seal exhaust vents were not 4 made. However, results of another study (7) made at the Point Beach j Nuclear Facility revealed that the gland seal exhaust was discharging 213 , 1 l I
TABLE 6.18 FRACTION OF FEED ACTIVITY COMING FROM HP DRAINS FOR UNIT #4 i Fractional Activity of Radionuclide Date 24Na 131I 133I 134Cs 137Cs 1/22/78 0.6510.11 0.67 0.02 0.65 0.03 0.60 0.23 0.68 0.11 1/23/78 0.85 1 0.02 0.83 i 0.01 0.84 0.01 0.87 0.04 0.83 0.03 1/24/78 0.76 0.05 0.7110.02 0.70 0.02 0.60 0.10 0.74 0.05 1/26/78 AM
- 0.95
- 0.01 0.95 0.01 >0.88 1/26/78 PM 0.76 1 0.03 0.85 i 0.02 0.55 0.34 <0.68 2/3/78 * >0.36 * *
>0.74 '
Averaget 0.75 0.04 0.74 0.01 0.76 i 0.01 0.66 1 0.11 0.75 0.04
- Insufficient data to perform calculation.
t Average does not include data obtained on 1/26/78 AM because of nonequilibrium conditions. t l l l 214
TABLE 6.19 AIR EJECTOR IODINE SPECIES SAMPLE Activity Release Rate Est. % Activity (uCi/sec)* Released Through AEt Date Unit 131I 133I 131I 133I 11/9/77 3 8.7 i 1.1(-4 7.2 0.9(-4) 2.3 0.8(-1 1.6 1.1(-1) 11/14/77 3 6.7 i 0.8(-5 3.5 0.6(-5) 5.1 1.7(-2 9 6(-2) 1/18/78 4 9.2 i 1.0(-6 8.5 i 1.1(-6) 1.6 1 0.2(-1 9.1 1.4(-2) 1/20/78 4 1.6 i 0.5 - 2.5 0.6 - 2.0 0.6 -2 2.4 0.7 -2 1/24/78 4 4.2 1 0.3 - 5.9 i 0.4 - 1.5 0.2 -1 1.4 0.1 -1
- Activity release rate via air ejector determined using equation F
AAE = CAE AE where AAE - activity release rate through air ejector (uCi/sec) CAE - radionuclide concentration in air ejector vent (pCi/cc) FAE - flow rate throigh air ejector vent (cc/sec) t Percent activity in main steam released through air ejector detennined using the equation
% Activity = (AAE/Ag3) x 100 where AAE - activity release rate through air ejector (pCi/sec)
Ag3 - activity transport rate in main steam (pCi/sec) A g3 can be determined from the average radionuclide concentration in steam generator water (C SG
), the average steam generator DF for the radionuclide (F), and the total main steam flow (Fg3):
bS
- SG g3 F /F i
t , 215 ( l _
i 4 iodine at a rate 20 times greater than the air ejector exhaust. If l this factor is applied to the highest air ejector activities (Table 6.19 - column 1) for each unit, then the worst possible case for Unit #3 is - a 4.9% iodine discharge and for Unit #4 a 3.0% iodine discharge. The
- Point Beach study revealed that about 60% of the radioiodine discharged from the gland seal exhaust was particulate iodine and that less than 35% of the iodine discharged was volatile (7). Assuming the activity ,
discharge situation is analogous between Point Beach and Turkey Point i then there should be a total of less than 2% volatile and less than 5% i total iodine being discharged from the air ejector and gland seal exhausts. The high percentage of particulate iodine in the gland seal exhausts indicates a possible deposition mechanism for iodine in the turbine train. 6.6 Conclusions I The following conclusions can be drawn from the data obtained from the secondary systems at Turkey Point: l , 1. The secondary system chemical treatment is successful in keeping iodine in solution. 1 The majority of the radioiodine (> 95%) leaving the
~
2. the steam generator is non-volatile iodine which leaves in the entrained moisture. I
- 3. Iodine exists in the secondary system water as predominently i iodide (I') which is in solution as a result of secondary system chemical treatment.
]
- 4. The principal losses of iodine in the secondary system is due to radioactive decay and steam generator blowdown. i
- 5. Volatile iodine losses are quite small (f 2%).
- 6. Air ejector radioiodine discharges may be related to iodine conversion in the LP turbine or the main condenser.
- 7. Iodine exists in the reactor coolant as predominantly iodide (I").
- 8. Steam generator moisture entrainment fraction is about 0.3%.
- 9. Steam generator iodine partition factor is about 0.001.
216
- 7. WASTE GAS PROCESSING AND CONTAINMENT BUILDING SYSTEMS 7.1 Waste Gas Processing System 7.1.1 System Description During plant operations, waste gases are generated from the following processes:
- 1. Degassing of reactor coolant in the chemical and volume control system (CVCS),
- 2. The addition of hydrogen to the CVCS for corrosion control in the reactor coolant system,
- 3. Displacement of cover gases as liquids accumulate in various t
tanks, e.g., holdup, CVCS, pressurizer, etc. ,
- 4. Miscellaneous equipment vents and relief valves.
Of the above waste-gas sources, a major fraction of the waste is generated by the displacement of cover gases in various CVCS tanks as they fill with liquid. A schematic of the Waste Gas Processing System (WGPS) is shown in Figure 1.2. A more detailed description is shown in Figure 7.1 The WGPS system is common to both Units #3 and #4. It consists of a collection header, two compressors, and six waste gas decay tanks (WGDT). Gas is collected via the waste gas collection header, compressed and then transferred to one of the WGDT's. Once in a WGDT, the waste gas is released to the environment or returned to one of the process tanks as the tank is being emptied. Before the contents of a WGDT may be discharged to the environment, its contents must be sampled and analyzed to determine activity. Depending upon the results of the
, analysis, the tank is discharged at a controlled rate through HEPA filters into the plant stack or isolated for further decay. The designed capacity of the six WGDT's is sufficient to permit 45 days of decay during nonnal plant operation before a WGDT must be released. The average decay time prior to release during the in-plant measurement period at Turkey Point was two days. This was due to the large quantities of gas generated during the frequent reactor startups and shutdowns.
However, this does r.ot imply that a WGDT was released every two days but that the radioactivity was sufficiently low to be released within two days. 7.1.2 Measurement Data and Methods Various types of samples taken of the waste gas decay tanks included: iodine species, particulates, noble gases, I"C, and 3 H The iodine species and particulate samples were taken using the particulate-f odine species sampler (4). Noble gas samples were taken by filling an evacuated 250 cc glass cylinder with sample gas. The particulate, noble gas, and iodine samples were analyzed for radionuclides using the NRC mobile laboratory. 217
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Carbon-14 and 3 H samples were collected by . filling two 150 cc i stainless steel cylinders to approximately 10 psig. These volumes are
<- required to provide the required sensitivity for H analysis. The cylinders were returned to INEL, where the contents were mixed with air to provide oxygen for the catalytic oxidizer. The amount of air added was too small and its activity too low to contribute any significant activity to the sample. The samples were then processed in the same manner as the ventilation samples (4).
Prior to sample collection, sample lines were grged for at least C, 3H and noble gas 30 minutes. The cylinders used for collecting the samples were evacuated and purged with sample gas for a minimum of 5 i minutes before the sample was collected. Collection time for iodine species-particulate samples ranged from 30 minutes to 60 minutes. The results of analysis of waste gas decay tank samples are shown in Appendix B Tables 8.56-B.58. 7.1.3 Results and Discussion I During the measurement period at Turkey Point a total of 10 wastegasdeca{C3 analyzed for 1 tank H,10samples were for iodine taken. Of these 10and species-particulate samples, 7 were 7 for noble gas. The samples were taken over the period 1/31/78 to 5/2/78. Samples were obtained during the Unit i; refueling outage, power operation after refueling, Unit #4 power operation, Unit #4 shutdown for steam generator repairs, and transfer of Unit #3 fuel pit water to the holdup tanks in preparation for repair of the SFP liner.
! In an attempt to increase the data base for calculating extrapolated annual releases from the WGPS, a comparison of the in-plant measurement program analyses and Florida Power and Light (FPL) analyses was made.
Of the 10 iodine species analyses performed by INEL personnel, 8 corresponding FPL analyses were available. The average of the INEL measurements was 1.43 + 0.09(-7) pCi/cr. The average of the eight corresponding FPL analyses was 1.32(-7) pCi/cc with no quoted uncertainty. The quoted uncertainty for the INEL data includes counting statistics only. The
- INEL and FPL results, therefore, agree to within 2 sigma without including systematic errors. Table-7.1 lists the comparison measurements along with the FPL 131 I data associated with the 29 WGDT's released at Turkey Point through May in 1978. Similarily, corresponding FPL and INEL noble gas measurements were available for five waste gas decay tanks.
1 The noble gas comparative values, corrected to the time of the WGDT release, are included in Table 7.2. The INEL noble gas data were taken from Appendix Table B.56. The FPL data were used to calculate 1311 and noble gas extrapolated I annual releases presented in Table 7.4 using the following equation: I i l l 219 t l
TABLE 7.1 131I ANALYSIS OF WASTE GAS DECAY TANKS (FPL and INEL Measurements) Sanple Release pCi/cc WGDT Date; Tire Date; Time at pCi/cc corrected Volumes Released Ci per Release ( Ts To (hrs) at Te to Tp (x 106 cc) Release 78-1 1/28/78; 0850 1/31/78; 1011 74.2 7.0 -8) 5.4 -8) 88 4.7(-6 78-2 1/30/78; 1500 1/30/78; 1320 22.3 2.1 -7) 1.9 -7) 81 1.6(-5 78-3 1/31/78; 1150 2/1/78; 0533 17.7 1.4 -7) 1.3 -7) 88 1.2(-5 1.29(- ) [1] 78-4 2/1/78; 1002 2/3/78; 1433 52.5 1.2 - 9.9 - 90 8.9-6) 78-5 2/1/78; 1030 2/3/78; 1810 55.7 5.4 - 4.4 - 88 3.9 -5) 78-6 2/4/78; 0045 2/5/78; 0205 25.3 6.7 - 6.1 - 90 5.5 -5) 6.35(-7) [1] 78-7 2/4/78; 0400 2/5/78; 0538 25.6 1.1 -7) 1.0(-7 88 8.8 - 78-8 2/5/78; 0815 2/5/78; 1818 10 4.7 - 92 4.2 - y ro 78-9 2/5/78; 2020 2/7/78; 0950 37.5 4.5 - 4.5((-7 3.9 -7 86 3.4 - 78-10 2/6/78; 0835 2/7/78; 1100 26.5 1.6 - 1.5(-7 86 1.3(-5 2.08(- ) [1] 78-11 2/7/70; 0910 2/13/78; 0829 143.4 1.5 - 9.0 -8 83 7.4(-6 78-12 2/13/78; 0820 2/14/78; 0857 24.6 1.4 - 1.3 -6 78 1.0(-5 78-13 2/13/~d; 0955 2/14/78; 1403 28.1 3.2 - 2.9 - 83 2.4(-5 78-14 2/14/78; 1345 2/14/78; 1746 4 1.9 - 1.9 - 88 1.7(-5 78-15 2/14/78; 1418 2/16/78; 1445 48.5 9.6 - 8.1 - 77 6.2 - 78-16 3/9/78; 0955 3/9/78; 1950 10 4.1 - 4.0 -7 89 3.5 - 78-17 4/5/78; 1500 4/11/78; 1243 141.75 1.4 - 8.4 -8 83 7.0 - 78-18 4/11/78; 1343 4/12/78; 0155 12.2 6.0(-8 5.7 - 91 5.2(-6) 5.4 - [1] 78-19 4/15/78; 1400 4/20/78; 1345 119.75 3.6(-8) 2.3 - 78 1.8(-6) 2.86(-8) [1] 78-20 4/20/78; 1400 4/27/78; 0923 163.4 7.4(-8) 4.1(-8) 84 3.5(-6)
TABLE 7.1 (cont'd) 131I ANALYSIS OF WASTE GAS DECAY TANKS (FPL and INEL Measurements) 1311 1311 Sample Release WGDT Date; Time pCi/cc Date; Time at pCi/cc corrected Release # Ts Volumes Released Ci per TR (hrs) at Ts to Tp (x 106 cc) Release 78-21 4/24/78; 1415 4/28/78; 1000 91.75 2.2(-8) 1.6(-8) 93 1.5(-6) 78-22 4/28/78; 1120 5/2/78; 0023 1.68(-8) [1] 85 2.l(-8) 1.5(-8) 86 1.3(-6) 78-23 4/28/78; 1445 5/2/78; 0535 1.55(-8 78-24 5/2/78; 0840 5/2/78; 1423 86.9 9.0(-8) 9.0(-8)) [1] 90 8.1(-6 5.66 8.95(-8) 8.77(-8) 88 7.7(-6 78-25 5/2/78; 0905 5/2/78; 1625 7.33 78-26 5/2/78; 1001 5/4/78; 0645 9.8(-8) 9.5(-8) 89 8.5(-6 44.75 6.8(-8) 5.8(-8) 86 to 5.0(-6) U 78-27 5/4/78; 0840 5/4/78; 1055 5.60(08) [1] 2.25 5.4(-8) 5.4-8) 77 78-28 5/4/78; 1023 5/4/78; 1253 2.5 4.1 -6) 78-29 5/4/78; 1244 5.8(-8) 5.7 -8) 89 5.1-6) 5/9/78; 2115 104.5 5.2(-8) 3.6 -8) 88 3.1 -6) Ave. 2.21(-7) 8.6(7) i [1] INEL analyses Note: INEL Avg. = 1.43 1 0.17(-7) pCf/cc FPL Avg. = 1.32(-7) pCi/cc
TABLE 7.2 NOBLE GAS ANALYSIS DATA 0F WASTE GAS DECAY TANKS j (FPL and INEL Measurements) asgp 131mye 133Xe 133mXe 13sXe R91 ease # (pCi/cc) (uC1/cc) (pCi/cc) (uC1/cc) (pCi/cc) 1:;: - (-4) H:1 *:?: i::: ) 78-3 1.7 - 3.2 - 3.6 - i:':8l 1.5 -4 H: 5 1:': i:!: i:': !: :l H:' U: 1:l{:23 S:1: ?:!: 1::: H:! 78-10 1:!{:i} 1.8 -3 1.61 1.0(-3) 0.08(-3) 5.3 0.6(-4) 1:{:!l 4.8 5.55
-2 i:if:!l 5.7(-4 0.03(-2) 6.1 0.2(-4)
!:l{:1}
1.3 -4 1.55 0.02(-4) 78-11 9.0 - 7.7 - 2.4 - 1.2(-4 78-12 7.0 - 3.8 - 2.6 - 1.2(-6 78-13 1.4 - 5.7 - 3.1 - 1.5 - 78-14 2.3 - 3.9(-4) 2.0 - 1.4 - 78-15 3.2 - 3.0 - 1.5 - 4.9-} 6.3 - H:l' :!{:21 :*{:21 1:1: :!{:'l M{:ij 78-18 2.9(-4) 4.0 - 4.5(-4) 2.6 -4 4.9 1 0.5(-4) 3.8 0.3(-2) 3.6 0.3(-4) 2.1 1 0.2(-4) 78-19 5.8(-4) 4.3(-4) 1.4(-2) 3.7(-5) 1.8(-7) 6.8 1 0.5(-4) 3.610.1(-4) 1.60 0.02 3.1 0.2(-5) 9.6 0.6(-8)
' 2(-4) 4 '(-')
H:!? U{:!) i:!{:!! (-2) H{:ll 78 22 9 k) 2.9(-4) . -4) 5.6(-9) 7.8 1 0.6(-4) 2.07 0.08(-4) 4.16 0.02(-3) 5.0 1 0.2(-10) H:!! 78-25 1:!{: 3.4 - 1:': 1.6 - B: 1.1 - 1.2 - U{:' 6.3 - 78-26 4.0 - 1.7 - 8.6 - 7.9 - 1.5 - 78-27 1.0 - 1.3 - 8.9 - 8.2 - 7.4 - 78-28 3.5(-4 1.3 - 9.0 - 9.0 - 3.2 - 78-29 3.1(-4 1.2 - 5.6 - 2.9 - 4.7(-7) Average 8.76(-4) 2.6(-4) 4.4(-2) 2.3(-3) 1.2(-4)
- INEL Measurements 222
f RWG = CT xA yxR xf 1 x 10-6 C1/ year . where 1 i Ryg = Extrapolated annual release (Ci/ year) C T
= Average nuclide concentration in tank (uCi/cc)
Ay = Average volume released per tank (cc) R f = release frequency (ycar-I) The 183 1 and noble gas average concentrations employed in the equation were taken from Tables 7.1 and 7.2. The average gas volume released per tank was 8.6(7) cc (Table 7.1). The WGDT release frequency used was 64 tanks per year. The value of 64 is based on the number of WGDT's released at Turkey Point in 1977. This value is in good agreement with an extrapolated annual WGDT frequency of 65 obtained from the number of WGDT's released (38) during the seven-month in-plant measurement period at Turkey Point (Table 7.3). Only INEL data were used to extrapolate annual 14C, 3H and particulate releases. The respective average concentrations for these radionuclides can be _obtained from Appendix Table B.56. Again an average relcase l frequency of 64 and an average tank volume of 8.6(7) cc were used. Since the average WGDT radionuclide concentrations in Table 7.4 , include radionuclide concentrations measured during both the refueling and non-refueling intervals, the average release for the combined refueling-non-refueling interval can be obtained (Table 2.5) by con- ! verting the reported C1/y values to Ci/sec and applying a decontamination factor of 100 for particulates for the exhaust ilEPA filters. Likewise, the release rates specific to the refueling (Table 2.8) and non-refueling (Table 2.7) intervals can be obtained (Table 7.5). The radionuclide concentrations used in Table 7.5 were taken from Appendix Table B.56 and Table 7.1. The Ci/ year values were calculated from the above equation; however, the waste gas decay tank annual release frequencies used were
- 33 for refueling and 31 for non-refueling. These values are based on the number of WGDT's released which were associated with the refueling i
and non-refueling intervals during the January to May,1978 interval at Turkey Point (Table 7.1). Releases 78-1 to 78-15 were associated
, with refueling and the balance was during non-refueling. Also it should be remembered that a decontamination factor of 100 must be applied for
- the particulates due to the exhaust HEPA filters.
i
- The iodine species detected in the waste gas decay tanks were: 1, 2 H0I, and organic iodides. Only one sample showed any particulate iodine ,
present (0.1%). The Dredominant species was found to be organic iodide, l averaging 93.7% for l'1I 3 with a range of 89.4% to 96.8% (see Table l 7.6). Other species of I31 1 detected were: 12 (0.8% average, range l 0.1-2.5%) and H0I (5.5% average, range 1.9-9.0%). The data shown above l l l 223
i TABLE 7.3 i TURKEY POINT WASTE GAS TANK RELEASES I DURING IN-PLANT STUDY PERIOD Release Number Start Time Sto'p Time Volume (cm3) i 77-56 11/3/77 - 0200 11/3/77 - 0745 6.9(7) j 77-57 11/3/77 - 1725 11/3/77 - 2315 7.6(7) 77-58 11/4/77 - 0040 11/4/77 - 0600 8.1(7) l 77-59 11/4/77 - 0840 11/4/77 - 0940 8.8(7) 77-60 11/4/77 - 1105 11/4/77 - 1800 9.2(7) ! 77-61 11/4/77 - 2300 11/5/77 - 0100 8.7(7) ! 77-62 11/13/77 - 2200 11/13/77 - 2345 7.7(7) ! 77-63 11/14/77 - 2000 11/14/77 - 2315 8.6(7)
- 77-64 12/9/77 - 1250 12/9;77 - 1420 8.8(7) 78-1 1/31/78 - 0955 1/31/78 - 1028 8.3(7) 78-2 1/31/78 - 1305 1/31/78 - 1335 8.1(7) 78-3 2/1/78 - 0510 2/1/78 - 0555 8.8(7) 78-4 2/3/78 - 1405 2/3/78 - 1500 9.0(7) 78-5 2/3/78 - 1750 2/3/78 - 1830 8.8(7) l 78-6 2/5/78 - 0120 2/5/78 - 0250 9.0(7) l 78-7 2/5/78 - 0515 2/5/78 - 0600 8.8(7) 78-8 2/5/78 - 1755 2/5/78 - 1840 9.2(7)
! 78-9 2/7/78 - 0930 2/7/78 - 1010 8.2(7) l 78-10 2/7/78 - 1040 2/7/78 - 1120 8.6(7) l 78-11 2/13/78 - 0810 2/13/78 - 0845 8.3(7) l 78-12 2/14/78 - 0840 2/14/78 - 0915 7.8(7) 78-13 2/14/78 - 1340 2/14/78 - 1425 8.3(7) i 78-14 2/14/78 - 0527 2/14/78 - 0605 8.8(7) l 78-15 2/16/78 - 1430 2/16/78 - 1500 7.7(7) 78-16 3/9/78 - 1930 3/9/78 - 2010 8.9(7) 78-17 4/11/78 - 1215 4/11/78 - 1310 8.3(7) 78-18 4/12/78 - 0130 4/12/78 - 0220 9.1(7) l 224 I
TABLE 7.3 (cont'd) TURKEY POINT WASTE GAS TANK RELEASES DURING IN-PLANT STUDY PERIOD i Release Number Start Time Stop Time _ Voltane (cm3 ) ] 78-19 4/20/78 - 1325 4/20/78 - 1405 7.8(7) 78-20 4/27/78 - 0845 4/27/78 - 1000 8.9(7) 78-21 4/29/78 - 0930 4/29/78 - 1030 9.3(7) 78-22 5/2/78 - 0005 5/2/78 - 0040 8.6(7) 78-23 5/2/78 - 0520 5/2/78 - 0550 9.0(7) 78-24 5/2/78 - 1355 5/2/78 - 1450 8.8(7) 78-25 5/2/78 - 1550 5/2/78 - 1700 8.9(7) 78-26 5/4/78 - 0615 5/4/78 - 0715 8.6(7) 78-27 5/4/78 - 1030 5/4/78 - 1120 7.7(7) 78-28 5/4/78 - 1230 5/4/78 - 1315 8.9(7) 78-29 5/9/78 - 0850 5/9/78 - 0940 8.8(7) 1 225 E__
TABLE 7.4 EXTRAPOLATED ANNUAL RADIONUCLIDE RELEASES FROM WGDTEI3 ! (Conbined Refueling and Non-refueling) l l C T WG Nuclide (pC1/cc) (Ci/ year) 85Kr 8.8 - '51 4.8 ! 131mXe 2.6 - '5' 1.4 l 133Xe 4.4 - '5' 2.4 133mXe 2.3 - 'S' 1.3 4 13sXe 1.2 - l5 6.6 - ) ' seCo 1.6 - 8.8 - 60Co 5.4 - 2.9 - l 134Cs 6.7 - 3.7 - 1 137Cs 1.6 - 8.8 - 59Fe '2' -- ( I 54Mn l2 14C 3.0(-4) 1.65[3] 3H 4.4(-6) 2.4(-2)[4] 131I 2.2(-7)[5] 1.2(-3) i [1] Based on release frequency of 64 WGDT releases ! per year and avera9e tank volume of 8.6(7). 1 [2] Insufficient data for calculation. I l [3] Includes total 14C as 14C0 ,2 R140, or 14C0. l [4] Includes total 3H as HT, HTO, or RT. [5] Based on FPL and INEL analysis of 29 tanks from . January to May,1978. l I 226
TABLE 7.5 EXTRAPOLATED ANNUAL RADIONUCLIDE RELEASES FROM WGDT's (Refueling and Non-refueling) Refueling EI3 Non-refueling E23 ! C Rgg C R yg T T Nuclide (pCi/cc) (Ci/ year) (uCi/cc) (Ci/ year) seCo 2.4 - 6.8 - 5.3(-11 ) 1.4(-7 60C0 1.4 - 3.9 - 1.0 - 2.7 - 134Cs 3.3 - 9.4 - 1.2 - 3.2 - 137Cs 1. 5 - 4.3 - 1.5 - 4.0 - 14C 3.0(-4)'3' 8.5(-1 ) 3.0 -4 [3' 8.0(-1) 3H 6.4(-6) 4 1.8(-2) 2.9 -6 [4 7.7(-3) 1311 3.5(-7)[5] 9.9(-4) 6.8(-8)[5] 1.8(-4) [1] Based on 33 waste gas decay tanks released per year during refueling. [2] Based on 31 waste gas decay tanks celeased per year during rion-refueling. [3] Includes total 14C as 14C0 ,2 R14C, or 14C0. [4] Includes total 3H as HT, HTO, or RT. [5]- Based on FPL analysis of 29 waste gas decay tanks in Table 7.1. Tanks 78-1 to 78-15 were during refueling and tanks 78-16 to 78-29 were during non-refueling. 1 227
l are averages and ranges for the 10 waste gas decay tanks isolated and sampled over the period 1/31-5/3/78 (see Table 7.6). The shorter-lived 133I was detected in five of the ten samples mentioned above. Although the predominant species was still organic iodide (88.6% average), the i range (56-100%) was much greater than that observed for 131 1 Individual sample species are shown in Appendix B Tabte B.57. 14 The distribution of chemical species for C and 3H is shown in Appendix B Table B.58. The average for the two radionuclides was 93% oxidizable 14C and 43% oxidized 3H t 7.2 Containment Building System I 7.2.1 System Description The containment purge, internal cleanup, and cooling systems are identical for both units. Both have a two-inch continuous vent line to the stack. The continuous vent lines are used for pressure control inside the reactor containments. Based on pressure buildup in containment, FPL personnel estimate the continuous vent line flow rate to De 20-30 f t 3/mir.ute. The containment purge system consists of one supply and one exhaust fan. The rated flow of both the supply and exhaust fans is 35,000 cfm. This flow is equivalent to 1.33 volume changes per hour. The containment purge exhaust duct feeds through a bank of 40 roughing filters (Dustfoe-35) into the plant's main stack (Figure 1.2). Each Dustfoe-35 filter has an f accordian-type fold in the body of the filter and is 24" square by 12" deep. i The containment internal cleanup system consists of three emergency filter units. Each unit, in order of flow, consists of a prefilter, a HEPA filter, and a charcoal filter. Each unit has a rated flow of . 37,500 cfm at 50 psig and 205 F. This system was not operated during the in-plant measurement program at Turkey Point. In fact, it has never been operated, except for efficiency testing. The system is held in standby for emergency use only. No decontamination factor measurements were made on the internal cleanup system. The containment's internal cooling system has a design flow of 137,000 cfm for each of the four fan-coil units. During normal operation, three of the four units are operational. Air flow, in sequence, is through the cooling coils, fans, and discharge header. This system is designed to keep the containment temperatures at or below 120 F during normal operation. 7.2.2 Measurement Methods Samples were taken from containment atmospheres of both l Units #3 and #4 to determine iodine species, 14C, 3H and noble gases. 1 The samples were taken using the plant sample lines. These sample lines i are used by plant personnel to take samples on a weekly basis and also i i I 228
5 f l TABLE 7.6 l WASTE CAS DECAY TANKS I0 DINE SPECIES Radionuclide & number Particulate (%) 12 (%) HOI (%) Organic (%) , of samples radio- T nuclide detected l/2 Avg Rance Ava Rance Ava Rance Avn Rance j 10 1311 8.05 days 0.01 0-0.1 0.8 0.1-2.5 5.5 1.9-9.0 93.7 89.4-96.8 133I 20.8 hrs 0
- 8.8 0-44 4.6 n-22.9 88.6 56-100 ,
5
*
- 0
- 100 ---
1 1351 6.6 hrs 0 0
* - Iodine activity not detected ca this sampler component.
M e i 1
)
i b -~
prior to any release due to a containment purge or vent. The sample lines were purged for at least 10 minutes prior to pulling any sample. A 250 cc glass cylinder was used for noble gas samples; the particulate-iodine sgplqr was used for the iodine species and particulate samples; and the C- N sampler was used for 14C and 3 H samples (4). All samplers
- are described in reference 4. The sampling period ranged from 2 to 24
- hours for the fodine species samples and 18-24 hours for the 14 C and 3H samples.
Five samples were obtained from Unit #3 containment. One sample (11/10/77) was taken just prior to shutdown for refueling, one sample (2/23/78) just after startup following refueling, and three samples (3/15, 4/18, and 5/3/78) during full power operation following refueling. The five samples were taken for various times, ranging from 3 to 20 days, i after a prior purge. The results for Unit #3 containment airborne activities are shown in Appendix B, Tables B.59-B.62. Three samples of Unit #4 containment atmosphere were taken over the period 1/12/78 to 5/10/78. These samples were taken from less than 1 day to as long as 27 days after a containment purge. The data obtained from analysis of these samples are presented in Appendix B, Tables B.63-B.65. l 7.2.3 Results and Discussion 7.2.3.1 Reactor Coolant Effective Radionuclide Inventory Leakage Rates. Effective Partition Factors, and Iodine Species To predict radionuclide releases from the containment building via the gaseous pathway, for a given radionuclide reactor 1 coolant concentration, one must know the radionuclide leakage rate into the containment building, the time since the last containment purge, the radionuclide partition factor, and the performance characteristics of any effluent treatment systems (ETS). These data would allow one to estimate a containment radionuclide gaseous inventory and the subsequent
- radionuclide release rate at the time of a containment purge. However, these data are not easily obtained. In particular, the amount of partitioning (i.e., the gaseous to liquid distribution) that occurs for j the different radionuclides is not known for conditions that exist in the containments. Consequently, actual reactor coolant leakage rates cannot be determined. Instead, effective reactor coolant leakage rates can be determined. Here an effective reactor coolant leakage rate is 1
- defined as the percent of a given radionuclide inventory in the reactor '
l coolant that leaks into the containment building per day and becomes airborne. The utility of effective reactor coolant leakage rates can i be seen when one considers that effective reactor coolant leakage rates, i corrected for the time since last containment purge, yield the airborne radionuclide inventory in containment. 230 l
. At Turkey Foint results from samples taken of the containment
- atmosphere of both units, along with radionuclide reactor coolant con-centrations, were used to calculate effective leakage rates of reactor coolant radionuclide inventories into the containment buildings. When possible, reactor coolant radionuclide concentrations used in the calculations were for the same day the containment atmosphere sample
; was taken. At other times the radionuclide reactor coolant concentration l was interpolated to the time of the containment atmosphere sample. To 1 increase the data base, a comparison of FPL and INEL measurements of the containment atmosphere was made. From the comparison it was concluded
, that measurements by both FPL or INEL personnel yield statistically the
- same results. Therefore, FPL's measurements have been included. Effective reactor coolant radionuclide inventory leakage rates for 3H, radioiodines, l
and noble gases are presented. Also, experimentally derived effective l partition factors for the radiofodines are included. Since the only containment effluent treatment system at Turkey Point consists of j roughing filters, the ETS perfomance was not considered (cf section
- 7.2.3.3).
Depending on the radionuclide half-life, reactor coolant radionuclide
- inventory leakage rates were calculated by one of two methods. For radionuclides with half-lives on the order of a few days the leak rates were calculated using equation (1) and leak rates, based on 3 H, were calculated using equation (2). In both cases it is assumed a steady- '
state reactor
' Cc xV c (1) 1-e-Aat)1/2(tin 2} ,g j CR xV R Cc Ye000%)
(2) At C V LR RR I where C = Concentration of nuclide in containment atmosphere (uCi/cc) c V c
= Free volume of contairment (4.4(10) cc)
At = Time between prior long-tem purge and sample time (days) 1 l t 1/2 = Nuclide . half-life (days) C = Concentration of nuclide in primary coolant (uCi/cc) f R V = Volume of primary coolant (3.8(8) cc) R , LR = Percent radionuclide inventory in primary coolant leakage to containment atmosphere per day (percent per day) l 231 l l
l l
! coolant leak rate exists and that essentially all airborne activity is removed from containment during a long-term containment purge. Based e
on the long-term purge durations, flow rates, and the containment free volume it can be shown that the latter assumption is valid (cf section 7.2.3.3), i.e., greater than 99% of the airborne radioactivity is removed , during a containment long-tenn (>4.0 hours) purge. ' The reason two equations were used to calculate the radionuclide leak rates to containment is due to the time required for the different radionuclide gaseous concentrations to reach equilibrium following a containment purge. For example, in equation (1) equilibrium airborne radionuclide concentrations in containment are required to exist; but, on occasions the time between sampling and the previous purge is not sufficient to allow equilibrium radionuclide concentrations to be established. The term, 1-e-AAt, corrects short-lived radionuclides to a equilibrium 3 conditions. However, due to the relatively long half-life of H, in comparison to times between samples and containment purges i involved, equation (1) is inappropriate to calculate 3H leak rates since large errors in the equilibrium correction tenn (1-e-Aat) can occur. As a result eguation (2), which does not require equilibrium conditions was used for H leak rate calculations. i i The average reactor coolant radionuclide inventory for 3H and the radiofodines for Units #3 and #4 are presented in Table 7.7. Individual leak rates for 3H , radioiodines and the noble gases are shown in Tables 7.8 and 7.9. Leak rates presented in Tables 7.8 and 7.9 are based on INEL measurements (Appendix Tables B.59-B.65). Leak rates listed in Tables 7.16 and 7.17 are based on FPL analyses and are included in the averages in Table 7.7. As indicated, all leak rates were calculated from equations (1) and (2) above. The containment free volume and primary i coolant volume used were respectively 4.4(10) and 3.8(8) cc. In all
; cases the At values used were the time between the prior long-term purge for the respective containment and the sample time. The radioiodine reactor coolant concentrations included radioiodine spike concentrations.
From these data, it can be concluded that on the average 3.4(-2) and 8.7(-4) percent of the 3H and 131 1 reactor coolant inventories leak into the containment buildings, respectively. This should not be interpreted as two different actual reactor coolant leakage rates, but as effective leak rates of the two radionuclides, i.e., the quantity of the two radionuclides present in the reactor coolant that leak to the containment building and is converted to the vapor state. The difference in the effective leak rates for the two radionuclides is due to the different partition factors (gaseous to liquid distributions) for 3 H and radioiodine. Table 7.10I gresents I the effective partition factors (EPF) based on containment I and 3 H atmosphere concentrations. The EPF is l dehned as the ratio of the containment normalized 1311 and 3H concen- ' l trations where the 1311 and 3H containment nonnalized concentrations are airborne radionuclide concentrations divided by the reactor coolant 232 l l
.c._ _
TABLE 7.7 RADIONUCLIDE AVERAGE REACTOR COOLANT EFFECTIVE LEAK RATES TO THE CONTAINMENT BUILDING (percent per day) Total Number of i Nuclide Unit #3 Unit #4 Combined Samples 1311 1.1(-3) 3.8(-4) 8.7(-4)[1] 29 1321 3.8(-4) [2] [3] 3.8(-4) 3 133I 7.8(-4) 4.6(-4) 6.7(-4) [4] 6 135I 5.8(-4) 4.2(-4) 5.3(-4) [4] 6 4 3H 3.4(-2) 1.3(-2) 2.7(-4)[4] 6 [1] The conbined average is a weighted average based on 19 samples from l Unit #3 and 10 samples from Unit #4 which includes FPL data (Tables 7.16 and 7. 7). [2] Based on three samples only. [3] Radionuclide not detected. [4] The combined average is a weighted average based on 4 samples from Unit #3 and 2 samples from Unit #4. [5] Weighted average based on number of samples. i 1 233 I l
i ^ TABLE 7.8 RADIONUCLIDE EFFECTIVE LEAKAGE RATE TO CONTAINMENT BUILDING ' UNIT #3 l 1 l Percent reactor coolant inventory leakage per day ' 2/23/78 EI3 4 Nuclide 11/10/77 3/15/78 4/lR/78 5/3/78 1311 5.4(-3) 1.6(-4) 1.5(-3) 1.5(-3) 1.2(-3)
; 1321 4.0(-4) 2.1(-4) 5.3(-4) 133I 7.4(-4) 5.0(-4) 9.3(-4) 9.5(-4)
I 135I 2.8(-4) 2.1(-4) 9.8(-4) 8.4(-4) asmKr 5.3(-1) 6.l(-1)
- i 133Xe 9.8(-2) 9.0(-1) 13sXe 1.8(-1) 1.5(-1) 4.9(-1) 3H 6.8(-2) 1.3(-3) 2.4(-2)
- 3.0(-2)
* - Insufficient data for leakage rate calculation.
[1] This sample was taken approximately 5 days after startup from refueling, equilibrium in primary coolant or containment not ,i yet reached. i I 234
l TABLE 7.9 , l l RADIONUCLIDE EFFECTIVE LEAKAGE RATE TO CONTAINMENT BUILDING UNIT #4 e Percent reactor coolant inventory leakage per day Nuclide 1/12/78 5/10/78 131I 1.3(-3) 8.9(-5) i 133I 7.4(-4) 1.7(-4) l 135I 6.2(-4) 2.1(-4) 8smKr *
- 133Xe 1.7(1) 1 l
- 133mXe 1.5(1) l 13sye
- 9,4(o) i 3H 9.2(-3) 1.6(-2)
* - Insufficient data for leakage rate l
calculation. i i a 235
TABLE 7.10 CONTAINMENT EFFECTIVE PARTITION FACTORS EI3 3H g 3Hg 1311 g 131I L Uni t Sample Date (uCi/cc) (uC1/cc) (uCi/cc) (uCi/cc) EPF 3 11/10/77 9.9(-7) 1.7(-1 ) 1.5(-8) 2.0(-2) 1.3(-1) 3 2/23/78 1.9(-7) 4.2(-2) 8.0(-10) 5.4(-3) 3.3(-2) 3 3/15/78 7.6(-6) 1.9(-1) 6.8(-9) 1.5(-2) 1.1(-2) 3 5/3/78 9.6(-6) 3.0(-1) 6.0(-9) 7.5(-3) 2.5(-2) 4 1/12/78 3.4(-6) 1.6(-1) 7.3(-9) 7.0(-3) 4.9(-2) 1 4 4/11/78 1.2(-6) 2.1 (-1 ) 2.5(-9) 9.0(-2) 4.8(-3) ) 4 5/10/78 7.4(-6) 2.0(-1) 6.6(-10) 7.5(-3) 2.4(-3) 0 l [1] EPF = l 3HG /3H L where the subscripts G and L refer to the gas and 11ould phases, respectively. 9 236
, _ . .- _ _-_n.-
- i. l l
concentrations. Again, it should be emphasized that the effective partition factors do not represent actual partition factors but represent j the amount of partitioning that occurs between the radioiodine and 3H in the reactor coolant. However, it should be pointed out that the ; EPF can be used to predict radiciodine or tritium containment airborne inventories for given reactor coolant concentration if either the radiciodine or tritium effective leak rates are known. Tables 7.11 and 7.12 present the average iodine species distributions and ranges for the different radioiodines observed in the Unit #3 and Unit #4 containment buildings. These data are taken from Appendiy Tables B.61 and 0.64. Examination of the data in Tables 7.11 and 7.12 indicate a relationship between the half-lives and the fraction of organic iodine in the two containment buildings. This observation is ; consistent with data in other studies (2,3,7), i.e., the shorter the radioiodine half-life, the less highly enriched in organic iodine the ; radioiodine nuclide is. The reason for the exception in these data, high organic 1351 levels in Unit #4, is unknown. Also examination of appendix Tables B.61 and B.64 shows a definite trend in the percent organic iodide and the time since the previous containment purge. For Unit #3 this is illustrated in Figure 7.2, where the percent organic 1811 1 increases with the time since the last containment purge. This same trend was observed in the Unit #4 containment. For the sample collected on 4/11/78, less than one day after a containment purge, the o ganic fractions for 131I and 133I were respectively 35.0 and 30.8 percent. For the samples taken on 1/12/78 and 5/10/78, 27 and 26.5 days after a containment purse, a significant increase in the fraction of organic iodine is seen for both 1311 and 133I. In addition, the fraction of 133I in the organic form did not increase as much as did the organic 1311 fraction. The reason for the inconsistency of the lasI organic to follow this tend is unknown. The fact that radioiodines become more
, organic as the length of time they persist in the cm.ainment atmosphere increases (i.e., longer half-lives and times since last containment purge) supports the conclusion that surfaces in containment, in particular concrete, can play an important role in the conversion of the more reactive iodine species, such as I2 and H0I, into the organic form (13).
7.2.3.2 Containment Purge Frequency As noted in the system description, both containments have 2-inch continuous vent lines for pressure control inside the containment buildings. However, the continuous vent line on Unit #3 had a tendency to become blocked. Alternately, pressure control in Unit #3 containment was maintained by turning on the exhaust fan for approximately five minutes. This mode of operation is hereafter called short-term purging and was characteristic of Unit #3 only. Tables 7.13 and 7.14 present the purge histories of Units #3 and #4 during the in-plant measurements
- at Turkey Point. The short-tenn purges for Unit #3 are indicated in Table 7.13.
1 1 237
_ _ _ - _ -- ._ _ - - _ _ . - = _ . - _ . - - . . l TABLE 7.11 UNIT #3 CONTAINENT ATMOSPHERE IODINE SPECIES AVERAGES Particulate (%) I2 (%) H0I(%) Organic (%) Number t of Sarelos Isotope 1/2 Avg Range Avg Range Avg Rance Ava Range 5 1811 8.05 days 0.7 0.3-2.1 2.0 0.3-4.8 4.0 0.1-8.3 93.2 84.7-98.4 4 132I 2.3 hours 0.0 --- 14.6 0-25.3 20.8 0-73.7 64.4 0-100 5 133I 20.8 hours 1.1 0-3.2 2.1 0-3.6 5.2 0.1-10.8 91.0 86-95.1 4 134I 6.6 hours 2.7 0-8.1 7.3 0-23-8 12.6 0-28.5 77.6 39.7-100 to
TABLE 7.12 UNIT #4 CONTAI,NMENT ATMOSPHERE IODINE SPECIES AVERAGES Particulate (%) 12 (%) HOI (%) Organic (%) Number t of Sarples Isotope 1/2 Av2. Range Avg _ Range Range Avg Range Avo_
~
3 131I 8.05 days 0.5 0.3-1.6 4.7 4.4-9.4 7.2 5.2-54.0 86.5 35.0-86.7 3 133I 20.8 hours 1.2 1.2-2.0 10.7 8.1-13.2 15.1 8.1-58.4 73.2 30.8-82.7 3 135I 6.6 hours 0 --- 9.0 0-17.9 0 --- 91.0 82.1-100 C' Note: The sample of 4/1/78 was not included in average as it was taken less than one day after purging. Equilibrium had not been established. i i .l i
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. 0 m C y g- O Af 5oe _mm~ 13C c0 3mME 1p.? " Ouo e I ,lii j 1' l, [ 'lI I i l l
_ _ . = _ _ . _ . TABLE 7.13 TURKEY POINT UNIT #3 CONTAINMENT PURGES AND VENTS [1] Purge # Start Time Stop Time 77-43 11/7/77 - 1821 11/8/77 - 0235 77-44 11/19/77 - 1355 11/19/77 - 1845 77-45 11/24/77 - 0040 1/21/78 - 1130 78-1 1/25/78 - 2300 2/17/78 - 1912 78-2 2/25/78 - 0600 2/25/78 - 1000 78-3 2/28/78 - 0935 [2] 2/28/78 - 0938 78-4 3/9/78 - 1345 [2] 3/9/78 - 1346 78-5 3/13/78 - 1120 [2] 3/13/78 - 1123 , 78-6 3/15/78 - 1703 [2] 3/16/78 - 1705 78-7 3/21/78 - 2035 3/22/78 - 0804 . l' 78-8 3/31/78 - 0915 3/31/78 - 1315 78-9 4/4/78 - 0532 4/4/78 - 1138 78-10 4/10/78 - 0330 4/10/78 - 0730 78-11 4/12/78 - 0600 4/12/78 - 1442 78-12 4/19/78 - 0600 4/21/78 - 0930 78-13 5/4/78 - 1130 [2] 5/4/78 - 1135 78-14 5/15/78 - 0905 [2] 5/15/78 - 0907 78-15 5/19/78 - 2145 5/20/78 - 0745 [1] Data for measurement program period only. [2] Short-tenn purge. l l l 1 l 241
1 l
- TABLE 7.14 TURKEY POINT UNIT #4 CONTAINENT PURGES AND VENTS [1]
Purge # Start Time Stop Time ! 77-19 11/5/77 - 0520 11/5/77 - 1512 77-20 11/10/77 - 1145 11/10/77 - 1515 l 77-2i 12/15/77 - 0645 12/16/77 - 0840 78-1 1/20/78 - 1025 1/21/78 - 1350 j . 78-2 1/25/78 - 0620 1/25/78 - 1732 78-3 2/14/78 - 1610 [2] 78-4 3/29/78 - 0655 3/29/78 - 1105 78-5 4/11/78 - 0540 4/11/78 - 0940 78-6 4/14/78 - 0600 4/14/78 - 1125 , j 78-7 5/15/78 - 0600 5/15/78 - 1350
- 78-8 5/26/78 - 1150 5/26/78 - 2010
[1] Data for measurenent program period only. , l [2] Approximately 20 days during outage, l I i 242
Using data for the first five months of 1978 (Tables 7.13 and 7.14) the extrapolated annual purge frequencies for Units #3 and #4 are 40 and 19 times,.respectively. The total of 40 for Unit #3 includes 16 short-term purges. The durations of the purges are discussed in the next section. 7.2.3.3 Extrapolated Annual Radionuclide Release The extrapolated annual releases for 3, H 131 1, and I"C and particulates are shown in Table 7.15. The release rates (pCi/sec) for the different radionuclides during the different modes of plant operation (refueling and non-refueling) are presented in section 2. The intent of this section is to present the calculational methods employed and the data used in the calculations. The method used for calculating 131I and 3H releases from a given containment building was identical; however, the method for calculating 3H and 131 1 released from the two containment buildings was different. The differences were to include a factor for continuous venting (2-inch line) of Unit #4 containment and not include it for Unit #3; and to include short-tenn purging for Unit #3 and not include it for Unit #4. In calculating the release from both containments, the leakage of 131I and 3 H into the containment during a long-term purge (>4.0 hrs) was considered. In addition, all airborne activity was assumed to be removed from the containment during a long-tenn purge (>4.0 hrs). This is reasonable if we represent the concentration in containment during purge by the following relationship: A t -Qt/V y=e o where A = Activity remaining at time t t Ag = Activity present at start of purge Q = Containment purge rate (cfm) t = Purge duration (minutes) V = Containment volume (ft3) Using this relationship, A is equal to one percent of Ag after a 3.3 hour (197 minute), purge. t The following are the equations (with respective containment unit denoted) used to calculate the 3H and 1311 annual releases: 243 I
)
4 TABLE 7.15 EXTRAPOLATED ANNUAL CONTAINMENT BUILDING RADIONUCLIDE RELEASES (Ci/ year) Nuclidq Unit #3 Unit #4 Total 131I 4.1(-3) 2.0(-3) 6.1(-3) 3H 5.2(0) 3.9 0) 9.1(0) 14C 9.7(-2) 4.6{,-2) 1.5(-1) i 134Cs 2.1 - 4.7(-5) 7.7 - 137Cs 8.4 - 9.2(-5) 1.9 - seCo 3.2 - [1] 3.2 - 60Co 1.4 - 1.3(-5) 3.0 - 1 54Mn '1' '1 ' sspe '1
, l3
[1] Insufficient data l l 244
Equation U .st P LR dh (1) R Rc y (TOU) 74- + Cc Cy =R p 3, 4 (2) CcYFYdm =R stp 3 (3) Cc Yc O y =R cv 4 (4) R cv + (R p F)=R p a 4 (5) R stp Fy +R p Fp =R 3 where: R y - Reactor coolant volume (3.8(8) cc) R c - Radionuclide concentration in reactor coolant (uCi/cc) (LR)- Effective radionuclide reactor coolant to containment leak rate (%/ day) Pdh - Long-tem purge duration (hrs) C c . Containment radionuclide airborne concentration (vCf/cc) Cy - Containment free volume (4.4(10) cc) R p
- Average release for long-term purge (pCi)
V F
- Short tem purge flow (7.l(8) cc/ min).
Vdm - Short tem purge duration (2.7 min) R stp
- Release for short-term purge (pCi)
V c
- Continuous vent flow (1.0(9) cc/ day) l D y - Continuous vent duration (250 days) cv - Release for continuous vent (uCi)
R F p
- Long-term purge frequency (yr-l)
.F y - Short-tem purging frequency (yr-I) l R a
- Radionuclide extrapolated annual release (pCi) l l 245
l i Tables 7.16 and 7.17 list the appropriate values used in the calcu- l lations for 131 1 For all release calculations a reactor coolant volume of 3.8(8) cc was used for both Units #3 and #4 together with a containment free volume of 4.4(10) cc. Specific to Unit #3 for a short-tem purge, i a flow of 25,000 cfm (7.l(8) cc/ min) and an average duration of 2.7 I minutes was used. The short-tenn and long-tenn purging frequencies were ) 16 and 24 times per year, respectively. Specific to Unit #4, the continuous vent flow was 1.0(9) cc/ day (25 cfm) and the continuous vent duration was assumed to be for 250 days / year. The long-tenn purge frequency used for Unit #4 was 19 times per year. The long-term purge durations used were 7.2 hours for Unit #3 and 8.9 hours for Unit #4. Iodine-131 effective leak rates are from Table 7.7. The average long-tenn purge flows, durations, and frequencies described above for the respective units were used to obtain the 3H results shown in Table 7.18. In addition, the same flows, durations, and frequencies were used for the continuous vent and short-tenn purges of Unit #3 and Unit 3
#4. Tritium effective leak rates were taken from Table 7.7. The H containment airborne concentrations were taken from Appendix B, Tables B.62 and B.65.
i The extrapolated annual release of 14 C aM particulates from Unit #3 was calculated using the following relation. No corrections for leakage of reactor coolant into containment during a purge were made in the
- calculation of 14 C and particulate releases.
[(Cy x Fy) + (Cp x F p)] x C Ave xK=R C where Cy = Average volume of a vent short-tenn purge (cc) l Fy = Short-term frequency (year-I) Cp = Free volume containment (cc) Fp = Long-tenn purge frequency (year-l) Cave = Average concentration of nuclide in containment (uCi/cc) K = Conversion factor (10~0 Ci/pCi) R C
= Extrapolated annual release from ccatainment (Ci/yr)
For Unit #4 calculations, the first tera
- the equation, (Cy x F y), is replaced by (CCP
- ICP). C CP is equal to the average volume released per l
l 246
TABLE 7.16 DATA FOR UNIT #3 CONTAINENT ANNUAL 1311 RELEASES Days Containment Reactor Purge Purge Since Atmosphere Coolant
'urga Start Duration Previous Concentration Leak Rate Concentration Release
# _Date (hrs ) Purge (uCi/cc) (f/ day) (pCi/cc) (pCi )
3-1 During shutdown (no data) 3-2 2/25/78 4.0 8 8.9(-10) 3.3 - 5.4 - 40.0 3-3 2/28/78 3 min. 3 1.4 - 7.8 - 8.0 - 4.2 3-4 3/9/78 1 min. 12 3.4 - 5.3 - 1.0 - 3.4 3-5 3/13/78 3 min. 16 2.5 - 2.3 - 1. 5 - 7.4 3-6 3/16/78 2 min. 19 5(-) 4.2 - 1.5 - 9.9 3-7 3/21/78 11.5 24 3.4 - 3.61 - 1.1 - 154.3 3-8 3/ 31/78 4.0 10 4.0 - 7.0(, - 10- 179.5 3-9 3-10 4/4/78 4/10/78 3.1 4.0 4 3.8 - 3.5 - 1.7I ,- 8-3)) 171.9 6 1.11 - 8 -3) 158.3 3-11 4/1 2/ 78 21 2 1.8 - 1.6(- 7-) 92.8 3-12 4/19/78 3.5 7 3.6 - 1.2I - 7-) 184.9 3-13 5/4/78 5(min.) 14 4.3 - 8. 3 L,- 7.5 -3) 21.5 3-14 5/15/78 2 (min.) 25 7.7 - 1.2(-3 7.5 -3) 15.3 3-15 5/19/78 6.8 30 7.5 - 1.0(-3) 8(-3) 337.7 Av2 rages :1] 7.2 hr [3] 3.8(-9) [4]8.5(-4) 8.5(-3) :5; 165 2] 2.7 min :6 10.2 [1] Long tenn purge average duration. [2] Short tenn purge average duration. [3] Avsrage containmer;t airborne concentration from FPL analyses at time of purge. [4] Calculated by same methcd in section 7.2.3.1 from FPL 1311 measurements and reactor coolant 131I concentrations for day nearest to purge. Data included in combined average for Section 7.2.3.1. [5] Average release rate for long-tenn purge (pCf/ purge). [6] Avorage release rate for short-term purge (pC1/ purge). 247
TABLE 7.17 DATA FOR UNIT #4 CONTAINMENT ANNUAL 1311 RELEASES Days Containment Reactor Purge Purge Since Atmosphere Coolant Sta rt Duration Previous Concentration Leak Rate Concentration Relez P# Date (hrs) Purge (uCi/cc) (%/ day) (vCi/cc) (pC-8-4-1 1/20/78 3.4 26 2.6(-9) 3.2 - 7- 123. 8-4-2 1/25/78 11.2 5 2.1(-9) 8.7 - 7- 96. 8-4-3 2/14/78 24 30 2.0 - 3.1 - 7- 98.f 8-4-4 8/29/78 4.2 20 3.1 - ) 6.4 - 6- 14.1 8-4-5 4/11/78 6.5 13 1. 8 - 3.0 - 9- 84.< 8-4-6 4/14/78 5.4 3 2.1 - 1.0 - 9- 95. 8-4-6 5/16/78 7.8 32 9.3-10) 1.3 - 7.5 3) 45. 8-4-8 5/26/78 8.4 10 2.3 -9) 5.8(-4) 7(-3) 107.i Averages [1] 8.9 hours [2]1.8(-9) [3] 3(-4) 2.8(-2) [4] 83.' [1] Long-Tem purge duration. [2] Average containment airborne concentration from FPL analyses at time of purge. [3] Calculated by same method in Section 7.2.3.1 from FPL 1311 analyses and reactor coolant 131I concentrations for day nearest to purge. Data included in contined average for Section 7.2.3.1. [4] Average release rate for long-tem purge (uti/ purge). 248
TABLE 7.18 , DATA FOR ANNUAL 3H RELEASES FROM CONTAINMENT PURGES Containment Reactor Prior Atmosphere Coolant Purge Concentration Concentration Release Sample Date Unit (dgggl (pC1/cci (vCi/cc) _ _ (pC1) 11/10/77 3 2.7 9.9 - 1.7(-1) 2/23/78 3 4.3 1.9 - 4.2 - 3/15/78 3 19.3 7.6 - 1.9 - 5/3/78 3 12.5 9.6 - 3.0 - Average 4.4(-6) 1.8(-1) [3] 1.7(5) 1/23/78 4 27 3.4 - 1.6 - 4/11/78 4 0.7 1.2 - 2.1 - t 5/10/78 4 26.5 7.4 - 2.0 - Average [1] 4.0(-6) [2]1.9(-1) [3] 1.8(5) [1] Based on Source Term Program measurements. (Appendix B Tables B.62 and B.65). [2] Average reactor coolant concentraticas based on smnples on day nearest to purge. [3] Average released due to long-term purge based on average long-term urge duration of 7.2 hours for Unit #3 and 8.9 hours for Unit #4 p(Tables 7.16 and 7.17). Uses leak rates of 3.4(-2) and 1.3(-2) percent per day for Units #3 and #4 respectively (Table 7.5). I l l 249
day for a continuous venting (1.0(9) cc/ day) and FCP is equal *.o the continuous vent duration (250 days / year). All other parameters are described in the 3H and 1 II annual release discussion. The radionuclide averages used in the calculations are in Table 7.19. 250
TABLE 7.19_ i 14C AND PARTICULATE RADIONOCLIDE AVERAGES , i Unit #3 Unit #4 i Nuclide (uCi/cc) (uCf/cc) ! 134Cs 1.9(-11) 4.3(-11) 137Cs 7.6(-11) 8.3(-11) 58Co 2.9(-12) [1] 6000 1.3(-11) 1.2(-11) l 14C [2] 8.9(-8) 4.2(-8) [1] Radionuclide not detected. ) t [2] Includes total 14C. 0 j f l 1 I i
, 251
, , - - - - - - . - . - , - -,.,r,_. --e... w.,- - , . . w--~w--+rm. -r ----e,,
r l l i
- 8. AUXILIARY BUILDING VENTILATIJN SYSTEf1 i
l 8.1 System Description and Sampling Methods i
- Figure 1.2 presents a schematic diagram of the overall gas waste I
- disposal system at Turkey Point. Figures 2.9 and 8.1 show the auxiliary
; building vantilation system in more detail. The auxiliary building ventilation system exhausts air from both Unit #3 and Unit #4 equipment i rooms and the open areas of the auxiliary building. The exhaust system includes a bank of high efficiency particulate air (HEPA) filters, i consisting of twenty individual fillers each 24" x 24" x 11-1/2". Air is moved through the filter bank by two exhaust fans rated at 40,000 cfm each, discharging to the atmosphere via the plant vent. The system is l' designed to provide a minimum of five air exchanges per hour for each of the equipment rooms and open areas of the building. Supply air to the j auxiliary building enters via two fans (each rated at 13,500 cfm). The 2
balance of supply air is from inleakage. This system is designed to j assure proper direction of air flow for removal of potential airborne radioactivity and adequate heat removal from operating equipment. Additional buildings and systems exhausting to the plant stack include Unit #3 and Unit #4 containment buildings, Unit #4 fuel pit
; area, the new radwaste building, and the waste gas processing system.
l The secondary system off-gas system and the Unit #3 spent fuel pit area discharge to the atmosphere via their own individual vents. t The containment buildings and the Unit #3 fuel pool area exhaust systems as well as the secondary system associated off-gas and the waste 4 gas processing systems are described and discussed in other sections of this report. The Unit #4 fuel pit area exhaust system consists of 4 prefiltcrs, HEPA filters, and a 20,000 cfm rated exhaust fan. The supply air to the Unit #4 spent fuel pit area enters via two fans with t a combined rated capacity of 3000 cfm. The balance of the supply dir is supplied by inleakage. The new radwaste building ventilation exhaust system is rated at 15,000 cfm via two exhaust fans. The supply air to j this building is provided entirely by inleakage. Figure 8.1 also shows sample points and types of samplers employed 4 at each location. Specific plant areas whose ventilation exhaust feeds t each sample point are listed in Table 8.1. Also, included in Table 3.1 are the design duct flows at the sampler locations. Design flows were taken from Figure B.16 in Appendix B. As expected, the plant vent and auxiliary building duct flow rates j changed with different fan alignments, i.e., with the number and/or with , particular fans feeding the plant vent. To assess the significance of changes due to different fan a' , ments, two sets of fic,w rate measurements were made at each long-term sa.Oi ig location using a aaltiple-injection, i 252
o k l 4
-\ J d' s w, Yfp a a a ir .
nw. u r
~
UNIT 4 {' f[-, l CONTAINVENT ;
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i nocu rm j l s co'a O
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i r-N / Of5h O !jt ' a
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- li ! ]g i A l_ ' , l "j; t' 4
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It LT- [49 DRU H F'I N G i AREA lL U ROO?A [0" n o.j n E ^"E" f; {
- 4 I O (
y \0 o nw
,- c-- _ _
3,,,, {4,A E'i X. ) m
] R COM P. COOLING AREA 4B 4A- ,
~
( 40 C C.H. A . ) o ~4 ~J i [ ) o ( 4 C C.C.H.X. I 253
\
b
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1
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Oi I ( h CONTA J'AE 4T
- n. VALv 6
--- -B ORlC ACID l l T R A.J S F :_' H j c a'v 7 5 N'f?tdS .l .. '. *;- ..-
i I pu ; .
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VALV E ROOM NNj m ,_. m _. .
. 7, , ' g/ _.
( (' e , RO g - 4_G , 1 I ' \ F. P. h 4 J' ' O C! I3 g O THac,1No "- o u _d o urC,j_; B (.r , U{ l s , ,, g , x,, MO NI T OR m 3,g s .) 1 t t' i py
~
T K. ROvu j I '~ ' Lu 30 Lg a- , [. !r u t t. ; j usH B A lOOM
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O il I n c ota --- A7tA i t_i z us O l
' l l -
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,, i3A 30 C O M P. COOLING t,REA j
' i g, i,b -
goc.CHX. )
- i 2J 3C O l l L_.L_J k d
g- (3C C.C. H. X. ) .9 _u - O l Figure 8.1 Auxiliary Building Ventilation System 4 i
\.
I - - . . . - . - .: . .- ,. I AUXILIARY BUILDING SAMPLING S TI0t FEEDS WITH DESIGN DUCT FLOWS Main Stack - 85.000 cfm
- 1. Auxiliary building ventilation system i
- 2. Containment source
- 3. Unit #4 fuel pit area
- 4. New radwaste building
- 5. Waste gas decay tanks (Units #3 and (4)
Station 1 - 4,570 cfm
- 1. Hot machine shop
- 2. Unit #4 electrical penetration room Station 2 - 5,375 cfm
- 1. Unit f4 residual heat exchanger room
- 2. Unit #4 pipeways
- 3. Unit #4 non-regenerative heat exch3nger room
- 4. Unit #4 seal water heat exchanger room
- 5. Unit #4 charging pump room
- 6. Unit #4 sample room
- 7. Unit #4 pipe and valve room i I 8. Unit #4 containment spray pump room i Station 3 - 7.550 cfm
- 1. Unit #3 charging p ep room
- 2. Monitor ta'nk room
- 3. Ion exchange room (Units #3 and #4) j
- 4. Unit #4 residual heat renoval pump rooms
- 5. Safety injection pmp room
- 6. Deborating demineralizer room (Units #3 and #4) .
- 7. Solid waste druming room Station 4 10,855 cfm
- 1. Boric acid tank room
- 2. Unit #3 residual heat exchanger room
- 3. Unit #3 non-regenerative heat exchanger room
- 4. Unit #3 seal heat exchanger room
- 5. Unit #3 pipe and valve room
- 6. Unit #3 containment spray pump room
- 7. Concentrate holding tank room Residual heat exchanger rooms Gas stripper rooms (Units #3 and #4)
I8. 9.
- 10. Unit #3 sample room
- 11. Unit #3 Pipeways
- 12. Basement tank room
- 13. Basement p sp room
- 14. Boric acid evaporator package rooms (Units #3 and #4) 254
I , TABLE 8.1 (cont'd) l AUXILIARY BUILDING SAMPLING STATION FEEDS WITH DESIGN DUCT FLOWS l l [
- Station 4A - 5505 cfm includes items 1-8 t
Station 5 - 14.310 cfm
- 1. Unit #3 electrical penetration room
- 2. Holdup tank rooms
- 3. Waste gas decay tank rooms Unit #3 fuel pit area - 20,000 cfm
- 1. Unit #3 spent fuel pit heat exchanger pump room
- 2. New fuel storage room
- 3. Spent fuel pit room 255
- - . _ . , .ny.. - - y. - + -
2 helium dilution technique (4). Table 8.2 lists the alignments considered and the results of these measurements. Also, included in Table 8.2 are plant stack flow rate measurements obtained by the standard pitot tube traverse method. At locations where both helium and pitot tube measurements
; were made, the results agree within two percent. Ths is excellent agreement and indicates valid sampling locations. Duct flows used for short-tenn samplers (see Figure 2.9) were calculated from the ratio between design and measured duct flow rates at downstream sample locations. The flows employed were 205 cfm for the gas stripper and 150 cfm for the Unit #3 sampling room.
As indicated in Figure 2.9, the sampling system installed at each sample point was a low volume iodine species sampler and/or a 14C and 3H sampl er. The samplers are described in reference 4. t 8.2 Measurement Data The radionuclide release rates (vCi/sec) for the auxiliary building ventilation system and Unit #3 spent fuel pit are presented in Appendix B Tables B.66-B.74. These tables contain the following information: Isotopic Concentration Sample Location Measurement Period Data Presented for Stations 1-5 11/10/77-6/1/78 Pa rticul ate, 1311 species Station 4A 1/20/78-6/1/78 Parti cul ate, 131I species Unit #3 fuel pit a"ea 12/8/78-6/1/78 Particulate, 1311 species 1"C and 3H Unit #3 primary 4/3/78-5/18/78 Pa rti cul ate, 1311 species sample room Boric acid evaporator 5/18-6/1/78 Pa rticul ate, 1311 species
- gas stripper room Main Stack 11/10/77-6/1/78 Particulate, 1311 species, 14C, 3H.
The stack sampler release data are presented in Appendix B, Table B.75. Particulate, "C 3H , and 1311 release rates are given. Gaseous iodine
- species data are presented in Appendix B, Tables B.76-B.85. In addition to radionuclide release rates, upper limit values, based on experimental limits of detection are presented for a selected list of nuclides. Table B.86 in Appendix B presents a summary of sample station data for 131 1 l
l 8.3 Results and Discussion i It is the' intent of this section to present the basis for the major conclusions based upon studies of gas systems at Turkey Point as well as ! 256 l I l
TABLE 8.2 VENTILATION DUCT FLOW MEASUREMENTS Exhaust Fans in UseEI3 ' Measured Flows Auxiliary Containment (cfm) Building Purge Main Stack [2] (40,000 (35,000 Pitot Tube Station Station Station Station Station Station cfm each) _cfm each) Measurement Main Stack #1 #2 #3 #4 #4A #5 1 58,000 [3] 5900 9900 7200 16,200 13,300 5900 2 68,500 69.300 6800 12,800 8300 21,300 17,800 6500 1 1 90.100 88,500 6000 10.300 7300 16,80r. 14,000 5700 m 1 2 [] 95,500 5200 7800 6300 10,000 12.200 5000 [1] All measurenents made with #4 fuel pit exhaust fan and both "new" radwaste building exhaust fans operating. [2] Plant measured flow rate. [3] No plant measurement made. Note: Unit 3 spent fuel pit area duct flow is 20,000 cfm.
4 i i to present the results of additional measurements. The following topics, 2 which are pertinent to the analysis of the auxiliary building sampler data, are presented:
- (1) Normalized 1311 and 3H Release Rates.
)< (2) 131I Source Term and Annual Releases. , (3) Particulate Source Term and Annual Releases (4) Tritium and I"C Annual Releases (5) Decontamination Factors for HEPA Filters t (6) Effective Reactor Coolant Leakage Rates, Partition Factors and 131 I Species. (7) Stack Release Rates of Selected Beta-Emitting Radionuclides. (8) Stack Release Rates of 133Xe. 8.3.1 Normalized 1311 and 3H Release Rates As a means of comparing radioactive effluents between different A PWR's, it is advantageous to nomalize the radionuclide release rates to their respective reactor coolant concentrations (5). The normalization formula is: i I G N= where N = normalized release rate (uCi/sec)/(pCi/gm) ; Ig = radionuclide gaseous release rate (pCi/sec). Ig = average radionuclide concentration in reactor coolant (uCi/gm). j l I During a reactor power transient the 131I reactor coolant concen- 1 i tration becomes elevated (i.e., spikes). The 3H concentration is unaffected by the power level change. Tables 8.3 and 8.4 present the 1311 reactor coolant concentrations excluding and including spikes.
- Also in Table 8.4 are the H3 and 13gI (including spikes) nomalized release rates from the main stack. Table 8.5 presents the corresponding 131 1 normalized release rates (excluding spikes) for the main stack and j ,
4 the auxiliary building. Since the 1311 nomalized release rates i
- including spnes wergalculated using higher 1311 reactor coolant concentrations, the I normalized release rates including spikes l were lower. The average 131 1 normalized main stack release rates i I l
I 258 l
- . _ - -.___-I
TABLE 8.3 AVERAGE 1311 REACTOR COOLANT CONCENTRATIONS (pCf/gm) (Excluding Spikes) Date Unit #3131If1] Unit #41311[1] Averagef2] 11/10-11/21 1.4(-2) 6.0(-3) 1.0(-2) l 11/21-12/4 7.6(-3) 7.4(-3) 7.5(-3) 12/4-12/14 [3] 8.0(-3) 8.0(-3) 12/14-12/28 [3] 6.6(-3) 6.6(-3) 12/28-1/11 [3] 8.2(-3) 8.16(-3) 1/11-1/25 [3] 6.2(-3) 6.5(-3) l 1/25-2/8 [3] 7.2(-3) 7.2(-3) 2/8-2/22 2.8(-3) [4] 6.7(-3) [4] 4.8(-3) [4] 2/22-3/9 1.0(-2) [3] 1.0(-2) 3/9-3/21 1.8(-2) 6.2(-3) 1.2(-2) 3/21-4/3 1.3(-2) 1.0(-2) ' 1.2(-2) 4/3-4/20 7.3(-3) 5.7(-2) 3.2(-2) 4/20-5/4 7.2(-3) 1.0(-2) 8.8(-3) 5/4-5/18 7.4(-3) 7.9(-3) 7.6(-3) 5/18-6/1 7.3(-3) 7.1(-3) 7.2(-3) [1] Reactor coolant average based on average of plant and source term measurements. Spikes are not included. [2] Average of Unit #3 and Unit #4 reactor coolant 1311 concentrations. Spikes are not included. [3]. Refueling Outage [4] Based on analyses during period of Unit #3 startup and Unit #4 shutdown. 259
4 TABLE 8.4 i MAIN STACK 3H AND 1311 NORMALIZED RELEASE RATES [10] AND EFFECTIVE PARTITION COEFFICIENTS Normalized 3H Normalizeo Reactor Coolant Release Rate Reactor Coolant 1311 Release [8]
- Average 3H (pCi/sec)[6,7] Average 1311 pCi/sec g' Date (pCi/ml)f91 pCi/nm _
_(pCi/ml)[3] _ Rate (Eti/on EPF) l 11/10-11/21 7.l(-2) 0.44 1.3(-2) 3.08 7.0 11/21-12/4 1.9(-1) 0.83 1.4(-2 0.74 0.09 12/4-12/14 1.l(-1)[1] 4.4 1.3(-2 [1] 1.34 0.30 12/14-12/28 1.5(-l [1] 4.5 1.2(-2[1] 5.64 1.25 12/28-1/11 1.6(-l ;1] 0.94 8.3(-3)[1] 5.29 5.62 j 1/11-1/25 2.3(-1 _11 0.83 6.5(-3)[1] 0.52 0.62
- 1/25-2/8 1.6(-1)[l 1.5 1.4 -2)[1] 5.47 3.65 2/8-2/22 8.0 - 3.0 [4] [4]
2/22-3/9 1. 3 - [2] 3.5 1.1 -2)[2] 1.75 0.50
, 3/9-3/21 1.4 - 1.6 1.2 -2) 1.19 0.74 t
3/21-4/3 1.5(-1) 1.8 1.2 -2) 0.21 0.12 4/3-4/20 2.2(-1) 7.7 3.2 - 0.15 1.9(-2 4/20-5/4 2.3(-1) 1.4 1.2 - 2.13 1.52 5/4-5/18 1.85(-1) 0.11 1.0 - 0.11 1.0 5/18-6/1 2.0(-l) 5.0 8.6(-3) 0.21 e.2(-2 1 [1] Unit #4 shutdown; Unit #3 reactor coolant concentraticns used.
- [2] Unit #3 shutdown; Unit #4 reactor coolant concentrations used.
[3] Reactor coolant average 1311 concentr tion includes spike concentrations. Also includes plant analyses. [4] Insufficient reactor coolant analyses to calculate average reactor coolant concentration of release rate. [5] Effective Partition Factor (EPF) is the ratio of nomalized 131I release rate to the nomalized tritium release rate. [6] The 3H releases from the Unit f3 spent fuel pit (SFP) are not included in the above nomalized 3H release rates. Including the Unit #3 SFP area i would increase Therefore, the total plant the corrected nomalized nomalized ratesrelease total plant by approximately rate is 2.9517 p(ercent.pCi/sec)/(p [7] One ml of reactor coolant sample weighs one nram, i.e., oran and ml are essentially interchangeable. [8] The 1311 releases from the Unit #3 spent fuel area are not included in the above i nonnalized 1311 release rates. However, since 1311 releases from the Unit #3 are i less than 2 percent of the total plant 1311 releases the main stack 1311 normalized release rates represent total plant normalized rates. j l [9] Average Unit 83 and #4 3H reactor coolant concentrations except as noted. l [10] Normalized release rates are for both Units #3 and #4. I l i a 260 l
l TABLE 8.5 NORMALIZED 131I RELEASE RATES [1] (uCf/sec)/(uCf/gm) Main Date Stack _ Auxiliary Building Sum 11/10-11/21 4.0 4.6 11/21-12/4 1.4 0.6 12/4-12/14 2.2 0.3 12/14-12/28 9.5 5.8 12/28-1/11 5.4 4.4 1/11-1/25 0.5 0.3 1/25-2/8 10'.6 9.9 2/8-2/22 [2] [B] 2/22-3/9 1.9 1.2 3/9-3/21 1.2 1.6 3/21-4/3 0.2 0.1 4/3-4/20 0.15 0.1 4/20-5/4 2.9 3.2 5/4-5/18 0.15 0.14 i 5/18-6/1 0.25 0.05 ) [1] Average reactor coolant concentrations do not include 131I values due to reactor power transients (spiking). [2] Few analyses taken during period of Unit #3 startup and Unit #4 shutdown. l 261
l l including spikes was 2.0 (uCi/sec)/(UC1/gm) while the average 1311 normalized main stack release rates excluding spikes was 2.9 (uCi/sec)/ (uC1/gm). It should be noted that since the Unit #3 SFP 1311 release rates are less than two percent of the total plant 1311 releases the i main stack 1311 nonnalized release rates represent total plant 151 1 normalized release rates from both Units #3 and #4. The average auxiliary building 131 1 normalized release rate excluding spikes was 2.3 (vC1/sec)/ ( UCi/gm) . The average3 H normalized release rate from the main stack was 2.5 (uCi/sec)/(vCi/gm). Since the Unit #3 spent fuel area contributes approximately 17% of the3 H releases from the total plant (both Units
#3 and #4), the total plant H3 nonnalized release rate is 2.95 (UCi/sec)/
(U Ci/ gm) . Comparison of these normalized release rates to those observed in another study (7) indicate that the H3 release rates from Turkey Point are nominally 1.5 times lower and the 1311 release rates are nominally a factor of 40 higher than those at the other PWR's studied. 8.3.2 Iodine-131 Sources and Annual Release Rates Table 2.6 shows the auxiliary building 1311 release rates in uCi/sec from each sample location as percentages of the sum of the i auxiliary building 131I activity (data for each sampling location may be found in Table B.86 of Appendix B). Sample station 4A is not included in the sum as it is a component of station 4, but the percentage of 131I i activity contributed to the auxiliary building total activity by station 4A is shown. Consequently, the sum of the six sample stations is greater than 100 percent. In addition, the stack release rate for 131 I is also
, presented. The difference between the auxiliary building total release rates and the stack release rates can be attributed in part to waste gas decay tank (WGDT) and containment purge releases which occurred during the measurement periods. The dates and times of these releases are listed in Tables 1.3, 7.13 and 7.14. An estimated 20% uncertainty should be applied to the release data due to variations in duct flows, sampler flows, and unknown effects of plant operations, e.g., maintenance on steam e nerators.
, As is apparent from the data presented in Table 2.6, the major part of the auxiliary ouilding release of 131 I originates between stations 4
=nd 4A (see Figure 2.9 and Table 8.1). Based on previous experience (2), a particulate-iodine species sampler was installed in the Unit #3 primary sample room. After detennining that the Unit #3 primary sample room was not the major source of 131 I activity (see Table 8.6), the
~
I ' sampler was moved to the BAE's gas stripper room. The data indicated that neither the Unit #3 primary sample nor the gas stripper room are l the major source of activity in the system. Discussions with FPL personnel ! indicated that the boric acid evaporator rooms might be the primary ! source of activity; but the BAE rooms could not be sampled due to room - access limitations. For the same reason the Unit #3 pipeways could not be sampled. 262
l l TABLE 8.6 STATION #41311 AIRBORNE ACTIVITY RELEASE RATES (uCi/sec) Sample Unit #3 Unit #3 Sample Station Primary Gas Stripper Station Sample Period #4 Sample Room Room #4A 4/3-4/20/78 2.84 1 0.04(-3) 6.6 0.3(-6)[1] 6.5 0.2(-4) 4/20-5/4/78 2.72 0.01(-2) 3.25 0.03(-5) 9.61 0.06(-3) 5/4-5/18/78 2.92 0.07(-4)[2] 1.34 0.07(-6) 1.04 0.02(-3) 5/18-6/1/78 2.9 i 0.l(-4)[2] 8.6 0.8(-7) 3.5 0.l(-4) [1] Short sample period 4/14-4/20/78. [2] Air flow to sampler off for a portion of the sample period. l l 263
l l Figure 8.? presents correlation between the percent operation of the boric acid evaporators during each sample period and the normalized 131 1 release rate for the auxiliary building. The auxiliary building sum was used for comparison purposes, as station #4 accounts for an average of 87% (excluding last two sampling periods) of the auxiliary building release. These data indicate a correlation between operation
- of the #3 boric acid evaporator and changes in the 1311 release rate from the auxiliary building. This correlation was observed in all measurements except two (the periods 4/3-20/78 and 5/4-18/78). It is not known why the release did not increase with the increased usage of the BAE during the 4/3-20/78 sampling period. The lack of correlation during the 5/4-18/78 period, however, can be explained by the lower level of iodine (factors of 10 2-10 3lower) in the feed processed by the BAE. No correlation is indicated between the #4 BAE (i.e., radwaste evaporator) operation and the auxiliary building release. These data support the conclusions that the #3 BAE was the primary 1311 source term at Turkey Point.
i
- In addition to the above, the values of the effective partition factorsQeeTable8.4)supporttheconclusionthattheBAEisthe I I source. The effective partition factor (EPF) is defined primary as the ratio of the nomalized 1311 release rate to the normalized 3H release rate, where a nomalized release rate for a radionuclide is the ,
absolute release divided by the radionuclide concentration in the source of the radionuclide. The source is normally the reactor coolant. The EPF should not be greater than 1.0. An EPF value greater than 1.0 indicates that the source to which the absolute release rates were nomclized (reactor coolant in this case) was not the true total source. When the 1311 release rate is nomalized to the 1311 concentration in the BAE bottoms for the period 1/25-2/8/78 (the only period for which i data exist to make a comparison), the EPF value obtained is lowered from 3.65 to 0.46. An EPF of 0.46 is more reasonable and implies that the BAE is the primary 131 I source at Turkey Point. 131 I release rate for different systems or buildings
.The average
! at Turkey Point during the combined refueling-non-refueling interval, the non-refueling interval, and the refueling interval are presented
- in Tables 2.5, 2.7, and 2.8. For the combined refueling-non-refueling interval the data indicate that the annual stack release is 0.80 Ci/ year, and the auxiliary building annual release is 0.74 Ci/ year. This indicates that approximately 92% of the 131 1 released by Units #3 and #4 via the !
air pathway come from the auxiliary building, since the release from the Unit #3 spent fuel area is insignificant (0.02 Ci/ year). Within a the auxiliary building, station #4 (BAE) contributes 0.65 curies per year (Table 2.6). These releases were obtained by converting the l uCi/sec release rates in Tables 2.5 and 2.6 to C1/ year. Only sample ! periods where no sampler malfunctions occurred were used in the calcu-
! lations. For the refueling and non-refueling intervals (Tables 2.7 and 2.8) similar calculations could be performed, however, the VCi/sec
, release rates should be weighted for the refueling and non-refueling , durations. I i i 264 1
Figure 8.2 Auxiliary Buildina 1311 Release Rates 10 , , , , , , , , , , , , , 9 -
~ ~ ~ ~1 -
1 I O ~ l Auxiliary building lodine-131 release - g (normalized to average R.C. w/o lodine g 7 - spikes -' I ----- Reactor coolant values not well 6 - known, used extrapolated value 7* l - I _.y5 - l B i g 4 - 1 - m i h 3 - 2 - l - 1 - 0 ' ' ' ' ' ' 100 I ' i ' ' ' ' F-~ 3 3 I 3 I 8
'l Unit 3 boric acid evaporator
# 75 - p __ , ,, l l --- Unit 4 boric acid evaporator -
g _ __s - c____ - - - - g 50 _ _ ____ e I
$ 25 -
0 ' ' ' ' ' " L- ' -- 11/10 11/21 12/4 12/14 12/28 1/11 1/25 2/8 2/22 3/9 3/21 4/3 4/20 5/4 5/18 6/1 Sample period INEL-A-10 985 i
l 8.3.3 Particulate Source Terms and Annual Releases for Gaseous Effluents Table 8.7 presents the extrapolated average and annual l releases of radioactive particulates in the gaseous effluents from the auxiliary building and stack. Data used in the calculation of releases from the auxiliary building were obtained by summing contributions from sampling stations #1-#5. Only data obtained when all samplers were operational were used. Also, only measured values (i.e., no detection limit' estimates) were uscA. For the auxiliary building the average release rate was then multiplied by the number of seconds per year, corrected for filtration, and converted to curies to obtain annual releases. Release rates from the stack were handled in a similar manner except no filtration correction was made. The annual release of particulat.e nuclides from the auxiliary building is one percent, or less, of the total stack release for the nuclides shown in Table 8.7. Release paths to the stack that were not monitored in this study were the Unit #3 and Unit #4 containment purge exhaust ducts, the Unit #4 fuel pit area duct, the new radwaste building duct, and the waste gas decay tanks release duct. Using the release rates in Table 2.5, Table 8.8 presents a mass balance of particulate releases for an arbitrary two-week period. In the mass balance it was assumed one containment purge and one waste gas decay tank release occurred during the sample period. From the data in this table, it is apparent that none of the measured releases is the major source of plant particulate release. Therefore, the major release point for particulate radionuclides is either the new radwaste building, which has no exhaust HEPA's, or the Unit #4 fuel handling area, which has HEPA filters on the exhaust vent. For this reason (also see Unit #3 SFP particulate releases, Table 2.5), it appears that the new radwaste building is the likely primary source for the release of particulates. The new radwaste building system consists of a radwaste evaporator and j associated storage tanks area, the test demineralizer (see section 4.1.2), as well as the solid waste drumming area. The radwaste area of ' I the building was not used during the measurement period except to house the test demineralizer system used during the latter stages of the study (April-June,1978). However, the particulate releases thought to be due to the new radwaste building vent were not correlated with the times the test demineralizer system was in use. Particulate releases were observed throughout the measurement period. Consequently, the primary source of particulates at Turkey Point is believed to be the solids drumming area l in the new radwaste building. i 8.3.4 Stack Gaseous 3H and 14C Release Rates The annual 3H and 14r, release (including both oxidized and non-oxidized species) via the vapor pathway from all plant areas exhausting to the main stack is presented in Table 8.9. The extrapolated annual releases do not include releases from the Unit #3 fuel pit area because. this area exhausts separately (see Table 2.5). The average total release 266
TABLE 8.7 EXTRAPOLATED ANNUAL STACK AND AXULIARY BUILDING PARTICULATE RELEASES FOR GASE0US EFFLUENTS Auxiliary Building _ gg Average Release Rate Extrapolated [1] Sum of Stations 1 '; Annual Releases Average Extrapolated Refme Filters After Filters Release Rate Annual Releases Nuclide (uC1/sec) (C1/ year) (uci/sec) (Ci/ year) 134Cs 2.5 (- 5) 7.9(-6) 2.6(-4) 8.2(-3) 137Cs 4.4(- 5) 1.4(-5) 4.2(-4) 1.3(-2) saco 3.5(- 5) 1.1 (- 5) 3.4(-3) 1.1(-1) 60Co 1.9(- 5) S.0(-6) 1.4(-3) 4.4(-2) y 54Mn 2.0(- 6) 6. 3(- 7) 2.0(-4) 6.3(-3) 59Fe [2] [2] [2] [2] [1] Annual and average releases are for the auxiliary building only, decontamination factor assumed to be 100. [2] Not sufficient data to make extrapolation of annual release. i
TABLE 8.8 AVERAGE PARTICULATE RELEASE MASS BALANCE Stack bl3 Auxiliary EI3 Containment [2] Waste Gas [3] Nucl Me (pC1) Buildino (uC1) Purge (pCi) Decay Tanks (pCi) 134Cs 3.1(2) 3.0 (-1) 3.0 1.4(-5) 137Cs 5.1(2) 5.3 (-1) 7.3 3.4(-5) seCo 4.1(3) 4.2 (-1) 1.2(-1) 3.4(-4) 60C0 1.7(3) 2.3 (-1) 1.2(-1) 1.1(-4) 54Mn 7.6(3) 2.0(-2) [4] [4] [1] Average stack and auxiliary building releases are calculated for a two-week sample period [1.21(6) seconds]. Decontamination factor of 100 assumed for auxiliary building HEPA filter banks. [2] Calculated average releases for one containment purge (see Section S.3.3). [3] Calculated average releases for one waste gas decay tank release. Decontamination factor of 100 assumed for HEPA filter banks (seeSection2.3.1). [4] Insufficient data to calculated average release of this nuclide. i I l 268 1
TABLE 8.9 EXTRAPOLATED ANNUAL STACK RELEASES OF GASEOUS 3H AND 14C l Average Extrapolated l Release Rate Annual Release ! Nuclide (uCi/sec) (Ci/ year) 3H 4.3(-1) 13.5 14C 2.2(-1) 6.9 i TABLE 8.10 CONTAINMENT PURGE AND WGDT[13 CONTRIBUTIONS TO ANNUAL STACK RELEASE OF 14C AND 3H Fractional contribution to annual stack release (%) Unit 3 & 4 Containment Waste Gas Decay Nuclide Purge Exhaust Tanks 3H 67.4 0.2 14C 2.1 23.6 [1] Waste gas decay tank release
\
269
rate was multiplied by the number of seconds in a year to obtain annual , l release. Based on calculations described in sections 8.3.1 and 8.3.3, l
- the contribution of 14C and 3H in waste gas decay tank releases and containment purges to the stack release may be obtained. Values are presented in Table 8.10.
The distribution of oxidized and oxidizable 3H and 14C species is presented in Table 8.11. The average release was found to be predominantly oxidized 3 H (91% HTO) and oxidizable 14C (82% R14C, etc.). An estimation of the auxiliary building radionuclide source term was included in Table 2.5. It was assumed that the contribution of 3H and 14C to the annual release from the Unit #3 and Unit #4 spent fuel pit areas was the same. In addition, it was assumed that the new radwaste building contribution to the 3H and 14C source term was insig-
- nificant. With these assumptions, the auxiliary building is the predominant source of 14C, accounting for 64%, of the total plant release. The fuel pit areas combined contribute 10% to the total plant release of 14C.
The balance of the 14C release is via the waste gas processing system. The estimated contribution to the total plant release of 3H from the i auxiliary building is approximately 10%. The predominant source of 14C was the containment buildings (56% of plant total), however, a very significant source of 14C was the two spent fuel areas, which contributed approximately 34% of the total 14C plant releases. 8.3.5 Decontamination Factor for HEPA Filters l I Figure 1.2 presents a schematic drawing showing the filter i banks in the plant gaseous waste system. The only systems which incorporate ! HEPA filter banks are the Unit #4 and Unit #3 fuel pit area exhaust i ducts ad the auxiliary building ventilation duct. Due to a lack of I sample locations, the fuel pit area ducts were not sampled. Therefore, l no decontamination factor measurements could be made for the HEPA filters l in these systems. l Continuous measurements were made of all exhaust streams entering the auxiliary building HEPA filter banks, except the waste gas decay tank releases which were sampled periodically. The stack was sampled on the exhaust side of the auxiliary building HEPA filter banks. In addition to the auxiliary building exhaust, the new radwaste building duct, the Unit #4 fuel pit area duct, and the containment purge exhaust ! duct also vent to the stack sampling point (see Figure 1.2). No samplers ! existed on these ducts. No decontamination factor could be calculated ! for the auxiliary building HEPA filters as the stack releases for the i radionuclides listed in Table 8.8 are approximately 100 times greater
- than the releases from the auxiliary building.
1 i 8.3.6 Effective Reactor Coolant Leakage Rates, Partition Factors, and '3'I Species The distribution of iodine chemical species measured for the auxiliary building and stack species samplers is given in Tables B.76-B.85 of Appendix B. Table 8.12 shows the average iodine species distribution 270
l l TABLE 8.11 STACK GASEOUS 3H AND 14C SPECIES Sample Period HT0 (%)E' 3 CO2 (%)[B] 11/10-11/21/77 -- -- 11/21-12/4/77 87 6 12/4-12/14/77 98 24 12/14-12/28/77 96 23 12/28-1/11/78 90 18 1/11-1/25/78 -- -- 1/25-2/8/78 94 7 2/8-2/22/78 96 9 2/22-3/9/78 99.6 47 3/9-3/21/78 100 22 3/21-4/3/78 98 5 4/3-4/20/78 88 35 4/20-5/4/78 90 7 5/4-5/18/78 62 16 5/18-6/1/78 92 9 [A] 0xidized fraction (") of total tritium in sample. Balance is oxidizable. [0] 0xidized fraction (%) of total 14C in sample. Balance is oxidizable. 271
TABLE 8.12 AVERAGE FRACTIONAL PERCENTAGE FOR 131I SPECIES Particulate, Filter 12 HOI Organic Ave Range Ave Range Ave Range Nisnber of Ave Range Station Sanoles M (%) (%) (%) (%) (%) _ (%) (%) 1.3 ( .05-6. 3) 5.9 (.7-17.2) 27.9 (5.6-54.1) 64.3 (29.8-93.8) Stack 15 2.0 (0-13.3) 8.8 (0-19.7) 27.2 (7.6-46.9) 61.8 (37.2-86.3) Station #1 15 4.0 22.3 (2.6-51.3) 33.1 (0-49.2) 40.6 (20.5-91.7) Station f2 15 ( .1-9. 3) 0.4 (.1 .9) 4.1 (1.4-10.0) 27.0 (8.3-65.3) 68.0 (27.2-90.1) Station #3 13[1] 4.7 (.6-15.1) 19.7 (3.3-50.7) 74.5 (34.5-96.0) Station #4 13 [2] 1.2 (.2-5.1) 6.3 ( .04-13.6) 57.8 (12.6-74.4) 33.0 (14.5-85.1) Station #4A 10 2.1 (.4-7.7) 2.0 6.2 (0-40.4) 14.7 (0-31.6) 77.1 (25-100) g Station #5 15 (0-14.8)
" 1.8 11.8 (0-25.0) 57.9 (0-89.3) 28.5 (12.7-100)
#3 Fuel pit duct 12 [3] (0-3.9) 2.5 9.2 24.3 64.0 Auxiliary Building Avg [4]
D] Sa gle periods 5/4-5/18/78 and 5/18-6.1/78 not included due to possible partial sample line blockage. [2] Sa gle periods 5/4-5/18/78 and 5/18-6/1/78 not included due to closed sample line for part of sample period. [3] Sagle period 12/8-12/12/77 not included due to possible broken sample line for part of the sample period. [4 ? Auxiliary Building Average is average of stations 1-5.
l l l and the measured range for each sampling location and average values for the entire sampling program. As indicated, the predominant chemical form of iodine for both the stack and the auxiliary building is organic. This was the case for all sampling stations except sampling stations #2,
#4A, and the Unit #3 spent fuel pit area duct. The latter two stations were enriched in HOI while station #2 analyses indicated iodine in the i
elemental form. The reason for the presence of H0I in station #4A is unknown. The presence of H0I as the predominant species in the spent fuel storage area is consistent with measurements in another PWR measurement program (7). In that study, 131I of unknown origin was observed in gaseous effluent prior to the plant's first refueling. In this study, similar observations were recorded - the presence of 1311 was detected in the fuel pit area effluent prior to movement of fuel into the fuel pit. Samples of both the spent fuel pit (SFP) and refugling water storage tank (RWST) waters indicated that the source of I I I prior to fuel movement was the RWST water as 131 1 was detected in the RWST water and not in the SFP water. The RWST water is processed through a mixed-bed demineralizer which is common to the SFP and is located adjacent to the SFP. Subseguenttofuelmovement,theSFPwasthesourceof 1311, as i both the 13 I in the SFP water and the gaseous effluent increased. Tritium concentration in the fuel pit water and fuel pit exhaust was relatively constant until the SFP water was removed for SFP liner repairs (see section 5). After the water was removed from the SFP, 3H showed a gradual decrease, and a drastic drop in 131 I concentrations was observed, but the predominant species was still H0I. During the latter intervul, RWST water was still being processed through the demineralizer. The presence of significant quantities of elemental iodine in sampling station #2 is believed to be due to a source in the Unit #4 sampling room. This conclusion is consistent with data from other plants (2,7) where 131 I was detected in the primary system sampling rooms. It is an indication that the source is due to an active leak, i.e.', reactor coolant. However, the predominant iodine species in the Unit #3 sampling room were organic (60%) and HOI (29%). This indicates
! a source further removed in time from the reactor coolant. It further indicates tal that the Unit #3 sampling room was not the primary source of l, since the primary source was principally organic.
Values for effective reactor coolant leakage based on stack release and reactor coolant concentrations are given in Table 8.13. To obtain a leakage rate of reactor coolant into the auxiliary building, the average stack leakage rate (475 lbs/ year) must be multiplied by 0.12. This is based on extrapolated annual 3 H release data, where the auxiliary building contributes twelve percent of the total 3 H stack release. Also included in Table 8.13 are the effective partition factors (EPF's) calculated from the ratio of the 131I and 3H nomalized release rates (Table 8.4). Because unrealistic EPF values (>l.0) are often obtained, it is felt that effective reactor coolant leak rates should be determined from 3H 4 measurements only (see section 8.3.2). a
, 273
- - - , r--y - --r- --e-- - - - - - - - - - - - - - - - - - - - - --
1 i i TABLE 8.13 l MAIN STACK EFFECTIVE REACTOR COOLANT LEAK RATES Leak Rate Stack 3H ' Based on Tritium Release Rate l Date (1bs/ day) [1] EPF [2] (uCi/sec)_ l 11/10-11/21 83 7.0 3.1(-2) [3] l 11/21-12/4 158 0.89 9(-2) 12/4-12/14 848 0.30 4.9(-1) 12/14-12/28 863 1.25 6.8(-1 ) 12/28-1/11 178 5.62 1.5(-1) 1/11-1/25 157 0.62 1.9(-1) [3] 1/25-2/8 286 3.65 2.4(-1) 2/8-2/22 567 ---- 2.4(-1) 2/22-3/9 658 0.50 4.6(-1) 3/9-3/21 299 0.74 2.2(-1) 3/21-4/3 338 0.12 2.7(-1) 4/3-4/20 1462 1.9(-2) 1.7(0) 4/20-5/4 256 1.52 3.1(-1) 5/4-5/18 22 1.0 2.1(-2) 5/18-6/1 952 4.2(-2) 1.0(0) l [1] Leak rates tabulated are calculated from the reactor coolant and stack concentrations. To obtain an average value for leakage of reactor coolant into the auxiliary building the average of the values above must be multiplied by 0.12. [2] EPF is the effective partition factor calculated from the ratio of the nonnalized 1311 to 3H release rate values. [3] 0xidized tritium only, i I 274
Effective leak rates based on 3H listed in Table 8.13 were calculated using the following relations. The 3 H values gaseous and reactor coolant concentrations used are from Tables 8.13 and 8.4. Cg x 60 sec/ min x 1440 min / day = M l M C x 454 gm/lb
=L where:
l Cg = airborne nuclide activity per unit time (uCi/sec) i l M = airborne nuclide activity per day (pCi/ day) C = nuclide activity per unit mass in reactor coolant (uCi/gm) L = primary coolant leakage rate (lbs/ day) The utility of the effective leak rates is discussed in more detail in sections 2 and 7.2.3.1. 8.3.7 Stack Release Rates for Selected Beta-Emitting Nuclides Table 8.14 lists results of beta analysis including data from several sample periods. The samples analyzed were particulate filters from the partkulate-iodine sampler installed on the stack sample point. Two sample periods indicateu increased activity levels for 63Ni and 55 Fe. During the first sample pt.-iod (12/14-28/78) repairs were being made on a Unit #3 steam generator a.M during the second time period (2/22-3/9/78) repairs were being made an Unit #4 steam generator. 8.3.8 Stack Release Rates for l 3Xe Table 8.15 tabulates stack release rate (uCi/sec) for 133Xe. Samples were taken using a cryogenic air sampler (14), which produces a concentrated noble gas sample which was subsequently analyzed by gamma-ray spectrometry and gas chromatography. The release rate for 133Xe was calculated using the relation: 1 (f)xAxD=R M = activity of 133 Xe in sample (uC1) Q = quantity of xenon per can (cm3 ) 275
~
TABLE 8.14 STACK RELEASE RATES FOR SELECTED BETA-EMITTING NUCLIDES (uCi/sec) Sanple Date 90Sr 89Sr 63Ni 55pe 12/4-12/28/77 1.410.3(-6) 3 1(-5) 3.9 0.1(-4) 5.49 0.06(-3) 1/25-2/8/78 1.7 0.3(-6) <3.5(-6) 1.0 0.1(-5) 1.6 0.1(-4) 2/22-3/9/78 1.110.2(-6) 3.0 0.3(-5) 3.05 0.03(-3) 1.50 1 0.02(-2) 5/18-6/1/78 2.310.3(-6) 8 1(-6) 1.2 0.1(-5) 1.0 0.1(-4) l l 1 276
TABLE 8.15 STACK 133Xe RELEASES (4/29-6/7/78) Date 133Xe (uCi/sec) Date 133Xe (pCf/sec) ! 4/29/78 2.4 0.5(2) 5/19/78 [1] 4/30/78 [1] 5/20/78 8 2(1) 5/1/78 [1] 5/21/78 [1] I 5/2/78 3.0 0.4(2) 5/22/78 [1] t 5/3/78 2.0 0.4(2) 5/23/78 [1] , 5/4/78 2.1 0.3(2) 5/24/78 [1] 5/5/78 5.6 0.8(1) 5/25/78 4.9 i 0.6(1) 5/6/78 1.3 0.2(2) 5/26/78 1.0 0.1(3) 5/7/78 [1] 5/27/78 [1] 5/8/78 [1] 5/28/78 2.97 0.04(1) 5/9/78 3.1 0.5(1) 5/29/78 1.22 0.01(2) 4 ^ 5/10/78 [1] 5/30/78 2.6 0.3(2) 5/11/78 [1] 5/ 31 / 7 8 1./ 0.2(1) 5/12/78 6.4 2 0.6(1) 6/1/78 [1] 5/13/78 6.8 0.9(1) 6/2/78 2.6 0.5(2) 5/14/78 [1] 6/3/78 1.4 0.2(2) 5/15/78 [1] 6/4/78 3.3 0.5(2) 5/16/78 4.7 0.5(1) 6/5/78 1.8 0.3(0) 5/17/78 [1] 6/6/78 6 1(0) 5/18/78 [1] l [1] No detectable xenon concentration. l 4 277
A = abundance of xenon in air (0.084 x 10-6 cc Xe/cc Air) D = duct flow rate (cc/sec) R = release rate of stack 133Xe(pCi/sec) 1 1 l l 1 k l 278
1 1 REFERENCES
- 1. " Calculation of Releases cf Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code),"
Office of Standards Development, U. S. Nuclear Regulatory Commission, NUREG-0017, April 1976.*
- 2. N. C. Dyer, J. H. Keller et al., "In-Plant Source Tem Measurements at Fort Calhoun Station - Unit 1," Idaho National Engineering Laboratory, EGAG Idaho, Allied Chemical Corp., NUREG/CR-0140, 1
July 1978.* l 3. N. C. Dyer, J. H. Keller et al., "In-Plant Source Tem Measurements at Zion Station," Idaho National Engineering Laboratory, EG8G Idaho, Allied Chemical Corp., In Press.
- 4. N. C. Dyer, J. H. Keller et al., " Procedures, Sourcc Tem Measurement Program," Idaho National Engineering Laboratory, EG&G Idaho, Allied Chemical Corp. , NUREG-0384, December 1977.*
- 5. C. A. Pelletier, " Study of the Point Beanch PWR Secondary System and Shutdown Primary Spiking," Nuclear Environmental Services Division, Science Applications Inc., EPRI NP-939, August 1978.
- 6. G. W. Hood, " Steam Side Icdine Transport at Point Beanch Unit No. 1 of the Wisconsin Electric Power Company," Westinghouse Electric Corp. , S. O. No. DGRP-41102, October 1973.
- 7. C. A. Pelletier et al., " Sources of Radiciodine at Pressurized i Water Reactors," Nuclear Environmental Services Division, Science Applications, Inc., EPRI Report NP-939, November 1978.
- 8. J. M. Skarpelos and R. S. Gilbert, " Technical Derivation of BWR 1971 Design Basis Radioactive Material Source Tems," General Electric Company, NED0-10871, March 1973.
- 9. "American National Stendard, Source Tem Specification," American Nuclear Society, ANS-18.1, ANSI N237-1976, May 1976.
l
- 10. " Final Safety Analysis Report, Turkey Point Units 3 & 4," Florida Power and Light Company, U. S. Atomic Energy Commission Docket l Nos. 50-250 and 50-251.
- 11. J. A. Martucei, " Iodine Decontamination Factors During PWR Steam Generation and Steam Venting," Combustion Engineering Inc., CENPD-67, September 1973.
- 12. S. W. Duce, J. H. Keller, and J. L. Thompson, "An Atmosphere Tritium and Carbon-14 Monitoring System," Idaho National Engineering i Laboratory, Allied Chemical Corp. NUREG/CR-0386, September 1978.*
279
i i I t
- 13. R. T. Hemphill and C. A. Pelletier, " Surface Effects in the Transport of Airborne Radioiodine at Light Water Nuclear Power Plants,"
Nuclear Environmental Services Division, Science Applications Inc., 4 j EPRI NP-876, September 1978.
; 14. J. H. Keller et al, " Measurement of 133Xe in Air and 1311 in Vegetation ar.d Milk Around the Quad Cities Nuclear Power Station," Idaho ,
National Engineering Laboratory, Allied Chemical Corp.,1978, In Press. l l
- 15. C. A. Pelltcier, Nuclear Environmental Services, Division, Science '
Applications, Inc., Private Correspondence, January 1979.
! 16. O. Jonas, " Turbine Steam Purity," pp.11-27, Combustion, December 1978.
,1 1 l 1 l 1 i l i Available for purchase from the NRC/GP0 Sales Program, U.S, Nuclear Regulatory Comission, Washington, DC 20555, and the National Technical Information Service, Springfield, VA 22161. 280 i
. , - - ,n, -. - , . . - - . - - .,-n. - , . . , -- . - .
APPENDIX A. SAMPLE AND DATA HANDLING A.1 Sample Handling A.l.1 Sample Volumes and General Sampling Procedures The general sampling procedures that were followed have been described in the report " Procedures, Source Term Measurement Program" (4). Brief descriptions that include details specific to Turkey Point (e.g., sample points, purge times, sample volumes, etc.) are included in the appropriate sections of this report. A.l .2 Sample Validation Although it is impossible to prove that a given sample is an actual representation of the liquid vr gas under study, some confidence can be placed on the quality of the sample by showing how reproducible the results are for replicate samples or for samples from a given s); tem taken under different conditions. Validation of liquid and gaseous samples at Turkey Point was handled as described below. A.l.2.1 Liquid Samples During the initial phase of the measurements at Turkey Point, each sample point was investigated to determine whether or not a valid sample could be obtained. This was done by first discussing the sample point with plant personnel and determining the sampling procedures they recommend (e.g., recirculation times for tanks). The sample line was then purged for a time at least as long as recommended by plant personnel, a sample was taken, and the results were compared with expected activities (e.g., with earlier neasurements made by the plant) and with other associated samples. Replicate samples obtained under identical conditions over a period of time were compared in order to give final verification of sample validity. During the course of sampling at Turkey Point, two problem areas were identified. The first involved the measurement of crud radionuclides (e.g., radioisotopes of Co, Mn, Fe, etc.) in reactor coolant. The range in the neesured concentrations for the cruds was found to be much larger than that for soluble radionuclides (e.g., 24Na, iodines, rubidiums, cesiums). Measurements made by FPL also indicated this large range in crud concentrations. In order to investigate this variation in crud concentrations in more detail and determine if it was due to sampling procedures, series of replicate samples were obtained under identical conditions. The sample lines were purged for longer than normal durations (about one-half to one hour). Then without altering any valve settings, samples were obtained separated in time by 1 to 19 minutes (depending on the series). A-1
Six replicate series were taken: (1) two samples from Unit #3 at 0805 and 0806 on 4/9/78; (2) two samples from Unit #3 at 1000 and 1004 on 4/17/78; (3) four samples from Unit #3 at 1006, 1007, 1008, 1009 on 4/25/78;(4) two samples from Unit #4 at 1916 and 1935 on 3/25/78; (5) two samples from Unit #4 at 0903 and 0904 on 4/9/78; and (6) two samples from Unit #4 at 1045 and 1046 on 4/17/78. Results of analyses of these samples are given in Appendix B Tables B.3 and B.4. Comparisons of the replicate samples indicated that the results for the soluble radionuclides (e.g., Na, iodines, rubidiums, cesiums) did not change from sample to sample in a given series. The measured crud concentrations for a series, however, varied from factors of 2 to 10, depending upon the series. These variations were not always in the same direction - sometimes the earlier samples had higher crud concentrations (replicate series 1, 2, and 6) while in other series (replicate series 3, 4, and 5) the later samples showed higher crud concentrations. The conclusion was that nothing could be done to eliminate the wide variation in measured crud concentrations. The best estimate of the average crud concentrations in the reactor coolant could be obtained by taking many samples and averaging the results. The second problem area was sampling the outlet of the radwaste evaporator (i.e., old boric acid evaporator #4). This sample point was at the bottom of a U-tube that was valved off on one end. Liquid, therefore, did not normally flow through this U-tube and large amounts of crud had built up on the sample valve. Replicate samples taken over a period of time indicated that the sampling procedure should include a 15-minute purge at a high flow rate to purge the U-tube and then another 15-minute purge after the sample valve had been adjusted to the lower flow required for sampling. T' ~ 7cond purge was required 3 insure that crud loosened when the sample setting was changed w- emoved and did not get into the sampls.. I A.l.2.2 Gaseous Samples Verification of the validity of gaseous samples was ; determined by two methods. The first involved pitot tube and helium i dilution measurements (see sections 2.2 and 2.3 in reference (4)). Ventilation duct flows were calculated based on the results of each of the two types of measurements. The two results agreed (to within + 10%), indicating that representative samples had been obtained. The second verification method involved taking replicate camples under identical conditions over a time interval. Analyses of these replicate samples agreed (to within + 2a), indicating that representative samples had been obtained. i A.l.3 Validation of Sample Analyses l In order to insure accuracy of analyses, efficiency tables ' for all systems, both those in the NRC Mobile Laboratory and those at INEL, were established through the use of NBS standards or standards l A-2 l
traceable to NBS. These efficiency tables were obtained under all geometric conditions which the samples were analyzed. During the l l measurement period at Turkey Point the efficiencies were checked using i standards on a routine basis. In addition to checks with standards, intercomparisons were made with the Source Term group and the DOE Radiological and Environmental Sciences Laboratory (DOE-RESL) at the INEL and with the FPL radiation chemistry group at Turkey Point. Tables A.1 and A.2 show the results of these intercomparisons. Table A.1 shows the results of two sets of reactor coolant samples obtained by FPL and INEL. In general, agreement is very good (within + 20) for 28.Na and iodines. Cesiums show slightly higher differences. The differences seen in the cruds in the samples of 5/1/78 are similar to the differences observed in the replicate samples discussed in section A.l.2.1 and are due to actual differences in crud concentrations in the pairs of samples.
- A.2 Cata Handling Analysis of each gamma-ray spectrum was performed either in the NRC Mobile Laboratory or at INEL using a program that searches the j spectrum for gamma-ray peaks, identifies the isotopes producing the detected gamma rays, decay corrects activities to the sample collection i time, and provides radionuclide concentrations together with error estimates. For radionuclides of interest that were not detected, the
, program determines lower limits' of detection. A discussion of this program i
can be found in reference 4.
- Uncertainties quoted for radionuclide concentrations in this report are la errors due to counting statistics only. An additional uncertainty of approximately 10% should be added to the quoted errors to account for calibration and volume measurement errors. Indeterminate sampling errors have not been treated, however, the total errors due to sampling, calibration, and volume measurement are estNted to be approximately 20%.
In the detemination of mean concentrations, both individual measured concentrations and lower limits of detection (when a nuclide was not detected) were included. The lower limics of detection were included so that the means would not be biased high. The procedure used to do this was as follows. Since a lower limit of detection indicates that the true concentration is somewhere between zero and the lower limit of detection, two mean concentrations were obtained - one assuming the concentration of the undetected radionuclide to be zero and the second assuming the concentration to be equal to the lower limit of detection. The quoted mean was then obtained by averaging the two results. In the detemination of radionuclide concentrations in reactor coolant, the sample was gamma counted 3-4 times over a period of 2 weeks in order to obtain data on both long- and short-lived isotopes. When a radionuclide was detected in more than one spectrum, the quoted concentration for that sample was obtained from the average of the individual results. A-3
t TABLE A.1 COMPARISONS OF RESULTS FROM REPLICATE SAMPLES OF REACTOR COOLANT ANALYZED BY FPi AND INEL INEL Sample FPL Sample INEL Sample FPL Sample l Unit #3 Unit #3 Unit #4 Unit #4 12:00 11:55 09:09 09:00 l 12/2/77 12/2/77 5/1/78 5/1/78 Nuclide (uti/ml) (pCi/ml) (uCi/ml) (pCi/ml) i 131I 6.4 2 0.1(-2) 6.4 1 0.2 - 1.06 t 0.01 - 9.6 i 0.2(-3) 1321 1.55 1 0.05(-2) 1.4 1 0.1 - 1.94 1 0.02 - 1.73 0.05(-2) 133I 4.7 0.1(-2) 4.2 1 0.1 - 1.73 0.03 - 1.5 1 34 I 1.21 1 0.08(-2) 1.17 i 0.08( 2) 2.18 0.06 - 2.14 0.1(-2) 0.05(-2 ) 135I 2.4 0.1(-2) 2.0 1 0.2(-2 1.89 1 0.05 - 1.46 0.07(-2)
** 1.2*1 0.1(-3) , 134Cs 1.1110.09(-3 1.61 0.05(-3
- 136Cs 9.4 1 0.9(-5) ) **
2.4 1 0.5(-3) ) 137Cs 2.19 i 0.08(-3) 3.0 0.1(-3) 2.4 1 0.1(-3) 24Na 1.28 0.07(-2) 1.37 1 0.05(-2) 8.6 0.3 - 8.5 0.4 - ! seCo 1.85 0.09( 4) ** 8.5 0.4 - 9.5 0.8 - I 60C0 1.4 1 0.1 - ** 4.6 0.9 - 2.0 0.2 - 99Mo 1.1 1 0.1 - ** 7.110.6- 1.310.3-139Ba 3.3 1 0.4 - 6 1(-3)*** 6.3 1 0.1 - 5.8 1 0.8 - )
- Radionuclide not detected
** Radionuclide not measured *** Parent-daughter correction not made i
1 1 A-4
. - - - , _ = _ - - - . --
_- . - - . . . - -- .-- - .. .=. - - - _ . . . _ - - - _ _ - _ _ - _ _ _ _ _ _ TABLE A.2 RESULTS OF SPLIT AND REPLICATE SAMPLES ANAlvZED BY DOE-RESL AND INEL Unit #3 Reactor Coolant Waste Holdup Tank #2 09:38; 6/1/78 12:50; 5/23/78 12:48; 5/23/78 08:58; 5/25/78 08:56; 5/25/78 INEL DOE-RESL INEL DOE-RESL INEL DOE-RESL Nuclide (uci/ml) (uCi/ml) (uC1/ml) (uC1/ml) (uti/ml) (uC1/ni) 131I 7.2 1 0.5(-3) 6.1 0.2(-3) 2.14 0.02(-4) 2.1 0.3(-4) 1.31 0.06(-4) 1.1 0.4(-4) 134Cs 3.95 0.07(-4) 3.6 0.1(-4) 1.98 0.02(-4) 2.06 0.08(-4) 1.5 0.03(-4) 1.93 0.08(-4) 137Cs 5.7 i 0.1(-4) 6.0 0.2(-4) 3.10 0.05(-4) 3.2 0.1(-4) 2.6 0.1(-4) 2.8 0.2(-4) . 54Mn 2.13 0.09(-5) 2.9 0.3(-5) 7.3 0.3(-5) 8.0 0.5(-5) 57C0 7.7 1 0.9(-6) 1.4 0.2(-5) seCo 1.6A 0.07(-4) 2.20 0.08(-4) 7.8 0.3(-4) 8.5 i 0.3(-4) a soCo 3.16 0.03(-4) 3.14 0.11(-4) 1.6 0.06(-3) 1.85 0.06(-3) , s. 65Zn 1.3 0.3(-5) 6.7 1.0(-5)
& 99Mo 1.02 0.04(-3) 1.810.4(-3) 110 mag 9.6 0.3(-5) 1.12 0.08(-4) 12sSb 1.5 0.2(-5) 2.2 0.8(-5) i Monitor Tank A Monitor Tank C 15:45; 5/25/78 15:45; 5/25/78 10:20; 5/25/78 10:22; 5/25/78 INEL DOE-RESL INEL DOE-RESL Nuclide (uCi/ml) (uCi/ml) (vCi/ml) (uci/ml) 54Mn 4.9 1 1.5 - 5.0 1.1(-7) 7.9 1.3 - 7.7 1 1.5 -7) seCo 9.5 i 1.0 - 1.16 0.15(-6) 2.7 0.2 - 3.0 0.3-6) l '60Co 2.6 0.1 - 2.7 0.2(-6) 6.1 1.5 - 7.7 0.4 -6) 110 mag 9 2(-7) 1.1 0.2(-6) 1 i
On occasion, several samples were taken from a tank or a system for the purpose of validating a sample point (e.g., see section A l.2.1). When average radionuclide concentrations were later obtained for this tank or system, the results of the closely spaced samples were averaged to obtain a single value for that time period. This was done to prevent the final average from containing a bias. A-6
l l , APPENDIX B l l The data obtained during the in-plant ineasurenent studies at I Turkey Point Units #3 and #4 are presented in this Appendix. The Piping and Instrument Diagrams (P&ID) for the systems studied are also presented. B-1 I
TABLE B.1 RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 POWER OPERATIONS PRIOR TO REFUELING 11/9/77; 15:20 11/14/77; 10:37 11/16/77; 16:23 11/17/77; 11:58 11/18/77; 17:35 11/21/77; 17:10 h3 clide (uCi/ml) (uCi/ml) (uC1/ml) (uci/ml) (uci/ml) (uti/ml) 131I 6.0 t 0.1(-2) 1.52 0.05(-2) 1.4 0.1(-2) 1.33 0.04(-2) 1.67 0.03(-2) 1.8 0.1(-2) 132I U 1 1 0.07(-1) 1.71 0.06(-1) 1.90 0.04(-1) 1.86 0.03(-1) 2.00 0.04(-1) 2.16 0.07(-1) 133I 1.04 1.12 0.03(-1) 1.13 0.03(-1) 1.10 0.02(-1) 1.28 0.03(-1) 1.2 0.1(-1) 0.03(-1) 0.06(-1j 3.4 3.4 0.1(-1) 3.3 0.1(-1) 1341 1.68 3.4 1 0.3(-1) 3.5 1 0.1(-1) 0.1(-1) 135I 1.16 0.05(-1) 2.12 0.06(-1) 2.00 0.04(-1) 2.00 0.05(-1) 2.16 0.05(-1) 2.01' O.04(-1) 88Rb 4.3 0.3(-21 1.3 1 0.2 - 1.1 0.1(-1) 1.1 0.1(-1) 8.2 0.5(-2) 8.2 i 0.6(-2) 89Rb 1.1 0.1 - 1.0 1 0.1{-1) 1.0 i 0.1(-1) 9.1 0.2(-2) 1.0 0.1(-1) 134Cs 4.9 1.10 :i 0.0E(-3 0.1(-2) ) 3.120.3- 1.41 0.04(-3) 1.43 0.03(-3) 1.47 0.03(-3) 1.58 0.04(-3) 136Cs 1.1 0.2 -4 1.4 i 0.1(-4) 1.3 0.2(-4) 2.2 i 0.5(-4) 1.4 0.1(-4) 137Cs 1.5 i 0.1(-4) 1.77 0.08(-3 ) 5.3 1 0.3 -4 2.39 0.08(-3) 2.38 0.07(-3) 2.46 2.9 0.06(-3) 2.75 0.05(-3) A, 138Cs 1.5 0.1 -1 3.3 0.1 -1 3.1 0.1 -1) 3.1 0.1(-1) 0.1 (-1 ) 2.8 1 0.1(-1) 139Cs 1.4 0.1 -1 3.3 0.5 -1 3.0 0.3 -1) 2.7 0.2(-1) 2.2 0.2(-1) 3.0 0.3(-1) 24Na 2.6 0.3(-3) 3.0 0.2 - 6.1 0.3(-3) 6.2 0.3(-3) 7.0 0.4(-3) 6.8 0.3(-3 51Cr <6(-5) 1.7 1.2 - <2(-5) <3(-5) <4(-5) 1.7 0.3(-4 54Mn 1.2 1 0.3(-5) 1.0 0.3 - 1.0 i 0.4(-5) <4 - <2 - 1.9 0.2(-5 56Mn 45-4) <9 - <5(-3) <8 - <7 - <7(-3) 59Fe <C-6) <2 - 1.1 0.7(-5) <2 - <2 - 9 5(-6) 5700 <4 -5) <2 - 1 1(-5) <2(-5) <2 - 6 4(-6) 58Co 5.6^0.5(-5) 1.30 0.08(-4) 1.10 i 0.07(-4) 4.7 0.1(-4) 1.3 0.2(-4) 1.51 0.07(-3) 60C0 1,8 1 0.5(-5) 6 1(-5) 6.7 0.8(-5) 1.2 0.1(-4) 4.5 0.8(-5) 3.3 0.3(-4) 4 652n <9(-6) <2(-5) <9(-6) 1.9 i 0.7(-5) 2 i 2(-5) 1.3 0.3(-5) 91Sr 6 3(-4) 1.0 0.3(-3) 6 6(-4) 8 2(-4) 1.3 0.3(-3) 3 3(-4) 91my t t t t , t 9 3Y 311(-3) <5(-4) <3(-3) 2 1(-3) 5 2(-3) 3 1(-3) 95Zr 1.6 1 0.6(-5) 6.5 i 0.6(-5) 1.4 1.0(-5) 1.9 0.6(-5) <2(-5) 1.5 0.4(-5) 95Nb <2(-5) 6.9 0.6(-5) <2(-5) <8(-5) <6(-5) 2.4 0.3(-5) 99Mo 3.2 0.1 (-3) 2.2 0.4(-4) 1.28 0.05(-3) 1.4 0.3(-3) 1.56 0.06(-3) 1.36 0.09(-3) t t t t 99mTc t t
<2(-5) <2(-5) 1.6 1.0(-5) <3(-5) 2.4 0.4(-5) lo3Ru <3(-5)
_ _ _ = . _ _ . TfARLE B.1(cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 POWER OPERATIONS PRIOR TO REFUELING 11/9/77; 15:20 11/14/77; 10:37 11/16/77; 16:23 1Y/17/77; 11:58 Nuclide 11/18/77; 17:35 11/21/77; 17:10 (uCi/ml) (uC1/ml) (uCi/ml) (uCi/ml) (uci/ml) (uC1/ml) 103mRh t t t t t t 106Ru lo6Rh
<3(-5) t
<2(-) <3(-5) ~ 8(- )
<6(-5) 6 2(-5) t t t t 11orAg t
<5(-4) <2(- <6(-4) <8f-124Sb 411(-6) <7(-4) <6(-4) 12sSb i
<91 -
<4(-6) <8f- 6 5(-6) 2.0 0.3(-5) 1(-5) <2 -
1.5 1.4(-5) <2I -
<4(-5 <2(-5)
<2 -5) <21 l-129ETe <2 -
<2(-5 <2 - 5 i 5(-6) 129Te <5 -3 <2 - <5(-3 <7l -
<7 - <6(-2)
<5(l, l
o u <3 ,,4 < 4 ,, <4(,,4 <5 ,,
, 3 i,2(-4) cn 132Te <9(-5) <6(-5) 4 2-
& <5(-5) <5(-6) 3 1(-6) 139Ba 4 l f-3) <2(-2) 6 3- 1.6 0.2 1.7 0.2(-2) <7(-3) 140Ba 9 3Y,-5) 1.9 14ela 3 0.3(-4) 4 2- 5 3(-5)(-2) 4 2(-5) 1.2 0.2(-4) 7(-4) 7 4(-4) <9-5) 2 2(-5) 4 1 7(-5) 141Ce <2(-4)1
<6(-5 <2 - <2(-5) <2 -5) 2 1(-5) 1.4 0.4(-5) 143Ce <2(-4) <1 - <8(-5) <7 -5) <1(-4) 144Ce <4(-5) <3 - <3(-4)
<2(-5) <3 -5) <3(-5) <9(-6) 144Pr t t t t t t 152Eu <5(-5) <4(-5) <2(-5) <3 -5 4 <4(-5) <3 -
<7(-6)
, <9 (,,6) <5(,,6) <9 -6 <2(,-5) <5 187W <g(-4) <2(-4) <2(-4) <1(-4) 239Np <7(-5) <5(-4)
<2(-4) 1.1 0.9(-4) 1.0 0.6(-4) 6 3(-5) <4(-5) <7(-5) t Radionuclide not measured directly. Concentration can be inferred from parent or daughter.
4
** Radionuclide not measured.
I i
TABLE B.2 RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 DURING REFUELING 12/2/7/; 11:45 12/30/77; 11:34 1/3/78; 11:10 1/3/78; 14:40 1/11/78; 11:36 i Nuclide (uC1/ml)tt (uCi/ml)tt (uCi/ml)tt (uCf/ml)ttt (uCf/ml)tt 131I 2.7 i 0.4(-4) 1.0 0.1(-6) 4.7 0.3(-5) 6.8 0.2(-5) 2.6 0.2(-5) 132I 5.8 0.1(-3) <5 - * *
- 1331 <1(_4) <8 - * *
- 134I <4 - <8- * *
- 1351 <4 - <4 - * *
- esRb * *-
<7(-5) <5(-6) *
<4(-4) * *
- 89Rb <3(-5) 13*Cs 4.6 0.7 - 3.1 0.3 - 9.0 0.9(-6) 5.2 0.7(-6) 5.4 1 0.3 -
136Cs 1. 7 i 0. 5 - 5.3 0.9 - 6 1(-6) <3(-6) 1.2 1 0.2 - 137Cs 7.5 0.9 - 3.9 0.2 - 1.0 0.2(-5) 1.04 i 0.08(-5) 5.4 0.4 - 13Ns <7(-5) * * *
<3(-6) 139Cs <4(-4) <5(-5)
- 24Na <2(-5) <2(-6) <2(-6)
- 51Cr 1.110.l(-3) 5.910.1(-4) 1.8 0.2(-4) 1.9 t 0.1(-4) 4.3 0.2(-4) 54Mn 1.3 i 0.2(-4) 2.4
- 0.1(-5) 1.9 i 0.l(-5) 4.5 0.1(-5) 3.9 0.2(-5) 56Mn 2 1 1(-5) <1(-6) <2(-6)
- 59Fe 2.011.5(-5) 1 .8 i 0.21 -
7 2(-5) 1.4 0.1(-5 7 5(-6) 57Co 4.6 i 0.5(-5) 9.1 0.7I - 8.0 1 0.8 -6) 9.3 0.5(-6 1.4 2 0.1(-5) seCo 3.05 1 0.06(-2) 5.3 0.2I - 4.3 1 0.1 -3) 4.3 0.1(-3 7.5 t 0.1(-3) 60Co 3.3 1 0.1(-3) 7.3 0.31 - 5.9 1 0.3 -4) 7.7 0.2(-4 1.58 0.02(-3) 1 65Zn 5 2(-5) 1.4 0.5(-l 8 i 1(-6) 1.1 i 0.2(-5 4.6 0.9(-5) 91Sr <9(-5) <7(-6) 91mY t t t t t 93Y <7(-5) <7(-6) 95Zr 3 2(-5) 2.6 1 0.2(-S) 1.2 1 0.4(-5) 2.9 0.1(-5) 3.3 0.4(-5) 95Nb 6 i 3(-5) 5.0 1 0.7(-5) 4.310.9(-5) 5.4 0.3(-5) 6.5 i 0.9(-5) 99Mo <9(-4) <4(-6) <4(-6) <3(-5) <9(-6) 99mTc ~t t t t t lo3Ru 311(-5) 3.5 1 0.2(-5) 1.2 1 0.2(-5) 3.8 0.1(-5) 2.2 1 0.3(-5) 10 3nRh t t t t t lo6Ru 1.0 0.7(-4) <3(-6) <1(-5) 2.2 i 0.6(-5) <2(-5) 106Rh t t t t t
. . __ _ = _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
TABLE B.2(cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 DURING REFUELING 12/2/77; 11:45 12/30/77; 11:34 1/3/78; 11:10 1/3/78; 14:40 1/11/78; 11:36 Nuclide (uti/ml) (uC1/ml) (pC1/ml) (uC1/ml) (uCi/ml) 110 mag 2.0 0.1(-4 4.9 0.5(-5) 7.6 0.6 - 1.88 0.08(-5) 7.7 0.4(-5) 12"Sb 5.810.2(-4 6.5 0.3 - 8.0 0.2 - 7.2 0.2(-5) 1.16 0.04(-4) 12sSb 3.4 0.9(-4 3.0 0.3 - 3.7 0.3 - 3.6 0.2(-5) 5.8 0.8(-5) 129mTe 7 4(-5) 2.8 1 0.9 - 4 2(-5) 8 2(-5) 6 5(-6) 129Te <2(-4) <8(-5) * * *
- 131mTe 4 1 3(-5) <5(-6) <6(-6) <2(-5) 6 6(-6) 131Te 132Te 5.8 1 0.2(-4) <7(-6) <4(-6) 2 2(-6) 5 4(-6) 13988 * *
<5(-5) <4(-6)
- 140Ba 1.0 0.2(-4) 3 2(-6) 3 2(-6) 1 2(-6) <2(-5)
= 140La 4 1(-5) 1.1 m
0.6(-6) 2.1 0.4(-6) 7 6(-6) 3.0 0.7(-6) 141Ce 1.8 0.8(-5) 1.9 1 0.1(-5) 1.18 0.08(-5) 1.26 0.06(-5) 2.3 0.3(-5) 143Ce <5(-5) <7(-6) <6(-6) <2(-5) <9(-6) 144Ce <2(-5) 2.1 0.4(-5) 1.4 1 0.2(-5) 1.5 0.3(-5) 5 3(-6) 144Pr t t t t t 152Eu <3(-5 <3(-6) <2(-6) <2 -6 <1(-5
]",Eu <2(,,5 <1(, ,6) <2(, ,6) <1 ,,6 <5(,,6 187W <8(-5) <9(-6) <5(-5) <4(-5)
- 239Np <3(-4) <4(-6) <4(-6) *
<4(-6) t Radionuclide not measured directly. Concentration can be inferred from parent or daughter.
- Radionuclide not detected.
** Radionuclide not measured.
tt Sample obtained from RHR system. ttt Dip sample obtained from reactor cavity, i
TABLE B.3 RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 POWER OPERATIONS AFTER REFUELING 2/21/78; 13:16 2/23/78; 11:33 3/17/78; 09:35 3/25/78; 18:10 3/25/78; 18:41 4/9/78; 08:05 Nuclide (uCi/ml) (uC1/ml) (uCi/ml) (uci/ml) (uCi/ml) (uC1/ml) 85mKr esgp 2.1 2 0.2(-2) 2.310.3(-2) 2.7 0.3(-2) **
<1(_4) <4(-5) < 2(-1 )
87Kr 3.9 1 0.3(-2) 4.1 0.3(-2) 4.4 0.4(-2) 5.5 0.4(-2) ** ** ** eeKr 6.3 1 0.5(-2) 6.1 0.1(-2) ,, ,, ,, 89gp * * , 131mXe <5 (-5-) <9(-5) <7(-3) ** ** 133mXe 2.9 i 0.1(-3) <4(-3) 133Xe 5.3 1 0.4(-2) 5.4 1.01 2 0.04(-1 0.2(-3) ) ** ** 135mXe <1(-1) 7 7(-2) < 2(-1 ) 13sXe 1.75 1 0.03(-1) 1.94 1 0.07(-1) 1.96 0.03(-1) 137Xe 138Xe 1.13 t 0.02(-1) 1.17 1 0.02(-1) 1.56 0.02(-1)
& e4Br ** ** ** ** **
1.6i0.2(-2) 131I 4.3 0.2(-3 5.4 0.3 - 2.4 0.1(-2) 1.49 0.03(-2) 1.62 0.04(-2) 7.9 0.5(-3)
*
- 1.04 132I 4.3 0.1(-2 6.5 1 0.2 - 1.02 0.01(-1) 0.02(-1) 133I 5.6 0.1 -2 6.1 1 0.2 - 6.0 3.8 0.1(-2) 4.1 1* 0.1(-2) 5.4 1 0.2(-2) 134I 1.7 1 0.1 -1 1.8 1 0.1 - 1.65 0.2(-2) 0.07(-1 )
- 1.79 0.04(-1) las! 9.8 0.3 -2) 1.05 0.02(-1 ) 8.9 1 0.2(-2) *
- 9.8 0.3(-2) serb 7.710.3-2) 7.910.4(-2) 8.7 0.5(-2) *
- 5.8 1 0.2 -2) 89Rb 5.7 i 0.1 -2) 5.6 1 0.2(-2) 7.1 0.1(-2) *
- 6.4 0.1 -
134Cs 2.8 0.1 -4) 5 1(-5) <7- 2.01 0.03(-3) 2.16 1 0.03( 3) 4.4 0.3 - 136Cs 6 1 3(-5) 5.8 0.9 - <2- 4.0 1 0.2(-4) 4.6 1 0.3(-4 3.9 0.3 - 137Cs 6 i 1(-4) 6.1 0.7 - <1 - 2.310.1(-3) 2.5 0.1(-3 5.8 i 0.3 - 13sCs 1.6 0.l(-1) 1.8 i 0.1 - 2.02 0.02(-1) *
- 1.70 1 0.04(-1) 139Cs 1.1 0.1(-1) 1.320.1-1) 2.0 1 0.2(-1) *
- 1.6 0.1(-1) 24Na 2.38 0.09(-3) 2.7 1 0.1 - 4.3 1 0.2(-3) 4.3 0.4(-3) 4.9 0.4(-3) 7.1 1 0.5(-3) 41Ar 2.310.1(-2) 1.7 i 0.1 - 1.6 1 0.9(-3) 51Cr 1.6 0.2 - < 8(-3) < 2(-?.) 1.09 1 0.04(-4) 3.0 0.9(-4) 54Mn 3.0 1.08 1i 0.04(-4 0.3(-4) ) 2.8 1 0.4 - ** 5.6 1 0.2(-4) 2.5 2 0.1(-4) 2.6 0.2(-5)
* * <3(-3) 56Mn <4(-3) <4(-3) <5-59Fe <2(-5) 1.4 i 0.5 -5) <2 - 6.6 1 0.5(-4) 2.6 i 0.3(- ) 1.2 2 0.2(-5) 57Co 4il(-6) 2.9 i 0.9 -6) <5- 4 3(-5) 3.1 0.9(- ) <2(-6) seCo 1.16 0.04(-3) 3.4 i 0.2 -4) <7- 1.95 0.03(-2) 1.01 i 0.02 -2) 3.4 0.3(-4) 60Co 5.5 2 0.2(-4) 1.7 i 0.2(-4) 4 3(-4) 5.6 0.1(-3) 2.9 1 0.1(-3) 1.7 0.3(-4)
TABLE B.3 (cont'd) PADIONUCLTDE CONCENTP.ATIONS IN REACTOR COOLANT - UNIT 03 POWER OPERATIONS AFTER REFUELING 2/21/78; 13:16 2/23/78; 11:33 3/17/78; 09:35 3/25/78; 18:10 3/25/78; 18:41 4/9/78; 08:05 Nuclide (pCi/ml) (pCi/ml) (pC1/ml) (pCi/ml) (uC1/ml) (uC1/ml) 65Zn 7 1 6(-6) 9 4(-6) <2(-3) 8 3(-5) 1.2 0.3(-4) <3(-6) 91Sr 1.0 1 0.1(-3) 3 2(-4) <2(-3) <2(-3) <1(-2) 91my t t t t t
<4(-4) 1.0 <3(-4) 9 3Y 0.3(-3 2.0 0.6 -3) <2(-3) <6 (-3) <6(-3) <3(-3) 95Zr 6.1 t 0.6 -5 3.8 0.4 - <7(-j) 4.8 0.1 - 1.27 0.03(-3) 7.5 0.9(-5) 95Nb 1.2 1 0.1 -4 5.6 0.6 - 3.6 0.1 - 1.18 1 0.03(-3) 7.7 0.9(-5) 99Mo 1.3 0.1 -4) 1.6 0.2 - <3(-3) 3.8 0.1 - 4.1 0.1(-3) 1.48 0.05(-3) 99mTc t t t t t t lo3Ru 3.1 0.4(-5) 2.1 0.3(-5) <8(-4) 1.56 i 0.03(-3) 1.03 0.03(-3) 2.4 103mRh t t t t t t 0.3(-5) 106Ru 1.0 i 0.2(-4) 5 3(-6) 1.2i0.2(-3) 5 2(-4) <2(-5) 106Rh t t t t t t c' **
N 110 mag <3(-4) <5(-5) 5 1(-4) <2(-3) 4.3 0.8(-6) 124Sb 9 4(-6) 4 3(-6) <7(-4) 8.4 0.4(-4) 6.520.3(-4) 1.4 0.2(-5) 12sSb 1.1 i 0.7(-5) 4 3(-6) <8(-4) 5.3 0.5(-4) 3.4 0.5(-4) 6 2(-6) 129prie <2(-5) <8 - 4(-4) <6(-5) 5
<5(-5) <5(-5) 129Te 2.1 1 0.2(-2) <3- <5(-2) *
<2(-2 131*Te <4 ( ,4,) <4 - <3(f) 1.33 0.3(-3) <3(-4,) -
13"Te 1.3 i 0.1 - 1.3 0.3(-4) 4.1 0.1(-3) 2.6 *i 0.1(-3) <9(-6 1.3
- 13988 l.4 1 0.1 - 0.1(-2) 5 2(-3) 3.0 0.7(-3) 140Ba 4.2 1 0.3 - 1.810.6(-4) <9 - 4.8 0.8(-4) 6.9 0.9(-4) 1.1 0.1(-4) 140La 3 i 1(-5) <8(-5) <7 - 3 1 2(-4) <4(-4) <9(-5) 141Ce 3 1(-5) 4 i 2(-6) <8 - 2.0 0.3(-4) 1.1 0.1(-4 <3(-6) 14 3Ce <7(-5) <2(-4) <1 -3) 2 i 1(-4) 1.6 0.9(-4 <2(-4) 144Ce 2.011.5(-5) 3 3(-5) <6(-3) 5.4 0.9(-4) 2.3 0.6(-4 <7(-6) 144Pr t t t t t t 152Eu <8(-6) <2(-5) <6(-4) <2(-4) <2(-4) <2(-5) 154Eu <6(-6) <4(-6) <9(-4) <5(-5)
<3(-5)**
<2(-6) 1ssEu ** ** **
<4(-6 187W <7(-5) 8 5(-5) <7(-4) <8(-4) 1.1 0.3(-3) <5(-4 239Np 9 5(-5) 1.8 1 0.6(-4) <5(-4) 5.0 0.7(-3) 2.1 1 0.4(-3) <8(-5 t Radionuclide not directly measured. Concentration can be inferred from parent or daughter.
- Radionuclide not detected.
** Radionuclide not measured.
TABLE B.3 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 POWER OPERATIONS AFTER REFUELING 4/9/78; 08:06 4/12/78; 18:24 4/13/78; 11:26 4/15/78; 11:23 4/17/78; 10:00 4/17/78; 10:04 Nuclide (pCi/ml) (pC1/ml) (uCi/ml) (uCi/ml) (pC1/ml) (pC1/ml) esmKr ** ** ** ** ** 1.9 i 0.1(-2) esKr <4(-2) 87gp ** ** ** ** ** g{,p) 4 seKr ** ** ** ** ** 4.8 0.1(-2) 89gp ** ** ** ** **
<3(-2) 131mxe <3(-3) 133mXe 7.4i0.2(-3) 133xe 2.4 1 0.2(-1) 135mye ** ** ** ** **
13sye ** ** ** ** **
<](_]}
1.5 0.1(-1) 137Xe 138Xe ** ** ** ** ** 2 l .11 2(-2)(-1) 0.1
*
- 1.4 1.5
?' e4Br 0.2(-2) 0.2(-2) 1.5 0.2(-2) 1.6 0.1(-2) m 131I 7.2 1 0.3(-3) 7.3 0.2(-3) 6.9 6.6 0.4(-3) 5.8 0.2(-3) 6.3 0.3(-3) 132I 1.02 1 0.02(-1) 1.07 i 0.02(-1) 1.15 0.8(-3)) 0.09(-1 1.07 i 0.02(-1) 1.00 1 0.02(-1) 1.01 0.03(-1) 133I 5.4 1 0.2(-2) 5.3 0.2(-2) 5.0 5.5 0.2(-2) 5.2 0.2(-2) 5.5 0.1(-2) 134I *
- 1.83 0.4(-2) 0.04(-1 ) 1.83 0.02(-1) 1.62 0.04(-1) 1.64 0.06(-1) lasI 1.00 1 0.02(-1) 9.9 1 0.1(-2) 9 i 1(-2) 9.8 0.2(-2) 9.7 0.2(-2) 9.8 0.2(-2) 88Rb *
- 7.3 0.6(-2) 7.2 0.5 - 6.2 0.2 - 5.7 i 0.2 -
89Rb *
- 6.5 0.2(-2) 6.5 0.2 - 4.6 0.2 - 5.1 0.2 -
134Cs 4.1 0.1 - 3.43 1 0.06(-4) 3.19 0.05(-4) 3.0 0.1 - 3.0 0.2 - 4.5 1 0.2 - 136Cs 3.110.5- 4.7 0.1(-5) 6 1(-5) 5.510.4(-5) 5.2 0.1 - 6.4 1 0.3 - 137Cs 5.4 1 0.8 - 5.1 0.1(-4) 4.5 1 0.2(-4) 4.3 0.1(-4) 4.2 0.1 - 6.5 1 0.3 -4) 138Cs *
- 1.83 0.05(-1) 1.79 i 0.04(-1) 1.61 0.04( 1) 1.72 0.05(-1) 139Cs *
- 1.3 0.2(-1) 2.4 0.3(-1) 1.0 0.1(-1) 9.0 1 0.7(-2) 24Na 7.510.4(-3) 7.110.3(-3) 7.3 0.8(-3) 7.6 1 0.4(-3) 7.9 1 0.4(-3) 0.2(-3) siCr 7.7 0.2(-4 ) 6 1 1(-5) 1.1 1 0.2(-4) 6 2(-5) <3(-5) 1.1 0.4(-4) 54Mn 1.04 1 0.02(-4) 1.7 1 0.4(-6) 7.110.5(-6) 1.8 0.2(-5) 1.5 1 0.1(-5) 2.1 i 0.2(-5) 56Mn <3(-3) <4 - <8(-3) <6(-3) <6(-3) <8(-3) 59Fe 5.110.3(-5) <2 - 5 1 -6) 4.7 0.9(-6) <3(-6) <7(-6) 57Co 5.3 1 0.6(-6) <2 - <2(-6{. <9(-7) <2(-6) <3(-6) seCo 1.65. 0.03(-3) 9.1 1 0.5(-5) 1.17 , 0.05(-4) 1.16 0.08(-4) 1.1 0.4(-4) 3.4 0.2(-4)
TABLE B.3 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 POWER OPERATIONS AFTER REFUELING 4/9/78; 08:06 4/12/78; 18:24 4/13/78; 11:26 4/15/78; 11:23 4/17/78; 10:00 4/17/78; 10:04 Nuclide (pCi/ml) (pC1/ml) (pC1/ml) (uC1/ml) (pci/ml) (uC1/ml) 60Co 1.2 1 0.2(-3) 1.5 1 0.2(-5) 5.5 0.4(-5) 9.1 0.5(-5) 6.4 1 0.9(-5) 2.3 0.2(-4) 65Zn 7 3(-6) <2(-6) <2-6) <2 - <7 -
<3(-6) 91Sr <4 - <4(-4) <3-4) <2 - <6-4) <5 -
91mY <1 - *
<5 -4) <3 - <4 -4) <4 -
93Y <3 - <2(-3) 9 4(-4) <2 - <2-3) <2 - 95Zr 1.7 i 0.1(-4) 9 2(-6) 2.1 0.2(-5 1.9 0.2(-5) 4.2 0.8 - 2.5 0.4 - 95Nb 1.8 i 0.1(-4) 9 2(-6) 1.9 0.1(-5 2.2 0.3(-5) 2.8 0.9 - 2.7 0.4 - 99Mo 1.15 1 0.02(-3) 1.19 i 0.04(-3) 7.6 0.6(-4 1.52 i 0.08(-3) 2.4 0.1 - 6.2 1 0.3 , 99mTc t t t t t t lo 3Ru 9.1 1 0.2(-5) 1.0910.09(-5) 9.4 0.8(-6) 5.9 0.7(-6) 2 1(-6) 7 1 3(-6) 10 3mRh t t t t t t 106Ru 6 1(-5) <1 (-5) <1 (-5) <1(-5) <2(-5) <4(-5) 106Rh t t t t t t
? 110 mag
- 1.710.2- 1.8 1 0.4 - <1(-6) <2(-6) <2(-6) 5 3(-6) 124Sb 6.3 1 0.2 - 8.5 0.8 - 3.9 0.6(-6) 2.9 0.6(-6) <2(-6) <4(-6) 125Sb 3.7 1 0.3 - 1.0 0.2 - <4(-6) <4 - <5(-6) 1.6 0.6(-5) 129mTe <7 -5) <3(-5) <3(-5) <3 - <4(-5) <1(-4)
* <7 -
129Te <2 -2) <2-1) <2(-1) <1(-1)
<3-3) <4 -
131mTe
<2(-4) <3 -4) <3(-4) <5(-4)
* <6 -
131Te <5 -3) <3(-3) <3(-3) 132Te 1.9 1 0.6(-5) <9(-6) 1.6 0.4(-5) <8 - 4 2(-6) 1.7 0.7(-5)
- 9i3- <3 -
139Ba <3(-1) 1.1 0.1 -2) 1.26 0.08(-2) 140Ba 1.710.1(-4) 1.9 i 0.4(-5) 6 2- 1.6 0.3(-5) 7.510.5-4) 2.1 0.2(-4) 14ola <9(-5) <3 - 2 1- 1 1(-5) 4.1 i 0.5 -4) 1.2 0.2(-4) 141Ce 8 i 4(-6) <3- <2 - 2 1(-6) <3(-6 <5-143Ce <4(-4) <5 - <6 - <7(-5) <8(-5 <5 - 144Ce 2.1 1 0.4(-5) 1.3 0.5(-5) <9 - <1(-5) <2(-5 <2 - 144Pr t t t t t t - 152Eu <2 - <2 - 44 - <3(-5) <4(-5) <3-5) 154Eu <4 - <2 - <2- <2(-6) <5-6) <6-6) - 155Eu <5 - <4 - <6- <4- <5 -6) <2 -5)
<1 -
187W <l - <6- <l - <2-4) <1 -4) 239Np <8 - <3- <4 - <4 - <3 -5) <6 -5) t Radionuclide not directly measured. Concentration can be inferred from parent or daughter.
- Radionuclide not detected.
** Radionuclide not measured.
_- - _ _ - - - - - _- _. - _ - - _ . - - =- TABLE B.3 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 POWER OPERATIONS AFTER REFUELING 4/25/78; 10:06 4/25/78; 10:07 4/25/78; 10:08 4/25/78; 10:09 5/1/78; 09:18 5/9/78; 10:13 Nuclide (uCi/ml) (pCi/ml) (pC1/ml) (pC1/ml) (uC1/ml) (pCi/ml) emKr 1.7 2 0.1(-2) ** ** 1.7 1 0.1(-2) 1.57 i 0.03(-2) 85Kr <2(-2) . <4(-2) <3(-3) 4.24 2 07(-2) b8 2) 0 55 2) 1 133mXe 1s6 2.0.4(-3) <3(-3) <2(-3? ! 133mXe 1.23 0.04(-3) <2(-3) 5 1[-3) 133Xe 2.9 i 0.4(-1) ** ** 2.5 i 0.5(-1) 2.320.1(-1) 135mXe <3(-1)
<2(-1 ) 3 i 3(-1)
]5 1.50,i0.07(-1) ,, 1.5,i0.1(-1) 1.4 ,1 0.1(-1) 13eXe 1.13 i 0.02(-1) ** 1.17 0.03(-1) 9.2 1 0.6(-2) 5 84Br 2.0 0.3(-2) 9 1 2(-3) 1.3 20.3(-2) 1.3 0.3(- 2.0 1 0.2(-2) 1.6 1 0.2(-2 o 1311 7.2 0.3(-3) 7.1 0.5(-3) 7.1 0.4(-3) 7.0 0.4(- 7.5 i 0.2(-3) 7.9 0.2(-3 132I 1.02 1 0.02 -1) 1.00 2 0.01 - 1.05 1 0.05 - 1.04 i 0.02 1.01 t 0.03 -1) 1.07 1 0.02( 1) 133I 5.8110.07- 5.62 0.07 - 5.72 0.09 - 5.75 0.08 - 5.45 0.08 -2) 5.9 0.1(-2) 1341 1.72 0.03 - 1.80 1 0.02 - 1.77 0.02 - 1.8A 0.03 - 1.81 0.03-1) 1.7210.02(-1) 135I 1.00 0.03 - 9.9 0.3 -2) 1.02 i 0.03(-1) 1.02 0.05(-1) 1.02 0.02 1) 1.0310.02(-1) 88Rb 7.110.5- 6.0 1 0.4 -2) 7 1(-2) 6 3(-2) 7.3 0.7 - 812(-2) 89Rb 6.4 1 0.3 - 6.4 0.4 - 6.4 0.7 - 7 1(-2) 5.8 0.2 - 5.1 1 0.3(-2)
- 134Cs 6.210.3- 4.9 i 0.2 - 5.7 2 0.3 - 4.6 0.2(-4 7.9 i 0.3 - 1.59 1 0.03(-3) 136Cs 1.0 1 0.1(-4 8.310.6- 1.7 0.6 - 7.8 2 0.6(-5 8.4 2 0.3 - 1.30 0.09(-4) 137Cs 4.2 1 0.2(-4 6.3 0.2 - 6.8 0.2 - 6.36 0.07(4) 1.05 i 0.03 3) 2.0 0.1(-3)
- 138Cs 2.00 i 0.06(-1) 1.7 1 0.1 -1) 1.70 i 0.04(-1) 1.85 1 0.05(-1) 1.72 0.07 -1) 1.9810.08(-1) 139Cs 2.0 0.5(-1) 2.0 0.2 -1) <2(-1) <4(-1) 2.6 i 0.4(-1) 3 3(-2) 24Na 7.8 0.4(-3) 7.0 0.3(-3) 7.2**1 0.6(-3) 7.0 0.3(-3) 8.0 i 0.3(-3) 1.26
- i 0.05(-2) 41Ar <7(-4)
** <7(-4)
SICr <9(-5) 8 i 2(-5) <4 - 1.8 0.8(-5) 1.8 0.3(-4) 9 i 4(-5)
- 54Mn 2.6 1 0.3(-5) 2.1 1 0.5(-6) <2 - 1.710.2(-6) 2.8 0.2(-5) 2.810.3(-5) 56Mn <5- <7(-3) <6 - <6(-3) <3(-3) <3(-3) 59Fe <1 - 3 i 2(-6) <7 - 2.1 0.8(-6) 1.0i0.3(-5) <8(-6) 57CO <4 - <9(-7) <3 - <7(-7) 3.6 0.9(-6) <3(-6) 58Co 1.73 1 0.09(-4) 1.5 i 0.2(-4) 1.0 0.2(-4) 3.1 0.1(-5) 5.9 0.1(-4) 2.1410.06(-4)
TABLE B.3(cont'd) RADIONUCLIDE CONCENTPATIONS IN REACTOR C00LMT - UNIT #3 POWER OPERATIONS AFTER REFUELING l l 4/25/78; 10:06 4/25/78; 10:07 4/25/78; 10:08 4/25/78; 10 09 Nuclide (pCi/ml) (pCi/ml) 5/1/78; 09:18 5/9/78; 10:13 (pC1/ml) (pC1/ml) (pCi/ml) (pC1/mi) 60Co 1.3410.09(-4) 2.0 0.3(-5) ssZn <8 - 6 2(-5) 9 1(-6) 2.59 0.03(-4) 2.1 i 0.2(-4)
<4 - <6 - <1 -
91Sr <7 - <2 - <5 - <4(-6) <8(-6)
<1 - <2(-4) 91mY <3 - <6 - <9 - <7 - * <3(-4) 93Y <4 - <8 -3) 95Zr <2 - 2.4 i 0.8(-3) <6(-3) <2(-4) <5(-4) 8.0 i 0.8 - <4(-5) 1.1 0.4-5) 6.6 0.4(-5) 3.6 95Nb 2.7t0.6(-5) 1.2 1 0.5 - <3(-5) 7.2 0.5 -5) 0.4 -6) 8.4 1 0.4(-5) 3.9 0.6-5) 99Mo 1.64 1 0.03(-3) 1.4 1 0.2 - 1.11 0.05(-3) 1.5 0.2-3) 1.20 7.4 ssmTc t t t t 0.02(-3) 0.3 -4) t t lo 3Ru 1.1 i 0.5(-5) 1.0 i 0.4(-5) <3(-5) 2.4 0.5(-6) 1.9 0.2(-5) 1.1 103mRh t t t t t 0.4(-5) losRu t
<5(-5) 1.3 0.6(-5) <7 -4) <7(-6) m losRh t t
<3(-5) <7(-5) t t
.L 11os%g <7 - t t
" 6 (-6) <4-5) 7 3(-7) 1.1 i 0.2 - <5(-6) 124Sb <4 - 1.12 0.08(-5) <5 -5) 1.5 12sSb <2 - 0.6(-6) 1.2 1 0.2 - <4(-6)
<4 - <7 - <3(-6) 1.9 0.7 -
129mre <2 - <6 - 1.7 0.8(-5)
<6 - <2(-5) <6(-5) <2 -
129Te <6 - <2 - <2 - *
- 131mTe <3 - <5 - <2 -
<7 - <4 - <3(-4) <4 -
131Te 7 2 1(-3) <4 - <6 - <2 -
- 132Te <3(-6) <3 - <4 - <2 - <2 -3) 139Ba 2i1(-3) 1.2 <3(-5) <2-5) 0.1(-2) 3 2(-2) <7 - 1.77 0.05(-2 1.8110.02(2) 140Ba 7.1 0. 3(-4) 2.0 0.1(-4) 1.6
<2 -
0.4(-4) 1.820.4(-4) 7.9 1 0.6(-5) ) 6.0 1 0.3(-4 140La 9 i 2(-3) <3(-4) <3(-4) 1.210.9(-4) 3.0 1 0.6(-4 141Ce <7 - <3-6) <5 - <2-6) 14 3Ce <8 - <1 -4) <2 - <4-6) <7(-6)
<2 -4) <5 -5) <2(-4) 14*Ce <3 - <7 -6) <2 - <4 -6) 144pr t t t
<2-5) <3(-5) t t t is2Eu <7(-5) <2(-5) <9(-5) 154Eu
<6(-6) <6 - <8 -5)
<1(-5) <3(-6) <7(-5) <2(-6) <4 -
issEu <2(-5) <4 - <2-5)
<4 -
<1 a) 2 1-6) <7 - <2 -5) 187W <3(-4) <2 -4) <3(-4 3 2(-4) 239Np 1.2 1 0.3(-4) <9 - <1 -3)
<2 -4) <9(-5 <4(-5) <7 -5) t Radionuclide not directly measured. Concentration can be inferred from parent or daughter.
Radionuclide not detected.
** Radionuclide not measured.
4
L TABLE B.3(cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 , POWER OPERATIONS AFTER REFUELING 5/16/78; 09:51 5/22/78; 12:00 5/25/78; 09:23 6/1/78; 09:38 Nuclide (uCi/ml) (uC1/ml) (pC1/ml) (uCi/ml) esmKr 1.5 ** 2.0 *1 0.1(-2) 7 2(-3) 0.3(-2) ** esKr <1(-2) e7KP 4 i l(-2) 1.7 1 0.9(-2) 3 1 (-2) seKr 4.98* 1 0.08(-2) 2.10 i 0.04(-2) 3.9 1 0.1(-2) ** 89Kp <5(-2) 3 1(-2) ** 131mXe <2(-3) <4(-3) <2(-3) ** 133mXe 8.7 i 0.6(-3) 6.3 0.5(-3) 3 1(-3) ** 133Xe 3.110.3(-1) 2.6 0.4(-1) 1.6 i 0.2(-1) ** ! 13smXe <2(-1) <3(-2) 5 4(-2) ** 13sXe 1.6 *i 0.1(-1) 3.05 0.04(-2) 1.20 0.06(-1) ** 137Xe 5 1 4(-2) 4.3 0.3(-2) 13eXe 1.33 0.07(-1) 5.82 i 0.07(-2) 1.03 0.04(-1)
?' 84Br 2.2 1 0.2(-2 6.7 0.8(-3) 1.4 0.1(-2 1.6 0.1(-2 M 1311 7.8 0.2(-3 3.16 0.08(-2) 8.1 0.4(-3 7.2 0.5(-3 132I 1.02 1 0.03( 1) 6.4 0.1(-2) 7.7 0.3(-2 1.09 i 0.02( 1) 133I 6.0 1 0.2(-2 2.95 0.06 -2) 6.0 0.2(-2 5.8 0.2(-2))
, 134I 1.86 0.03(-1) 1.05 2 0.01 -1) 1,69 0.04(-1) 1.97 0.05(-1 135I 1.06 t 0.02( 1) 5.16 1 0.05 -2) 1.03 i 0.03(-1) 1.09 0.02(-1) serb 8.0 1 0.7 - 3.6 0.1(-2) 5.9 1 0.2(-2) 6.510.3(-2) a 89Rb 7.6 0. 3 - 2.9 0.1(-2) 5.4 0.2(-2) 6.8 1 0.2(-2) i 134Cs 7.9 i 0.3 - 2.96 0.03(-3) 1.85 i 0.05(-3) 3.95 0.07(-4) 136CS 1,0 f g,1 - 8.3 0.3-4) 5.6 0.2(-4) 6.2 0.4(-5) 137Cs 1.12 1 0.09 3) 3.1 1 0.1 - 13eCs 2.48 0.07 -1) 8.7 0.3 - 2.610.1(-3)) 1.68 0.05(-1 .5.7i0.1(-4)) 1.87 i 0.05(-1 139Cs 1.2 i 0.5(-1) 6.6 0.9 - 1.12 0.08(-1) 1.8 0.1(-1) 24Na 8.5i0.3(-3) 4.610.2(-3) 8.2 0.3(-3) 9.810.9(-3) 41Ar <6(-4) <4(-4) <5(-4) SICr <4(-5) 1.7 -0.1(-3) 3.8 1 0.4(-4) <3(-5) 54Mn 1.7 i 0.2(-5) 7.6 0.7(-5) 4.7 0.3(-5) 2 1 1(-6) 56Mn <6(-3) <5(-3) <7(-3) <5(-3) 59Fe <4(-6) 8 1(-5) 2.6 0.4(-5) <3(-6) 57Co <1(-6) 1.0 0.5(-5) <2(-6) 3 1(-6) seCo 2.1 1 0.1(-4) 3.5 0.1 (-3) 8.0 1 0.2(-4) 1.3 0.1(-5)
TABLE B.3(cont'd) i RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #3 POWER OPERATIONS AFTER REFUELING 5/16/78; 09:51 5/22/78; 12:00 5/25/78; 09:23 6/1/78; 09:38 Nuclide (uC1/ml) (uCi/ml) (uC1/ml) (uCi/ml) 60Co 1.88 0.08(-4) 7 1(-4) 2.8 0.5(-4) 652n 4 2(-6)
<4 (- 6 7 2(-5) 9 3(-6) <4 -
91Sr <5 - <5 - 1 <5 - <4 - 91mY <3 - <3 - h <3- <4 - 9 3Y 95Zr
<2 -
1.9
<2-) <3 - 5 i 3(-3) 0.3(-5 5.8 1 0.3 - 1.05 0.05(-4) <3(-6) 9sNb 2.8 0.3(-5 4.6 0.3 - 1.16 1 0.05(-4) 5 1(-6) 99Mo 1.00 1 0.04( 3) 5.4 0.7 - 1.2 0.1(-2) 1.02 0.04(-3) 99mTc t t t t 103Ru 4 i 1(-6) 1.6 1 0.1(-4) 2.5 0.2(-5) <2(-6) loamRh t t t t losRu 3.6 0.8(-5) <2(-4) 4 1(-5) <4(-5) 106Rh t t t t ? 11onng 5 i 1(-6) 2.6 0.8(-5) 7 1 2(-6) <3(-6) g 124Sb 6 i 1(-6) 2.0 0.2(-5) 1.55 1 0.09(-4) 1.9 1 0.6(-6) 12sSb 1.2 i 0.3(-5) 7 3(-5) <2 - <5(-6) 129mTe <8 - <5 -4) <9- 5 3(-5)
~ 129Te <2 - <6 - <9 -
- 131nge <2 - <4 - <2 -
<5 - <2-4) 131Te <3 - <3- <4 -3) 132Te 3 1 -5) 2.9 0.7(-4) 1.9 0.4 <3-6) 139Ba 1.9 0.6(-2) 4 1(-3) 9 1(-3)(-5) 4 1-140Ba 8.4 i 0.6(-5) 1.81 0.08(-3 2.8 0.1(-4) 4i1-141La <2(-4) 5.2 0.7(-4)) 2.2 i 0.6(-4) 513-141Ce <3(-6) 3 2(-5) <3 - <3 -
143Ce <5(-4) <2(-4) <1 - <2 - , 144Ce <8(-6) <8(-5) <2 - <2 - 144Pr t t t t is2Eu <4 - <3 - <8 - <3 - 154Eu <4 - <3 - <4 - <3-1ssEu <4 - <5 - <6 - <6 - 187W <8 - <8 - <5 - <4 - 239Np <2- <3 - <6 - <2 - t Radionuclide not directly measured. Concentration can be inferred from parent or daughter. Radionuclide not detected. Radionuclide not measured.
t TABLE B.4 i RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #4 POWER OPERATIONS 12/2/77; 12:00 12/6/77; 15:06 12/8/77; 09:45 12/9/77; 09:47 12/10/77; 09:16 12/11/77; 10:52 Nuclide (pC1/ml) (pCi/ml) (pci/ml) (pCi/ml) (pCi/ml) (uci/ml) 8:>mKr ** ** ** 2.0 0.1(-2) 2.5 0.1(-2) 1.9 0.1(-2) 85Kr ** ** **
<7(-4) <3(-4) <2(-4) 87Kr ** ** ** 2.7 0.1(-2) 3.1 0.1(-2) 2.4 0.1(-2) >
, eeKr ** ** ** 4.2 0.1(-2) 5.0 0.1(-2) 3.5 0.5(-2) 89Kr 131mXe **
** ** <2(-3) 9 5(-4) <5(-4)
'133mxe ** ** ** 1.3 0.1(-2) 1.05 0.04(-2) 7.9 0.4(-3) 133Xe ** ** ** 5.1 i 0.5(-1) 3.5 0.3(-1) 2.7 0.1(-1) 13smXe ** ** ** <6(-1) <2(0) 2 2(-2) .
135Xe ** ** **
- 1.8 0.1(-1) 1.51 0.05(-1) 1.27 0.03(-1) 137Xe j 13 axe ** ** ** 3.2 0.2(-2) 8 2(-2) 3.1 1 0.1(-2)
$ 84Br ** ** ** ** **
8.6 1
- 131I 6.4 1 0.1(-2) 3.7 1 0.1(-2) 7.2 0.1 (-3) 9.8 0.3(-2) 2.08 0.09(-2) 0.2(-3) is21 1.55 1 0.05(-2) ** 9.9 0.2(-3) 3.36 0.04(-2) 1.21 0.02 -2) 1.18 0.03 -
133I 4.7 1 0.1(-2) 1.59 i 0.03(-2) 1.07 0.02(-2 ** 2.08 0.03 -2) 1.29 0.03 - 134I 1.21 0.08(-2) ** 1.09 0.03(-2 1.08 0.04(-2) 1.36 0.05-2) 1.10 0.03 - 135I 2.4 0.1(-2) 9.9 0.2(-3) 1.10 0.08(-2 5.2 0.1(-2) 1.4 0.1(-2) 1.17 1 0.07(-2 serb 5.8 0.3(-2) ** 7.4 0.6(-2) 6.3 0.8(-2) 2.2 i 0.5(-1) 5.6 0.3(-2) 89Rb
** 8.0 7.5 0.9(-3) <2(-2) 7.210.6(-3) 134Cs 1.1 1.11 1 0.09(-3 0.1(-2) ) 98 0.2(-4) 1.13 0.8(-3) 0.03(-3 ) 5.1 0.1 (-3) 1.51 0.05(-3) 1.1 0.1(-3) 136Cs 7.9 5.7 2.8 0.1(-3) 4.5 0.3(-4) 9 3(-5) 137Cs 9.4 10.9(-5) ) 1.99** 0.7(-5) 2.1910.08(-3 0.04(-3 ) 2.11 0.6(-5) 0.04(-3 ) 8.5 0.2(-3) 2.5510.06(-3) 1.74 0.03(-3) 13eCs 4.7 0.1(-2) 4.2 0.1(-2) 5.2 0.2(-2) 1.6 1.5(-2) 3.8 1 0.1(-2) 139Cs 2.1 0.4(-2) ** <2(-2) <2(-2) 1 1(-3) 24Na 1.2810.07(-2) 8.6 1 0.3(-3) 7.5 0.2(-3) 4.5 0.l(-4) 5.4 0.3 - 5.1 0.3(-3) 41Ar ** ** ** 6 2(-4) 1.6 0.2 -
51Cr <4(-5) <4-5) <3(-5) 2.4 0.3-3) 2.2 0.1 - 513(-4)(-4) 4.5 0.6 54Mn 5.0 0.5(-6) <5-5) <5 - 2.9 0.1 - 3.7 0.1 - 1.9 0.5(-4) 56Mn <4(-4) <5 -3) <3 - 3.5 0.4 - <5(-4) <5(-4) 59Fe 5.4 0.8(-6) 1.0 i 0.8(-5 <3 - 2.8 i 0.1 - 2.7 0.1 - 5.8 1 0.7(-5)
<3 - 2.0 i 0.3 - 9.2 i 0.7 - 9 1 2(-6) 57C0 212(-6) 1.1 0.2(-5 6.7 0.2 - 2.49 0.05(-3) seCo 1.85 0.09(-4) 2.5 0.1(-4 2.0 0.1(-4) 1.1 i 0. . , -
TABLE B.4 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR C00LAfC - UNIT 04 POWER OPERATIONS 12/2/77; 12:00 12/6/77; 15:06 12/8/77; 09:45 12/9/77; 09:47 12/10/77; 09:16 12/11/77; 10:52 Nuclide (pCi/ml) (uti/ml) (uci/ml) (uC1/ml) (pCi/ml) (uci/ml) 60Co 1.4 0.l(-5) 2.6 0.3 (-5) <2 -5) 9.3 0.7(-4) 5.1 0.2(-4) 3.6 0.3(-4) 652n 1.7 1.0(-6) <l (-5) <2 -5) 2.9 0.3(-5) 2.3 0.2(-5) 91
<3(-4) 4 4(-5) 6 3(-5) <1 -4) <2(-4) <6(-4) 2 2(-4) 93Y <3(-4) <1 -
612(-4) 1.4 0.4 - <4(-4) 3.0 0.2(-5) 952r 912(-6) 7 3L-6) <2 - 3.7 0.1 - 3.2 0.2(-4) 2.7 i 0.4(-5 ssNb 9 4(-6) <3(-5) <4 - 3.5 0.2 - 2.9 1 0.1(-4) 2.2 0.4(-5 99Mo 1.1 0.1(-4) 9.0 0.3(-5) 1.28 0.07(-4) 1.9 0.2(-3) 1.85 3.6 99Mc t 0.05(-3) 0.4(-4 t t t t t 10 3Ru 1.0 0.8(-6) <8(-6) <3(-5) 2.1 0.4(-5) 1.4 0.2(-5) IoamRh t
<2(-5) t t t t t 106Ru <7(-6) <8(-5) <2(-5) <2(-5) <9(-6) <2(-5) 106Rh t t t t t y' 110 mag t
g <4(_4) <5-4) <8(-4) <3(-3) 3.3 0.8(-5 <7(-4) 124Sb 2.0 0.4(-6) <8 -6) <2 - 1.4 0.l(-4) 7.3 1 0.9(-5 1.5 0.3(-5) 12sSb <4 - <8-6) <3 - 4.7 0.8(-5) 1.4 0.4 <2-129mTe <5 - <6 -5) <3 - <2(-5) 8 4(-5)(-5 <2 - 129Te <9 - **
<7 -3) <8(-3) <6(-3) <9 -
131mTe <1 - <1(-4) <9(-5) <2(-3) <2(-4) <2 -4) 131Te ** ** ** ** ** ** 132Te <7(-6) <2 (- 5) <2(-4) <5(-4) <2(-4) <2(-4) 139Ba 3.3 1 0.4(-3) ** 4 2(-3) 5 1(-3) 6 1(-3) 5.6 0.4(-3) 140Ba 8 4(-6) <2 - <3 -5) 5 4(-5) 4.0 0.4(-4) . 140La 1.0 0.6(-5) <1 - <1 -4)
# <4-5{
<2 -4, <2(-5) <8(-5 141Ce <6(-6) <8 - <3- <2-5) 4 2(-6) <2 -
143Ce 1.210.8(-4) <2 - <4 - <3 -4) <1(-4) <1 - - 144Ce <4(-6) <8 - <3- 3 3(-6) <6(-5) <1 - 144pr t t t t t t 152Eu <2(-5) <2(-5) <6(-5) <5(-4) <4(-5) <3(-5) 154Eu <2(-6) 5 4(-6)
<2(-5) <3(-6) <2(-6) <7(-6)
IssEu ** ** ** ** 167W <2(-4) <2(-5) <7(-5) <2(-4) <2(-4) <2(-a) 239Np <3(-5) <2(-5) 7 2(-5) 8 8(-4) 5 5(-5) <2(-4) t Radionuclide not directly measured. Concentration can be inferred from parent or daughter. Radionuclide not detected.
** Radionuclide not measured.
TABLE B.4 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #4 POWER OPERATIONS 12/12/77; 09:27 1/7/78; 16:18 1/11/78; 20:30 1/18/78; 11:27 1/19/78; 11:27 1/20/78; 13:05 Nuclide (uCi/ml) (uCi/ml) (pCi/ml) (pCi/ml) (uci/ml) (uti/ml)
** 1.6 1 0.1(-2) ** ** **
85mKr 1.75 1 0.03(-2) esKr ** ** ** **
<3(-5) <6(-2) 87Kr ** 2.1 0.1 2) 2.3 0.1(-2) ** ** **
seKr 89Kp 3.2 0.1((-2) 3.8 0.1(-2) 131mXe <3(-5) <2(-3)
** 5.3 0.1(-3) 7.1 ** ** **
133mXe 0.9 -3) ** ** ** 133Xe ** 1.8 0.1(-1) 2.4 0.1 -1) ,
** 1.7 ** ** **
13smXe <4(-2) 0.3 -2) ** ** ** 13sXe ** 1.02 1 0.02(-1) 1.15 i 0.02(-1) ** ** ** 137Xe ** * *
** 2.3 ** ** **
13eXe 3.3 1 0.1(-2) 0.1(-2) cn 84Br ** ** ** ** ** ** 1311 7.3 0.3(-3) 5.79 0.06(-3 6.1 0.1 (-3) 7.2 0.5(-3) 6.8 0.1(-3) 6.7 i 0.3(-3) 132I 1.06 0.02(-2) 1.25 0.02(-2 1.20 0.02(-2) 1.3220.02(-2) 1.36 0.02 -2) 1.2410.02(-2) 133I 1.19 0.02(-2) 1.13 i 0.04(-2 1.20 0.04(-2 1.19 0.04(-2) 1.12 0.02-2) 1.10 0.02 - 1341 1.1310.04(-2) 1.24 1 0.04(-2 1.19 i 0.03(-2 1.4 0.1(-2) 1.53 1 0.06 -2) 1.32 0.05 - 135! 1.15 1 0.07(-2) 1.1 1 0.1(-2) 1.20 2 0.03(-2 1.33 0.06(-2) 1.21 0.04 -2) 1.19 1 0.04 - serb 5.9 i 0.3(-2) 5.0 0.4(-2) 5.1 0.4 - 5.4 0.3-2) 6.2 0.5(-2) 5.1 0.1 -2) 09Rb 7.310.8(-3) 9 1(-3) 6.8 0.3 - 1.0 0.1 -2) 1.4 0.1(-2 9.3 0.3 - 134Cs 7.6 1 0.7(-4) 6.2 0.1(-4) 4.4 0.4 - 7.5 0,8 -4) 7.2 1 1.0(-4 8.3 0.6 - 136Cs 6 2(-5) 1.5 <2(-4) 4 1 2(-5) 2.9 0.8(-5 3.0 1 0.4 - 137Cs 1.4320.07(-3) 1.24 0.3(-5)) 0.02(-3 1.2 0.1(-3) 1.4610.08(-3) 1.5 0.1(-3 1.62 0.06(-3) 13ecs 3.7
- 0.1(-2) 3.6 0.1(-2) 3.8 1 0.1(-2) 4.1 0.1(-2) 4.0 1. 0.2(-2) 4.010.1(-2)
- 139Cs <5(-3) <3(-3) <2(-3) 1.4 0.2(-2) <2(-3) 1.4 0.3(-2) 24Na 5.2 1 0.2(-3) 4.8 1 0.2(-3) 5.5 i 0.3(-3) 5.6 0.2(-3) 5.5 0.2(-3) 41Ar ** <6(-4) 4.9 1.05 1 0.07(-3 0.2(-3) ) ** ** **
s1Cr 4.0 0.5(-3) 2 1(-5) <2-4) 4 1(-5) 3 i 1(-5) 4 1 3(-6) 54Mn 6.0 0.6(-4) 9 1(-6) <1 - 1.0 0.2(-5) 2.0 0.5(-5) 2.0 0.4(-6) 56Mn <3(-4) <6(-4) <4 - <6(-4) <6(-4) <5(-4) 59Fe 5.3 1 0.7(-4) 2.0 0.7(-5) <2- <4(-6) 5 2(-6) 1.6 1.1(-6) 57Co I .3 + 0.1(-5) 4 2(-6) 8 3(-5) 2 1(-6) <2(-6) <4(-6) 58C0 1.2 1 0.1(-2) 3.0 0.1(-4) 1.0 0.3(-4) 2.1 1 0.2(-4) 2.7 0.4(-4) 1.13 0.09(-4)
TABLE B.4 (cent'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT H POWER OPERATIONS 12/12/77; 09:27 1/7/78; 16:18 1/11/78; 20:30 1/18/78; 11:27 Nuclide 1/19/78; 11:27 1/20/78; 13:05 (uCi/ml) (uci/ml) (uC1/ml) (uC1/ml) (uci/ml) (uCf/ml) 60Co 1.0 0.2(-3) 4 1(-5) 7 3(-5) 3.1 65Zn 2.5 0.3(-5) 6.310.8(-5) 1.05 0.08(-5) 0.4(-5) <3(-6) <2(-4) <3(-6) <4(-6) <2(-6) 91Sr <2(-4) <5(-5) 91my ** **
<2(-4) <2(-4) <2(-4) <4(-4) ,
93Y 2.2 0.6(-3 5 1(-4) <2(-4)
<2 <2(-4) <5(-4)
! 95Zr 4.6 0.3(-4 5 i 1(-6) <2 -- } 8 2(-6) 4 1(-6) 1.0 4.4 0.5(-5) , 95Nb 0.2(-4 6 2(-6) <4-) 5 1(-6) 4.9 0.9(-6) 5 i 1(-6) 99Mo 1.0810.02(-3) 9.6 0.2(-5) <3(-4) 2.6 0.6(-5) 3 1(-5) 1.3 0.1(-4) 99mic t t t t t t 10 3Ru 3.5 i 1.0(-5) 4 2(-6) <2(-4) 3 3(-6) 3 1 3(-6) <5(-6) loanRh t t t t t t 106Ru <2(-5) <4(-6)
- ca 106Rh
<4(-6) <5(-6) <4(-6) t t t t t t 1 11cmAg 3.4 1 0.8(-5) <3f-4)
- N 1.5
<3(-4) <3(-4) <3(-4) 124Sb 0.1(-4) 4.4 0.9(-6) <2 -4) 1.3 0.5(-6) 12sSb 3.0 0.6(-5) <4 - <1 -
4 3-6) 1.9 1 0.6(-6) - 6t[-6)
<5 -6) 8 <4( -
129mTe <3 - <3- <2 - <4 -6 <4 -61 <4 d66 129Te <7 - <9 - <6 - <9 -3 <1 - <8 3 l 13 e <2 - < 5 (- 5, < 2 ,, <9(-5 <8 - ) <8((,, 132Te 5 2(-5) *
<7(-6) <6(-5) <5(-5) <3(-5) 139Ba 5.6 0.6(-3) 4.8 0.4(-3) 4.9 0.2(-3) 4.1 0.3(-3) 5.3 0.4(-3; 3.4 140Ba 1.3 5.0 0.6(-5) 0.4(-3) 1.0(-4) 3 2(-4) 1.0 0.2(-4) 1.0 0.3(-4) <2 - i 14cLa <5- <1 - <2 - <4-) <5 - <2 -
141Ce <2 - <4 - <2 -
- 14 3Ce <7 - <2 - <2 - <4-l
<4 i
<3-
<3 -
<3-
<5 -
244Ce <7 - <3- <4-4) 3 2(-5) 1.5 1.0(-6) <3-6) 144Pr t t t t t t 152Eu <8(-6) <4(-6) <2(-4) <9(-6) <1 (-5) <8(-6) , 154Eu <4(-6) <2(-6) <2(-4) <5(-6) <6(-6) <2(-6) IssEu ** ** ** ** ** ** 187W 1.0 0.2(-4) <3(-5) <2(-4) <6(-5) <6(-5) 1.6 <2(-4) 239Np '0.5(-4) 5 4(-6) <8(-5) <4(-5) <8(-5) <3(-5) i t Radionuclide not directly measured. Concentration can be inferred from parent or daughter. Radionuclide not detected.
** Radionuclide not measured.
,___r ___ _ -r----- ___
TABLE B.4 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #4 POWER OPERATIONS 1/22/78; 11:16 1/23/78; 11:14 1/24/78; 10:31 1/25/78; 04:14 1/26/78; 12:55 2/3/78; 09:43 Nuclide (pC1/ml) (uC1/ml) _ (pCi/ml) (pCi/ml) (uci/ml) (pci/ml) esmKr ** ** ** ** ** ** esKr - ** ** ** ** ** ** 87Kp ** ** ** ** ** *w 88Kr ** ** ** ** ** ** 89Kp ** ** ** ** ** ** 131mye ** ** ** ** ** ** 133mXe ** ** ** ** ** ** 133Xe ** ** ** ** ** ** 135mXe ** ** ** ** ** ** 13sxe ** ** ** ** ** ** 137xe ** ** ** ** *. ** 138Xe ** ** ** ** ** ** tjo 84Br ** ** ** ** ** ** 5; 131I 5.8 0. 3(-3) 6.7 0.2(-3) 6.0 0.4(-3) 6.3 0.3(-3) 4.7 0.1(-2) 5.9 i 0.2(-3) 132I 1.26 i 0.02(-2) 1.28 0.02(-2 1.23 0.02 -2) 1.29 0.03(-2 1.4310.03-2) 1.37 0.02 - 133I 9.9 i 0.2(-3) 1.17 i 0.02 - 1.02 0.06-2) 1.07 i 0.06(-2 2.86 0.05 - 1.27 0.04 - , 134I 1.42 0.05(-2) 1.38 0.05 - 1.38 i 0.04 -2) 1.37 0.03(-2 1.40 0.06 - 1.46 0.06 -
)
lasI 1.09 1 0.09(-2) 1.27 i 0.04 - 1.23 0.03(-2) 1.26 0.03(-2 1.54 0.04 - 1.38 0.03 - serb 5.1 0.1(-2 5.2 0.3 - 5.3 0.2(-2 5.2 0.3(-2) 4.710.2(-2) 4.6 1 0.2(-2 89Rb 9.4 i 0.2(-3 8.6 1 0.3 - 1.0 0.1(-2 8.8 0.3(-3) 7.8 8.8 0.2 - 134Cs 7.9 i 0.3(-4 9.7 1 0.5 - 1.02 0.09( 3) 9.6 0.6(-4) 1.90 0.3(-3)) 0.07(-3 7.7 0.7 - , 136Cs 5.2 1 0.9 - 3.3 0.3(-5) 6.8 0.5(-4) 2.5 0.2 - i 137Cs 2.910.l(-5) 1.61 1 0.06(-3 ) 1.78 0.04 3) 3.5 1.79 1 0.08(-3 0.2(-5) )1.85 1 0.04(-3) 2.62 0.09(-3) 1.41 0.05 -3) 13eCs 3.9 1 0.1(-2) 4.0 1 0.2(-2) 4.1 1 0.1(-2) 3.9 0.1(-2) 3.8 1 0.1(-2) 3.6 0.1(-2) 139Cs 2.110.6(-2) 1.610.3(-2) 2.1 0.4(-2) 1.0 0.7(-4) 6 5(-4) 1.2 0.2(-2) l 24Na 5.3 0.1(-3) 5.6 0.2(-3) 6.3 **t 0.2(-3) 6.0 0.2(-3) 3.6 0.1(-3) 4.2 1 0.3(-3) 41Ar 51Cr 7 2(-5) <3(- 5) 411(-5) 6 3(-6) *
<8(-6) i s4Mn 2.2 0.4(-5) 8 i 5(-5) 1.9 0.5(-6) 1.5 0.7(-6) 2.1 0.2(-5) ssMn <6(-4) <5- <5 - <5(-4) 6.9 0.4(-3) <4(-4) 59Fe 5 2(-6) <4 - <3- <3(-6) 2 1(-4) 5 1 3(-6) 57Co 2.210.4(-6) <5- <3- 2 1(-6) <2(-4) <5(-6) 58C0 3.9 i 0.2(-4) 5.6 i 0.3(-4) 1.22 0.03(-4) 6 2 4(-4) 6.5 0.4(-3) 1.5010.06(-4)
TARIF B.4 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #4 POWER OPERAT!ONS 1/22/78; 11:16 1/23/78; 11:1A 1/24/78; 10:31 Nuclide 1/25/78; 04:14 1/26/78; 12:55 2/3/78; 09:43 (uCi/ml) (vCi/ml) (vCi/ml) __ (uC1/ml) (uci/ml) (vCi/ml) 60Co 7.010.3(-5) 3.0 1 0.6(-5) 8.2 0.9(-6) 3.2 0.6(-6) 65Zn <5f-6 2 2(-5 ) 5.4 2 0.8(-4) 2.0 0.2(-5)
<3(-6) 1.6 1.1(-6) 91S -
<3( ,4,); <2( ,4} <2( ,4,)
1.2 2 1.0(-4) <5(-6)
<2( ,4) <2( ,4) <1 (-4,)
93Y <2(-5) <1(-4) <3(-4) <3(-4) 952r 1.1 0.1 - 4.1 0.9 <2(-4) <1(-4) 1.0 2 0.3 - <3(-6) 3 1(-6) 3.2 0.8(-4) 9sNb 4 2(-5)(-5) 2 1(-6) 4 7 2(-6) 99Mo '1.0 1 0.2 - 2(-6) <6(-4) 5 3(-6) 92Tc 1.60 1 0.08(-4) 1.6 0.1(-4) 1.23 0.07(-4) <5(-4) 4.2 0.3(-5) t t t lo 3Ru t t
<6(-6) <3(-5) 2 1(-6) t
- 103mRh t t 4 3(-6) <2(-4) <9(-6) t t lo6Ru <5(-6) <3(-5) t t
, 106Rh <5(-6) <4(-6) *
<9(-6) t t t f., 110 mag t t e 124Sb
<3(-4) <7(-4) <4(-4) <4(-4'
- t ,
5 2(-6) 1.7 i 1.0(-5) 1.1 <3(-4) ' 125Sb 4 41,-6) 0.5(-6) 1.2 1 0.6(-6) <4 - 2 2(-6)
<3(-5) <6 - 1 <4 - <2 - <6 -
129mTe <4 - i <2(-5) <4-h <3- <2 - 129Te <8 - h
- 4-
<9 - h <7- <7 - <7 -
131mTe <5 - ) <2(-4) <3-) <2- <3 - <6 - 131Te ** ** ** ** ** ** 132Te <6(-6) <2(-4) i 139Ba 2.5 0.6(-3) <8(-6) <3(-5) *
<2(-5) 2.9 0.4(-3) 2.8 0.4(-3) 3.9 0.3(-3) 4.2 140Ba 4.1 0.5 0.2(-3) 7.6 0.3(-3)
<3(- <8(-6) 1.5 ! 0.5(-5) <6-4) 3.5 0.3(-4) 5 1 5(-6)(-5) 140La <31 ,-
141Ce
<3(-5) <2 -4) <2 -4) <1 -
<4 (-6) <3(- 3 2(-6) <5 - <2 -4 t 143Ce <2(-5) <5
<5 -
<3(-4) <2 - <2 -4 <4 -
144Ce 1.2 0.4(-5) < 3(I -
<3(-6) <4 -
144Pr t
<5(-4 <4 -
i t t t 152Eu t t
<4(-5) <3(-5) <3(-5) <5(-5) 154Eu <4(-5) <6(-4) <7(-6)
<3(-5) <3(-5) <5(-5) <3(-4) <5(-6) issEu ** ** ** ** **
187W <2(-5) 339Np
<5(-5) <3(-4) <2(-4) <2(-4) <6(-5)
<9(-6) <5(-5) <3(-5) <9(-5) <2(-4) <2(-5) t O Radionuclide not directly measured. Concentration can be inferred from parent or daughter.
0* Radionuclide not detected. Radionuclide not measured. e i
TABLE B.4 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #4 POWER OPERATIONS 2/7/78; 09:50 2/8/78; 09:55 2/9/78; 11:04 3/17/78; 10:36 3/25/78; 19:35 3/25/78; 19:16 Nuclide (pC1/ml) (pC1/ml) (vC1/ml) (uCi/ml) (uci/ml) (pC1/ml) 85mKr 1.6 0.1(-2) ** ** ** 85Kr <2(-4) ** ** ** ** 87Kr 2.3 0.2(-2) ** ** ** ** 88Kr 3.3 0.1(-2) ** ** ** ** 89xp * ** 131mXe <2(-3) ** ** ** ** ** 13 EXe 7.3 0.2(-3) ** ** ** ** 133Xe 2.25 0.05(-1) ** ** ** ** 135mXe 2 2(-2) ** ** ** ** ** 135Xe 1.08 0.03(-1) ** ** ** ** 137Xe ** ** ** ** 138Xe 4.4 0.1(-2)
?' e4Br 5.2 Es 131I 7.0 0.2(-3) 6.4 0.4(-3) 6.6 0.1(-3) 3.9 0.1(-3) 5.0 0.1(-3) **
0.1(-3) 132r 1,4510.03(-2 1.46 0.02(-2 133I 1.37 1 0.05(-2 1.16 0.02(-2 1.21**1 0.02(-2) 1.66**1 0.03(-2) 1.24 0.03(-2) 1.22 0.02(-2) 134I 1.72 1 0.04(-2 1.66 0.07(-2 ** ** ** ** 135I 1.42 0.06(-2 1.44 0.03(-2 esRb 5.4i0.4-2) 5.4 0.3 - ** ** ** ** 89Rb 1.4 1 0.1 -2) 1.110.1-8.7 1 0.9 - 8.0 1 0.2(-4) 1.42 0.04(-3) 1.47 0.03(-3) 1.5210.02(-3) 134Cs 1.4 i 0.1 -3) 4.2 0.5(-5) 3.9 0.6(-5) 3.3 0.5(-5) 136Cs 4.110.8-5) 2.4 1 0.6 - 2.3 0.6(-5) 1.6310.04(-3) 2.66 0.06(-3) 2.9 0.1(-3) 2.91 0.06(-3) 137Cs 2.210.1-3) 1.53 0.04(-3) 13sCs 4.4 0.1 -2) 4.010.2(-2) ** ** ** ** 139Cs 5 1(-2) 2.6 1 0.6(-2) 5.2 0.6(-3) 5.0 0.3(-3) 7.6 0.4(-3) 6.0 1 0.4(-3) 5.7 0.6(-3) 24Na 5.6 0.4(-3) ** ** ** ** ** 41Ar <6(-4) . 1.7 0.2(-4) <2(-4) <3(-5) 51Cr 1.9 0.4(-4) <4(-6) <6(-3) 1.3 0.3(-5) 54Mn 7.6 0.3(-5) 7.310.4(-6) 1.4 0.1(-5) 1.7 0.2(-6) 7.6 0.6(-5) ** 56Mn <7(-4) <6(-4) 59Fe 3.7 0.4(-5 6 1(-6) 3.3 0.6(-6) 2.1 0.5(-5) 4 i 1(-5) <2(-5) 1.3 0.9(-6) 5 3(-6) 1.4 0.6(-5) <2(-5) 57CO l.l i 0.4(-5 <2(-6) 1.15 i 0.04(-4) 1.62 0.04(-3) 5.1 1 0.1(-3) 7.4 i 0.2(-4) 58Co _ _ _ _ _ _ _ _ 1.810.6(-3 7.3 0.5(-5)
TABLE B.4 (cont'd) RADIONUCLEDE CONCENTRATIONS IN REACTOR COOLANT - UNfT 04 POWER OPEPATIONS 2/7/78; 09:50 2/8/78; 09:55 2/9/78; 11:04 3/17/78; 10:36 3/25/78; 19:35 3/25/78; 19:16 Nuclide (pCf/ml) (pCi/ml) (vCi/ml) (pC1/ml) (uC1/ml) (uC1/ml) 60Co 1.710.1(-4) 1.0 0.1 (- 5) 1.8 0.l(-5) 4.9 0.7(-5) 1.6 0.1(-4) 1.9 0.4(-5) ssZn <8(-6) <2(-6) <2(-5) <1(-5) <2(-5) <2(-5) 91Sr <4(-4) **
<2(-4) <2(-3) <4(-3) <3(-3) 91my ** ** ** ** **
<5(-4) **
93Y <2(-4) <2(-3) c4(-3) <3(-3) l ssZr 5.2 0.5 - 1.311.1(-6) <2(-6) 3.2 0.5(-5) 9 1(-5) 4.1 0.9(-5) 6.8 0.3 - 95Nb 2 1 (-6) 2.0 0.4(-6) 2.2 0.4(-5) 1.1 0.1(-4) 1.7 0.5(-5) 99Mo 1.0 0.1 - 4.9 0. 2(-5) 4.3 0.2(-5) 2.02 0.05(-4) 1.17 0.09(-4) 1.11 1 0.08(-4) 99mic t t t t t t lo 3Ru 1.1 0.2(-5) 1.9 0.7(-6) <3(-6) 1.0 0.3(-6) 2 2(-5) <3(-5) 103mRh t t t t t t lo6Ru <2(-5) <4 (- 6) <5(-6) <3(-5) <4(-6) <3(-5) lo6Rh t t t t t t i' 110 mag <7(-4) <3(-4) <3(-4) <8(-4) <1(-3) <7(-4) S 124Sb 1.0 0.3(-5) 8 5(-7) 1.0 0.5(-6) 3.7 0.7(-5) 1.0 0.1(-4) 1.5 0.7(-5) 12sSb <2(-5 <3(-6 <5(-6) 1.9 0.7(-5) <3(-5) <3(-5) 129mie <9(-6 <3- <4(-6) <2(-5) <3(-5) <2(-5) 129Te <9(-3 <9 - ** ** ** ** 131mTe <2(-4 <4 - <4(-5) <9(-5) <1(-4) <7(-5) 131Te ** ** ** ** ** ** 132Te <4(-5) 4 3(-6) <6(-6) <2(-5) 3.5 0.9(-5) <4(-5) 2.5 ** ** ** ** 139Ba <5(-3) 0.6(-3) 140Ba 1.5 0.1(-4) 8.3 0.5(-5) 9.4 0.6(-5) <2(-5) 4 2(-5) <4(-5) 140La 2 2(-4) <2(-4) <3(-4) <2(-5) <4(-5) <3(-5) 141Ce 2.510.4(-5) <3(-6) <3(-6) <1(-5) <4(-5) <2(-5) 143Ce <2(-4) <4(-5) <3(-6) <9(-5) <2(-4) 4 4(-5) l 144Ce <2(-5) <3(-6) <2(-6) <7(-6) <2(-5) <2(-5) I 144Pr t t t t t t l 152Eu <3(-5) <4(-6) <6(-6) <2(-5) <4(-5) <4(-5) 154Eu <5(-6) **
<2(-6) <2(-6) <6(-6) 1.0 0.7(-6) 9 6(-6) 155Eu ** ** ** ** **
187W <4(.4) <2(-4) <8(-5) <2(-4) <3(-4) 5 2(-4) 239Np 5 2(-5) <2(-5) <7(-6) <2(-5) 7 i 2(-5) <3(-5) t Radionuclide not directly measured. Concentration ca.. be inferred from parent or daughter.
- Radionuclide not detected.
** Radionuclide not measured.
TABLE B.a (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #4 POWER OPERATIONS 4/9/78; 09:03 4/9/78; 09:04 4/12/78; 17:53 4/13/78; 11:38 4/14/78; 13:55 4/17/78; 10:45 Nuclide (pCi/ml) (pCi/ml) (pC1/ml) (pC1/ml) (uCi/ml) _ (uti/ml) 85mgp ** ** ** ** ** ** 85Kp ** w* ** ** ** ** 87Kr ** 88Kp ** ** ** ** ** 89Kr ** 131mXe ** 133mXe ** 133Xe ** ** 135mXe ** ** 135Xe ** ** 137Xe ** ** ** 138Xe 84Br 1.9 i 0.5(-3) <2(-3) 2 1(-3) <3(-3) <2(-3) <2(-2) [ 131I m 8.0 1 0.1(-2) 7.8 0.1(-2) 9.7 0.6(-2) 1.01 0.05(-1) 6.6 1.04 0.06 - 132I 2.27 1 0.03 - 2.25 0.04 - 2.45 0.05(-2) 2.17 0.04(-2) 1.89 0.4(-2) 0.05(-2 ) 1.88 2 0.03 - 133I 5.21 0.08 - 5.07 0.08 - 5.0 0.2(-2) 5.2 0.1(-2) 3.3 1.78 0.04 - 134I 2.16 0.05 - 2.21 0.04 - 2.18 0.04(-2) 2.12 0.03{-2) 2.01 0.2(-2)) 0.03(-2 2.00 0.06 - 135I 3.02 0.08 - 2.95 0.08 - 3.05 0.06(-2) 2.84 0.07s-2) 2.0 0.2(-2) 1.76 0.04 - 88Rb 5.5 0.2 - 5.5 0.6 - 5.0 1 0.2(-2 8 i 1(-2) 6.2 0.4 - 5.0 89Rb 1.2 0.1 - 1.1 0.1 - 1.2 0.1(-2 8 2(-3) 9.8 0.6 - 1.14 0.2(-2)) 0.03(-2 134Cs 1.8 1 0.1 - 1.7 0.1 - 1.73 0.08(-3) 1.73 0.08(-3) 1.8 0.2 - 1.8 0.1(-3) 136Cs 1.710.2- 2.0 0.2 - 2.7 0.5(-4) 2.4 1.45 0.05(-4) 2.5 3.2 0.1 - 3.15 i 0.09(-3) 3.06 0.08(-3) 3.16 0.4(-4)) 0.09(-3 3.44 0.08(-3) 3.37 0.3(-5) 0.09(-3 ) 137Cs 138Cs 4.9 i 0.1 - 4.6 1 0.2(-2) 5.1 1 0.2(-2) 7 2(-2) 5.1i0.2(-2) 4.5 0.2(-2) 139Cs 1.6 1 0.3(-2) 311(-2) 1.5 0.4(-2) <2(-1) <3(-2) 8 3(-3) 24 a 2.45, i 0.09(-2) 2.4210.08(-2) 2.420.1(-2) 2.39 0.08(-2) 1.64 0.09(-2) 1.17, , 0.03(-2) SICr 1.710.6(-4) <2(-4) <3(-4) <2(-4) <2( 4) <4(-5) 7.8 0.6(-6) 1.4 0.1(-5) 5.3 0.7(-6) 2.1 0.1(-5) 7.6 0.6(-6) 54Mn 1.56 1 0.08(-5) <2(-3) 56Mn <8(-4) <8(-4) <2(-3) <2(-3) <1(-3) 6 i 1C-6) 7 2(-6) 2 2(-6) 1.6 1 0.2(-5) 2.2 1 0.9(-6) 59Fe 8 i 2(-6) <3(-6) <3(-6) <2(-6) 57Co <4(-6) <5(-6) <3(-6) 3.7 0.1(-4) 2.6i0.2(-4) 1.14 0.03(-3) 1.8 0.3(-4) 58Co 1.4 1 0.1(-3) 7.9 0.6(-4)
TABI E B.4 (cont'd RADIONUCLIDE CCNCENTRATIONS IN REACTOR COOLANT - UNIT #4 I POWER OPERATIONS ! 4/9/78; 09:03 4/9/78; 09:04 4/12/,8; 17:53 4/13/78; 11:38 Nuclide 4/14/78; 13:55 4/1',. (uCi/ml) (uCi/ml) (uC1/ml) (uC1/ml) (uCi/ml) (u, 60Co 3.7 0.3(-5) 2.5 0.4(-5) 4.0 0.2(-5) 1.4 0.1(-5) 4.5 0.4(-5) 1.4 0.1(-5) 65Zn 321(-6) <2 - <4 - 2.9 0.8(-6) <3 - <2 - ; 91Sr <3 - <3 - <4 - <3 - <2 - ' 91mY <4 - <2 - <5 -
<2 - <3 - <2 - <2 -
93Y <2 - <2 - <2 - <2 - l 95Zr 3.7 9 4(-4) <2 - l 0.6 - 1.1 0.1 - 5 1(-6) 1.0 0.2 - 1.8 0.2 - 2 1(-5) 95Nb 2.8 1 0.2 - 1.6 0.1 - 8.4 0.8(-6) 8.3 0.9 - 2.310.2- 3.4 99Mo 1.3 0.2 - 1.1 1.2 1.3 0.2 - 0.7(-6) 0.2 - 0.2(-4) 1.5' 1 0.3 - 1.5 0.2(-4) 998Tc t t t t 2 t 10 3Ru 312(-6) 4 1(-6) <4(-6) <3(-6) 9 1 2(-6) <3(-6) 103mRh t t t t - t t 10ERu <3(-5) <2(-5) <7(-5)
,106Rh <1(-4) <5(-5) <2(-5) , t t t t t t i n'o 11omAg 3.5 0.8(-6) 211(-6) 7 1(-6) 3.5 t 0.8(-6) 6 1(-6) <2(-6) u> 124Sb 2.6 0.2(-5) 2.6 1 0.9(-5) 4.3 0.7(-6) 5.2 0.7(-6) 1.8 0.2(-5) 2.5 0.6(-6)
- 12sSb <2 - <8 - <9 - <8(4' <1 -5) 129mTe <6 - <6 - <9 - <6(-6) i
<2( ~ <9-5) <7(-5) 129Te <5 - <4 - <2 - <3 - <2 -2) <7(-2)' i 131mTe 1.6 1 0.8(-3) <3- <4 - <3 - <2(-4) <3 -
l 131Te <9(-4) <2 - <2(-3 <3 - <2(-3) <2 - ; 132Te <3(-5) <3 - J <2(-5 <3 - 1.9 0.9(-5) <7 - l 139Ba 4.6 0.5(-3*; 3 [-3) 4.3 0.6(-3) 9 2(-3) 7 3(-3) 5.5 0.4(-3) 140Ea <2 - <2 - <2 - <2 - <2 - 2.9 0.6(-5) 140La <3 - <3-)> <4 - <2- <1 - <3-5) o 141Ce <7 - <6 - <8- <8 - <6 - <3 -6 143Ce <8 - <2 - <2 - <5 - <4 - <6 -5 144Ce <3 - <3 - <4 - <3 - <3- <2(-5 144Pr t t t t t t 152Eu <3- <9(- <2(-4) <8 - <2 - 4 154Eu <2- <21 -
<3(-6) <2t;-6)<21-4) <2 - <2 -
1ssEu <9 - <2 '- <2 - <2q- <1 - <6 - 187W <2 - <2 - <4 - <2I <6 - <2 - 239Np <1 - <2 - <2 - < 8 (, - <5-5) <4(-5) t Radionuclide not directly measured. Concentration can be inferred from parent or daughter.
- Radionuclide not detected.
** Radionuclide not measured.
}
TAELE B.4 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #4 POWEP OPERATIONS 4/17/78; 10:46 4/25/78; 10:15 5/1/78; 09:09 5/9/78; 09:21 5/16/78; 10:12 5/23/78; 10:30 Nuc1 Me, (uCi/ml) (pCi/ml) (pC1/rl) (pC1/n1) (pCi/ml) (pC1/ml) esmAr 1.5 0.1(-2) 2.2020.03(-2) 1.8 0.3(-2) 1.91 0.03(-2) 1.61 0.04(-2) 1.6 0.2(-2) esKr <9(-3) <5(-2) <8(-4) <2(-2) *
<2(-2) e7Kr 2.1 0.1(-2) 3.0 0.1(-2) 2.7 0.2(-2) 2.8 0.1(-2) 2.5 0.2(-2) 2.5 0.2(-2) 88Xr 3.2 0.1(-2) 4.76 0.09(-2) 4.12 0.08(-2) 4.43 0.06(-2) 3.63 0.05(-2) 3.70 0.08(-2) 89Kr * * * *
<2(-3) 131mXe <3(-3) <4(-3) 4 4(-5) 4 1(-4) <1(-3) <2{-2)
<2,-2) 133mXe 8.020.2(-3) 1.3 0.1(-2) 7.0 0.2(-3) 6.1 0.6 - 5.7 0.7(-3) 6.4 0.4(-3) 133Xe 3.0 0.1(-1) 6.1 0.3(-1 ) 2.3 0.2(-1) 2.0 0.2 - 1.9 0.1(-1) 1.97 0.04(-1) 135mXe <2(-2) <9(-2) 3 3(-2) 3.2 0.8 - <5(-2) <4(-2) 13sXe 1.1 0.1(-1) 1.5 *1 0.1(-1) 1.4 0.1(-1) 1.28 0.07( 1) 1.12 0.02(-1) 1.15 0.06(-1) 137Xe 6 5(-3) 1.9 0.4(-2) 13eXe 2.9 i 0.1(-2) 3.7 0.6(-2) 4.7 0.2(-2) 4.0 , 0.2(-2) 4.3 0.2(-2) 4.1 i 0.1(-2)
[ 84Br 2.0 0.6(-3) 5 2(-3) 3.8 0.7(-3) 5 1(-3) 2.2 0.9(-3) 3.6 0.6(-3)
- 131I 1.03 0.02 - 2.6 0.1(-2) 1.06 0.01 -2) 8.2 0.1 (-3) 7.4 0.2(-3) 7.1 0.2(-3) 132I 1.90 0.05 - 2.46 0.08(-2) 1.94 0.02 -2) 1.83 0.04-2) 1.98 0.02(-2) 2.06 0.03(-2) 133I 1.77 0.03 - 3.19 0.07(-2) 1.73 0.03-2) 1.72 0.04 -2) 1.8 0.1(-2) 1.68 0.07-2) 134I 1.93 0.04-2) 2.24 0.03(-2) 2.18 0.06-2) 2.0610.03-2) 2.39 0.03(-2) 2.43 0.05 -2) lasI 1.85 0.05(-2) 2.46 0.09(-2) 1.89 0.05 -2) 1.83 0.03-2) 1.96 0.03(-2) 1.98 0.03-2) serb 5.2 0.2 (-2) 1.1 0.3(-1) 6.2 0.5(-2) 6.6 0.8(-2) 6.2 0.5(-2) 5.1 1 0.2(-2) 89Rb 1.14 1 0.02(-2) 1.7 0.2(-2) 1.38 0.05(-2) 1.2a 0.08(-2) 1.63 0.07(-2) 1.47 0.04(-2) 134Cs 2.15 0.07(-3) 1.65 0.04(-3) 1.61 0.05(-3) 1.90 0.05(-3) 1.65 0.09(-3) 3.1 0.1 -3) 136Cs 3.9 3.9 0.2(-5) 2.4 0.5(-5) 4.8 0.5(-5) 3.4 137Cs 3.92 0.2(-5) 0.09(-3 ) 3.08 1 0.09(-3) 3.0 0.1(-3) 3.3 0.2(-3) 3.16 0.3(-5))
0.07(-3 5.210.4-4} 4.5 0.1 -3f 13eCs 4.8 0.2(-2) 6.2 0.5(-2) 7.8 0.2(-2) 5.6 0.2(-2) 5.3 1 0.2(-2) 5.0 0.1(-2) 139Cs 1.1 i 0.2(-2) <2(-1) 42 1(-2) 2 2(-3) <4(-2) 2.0 0.3(-2) 24Na 1.15 0.03(-2) 1.17 0.03(-2) 8.6 0.3(-3) 9.2 0.3(-3) 8.5 0.2(-3) 8.4 0.4(-3) 41Ar <2(-4) <3(-4) <2(-4) <2(-4) 51Cr 6 1 3(-5) 1.1 C .1 - 1.8 0.1(-3) 2.3 0.4(-4) 5 2(-4) 5.5 0.4(-4) 54Mn 2.2 1 0.4(-5) 3.6 C3- 5.0 0.3(-4) 7.0 0.3(-5) 2.0 0.2(-4) 4.310.3(-5) 56Mn <2(-3) 4.2 2 0.4 - <6(-4) <6(-4) <1(-3) <2(-3) 59Fe 1.0 1 0.3(-5) 2.1 0. -4 3.2 0.2(-4) 3.1 0.4(-5) 7.8 0.5(-5) 2.4 0.5(-5) 57C0 <4(-6) 1.0 0.1-5)) 1.9 0.5(-5) 4 i 2(-6) 4 2(-6) 1.3 0.5(-6) seCo 3.5 t 0.2(-4) 5.2 0.4 -3) 8.5 0.4(-3) 1.55 0.04(-3) 2.9 0.3(-3) 1.50 0.07(-3)
- - - - - - - - - - - - - - - - - - - - - - - - ~ ~
TABLE B.4 (cont'6J) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT - UNIT #4 POWER OPERATIONS 4/17/78; 10:46 4/25/78; 10:15 5/1/78; 09:09 5/9/78; 09:21 5/16/78; 10:12 5/23/78; 10:30 Nuclide (pCi/ml) (pC1/ml) (uCi/ml) (pCi/ml) (pC1/ml) (pC1/ml) 60Co 8 2(-5) 3.5 0.5(-4) 4.6 3.3 0.1(-4) 652n 0.9(-4) 2.9 0.2(-4) 2.0 0.5(-4)
<4(-6) 1.5 1 0.2(-5) 2.5 0.9(-5) <7(-6) 1.7 0.5(-5) 91Sr <3 - <3 - <1(-4) 2 2(-4) <2(-4) <3(-4) <3(-4) 91mY <2 - <4 - *
- 9 3Y <2 - <1 - <4(-4) <2(-4)
<3(-4) <3(-4) <1(-3) <2(-3) 95Zr 1.410.2-5) 1.2 0.1(-4) 1.6 0.2-4) 7.6 0.3(-5) 4.4 1 0.5(-5) 1.2 1 0.1 -
95Nb 1.9 0.3 -5) 1 .31 0.07(-4) 2.2 0.2 -4) 1.11 0.05(-4 9 1(-5) 1.4 0.4 - 99Mo 1.4 0.1 -4) 2.1 0.2(-4) 7.1 0.6 -5) 1.1 0.1(-4)) 1.7 0.2(-4) 1.5 0.1 - 99mic t t t t t t 10 3Ru 6 3(-6) 1.6 2 0.2(-5) <5(-5) 1.3 0.4(-5) 8 4(-6) 6.0 1 0.3(-5) IoamRh t t t t t t 106Ru <3(-5) <6(- 5) <5(-5) <5(-5) <6(-5) <3(-5) to 106Rh t t t t t t A, 110 mag 5 i 2(-6) 1.4 1 0.3 - <2(-3) 2.220.3(-5) 3 2(-5) 1.6 0.1(-5) 12a.Sb 6 2(-6) 4.0 0.3 - 8.3 0.9(-5) 1.9 0.5(-5) 2.4 0.5(-5) 6.5 1 0.3(-5) 12sSb <2 - 1.6 0.9 - <3 - <3(-5) <2(-5) <6 - 129.Me <8 - <9 - <3 - <2(-4) <2(-4) <7 - 129Te <3 - <6 - <1 - * *
<3 -
131mTe <2 - <3 - <2 - <2(-4) <1 - <1-3) 131Te <9 - <4 - * *
<2- <8 -4) 132Te <2 - <2 - 2 2(-5) <9(-6) <2 - 1.1 i 0.2(-4) 139Ba 5.1 1 0.3(-3) 9 i 4(-3) 6.3 0.1(-3) 6.8 0.1 8 5(-3) 4 1(-3) 140Ba 6 1(-5) <2 - <5(-5) 4 2(-5)(-3) <4(-5 7.6 0.3(-3) 140La <2(-4) <1 - <2(-3) 1.2 0.3(-4) <1(-4 3.4 1 0.1(-3) 141Ce <5(-6) <6- 111(-5) <8 - <9 - <8 -
143Ce <6(-5) <9 - <8(-5) <6 - <3 - <3 - 144Ce <2(-5) <4(-5 <3(-5) <3- <3 - <6 - 144Pr t t t t t t 152Eu <3- <3 - <3(-5) <4 - <3 - <3(-4) is4Eu <4 - <4 - <2(-5)
<7 - <7-{ <7-6) 15sEu <l - <3- <2 - <3 - ) <9 -6) 187W <2- 7 2(-4) <2(-4) <3- <4(-4) <4-4) 239Np <6 - <2(-4) <3(-5) 6 4(-5) <1(-4) <9 -5) 1 t Radionuclide not directly measured. Concentration can be inferred from parent or dauahter.
- Radionuclide not detected.
** Radionuclide not measured.
TABLE B.5 BETA-ONLY-EMITTING RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT UNIT #3 - DURING REFUELING 11:48* 11:32** 14:35*** 11:36** 12:24*** 12/2/77 12/30/77 1/3/78 1/11/78 1/11/78 Nuclide (uC1/ml) (uC1/ml) (pCi/ml) (pCi/ml) (pCi/ml) 3H 3.8 0.1(-2). 1.70 0.05(-2) 5.1 0.2(-3) 4.5 0.1(-3) 4.4 0.1(-3) 14C 1.0 1 0.2 - 1.9 0.2(-6) 2.7 i 0.3(-7) 4.7 0.5(-7) 8.5 0.9(-8) 55Fe 1.5 i 0.1 2.7710.04(4) 4.6710.06(-3) 3.93 0.02(-4) 5.21 0.03(-4 63N1 1.6 0.1 - 3.7 0.2 -4 1.20 0.01(-3) 6.1 0.1(-4) 7.4 0.3(-4) )
- 89Sr 1.01 2 0.01( 4) 5.8 0.2 -6 8.1 0.2(-6) 4.00 0.08 -6) 1.26 0.02(-5) 90Sr 2.66 0.07(-6) 9.5 8.0 0.1(-7) 2.62 0.03-6) 2.53 0.03(-6)
{l 91Y 9.9 0.4(-6) 1.39 0.4-7)) 0.01(-5 2.06 0.02(-5) 1.60 0.02 -5) 9.9 i 0.1(-5) os ;
- Sample obtained from inlet to CVCS.
** Sample obtained from RHR
*** Dip sample obtained from reactor cavity.
1
TABLE B.6 BETA-0NLY-EMITTING RADIONUCLIDE CONCENIRATIONS IN REACTOR COOLANT UNIT #3 - POWER OPERATIONS AFTER REFUELING 10:10 09:40 4/25/78 6/1/78
- Nuclide (pC1/ml) (uCi/ml) 3H 1.65 1 0.08(-1) 1.20 0.04(-1) 14C 2.5 0.3 - 6.7 i 0.7(-5) 32P 3.0 1 0.2 - 4.37 0.03(-3) 55Fe 2.1 0.2 - 9.7 1 0.5(-6 63Ni 9.1 1 0.8 - 1.9 0.2(-6 895r 2.11 1 0.04(-5) 1.76 1 0.02( 5) 90Sr 4.1 1 0.4 -7 1.7 0.4 -8 91Y 5.9 i 0.7 -7 5.8 0.2 -7 l
a i B-27
- ,..._,.r _ , , , , ,m ,. , , .-,,
- - - - - - %__ u----m. -- 4 as., , w a en _ -.a.. - - e.-4#e_',44 -4.,_4, . - --+e.eia., - . + .- M1-14--mu - -% r A__ - + -=4L-- E--w- uL 4 .* e- e- * =* h-I TABLE B.7 ,.
BETA-ONLY-EMITTING RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT UNIT #4 - POWER OPERATIONS 16:10* 09:29 09:33* 10:25 11/30/77 12/2/77 12/12/77 4/24/78 Nuclide (uC1/ml) (pCi/ml) (uCi/ml) (uci/ml) 3H 1.60 i 0.05(-1) 1.54 0.01(-1) 1.48 0.01(-1) 2.00 0.06(-1) 14C 3.2 1 0.3(-6) 1.22 0.03 - 1,41 0.03(-6) 9.5 1.0(-6) 32P t 1.58 0.05 - 3.6 i 0.1(-5) 8.2 1 0.2(-3) 55Fe 3.80 0.06 - 4.47 1 0.06(-6) 63Ni 5.3 1.49 i 0.5(-5) 0.04(-5 ) 1.68 0.07 - 1.25 1 0.07(-6) 9.8 1.76 i1 C.1(-4) 0.06(-4 ) ; ao 89Sr 3.2i0.1(-5) 4.810.5(-7) 5.1 0.2(-6) 5.9 0.2(-6)
/3 90Sr 6.3 0.6(-7) 2 1 1(-8) 8 1 2(-8) 4.310.5(-7) oo 91Y 4 2(-7) 5 1 4(-8) 1.1 0.4(-7) 1.35 0.09(-6) 4
- Sample obtained from inlet to CVCS.
t Sample not analyzed soon enough to detect this radionuclide. j
TABLE B.8 3H AND 131I CONCENTRATIONS IN REACTOR COOLANT UNIT #3 - FPL MEASUREMENTS , 3H 1311 Date (pCi/ml) Date Time (pCi/ml) 11/3/77 1.7 - 11/1/77 0200 1.2 - 11/4/ 77 1.8 - 11/8/77
- 2.0 -
11/5/77 1.9 - 11/12/77 0520 1.2 - 11/6/77 1.8 - 11/15/77 0105 1.1 - 11/7/77 1.8 - 11/22/77 0936 1.4-2) 11/9/77 1.5 - 11/24/78 0500 1.2 -1) 11/14/77 5.8 - 2/21/78 5.0 -3) 11/15/77 5.7 -2 3/6/78 1830 1.3 - 11/20/77 8.8(-2 3/7/78 0430 3.0-2) 11/21/77 8.4(-2 3/13/78 0910 2.0 -2) 11/28/77 5.1-6.9(-2) 3/13/78 1436 2.7 - 12/5/77 3/14/78
- 1.6 -
12/12/77 4.6-2l 4.3 -2., 3/17/78 0419 1.8 - 12/19/77 4.0 -2 3/17/78 1110 2.0 - 12/26/77 4.1 -2 3/21/78 0800 6.0 - 1/2/78 2.2 - 3/21/78 2120 6.2 - 1/9/78 5.3 - 3/25/78 Oa00 3.3 - 1/16/78 6.7 - 3/25/78 1125 2.2 - 1/23/78 6.0 - 4/4/78 0815 7.6 - 1/30/78 4.9(-3 4/11/78 8.8(-3) 2/6/78 2.8(-3 4/18/78 0835 6.5(-3) 2/13/78 7.0 - 4/20/78 0250 3.9 - 2/20/78 4.1 - 4/20/78 1015 6.6 - 2/27/78 1.1 - 4/26/78 1320 1.2 - 3/6/78 .l.9 - 4/26/78 2115 1.7 - 3/13/78 .1.9 - 4/28/78 1325 1.7 - 3/20/78 2.2 - 5/2/78 0835 6.4 - 3/27/78 1.2 - 5/9/78
- 6.S(-3) 4/3/78 1.7 - 5/11/78 0904 2.5(-2) 4/10/78 2.2(-1 5/11/78 1915 6.4(-2) 4/17/78 2.4(-1 5/12/78 0218 4.0(-2) 4/24/78 2.8 - 5/13/78 0025 2.0 -
5/1/78 3.2 - 5/13/78 0915 4.4 -2) 5/8/78 2.1 - 5/16/78 0849 7.4 -3) 5/15/78 1.1 - 5/20/78 0200 5.3 - 5/22/78 9.8 - 5/22/78 0925 3.1 - 5/29/78 1.1 - 5/22/78 1332 2.2 - 5/22/78 2045 1.4 - 5/23/78 1840 8.0 - 5/24/78 0235 8.1 - 5/30/78 0855 F.9 -
- Sample collection time not known.
B-29
TABLE B.9 3H AND 1311 CONCENTRATIONS IN REACTOR COOLANT UNIT #4 - FPL MEASUREMENTS 3H 131I Date (pCi/ml) Date Time (uCi/ml) 11/7/77 1.5 - 11/12/77 0525 2.8 -2 11/14/77 3.4 - 11/15/77
- 5.8 -
11/21/77 1.0 - 11/29/77 0835 7.8 - 11/28/77 1.6 - 12/1/77 0130 1.1 - 12/5/77 1.8 - 12/1/77 2200 3.0 - 12/10/77 1.5 - 12/9/77 0430 1.2 - 12/12/77 1.5 - 12/9/77 1100 9.8 - 1/2/78 1.5 - 12/13/77 0025 1.0 - 1/9/78 1.8 - 12/17/77 0945 1.9 - 1/16/78 2.2 - 12/17/77 1530 2.0 - 1/23/78 2.5 - 12/20/77
- 1.1 (-l 1/30/78 1.1 - 12/26/77 1400 1.4 -
2/1/78 1.3 - 12/27/77 0545 1.8 - 2/2/78 1.3 - 1/9/78 0630 1.0 - 2/4/78 1.4 - 1/17/78
- 7.0 -
2/6/78 1.5 - 1/25/78 0620 1.7 - 2/7/78 1.5 - 1/25/78 2200 2.0 - 2/8/78 1.6 - 1/31/78 0825 2.9 - 2/9/78 1.6 - 2/14/78
- 1.2 -
2/10/78 1.7 - 2/14/78 1730 1.7 - 2/13/78 1.9 - 3/10/78 0130 1.2 - 2/14/78 1.9 -1) 3/14/78 0832 3.6 - 2/20/78 8.8 -2 3/21/78
- 6.3 -
3/6/78 8.0 -2 3/28/78 0900 5.8 - 3/13/78 4.6 -2 4/11/78 0919 1.0 - 3/20/78 1.0 -1 4/18/78
- 9.3 -
3/27/78 1.6 -1) 4/25/78 0020 1.5 - 4/3/78 2.1 -1) 4/25/78 0250 1.8 - 4/10/78 2.1 -1) 4/25/78 0935 2.4 - 4/17/78 3.0 -1) 5/9/78 0920 7.6(-3 4/24/78 1.7 - 5/16/78
- 6.3(-3 5/1/78 1.6 - 5/23/78 0833 6.3(-3 5/8/78 2.4 - 5/31/78 0405 6.5(-3 5/15/78 1.7 - 5/31/78 1323 1.8(-2 5/22/78 3.1 -
5/29/78 3.2 -
- Sample collection time not known.
l l B-30
TABLE B.10 RADIONUCLIDE CONCENTPATIONS IN REACTOR COOLANT 1/24-26/78 SHUTDOWN OF UNIT #4 Time 10:31 04:14 05:04 06:20 07:16 Date 1/24/78 1/25/78 1/25/78 1/25/78 1/25/78 Power 100% <100% 65% 0% 0% Nuclide (uC1/ml) (uC1/ml) (vCi/ml) ___(pci/ml) (uC1/ml) 131I 6.0 i 0.4(-3) 6.3 0.3(-3) 7.5 2 0.4(-3) 2.0 0.1(-2) 1.29 0.02(-1) 1 32I 1.23 0.02 - 1.29 0.03-2) 1.32 0.03(-2 2.11 0.02(-2) E.3 0.1(-2) 133I 1.02 0.06 - 1.0710.06-2) 1.19 0.09(-2 4.0 ~ 0.3(-) 1.19 i 0.01(-1) 134I 1.38 i 0.04 - 1.3710.03- 1.3410.02(-2 1.30 0.03 -2 2.2 1 0.1(-2) 1351 1.2310.03- 1.26 0.03 - 1.28 2 0.02(-2 2.0 $ 0.1(- ) ) 7.8 0.1(-2) 8aRb 5.3 0.2(-2) 5.2 0.3-2) 4.9 0.1(-2) 5.3 0.2(-2) 4.4 0.2(-2)
, 89Rb 1.0 0.1(-2) 8.8 2 0.3 -3) 6.2 0.2(-3) 1.4 0.2(-3) <6(-4)
, , 134Cs 1.02 0.09(-3 3.5 0.2(-5) 9.6i0.6-4) 1.1 i 0.1(-3) 1.1 0.1(-3) 2.6 i 0.1 -3) 136Cs ) 3.3 0.3 -5) <3(-4) <5(-4) 9.9 1 0.6 -4) 137Cs 1.79 i 0.08 3) 1.85 1 0.04 3) 1.76 2 0.04( 3) 1.79 0.07(-3) 3.5 1 0.1 -3) 138Cs 4.1 1 0.1 - 3.9 i 0.1 - 3.5 0.1(-2 2.23 + 0.02(-2) 1.01 1 0.02(-2) 139Cs 2.1 0.4 - 1.0 i 0.7 - 1.0 0.3(-2 <6(-4{ <2(-3) 24Na 6.3 0.2 6.0 0.1(-3) 5.6 0.1(-3) 5.1 1 0.1(-3) 4.7 1 0.1(-3) 51Cr 4 i 1(-5)(-3) 6 i 3(-6) 1.0 0.7(-5) * <2(-4) 54Mn 1.9 i 0.5(-6) 1.5 2 0.7(-6) <8 - *
- 56Mn <5 <5(-4) <4 - 3 2(-4) <6 -
59Fe <3 - <3(-6) <3 - <4(-4) <1 - 57C0 <3 - 2 2 If-6) <2 - <2(-4) <3 - 5800 1.22 _ 0.03(-4 6i4(-4) 8 + 4(-4) 3.8 1 0.3 - 8.3 1 0.9(-3) 60Co 3.2 1 0.6(-6) <2{-4) 1.6 0.5 - <2(-4
- 65Zn 91Sr 8.2
<3(-6 i)0.9(-6) 1.6 i)1.1(-6) 9 6(-5) 1.2 0.9 - <3(-4
<2(-4) <2(-4) <2(-4) <3(-4) <3(-4 91my ** ** ** ** **
93Y <3(-4) <3(-4) <2(-4) <4(-4) <2(-3) 95Zr <3(-6) 3 i 1(-6) <2(-4) <2(-4)
<2(-4 95Nb 2 1(-6) 4 2(-6) *
<8(-4 99Mo 1.6 0.1(-4) 1.23 0.07(-4) <3(-4) <3(-4) <2(-4
TABLE B.10 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 1/24-26/78 SHUTDOWN OF UNIT #4 Time 10:2' 04:14 05:04 06:20 07:16 Date 1/24,/8 1/25/78 1/25/78 1/25/78 1/25/78 Power 100% <100% 65% 0% 0% Nuclide (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) 99mTc 2 ,1(-6) 4 ,,3(-6) <2(,,4) <2(,,4) <2(,,4) 106Ru <5 (-6)
<4(-6) 106Rh
* * <2 -
11ong9 <4(_4) <4(-4) 124Sb 1.1 0.5(-6) 1.2 1 0.6(-6) <2 - <3 - <3 - ca 12sSb <6 - <4 - <2 - <2 - <4 - 2, 129mTe <4 - <3 - <2 - <2 - <2 - N 129Te <9 - <7 - <7 - <5 - <2 - 131 e <3 ,- <2 - <2 - <3 - <5 - 132Te <8(-6) <3(-5) 139Ba 2.8 0.4(-3) 2.9 i 0.3(-3) 3.0 0.4(-3) 3.8 0.3(-3) 3.210.2(-3) 140Ba <8(-6) 1.5 0.5(-5) <6(-4) <7(-4) <2(-3) 140La <3(-5) <2(-4 <2 -4) <2(-4) <2(-4) 141Ce 312(-6) <5 - <2 - <2-4) <3(-4) 143Ce <3(-4) <2 - <2- <2-4) <4(-4) e <3(,,6) <4 <4 ,, <5 ,,4) <8(,,4) 152Eu <3 -5 <5 -5 <3- <3- <5 -
}sy <3 ;*5 #5 ;*5 <2 ;, <2 ;, < 2 ,,
187W <3(-4) <2(-4) <2(-4', 4 2(-4) 2 1(-4) 239Np <3(-5) <9 (-5) <9(-5) <2(-4) <2(-4)
- Radionuclide not detected
** Radionuclide not measured
TABLE B.10 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 1/24-26/78 SHUTDOWN OF UNIT #4 Time 08:18 09:17 10:24 10:05 12:55 D=t te 1/25/78 1/25/78 1/25/78 1/26/78 1/26/78 Power 0% 0% 0% 100% 100%
.Nuclide (uC1/ml) (uci/ml) (uti/ml) (uti/ml) (uci/ml) 131I 2.1 2 0.1(-1) 2.4 i 0.1(-1) 2.3 0.1(-1) 5.7 132I 0.1(-2) 4.7 0.1(-2) 1.23 2 0.03(-1) 1.17 0.03(-1) 9.2 1 0.1(-2) 1.54 0.01(-2) 1.43 0.03(-2) 133I 2.1 1 0.1(-1) 2.310.1(-1) 2.1 0.1(-1) 2.86 0.05(-2) 1341 1.7510.04(-2) 9.5 1 0.2(-3) 3.9 t 0.1(-3) 3.4 1.42 1 0.03(-2 0.1(-2) ) 1.40 0.06(-2) 135I 1.2310.02(-1) 1.25 0.02(-1) 1.04 1.59 0.01(-1) 0.04(-2) 1.54 1 0.04(-2) '
88Rb 3.4 0.1(-2) 2.6 1 0.1(-2) 1.7 0.1(-2) 5.4 0.2 - 4.7 0.2(-2) cm 89Rb <2(-3) <2(-3) 8.3 0.4 - 4, <1(-3) 7.8 0.3(-3) 134Cs 4.7 0.2 -3) 5.7 1 0.1 - 5.7 1 0.1 - 2.2 0.1 - 1.90 0.07(-3) USCs 2.8 1 0.2 - 3.6 0.2 - 3.8 2 0.2 - 9.8 0.5 - 137Cs 6.5 0.2 - 8.1 0.2 - 8.2 1 0.1 - 3.110.1- 6.810.5(-4)) 2.62 2 0.09(-3 13sCs 3.5 1 0.2 - 9.8 1 1.0 - <4(-4) 3.9 i 0.1 - 3.8 i 0.1(-2) 139Cs <2 (-3) <2(-3) <9(-4) 1 i 1(-3) 6 5(-4) 24Na 3.9 t 0.1(-3) 3.610.1(-3) 2.9 i 0.1(-3) 3.6 0.1(-3) 3.6 1 0.1(-3) 51Cr <3(-4) 5.4* i 0.8(-3) 3 i 1(-4) <2(-3)
- 54Mn * *
<6(-3)
- ssMn <5(-4) 4.6 1 0.3(-3) 2.9 <5 -
59Fe
- 0.3(-3) 6.9 t 0.4(-3)
<2(-3) <2(-3) <4 - 2 1(-4) 5700 <2(-4) <3(-4) <2(-4) <4 - <2(-4) seCo 1.010.2(-2) 1.89 0.08(-2) 1.58i0.07(-2) 2.0 0.2(-3) 6.5 0.4 -3) 60C0 1.5 1.0(-4) 5.910.6(-4) 5.2 i 0.4(-4) <6(-4) 5.4 1 0.8 -4) 652n <3 -4 <2(-4) <3(-4) <2(-4) 1.2 1.0 -4) 9 <3 -
g 1.0 0.9(-4) <2(-4) <2(-4) <2(-4) 93Y <2( 2) <2(-3) <8(-3) <3(-3) <2(-4) 95Zr 1.4 i 1.0(-4) 5.8 0.9 5.6 0.9(-4) <4(-4) 3.2 i 0.8(-4) 95Nb <8(-4) 4 1 3(-4)(-4) 3 3(-4) 3 2(-4) <6(-4) 99Mo 211(-4) 1.09 1 0.06(-3) 1.2 0.1(-3) 2 2(-4) <5(-4)
TABLE B.10 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 1/24-26/78 SHUTDOWN OF UNIT #4 Time 08:18 09:17 10:24 10:05 12:55 Date 1/25/78 1/25/78 1/25/78 1/26/78 1/26/78 Power 0% 0% 0% 100% 100% Nuclide (pCi/ml) (pC1/ml) (pCi/ml) (pCi/ml) (pCi/ml) 99erc ** 10 2 1, 1(-4) <2(-4) <2(-4) <2(-4) 3, 1 i] -4)
- 106Ru *
- 106Rh
- 11ongg <4 -3) <4 - <5 - <2-124Sb <2 - <2 - <2 - <2 - <4 -
m 125Sb <4 - <3 - <3- <3 - <2 - w 129mTe <3 - <3 - <2 - <3 - <2 - 129Te <8 - <4 - <4 -4 <7 - <7 -
<3 -3 <3 - <3 -
]l e <4 ,- <4 -
132Te <4(-4) 4.2 0.2(-3) 139Ba 2. i 0.4(-3) 1.8 i 0.2(-3) 1.5 0.2(-3) 5.1 0.3(-3)
<2 - <1(-3) <6 - <7 - <6 -
140Ba
<2 - ' i 7(-5) <2 - <2 - <2 -
140La
<4 - <3 -4 <3 - <2 - <2 -
141Ce
<6 - <7 -4 <5 - <3- <2 -
14%e 144Pr **
<1(-2) <2(-3) <6(-4) <6(-4) 152Eu <9(-4) <3(-4) is4Eu <2(-4) <2(-4) <2(-4) <2(-4) **
155Eu 187W 5 1(-3) 3.010.7(-3) 1.6 1.0(-4) <2(-4) <2(-4) 239Np <2(-4) 6 i 2(-4) <2(-4) 8 7(-5) <2(-4)
- Radionuclide not detected
** Radionuclide not measured -
TABLE B.11 RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT #3 Time 09:51 21:41 22:51 23:52 00:51 01:46 Date 5/16/78 5/19/78 5/19/78 Power 100% 5/19/78 5/20/78 5/20/78 i 70% 26% 0% 0% 0% Pressure (psi) 2240 2240 2240 2240 2260 Temp. (*F) 570 2260 565 553 548 540 F1owt (gpm) 540 54 52 52 50 50 50 Nuclide (uCi/ml) (uCi/ml) (uC1/ml) (uC1/ml) (uci/ml) (uci/ml) 85mKr 2.0 0.1(-2) 1.4 0.2(-2) ** ** 9.6 ** esKr ** ** ** ** ** 0.2(-3) 87Kr 4 i 1(-2) 2.5 0.9(-2) ** ** 6 4(-3) ** 88Kr 4.98 2 0.08(-2) 3.53 2 0.07(-2) ** ** 89Kp ** ,, 1.78 0.06(-2) ** 131mXe <2(-3) <8(-2) ** **
<2(-2) **
cn 133mXe 8.7 i 0.6(-3) <1(-2) ** ** 8 4(-3) ** d> 133Xe 3.1 1 0.3(-1) 2.7 0.1(-1) ** ** 2.9 m lasmXe 0.1(-1) **
<2(-1) <1 (-1 ) ** **
<6(-2) **
135Xe 1.6**i 0.1(-1) 1.320.1(-1) ** ** 1.23 1 0.03(-1) ** 137Xe ** ** ** ** ** 13eXe 1.3310.07(-1) 6.8 0.1 (-2) ** ** 7 3(-4) ** e4Br 2.2 i 0.2(-2) 1.3 i 0.1(-2) 4.8 i 0.7(-3) 2.4 7.8 0.4(-3) <1(-3) 814(-4) 1311 0.2(-3) 9.2 i 0.5(-3) 9.9 0.6(-3) 1.97 1 0.06 -2) 5.1 0.1(-2) 6.8 0.2(-2) 132I 1.02 i 0.03(-1 1.03
- 0.03(-1) 1.11 1 0.02 - 1.04 0.03 1) 0.05 -
1.3010.02-1) 1.41 1331 6.010.2(-2) ) 5.7 0.2(-21 5.87 0.04 - 6.1 0.1(- 9.94 2 0.06 -2) 1.15 i 0.02 - 134I 1.86i0.03-1) 1,61 0.03(-1 1.23 0.01 - 6.9 i 0.1(-2 4.17 0.05 -2 2.45 t 0.02 - 1351 1.06 i 0.02 -1) 9.9 0.2(-2) ) 9.2 t 0.1(- ) 7.8 0.2(-2 9.1 0.1(- ) ) 9.2 0.1(-2) 88Rb 8.0 1 0.7 - 5.8 i 0.1(-2) 6.9 i 0.2(-2) 3.7 0.1(-2) 3.1 0.1(-2) 2.1 0.1(-2) 89Rb 7.6 1 0.3 - 3.9 0.1 (-2) 1.8 0.1(-2) 1.2 0.4(-3) <9(-4) <9(-4) 134Cs 7.9 i 0.3 - 8 1(-4) 411(-4) 8 1(-4) 1.7 1 0.2(-3) 2.3 0.1(-3) 136Cs 1.0 0.1 - <4(-4) <4(-4) <2(-4) 5.5 0.7(-) 1.11 i 0.07(-3) 137Cs 1.1210.09-3) 7 2(-4) <9(-4) 7.0 0.8(-4) 1.5 0.1(- ) 2.1 0.2(-3) 138Cs 2.48 0.07 -1 1.60 1 0.04(-1 8.3 1 0.2(-2) 3.07 i 0.06(-2) 1.11 1 0.04 -2) 5.2 0.4(-3) 139Cs 1.2 i 0.5(- ) ) 6.510.4(-2) ) 3.1 0.3(-2) <5(-3) <5(-3) <5(-3)
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR C00t. ANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT #3 Time 09:51 21:41 22:51 23:52 00:51 01:46 Date 5/16/78 5/19/78 5/19/78 5/19/78 5/20/78 5/20/78 Power 100% 70% 26% 0% 0% 0% Pressure (psi) 2240 2240 2240 2240 2260 2260 Temp. (*F) 570 565 553 548 540 540 Flow t (gpm) 54 52 52 50 50 50 Nuclide (uCi/ml) (uCi/ml) (pC1/ml) (uCi/ml) (pCi/ml) (pCi/ml) 24Na 8.5 i 0.3(-3) 8.4 i 0.2(-3) 8.8 0.2(-3) 6.7 0.2(-3) 6.6 i 0.2(-3) 5.9**i 0.1(-3) 41Ar <6(-4) <6(-4) <7(-4)
- 51Cr <4(-5) 54Mn 1.7 i 0.2(-5) * * * <3 - <2 -
56Mn <6 -3) <6 -3 <5 - <3(-3 <2 - <1 -
<8-4) <9 - <7(-4 <7 - <1 -
59Fe <4-6) <7 - m 5700 <1 -6) <2 -4) <4 - 4(-4 <7 - 2.1 0.1(-4) 2.7 0.2(-3) <4 - 3.7 0.2(-3) <4 - <5 - h 58C0 60C0 1.88 0.08(-4) <2 - <2 - 3.8 i 0.9(-4) 1.3 0.6(-4) <2 - 6sZn <5 - <6 - <2(-4) <4(-4) <2 -
<4(-6 <3 -
91Sr <5(-4 <5 - <6 - <3(-4) <4(-4)
<3 - <3 - 1.6 1 0.7(-4) <2 -4) <2 -
91mY <3(-4 <2 - <3 - 9 3Y <2(-3 <4 - <3 - <2(-3) 1.9 i 0.3 -5 <3 - <4 - 2 1(-4) <3 - <3 - 95Zr
. ib04-3) <5(,- 3) <4(, 3) <3(, 3) <
,, 2.1,,0.3(-4)
<2(,,4) <2( ,4) <2(,,4) 103R 4 i,1(-6) <3( ,4) <3(,,4) 106
- 3.6,i 0.8(-5) ,,
11onMg 511(-6) <3(-4) <2(-4) <2(-4) 124Sb 6 1(-6) <3(-4) <4(-4) 1.2 <9(-4) <7(-4) <6(-4) <5(-4) <8(-4) 12sSb 0. 3(-5)
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT #3 Time 09:51 21:41 22:51 23:52 00:51 01:46 Date 5/16/78 5/19/78 5/19/78 5/19/78 5/20/78 Power 5/20/78 100% 70% 26% 0% 0% 0% Pressure (psi) 2240 2240 2240 2240 2260 2260 Temp. (*F) 570 565 8iS3 548 540 540 Flowt (gpm) 54 52 a2 50 50 50 Nuclide (uC1/ml) (uci/ml) (uci/ml) (uti/ml) (uci/ml) (pci/ml) 12sMe <8 - <5 -3) <6 -3) <4 -3) <4(-3 <3 - 12sTe <2 - <2 - <8 -2 <4 -2) <2 - <8 - 131Me <2 - <2 - <2 -2 <1 -2) <7 - <3 ~ 131Te <5 - <3 - <4(-3 <2 -3) <3 - <2 - 132Te 3 1(-5) ** ** ** <2 - <2 - 139Ba 1.9. 0.6(-2) 6.9 i 0.5(-2) 5.820.5(-3) 4.2 0.5(-3) 2.7t0.4(-3) 2.8i0.6(-3) 7 140Ba 8.410.6(-5) <3- <1 - <5 -3) <5(-3 <2 - y 140La <2(-4) <2 - <2 - <2-4) <2 -4 <8 - 141Ce <3(-6) J- <3- <4 -4) <4 -4 <2 - 143Ce <5 - <5(-4 <4 -4) <5 -4) <5 -4) <5 -4) 144 e <8 - <2( ) <2-2) <7(j) <7(-3)
<4 ,-3) 152Eu <4 - <2 - <6 - <6 - <6 - <6 -4) 154Eu <4 - <7 - <6 - <5 - <5 - <4 -
t issEu <4 - <4 - <9 - <4 - <4 - <9 - 187W <8 - <6 - <2-3) <7 - 1.2 1 0.3(-3) <1 - 239Np <2 - <3(-4 <7 -4) 4 2(-4) 3 i 1(-4) 5 1 2(-4) t t Letdown flow
- Radionuclide not detected
** Radionuclide not measured i
TABli B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS 10 REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F U41T #3 Time 02:48 03:40 04:40 05:40 06:40 07:40 Date 5/20/78 5/20/78 5/20/78 5/20/78 5/20/78 5/20/78 Power 0% 0% 0% 0% 0% 0% Pressure (psi) 2260 2250 2250 2250 2250 2250 Temp. (*F) 540 546 546 546 546 546 F1out (gpm) 50 51 51 50 50 50 Nuclide (pCi/ml) (pCi/ml) (pCi/ml) (uci/ml) (pC1/ml) (pC1/ml) 5.2 ** 2.9 ** esmKr ** 0.2(-3) ** 0.2(-3) ,, 85Kp ** ,, ,, ,, 1.2 ** ** 1.0 ** 87Kr ** 0.2(-3) 0.2(-3) ** 88Kr ** 7.5 0.3(-3) ** ** 3.5 0.2(-3) 89Kr ** * ** ** * ** 131mXe ** <2(-2) <5(-3) T133mXe ** 4.8 i 0.7(-3) ** ** 5.9 i 0.4(-3) **
** 2.37 **
M133Xe ** 4.6410.05(-1) ** 0.09(-1) 135mXe ** <5(-2) <5(-2)
** ** 8.7 0.1(-2) **
135Xe ** 1.00 i 0.04(-1) ** ** 137Xe ** <2(-3) ** <4(-2) 138Xe ** <7(-4) <9(-4) 84Br 7 2 3(-4) 6 1 3(-4) <5(-4) 4 2(-4) <6(-4) <3(-4) 1311 6.310.2(-2) 6.1 0.2(-2) 5.9 0.1(-2) 5.6 0.2(- ) 5.5 0.1(-2) 5.1 i 0.1(-2) 132I 1.29 i 0.04 - 1.19 0.02(1) 1.13 0.02 - 1.09 0.02 - 1.0510.01(-1) 1.00 0.01(-1) 133I 1.03 0.01 - 9.7 1 0.1 - 8.88 0.04 - 8.17 0.04 - 7.58 0.04(-2) 6.86 0.02(-2) 134I 1.27 0.03 - 6.5 0.1 - 3.99 i 0.09 - 2.48 0.08 - 1.4 i 0.1(-3) 7.1 1 0.6(-4) lasI 7.7 i 0.1(-2) 6.3 0.1 - 5.43 0.08 - 4.65 i 0.04 -2) 3.91 c 0.06(-2) 3.3710.04(-2) 88Rb 1.6710.05(-2) 1.4 z 0.1(-2) 9.6 0.3(-3) 7.8 i 0.3(-3) 6.4 0.7(-3) 3.9 1 0.2(-3) 89Rb <7(-4) <8(-4) <6(-4) <6(-4) <1(-3) <4(-4) 134Cs 2.2 1 0.2(-3) 2.98i0.09(-3) 2.14i0.07(-3) 2.2 0.1 - 3.50 0.06(-3) 2.17 0.05(-3) 136Cs 1.0010.09(-3) 1.18 0.09(-3) 1.15 0.08(-3) 1.1 0.1 - 1.6 0.l(-3) 1.03 1 0.08(-3) 137Cs 2.2 i 0.2(-3) 3.6 0.1(-3) 2.420.1(-3) 2.3 0.1 - 4.4 0.2(-3) 2.4 0.l(-3) 13eCs 3.2 0.3(-3) 2.0 1 0.4(-3) 1.9 i 0.4(-3) 1.0 1 0.3 - <3(-4) <3(-4) 139Cs <4(-3) <9(- 3) <3(-3) <2(-3) <8(-3) <2(-3)
/
TABLE B.ll (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT #3 Time 02:48 03:40 04 40 05:40 06:40 Date 07:40 5/20/78 5/20/78 5/20/78 5/20/78 5/20/7't 5/20/78
' Power 0% 0% 0% 0% 0% 0%
Pressure (psi) 2260 2250 2250 2250 2250 2250 Temp. (*F) 540 546 546 546 546 546 F1owt (gpm) 50 51 51 50 50 50 Nuclide (uC1/ml) (uC1/ml) (uCi/ml) (uci/ml) (pCi/ml) (uC1/ml) 2 a 5.3 1 0.2(-3) 610.1(-3) 4.3 1 0.1(-3)
, 3.9 i 0.1(-3) 6 2 0.1(-3) 3.3,3 , 1 0.08(-3)
SICr <2 - <8 - <1 -3) <7 -4) <9 -
<2 -3 <7(-4) 56Mn <2 - <2 -3) <2 -3) <1-3) <1(-3)
SSMn <8 -4 <5 - <4 -4) <2 -4) <2 -4) <2(-4) 59Fe <7 - <5(-4) <4(-4) <4 -4 <5(-4) <3 - ai, 57C0 <8 - <3(-4) <6(-5) <9 - <6(-5) <5 - L, ssCo <4 - 4.0 <4 -
<8 -
0.2(-3) 4.410.1(-3) 3.8 0.1(-3) <3 - 60Co 3.2 0.4(-4) 2.8 0.3(-4) <3 - 4.4 1 0.4(-4) 1.8 i 0.3(-4) 652n <2 - <2(-4) <2 - <2 - <2(-4) <2 - 91Sr <4 - <3(-4 <3 - <3- 9 i 2(-4) <2 - 91mY <2 - <2 - <1 - <1 - 1.5 0.7(-4) <1 - 93Y <3 - <3 - <2 - <3- <3 - <3(-3) 95Zr <2 - <2 - <2 - <2 - 2.7 0.6(-4) 95Nb <3 - <2 - - }
<2 <2 - <2 - <2- <9(-5) 4.lg 0.4(-4) 2.7 0.3(-4) 3.7 i 0.3(-4) 2.6 0.4(-4) 4.0 0.3(-4) 6.3 1 0.2(-4) 103
<2(-4) <2(,,4) <1(,,4) <2(,,4) <2(,,4) <9(,,5) 106Ru * * * * *
- 10sRh ** ** ** ** ** **
110l%g * * * * *
- 124Sb <2 -4 <2 - <1 -4 <9(-5) 1.5i0.6(-4) <8(-5) 12sSb <3 -4 <4 - <3 -4 <3(-4) <4(-4) <3(-4)
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT #3-Time 02:48 03:40 04:40 05:40 06:40 07:40 Date 5/20/78 5/20/78 5/20/78 5/20/78 5/20/78 5/20/78 Power 0% 0% 0% 0% 0% 0% Pressure (psi) 2260 2250 2250 2250 2250 2250 Temp. (*F) 540 546 546 546 546 546 Flowt (gpm) 50 51 51 50 50 50 Nuclide (uCi/ml) (uC1/ml) (uCi/ml) (uci/ml) (uCi/ml) (uci/ml)
<3 - ) <4 <4 - ) <3 - <3 - <2 -
129mTe
<3
') <2 - ) <3 - <2 - <2 -
12sTe 'l
<2 - <1 - <7 - <4 -
- 131mTe 131Te
<5-l)
<2
<2
- l
,) h l <1
<8 - '!
l l <1 - <1 - <1 - 132Te <9 - 1 <2-ll
<2 - 1 I <7 - 1.911.3(-4) 1.0 0.3(-5) 189Ba <2 ,,l 1.3 1.0(-3) 9 3(f-5) -4) <5 - <6f- <6 -
140Ba <2 <2 - <2 -3) <2 - <2/ - <2 -
'I <6 - ~Sf - <6 -
?' 140La <7 'l
<7 - <7 -5)
$ 141Ce <3 l <3 - <3-4) <3-4) <3/-4) <3-4) 1%sCe <6 - h <4 - <3 -4) <3 -4) <3(-4) <2 -4) 4 e <3 -3) <1 (-3) <7(-4) <5(-4) <4(-4) <4(-4) 152Eu <4 - <6-4) <9-4) <9(-4) <2(-3 <8(-4) 154Eu <4 - <3 -4) <4 -4) <3(-4) <4(-4 <3(-4) 1ssEu <3 - <3 -4) <2 -4) <4(-4) <3(-4 <3(-4) 16P' 2.910.5(-3) 1.7 0.3(-3) <2(-3) <7(-4 1.7 0.3(-3) 2.410.5(-3) 2 i 1(-4) 2ssNp <3(-4) <2(-4) 1.9 0.9(-4) 211(-4) <3(-4) t Letdown flow
- Radionuclide not detected
** Radionuclide not measured
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTOOWN, STARTUP OF UNIT #3 Time 08:40 09:40 10:45 11:40 12:47 13:45 Date 5/20/78 5/20/78 5/20/78 Power 5/20/78 5/20/78 5/20/78 0% 0% 0% 0% 0% 0% Pressure (psi) 2246 2252 2235 2235 2238 Temp. (*F) 2235 547 547 547 547 547 542 Flow t(gpm) 52 52 53 53 53 52 Nuclide (uct/ml) (uci/ml) (uci/ml) (uci/ml) (uci/ml) (pci/ml) asmKr ** 1.9 ** 0.1(-3) ** 1.1**1 0.1(-3) ** 85Kr ** ** ** 87Kr ** ** **
<2(-4) <3(-4) **
seKr 1.9**1 0.1(-3) ** ** 1.2 ** 89Kr ** ** ** ** 0.2(-3) 131mXe ** ** **
<6(-3) <3(-3) **
y 133mXe 6.1 2 0.3(-3) ** 6.6 0.7(-3) ** 133Xe ** 2.7* 2 0.1(-1) 2.5 0.2(-1) ** 3 135mXe ** ** ** * ** 13sXe ** ** ** ** 8.32 1 0.07(-2) ** ** 6.69 i 0.07(-2) 137Xe * ** 138Xe ** * ** ** * ** 84Br 5 1 2(-4) <5(-4) <3(-4) <3(-4) <5(-4) <2(-4) 131I 4.8 0.1(-2) 4.8 0.1(-2) 4.27 i 0.09(-2) 3.88 0.06(-2) 3.94 1 0.04(-2) 3.52 i 0.07(-2) 132I 9.6 i 0.1(-2 9.6 i 0.1(-2) 1.0110.02(-1) 9.1 0.1(-2) 9.2 0.2(-2) 9.1 0.2(-2) 133I 6.31 0.02( 2) 6.03 1 0.07(-2) 5.19 0.05(-2) 4.57 0.07(-2) 4.41 0.06(-2) 3.81 0.07(-2) 134I 3.7 i 0.6(-4 1.6 0.6(-4) <2(-4) <1(-4) <2(-4) <9(-5) lasI 2.80 0.06(-2) 2.43 0.07(-2) 2.05 0.02(-2) 1.67 0.02(-2) 1.41 : 0.03(-2) 1.18 0.02(-2) serb 2.5* i 0.3(-3) 2.1* 1 0.2(-3) 1.8 0.3(-3) 1.5 0.2(-3) 6 1(-4) 9 1(-4) 89Rb
- 13"Cs 1.94 0.07(-3 3.3110.06(3) 1.85 0.06(-3) 1.76 0.05(-3) 1.95 0.05(-3) 1.6010.04(-3) 136Cs 9.5 i 0.9(-4) ) 9.9 0.4(-4 9.8 0.3(-4) 8.1 0.4(-4) 7.9 0.6(-4) 7.8 0.4(-4) 137Cs 2.2510.08(-3) 4.0 0.1(-3 2.10 0.06(-3) 2.01 0.05(-3) 2.37 0.07(-3) 2.07i0.05(-3) 13sCs <2(-4) <3(-4) <3(-4) <2(-4) <3(-4) <2(-4) 139Cg
. . . _ . ~ = _. _ _ _ .
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT 13 Time 08:40 09:40 10:45 11:40 12:47 13:45 Date 5/20/78 5/20/78 5/20/78 5/20/78 5/20/78 5/20/78 Power 0% 0% 0% 0% 0% 0% Pressure (psi) 2246 2252 2235 2235 2238 2235 Temp. (*F) 547 547 547 547 547 542 Flow t (gpm) 52 52 53 53 53 52 Nuclide (uC1/ml) (pC1/ml) (uci/ml) (pci/ml) (pCi/ml) (uci/ml) 24Na 3.1**i 0.1(-3) 2.76 1 0.08(-3) 2.49 i 0.06(-3) 2.12 0.07(-3) 2.07 0.09(-3) 1.87 0.06(-3) 41Ar <5(-4) <5(-4 51Cr * * * * <8(-4 <7 - 54Mn <8 - <7(-4 <5-4) <4 - <4(-4 <4 - 56Mn <2 - <2(-4 <2 -4) <2 - <2(-4 <1 - 59Fe <2 - <3(-4 <2 - <2 - <2(-4 <2 - 57Co- <6 - <5(-5 <4 - <5 - <7(-5 <5 - 6.4 0.3(-3) 3.7 0.1(-3) <3 - <3(-3 <3(-3) <3-3) 7 seCo 1.1 1 0.2(-4) 3.4 i 0.4(-4) 2.0 0.3(-4) 5 i 2(-5) g; 60Co 4.5 1 0.4(-4) 3.2 i 0.4(-4) 65Zn <2(-4) <2(-4) <1(-4) s2(-4) <2(-4) <2(-4) 3.7 6 i 1(-4) 1.5 0.1(-3) 7.1 i 0.9(-4) 9 1(-4) 4.7 0.1(-3) 91Sr 0.9(-4) 5.3 0.6(-4) 91mY 1.3 0.5(-4) <2(-4) 1.9 i 0.5(-4) 1.7 0.5(-4) <2(-4) 93Y <3(-3) <3 -3) <3(-3) <3(-3) <4(-3) <3(-3) 95Zr 2.4 0.6(-4) <2 -4) <2(-4) 2.2 0.5(-4) 2.1 0.6(-4) <2(-4) 95Nb <2(-4) <8 -5) <2(-4) <9(-5) <2(-4) <7(-5) 4.1 0.3(-4) 3.1 0.2(-4) 4.5 0.5(-4) 4.2 0.3(-4) 2.3 0.2(-4) 99Mo 7.1 0.2(-4) ** ** ** ** ** 99mTc **
<2(-4) 9 4(-5) <1(4) <1(-4) <8(-5) 103Ru 4 2(-4) ** ** ** ** **
103mRh
~106Ru ** ** **
106Rh <7 - 110 nag * <1(-4) <9 - <7(-5) <1(-4)
<7 - 8 i 3(-5) <8(-5) <6 -
124Sb 1.7 0.4(-4) 1.4 0.5(-4) <3 - 12sSb <3(-4) <5(-5) <3- <3(-4) <3(-4)
TABLE B.ll (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP OF UNIT #3 Time 08:40 09:40 10:45 11:40 12:47 Date 13:45 5/20/78 5/20/78 5/20/78 5/20/78 5/20/78 Power 0% 0% 5/20/78 0% 0% 0% 0% Pressure (psi) 2246 2252 2235 2235 2238 Temp. (*F) 2235 547 547 547 547 547 542 Flowt (gpm) 52 52 53 53 53 52 Nuclide (uCi/ml) (pCi/ml) (uC1/ml) (uCi/ml) (pC1/ml) (uCi/ml) l'We- :!{:11 :!{:11 :!{:1 :!{:l e <7(-4 <6(-4) <3(-4) <4(-4) :3{:il
<8(-4 :!{::1
<8 -4) 132Te <1 - 3.0 0.5(-4) 2 1(-5) 2 2(-5) 8 -5) 139Ba <6 - <6-6)
<5 -4) <4 -4) <4 -4) <4 - <5 -4) 140Ba <2 - <2 - <2 - <2 - <2 - <2 -
ca 140La <6 - <7 - <4 - <4 - <5 - <4 - 1 w 141Ce <3 - <3 - <2 - <3- <3- <3-14 3Ce <3- <2 - <2 - <2 - <2 - <2 - 4Ce <4 - <5 - <4-4) <5 - <7 - <4 - 152Eu <8 - <8 - <7 - <8 - <7 - <7 - 154Eu <3 - <4 - <3 - <3 - <3 - <3-1ssEu <8 - <3 - <2 - <3- <4 - <2 - 187W <6 - <g - <7(-4 3.6 0.5(-3) 1.3
<6(-4 0.4(-3) 239Np 5 i 2(-4) <3 - <2(-4 <3(-4) <3(-4 <2(-4) t Letdown flow
- Radionuclide not detected
** Radionuclide not measured
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP OF UNIT 73 Time 14:40 20:05 23:45 04:50 05:50 Date 5/20/78 5/21/78 5/21/78 5/22/78 5/22/78 Power 0% 0% Just Critical 36% 99% Pressure (psi) 2235 2255 2240 2256 2260 Temp. (*F) 542 545 545 547 570 Flow t (gpm) 52 55 52 61 90 Nuclide (uci/ml) (uC1/ml) (uci/ml) (pCi/ml) (pCi/ml)
** ** 5 2(-3) 8smKr ** <9(-5) **
** ** ** c*
85Kr 87Kr ** ** <2(-4) ** 2.1 0.3(-2)
** 1.65 0.04(-2) 88gr ** ** 2.9 0.7(-4)
** ** ** 2 89Kp <2(-2) <Tg 131mXe
** ** <2(-2)
** <6(- 2 13mXe ** ** 3.4 0.1(-3) **
5 3-3)
** ** 1.8 0.1(-1) 3.8 0.2(-1)
? 133Xe ** <2(-2) g 13smXe ** ** <2(-2) **
13sXe
** ** 6.64 0.04(-3) 1.65 0.02(-2)
** <7(-3) 137Xe
** ** <5(-3) 138Xe
** ** 7 3(-4) ** 5.86 i 0.04(-2) 84Br * *
- 4.0 0.4(-3) 8.5 0.6 -
2.04 0.02(-2) 1.87 0.02 -2) 6.2 0.2(-2) 6.1 0.2 - 131I 3.29 i 0.07(-2) 6.86 i 0.10 - 6.5 0.1 - 132I 8.8 1 0.2(-2) 7.7 1 0.1(-2) 6.62 0.08-2) 7.58i0.08(-3) 6.14 0.10 -3) 1.82 1 0.02 - 2.04 0.03 - 133I 3.44 1 0.06(-2) 2.85 1 0.03 - 3.87 0.07 - 134y <1(_4) <6(-5) <6(-5) 3.0 1 0.4(-4) 3.5 0.5(-4) 9.9 0.4(-3) 2.13 i 0.04 - 135I 1.02 0.02(-2) 88Rb 5 1* 2(-4) <2(-4) <2(-4) 7.7 0.6(-3) 1.89i0.09(-2)
*
- 2.28 0.06(-2) 2.54 1 0.05(-2) 89Rb 5.6 i 0.1 -
134Cs 1.57 1 0.07 3) 1.66 0.03(-3) 1.76 0.05(- 3) 5.8i0.2(-3) 6.0 1.9 1.9 0.1 - 136Cs 6.7 1 0.3(- 6.5 1 0.4(-4) 5.3 1 0.2 - 137Cs 1.9 i 0.1(- 2.05 A 0.06(-3) 2.33 0.5(-4) 0.05(-3 ) 5.70 0.1(-3)) 0.0R(-3
- 9 1 3(-4) 6.0 0.3 - 1.1410.03(-1) 138Cs <2(-4) *
- 4.3 1 0.2 - 3.7 1 0.3(-2) 139Cs
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP OF UNIT 73 Time 14:40 20:05 23:45 04:50 05:50 Date 5/20/78 5/21/78 5/21/78 5/22/78 5/22/78 Power 0% 0% Just Critical 36% 99% Pressure (psi) 2235 2255 2240 2256 2260 Temp. (*F) 542 545 545 547 570 Flow t (gpm) 52 55 52 61 90 Nucl ide (pC1/ml ) (pC1/ml) (pCi/ml) (pCi/ml) (pCi/ml) 24Na 1.67 0.05(-3) 3.5 0.2(-4) 2.9 0.3(-4) 1.10 0.04(-3) 2.05 0.08(-3) 41Ar ** ** **
<4(-4) <4(-4) 51Cr <7 - * * *
- 5*Mn <3 - <3(-4) 6 2(-5) *
- 56Mn <l - <7(-5) <8 - <2(-3 <3-39Fe <2 - 1.2 0.3(-4) <7 - <3(-4 <4 -
5700 <6 - <3(-5) <3- <5(-5 <1 -
? seCo <3- 1.5 0.1(-2) 4 1(-3) 1.03 0.09(-2) 1.1 0.l(-2)
$ 60Co 2.4 1 0.3(-4) 1.50 0.04(-3) 6.3 0.2(-4) 4.2 0.4(-4) 9.7 1 0.6(-4) 65Zn <2(-4) <9(-5) <8(-5) <4(-4 <6(-4) 91Sr 1.03 0.09(-3) 1.7 0.4(-4) 2.2 0.6(-4) <4(-4 5 2(-4) 91mY 1.2 1 0.4(-4) <7(-5) <7(-5) <2(-4 1.7 0.8(-4) 93Y <3-3) <4(-3) <4(-3) <3(-3) <3(-3) 95Zr <1 -4) 1.37 0.07 -3) 3.5 0.4 -4) <2(-4) 9 1(-4) 95Nb <8-5) 1.2710.03-3) 3.9 0.7-4) <6(-4)
- 99Mo 4.2**1 0.2(-4) 1.06 i 0.01 -3) 4.3 0.2 -4) <6(-4) <1(-3) 99mTc ** ** ** **
103Ru <8(-5) 4.5 0.4(-4) 1.5 0.3(-4) <1(-4) 2.5 0.7(-4) ! Io3mRh ** ** ** ** ** l 106Ru * * * *
- 1 106Rh ** ** ** ** **
3.7 *
- 11orAg <7(-5) <5(-5) 0.9(-4) 124Sb <6(-5) 4.2 1 0.3(-4) 1.5 0.3(-4) <2(-4) <4(-4) 12sSb <3(-4) <2(-4) <2(-4) <5(-4) <7(-4)
- TABLE B.11 (cont'd)
RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT #3 Time t?:40 20:05 23:45 04:50 05:50 Date 5/20/78 5/21/78 5/21/78 5/22/78 5/22/78 Power 0% 0% Just Critical 36% 99% Pressure (psi) 2235 2255 2240 2256 2260 femp. (*F) 542 545 545 547 570 i Flow t(gpm) 52 55 52 51 90
- l. Nuclide (uCi/ml) (pC1/ml) (pC1/ml) (uci/ml) (uC1/mi) 129mTe <3 - <2(-3) <2- <3 - <4 -
129Te <2 - <l(-3) <1 - <4 - <6 - 131mTe <3 - 4.7 0.1(-4) <5 - <8 - <2 - 131Te <6(-4 <1(-3) <5 -4) <1(-3 <1 - 132Te <7(-6 139Ba < 3(-4 <2(-3) <2(-3) 6.4 0.4(-3) 1.12 0.06(-2) 140Ba <2 - 5.9 0.9(-3) 5.4 0.7(-3) 6.0 0.4(-3) 9.4 0.6(-3)
? 140La <4 - 1.4 1 0.1(-4) 1.4 0.2(-4) <7(-5 4 1(-4)
$ 141Ce <3 - <2 -
~
<2 -4) <3(-4 <2 -
143Ce <2 - <8 - <9 -5) <3(-4 <3 - 144Ce <4 - <3 <2 -4) <3(-3 <6 - 144Pr 152Eu <8 - <8 - <7 - <2(-3) <2(-31 , 154Eu <2 - <2 - <2- <3(-4) <4(-4 l IssEu <2 - <2 - <3 - <2- <7(-4l I : 187W 2.510.3(-3) <7 -4) 7 1(-4) <5 - <6(-4ll 2 39Np <2(-4) I .l i 0.1(-3) 2.3 0.6(-4) <2 - 4 2(-4) I t Letdown flow
- Radionuclide not detected
** Radionuclide not measured
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTD0t'N, STARTUP OF UNIT 03 Time 06:50 07:55 08:44 Date 5/22/78 11:04 12:00 Power 5/22/78 5/22/78 13:25 100% 100% 5/22/78 5/22/78 5/22/78 Pressure (psi) 100% 45% 2235 2250 45% 45% Temp. (*F) 574 2248 2250 571 2240 2240 Flow t(gpq) 570 570 90 90 558 559 89 90 92 90 Nuclide (uCi/ml) (pCi/ml) (pCi/ml) (uci/mi) (pCi/ml) BsmKr ** (uCi/ml) 4.7 0.3(-3) ** ** 85Kr ** ** ** 7 2(-3) ** B7Kr ** ** 88Kr ** 1.2i0.4(-2) ** ** 1.710.9(-2) ** 1.43 0.03(-2) ** ** 89Kr ** 2.10 0.04(-2) ** 131mXe ** <2( 2) ** **
<6(-2) ** ** <5 -2) **
133mXe ** cm ** <5(-3) ** ** <5-2) ** 133Xe 2.9 ** <2 -2) ** 1 13smXe ** 0.1(-1) ** 2.7 0.1(-1) ** 135Xe **
<2(-2) ** **
1.80 0.02(-2) ** ** <4(-2) 137Xe ** 3.18 0.03(-2) **
<9(-3) ** **
138Xe ** ** 3.3 2 0.8(-3) ** ** <9(-2)
'5.82 0.07(-2) **
84Br 3.320.4(-3) 2.7 1311 5.2 0.1(- 0.6(-3) 4.2 1 0.7(-3 6 1(-3) 6.7
) 4.66 i 0.08 - ) 4.3 0.2(-2 3.0 0.8(-3) 711(-3) 1321 4,79 1 0,09 4,44 0.07 - f 4.97 0.2(-2) 3.11 0.07(-2 2.61 0.09(-2 133I 1.84 1 0.01 - 2.07 0.02 0.05(2) 6.09 0.07 - 6.4 i 0.1(-2) ) 6.6 0.1(-2) )
134I 2.42 0.04(-2 2.86 0.06 - 2.89 0.04 - 3.69 0.05 - 4.88 0.04 - /l 2.71 1 0.06(-2 135I 2.22 0.04 - 7.2 0.2(-2) ) 1.19 0.04 - 1.05 1 0.01 - 9.5 t 0.2(-2) ) 3.2010.07-) 4.38 1 0.05(-2) 5.42 1 0.08 - 5.16 i 0.05 - 88Rb 4.87 i 0.04(-2) 2.01 1 0.08(-2) 2.4510.09(-2) 2.6 0.1(-2 2.55 89Rb 3.010.2(-3) 2.6 0.07(-2) 3.4 0.1(-2) 3.1 1 0.1(-2) 0.2(-3) 4.4 1 0.2(-3 2.03 0.02(-2) 134Cs 4.83 1 0.07(-3) 5.1510.08(-3) 4.19 0.07( 3) 2.8 0.2(-3) 2.94i0.04(-2) 3.85 0.04(-2 136Cs 1.5 i 0.1(-3) 1.7 1 0.1 - 2.99 0.09(-3) 2.2 0.2(-3)) 1.3 0.2(-3 5 2(-4) 137Cs 4.57 0.08(-3) 5.0 1 0.1 - 3.90 9 i 2(-4) 8.3 t 0.9 - 138Cs 0.07(-3) 2.4 0.2(-3) 2.5 1 0.2(-3 1.7 1 0.1 - 5.810.1(-2) 5.7 0.1 - 7.6 0.2(-2) 9.3 1 0.2(-2) 8.6 0.3(-2 139CS <4(.3) <5(-3) <4(-3) 8.2 1 0.2 - 3.6 2 0.2(-2) 6.6 0.9(-2 1.06 2 0.03(-1)
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR C00LANi 5/19-22/78 SHUTDOWN, STARTUP OF UNIT #3 Tirse 06:50 07:55 08:44 11:04 12:00 13:25 Date 5/22/78 5/22/78 5/22/78 5/22/78 5/22/78 5/22/78 100% 100% 100% 45% 45% 45% Power Pressure (psi) 2235 2250 2248 2250 2240 2240 Temp. (*F) 574 571 570 570 558 559 Flow t(gpm) 90 90 89 90 92 90 (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) (pC1/ml) (pC1/ml) Nuclide 24Na 2.16 1 0.06(-3) 3.0t0.2(-3) 3.99 0.09( 3) 5.00 0.09(-3) 4.6 0.1(-3) 4.4 0.1(-3) 41Ar ** <3(-4) ** <4(-4)
- SICr * *
- 54Mn 56Mn <2 - <3- <4(-3) <6(-3) <5(-3) <4 -
59Fe <2 - <3 - <5(-4) <5(-4) <6(-4) <4 -
<5 - <6 - <6(-5) <1(-4) <8(-5) <7 -
57Co 6.9 0.4(-3) 7.8 0.8(-3) 2.810.2(-3) 9.610.9(-3) <2 - i 58C0 7.210.6(-3) 7 1(-4) <2 -4) m 60C0 8.3 1 0.6(-4) 8.9 0.5(-4) 1.01 0.05(-3) <2(-4)
<2(-4) <2(-4) 5 3(-4) <5 - <6 -4) ssZn <2(-4) <4 - <5 - 4 1 2(-4) 91Sr 4 1 2(-4) 8 1 2(-4) 9 2(-4) 2.610.7(-4) 2.4 0.7(-4) <2 - <3 - 2.6 0.8(-4) 91mY <2(-4) <4 - <2.-3) <2(-3) 93Y <2(-3) <2(-3) <2(-3) 1.8* i 0.8(-4) 6.7 0.9(-4) <4 -4) 9 2(-4) <3(-4) 95Zr 7 1* 1(-4) * * *
- 9sNb
<8(-4) <8( ,4) <6(-4)
<5(-4) <5(, ,4) <6(,,4) , ,
<2(-4) 1.7,+0.5(-4) <1(-4) <2(,,4) <2( ,4) <2(,,4) 106py ** ** **
106Rh * * *
- 110 mag
<2 -4 <3 -4 <3 -4 <3 -4 <4 -4 <2(-4) 124Sb <4 -4 <4(-4) 12sSb <3 -4 <6 -4 <4 -4 <4 -4
TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS 1N REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP OF UNIT #3 Time 06:50 07:55 08:44 11:04 12:00 13:25 Date 5/22/78 5/22/78 Power 5/22/78 5/22/78 5/22/78 5/22/78 100% 100% 100% 45% 45% Pressure (psi) 2235 2250 2248 45% 2250 2240 2240 Temp. (*F) 574 571 570 570 558 559 Flow t (gpm) 90 90 89 90 92 90 Nuclide (uCi/ml) (uci/ml) (uCi/ml) (uCi/ml) (uCf/ml) (uci/ml) 129mTe <3(-3) <3 - <5 - <4 - <4 - 129Te <3 -2 <3 - <5- <7- <5-3)
<9 -3 <6 - <6 -2) 131mTe m <2 - <2 - <4 - <2 - <2 -
<7 -4 <1 - <1 - <2 - <3 -
, <4-139Ba 1.32 0.03(-2 1.20 1 0.02(-2) 8.9 1.87i0.03(-2) 0.5(-3) 4 1(-3) 1.4 1 0.4(-3)
'?
140Ba 140La 7.5 z 0.5(-3) ) 1.61 0.03(-2 1.14 1 0.05(-2) <7- 1.8 0.7(-3) 2.3 0.4(-3} 8
<2 -
4(-5) 6.5 0.5(-4)) <9 - <2 - 5.3 0.7(-4) <2 -
'$ 141Ce <2 - <2 - <3 - <2 - <3-143Ce <2 - <4 - <3 - <4 - <5 -
<4 - <4 -
144Ce <6 - <8 - <2 - <1 - <9 - 144pp ** ** ** ** ** ** 152Eu <2(-3) <2(-3) <1 - is4Eu <9(-4) <8- <7(-4
<4(-4) <3(-4) <3 - <5 -4) <5 - <3(-4 155Eu <4(-4) <3(-4) <5 -
187W
<4 -4) <6 - <5 -4 5 2(-4) 4 2(-4) <8(-4) <8 -4) <8(-4) <8 -4) 239Np <3(-4) 1.9 0.9(-4) <4(-4) <3 -4) <5(-4) <4-4) i t Letdown flow Radionuclide not detected Radionuclide not measured
1 i TABLE B.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT ir3 Time 14:20 Date 5/22/78 Power 45% Pressure (psi) 2240 Teep. (*F) 558 Flow t (gpm) 92 Nuclide (pCi/ml) 85mKr ** 85Kr .** 67Kp ** 88Kp ** 89gp ** 131mXe 133mXe 4 o 133Xe 135mXe 135ye ** 137Xe 138Xe 84Br 5 i 1(-3) 131I 2.50 0.08( 2) 132I 6.1 0.2(-2 133I 2.61 1 0.05( 2) 134I 7.9 i 0.1(-2 135I 4.63 0.07(2) . 88Rb 2.64 1 0.08(-2) 89Rb 2.3 0.1(-2) 134Cs 2.0 0.1(-3) 136Cs 6.4 1.0(-4) 137Cs 1.6 i 0.1(-3) 138Cs 7.4 1 0.2 - 13sCs 3.8 i 0.4 -
- ~
4 4 4', e p , s: r . l,/ - R,).. ,s;Q47% t s,;ip,.% k IMAGE EVALUATION NNNN TEST TARGET (MT-3) I. i 1.0 lll m E
=
s= mk!=a m j;; g 3 l-l -
'bb
' "l=I .8 ll =
1.25 1.4 1.6
- 6" >
MICROCOPY RESOLUTION TEST CHART & n, K*% e'p
/!b
<>4;m sp
,v65
___T o F~ ffge.4c 3', l 4p l _ . _- _. .
I si .f As . 9p i t ,F$// ln p ' $ $ k" g )f> / g(Ryvyf' j[h' lg?
! \ IMAGE EVALUATION
'N"//l
/,
i TEST TARGET (MT-3) 4 t J i, l.0 ;;;"2" E a a jg *2 2
! ,2 :36 l
1.1 t m g 2.0
- L -
[14 l.25 l.4 ll,_1.6_ f f MICROCOPY RESOLUTION TEST CH ART
+ rp m& %w,/%'A, + <>+ .
a fAs \ V - /Yw -r-. 6p e t ,e o g. ,sy y f
%' V'
TABLE B.11 (ctnt'd) RADIONUCLICE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP OF UNIT #3 Time 14:20 Date 5/22/78 Power 45% Pressure (psi). 2240 Temp. (*F) 558 Flowt (gpm) 92
'Nuclide (pCi/ml) 24Na 41Ar-4.2**i 0.1(-3) s1Cr
- 56Mn
- ssMn <4 -
59Fe <5 - 57CO <l 58C0 *
? 60Co 2.6 i 0.5(-4) 50 ssZn <3f-91Sr <4 -
9Imy <2l . 93Y <3 95Zr <2(d- )I 9sNb
- 99Mo <6(-4)
' 99mTc **
103Ru 2.6 0.6(-4) 103mRh ** 10sRu
- 10sRh **
, 110 mag
- 4 124Sb <3 -4 12 sSb <3 -4
l TABLE R.11 (cont'd) RADIONUCLIDE CONCENTRATIONS IN REACTOR COOLANT 5/19-22/78 SHUTDOWN, STARTUP 0F UNIT #3 Time 14:20
.Date - 5/22/78 Power- 45%
Pressure (psi) 2240
. Temp. ("F)' .558-Flowt (gpm) 92 Nuclide (uCi/ml) 129mie <4(-3) 129Te <5 -
131mie- <3 - 131Te <2 - 132Te *
- as 139Ba 7.5 0.4(-3) '
- 4, 140Ba <4I'-
N , 4 14oLa <11l-141Ce 143Ce <3(1
<3- -
144Ce <8(- 144Pr ** 152Eu <5 - 154Eu <4 - IssEu <3 - 187W <5 - 239Np <3 - l t Letdown flow
- Radionuclide not detected
** Radionuclide not measured l
4 4
TABLE B.12 UNIT #3 CVCS PURIFICATION DEMI %ERALIZER Demineralizer B 13:52; 2/21/78 Used for: 25 days Bed Volumes Thru: 7.5(3) Letdown Flow Rate: 45 gpm Reactor Coolant Boron: 867 ppm Inlet .;ctivity Outlet Activity Decontamination Nuclide (uCi/ml) __ (uCf/m ) Factor
. 131I 3.37 1 0.03 - 1.1 0.1 - 3.1 1 0.3(3)
- 132I 3.45 2 0.08 - 1.5 0.1 -
- 2.310.2(4) i 133I 1.01 1 0.02 - 1.6 0.1 - 6.3 1 0.4 2) 134I 1.27 t 0.03 - 2.8 1 0.2 -
- 4.5 0.3 3) 1351 7.9 t 0.1(-2) 2.5 0.4 -
- 3.2 1 0.5 3) l 88Rb 5.9 0.3(-2) 2.2 0.2(-2) 2.7
- 0.3(0 l
89Rb 2.7310.04(-2) 1.3 0.2(-3) 2.1 1 0.3(1 13'ts 1.92 0.03(-4) 1.43 0.03(-5) 1.34 i 0.04 1)
- 136Cs 9.4 1 3.2(-5) 1.0 0.3(-7)^ 9.4 4.3(2) 137Cs 2.38 0.04(-4) 1.88 0.04(-c. 1.27 i 0.03(1) lats 9.010.2(-2) 6.1 0.5(-3) 1.5 0.1(1 139Cs 6.9 0.7(-2) <1.1(-3) >6. 3(1 )
24Na 1.70 0.03(-3 <2.5(-8)* >6.8(4) 51Cr 1.80 0.07(-4 6.0 2.4 -
- 3.0 1.2(2 l 54Mn 1.14 1 0.01( 4 3.4 1 0.5 - 3.4 i 0.5 j 59Fe 2.0 1 0.1(-5 3.0 0.5 -
- 6.7 1 1.2 57Co 4.5 9.2(-6 9.9 3.5 - 4.5 1.6 seCo 1.45 0.02( 3) 6.0 0.1(-6) 2.4 i 0.1 f'
60Co 8.7 i 0.1(-4 2.0 1 0.1(-6) 4.4 1 0.2 65Zn 1.3 i 0.1(-5 <1.1(-7)* >1.2(2) i 84Br i 91Sr 5.25 1 0.07(-3) <1.5(-7)* >3.5(4) ; l 91mY ! ! 93Y 1.3 0.2 - <1.8(-7)* >7.2(3) i 95Zr 3.1 i ';.1 - 5.0 0.5 -
- 6.2 0.7 i 95Nb 1.4 2 0.5 - 6.7 0.7 - 2.1 1 0.8 l- 99Mo 4.2 i 0.1 - 1.5 0.1 -
- 2.0 0.2 103Ru 3.12 0.05 5) 3.8 1.2 - 8.2 i 2.6
- lo6Ru <7.5(-6) 5.7 1.3 - * <1.3(1) 1.7 i 0.4 3 110 mag 3.3 ' 0.7(-4) 1.9 i 0.4 -
124Sb 5.3 0.2 - 9.4 0.5-) ) l 12sSb 5.0 6.92 i O.2(-5) 0.09(-5 ) 7.38 1 0.07(-5) 9.4 1 0.2 - ) i 129Te <1.2(-2) <1.7(-4)*
- 131mTe <2.3(-4) <1.6(-7)*
l 131Te 132Te 2.610.4(-6) 5.8 1.0(1) l 139Ba 1.520.1(-)) 1.73 1 0.04 -2 7.2 t 0.1(-3) 2.4 0.1(0) i 140 Ba' 5.90 0.04 -4) <8.5(-8)* >6.9 l 14oLa 4.0 1.5(-5) <8.4(-8)* >4.8 l 141Ce - 3.220.7(-6) <8.0(-8)* >4.0
' 144C, <2.2(-6) 2.0 0.7(-7)* <1.1 187W 1.9 0.2(-4) <1.7(-7)* >1.1 239Np 1.0 1 0.3(-4). 2.6 0.8(-7)* 3.8*1.7(2) 78-53
v TABLE D.12 (cont'd) UNIT #3 CVCS PURIFICATION DEMINERALIZER Demineralizer B 12:21; 2/23/78 Used for: 27 days Bed Voluires Thru: 8.0(3) Letdown Flow Rate: -J3 gpm Reactor Coolant Boron: 848 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uCi/ml) (pCi/ml) Factor ._ 1311 4,43 1 0.03( 3) 1.6 0.4(-6) 2.8 1 0.7(3) . 1321 5.1 0.1(-2 <1.2(-4) >4.3(2) l- 133I 5.0 1 0.1(-2 2.0 0.2(-5) 2.5 i 0.3(3) 134I 1.29 1 0.03( 1) <1.6(-4) >8.1(2) l 13sl 8.7 1 0.1(-2) . 6 i 0.5(-5)* 5.4 1.7(3) ] serb 2.7 i 0.3(-2 2.6 0.3(0) 7.0 1 0.08(-2 0.4(-2) ) 1.3 0.2(-3 ~ 89Rb 2.41 1.9 0.3(1) 134Cs 4.99 0.08( 5) 1.93 i 0.04( 5) 2.59 i 0.07(0) [ 136Cs 3.5 1.0(-5 9.7 i 8.6(-8
- 4 137Cs 7.4 1 0.1(-5 2.60 0.05(-5) 3.6 1i 0.07(0 2.85 3.4(2) )
13eCs 1.01 0.02(1) 1.16 1 0.05(-2) 8.7 i 0.4(0) 1 139Cs 5.4 1.2(-2 <1.3(-3) >4.2(1) l 24Na 2.34i0.06(-3) <1.5(-7)* >1.6(4) 51Cr 1.06 0.09( 4) <6.7(-7) >1.6(2) 54Mn 6.6 1 0.2(-5 2.0 1 0.5(-7) 3.3 0.8(2) 59Fe 1.66 0.07( 5) <1.9(-7)* >8.7(1) 57Co 8.3 1.6(-7) 1.3 0.5 - 6.4 2.7 ) saco 3.6 i 0.1(-4) 6.6 1 0.1 - 5.5 0.2 0.4 g) 1 60Co 2.06 i 0.04(-4) 2.4 0.1 - 8.6 1 6 sZn 3.0 i 0.6(-6) <1.6(-7)* ~> 3.1(1) l 84Br 91Sr 3.9610.09(-3) <7.5(-7) >5.3(3) 91my , 9 3Y 1.4' O.4(-3) <8.4(-7)* >1.7(3)
, 9sZr 2.1310.04(-5) 5.0 0.9(-7) 4.3 0.8 95Nb 4.3 0.8(-5) 1.0 0.1 - 4.3 0.9 99Mo 6.0 0.1(-4) 3.0 0.5 -
- 2.0 0.3 103Ru 2.21 0.03(5) 5.1 0.9 - 4.3 i-0.8 .
lo6Ru 4.1 1 0.5( <1.4(-7)* >2.9(2) 110 mag 5.0 0.2(-5 <6.6(-6)* >7.6(0) 124Sb 1.3710.04(-5) 1.9 0.1(-1) 12sSb 1.64 0.05(-5) 7.3i0.2(-5)) 1.05 1 0.02(-4 1.56 1 0.06(-1) 129Te 1.1 1 0.2(-2)- <3.4(-3) >3.2(0) 131mTe <2.5(-4) <2.4(-7)* 131Te-132Te_ 1.3 0.1(-4) 1.7i0.5(-5) 7.6 2.3(0) 13984 1.77i0.02(-2) 7.2 0.2(-3) 2.5i0.1(0) 4 14cBa .7.01 1 0.07(-4) <1.7-7)* >4.1 140La 2.8 .1.7(-5) <1.8 - * >1.6 141Ce; 2.0 1 0.3(-6) <1.1 - >1.8 144Ce <2.1(-6) <1.0 -
- 187W 2.9i0.6(-4) <6.0 - * >4.8(2) 239Np 9.1 3.0(-5) 7.5i2.0(-7)* 1.2 0.5(2)-
B-54
- t 4 ,-r
TABLE B.12 (cont'd) UNIT #3 CVCS PURIFICATION DEMINERALIZER - Demineralizer B 18:21; 4/12/78 Used for: 75 days Bed Volumes Thru: 2.43 (4) Letdown Flow Rate: 55 gpm Reactor Coolant Boron: 712 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uci/ml) (uCi/ml) Factor 131I 2.5 0.4 - 2.710.4(3) 1321 6.7 1.04 1 0.03(-1 0.2(-3) ) 2.5 1.0 - 4.0 1 1.5(2) 133I 5.2 0.2(-2) 2.0 0.4 - 2.6 0.5(3) 1 34I 1.62 0.03(-1) <8.5(-5) >1.9(3) 1351 9.4 1 0.2(-2) <1.2(-4) >7.8(2) 88Rb 3.97 1 0.05(-1) 3.50 i 0.06(-1) 1.13 0.02(0) 89Rb 5.2 0.7 - <1.l(-3) >4.7(1) 134Cs 3.5 0.1 - 3.2 0.1 - 1.1 0.05(0) 136Cs 4.4 0.8 - 7.1 0.4 - 6.2 1.2(0) 137Cs 4.8 1 0.1 - 4.6 0.2 - 1.04 0.05(0) 13sCs 2.14 0.04(-1) 5.1 0.1 - 4.2 0.1(0) 139Cs <2.3(-1) <5.1(-2) 24Na 7.2 0.2 - 4.8 1.8(-6) 1.5 1 0.6(3) 51Cr 3.8 1 1.6 - <2.7(-6) >1.4(1) 54Mn 7.1 0.3 - 8.8 0.2(-8) 8.1 0.4(2) 59Fe 1.2 0.2 - <2.1(-7) >5.7(1) 57Co <l.5(-6) <3.1(-7) seCo 1.24 0.06(-4) 7.8 0.8(-7) 60Co 1.06 i 0.02(-4) 4.6 01(-7) 1.6 2.30 1 0.07(2 0.2(2) ) 65Zn <1.9(-6) <2.4 - e4Br 1.7 0.4 -2) <1.5 - >1.1 91Sr 2.710.2-3) <1.9 - >1.4 91my 2.2 0.2 - <6.0 - >3.7 93y 2.8 1 0.6 - <2.7 - >1.0(2 95Zr 4.9 i 0.7 - >1.6(1 95nb- 3.1 0.3 - <3.0(-7)(-7) 3.6 1.6 8.6 3.9(0) 99Mo 8.1 i 0.9(-4 <2.3(-6) >3.5(2) lo 3Ru 3.2 i 0.6(-6 <3.8(-7) >8.4(0) 106Ru <1.6(-5) <2.5(-6) 110 mag <9.5(-7) <1.7(-7) 124Sb 2.210.4(-6) 6.110.8(-7) 3.6 0.8(0) 12sSb <2.5 - 1.5i07(-6) >1.7(0) 129Te <5.8 - <1.4 - l 131mTe <1.4 - <5.4 - 131Te 5.2 2.1 - <6.6 - >7.9(0) 132Te 1.2 i 0.3 - <7.5 - >1.6(1) 139Ba 4.2 1 0.2 - 3.0 1(-3) 1.4 1 0.1(0) 140Ba 1.03 2 0.03 3) <1.7 - >6.1(2) j 140La 7.66 0.07 -5) <4.4 - >1.7(1) 141Ce <6.5(-6) <1.3 - 144Ce 1.710.4(-4) <2.4 - >7.1(1) 187w 9.6 i 1.1 <2.5 - >3.8(1) 239Np .<1.9(-5) (-4) <8.8 - B-55 9 -- TABLE B.12 (cont'd) ! UNIT #3CVCS PURIFICATION DEMINERALIZER j l -Demineralizer B 11:21; 4/13/78 Used for: 76 days Bed Volumes Thru: 2.45(4) Letdown Flow Rate: 53 gpm l Reactor Coolant Boron: 732 ppm Inlet Activity Outlet Activity Decontamination Nuclide (pCi/ml) (uC1/ml) Factor l 1311 7.5 1.2 1 0.2(-5) 6.2 1.1(2) 1321 1.04 0.3(-3) 0.05(-1 ) <1.3(-4). >8.0(2) 133I 5.4 0.1(-2) 6.5 1 0.4(-5) 8.3 0.5(2) 134I 1.69 2 0.03(-1) <1.2(-4) >1.4(3) 1351 9.0 0.4(-2) 6.6 2.6(-6) 1.4 1 0.5(3) serb 4.8 t 0.2 - 1.30 0.02(-1) 3.7 i 0.2 0) 89Rb 5.3 1 0.4 - 1.7 1 0.2 - 3.1 0.41) 134Cs 3.2 i 0.1 - 3.3 0.2 - 9.7 i 0.7 - ) 136Cs 4.0 0.3 - 8.2 1 0.7 - 4.9 t 0.6 0 137Cs 4.6 0.1 - 4.4 0.3 - 1.0 0.1(0 13eCs 2.3 1 0.1 - 4.3210.08-1) 5.3 0.3(0 i 139Cs <2.4(-1) <1.3(-2)
-24Na 7.5 i 0.2 - 7.5 1 2.6(-6) '
1.0 0.3(3) 51Cr 6.1 t 1.3 - <4.8(-6) >1.31) 54Mn 5.8 0.1 - <3.5(-7) >1.72) 59Fe 7.7 1 1.1 - <3.9(-7) >2.0 1) 57CO <l .4(-6) 58C0 1.16 1 0.02(-4) <1.0(-6)(-6) 2.9 0.2 4.0 6.3 0.3(1) 60Co 8.8 1 0.1(-5) 1.4 01(-6) 0.5(1) 65Zn <2.4(-6) <4.4 - 84Br 1.710.4-) 2.3 0.2 -
<1.4 -
<1.8 -
>1.2
>1. 3 91Sr I simy 4.5 0.3 - <6.8 - >6.6 93Y 1.1 0.4 - <2.1 - >5.2(0 95Zr 5.311.3- <5.6 - >9.5(0 9sNb <4.9 -
99Mo <4.3(-6)(-4) 1.4 1 0.3 <3.8 - >2.5(2) 10 3Ru 3.1 0.5 <6.0 - >5.2(0) ' 106Ru <7.3(-5) (-6) <4.7 - l 110 mag <1.2(-6) <4.6 - 124Sb 1.9 1 0.7(-6) 8.0 6(-7) 2.4 1.0(0) l l 12sSb <1.8 l 129mTe <1.3-))
<6.5 - <1. 5
<2.3 129Te <3.3 - )
131mTe <1.5 <6.5 - 131Te <4.9 - ')i <6.1 -
- 132Te '1.1 <1.0 - >1.1(1) t 139Ba 1.30 0.3(-5) 0.02(-2 ). 6.1-i 2(-3) 2.1 0.1(0) 140Ba. 1.010.1(-3) <3.2 - >3.1(2) 140La 3.7 i 0.1 <1.7 - >2.2(1) ',
141Ce <3.9(-6) (-5) <2.4 -
~144Ce 1.0 1 0.2(-4) <5.9 - >1.7(1) 187W 1.1 0.1(-3) <1.9 - >5.8(1) 239Np. <2.0(-5): <1.1 -
B-56
TABLE B 12 (ctnt'd) UNIT #3 CVCS PURIFICATION DEMINERALIZER Demineralizer B 11:17; 4/15/78 Used .for: 78 days Bed Volumes Thru: 2.53(4) Letdown Flow Rate: 53 gpm
. Reactor Coolant Boron: 703 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uCi/ml) (uCi/ml) Factor 131I 6.02 1 0.02(-3) 1.6 1 0.8(-6) 3.8 1 1.9(3) 132I 1,04 1 0.01 - <8.1 (-5) >1.3(3) 133I 5.26 2 0.03 - 9.3 1 2.5(-6) 5.7 i 1.5(3) 134I ~ 1.73 10.02 - <1. 2 -4 >1. 4 "'
lasI 9.7 1 0.1(-2) <1. 3 -4 >7.5 2 esRb 1.27 i 0.04(-1) 5.6 i 0.1(-2) 2.3 i 0.1 0) 89Rb 2.1 0.2(-3) 2.8 0.3 1) 134Cs 5.9 2.89 i 0.2(-2) 0.02f 4 ) 2.97 1 0.03 - 9.7 0.1 -1) 136Cs 5.5 0.1(-5) 1.03 0.04 - 5.3 0.20) 337Cs 4.1610.02(-4 3.92 0.07 - 1.06 0.02(0) 13eCs 2.35 1 0.02(-1 2.5 0.6(-2) 9.4 2.3(0) 139Cs 1.2010.09(-1 <4.4 (-3) >2.7(1) 24Na 7.1 0.1(-3) <3.7 (-6) >1.9(3) 51Cr <4. 0 (-6) <6.6 (-6) 54Mn 6.3 i 0.1(-5) 4.2 2 1.5(-7) 1.5 i 0.5(2) 59Fe 9.6 i 1.1(-6) <5.4 (-7) >1.8(1) 57Co <1. 7 (-6) <5.0 (-7) 58Co 1.12 2 0.02(-4) 2.3 1 0.3(-6) 4.9 0.6(1) 60Co 9.3 0.1 - ) 8.6 2.3(-7) 1.1 i 0.3(2) 6sZn <2.5(-6) <5.1 - e4Br 1.5 0.1 - <8.1 -
>1.9(1) 91Sr 2.2 1 0.1 - <3.9 -
>5.6(1) simy 9.9 i 1.3 - <6.6 -
>1. 5 93Y 3.8 0.6 - <3.8 -
>1.0 1.1 1 0.1 -
~
952r <7.4 - >1.5 95Nb 7.2 0.4 - 8.4 1 2.6(-7) 8.6 2.7(0) 99Mo. 9.19 0.04(-4 <3.7 -
>2.5(2) lo3Ru 2.4 2 0.5(-6) ) <1.1 -
>2.2(0) 106Ru <2.8(-6) <6.8 -
110ngg 4.4 2 2.0 - <3.7 -
>1.2(1) 124Sb 3.0 1 0.4 - 8.8 i 3.3(-7) 3.4 i 1.4(0) 12sSb 3.9 t 1.5 - <2.3 -
>1.7(0) 129Te <6.3 -2) <2.8 -
131mie <1.9 -4) <5.3 - 131Te <2.7 <1. 3 - 132Te 1.7 0.2 <4.1 >4.1(0)
-3)(-5) ) <4.3 13988 l.96 1 0.04(-2 6.1 0 (-3) 3.2 0.1(0) 140Ba- 9.4 1 0.1(-4) -
>2.2(2) j 140La 8.310.2(-5) <5.9 - >1.4(2)
- 141Ce . <2.5 (-6) <1.0 -
144Ce <3.7 - 187W <2.4 (-6)(-4) 7.8 1 0.5 <3.8' -
>2.1(1) 239Np <1.8 (-51 <l.6 -
B-57
r r TABLE B.12 (cont'd) UNIT #3 CVCS PURIFICATION DEMINERALIZER 1 Demineralizer B 13:54; 4/27/78 Used for: 90 days Bed Volumes Thru: 2.99(4) l Letdown Flow Rate: 55 gpm Reactor Coolant Boron: 673 ppm Nuclid' Inlet Activity Outlet Activity Decontamination l (uCi/ml) (uCi/ml) Factor 131I 1.25 0.01 - 3.6 0.7(- 6 3.5 0.7 1321 9.79 0.07 - 2.7 1 0.9 1331 1341 5.82 1.74 0.04 - 0.01 - 3.611.2(-5)h' 1.0
<3.7(-5) 0.3(-4 5.8 1.7
>4.7(3)
! 1351 9.7 0.1(-2) <9(-5) >1.1(3) serb 1.18 0.03(-1 7.5 1 0.1(-2) 1.57 0.05(0) 89Rb 6.0 0.2(-2) ) 8.7 0.8(-4) 6.9 0.7(1) i 134Cs 5.32 1 0.09(-4) 5.09 0.06( 4) 1.05 0.02(0) 13sCs 5.0 1.2(-5) 2.1 0.4(- 2.4 0.7(0) 137Cs 7.54 0.09 - 6.48 i 0.07 4) 1.16 0.02(0) 13eCs 2.03 0.02 - 2.00 0.03 -2) 1.02 0.02(1) 139Cs 1.26 i 0.08 - <4.2(-3) >3.0(1) 24Na 9.5 0.3(-3) <6.9(-6) >1.4(3) 51Cr <4.7(-6) 1.7 0.3(-5) <2.8 (-1 ) ! 54Mn 6.7 0.2(-5) 9.2 1 1.6(-7) 7.3 1.3(1) l- s9Fe 1.0 0.1(-5) <8.7(-7) >1.1(1) s7Co <1,1(.6) <4,4( 7) SECo 1.8 i 0.07(-4) 2.65 0.06(-5) 6.8 0.3(0) ,. 60Co 1.11 0.02(-4) 9.0 0 3(-6) 1.23 0.05(1) ssZn <1.3(-6) <1.5(- 84Br 1.3 0.1 - <4.7I,- >2.8 91Sr 6.0 1.1 - <6.6 -
>9.1 91mY 8.2 2. ' - <2. 3(I,- >3.6 93Y 2.1 t 0.! - <2.1(- >1.0 95Zr 1.710.1- 3.5 0.4 - 4.9i0.60).
9sNb 7.8 2 1.4 -6 2.6 0.3 - 3.0 1 0.6 0) 99Ho 8.820.9- 1.5 0.2 - 5.9 0.81) 103Ru 5.5 1.5 - 1.7 i 0.4(-6 3.2t1.20) lo6Ru 2.9 1 1.6 - <5.7(-8) >5.1(1) 110 mag <1.4(-5) <9.4(-7) 124Sb- .7.9 i 1.2(-6)' 1.9 0.3(-6) 4.2 i 0.9(0) j 12sSb <4.7(-6) j 129Te 3.5 ii0.3(-2) <7.6I;-7))
<2.7(-3 >1.3(1) l 131mie 1.1 i 0.7 <9.8' >1.1(2) i 131Te <4.0(-3) (-4) <2.0 - ')i 1MTe 2.7i0.3(- >1.2(1) 139C3 9.5 0.5(- <2.2-)(-3) 2.6 0.3 3.7 i 0.5(0) 140Ba 9.04 1 0.09 4) 9.2 1 0 3(-5) 9.8 0.3(0) 14cLa <3.1 - <2.3 -
141Ce <2.9 - <1.3 - 144Ce <3.0 - <2.4 - 187W <3.8 - <8.0 - 239Np ~< 1.1 - <2.9 - B-58 r
. TABLE B.12 (cont'd)
UNIT #3 CVCS PURICICATION DEMINERALIZER Demineralizer B 09:45; 4/29/78 Used for: 92 days Bed Volumes Thru: 3.05(4) Letdown Flow Rate: 55 gpm Reactor Coolant Boron: 681 ppm Nuclide Inlet Activity Outlet Activity Decontamination (uCi/ml) (pC1/ml) Factor 131I 1.25 0.01(-2) <1.4(-6) >R.9(3) 132I 9.9 1 0.2(- ) >1.1(3) 133I 4.97 0.03 - <8.9(-5) 1.2 0.4 (-5) 4.1 1.4(3) 134I 1.75
- 0.03 - <1.1 -4 >1.6 135I 1.06 0.01 - <2.3 -4 >4.6 esRb 1.72 0.05(1) 1.0510.01(-1) 1.64 0.05(0)
- '89Rb 5.4 1 0.2(-2 1.6 0.2(-3) 3.4 0.4(1) 134Cs 5.27 i 0.08( 4) 5.77 0.06(-4) 9.1 0.2(-1) 13sCs <9.1(-5) 2.7 0.2(- <3.4(0) t 137Cs 6.1 0.1(-4) 7.8 0.l(- 9.1 1 0.2 )
13sCs 2.19 0.04(-1) 3.74 0.06 2) 5.9 1 0.1 1390s 1.6 i 0.3(-1) 9.9 4.7(-3) 1.6 i 0.8 24Na 1.03 t 0.01(-2) < 3.9 f,-6) >2.6(3) s1Cr <4.5(-6) <1.7I -7) 54Mn 6.7i0.2(-5) <5.2/-8 >1. 3( 3) 59Fe 9.8 1.4(-6) <1.9I!-7 >5.2(1) 57Co <1,7(-6) <1,4(-7 seCo 1.21 0.09(-4) 1.2 2 0.2(-6) 1.0 0.2(2) 60Co 9.8 0.3 - 1.3 01(-6) 7.5 0.6(1) 65Zn 7.3 1 2.2 - <4.1 - >1. 84Br 1.3 0.1 - <1.1 - >1.2 91Sr 4.2. 0.2 - <1.7 - >2.5 91mY 1.5 i 0.2 - <6.9 - >2.2 9 3Y <2.5(-3) <7.1 - 95Zr 6.8 i 1.5(-6) 7.3 1 3.5(-7) 9.3 4.9(0) 9sNb 5.9 2 1.l(-6) 8.7 2.4(-7) 6.8 2.3(0) 99Mo 7.83 1 0.07 4) <2.3(-6) >3.4(2) 103Ru 5.5 1.2(- >8.7(0) 106Ru 3.7 1.6(- <6.3(-7)(-6) 1.1 0.8 3.4 2 2.9(0) 110 mag <7.5(-5) <5.4(-5) 124Sb 3.2 1 0.8 - 8.4 .4(-7) 3.8i1.1(0) 12sSb 2.7 i 1.9 - <1.7 - >1.6
- 129Te 3.7 1.7 - <2.1 - >l.8
! 131mTe 1.3 0.6 - <3.8 - >3.4 131Te <7.2 - 132Te <3.2(-3) 1.320.3 (-5) <3.3 - >3.9(0) i 139Ba 7.7 i 0.09(-3) 4.5 0.3 - 1.7 0.1(0 140Ba l' 1.12 0.01(-3) 3.2 0.2 - 3.5t0.2(1)) 1 ! La <2.9(-4) 3.8 i 0 1 - <7.6(1) i i 141Ce <3.2(-7) <1.3 - ) 144Ce 3.0 i 1.3(-6 <3.3 - >9.1(0) ; 187W 6.9i1.0(-4)) <1.8 - >3.8(0)
]
239Np- <1.1 (-5) <3.0 - ; B-59 l . l 4
i TABLE B.12 (cant'd) l UNIT #3CVCS PURIFICATION DEMINERALIZER ~ Demineralirer B 10:15; 5/9,78 Used for: 102 days Bed Volumes Thru: 3.40(4) Letdown Flow Rate: 53 gpm Reactor Coolant Boron: 618 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uCi/ml) (uCi/ml) Factor 1311 7.3 0.1(-3 1.5 0 7(-6) 4.9 2.3(3) 1321 1.05 t 0.02( 1) 133I 5.320.1(-2 <1.4 - >3.8 134I 1.72 2 0.03( 1) <1.3 - >1.3 135I 1.010.01(-1) <7.2 - >1.4 88Rb 1.25 0.04(-1) 4.14 0.08(-1 ) 3.0 0.1(-1) 89Rb 5.4 0.2 - <1.7(-3) >3.2(1) Ws 8.0 0.1 - 8.5 0.1(-) 9.4 0.2(-1) 136Cs 7.4 1.0 - 2.12 0.04 - 3.5 1 0.5(0) 137Cs 1.15 0.02(-3) 1.13 0.01 - 1.02 0.02(0) 138Cs 2.03 1 0.03(-1) 4.80 1 0.03 - 4.23 i 0.07(0) 239Cs 8.9 1.9(-2) <1.5(-2) >5.9(0) 2' Na 1.12 ~0.01(-2) <3.6(-6) >3.1(3) 51Cr <2.2(-5) <2.0(-6) 5% 6.9 0.1(-5) <4.2 - >1.6(2) 59Fe 1.3 i 0.1(-5) <7.6 - >1.7(1) 57C0 <l.8(-6) <8.0 - 58Co 1.33 0.03(-4 8.3 1.4(-7) 1.6 0.3(2) 60Co 9.9 2 0.2(-5) ) 1.0 0.1(-7) 9.9 1 1.0(2) 65Zn <3.7(-6) <8.3(-7) 84Br 1.5 1 0.1 - 91Sr 2.7 0. 3 - <6.5(-5) >4.1(1) 91mY 1.1 1 0.2 - 93Y 1.5 i 0.2 - <4.6 - >3.3 95Zr 4.2 1.0 - <1.1 - >3.8 95Nb 1.08 0.06 6) <5.3 - >2.0 99Mo 9.72 0.07 -4) <1.6 - >6.1 103Ru <2.3 - <1.2 - 106Ru <3.4 - <1.6 - 110 mag <2.4 - <5.1 - 124Sb 3.0 1.2(-6) 6.4 1.1(-7) 4.7 i 2.0(0) 12sSb <2.3(-5) <1.9(-6) 129Te 131mTe <2.6(-4) <8.5(-6) >8.5(-6) 131Te. 132Te 6.8 i 2.6(-ti) <1.5(-6) >4.5(0) 139Ba 1.15i0.02(-2) 5.7 i 0.1 - 2.010.05(0) 1408a 1.07 0.02(-3) 5.6 2.5 - 1.9 0.9(2) 140La 51 03- <3.42) 141Ce <2.0(-4) 2.6 0.9 (-6) <1.8 - >1.4 0) 144Ce 1.5 0.1(-4) -<7.1 - >2.11) 187W <4.3 - 239Np <1.4(-5) <6.0 - B-60
TABLE B.12 (cont'd) i UNIT #3 CVCS PURIFICATION DEMINERALIZER Demineralizer B 09:55; 5/16/78 Used for: 103 days l Bed Volumes Thru: 3.66(4) l Letdown Flow Rate: 54 gpm ! Reactor Coolant Boron: 605 ppm Inlet Activity Outlet Activity Decontanination Nuclide (pCi/ml) (uC1/ml) Factor 131I 7.5 1 0.1 - 6.6 0.6 - 1.1 0.1(3)
- ' 1321 g,9 2 0,1 - 3.8 1.4 - 2.6 1.0(3) 133I 5.7 1 0.1 - 2.0 0.3 - 2.9 0.4(3) 134I 1.75 0.01(-1) <1.3(-4) >1.4(3) l lasI 1.0310.01(-1) <4.5(-5) >2.3(3) eeRb 2.92 1 0.05(-1) l' eMb 5.6 <4.7(-2) >1.2(0) 134Cs 4 . 31 0.5(-2))
0.03(-4 4.16 0.05(-4 1.04 0.01(0) i 136Cs 6.1 0.2(-5) 9.3 0.2(- ) ) 6.6 0.3(0) 137Cs 6.02 1 0.04(-4) 5.63 0.04 -4) 1.07 0.01(0) 138Cs 2.24 0.02(-1) 4.96 0.06 -2) 4.52 0.07(0) 139CS <1,7(-1) 24Na 7.5 0.1(-3) 4.4 i 1.5(-6) 1.7 i 0.6(3) 51Cr 5.2 1.6 - <4.7(-6) >1,1 (1 ) 54Mn 9.0 0.2 - 5 1.2(-7) 1.8 0.4(2) 59Fe 1.6 0.2 - <4.1(-7) >3.9 (1 ) 57Co <l .4(-6) 9.9 i 4.4(-7) <l.4(0) seCo 1.86 0.02(-4) 1.1 0.09 1.7 0.1(2) 60Co 1.44 i 0.02(-4) 9 2.(-7)(-6) 1.6 i 0.4(2) . csZn <4.1 (-6) <3.9(-7) 84Br 91Sr 4.1 0.3(-3) <3.3(-5) >1.2(2) 91mY 2.8 0.2(-3) <l.l(-4) -
>2.6(1) 93Y <7.3(-3) <9.7 -
95Zr <5.1 - 95Nb <3.5(-6) 9.5 ' 1.1 (-6) <3.1 - >3.1 99Mo 8.92 0.03(-4) <l.9 - >4.7 l lo3Ru 4.4 i 1.0(-6) <5.6 - >7.9 lo6Ru <2.1(-5) <3.2 - 1 110 mag <2.2(-6) <2.7 - 124Sb 4.1 0.9(-6) 6.5i0.9(-7) 6.3 1.6(0) 12sSb <9.1(-6) <l.2 - l 129Te 2.7 2.4(-2) <8.9 - >3.0(0)
- 131mTe <3.5(-4) <l.1 -
131Te 8.9 1.7(-3) <5.5-3) >1.6(0)
; 132Te 7.8 1.9(-6) <1.5 -6) >5.2(0) l 139Ba 2.6710.04(-2) 8.2*0.4(-3) 3.3 0.2(0) l 140Ba 1.45 0.03(-3) 5.9 1.2(-6) 2.5 0.5(2) 140La <1.3 - 1.01 0.4(-7) >1.3(3) i 141Ce <3.6 - <9.5 -
144Ce <1.3 - <2.3 - 187W <5.0 - <3.4 - 239Np 1.4 0.3(-4) <9.3 - >1.5(1) B-61
TABLE'B.12_(cont'd)
+
UNIT #3CVCS PURIFICATION DEMINERALIZER Demineralizer B
-21:20; 5/19/78 l Used for: 112 days i
Bed Volumes Thru: 3.78(4) Letdown Flow Rate: 52 gpm Reactor Coolant Boron: Inlet Activity Outlet- Activity . Decontamination Nuclide (uCi/ml) (vCi/ml) Factor l' 131I 6.8- 0.1(-3) 3.0 0.08(-6) 2.3 0.1(3) 1321 1 .05 0.02 -1) <1.3(-6) >8.1(4) l 133I 5. O.04 -2) 2.3 i 0.7(-6) 2.5 0.8(4) i 13"I 1. ) 0.02 -1) <1.2 -6 >l.4(5) 135I 1. 2 0.02(-1 ) <1.3 -6 >7.8(4) - esRb 1.12- 0.02(-1 3.7 0.7 3.0 i 0.6(3) 89Rb 4.22 0.07(-2 <5.7(-6) (-5) >7.4(3) - 134Cs 3.98 1 0.02 -4 4.1 0.1(- ) 9.7 0.2(-1 , 136Cs- 6.3' 7.4 0.3(- ) 8.5 1 0.4(0)) 5.68 O.2(-)) 137Cs 0.04 -4 5.43 1 0.02 -4) i.05 t 0.G1(0 13eCs 1.65 i 0.03 -1 6.9 0.9 2.4 i 0.3(4) ) lats 6.6 i.0.7(-2) ) <4.6(-5) (-6) >1.4(3) L 8.2 >2.2 24Na 0.." -3) <3.8 -7 ' , SICr 8.010.6[-5) . 7.5 -6
< >l .l((4) l)
l 54Mn 1.01 i 0.01(-4) 3.2't0.3(-6) 3.2 2 0.3(1) 59Fe 1.7 0.1(-) <7.3(-7) >2.3(1) 57Co 1.7 0.5(- ) <5.2(-7) ' >3.3(0) seCo 2.98i0.03-4) 6.94 0.03(-6) 4.29 2 0.05(1) ! 60Co 1.91 1 0.04 -4) 1.8 0.6 1.1 0.4(2) , l i-ssZn e4Br
<3.5(-6) <8.9(-7) (-6) '
91Sr 2.9 i 0.7(-3) <1.4 - >2.1(3) 'l 91mY 9.3 i 1.8(-4) <2.0 - >4.7(2) l 93Y <1.3 - l 95Zr <3.3(-3)(-5) 2.6 0.1 .
<9.0 - >2.9(1) l i
ssNb- 1586 i 0.08(-5) 2.3 05(-6) 8.1 1.8(0) 99Mo 7.15 i 0.06(-4) <1.4 - >5.1(2) i [ 103Ru 6.6*0.8(-6) <1.1 - >6.0(0). 106Ru - <1.3(-5) <6 . 3 - 11on#9 2.62.1.0(-6) -<4.7 - >5.5(0) ! 124Sb 7.4-i 0.6(-6) 1.2 1 0.2(-6) 6.2 1.1(0) _ 12sSb- .<1.8(-5) <2.5(-6)- 129Te' <2.5(-5) 131mTe <1.9(-4) 2.8211(-5) <6.8(0)
. 131Te <1.8(-3) <1.8 -
132Te ~< 1.4 - >1.2(1)
'139Ba 1.7 2 0.3(- ) }2.2 t 0 2(-5) 1.1310.04-2 <4.9 - >2.3(2) 140Ba 1.35 i 0.01 ' 6.1 i 0.6(1) 14oLa' 6.38i0.08-3l 141Ce -<2.7 <1.1 -
c144Ce- <1.2 - <5.9 - 187W gg,g . <g,4 2 9Np : <1.9 5) <8.0 - B-62
-,---c---_a- - - - - _----_.-s
TABLE B.12 (cont'd) UNIT #3CVCS PtJRIFICATION DEMINERALIZER Demineralizer B 20:00; 5/21/78 Used for: 114-days Bed Volumes Thru: 3.85 (4) Letdown Flow Rate: 55 gpm
- Reactor Coolant Boron:
Inlet Activity Outlet Activity Decontamination Nuclide (uC1/ml) (uCi/ml) Factor 131I 1.88 i 0.03(-2) 3.5 0.2(-6) 5.4 i 0.3(3) 132I 7.16 0.07(-2) <1.5(-6) >4.8(4) 133I 7.2 0.1 (-3) <1.2(-5) >6.0(2) 1341 135I 3.4 0.6(-4) <9.0(-7) >3.8(2) serb 89Rb 134Cs 1.38 0.01 - 3.78 0.05(4) 3.65 0.06(0) 136Cs 5.65 0.05 - 9.5 0.2(-6 6.0 0.1(1) 137Cs 1.90 0.02 - 5.18 0.01( 4) 3.67 0.04(0) 13eCs 139CS 24Na 3.2 0.2-4) <4.l(-7) >7.8(2) s1Cr 1.8 0.2 -4) <2.3(-6) >7.8(1) 54Mn 1.74 0.01(-4) 4.0 1 0.7(-7) 4.4 0.8(2) 59Fe 4.0 0.2(-5) <2.7(-7) >1.E(2) 57C0 2.6 i 1.2(-6) <1.3(-7) >2.0(1) ssCo 8.08 1 0.07(-4) 4.03 0.01(-6) 2.00 0.02(2) 6000 4.7i0.1(-4) 3.2 0.3(-6) 1.5 0.1(2) 652 , 1.1 0.2(-5) <2.1(-7) >5.2(1) BW37 91Sr 3.6 0.4 - <3.1 - >1.2 2) , 91mY 5.6 2 0.7 - <2.0 - >2.8 3) 93Y 3.6 i 1.4 - <2.0 - >1.8 3) 95Zr 3.0 0.1 - 3.7i1.6(-7) 8.1 3.5(1) 9sNb 2.9 0.1(-5) 9.0 0 8(-7) 3.2 0.3(1) 99Mo 4.59 i 0.01(-3) <1.5 - >3.l(3) 10 3Ru 2.1 0.1 <2. 3 - >9.1(1) 106Ru <2.0(-5)(-5) <1.6 - tiongg 4.8 1.6(-6) <l.4 - >3.4(1) 124Sb 2.4 2 0.1(-5) 1.93 0.07(-6) 1.24i0.07(1) 12sSb <6.3(-5) 2.1 0.4(-6) <3.0(1) . 129Te 131mTe <7.5(-5; <2.4(-6) l 131Te 132Te 4.9 0.4(-5) <1.1(-6) >4.5(1) t .13988 l 140Ba 3.11 0.03(-3) 3.7 0.3(2) l 140La 4.76 0.04(-3) -i.510.6(-6)) 1.37 0.04(-5 3.5 0.1(2) l 141Ce <7.6 - <4.6 -7) l 144Ce <2.6 - <1.0 -6 187W <],7 - <6.0 -6 239Np <2.7 - <9.3(-6 I B-63 i
TABLE B.12 (cont'd) UNIT #3CVCS PURIFICATION DEMINERALIZER Demineralizer B 09:10; 5/25/78 Used for: 118 days Bed Volumes Thru: 3.97(4) Letdown Flow Rate: 50 gpm , Reactor Coolant Boron: Inlet Activity Outlet Activity Decontamination
- Nuclide (uCi/ml) (uCi/ml) Factor l- 131I 7.65 0.06(-3) 1.6
- 0 6(-6) 4.8 1 1.8(3) 132I 8.5 1 0.3(- <1.1 - >7.7(2 133I 6.0 0.1(- <1.9 - >3.2(3 134I 1.92 0.0a -1) <l.0 - >1.9(3 135I 1.11 0.02(-)) <2.0 - >5.6(2 l esRb 3.70 0.09-1) 1.05 0.01(-1) 3.5210.09(0) l 89Rb 6.3 1 0.5(- ) 2.2 2.9 0.3(1) 0.2(-3) 0.04(-4 )
134Cs 3.99 0.04 -4 3.32 1.20 0.02 136Cs 1.12 s 0.05 -4 9.220.3(- 1.2110.07 137Cs 5.6510.08(-4 4.2 1 0.3(- 1.34 0.09 l 13eCs 2.44 0.05(-1 4.20 0.07 2) 5.81 0.2(0) l l 139Cs 1.5 1 0.7(-1) <7.9(-3) >1.9(1) 24Na 8.8 0.1(-3) <1.3(-5) >6.8(2) 51Cr 7.2 1 0.1(-5) 8 3(-7) 9.0 3.4(1) 54Mn 6.6010.09(-5) 2.9 0.1(-6) 2.30 0.08(1) 59Fe 1.12 0.06(-5) <5.4(-7) >2.1(1) 57Co <6.(-7) 7.2 3.0 <8. (-1 ) ssCo 2.17 0.03(-4) 5 1 2(-7)(-7) 4.3 i 1.7(2) 60Co 1.38 t 0.05(-4) <8.4 - >1.6 3.710.4(-6) <4.6 - >8.0 l 65fn 84 [ 91Sr 2.2 1 0.2 -3) <1.5 - >1.5 l 91mY 1.8 0.2 - <8.4 - >2.1 l 9 3Y 6.2 1.2 - <7.8 - >8.0 l 95Zr 6.410.3- 1.4 0.6(-6) 5 2(0) l 95Nb 6.5 0.3 -6) 1.6 0 2(-6) 4.1 0.5(0)
- 99Mo 8.61 0.0f(-4) <2.4 - >3.6(2) 103Ru 4.1 i 0.3' -6) <3.6 - >1.1(1) lo6Ru <5.6 6' <5.4 -
110 mag <8.9 -
<1.8{.6)(-6) l 124Sb 5.3 1 0.4 1.33 0.09(-6) 4.0 0.4(0)
I 12sSb 1.8 0.6(-6) >2.6(0) 129Te <6.3(-2) <3.5(-2 i 131mTe <4.0(-4) <7(-7) 1.1 0.8 - ) ( 4)<3.6(0 131Te 3.3 1 1.5(-3) <5.2(-4) >6.4(0 132Te 1.A t0.3(- >1.2(0 139Ba 3. e : 0.1(-
<1.2(-6)(-3) 7.2 2 0.4 4.6i0.3(0).
140Ba i.18 0.01 3) 4.1 i 0.9(-6) 2.9 0.6(2) 140La <1.2 -4 <6.4(-5) 141Ce <9.4 7 8 i 4( ) <1.2(0) 144Ce <2.6 - <4.0 - 187W - gg,7 , 44,6 - 239Np <2.7 - <8.8-6) B-64
TABLE B.13 UNIT #4 CYCS PURIFICATIOh sdMINERALIZER Demineralizer A 16:09; 11/30/77 t' sea kr: 180 days Bed Volumes ihru: 6.3(4) Letdown Flow Rate: 60 gpm Reactor Cooiant Boron: 651 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uci/ml) (uCi/ml) Factor 131I 5.97 i 0.03(-3) <1.6 - >3.7(3 1321- 6.8 0.1(-3) <1.0 - >6.8(1 1331 7.39 1 0.06(-3) <6.2 - >1.4(2 1 34I 135I 6.9t0.3(-3) 51.4 - >4.9(1 8.0 1 0.2(-3) <1.2 - >6.7(1) 88Rb 4.4 0.7(-2) 2.5 0.3 1.8 1 0.4(0) 89Rb 5.7 0.5(-3) <6.2(-4) (-2) >9.2(0) 134Cs 1.1310.01(-3) 5.77 0.07(-4) 1.96 0.03(0) 136Cs 5.0 0. 3(-5) 3.08 0.04(-5) 1,62 1 0.10(0) 137Cs 2.19 1 0.02(-3) 9.5 0.l(-4) 2.31 0.03(0) 138Cs 3.2 1 0.l 3.2 0.3(-3) 1.00 1 0.10(1) 139Cs <5.2(-3)(-2) 3H 1.60 0.05(-1) 1.63 0.05(-1) 9.8 t 0.4(-1) 14C 3.2 0.3 - 7.9 0.8(-5) 4.1 0.6(-2) 24Na 6.9 0.1 - <2.6(-5) >2.7(2) 51Cr 1.3 0. 3 - <l.2(-6) >1.1(2) 54Mn 1.68 0.03(-4) 2.2 0.1(-6) 7.6 0.4(1) ssFe 5.3 1 0.5 1.45 0.04(-5) 3.7 i 0.4(0) 59Fe <1.1(-5) (-5) <6.9(-7) 57Co 2.2 0. 2(-5) <l.3(-6) >1.7(1) 58Co 1.49 0.02 - 1.03 0.01(-4) 1.45 0.022) 60Co 2.45 0.04 - 1.98 0.04(-5) 1.24 i 0.03 2) t 63Ni 1.49 i 0.04 - 2.8 0.1(-6) 5.3 0.2(0 65Zn <1.2(-5) <4.8(-7) 89Sr 3.2 0.1(-5) 1.4 0.3(-7) 2.3 i 90Sr 6.3 i 0.6 0.5(2) l 7 3(-9) 9.0 4.0(1) ! 91Sr <l.6(-4) (-7) <l .0(-4) 91Y 4 1 2(-7) 2.4 0.2(-7) 1.7 1 0.8(0) 93Y <1.5(-4) <5.8(-5) 95Zr 1.6 0.4(-5 <7.2(-7) >2.2(1) 95Nb 3.7 0.9(-5 1.4 0.4(-6) 2.6 1.0(1) 99Mo 4.6 0.2 <2.5(-6) >1.81) 103Ru <1.1(-5)(-5 1.1 0.4(-6) <1.0 l) , 110 mag 5.7 0.9(-5) <1.3(-4) >4.4 -1) ! 124Sb 1.06 0.03(-4 1.6 0.1(-6} 6.6 0.5(1) l 12sSb 129mie 3.1 0.7(-5) ) 6.3 2.4(-6) 4.9 2.2(0) 1.8 1 0.8(-4) <8.4(-7) >2.1(2) 132Te <2.4(-6) l 139Ba <1.3(-5) 4.7 0.1 (-3) 3.61 t 0.09(-3) 1.30 1 0.04(0)
- 140Ba 5.2 1 1.4 <2.1 -6 >2.5(1) i 140La <1.6 - ) (-5) <8.6-7)f
' 141Ce <1.1 -6 143Ce <5.0- - )I
<3.5 <l.0-5) 144Ce <9.i -6ll <9.7 -
187W l.7 0.2 <3.8 - >4.5(1)
.239Np <1.2(-5) (-3) <3.6 -
.B-65
TABEE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 14:55, 12/6/77 Used for: 186 days Bed Volumes Thru: 6.5(4) Letdown Flow Rate: 60 gpm Reactor Coolant Boron: 630 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uC1/ml) (uCi/ml) Factor 131I 3.83 0.08(,- <4.2 - > 9.1 (, 1321 9.06 0 . 0 91 -
<9.2 - >9.81 ,
l ' 3I 1.63 0.02Il- <1.1 - >1. 51 , 1 34I <1.2 - >8.21 1
;2 '30 9.8 1.05 1 0.01(-2 0.4(-3) ) <1.7 - > 6 . 2 I, eaRb 5.8 1 0.5(-2) 1.5 0.2(-2) 3.9 0.6(0) 89Rb 7.2 0.3(-3) <3.6(-4) >2.0(1) _
134Cs 1.02 0.01(-3 1.13 0.02( 3) 9.0 0.2 - ) 13sCs 7.7 i 0.8(-5) ) 4.7 0.2(-5 1.6 0.2 0 137Cs 2.06 i 0.03(-3) 2.12 0.03( 3) 9.7 0.2 - ) j 13sCs 4.0 0.1 3.2 0.2 1.25 i 0.08(1) 139Cs <2.1(-3)(-2) <8.4(-4)(-3) l
- 2'Ha 9.1 0.2(-3) <3. 0(- 6) > 3.0(3) 51Cr <3.5(-5) <3.9(-6) 54Mn <5.7(-5) 1.06 0.07(-5)- <5.4(0) l 56Mn <2.7-4) <1.3 -
L 59Fe <1.5 -5) <1.7 - 1 I 57Co <6,6 6) <6,6 - I seCo 4.0 1 0.1(-4) 5.8 0.6(-5) 6.9 0.7(0) ) 60C0 4.4 i 0.7(-5) 4.4 0 7( F) 1.0 0.2(1)
- 6 sZn <l.0(-5) <1.4 -
91Sr <5.2(-5) <1.1 - l 93Y 6.1 1.6(-4) <2.0 - >3.1 95Zr 1.1 0.3(-5) <2.4 - >4.6 95Nb <2.9(-5) 4.6 1 0(-6) <6.3 99Mo 6.3 0.3(-5) <8.0(-6 >7.9 10 3Ru <7.7(-6) <4.5(-6 110 mag 1.8 1 0.8(-4) <3.0(-4 >6.0(-1) 124Sb <6.8(-Q 6.1 1 2(-6) <1.1(0) 125Sb <6.9(-6 <4.2 - 129mie <2.8 - 132Te 13988
-6.810.1(-3)
<5.1 (-5)))4.9
<1.8(-5 <7.9 -
01(-3) 1.3910.03(0) 140Ba 4.5 1.1(-5) <4.0 - >1.1(1) 14cLa <1.2-) <1.9 - l 141Ce <1.1 - ) <1.1 - , 143Ce <1.5 <6.4 - l 144Ce <7.3 - ')i <6.6 - 187W 1.3 0.1(-3) <1.1(-5 >1.2(2) j 239Np <9.3(-6) <8.4(-6 B-66
TABLEB.13(cont'd) LNIT #4 CVCS PURIFICATION DEMINERALIZER , Demineralizer A 12:42; 12/8/77 I Used for: 188 days Bed Volumes Thru: 6.6(4) !- Letdown Flow Rate: 60 gpm
- Reactor Coolant Boron
- 637 ppm l Inlet Activity Outlet Activity Decontamination Nuclide (uC1/nl) (pCi/ml) Factor i 131I 6.8 0.1 - 1.1 0.3(-5) 6.2 1.7(2) 1 32- g,1 i o,1 -
' >2.0(2) 1 9.8 0. 3 - <4.6(-5) 1.1 0.4 (-4) 8.9 1 3.2(1) 134I 9.7 0.2 - <4.6(-5) >2.1(2) lasI 9.7 1 0.2 - <2.9(-7)*. >3.3(4) serb 5.3 0.3(-2) 1.18 0.09(-2) 4.5 0.4(0) 89Rb 7.5 1 0.2(-3) 3.8 0.9(1) 134Cs 136Cs 1.14 i 0.02(-3) 2.0 1.09 1 0.01(-3 0.5(-4) ) 1.05 0.02(0) 4.3 0.2(-5) 1.2 1 0.1(0) l 137Cs 5.2i0.5(-)).
2.07 3.77 0.02 -3 1.98 0.02(-3) 1.05 i 0.01(0) 138Cs 0.09 -2) 2.4 0.1(-3) 1.57 1 0.08(1) 139Cs 8.7 2.1(-3) <3.6(-4) >2.4(1) 24Na 7.7 1 0.2 - 3.4 0 1(-6)* 2.3 0.1 siCr 1.6 0.2 - 1.0 0.1(-5
- 1.6 1 0.3 54Mn 5.5 0.3 - 7.1 1.2(-7
- 7.7 1 1.4 59Fe 1.1 1 0.4 - 3.7 1.5(-7
- 3.0 1.6 57Co 6.2 3.4 - * <3.3(-7)* >1.9(0) l seCo 3.5 i 0.1 - 4.7 0.2(-5) 7.4 0.4(0) 60Co 5.2 0.5 - 6.9 1.6(-6) 7.5 i 1.9(0) 6sZn <g,2(-7) <2,4 -7)*
l 91Sr 1.11 0.08(-4)* <3.6 -6)* > 3.1 C ) 93Y' 6.1 0.3 - * <1.5 -7)* >4.1(3)
- 95Zr 7.9 0.2 -
- 9.1 i 1.8 -7)* 8.7 1 1.7 95Nb 7.7 1.9 - 4.6 1.1 - 1.7 0.6 99Mo 8.3 1 0.7 - 6.6 0.1 -
- 1.310.1 103Ru 8.6 3.5-7)* 6.6 1.4 -
- 1.310.6 110 mag <5.9(-4) <3.1(-4)*
124Sb 6.2 i 1.2(-6)* 2.1 0.7-6) 3.0 1.1(0) l 12sSb <l.3-6)* 4.0 3.6 -7)* <3.3(0) l 129mTe <6.5-6) 2.2 1.9 -7)* <3.0(1) 132Te <1.5 -6)* <1.0(-6)* 139Ba 7.7 0.2(-3) 4.32 i 0.09(-3) 1.8i0.1(0 i 140Ba 4.9 1.4(-5) 2.6 0.7(-5) 1.9i0.7(0)) 140La 3.53 1 0.05(-5)* 1.9 0.5(-7)* 1.9 i 0.5(2) 141Ce <1.0(-6)* <4.0(-7)*
- 14 3Ce <2.4(-6)* <1.4(-6)*
144Ce <4.0(-5)* <3.2(-7)* 187W 1.15
- 0.05(-3)* <1.5(-5)* >7.7(1) 239Np 6.2 1.0(-6)* 2.2 l 0.2(-6)* 2.810.5(0) i i B-67 L
TABLE B.13 (cont'd) i UNIT #4 CVCS PURIFICATION DEMINERALIZER ! Demineralizer A 08:38; 12/9/77 Used for: 189 days Bed Volumes Thru: 6.7(4) Letdown Flow Rate: 62 gpm l Reactor Coolant Boron: 638 ppm r l Inlet Activity Outlet Activity Decontamination l Nuclide (uCi/ml) (pCi/ml) Factor i 1311 9.810.l(-2) 2.33 0.06(-5)* 4.2 0.1(3) 1321 3.5210.05-2) <8.2(-5) >4. 3(2) 133I 1.14 0.01 -1) 2.27 0.03(-5)* 5.02 0.08(3) l 134I 1.03 0.04 -2) >1.3(2) i 135I 5.2 0.2(,2) <8.2(-5) 7.5 0.4 (-6)* 6.9 0.5(3) 88Rb 5.7 0.7(-) 1.6 0.2(-2) 3.6 0.6(0 89Rb 7.9 0.5(- ) 8.1 1.7(-4) 9.8 2.1(0 134Cs 4.82 0.05 -3 8.2310.09(-4) 5.86 0.09 0) 136Cs 2.65 0.04 -3 3.6 7.4 0.4(1) l 137Cs 7.71 0.09(-3 1.52 0.2(-5) 0.02( 3 ) 5.07 0.09ll0) ! i 13eCs 5.0 0.1(-2) 3.2 0.3(-3) 1.6 i 0.1(1) l 139Cs <2.6(-3) <1.0(-3) 24Na 4.510.2-3) 6.1 1 0.4 -
- 6.4 1 0.6 )
t SICr 2.1 0.3 -3) 4.2 1. 3 - 5.0 1.7 / , 54, 3.3 0.1-4) 5.7 0.9 - 5.8 0.9 l 59Fe 2.1 0.1 -4) 1.2610.06(-6)* 1.7 0.1 ') 57Co <9.4(-6)* <g,g(.7)* 56Co 5.06 0.04(-3) 1.2710.03(-4) 4.0 0.1 GoCo 3.9 0.2 - 3.0 0.2(-5) 1.3 0.1 65Zn 5.6 1.5 - 2.6 i 1.0(-7)* 2.2 i 1.0 0.4
- 91Sr 1.1 <8.2(-6)* >1.3(1) 93Y 5.1 1 0.7 - <2.1(-7)* >2.4(4) 95Zr 1.3 0.1 - 1.0 0.2 - 1. 3 i 0. 3 )
9sNb 1.310.2- 9.4 1.8 - 1.4 0.3 99Mo 2.4310.02(-3) 4.3 0.2 -
- 5.7 0.3 ')l 103Ru <l.l(-5)* <6.0(-6)*
l 110 mag 2.9 1.3(-4) <2.2(-4)* >l.3(0) 124Sb 5.210.7(-5) 7.9 1.4(-6) 6.6 1.5(0) 12sSb <1.0 -
- 7.2 i 3.2(-7)* <1.4(1) 129mTe <3.2 - <3.3(-7) 132Te <1.7 - * <6.2(-7)*
139Ba 6.3 0.4 - 4.7 i 0.2(-3) 1.3 0.1(0) 140Ba 8.8 4.1 - * <7.4(-7)* >1.2(2) 140La 2.0 0.2 - 4.6 0.4(-7)* 4.310.6(2) 141Ce <1.8 - * <3.1(-7)* 143Ce <2.4 - * <1.3(-6)* 144Ce <5.2 - 2.1
- 1.3(-7)* <l.5(2)
- 187W 2.3i0.3(-3) <4.2(-5)* >5.5(1) 2 39Np - - 1.9 1 1.1 (-5)* 1.9 1 0.2(-6)* 1.0 1 0.6(1)
B-68
,.D TABLE B.13 (cont'd)
LMIT #4 CVCS PURIFICATION DEMINERALIZER Domineralizer A 09:18; 12/10/77 Used for: 190 days Bed Volumes Thru: 6.7(4) Letdown Flow Rate: 65 gpm Reactor Coolant Boron: 644 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uC1/ml) (vCi/ml) Factor 1 I 2.00 0.04Il- 1.58 0.03(-6)* 1.27 0.03(4) 33] I 1.00 0.05i -
<8.0(-5)- >1.3(2) 133{
134 1.88i0.03(- 8. 4 1 1. 3(-7)* 2.2 0.3(4) issI <1,2(d 1.23 0.02(-2) <1.2(-6)* >l.0(4) 88Rb l 3.3 0.6(-2) 89Rb <1,)(.3) 134Cs 1.19 136Cs 0.01(-3) 7.1 0.1(- ) 1.68 i 0.03(0) ' 3.35 2 0.09(-4)- 2.31 0.05 -5) 1.45 0.05(1) 137Cs 13eCs 2.2210.02(-3) 1.38 0.02 -3) 1.61 0.03(0) 139Cs 3.9 i 1.2(-3)
<8.3(-3) 24Na 51Cr.
5.0310.09(3) 6.7 1.7 -
- 7.5i1.92 2.5 i 0.5 -4) 2.9 0. 3 - 8.6 i 1.9 0
! 54Mn 8.2 1 0.2 - 4.9 0.2 - 1.67 0.08(1)- i 59Fe 2.9 i 0.2 - 2.1 i 0. 3(-6 1.4 1 0.2(1) s7Co 1.9 0.4 - *
<3.8(-7)* .>5.0(0)
- ssCo 8.2 0.1 - 1.01 0.02(-4) 8.1 0.2(0)
i 60Co 8.5 i 0.2 - 1.70 0.06(-5) 5.0 0.2(0) 65Zn 7.3 1 0.3 - * <2.8 - * >2.6
- 91Sr 2.2 0.2 - <5.6 - >3.9
'say 3.8 t 0.8 -
<6.0 - * >6.3 95Zr 2.2 0.2 - 2.9 i 0. 3 95Nb
- 7. 7
- 0.4 (-
1.7 0.4 - 9.0 0.6(- 1.9 . 0.5 8'M0 7.14 = 0.06( 4 1.05 i 0.01 6)* 6.8 0.1 i 103Ru 11 "A9
.3.8 'i 1.4(-6) ) <1.2(-6)* >3.2(0) 124Sb
<3.8(-4)* <2.2(-4)*
l 7.9- 1.1 6.310.4(-6) 1.3 0.2(0) 12sSb 129mTe
<4.5(-6) (-6) - 3.2 0.6(-6) <1.4(0) l <4.5(-6)*- <4.6(-7) l l 132Te 7.7 i 2.3 9.3 + 3.0(-7)*- 8.3 3.6 l 139Ba 140Ba
<1.4(-3) (-6) 4.07 C 09(-3) <3.4(-1) (0) 7.9 0.6(-5)- <1,1(-6)* >7.2(1) u 140La 141Ce 2.03 i 0.07(-5) 6.2 1.1(-7)* 3.3i0.6(1)
-<2.8(-6)* <4.3(-7)*
i 143Ce
<8.3(-6)* 5.5 2.2(-6)* <1.5(0)
- 144Ce
<2.5(-6)* <4.0(-7)*
187W 4.6 0.5(-4). <2.8(-5)* >1.6(1) 239NP -1.4 0.6(-5) 1.5t0.8(-6)* 9.3i6.4(0) i B-69 I-
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 10:54; 12/11/77 Used for: 191 days Bed Volumes Thru o./(4) Letdown Flow Rate: 60 gpm Inlet Activity Outlet Activity kcontamination Nuclide (uti/ml) (pCi/ml) Factor 1311 7.96 0.09 - 8.9 0 1(-7)* 1.11 i 0.02 - 8.9 i 0.1(3) 1321 <1.0 - >l.1 1331 1.28 0.02 - <6.0 - * >2.1 134I 1.09 0.03 - <1.1 - >9.9 1351 1.18 0.04 - <1.3 - * >9.1 88Rb 6.1 1.6 -2) 1.7 0.3(-2) 3.7 89Rb 4.2 1. 5 - 1.2(0)
<4. 3(-4 ) >9.8(0) 134Cs 7.7 0.2 -
6.8 i 0.7 - 7.4 1 0.1(-4) 1.04 2 0.03(0) 136Cs 2.36 0.05(-5) 137Cs 1.51 0.03(-3) 2.9 t 0.3(0) 1 .39 i 0.02(- 3) 1.09 t 0.03(0) 13eCs 4.1 1 0.3(-2) 1.9 0.2 2.1 0.3(1) 139Cs <1.8(-3) (-3) 24Na 5.7 0.2(-3) 3.8 1.1 -
- 1.90 0.07(-4) l.5 0.4(3 51Cr 9.5 5.7 - 2.0 1.2(1 54Mn 6.5 0.o-5) 9.6 0.9 - 6.8 0.7(1 59Fe 2.3 0.1 -5) 9.7 1 2.4(-7 57Co I.I i 0.2 - 2.4 1 0.6(1 6.8 0.2 - <4.8(-7)* >2.3(0) 58Co 2.13 0.05(-5) 3.2 0.1(1) 60Co 6.3 i 0.2 - 2.7 0.1(-6) 2.6 0.7 - 2.3 0.1(1) 65Zn <3.2 - * >3.1 903p 7.9 i 0.7 -4)* <6.6 - * >1.2 91y 3.1 0.2 -3)* <l.0 - * >3.1 95Zr 1.80 0.08(5) 3.310.3-1.8 1 0.2 - 5.4 0.6 95Nb 4.3 0. 3 - 4.2 0.5 99Mo 7.9 2 0.2 -
- 6.1 0.1 -
- 1.3 0.1 10 3Ru 1.9 0.3 - * <1.1(-6)*
110 rag <2.6(-4)*
>l.7(0)
<2.2(-4)*
124Sb 8.1 0.7(-6) 2.2 0.2 3.6 0.5(0) 12sSb <1.4(-6)* <l .0(-6) (-6) 129mTe <2.3(-6) 2.2 0.7(-5) <l.0(-1) 132Te 5.6 i 0.6 - * <l.9(-6)* >2.9(0) 139Ba 5.1 0.2 - 4.5 0.1(-3) 140Ba 7.8 1 0.7 - 1.1310.06(0)
<1. 3(- 6)* >6.0 140La 6.7 i 5.5 - * <4.2(-7)* >1.6 i
141Ce <l.9(-6) 1.2 0.4(-6) >1.6 14 3Ce 2.6 1 2.4(-6)* <4.2-6)* >6.2-1) 144Ce <1.5(-6) <5.2-7)* 187W 9.0 2.2(-4) <2.6-5)* 239Np <1. 8(- 6)* 1.2 >3.5(1) 0.6(-6)* <1.5(0) B-70
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 09:31; 12/12/77 Used for: 192 days Bed Volumes Thru: 6.8(4) Letdown Flow Rate: 53 gpm Reactor. Coolant Boron: 616 ppm Inlet Activity Outlet Activity Decontamination
-Nuclide (vCi/ml) (uci/ml) Factor 1311 8.1 4.0 01(-7)* 2.0 1 0.1(4) 132I 3.48 0.1(-3) 0.07(-3 ) <8.4 - >4.1(1 1331 7.7 1 0.1 - <9.7 - * >7.9 134I 3.6 0.1 - <3.3 - >1.1 135I 4.8 0.2 - <4.2 - * > 1.1 88Rb 5.0 1.0(-2) 1.5 1 0.7(-1) 3.3 i 1.7(-1) 89Rb . <2.2(-2) 134Cs <4.4(-3) 7.3 1 0.2 (-4 7.22 1 0.06 - 1.01 0.020) 136Cs 5.7 0.7(-5 2.3210.05- 2.4 0.3(0 137Cs 1.50 0.03( 3) 1.39 0.01 - 1.08 2 0.02 0) 13eCs 1.3 0.2(-2) 1.34 0.05 - 9.9 1.5(-1)-
139Cs <4,6(.2)- 3J 1.48 0.01(-2) 1.27 0.01(-1) 1.17 0.01(-1) IN 1.41 0.03( 6) 6.82 0.02 5) 2.07 0.04(-2) 24Na 3.26 0.09( 3) 2.4 0.4 -
- 1.4 0.2(3 32P 3.6 2 0.1 - 1.6 0.5 - 2.3 0.7(2 51Cr 8.3 1.0 - 4.5 0.3 -
- 1.8 0.3(1 54Mn 6.2 0.2 - 1.3 1 0.2 - 4,9 1 0.8(1 55Fe 4.47 t 0.06(-6) 4.43 1 0.06 6) 1.0110.02(0) 59Fe 1.61 0.08(-5) 2.2 i 1.2(-
- 7.3 4.0(1)
'57Co 1.1 0.3(-6) <3.6(-7)* >3.1(0) seCo 4.210.1(-4) 4.28 i 0.04(-5)~ 9.8 0.3 0 60Co 7.7 0.2(-6) 8.5 1 0.3 0 63Ni 6.5 1.25 i 0.1(-5) 0.07(-6 ) 2.27 1 0.07(-6) 5.5 0.4 - )'
65Zn 1.9 1 0.4(-6)* <2.3(-7)* >8.3(0) 895r 5.1 0.2(-6) 3 1(-7) 1.7 0.6(1) 90Sr 8 1 2(-8) <1.4(-7) >5.7( ) 91Sr .1.810.3(-4)* <1.2(-5)* >1.5( 91Y 1.1 0.4(-7) 7 i 4(-8) 2 1 ) 93Y 1.3410.09-3)* <5.1(-7)* >2.6(3) 95Nb E.0 1 0.9 - 3.0 i 0.3(-6) 1.7 1 0.3(0) 99Ho 6.5 i 1.8 - 4.29 1 0.05(-7)* 1.5 0.4(2) lo3Ru 8.3 3.7 - * <7.2(-7)* >1.2(0) 110 mag <2.3(-4)* <2.1(-4)* 124Sb 6.210.5(-6) 2.4 1 0.2(-6) 2.6 0.3(0) : 125Sb 1.2 1 0.8 <1.1(-6)* '1.1(0)
-129mTe <2.1(-6) (-6)* 3.6(-7)*
132Te 3.8 i 0.5(-6)* <9.0(-7)* >4.2(0) 139Ba 1.7 0.2(-3) 2.90 1 0.06(-3) 5.9 i 0.5(-1) 140Ba 1.06i0.07(-4) <8.1(-7)* >l.3(2) 140La 1.00 1 0.03(-4)* 9.7
- 7.5(-8)* 1.0 0.8(3) 141Ce <1.4 - * < 3.4 (-7)*
- 143Ce <2.2 -
- 4.4i1.5(-6)* <5.0(-1)
, 144Ce - <1.3 - <3.2(-7)*
- 187W 3.7 0.6(-4)* <1.2(-5)* >3.1(1) 239Np 2.6 1.8(-6)* 1.2 1 0.4(-6)* 2.2 i 1.7(0)
.B-71
- j
I l- TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER l Demineralizer A 14:14; 12/13/77 l- .Used for: 193 days Bed Volumes Thru: 6.8(4) L Letdown Flow Rate: 62 gpm Reactor Coolant Boron: 620 ppm
; Inlet Activity Outlet Activity Decontamiration ! Nuclide (uCi/ml) (vC1/ml) Factor 1311 5.93' O.07(-3) .8.3 0.2 7.1 i 0.2(3) 132I 133I 1.09 0.02(-2) <2.2(-5) (-7)* >5.0(2) l 7.01 0.2(-3)* '2.1 0.3(-7)* 3.3 0.5(4) 134I 1.09 0.05(-2) <2.2(-5) >5.0(2) 13sl 1.14 0.03(-2) <3.4(-7)* >3.4(4) serb 5.0 0.1-2) 7.5 0.6 - 6.7 0.5(0) 89Rb 7.3 0.2 -3) 9.8 2.1 - 7.4 1.6(1) 134Cs 7.0 0.2 -4 7.2 1 0.1 - 9.7 i 0.3(-1) 13sCs 3.0 1 0.8 -5 2.19 0.05 5 1.4 i 0.4(0) l 137Cs 1.40 0.03 3) - 1.36 1.03 t 0.03(0) 13ECs 0.02(-3 l i
3.67 1.0.05 2) 1.47 0.07(-3 2.5 0.1(1) l 139Cs 1.3~10.2(- >8.7(1)
<1.5(-4) 24Na 5.0 0.1(-3) 3.1 0.3 -
- 1.6 0.2 i
s1Cr 2.73 0.09(-4 1 ~. 2 - 0.3 - 2.3 1 0.6 54Mn 5.9 0.1(-5)) 9.5 0.6 - 6.2 1 0.4 59Fe 1.70 0.09(-5) 7.7 1.6(-7 2.2 i 0.5 57C0 9.9 2.2 - <4.4(-7)* >2.3(0) 58Co 6.5 1 0.2 - 2.35 0.03(-5 2.8 i 0.1(1) ;
- 60CC 5.5 1 0.1 - 4.310.1(-6) ) 1.3 0.1(1) i ssZn 2.5 0.5 - <3.6 - * ~
>6.90) l 91S r. - 5.5 0.7 - * <3.8 - * >1.4 1) 93Y 4.2 0.3 - * - <1.2 - * >3.5 3) 93Zr 1.89 0.09 5) 2.5- 0.2(-6) 7.6 1 0.7 i
95Nb 1.8 1 0.1 - 3.1 0.3(-6) 5.8 0.6 99Mo 4.0 0.8 - 3.36 i 0.07(-7)* 1.2 0.2 i 10 3Ru 1.4 0.2 - * <7.4(-7)* >1.9(0). 11cmA9 <2.5(-4)* .
<2.3(-4)*
124Sb 7.1 i 0.5(-6) 2.1 3.4 1 0.3(0) 0.1(-6) 12sSb 1.6 0.6(-6)* 129mie
<7.8-7)* >2.1(0)
<2.2(-6) <3.7 -7)*
132Te 7.8 3.9 - * <6.8 -7)* >l.l(0) 139Ba 5.3 0.3 - 2.17 0.02(-3) 2.4 0.1(0) 18.o Ba -7.4 0.7 - -
<7.8(-7)* >9.5(1)-
140La -6.0- 0.1 ;-
- 2.5 1 0.3(2) 2.4 1 0.3(-7)*
141Ce <1.3 -6)* <3.6(-7)* 14sCe - <2.3 -6)* : <1.3(-6)* 14Ce <9.0 -7) <5.0(-7) 18 % 8.2 i 1.8(-4) <9.0(-6)* >9.l(1). 239Np <1.3(-4) 1.5 0.3(-6)* <8.7(1)
*: Resin Coltan Data' B-72
TABLE E.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 11:41; 2/10/78 Used for: 252 days Bed Volumes Thru: 9.2(4) Letdown Flow Rate: 65 gpm Reactor Coolant Baron: 448 ppm Inlet Activity Outlet Activity Decontamination Nuclide (pCi/ml) (pCi/ml) Factor __ 131I 6.6810.06-3) <1.5(-6) >4. 5(3) 1321 1.45 1 0.01 -2) <4. 2(-5) >3.5(2) 133I 1.25 1 0.02 -2) 4.8 1.1(-5) 2.6 0.6(2) 1 34I 1.n5 10.03 -2) <6.0 - >2.4 2 135I 1.35 1 0.02(-2) <1.2 - >1.1 3 88Rb 4.9 0.2(-2) 1.0 0.2(-2) 4.8 0.8 0) 89Rb 4.1 0.7(- ) 2.0 0.4 l) 134Cs 8.0 7.60 1 0.09(-4 0.2(-3) ) 7.86 0.05 - 9.7 0.1 -1) 13sCs 1.20 0.02 - 2.2 1 0.3 0) 137Cs 2.6 1 0.01(-3 1.51 0.4(-5) ) 1.53 1 0.01 - 9.87 i 0.09(-1) 138Cs 3.65 0.08(-2) 2.3 0.3(-3) 1.6 0.2(1) 139Cs 1.4 1 0.3(-2) <4.4(-4) >3.1(1) 24Na 4.81 1 0.08(-3) >2.8(3) 51Cr 6.3 1.0 -5) <1. 7(-6}
<1.0(-6, >6.3(1 )
54Mn 4.9 0.2 -5) 5.810.4(-6) 8.5 0.6(0) 59Fe 8.5 i 1.1 -i) <4.4(-7) 57Co <2,0(-5) <9.4(-7) 58C0 2.93 0.05(-4) 2.610.2(-6) 1.11 0.07(2) soCo 3.0 0.1(-5) 4.5 1.3 6.6 i 1.9(1) G sZn <2.3(-6) <4.7 -7) (-7) 91Sr 1.8 0.4 - <6.1 - 6) >2.9 'l 93Y 3.7 2.2 - <1.4 -5) >2.6 l 95Zr 5.4 1 0.9 - <7.2 - >7.5 ssNb 6.2 1 0.8 - <6.8 - >9.1 99Mo 7.410.3-5) <3.6 - >2.1 103Ru <4.4(-6) <1.2 - 110 mag <2.9(-4) <2.0 - 124Sb 1.7 1 0.4 5.6 0 8(-7) 3.1 0.9(0) 12sSb 6.112.7j-6) .-6) <1.8 - >3.4(0) 129mie <2.6(-6) <7.6 - 132Te <1.1(-5) <2.2 - 13988 6.1 1 0.1(-3) 3.40 0.04(-3) 1.80 0.04(0) 140Ba 6.6 0.5 >4.7(1) l_ 140La <1.1 -5) (-5) <1.4
<5.0 - - )>
141Ce <2.8 - <1.2 'l l 143Ce <1.0 - <4. 3 - l
<2.2 - <1.2 -
l 144Ce ' 187W 5.510.7(-4) <1.6 - >3.5(1) 239Np <6.0(-6) <5.0 - l B-73 ' c
i TABLE B._13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 16:03; 2/10/78 Used for: 252 days Bed Volumes Thru: 9.2(4) Letdown Flow Rate: 65 gpm Reactor Coolant Boron: 448 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uCi/ml) (uci/ml) Factor 1311 6.31 0.06 -3) <2.8 -6) >2.3(3) ! 132I 1.32 0.01 -2) <6.8 - >1.9(2) ~ , 133I 1.22 1 0.02 - <6.2 - >2.0 134I 1.25 0.04 - <8.8 - >1. 4 las! 1.26 i 0.01 - <1.2 - >1.1 88Rb 4.9 0.3(-) 1.2 1 0.2(-2) 4.1 0.8(0) 89Rb <2.1(-4) >4.0(1) 134Cs 8.4 7.29 t2.0.3(- 0.07 -4) ) 7.72 0.08 - 9.4 0.1(-1) 136Cs 2.7 1.30 0.04 - 2.1 0.3(0) 137Cs 1.47 0.4(-)) 0.01 -3 1.49 i 0.02 - 9.9 0.1(-1) 13eCs 3.94 0.08 -2) 2.5 0.3(-3) 1.6 0.2(1) 139Cs <8.8(-4) <6.2(-4) 24Na 4.61 0.07( 3) 1.410.4(-6) 3.3 1 1.0(3) 51Cr 4.7 1.2 - >2.8(1) 54Mn 5.2 0.1 - <1.7(-6) 1.2 0.1 (-6) 4.3 0.5(1) 59Fe 1.5 0.6 - <6.8(-7) >1.6(0) 57C0 1.5 1 0.6 - <9.4(-7) >1.6(0) seco 3.37 0.05(-4 2.89 0.05(-5) 1.17 0.03(1) 60Co 3.4 0.1(-5)) 2.5 0.1(-6) 1.33 i 0.09(1) ssZn <2.6(-6) <6.1(-7) 91Sr 1.18 0.08(-4) <5.6(-6) >2.1(1) 93Y 2.4 0.6 - >1.3(1) 9sZr 8.4 1.0 - <1.8(-5)(-6) 2.1 0.6 4.0i1.1(0) 95Nb 7.5 i 1.1 - 2.1 0.2(-6) 3.5 i 0.6(0) 99Mo 6.6 i 0.2 - <6.5 - >1.0(1) lo3Ru <1.9 - 110 mag .<4.4(-6)(-4) 2.9 0.9 <2.4 - >1.2(0) 124Sb 2.8 0.7(-6) 8.3 1 3(-7) 3.4 1.0(0) 12 sSb <3.4 - <1.7 - 129mie <2.7 - <l.2 - 132Te <8.8 - <9.5 - 139Ba 5.7 0.1(-4) 4.07 0.08(-3) 1.40 0.04(-1) 140Ba 5.610.5(-5) <2.6 - >2.1(1) 14cLa 1.3 0.4(-5) <1.0 - >1.3(1) 141Ce <3.8 - <1.5 - 143Ce <1.1 - <6.4 - 144Ce <2.3 - <9.9 - 187W 4.7 0.3(-4) <1.9 - >2.5(1) 239Np . <5.6(-6) <6.0 - B-74
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Oemineralizer A 13:25; ?/14/78 Used for: 256 Jod Bed Volumes Thru: 9.4(4) Letdown Flow Rate: 62 gom Reactor Coolant Boron: 4?8 ppm Inlet Activ ty Outlet Activity Decontamination Nuclide (uCi/ml) (uCi/ml) Factor I 8.11 1 0.08 - 1.4 1 g(_7)* 5.8 4.1(4) [I 1.49 1 0.02 - <3.2 - * >4.7 233I 1.34 0.02 - >4.5
<3.0 -
- 134l 1.64 1 0.06 - >3.6
<4.5 -
- 13sI 1.35 0.01 - <g.7 - * >l .4 esRb 4.9 0.3 - 1.2t0.1(-2) 4.1 0.4 0 89Rb 8.7 0.3 - 3.9 i 0.8(-4) 2.2 0.5 1 134Cs 8.0 i 0.1 - 7.69 1 0.06(-4 1.04 1 0.0 0) 136Cs 2.310.2- 2.1 _0.2 )
1.09 0.02(-5 137Cs 1.51 1 0.02 -3 1.54 0.01(-3 9.8 0.1 ) 13eCs 139Cs 3.6i0.1(-)) 2.7 0.2(-3) 1.3 0.1 8.0 1 2.1(- ) <6.8(-4) >1.2(1) 24Na 4.73 1.9 0.3(3) 0.05(-3) 2.5 0.4(-6) 51Cr <5.9 i (-6 2.7 <2.2 0 0.4(-6)* 54Mn 59Fe 4.9 2 0.1(-5)) <8.4(-7) >5.8 1 1.05 0.09(-5) 1.3 1 0.3(-7)* 8.1 2.0(1) 57Co <3.2 i (-6) <6.8(-8)* seco 2.4320*07(1) 3.52 0.04(-) 1.45 i 0.04( 5) soCo 3.21 1 0.07( )- 1.210.2(-6 2.7 0.4(j) 652n- <1.7 (- 4.2 <4.0 1) 91Sr 3.2(-8 1.2 0.2(-4 <2.4(-7)* >5.02) 9 3Y <6.4 t (-5) <3.1(-7)* 95Zr 6.3 1.0-6) 4.2 0.3 -
- 1.5 0.3(y 95Nb 6.2 0.7 -6) 5.6 2 1.2(0) 9No 1.1 i 0.2 _
7.7 0.3-5) 5.2 4.2 -
- 1.5 1.2(3) 103Ru
<1.9(-7)*
110 mag 124Sb
.<5.1(-6)(-4) 2.8 0.3 <1.7(-4) >1.6(0) 3.3 2.5 0.4(-6) 7.5i1.9(-7) 1.0(0) 125Sb <4.4 - 3.1 1 2.0(-7)* *I 4(I) 129mTe <2. 7 -
132Te
<1.2(-7)*
<1.0 - 2.310.9(-7)* <4.3(1) 139Ba 4.4 1.0010.05(0) 140Ba 0.2(-3) 4.38 0.06(-3) 5.5 1 0.4(-5) <1.6(-7)* >3.4(2) 140La 1.17 6.9 i 2.4(1) 141Ce <3.8 -
0.05(-5) 1.7i0.6(-7)* <3.8(1) 143Ce <1.1 - 9.9 i 4.l(-8)*
<2.1(-7)*
144Ce <4.4 - 9.013.4(-8)* 18 4 6.0 1 0.7(-4) <6.9(-6) #49fl
>8.7 1 239NP <5.5 (-6)
<1.4(-7)*
{ t B-75 l
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 17:55; 4/12/78 Used for: 299 days Bed Volumes Thru: 1.10(5) Letdown Flow Rate: 65 gpm Reactor Coolant Boron: 337 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uti/ml) (pC1/ml) Factor 1311 8.6 0.2(-2) 9.3 0.2(-2) 9.2 0.3(-1) 1321 1,97 i o,04(_2) 2.33 0.07(-2) 8.5 0. 3 -1 ) l 133I 4,4 5.0 0.l(-2) 8.8 0.3-1) ' 134I 1.721s0,1(.2) 0.05(-2 ) 2.11 i 0.05(-2) 8.2 0.3 -1) 1 35I 2.59 0.03(-2) 2.72 i 0.05(-2) 9.5 0.2(-1) ) i esRb 4.3 0.1(-2) 6.5 0.8(-) 6.6 0.8(-1) ! 89Rb 9.5 6.6 1.4 0.3(0) 134Cs 1,66 0.3(-3) 0.03(-3 ) 1.65 1.3(- )) 0.05 -3 1.01 0.04() 136Cs 2.2 2.6 8.5 1.0 - < 137Cs 2.83 0.2(- )) 0.04 -3 2.94 0.2(-4) 0.05(-3 ) 9.6 0.2 - l 13sCs 4.06 0.07-2) 4.3 0.2 9.4 0. 5 - 139Cs 1.4 0.4(-2) <1.8(-2) (-2) >7.8(-1) 24Na 2.3? i 0.03(-2) 2.34 0.04(-2) 1.00 1 0.02(0) 51Cr <5.L(-5 <9.6 - s4Mn <2.0 - <2.3 - 59Fe <4.5 - <4.6 - 57Co <1,9 _ <6,4 seCo 4.6 0.2(-4) 6.4 0.2(-4) 7.2 0.4(-1) 60Co 4.0 2 0.6(-5) 6.5 1.0(-5) 6.2 i 1.3(-1) 65Zn <1.5(-5) <4.1 -5) 91Sr <6.4(-5) <1.2 -4) 93Y 2.0 0.4(-3) <1.0 - >2.0(1) ssZr <2.8 - 95Nb <1.1
<6.4(-5(-5'h <9.4 -
99Mo 9.6 i 1 9(-5) 1.1 01(-4) 8.7 1.9(-1) 10 3Ru <2.4 - <9.6(- 110 mag <7.0 - <1. 0 f,- 124Sb <1.0 - <3.3 - 12sSb <1.4 - <5.8 -5) 129mTe <1.2 - <2.2 -4) 132Te <2.7 - 139Ba 3.1 0.4(-3) <2.403 4.1 -4) (-3) 7.6t1.1(-1) 140Ba <6.0(-5) <2.1(-4 140La -2.9 0 7(-5) <3.5(-5 >8.3(-1) 141Ce <2.2 - <8.1(-5 143Ce <2.4 - <1.2(-4 144Ce <1.1 - 187W 1.9210.06(-3) <1.0(-4) 2.5 1 0.1 (-3) 7.7 i 0.4(-1) 239Np <2.6(-5) <4.9(-5) B-76
TABLE B.13 (cont'd) UNIT #4 CYCS PURIFICATION DEMINERALIZER Demineralizer A 11:33; 4/13/78 Used for: 300 days Bed Volumes Thru: 1.11(5) Letdown Flow Rate: 65 gpm l Reactor Coolant Boron: 336 ppm Inlet Activity . Outlet Activity Decontamination Nuclide (uCi/ml) (pCi/ml) Factor 131I 1.01 1 0.02(-1) 1.03 0.03(-1) 9.8 1 0.3(-1 132I 2.08 0.02(-2) 2.15 i 0.06(-2) 9.7 0.3(-1 133I 4.7 1 0.1(-2) 5.0 0.1(-2) 9.4 0.3(-1 1 34I 1.94 0.03(-2 2.07 0.06(-2) 9.4 0.3(-1J 13sl 2.61 i 0.02(-2 2.72 0.03(-2) 9.6 2 0.1(-1) ! esRb 1.00 0.03(-1) 6.1 0.6(-2) 1.6 i 0.2(0) 89Rb 9.3 2.4(-3) 8.6 1.0(-3) 1.1 1 0.3(0) 134Cs 1.67 i 0.02(-3) 1.67 0.03(-3) 1.00 1 0.02(0) 136Cs 2.2 0.2(-4) 2.4 0.2(-4) 9.2 1.1(-1) 137Cs 3.06 1 0.03(-3) 2.99 i 0.05(-3) 1.02 i 0.02(0) 138Cs 6.4 0.1(-2) 4.3 0.1(-2) 1.5 0.1(0) 139Cs <5.8(-2) <1.0(-2) 24Na 2.36 1 0.02(-2) 2.38 0.04(-2) 9.9 0.2(-1) 51Cr <6.1 -5) 54Mn <1.3-4) <5.0
<7.5 - - )l 59Fe <3.0 -5) <2.4 ,,1 57Co 1.2 (.4 - <1.2 - ) >1.0(0) seCo 5.7 0.1 - 5.4 i 0.2(-4) 1.1 0.1(0) 60C0 5.4 0.6 - 4.410.6(-5) 1.2 0.2(0) 65Zn <2.0(-5) <1.3(-5) 91Sr <9.4 (-5) <7.7(-5) 93Y 1.2 0.3 2.8 0.7(-3) 4.3 1.5 95Zr <1.2(-5) (-3) 4.4 0.9(-5) <2.7(-1) (-1) 9sNb <5.8(-5) <5.6(-5) 99Mo 1.10 0.07(-4) 9.4 1.4(-5) 1.2 0.2(0) lo 3Ru <1.4 - 1.5 0.6(-5) <9.3(-1) 110 mag <8.4 - <6.8-}
124Sb <1.2 - <9.0 - 12sSb <1.3 - <2.0 - 129mTe <1.4 - s1.2 - 132Te <1.8 - <1.8 - 139Ba 5.8 .1(-3) 5.0 0.2(-3) 1.2 1 0.1(0)
; 140Ba <3.7 -
! 140La <1.9 - <4.7(-5) 2.7 07 (-5) <7.0(-1) 141Ce <2.3 - <2.7(-5 143Ce <2.6 - <3.1 (- 5 l 144Ce <1.1 - <8.9(-5 2.0 0.1 (-3) 1.0 i 0.1(0) 187W 2.010.1(-3) 239Np <1.6(-5). <2.0(-5) B-77
.y'.
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 13:53; 4/14/78 Used for: 301 days Bed Volumes Thru: 1.11(5) Letdown Flow Rate: 65 gpm Reactor Coolant Boron: 320 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uCi/ml) (uC1/ml) Factor 131I 6.91 0.02 - 6.8 1.5(-6) 1.0 0.2(4) 1 32I 1.85 0.02 - <9. -1 >1.9 133I 3.23 0.03 - <9.7-) >3.3 1341 1.87 0.02 - <8.2 >2.3 lasI 1.98 0.02 - <4.4 - ')I >4.5 esRb 1.32 0.02(-1) 3.9 1.l(-4) 3.4 1 1.0 2 89Rb 9.1 0.5(- ) 5.6 1.3(-4) 1.6 0.4 1 13'+C s 1.63 1 0.02 - 1.98 i 0.03(-3) 8.2 0.2 - ) 136Cs 1.29 1 0.03 - 1.7 1 0.4 1 137Cs 3.1510.03- 7.81 3.50 1 1.8(-6) 0.06(-3 ) 9.0 0.2 - ) 138Cs 5.82 0.07(-2 1.00 0.01(-2) 5.82 0.09 0) 139Cs <1.3(-4) <4.5(-3) 24Na 1.56 0.01(-2) 2.2 1 1.0(-6) 7.1 3.2(3) 51Cr 2.3 1.1-4) 2.4 1.2(0) 54Mn 5.1 0.1 -5) 9.4 2.60 1 0.07(-5 0.8(-5) ) 1.96 0.07(0) 59Fe 1.7 0.1 -5) 1.1 1 0.1 - 1.5 1 0.2(0) 57Co <3.6(-6) 2.0 0.7 - <1.8(0) ssCo 1.0310.01(-3) 9.4 0.3 - 1.10 0.04(0) 60Co 6.110.1(-5) 4.6 0.1 - 1.33 i 0.04(0) 65Zn 2.8 1.0(-6) <1.5 - >1.9(0) 91Sr <2.2(-4) <2. 3 - 93Y <9.0(-4) <2.4 - esZr 2.0 0.1(-5 1.8 0.1(-5) 1.11 0.08(0) , 95Nb 2.0 0.1(-5 2.3 0.1 8.7 i 0.6(-1) 99Mo 1.13 0.07(4) <5.6(-6)(-5) >2.0(1) 103Ru .7.9 1.7 -6) >3.0(0) 110 mag 4.410.9-6) <2.6(-6)(-6) 2.3 0.7 1.9 0.7(0) 124Sb 2.0 0.1 -5) 2.1 1 0.1(-5) 9.5 0.7(-1) 125Sb ~< 6.0 - <1.1 -
<7.2 -
129mre <4. 3 -
- 132Te <l.1 - <4.6 -
1398a- 5.5 1 0.3 - 4.8'i0.2(-3) 1.1 1 0.1(0) 14cBa 3.8 0.6 - <1.1 - >3.5(0) 140La 1.4 0.2 - <1.1 - >1.3(2) 141Ce <8.0 - <2.7 - l 143Ce- '<3.2 - <9. 2 - I 144Ce <5.2 - <1.6 - 187W 1.61 0.04(-3) <9.5 - >l.7(l) -l 239Np <4.9(-5) <2.6 - 1 B-78 i i __m - . - , -
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 09:57; 4/15/78 Used for: 302 days Bed Volumes Thru: 1.11(5) Letdown Flow Rate: 60 gpm Reactor Coolant Boron: 323 ppm Inlet Activ ty Outlet Activity Decontamination Nuclide (uCf/ml) (pCi/ml) Factor 131I 1.70 0.01 - 6.7 2.0(-6) 2.5 0.8(3) 1321 1.7410.03- <5.3 - >3.3 1331 1.51 1 0.01 - <1.6 - >9.4 134I 1.94 i 0.03 - >2.6 lasI 1.54 0.02 - <7.5-)
<1.8 - >8.6 88Rb 1.66 0.02 - 5.01 0.07(-2) 3.31 89Rb 1.15 i 0.04 - 4.2 1.0(-4) 2.7 0.71()0.06 0) 134Cs 1.77 1 0.02 1.92 0.01 - 9.2 i 0.1 -1) 136Cs 4.1 1 0.1(-5 1,42 0.07 - 2.9 0.2 0) 137Cs 3.35 0.02( 3 3.54 0.04 - 9.5 0.1 -1) 4 138Cs 5.9 1.5 0.1(-2)) 8.8 0.1(-3) 6.7 0.1 0) 139Cs 0.6(-2) <4.3(-3) >3.5(0) 24Na 51Cr 9.310.1-3) 5.4 0.6-5) 1.7 0.2(2 1.8 1 0.4 -4) 2.8 0.8 -5) 6.4 i 2.3(0 54Mn 4.310.1-5) 5.9 0.3 -6) 7.3i0.4(0 59Fe 1.2 0.1 -5) 2.8 0.6 -6) 4.3 1.0(0 57Co <2.0(-6) <1.1 (- 6) 58C0 5.61 1 0.02(-4) 2.28 3.78 0.02(-4) 2.46 0.02(0) 60Co 0.06(-5) 1.22 0.05(-5) 3.1 0.1(0) 65Zn <1.5 - <1.2 -6) 913p <2.4 - <1,1 93y - 4)
<1.0 - <6. 3 -4) 95Zr 1.21 0.07 - 4.9 0.8(-6) 2.5i0.4(0) 95Nb 1.34 0.06 - 6.4 0.6(-6) 2.1 0.2(0) 99Mo 1.13 0.04 - <5.5 - >2.1 103Ru 4.4i1.1(-6) <3.2 - >l .4 110 mag 1.3 0.6(-6) <9.4 - >1.4 ,
124Sb 1.0710.06(-5) 9.7 08(-6) 1.1 0.1(0) 12sSb <4.2(-6) <6.3 - 129mTe <3.3(-6) <3.5 - 132Te 8.7 1 2.5 - <5.0 - >l.7(0) 139Ba 5.3 0.3 - 4.9 0.1(-3) 1.1 1 0.1(0) 140Ba 3.3 0.5 - <1.1(-5) >3.0(0) 140La 7.9 1 0.4 - <7.4(-6) >l.l(l) l 141Ce <5.8 - <2.3-6) 143Ce <2. 3 - <1.3 -51 144Ce <3.0 - <8.3 - 187W 1.27 1 0.04(-3) <1.5 - >8.5(0) 239Np 5.9 3.6(-5) <2.0 - >3.0(0) B-79
l l TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Domineralizer A ! 14:04; 4/27/78 l Used for: 313 days l Bed Volumes Thru: 1.15(5) Letdown Flow Rate: 45 gpm Reactor Coolant Boron: 288 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uC1/ml) (uC1/ml) Factor 131I 1.27 0.01 - 1 3.3 1 0.1 3.8 i 0.1(2) 1321 2.01 0.04 h <1.9(-5)(-5) >1.1(3) 133I 1.9
. O.02 - 4 2.5 i 1.1(-5) 7.8 i 3.4(2) 134I 2.09 <4.8(-5) >4.4(2)
[ i 135I 1.93 0.06-h) 0.05 - <6.6(-5) >2.9(2) serb 7.4i0.2(-1) 1.39 i 0.02(-1) 5.3 i 0.2(0) ! 89Rb <2.7(-4) >4.4(1) i 134Cs 1.2 1.22 1 0.3(-2) ) i 0.01(-3 1 .24 0.01(-3 9.8 0.1(-1) i 136Cs 2.4 5.4 i 0.4(-6) ) 4.4 0.6(OD l 137Cs 2.30 0.3(-5) 0.03(-3 ) 2.24 i 0.03(-3) 1.03 i 0.02CO) i 13eCs 7.9 0.1 5.6 i 0.1(-3) 1.41 0.03(1) 139Cs <4.7(-1)(-2 <1.6(-2) 24Na 9.4i0.2(- 1.9 i 0.4(-5) 4.9 1.0(2) l 51Cr 1.3 0.4( <1.1 (- 5) >1.2(2) 54Mn 4.59 i 0.07 5) 4.4t0.2(-6) 1.04 i 0.05(1) 59Fe 2.0 0.1 - 1.8 . 0.4(-6) 1.1 0.2 57Co 8.9 i 3.8 - <1 *( 8) >8.1(-1) (1) seCo 6.3 2 0.1 - 9213.l(-5) 6.8 0.1(0) , 60Co 4.6 0.1 , , 1.17 0.06(-5) 3.9 0.2(0) j 65Zn <6.8 - <9.6 - 91Sr <4.0 - <3.6 - 93y <4.0 - <1.4 - 95Zr 7.6 i 0.7(-6) 2.9 1.1(-6) 2.6 1.0(0) 35Nb 7.7 0.7(-6) 1.610.3(-6) 4.8 1.0(0) 99Mo 1.14 i 0.03(-4 <l.1 - >1.0(2) 103Ru 2.8i9.7(-6)) <1.2 - >2.3(0) 110 mag <1.6(-4) <l.1 - l 124Sb 7.9 i 0.6(-6) 4.3104(-6) 1.8 i 0.2(0) 125Sb <2.3(-7) <1.6 - 129mTe 1.711.1(-6) <2.1 - >8.1(0) 132Te <3.3(-6) <3.8 - 139Ba 1.07i0.04(-2)- 2.5 1.6(-3) 4.3t2.7(0) l 140Ba 4.7 i 0.5(-5) 9.5 2.8(-6) 5.0 i 1.6(0) 140La <5.7(-6) 141Ce <2.6 - <9.9(-7) 1.7 1.5 (-6) <1.5(-1) 143Ce <9.5 - <l.9-5) 144Ce <1.4 - <5.1 - 187W <6.6 - <4.2 - 233Np <5.0 - <5.0(- B-80
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 09:53; 4/29/78
' Used for: 315 days Bed Volumes Thru: 1.16(5)
Letdown Flow Rate: 45 gpm Reac'mr Coolant Boron: 288 ppm Inlet Activity Outict Activity Decontamination _Nuclide (uct/ml) (uCi/ml) _ Factor 131I 1.06 1 0.01 - 1.7 0 5(-6) 6.211.8(3) 1321 1.75 0.02 - <6.5 - >2.7 133I 1,76 i 0.02 - <1.6 - > 1.1 134I 1.89 1 0.05 - <1.0 - >1.9 1351 1.81 0.05 - <1.e(-4 >1.0 esRb 1.9720.04(-1) 1.67 0.03(-1) 1.18 0.03(0) 89Rb
<5.6(-4) >1.7(1) 134Cs 9.5 1.13 i10.8(-3) 0.01(-3 ) 1.22 0.01(-3 9.3 1 0.1 -i) 13sCs 5.4 0.2(-6) ) 4.8 0.6 0) 137Cs 2.6 ii 0.3(-5) 2.26 0.03(-3 ) 2.31
- 0.0 3(-3) 9.8 0.2 -1) 13eCs 6.1 0.1(-2) 1.18 0.04(-2) 5.2 i 0.2 0) 139Cs <2,9(_2) <j,g(_2) 24Na 8.ti 0.2 - 3.2 0.6 2.7 0.5(2) 52Cr 8.4 i 3.3 - <6.2(-7) (-5) >1.4(2) 54Mn 4.2 0.2 - 8.9 1.1(-7) 4.7 0.6(1) 59Fe 1.32 1 0.08 -5) 6.4 1.5(-7) 2.2 0.5(1) 57Co <1.3p6) <3.2(-8) seCo 4.2310.07(-4) 2.22 0.03(-5) 1.91 0.04(1) 60Co 4.65 0.09(-5) 3.50 i 0.02(-6) 1.33 0.03(1) 652n <1.5 - <4.2(-8) 91Sr <1.3 - 5.2 4.8 <2.5(0) 93Y 95Zr
<1.5 -
6.3
<4.9(-5)(-5) 0.7(-6) 1.1 0.2(-6) 5.7 i 1.2(0)
C5Nb 5.5 0.4(-6) 5.4 2.0(-7) 1.0 1 0.4(1) 99Mo 1.12 0.03(-4) <3.8 - >2.9(2) 103Ru <6.9(-7) <7.0 - 110 mag <1.5(-4) <1.7 - 6.9 i 0.5( ',) 124Sb 2.3102(-6) 3.0 1 0.3(0) 12sSb <1.7 - 129 Dire <4.6(-7) 1.6 0.9 (-6) <1.5 - >1.1(1) 132Te <3.5(-7) <4.5 - 139Ba 8.0 t 0.4(-3 5.9 0.1 (-3) 1.4 0.1(0) 140Ba 6.0 i 0.3(-5 1.7 1.0(-5) 3.5 0.6(0) 140La 6.4 1,2(-5 <8.1(-7) >7.9(1) 141Ce <1.6(-7) 1.0 i 0.7(-6) <1.6(-1) 143Ce 4.9 i 4.3(-5) <4.1(-6) >1.2(1) 144Ce <9.2(-7) <2.3 - 187W 6.1 i 2.0(-4) <5.4 - >1.l(2) 239Np <1.8(-6) <5.0 - 8-81
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 09:27; 5/9/78 Used for: 325 days Bed Voltsnes Thru: 1.19(5) Letdown Flow Rate: 55 gpm Reactor Coolant Boron: 250 ppm
-Inlet Activity Outlet Activity Decontamination Nuclide (uC1/mli (uCi/ml) Factor 131I 7.7 <1.5(-6) >5.1(3)-
132I 2.08 0.1(-3) 0.02(- 1 133I 1.73 i 0.03l l 2.2 0.8(-5) 7.9 i 2.9(2) 134I >2.5(2) l 13s! 2.39 2.01 1 0.03d,- i 0.05Il- l')I <9.4(-5)
<3.4(-5) >5.9(2) l ssRb 3.34 0.06(1) 3.41 i 0.02(-1) 9.8i0.2(-1) t 89Rb 1.5 1 0.1(- <6.8(-4) >2.2(1) l 134Cs 1.19 0.02 ? 1.26 0.01(-3) 9.4 0.2(-)
136Cs 1.62 i 0.06 3.7 0.3(-6) 4.4 1 0.4(0 l 137Cs 2.39 i 0.03 - 2.3910.02(-3) 1.00 1 0.02 0) 138Cs 7.65 0.08 - 1.20 1 0.04(-2) 6.4 t 0.2(0) 13sCs <9.8(-3) <6.4(-3) , l 24Na 8 . 9 i 0 .1 (,- 2.9 0.2(-5) 3.1 1 0.2(2) l 51Cr 7.2 t 0.51,- >2.8(1) l 5Mn 4.0 0.11 -
<2.6(-6)(-6) 1.7 1 0.2 2.4 1 0.3(1) ssFe 1.210.l(- 1.3
- 0.4(-6) 9.2 1 2.9(0) 5700 <1.3(-6) 5800 3.14 0.0U-4) <7.S(-7) 8.0 S2 (-5) 3.9 7.6 0.1(0) 1 60Co 2.89 1 0.08(-5) 3.8 i 0.i f-6)'
0.3(0) ssZn <7.0 f 'l 913p <1.6(j cl ,7 . 43,4 j',( 93Y <4.9I,- <2.01l- )p 95Zr 3.6 i 0.7(-6) 5.5 1.2(-1) 9sNb 4.9 i 0,4(-6) ) 6.5't0.5(-6)) 7.51 0.4(-6 6.5 0.6(-1) 9 8i40 1.19i0.01(-4 <1.5-6) >7.9(1) 10 3Ru <1.3(-6) <1.2-6) 11cp%9 <4.2 -7) 124Sb <1.2(-6) 5.1 1 0.5 (-6) 3.3 0 2(-6) 1.5i0.2(0) 12sSb <3.5(-6) <3.7 - l 129mTe <1.5 - 132Te <1.8 - >1.3(0) l
- 139Ba 140Ba
<3.4(-6) (I2.4i1.1-1ll 9.5 5.9 0.2 ,-
0.3I - I 4.79 i 0.07(-3)
<4.2(-
2.0 0.1(0)
>l.4(1) 140La 1.4 0.3I[ ') <4.01 - >3.5(2) l 141Ce <1.8f -
143Ce <1.3 - <9.0/-
<1.8 - 1l) i }
l 144Ce <3.0 - <1.l f-
- j. 187W 1.1 0.2(-3) <2. 0 l!- >5.5(1) 239NP 9.8 i 4.4(-6) 5.9 4.1(-5) 1.7 1.4(-1)
B-82
TABLE B.13 (cont'd) I UNIT #4 CVCS PURIFICATION 'DEMINERALIZER Demineralizer A 10:05; 5/16/78 Used for: 332 days Bed Voltaes Thru: 1.21(5) Letdown Flow Rate: 55 gpm Reactor Coolant Boron: 233 ppm Inlet Activity Outlet Activity Decontamination Nuclide (uC1/ml) Factor (uCf/ml) 131I 7.50 0.04 - ) <2.2f- >3.4 , 1 32I 2.10 i 0.03 - I <2.8/- >7.5 183I 1.7410.01-134I I <3.2/- >5.4 2.33 1 0.02 - ) c5.4f- >4.3 135I 2.02 0.03 - ) <2.8(- >7.2 serb 89Rb
<9. 4 (-3) 134Cs 1.17 136Cs 0.02( 3) 1.13 0.01( 3) 1.04 i 0.02(0) 2.0 0.2(-5 3.7 0.3(-6 5.4 0.7(0) 13Es 2.30 0.03( 3) 138Cs 7.3 0.1(-2 2.1110.02(3) 1.09 0.02(0) 139Cs <1.2(-2) >6.1(0)
<5.6(-1) 24Na siCr 8.4 1 0.1(-3) 2.4 0.1 3.5 0.2(2)
<7.5(-6) (-5) 54 %
.59Fe
.<4.1(-5) 4.0 0.1 (-5) 7.8 1.4 5.1 i 0.9(1) 8.6 1.2(-6) <7.0(-7)(-7) >1.2(1) 57Co 4
<l .2 (- 6) <4.4(-7) 5800 2.11 0.03(-4 1.16 0.02(-5) 1.82 0.04(1) 60Co 2.8 0.1(-5) ) 2.0 3(-6) esZn <1.9 - 1.4 1 0.2(1) 913p <6.lf-
,3,g . g) ,3 ),
93Y <7.9 - <7.6 - 95Zr 7.4 3.3(- <9.7 - 95Nb
>7.6(0 9.5 1.1(- <4.9 - >l.9(1 99Mo 1.28 103Ru 0.04 4) <2.0 - >6.4(1
<2.1(-6) <9.2 -
11erM9 <1.2(-6) <4.7 - 124Sb 3.7 1 0.5(-6) 9.9
- 1 0(-7) 3.7 0.6(0) 12sSb <8.5 - <3.7f-129mTe 132Te
<4.6-)! <1.6f-
<5.0 - )' <2.1d-139Ba 1.00 0.02 2) 1.2 8.3 0.7(-1) 140Ba 8.5 0.1(-2) 0.5(- -
<5.0 - >1.7(1) 14cLa- 3.8 0.1(- <6.2 - >6.1(2) 141Ce <2.5 - <2.1 -
143Ce <4.5 - <5.2 -6'l. 144Ce <l.4 - <5.1 'l 187W 239Np 1.3*0.2(-3) ~ <1.2 'I >l.l(2) i
<2.4(-5) . <9.8-)
B-83
, _. - ~ . .
TABLE B.13 (cont'd) UNIT #4 CVCS PURIFICATION DEMINERALIZER Demineralizer A 10:23; 5/23/78 Used for: 339 days Bed Volunes Thru: 1.23(5) Letdown Flow Rate: 47 gpm Reactor Coolant Boron: Inlet Activity Outlet Activity Decontamination Nuclide (uCi/ml) (uCi/ml) Factor 131I 6.49 1 0.04 - >1.0 3;l 132I 2.04 1 0.04 - <6.4-)i >4.1 2 l 133I 1.60 0.01 - <1.0 -
<5.0-l l >1.6 ll 134I 2.41 0.03 - <7.5 >3.2 135I 2.03 0.03 - <1.9 - ')I >1.1 )l serb 1.93 0.05 - 6.18 0.09(-2) 3.1210.09(0) 09Rb 1.31 1 0.05 - 2.5 5.2 1 1.7 1) 13'Cs 1.07 1 0.01 - 1.17 0.8(-4))
0.03(-3 9.1 0.2 -1) 13sCs 2.2 0.2(- ) 4.1 0.7(-6) 5.4 1.0 0) i 137Cs 2.1810.02-3) 2.23 0.01(-3) 9.8 i 0.1 -1) 13eCs 6.67 0.10 -2) 9.1 0.1(-3) 7.3 0.1(0) 139Cs <2.4(-2) <3.5(-3) 24Na 8.3 0.2 - 2.6 0.8 3.2 1.0(2) 51Cr 6.8 0.7 - <3.0(-6)(-5) >4.5(0) 1 59.n 4.6 2 0.1 - 5.0 0.6(-6) 9.2 1.1(0) 59Fe 8.7 0.7 - <1.8(-6) >4.8(0) 57Co 1.310.5-seCo 2.95 0.03-4) <1.6(-6) 5.9 *O2 (-5) >8.1(-1) 5.0 0.2 (0) 60C0 2.810.4(-5) >7.4(0) i 65Zn <2.0(-6) < 3. 8 f',-
<1.41 -
l 91Sr <1.0I 9 3Y <2. 2 (-4,)1
<l.4(-3 <3.91,-
l 9 5Zt- 5.7 1 0.9(-6) 8.5 1 0.2(-5) 6.7 0.1(-1) 1 95Nb 8.5 0.05(-6) 8.9 0.2(-5) 9.6 0.2(-2) 99Mo 1.24 1 0.02(-4) <7.3(-6) >1.7(1) lo 3Ru 3 1(-6) 1.34 0.08(-5) 2.2 0.8(-1) 1 11ongg <1.3(-6) l 124Sb <2,4(-6)(-6) 5.0 04 2.0 0.2(-6) 2.5i0.3(0) 12sSb <4.0(-6 <3.0 - l 129mTe <2.2(-6 <2.6 - 132Te <3.8(-6 <9.9 - 139Ba 1.0310.02(-2) 5.1 0.2(-3) 2.0 1 0.1(0) 140Ba >5.710.3(-5) 5 2(-6) 14cLa 141Ce <2.8 - 1 4.4 0.2(-5) <6.4(-2) 143Ce <3.2 144Ce <1.4 - )l <4.4(-5) 7.1 0.5 (-5) <7.5-2) 187W 7.7 2.1(-4) >7.0 0) 239Np 3.4 1.1(-5) <l
<3.6.l -5)('-4) >9.4 -1)
B-84
TABLE B.14 RADIONUCLIDE CONCENTRATIONS IN BASE CATION DEMINERALIZER C INLET
'2/16/78;.15:10 2/17/78; 13:15 2/20/78; 18:15 2/22/78; 16:05 5/2/78; 09:05 5/3/78; 09:52 Nuclide (uci/m1) (uci/m1) (uci/ml) (uCi/ml) (uci/m1) (uci/mi) 131I 3.47
~
0.07(-4) 5.93i0.'10(-4) 1.0010.02(-2) 8.7i0.1(-3) 1.12 0.01(-4) 1.04 0.02(-4) 136Cs 9.83 0.09(-4) 1.5810.01(-4) 2.02 i 0.01(-3) 2.16 135Cs .~ 0.02(-3) 3.5 i 0.1(-5)- 3.14 1 0.06(-5) 6.0 1 0.3(-4) '6.0 2 0.3(-4) 137Cg 2,05g.0.02(-3) 2.98 0.04(-4). 3.19 0.03(-3) 3.24 i 0.04(-3) 4.64 i 0.09(-6) 4.2 0.1(-5) 51Cr .'4.9 i 0.3 - 1.93 i 0.07(-6 8.2 1 0.5(-4) 7.9 i 0.4(-4) 2.1 0.4(-5) 56Mn. 6.4 'O.1 - 8.2 1 0.1 (-5) )2.49 i 0.03(-3) 2.57 0.04(-3) 1.50 1 0.09(-5) 7 1 2(-6)(-6) 5.8 0.3 59Fe 8.8 i 1.0 - 4.30 0.20(-5) 2.5i0.1(-4) 2.44 0.03(-4) 2.2 0.8(-6) 1.0 1 0.4(-6) 57C0 9.1 i 0.5 - 1.2 0.1(-5) 5.6 i 0.3(-5) 5.3 0.2(-5) 8.5 2.4(-7) 4 i 1(-7) ssCo 2.8710.04(-2) 3.39i0.07(-3) 4.27 1 0.07(-2 4.29 i 0.08(-2)- 1.83 i 0.0a(-4) 7.4 1 0.2(-5) p? soCo 1.35 0.01( 2) 1.70 i 0.02( 3) 6.08i0.09(-3) ) 6.3 i 0.1 - 1.17 1 0.04(-4) 7.3 1 0.1(-5) 55Zn 5.1 1 0.8(- 8.8 i 1.2 - 3.5 0.6 - 3.4 i 0.3 - <1.6(-6) <9.1(-7) 95Zr- 1.15.i 0.06 4)- 3.4 i 0.2 - 5.4 i 0.6 - 5.0 t.0.2 - 5.0 i 1.0( - 3.2 1 0.4 - 95Nb 2.02 i 0.06 -4) 6.1 0.2 - 8.5 1 0.5 - 7.5 i 0.3 - 9.7 i 0.05 ) 4.8 i 0.5 - 99Mo <9.6(-6) <1.5(-6)' 6.13 1 0.09(-4 3.79 i 0.02( 4 1.5 i 0.2( 1.'.i 0.2 - 103Ru~ 2.1 0.2(-6) 3.45 1 0.08( 5). 8.2 1 3.4(-6) )9.2i1.8(-6)) <7.5(-7) 9 2(-7) lo6RuD <1.5(-5) 3.7 i 1.2(-5 <1.0(-5) <9.4(-6) <7.1(-6) 7 2(-6) 11oung <5.8(-4) 2.6 i 0.7(-5 <5.5(-4) <6.2(-4) 4.0 0.8 - 2.4 i 0.3(-6 126Sb 6.2 i 0.1(-4) 4.17 i 0.06( 4) 1.89 i 0.05(-4) 1.89 0.03 - 9.3 1 0.8 - 8.1 1 0.5(-6 12sSb 1.0610.02(-3) 5.55 1 0.66(-4) 8.410.8(-5 1.14 0.03 - 1.5 1 0.2 - 1.44i0.07(5) 140Ba 2.9 i 0.1(-4 2.87 0.05 - 8.4 1.4 - 1.3410.1(-) ~, 14oLa 2.4 1 0.2(-5) 4.7 0.3(-6) 3.0610.03(4) 3.77i0.09(-4 2.6 0.3 - 4.9 i 0.3(-6) 141Ce 7.4 i 3.4(-6) 3.0 1 0.9(-6) <9.1(-6) <3.2(-6) <9.1(-7) <6.2(-7) 144Ce <7.7(-6) <3.4(-6) <2.9(-6) <2.5(-6) <4.9(-6) 211(-6) ;
TABLE B.15 RADIONUCLIDE CONCENTPATIONS IN BASE CATION DEMINERALI7ER A INLET 5/4/78; 14:36 5/9/78; 14:47 5/11/78; 11:55 5/15/78; 10:30 5/17/78; 14:45 Nuclide (uCi/ml) (uci/ml) (uCi/ml) (uci/ml) (pCi/ml) 131I 1.85 0.04(-4) 1.55 0.05(-5) 1.01 0.01(-4) 4.7 0.2(-5) 2.1 0.2(-6) 134Cs 2.07 1 0.02(-4) 2.97 0.07(-5) 7.5 0.2(-5) 4.6 0.l(-5) 2.26 0.04(-4) 13sCs 2.0 4.0 0.2(-6) 137Cs 3.59 0.5(-6)) 0.03(-4 4.2 1 0.l(-5) 1.35 0.05(-4) 5.6 0.3(-5) 3.11 0.03(-4) 3H
** ** 3.5 i 0.l(-2) ** **
14C ** ** 8.0 0.8-6) 32p 3.2 0.5 -5) SICr 3.4 0.6(-5) 2.0 i 0.4(-5) 3.0 1.5 0.2(-5) 1.0 0.3(-5) 54Mn 9.3 0.5(-6) 1.95 i 0.05(-5) 1.10 0.2 -5) ) 0.10(-5 1.13 1 0.05(-5) 6.8**i 0.3(-6)
** ** 4.03 **
m ssFe 0.02( 4) g 59Fe 3.7 1 0.7(-6) 2.9 0.6(-6) 3.6 0.5(-6 3.0 i 0.6(-6) 7.7 ; 3.2(-7) 57C0 <8.l(-7) 7.3 2.5(-7) 4.9 1.4(-7 1.0 0.2(-6) 2.5 1.0(-7) 5800 2.51 0.02(-4) 1.66 0.03(-4) 1.97 0.05(-4) 1.81 0.04(-4) A.27 0.08(-5) 60Co 5.0 1 0.2(-5) 2.03 0.05(-4) 8.6 0.2(-5) 1.27 0.02(-4) 8.32 1 0.09(-5) 63Ni ** 3.79 0.03(-5) 652n <8.9(-7) <l.1(-6) <9.3(-7) <1.3(-6) 6.1 3.0(-7) 89Sr ** 1.36 0.02(-6) **
** ** 8.8 **
90Sr 91y ** ** 3.71 0.4(-8) n.07(-7 ) ** ** 952r 6.3 0.7(-6 8.6 0.8(-6) 4 .1 0.4(-6) 5.5 0.6(-6) 4.3 0.3(-6) 95Nb 8.9 0.5(-6 1.63 0.07(-5) 8.1 0.4(-6) 9.1 0.5(-6) r6 0.3(-6) 99Mo 2.7 0.8(-6 <4.0(-7) 5.9 2.l(-7) <3.3(-7} <3.3(-7) 103Ru 1.710.5(-6) 1.810.3- 1.1 0.4(-6) 2.6 0.4(-6 8.04 2.0(-7) 106RuD <l .2(-6) 9.2 6.6 - <4.5(-6) 1.4 0.3(-5 <3.3{-6) 110 mag 3.4 0.7(-6 4.4 0.5 - 6.4 0.7(-6) 5.7 0.4(-6 1.8 0.2(-6) 124Sb 1.1 0.1(-5 5.7 1 0.5(-6) 7.8 0.6 - 1.2 0.1 - 1.5 0.2 - 12sSb 1.5 0.2(-5 1.06 0.08(-5) 1.1 1 0.2 - 3.3 0.2 - 4.6 0.6 - 140La 2.0 0.3(-6 8.4 i 1.7(-7 2.0 0.3 - 9.2 0.6 - 1.8 0.7 - 141Ce <9.1 -7 1.0 0.3 - <5.1 7 <5.8 - <3.6(-7) 144Ce <8.3 -7 4.6 i 1.7 - <2.2 -6 <3.2 - 2.1 0.8(-6)
** Radionuclide not measured.
TABLE B.16 RADIONUCLIDE CONCENTRATIONS IN BAE FEED (Base Cation Demineralizer Effluent) 2/16/78; 15:10 2/17/78; 13:15 2/20/78; 18:15 2/22/78; 16:15 5/2/78; 09:09 Nuclida (pCi/ml) (pCi/ml) (pC1/ml) (pCi/ml) (pCi/ml) 1311 3.24 1 0.04(-4) 5.96i0.09(-4) 5.2 i 0.1(-3) 6.0 0.4(-3) 2.92i0.04(-4) 134Cs <6.4 -7 ' <1.1 - cl.3 -6 <1.4 -7 137C5 <5.1 -5 <2.9 - 6 3(-7)
<6.0 -6 <7.2 -7 <1.5(-6) 51Cr 1.39 i ').07(-4) 5.0 0.2(-4) 2.37i0.14(-4) 3.0 0.3'-5) 54Mn 2.5 1.47 1 0.03(-5 0.1(-4) ) 6.4 1 0.3(-6) 2.03i0.06(-5) 4.38 0.16(-6) 3.9 i 0.3(-6) 59Fe 5.2 1 0.1(-5) 3.36 t 0.07(-5) 1.44 1 0.02(-4) 1.10 1 0.02(-4) 2.5 1 0.6(-6) 57Co. 3.6 0.2(-6) 2.8 1.0(-6) 3.6 0.7(-6) 3.9 1 2.3(-7) <3.7(-7) seCo 6.5 i 0.2(-4) '3.53 0.05(-4) 1.07 i 0.02(-3) 2.66 0.06(-4) 1.25 0.02(-4) io 60Co -2.28 1 0.05(-4) 1.18 i 0.01( 4) 3.0t0.04(-4) 4.83 i 0.05(-5) 4.76 1 0.08(-5) e3 55Zn 8 1(-6) 3.110.6(-5 7,8 i 1.2 - 9.1 1.9(-7) <9.0(-7) ssZr 4.43 i 0.06(-5) 2.17 0.05(5) 4.9 0.1 - 9.5 0.2(-6) 5.4 0.6(-6) 95Nb 7.5 i 0.3(-5) 4.0 0.15(-5) 8.4 1 0.2 - 1.51 1 0.05(-5) 1.40 0.08(-5 99Mo 6.711.5(-7) <7.8(- 7) -2.12 0.03(4) 1.96 1 0.05(-4) 2.6 0.3(-6))
lo3Ru 4.18i0.04(-5) 2.74 0.04( 5) 3.93 i 0.07(-5) 3.41 0.17(-6) 2.1 0.3(-6) 106RuD 5.7i0.4(-5) 3.8 i 0.3(-5 4.0 0.8(-5) <2(-7) 6 1 3(-6) lion %g 1.09i0.03(-5) 7.1 1 0.3(-6 1.53 0.06 - 6.74 0.18(-6) 4.6 0.3(-6) 124Sb 5.5 0.2(-4) 4.25 0.08(-4) 1.94 0.03 - 1.25 0.02(-4) 9.9 i 0.5(-6) ^ 12sSb 7. 39 i 0.08 ( 4) 5.65 i 0.07( 4) 1.65 0.02 - 8.2 1 0.3(-5) 1.54 0.07(-5) 140La 4.110.2- 1.9 1 0.2(-6 1.10 0.02 - 4.16t0.04(-5) 1.08 1 0.06(-5) 141Ce 3.9 0.5 - 1.9 i 0.3(-6 '
<6.8(-6) <7.0(-7) <6.8(-7)
, 144Ce 1.320.2- <2.8(-6) <1.3(-6) <2.A(-7) <2.9(-6) f 1 I i 1
TABLE B.16 (cont'd) RADIONUCLIDE CONCENTRATIONS IN BAE FEED (Base Cation Demineralizer Effluent) 5/3/78; 09:50 5/4/78; 14:34 5/9/78; 14:50 5/11/78; 11:58 5/15/78; 10:35 5/17/78; 14:46 Nuclide (uCi/ml) (pC1/ml) (pCi/ml) (pC1/ml) (pCi/ml) (uci/ml) 1311 1.36 1 0.05(-4) 2.8 0.2(-6) 1.47 0.09(-6) 6.9 0.1(-6) 2.57 0.07(-6) 7.0 1 3.0(-7) 1.810.4(-6) 1.1 0.1(-6) 8.8 1 0.1(-7) 9.3 0.9(-71 8.5 0.8(-7) 134Cs <2.3(-6) 136Cs 1.6 0.5( 71 137Cs <1.5(-5) 5.7 0.3(-6) 3.0 t 0.1(-6) 3.0 0.2(-6) 3.010.2(-6) 2.3 1 0.2(-6) 2.5 1 0.2 - 8.8 0.7 - 1.92 2 0.06(-5) 6.5 0.5 - 4.L '.6 - s1Cr 4.9 i 0.7(-4) 2.5 0.1 - 1.7 t ,.3-54Mn 1.1310.03(-4) 4.6 1 0.2 - 7.2 0.2 - 3.0 1 0.1(-6) 59Fe 3.010.2(-5) 2.5 1 0.4 - 9.3 i 1.8 - 1.6 1 0.1(-6) 1.9 0.2 - L.6 t 1.4 - 57Co 7.8 i 0.4(-6) 6.2i2.6(-7 4.0 1 0.6(-7 2.4 0.4 -7) 3.0 1 0.5 - 1.4 1 0.5 - m 2.5 1.0 1 0.1(-4 5.9 i 0.1(-5 5.9 0.2-5) 5.0 1 0.1 -5) 2.0 1 0.1 - seCo 0.06(-3) 3.710.1-5) 2.5 g 60Co 7.310.1(-4) 2.28 1 0.07( 5) 7.7 1 0.2(-5 2.5 0.1 -5) 0.3 - 4.4 - 4.5 1 3.8(-7) 3.7 1.0 -7) 4.6 1.4 -7) 5.5 2.6(-7) 65Zn 1.710.2(-5) 8.1 2.0 0.1 -6) 2.0 0.1 -6) 1.6 0.3(-6) 95Zr 6.9 0.1(-5) 3.1 0. 3 - 3.3 0.2(-6) ssNb 1.23 1 0.10(-4) 5.1 0.4 - 6.1 1 0.5(-6) 3.6 t 0.2 -6) 4.5 t 0.3 -6) 3.1 0.6(-6) 99Mo 3.4 1 0.2 -5) 4.9 1 2.5 - <1.f'-7) <3.3(-8) <1(-7) 6.5 0.8(-7 6.8 0.6(-7) 1.09 i 0.07(-6) 7.6 1.5 - lo3Ru 1.9 1 0.2 -5 7.311.7(-7) 3.5 1.0 - 106RuD 4.2 i 0.7 -5 <6.1(-7) 6.1 1.3(-6 2.6 0.6(-6) 3.6 i 0.6 - 6.0 2 0.3 - 1.3 0.2 - 2.0 0.3(-6 2.27 0.07(-6! 1.7 0.1 - 1.2 1 0.2 - 110D%g 124Sb 3.2 0.2 - 1.9 0.3 - 7.5 1.1(-7 1.1 0.1 - 1.8 0.1 - 4 1 1(-7) 12sSb 3.6 1 0.2 - 1.5 1 0.7 - 1.7 i 0.2(-6 1.1 0.2 - 3.6 8 2(-7) 3.5 1 0.3 - 1.0 0.2 - 8.9 i 0.7(-7 7.3 0.9 . 2.26 0.2-6))0.09(-6 7.9 0.5 - luoLa 1.1 2 0.5 - 141Ce 1.7 i 0.6 -6) 1.210.3-} 2.1 0.5f6) 7)
<1.1(-7) 2.4 0.3 (-6) 1.6 1 0.5 -
144Ce 1.310.5-5) <3.7(-7) <1.6(-7) 4.3 0.6 (-6) 1.4 0.3 -
TABLE B.17 RADIONUCLIDE CONCENTRATIONS IN BAE DISTILLATE 2/16/78; 15:45 2/17/78; 13:45 2/20/78; 18:12 2/22/78; 16:05 5/2/78; 09:10 Nuclide (pCi/ml) (pCi/ml) (pCi/ml) (uC1/ml) (uti/ml) 131I 2.61 1 0.07(-5) 1.94 a.C5(-5) 7.5 1 0.2(-6)* 1.64 1 0.02(-4) 2.78 0.07(-5) 134Cs 8 4(-8) 1.0 0.5(-7) 5.7 i 1.6 - * <1.3(-7) 2.0 0.7(-7) 136Cs 6.6 1 0.9 -
- 137Cs <7(-8) 6.7 4.2(-8) 2.9 1 0.1 - * <1.4(-7) <1.3(-7) 51Cr. <6.4(-8) 1.0 0.6(-7) <3.7(-9)* 2.8 2.4(-7) s*Mn 59Fe
<7.2(-8) 6.5 1.8(-8) 1.14 0.09(-8)* <1.3(-7) <8.6(-7)(-7) 5.9 1 0.7
<6.8(-8) <5.0(-8) 2.5 1 0.2(-8)* 1.3
- 0.2(-6) <1,7(-7) n 57C0 <5.5(-8) <4.1(-8) <3.6(-9)* <8.5(-8) 4 58C0 2.0 0.5(-7) 2.010.3(-7) 8.310.2(-8)* <2.4(-7) 3.9 0 2(-6) 1.7 i 0.1(-6) 60Co <1.4 - 1.911.3(-7) 1.21 1 0.04(-7)* 1.3 i 0.2(-6) 2.9 0.2(-6) 65Zn <6.4-)I <4.7(-8) <1.8(-9)* <2.1(-7) 95Zr l <4.5(-8) 7.0 1 1.4 -9)* <1.7(-7) 1.3 0. 9(-7) 3 1(-7) 9sNb <5.9-j/
<6.6 - 8.2 i 5.6(-8) 2.1 i 0.8 -
- 2.6 1.1(-7) 2.4'i0.6(-7) 4 99Mo <7. 2 - i 2.5 i 0.9 -
- 2.1 0.3 <8.1(-8) lo3Ru <4.1 - <5.2(-8)(-8) 6.4 2.1 7.6 1 1.4 - * <1.5(-7)(-6) <1.0(-7) 106RuD <9.1(-8)h <4.8(-8) <1.7(-9)* 1.210.6I;-6 <1.0(-6) 110 mag 1.4 i 0.7(-7) 1.1 1 0.3 - 4.2 0.1 -8)* 4.3 0.5L-7 <2.7(-7) 124Sb ' <5.7(-8) 1.0 0.3 - 5.8i0.3-8)* 1.3 1 0.2I;- 2 i 1(-7) 12sSb 3.110.8(-7) 2.3 1 0.5 - 6.9 0.5 -8)* 1.4 0.1I -
<3.9-ll 140La <5.8 - 8.0 i 3.5 - <1.0 -
- 9.6 i 0.7(,- <5.9 - J 141Ce <6.0 - <4.1(-8) <2.3 - * <2.9(-7) <1.5 - J 144Ce <5.8 - 5.3i4.1(-8) <6.4 - * <2.0(-7) <6.6(-7)
I
- Resin Concentration Samples
TABLE B.17 (cont'd) RADIONUCLIDE CONCEN DATIONS IN BAE DISTILLATE 5/3/78; 12:00 5/4/78; 14:45 5/9/78; 15:21 5/11/78; 12:57 5/15/78; 11:00 5/17/78; 15:12 Nuclide (uC1/ml) (uC1/ml) '(uci/ml) (pCi/ml) (pCi/ml) (uti/ml) 4 131I 7.9 0.2(-6) 9.0 0.8(-6) 1.6 i 0.1(-6) 3.6 0.2(-6) 2.7 0.8(-7) 2.6 0.6(-7) 134Cs <2.2(.-7) <2(-7) 1.4 0.1(-7)* 5.2 i 2.4 -9)* 4.7 3.6(-9)* 1.210.4(-7) 13cCs 3.9i2.2-9)* 137Cs 1.5 1 0.4(-8)* t
- z M (-8) 3.1 0.1(-7)* 5.612.6-8)* 1.4 1 1.0(-7) 4.0 2.8(-9)*
- s1Cr 4.312.3(-7) 9.4 1.4(-8)* 9.7 1.7-8)*- 5.4 2.0(-8)* 5.4 1 1.4(-8)*
54Mn <1.0(-6) 2.6 1 0.6 (-8)*2.2 0.6(-7) 1.2 0.1 1.4 1 0.5 -7) 3.7 1 0.3(-8)* 1.0 1 0.4(-7) 59Fe 4.6 i 2.2(-8)* <9.7(-8) <1.7(-7)(-7)* 1.7 0.5 - * <1.4(-7) <1.2(-7) 4.3 1.3 -
- 57Co <5.2(-8) <1.3(-7) 1.9 1.2 - * <8.9(-8) cg 58C0 <7.9(-8) 3.2 1 0.3 (-7) 2.8 1 0.4(-7) 3.8 i 0.8(-7) 1.4 i 0.6 - 2.2 0.5(-7) 1.6 1 0.4 -
& 60Co 1.2 1 0.3(-7)* 1.2 1 0.2(-7)t 8.20 1 0.05(-7) 3.0 0.1 -
- 3.8 0.2(-7)* 3.3 0.2 - t ssZn <1.8(-7) <8.0(-8) <1.4(-7) <2.7(-7) <,0(-7) <1.9( 7) 95Zr 712(-8)* 7.5 t 0.4 -
- 3.9i0.5(-8)* 3.0 1 0.4 -
- 2.5 0.9(-8)* 2.1 0.5 -
- 95Nb 5.5 i 3.5(-8 1.0 1 0.9 - 6.8 i 0.6(-8)* 4.6 1 0.4 -
- 3.3 0.4(-8)* 3.6 1 0.4 -
- 99Mo 1.1 0.5(-8
- 5.2 2.3 - <1.6(-7) 2.9 2.0 - * <8.3(-8) 1.5 0.9 -
- 103Ru 1.9 0.8(-8* <9.9(-8) 1.9 0.3 -
- 6.9 2.1(-9)* 9.4 1 2.2 -
- 106RuD <1.0(-6) 5.7 1 1.6(-8 * <1.1(-7)(-8)*
7.9 4.0 <8.3(-7) <6.7(-7) <7.3(-7) 4.310.8-
- 110 mag 6i1(-7) 3.0 0.5(-7 3.9 i 0.4(-8)* 8.0 i 0.6 -
- 6.5 1 0.7 -
- 124Sb 2 2 1(-8)* 3.7i0.5(-8* <9.3(-8) 1.9 0.5 -
- 1.2 0.4 -
- 1.7 0.3 -
- 12sSb 1.1 0.2(-7)* 7.4 0.5(-8)* 6.6 06(-8)* 7.7 1.0 -
- 4.5 1 0.5 -
- 5.3 0.6 -
- luota 2.3 1 0.8(-8)* <9.5 -PI <1.2 - <1.3(-7) 3.7 i 2.2(-9 * <6.6 -
141Ce <1.4(-7) <7'.0 -8) <1.3 - 7.9 2.1(-9)* <1.4(-7) <1.2 -
<6.3 -8) <1.3 - 1.1 0.2(-7)* <6.3(-7) <5.4 -
144Ce 7 3(-7)
* - Resin Concentration Samples t - Calculated from seCo ratios
TABLE B.18 RADIONUCLIDE CONCENTRATIONS IN BAE BOTTOMS
'2/16/78; 16:00 2/17/78; 14:00 2/20/78; 18:45 2/22/78; 16:00 Nuclide 5/2/78; 09:06 (vCi/ml) (uC1/ml) (uC1/ml) (uC1/ml) (uci/ml) 131I 7.8 1 0.1(-3) 7.5 0.3(-3) 1.34 0.02(-1) 1.14 0.03(-1) 2.53 0.06(-2) 134Cs 1.0 0.2(-4) 4.010.8(-5) <4.3(-5) <3.6(-5 <1.1 -
137CS 2.1 1 0.2(-4) 1.1 1 0.3(-4) 4.7 i 0.9(-5) <6.0(-5 <3.6 - 51Cr 3.5 i 0.2(-3 2.8 0.1 - 2.2 1 0.2(-3) 4.6 0.5(-3) 9.0 0.9(-4) 54Mn 4.5i0.2(-4 2.9 1 0.1 - i,10 0.09( 4) 2.3 0.1(-4) 1.2610.07(4) 4.310.6(-4 59Fe 4.6 i 0.6 - 7.7 0.3(-4 2.31 i 0.0F(-3) 1.1 1 0.1 57Co 8 2(-5) 512(-5) 7.0 2.7(-5 <4.4(-5) <1.0(-5)(-4) seco 8.22 0.15(-3) 7.3 i 0.3(-3) 7.25 i 0.10(-3) cm 5.03 i 0.07(-3) 8.48i0.06(-3) 2.46 0.05(-3) SOCo 3.58 i 0.18( 3) 2.01 1 0.03(-3) 2.53 i 0.05(-3) 1.20 0.02(-3)
$ 65Zn 95Zr
<4.9(-5) 7 t 4(-5) <2.7(-5) <2.3(-5) <2.1(-5) 6.0 t 0.4(-4) 5.0 1 0.2 - 2.4 0.2l 3.6 i 0.2(-4) 1.310.1(-4) ssNb 1.03 i 0.03(-3) 8.6 i 0.1 - 4.010.3l- -
6.1 2 0.2(-4) 2.3 1 0.1(-4) 99Mo <2.4(-5) 1.6 1 0.3 - 1.3 0.1 4.17 0.07 3) 2.12 4.5 i 0.1((- lo3Ru 0.06(-4) 6.1 i 0.2(-4) 4.84 1 0.05 4) 1
, 2.9 i 0.2(- 7.0 1 0.6 -
106RuD 9 1(-4) 6.8 1 0.1 - 4.9 i 0.8 - 6.8 1 2.9(- 2.6 1 0.6 - 110 mag 2.0 1 0.1(-4) 1.4 0.3 - 1.3 (1 - 1.16 0.11 - 1.7 i 0.1 - 124Sb 5.110.1(-3) 5.3 i 0.4 - 9.7 0.1(- 0.1 4.56 i 0.09 - 3.3 + 0.2 - 12sSb 6.62 0.09(-3) 6.3 i 0.8 - 1,32 1 0.01(-2) 4.61 1 0.06 - 5.0 0.6(-4 140La <2.5(-5 1.6 t 0.3 - 6.2 1.0(-5) 6.9 0.2(-4) 2.010.1(-4}, 141Ce <1.9(-5 2.8 i 0.5 - <2.2(-5) <2.6(-4) <1.8(-5) 144Ce <5.1(-5 <3.4(-5) <6.8(-5) <2.8(-5) <9.4(-5)
TABLE B.1R (cont'd) RA9IONUCLIDE CONCENTRATIONS IN BAE BOTTOMS 5/3/78; 11:55 5/4/78; 14:57 5/9/78; 14:40 5/11/78; 12:50 5/15/78; 11:25 i7/78; 14:29 Nuclide (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) (pC1/ml) _ (pCi/ml) I 'll 5.8 0.2(-3) 4.23 i 0.08(-3) 1.45 0.06(-4) 1.61 1 0.02(-4) 6.6 0.2(-5) 2.49 0.05(-5) 1340s <2.4(-6) 1.5 0.1 - 1.4 0.3(-5) 7.0 0.9(-6) 8.9 0.5(-6) 7.0 i 0.5(-6) 13cCs 7.8 1. 2 - 137Cs <2.2(-5) 5.4 0.2 - 3.6 0.4(-5) 2.6 0.1(-5) 3.36 0.05(-5) 2.33 0.06(-5) 3H 9.4 0.3(-3) 4.4 r 0.1(-2) ** ** 14C 1.210.1(-5) 1.3 i 0.1(-5) ** 32P 3.5 t 0.5(-4) 2.3 2 0.3(-4) ** ** ** 51Cr 2.7 1 0.3(-4) 2.15 0.09(-4) 8.2 2 2.0(-5) 7.7 0.2(-4) 1.02 0.06(-4) 4.9 0.4(-5) 54Mn 4.8 0.3(-5) 4.2 0.1(-5) 3.0**i 0.4(-5) 2.37 0.04(-4) 3.1210.08(-5) 1.75 0.05(-5) 4.76 2.31 ** ** ** 55Fe 0.01(-3) 0.01( 3)
? 59Fe 5.0 1 0.3(-5) 4.4 1 0.3(-5 <8.6(-6) 4.9 0.3(-5) 1.98 10.0R(-5) 1.01 i 0.09(-5) g 57C0 4.1 1 0.9(-6) 6.5 1. 3(-6 <4.1(-6) 1.1 0.1(-5) 2.9 0.2(-6) 2.320.5(-6) seCo 1.rJ 0.02 (-3) 1.11 0.09(-3) 6.2 0.2(-5) 1.8610.02(-3) 5.43 0.07(-4) 3.29 0.05(-4) 60Co 4.6 0.1(-4) 3.6 i 0.1(-4) 2.9**1 0.2(-4) 2.14 0.02(-3) 3.71 0.03(-4) 2.40 0.03(-4) 6 3Ni 1.7810.02(-4) 1.67 0.02(-4) 657n <5.0(-6) <2.5(-6) <6.7(-6) 2.4 0.2(.5) 7.5 1.5(-6) 3.1 0.9(-6) 0.04 - ** ** ** **
89Sr 3.3 i 0.5 -7) 1.19 90Sr 1.9 i 0.1 - 1.53 0.08 - 91Y 9.8 0.1 - 6.47 0.07 - ** ** ** ** 95Zr 5.5 0.3 - 2.4 i 0.3(-5) 2.0 0.5 - 2.00 0.04(-4) 2.54 1.52 0.07(-5) 95Nb 9.2 0.3 - 3.4 0.4 - 3.53 i 0.07(-4) 4.66 0.07((-5)0.07-5) 3.0 0.l(-5) 99Mo 5.7 i 0.1 - 2.49 2 0.09(-5 5.4 0.4(-5) ) 5.4 1.6 - 3.6 0.5-6) 5.5 1.6(-7) <4.9(-7) 103Ru 2.7 1 0.1 -5) 1.8 0.1(-5) 1.2 0.4 - 3.9 0.1 - 1.05 0.04(-5) 6.8 0.4(-6) 106RuD 8 1(-5) <5. 3(- 6) <5.8(-6) 1.5 0.2 - 4.0 0.4(-5) 2.7 0.4(-5) 110 mag 9.1 0.5(-5) 1.28 0.06 - 1.7 0.5 - 3.1 0.1 - 1.09 1 0.04(-4) 1.30 0.06(-5) 124Sb 1.45 i 0.09(-4) 1.05 0.06 - 1.3 0.5 - 3.21 0.08( 5) 1.73 0.06(-5) 1.08 3 0.10(-5) 125Sb 2.15 0.04(-4) 1.48 0.03 - 4.7 1.0 - 6.8 0.2 -5 4.2 0.l(-5) 2.2 0.1(-5) 140La 5.5 i 0.3(-5) 2.5 0.2(-5) 1.1 0.6(-5) 1.2 0.2 -5) 1.33 0.03(-5) 9.9 0.5(-6) 141Ce <5.1(-6) <1.6(-6) <3.9(-6) 6.1 0.6 -6) 9.6 3.0(-7) 1.1 0.4(-6) 144Ce 1.810.7(-5) <1.9(-6) <4.0(-6) 8.9 i 0.3(-5) 1.5 0.2(-5) 1.4 0.3(-5) l l ** Radionuclide not measured.
TABLE B.19 l RADIONUCLIDE CONCENTRATIONS IN BAE CONDENSATE DEMINERALIZER EFFLUENT 2/16/78; is:45 2/17/78; 13:35 2/20/78; 18:30 2/22/78; 15:15 5/2/78; 09:14 Nuclide (pCi/ml) _ (pC1/ml) (pCi/ml) (pCi/ml) (pC1/ml) 131I 4.02 0.09(-6) 4.12 0.08(-6) 8.2 0.l(-7)* 1.4010.03(-5) 2.1 1 0.1(-6) 134Cs 2.1 0.6(-7) 9.6 3.8(-8) 3.0 0.3(-8)* 5.1 3.9 - <1.2(-7) 136Cs 8 5(-8) 1.34 i 0.02(-8)* 2.1 1.3 - 137Cs 3.110.6(-7) 1.9 0.4(-7) 8.0 2.8(-8)* 5.0 1 2.8 - 2.2 i n.9(-7) 51Cr <5.5(-8) 5.5 3.8(-8) 4.0 1 0.7 -
- 5.E _ 0.8 - 6 i 3(-7) 54Mn 1.310.7(-7) 1.a 0.3(-7) 2.2 1 0.2 -
- 7.0 5.6 - <8.5-8) 59Fe <8.2(-8) <6.6(-8) 1.1 1 0.4 -
- 1.0 0.5 - <1.3-7) 57C0 <8. 7(-8) <3.7(-8) <5.5(-9)* <4.1(-8) <9.8-8) m 58C0 3.3 0 6(-7) 5.0 0.5(-7) 7.8 1 0.2(-7)* 4.8 1 0.2(-6) 3.5 1 0.6(-7) 60Co <1.9(-7 3.0 1 2(-7) 2.2 0.1(-7)* 9.9 1.0(-7 <1.3 -
G 5Zn <7.0(-8 <7.8(-8 <3.5(-9)* <4.2(-8) <2.4 - 95Zr <7.2 - <5.2(-8 6.6 2.4(-9)* 5.5 4.8(-8 <2.3 - 95Nb <8.7 - <1.0(-7 <l.5(-8)* 1.2 0.5(-8 <l.1 -7) 99Mo <8. 3 - 4.8 1. 5 - 3.9 0.9(-9)* 2.3 0.4 - <9.2(-8 103Ru <6.4(-8) 9.0 1 2.7 - 3.8 3.0(-9)* 3.4 4.2 - <9.7 - 106RuD <6.0(-8) 3.8 3.0 - <3.1(~9)* 1.9 4.3 - <6.2 - 110 mag 4.9 0.9(-7) 5.4 0.7 - 1.5 0.1(-7)* 2.4 0.4 - <l.2 - 124Sb <6.4(-8) 8.6 2.9(-6) 1.7 i 0.2(-8)* 2.5 3.9(-8) <1.5(-7) 12sSb <6.3(-8) 5.9 1 0.5(-9) 2.4 0.3(-8)* 6.9 4.7(-8) <2.1(-7) luota <8.7(-8) <4.4-8) 2.1 2.2(-9)* 2.3( 8 <5.2(-7 141Ce <5.5(-8) <3.4 -8) <l.l(-9) <4.2-) <l.5(-7 M4Ce <7.5(-8) <3.4 -8) <l.8(-9)* <4.0 - ) <5.4(-6
- Resin Concentration Samples
TABLE B.19 (cont'd) RADIONUCLIDE CONCENTRATIONS IN BAE CONDENSATE DEMINERALIZER EFFLUENT 5/3/78; 12:03 5/4/78; 14:48 5/9/78; 15:29 5/11/78; 13:07 5/15/78; 11:02 5/17/78; 15:21 Nuclide (uCi/ml) (pCf/ml) (uCi/ml) (uCi/ml) (pCi/ml) (pC1/ml) 1311- 2.1 2 0.7(-7)* 6.7 1 0.1(-7)* 2.0 0.4(-7) 1.6 0.3(-7) 8.5 1.8(-9)* 9.8 1.6(-8)* 134Cs 1.8 0.1(-7)* 9.5 0.3(-8)* 5.9 1.3(-9)* 6.9 0.4 -8)* 3.1 0.3 -8
- 1.110.4(-7) 137Cs 4.9 1 0.1(-7)" 2.2 0.1(-7)* 3.? 0.2(-8)* 5.4 3.7-8) 6.4 i 4.6 -8 1.2 i 1.1(-7) stCr <6.7(-8)* <3.9(-9)* 1. 5 1 0.2 (-7)* 4.6 1. 3(-8)* <3.2(-7) <6.9(-7) 54Mn <1.2(-8)* 5.4 i 0.7(-9)* <1.3(-7) 1.8 0.2(-8)* 6.7i3.8(-8) 3.0 i 2.3(-9)*
59Fe <l.8(-8)* <8.8(-9)* l.4 i 0.4 <1.2(-7) <8.4(-8) <2.6(-7) 57Co <8.0(-8)* <2.7(-9)* <9.9(-8) (-8)* <4.4(-8) <3.4(-8) <7.9(-8) 5800 7.0 1 0.4(-8)* 6.210.2(-8)* 8.5 0.1(-7)* 2.5 0.6(-7) 6.1 1 0.4(-8)* 8.6 1 4.1(-8) 60Co 8.8 1 0.4(-8)* 9.4 1 0.2(-8)* 4.8 1 0.1(-7)* 6.0 1 5.0(-8) 5.8 1.0(-8)* 6.8 1.0(-8)* i ssZn <2.2(-8)* <8.2(-9)* <1.5(-7) <9.9(-8) <2.1(-7) R 95Zr 5 2(-9)* 9.1 3.7-9)* 4.6 0.5(-8)* <1.0(-7) <9.6(-8) 4.3 2.8 (-9)* <1.4(-7) 95Nb <1.1 -
- 7.6 5.3 -9)* 8.1 i 0.5(-8)* 8.3 t 2.4(-8)* <6(-8) 5.1 2.2(-8)*
99Mo <7.6 - 2.2 0.8 -9)* <1(-7) 2.9 1.6(-9)* <4.2 - 9.1 1 3.1(-9)* 103Ru <1.0 - * <4.9(-9)* 1.2 0.2(-8)* <4.8(-8) <3.6 - <6.9(-8) 106RuD 1.0 0.4(-7)* <5.8(-9)* <4.4(-7) <3.5 - 110 mag 2.1 0.2(-8)* 2.9 1 0.3(-8 * <1.0(-7) 2.0 1 0.1 (-7)* 2.6 1 0.4 -8)* 1.1 0.3 - <7.4(-7)(-8)* 1.8 0.4 124Sb <1.7(-8)* 2.4 0.4(-8
- 4.0 1 0.7(-8)* 7.2 3.2 -9)* 6.4 1 3.3 -
- 6.5 3.7(-9)*
12sSb <4.2(-8)* 6.1 0.4(-8
- 7.0 0.6(-8)* 1.8 0.6 -7) 1.4 0.5 - 2.7 i 0.5(-8)*
luota 3 1(-9)* <4.7(-9)* <9.6(-8) <3.5(-8) <5.1(-8 <7.8(-8) 141Ce <1. 3(-8) * <2.7(-9)* <9.7(-8) <7.4(-8) <5.7(-8 <2.2(-7) 144Ce <5.7(-8)* <3.1(-9)* <1.3(-7) 2.5 1 2.0(-8)* <3.6(-7 <5.9(-7)
- Resin Concentration Sanples
TABLE B.20 RADIONUCLIDE CONCENTRATIONS IN BAE CONDENSATE DEMINERALIZER FILTER EFFLULaT 2/16/78; 15:30 2/17/78; 13:25 2/20/78; 18:05 2/22/78; 13:55 5/2/78; 09:20 Nuclide (pC1/ml) (pCi/ml) (pCf/ml) (pCi/ml) (pCf/ml) 1311 4.210.1(-6) 3.85 1 0.09(-6) 8.2 0.1(-7)* 1.28 0.03(-5) 2.25 0.05(-6) 134Cs 7.7 3.7(-8) 8.7 1 6.6(-8) 3.2 1 0.3(-8)* <1.7(-7) <9.4(-8) 136Cs 8.713.4(-8) 6.5 i 2.5(-8) 1.3 0.2(-8)* 137Cs 2.5 1 0.5(-7) 2.9 i 0.4(-7) 8.8 1 0.2(-8)* 8.5 0.6(-8) 2.'1 1 0.5(-7) s1Cr (4.7 - 5.9 1 3.6 -9) 1.0 0.3 - * <1.4(-7) <3.9(-7) 54Mn <4.1 - 5.3 3.5 -8) 4.7 i 1.0 - * <1.3(-7) 5 2(- ) 59Fe. <6.5 - 7.1 5.6 -8) 2.711.5-
- 1.7 1 0.6(-7) <9.1(-8 m 57Co <7.8(-8) <3.5(-8) 8.8 1 2.4 - )* 1.9 1.0(-8) <5.1(-8 a
m 5800 1.7 i 0.1(-6) 4.7i0.5(-7) 1.8 1 0.1 -
- 5.44 i 0.02(-6) 1.4 3(-7) (
60Co 4.1 1 2(-7) 1.9 1 1.2(-7) 6.3 0.4 -
- 9.2 0 9(-7) 1.8 i (-6) 652n < 8. 5 f'- <5.6(-8) <2.7(-9)* <9.9 - <1.2 -
95Zr <5.71 8.413.9(-9) <2.5(-9)* <9.8 - <7.6 - 95Nb <1.11 [- 1.310.9 4.1 1.2(6)* <8.3 - <5.2 - 99Mo <6.71 - <4.9(-9) (-8) <1.5(-9)* <1.1 - <5.6 ,- 10 3Ru <5.0(- 7.9 4.0(-9) 3.3 3.0(.')* 1.11 1 0.3(-7) <4.5( - 106RuD <5.6(- <4.8(-8) <4.7(-9)* <6.7(-8) <4.30- l 110 mag 4.9 1 0 5(-7) 5.2 0.4(-7) 1.8 1 0.1 - )* 3.0 i 1.0(-7) <8.0f - l 124Sb <3.5 -
<4.3 -8) 3.1 1.0 - )* <7.0(-8 <8 2 - l 125Sb <4. 3 (,- 1 <4.2 - 1 8.5 2.( 9 * <1.3-7) ; <1.2 - l 140La <6.11 -
<4.2 'l <1.9-9)* <6.5 -8" <6.0 - l 161Ce <4.0d- <3.6-l} <l.6 -9)* <1.1-7h <8.6 - l 144Ce 5.5 t 3.7(-8) <3.5 -8) <1.2-9)* <1.0(-7) <3.9 - l
- Resin Concentration Sanples
TABLE 8.20(cont'd) RADIONUCLIDE CONCENTRATIONS IN BAE CONDENSATE DEMINERALIZEP, FILTER EFFLUENT 5/3/78; 11:59 5/4/78; 14:40 5/9/78; 15:01 5/11/78; 12:10 5/15/78; 10:45 5/17/78; 14:54 Nuclide (pCi/ml) (uC1/ml) (pC1/ml) (uC1/ml) (uti/ml) (uC1/n1) 131I 2.1610.06(-7)* 7.1 0.8(-7)* 1.5 0.5(-7) 1.3 0.3(-7) 6.8 1 2.4(-8) 1.8 0.3(-8)* 134Cs 4.6 0.2(-8)*- 7.6 i 0.8(-8) 1.0 0.5(-7) 7 3(-8)t 7.1 2 3.0(-8) 137Cs 2.2 6.1 2 0.P 0.8{-8)*
.-8)* 9.3 0.9(-8)* 3.3 2.8(-9)* 8.1 4.1(-8) 1.5 0.2(-8)* 1.1 0.4(-8)*
3H ** ** 2.26 0.07(-2) 14C ** ** 5.9 1 0.6 -6) ** ** 32P ** ** ** 7.3 2 0.7 -7) 51Cr 4 1 3(-8)* <4.3(-9)* <4.4(-8) 1.8 1.6 - * <4.2(-7) <3.7(-7) s4Mn 7 i ** 5(-9)* <4(-9)* 7.8**1 5.9(-8) 1.0 0.2 -
- 6.1 0.6(-8) 6 2(-9)*
0.1 - ** ssFe ** 1.7 ca 59Fe <1.9(-8)* <1. 3(-8)* <9.5(-8) <8.7(-8) <9.7(-8) <1.2{-7)
& 57Co 311(-8)* <3.2(-9)* <5.9(-8) <5.1(-8) <6.0(-8) <7.ls-8) m 0.4 - 6.8 t 0.4(-8)* 1.3 0.3(-7) seCo 9.9 1. 3(-8)* 7.9 0.4(-8)* 4.8i0.6(-8) 1.4 3.0 1.3(-8)*
- 59Co 1.0210.10(-7)* 6.6 0.2(-8)* 2.3 1.0(-8)* 9.8 1.0 -
- 2.9 0.9(-8)*
6 3Ni ** ** ** 2.0 1 0.3 - 65Zn <2(-8)* <8 (-9)* <1.3(-7) 7) <9.8(-8) <1.}-7)
** ** 4 g 1(_g) ** **
903p ** 9ly ** ** <1(.9) 95Zr <5.4(-9)* <5.9(-9)* <5.3(-8) 5.7 3.6(-9)* <1.2(-7) <1.1(-7) 95Nb 2.4 06(-8)* 8.5 6.9(-9)* 1.0 0 8(-7) 1.3 0.2(-8)* 4.4 2.1(-9)* 6.9 2.8(-9)* 99Mo <5.7 - * <2.9(-9)* <7.6 - <7.8(-8) <7.6(-8) <5.4(-8)
- 103Ru <7.4 - * <5.2(-9)* <5.3 - 3.2 1.7(-9)* <4.d(-8) 2.9 1.6(-9)*
losRuD <8. 3 -
- 3.7 1.1(-8)* <5.0 - <6.7(-7) <4.0(-7) <d.0(-7) 4.5 4.1 0.9(-8)* 2.9 0.7(-8)* 3.2 0.6(-8)* 2.3 0.4(-8* 1.5 0.4(-8)*
110 mag 0.6(-8)* 1.0 0.4(-8)* 12"Sb 2.0 1 0.6(-8)* <5.4(-9)* <4.3(-8) 1.2 0.5(-8)* 5.7 3.9(-9
- 12sSb 4.1 1.3(-8)* 6.1 06(-8)* 2.9 0 3(-8)* 3.4 0.5(-8)* 4.0 0.5(-8* 2.5 0.5(-8)*
140La <8.1 -9)* <5.6 - * <7.6 - 7.3 3.6(-9)* <4.0 - <4.6(-8
; 141Ce <1.0 -8)* <2.6 - * <6.7 - <9. 5 (-8) <8.8 - <8.9(-8 1
144Ce <4.4 -8)* <3.8 - * <6.0 - <4.0(-7) <5.6 - <5.0(-7
- Resin Concentration Sayles
- _ _ _ _ _ _ _ _ _ _** lh00GkWD@ M Gagund. _____ ___
TABLE B.21 RADIONUCLIDE CONCENTRATIONS IN RADWASTE EVAPORATOR FEED 1/5/78; 13:15 1/F ~~8; 16:31 1/9/78; 09:46 1/9/78; 12:35 Nuclide (uC1/ml) , (uCi/ml) (uCi/ml) (pCi/ml) 131I 8.5410.07(-5) 8.0* t 0.1(-5) 5.610.1- 4.82 0.06(-4) 1321 <1 (-3) 7.4 1.2 - 9.1 1 1.4 - 133I 2.49 i 0.60(-6) 4.4
- 0.4(-6) 2.1 0.2 - 2.2 1 0.1 -
134I
- 1.2
- 0.2(-5) 6.4 1 1.8 - 1.2
- 0. 3 -
135I <5.6(-6) <6.9(-7) 1.3 s 0.2 - 1.5 0. 2 - 88Rb <3(-5) 6.1 1 1.2 - 3.6 1.0 - 134Cs 4.7 0.7(-5) 4.65 i 0.06(-5) 7.3 i 0.1 - 6.7 i 0.2 - 13sCs 2.510.3(-6) 2.5 i 0.2 - 3.010.7- 4.0 0.9 - 137Cs 2.4 0.6(-5} 8.9 0.2 - 1.33 0.02 4) 1.24 1 0.02 -4) 138Cs
- 5.311.0- 1.5 0.2(-5) 6.4 1.6(-6) 3H ** ** ** **
IT ** ** ** ** 24Na 1.710.2-6 2.3
- 0.2 -6 6.3 0.6(-6) 7.6 0.7 -6) 2.2 b04-5) .' 5 ib07-5) 1 0.1(-5) 1-59Fe 2.8 0.8(-6) <3(-6) <3(-6) 57Co 2.7 1 0.2(-6) <1.3(-6)(-6) 2.5 0.3 1.4i0.5(-) 2.4 0.5(-)
ssCo 1.16 0.03 - 1.00 0.03 - 6.2 i 0.1(- ) 6.6
- 0.1(- )
60 3.19 1 0.09 - 2.72,i 0.08 - 1.85 0.02 -4) 1.92,1 0.02 -4) ss 2.5 0.1(-5) 2.5 0 2.7{ 0.07(-5) 3 2(-5) 2.4, 1 0.2(-5) 90Sr ** ** ** ** 91Sr *
- 91Y ** ** ** **
95Zr 5.9 0.5(-6) 2.9 0.4(-6) *
<3(-6) 9sNb 9.8 0.7(-6) 4.7* i 0.3(-6) 3.1 0.6(-6) 2.2 0.6(-6) 99Mo * *
- 101Ru 4.7 i 0.3(-6) 3.4 1 0.3(-6) 2.3 0.6(-6) 2.6 0.6(-6) lo6Ru 8.6 2.8(-6) <1(-7) <3(-6) <4(-6)
- 2.9 110 mag 0.6(- ) *
- 124Sb 1.86
- 0.06(-5) 2.04 0.07-5) 3.0 0.2(-5) 2.6 0.2 -
12sSb 1.0510.11(-5) 1.11 1 0.07 -5) 2.0*1 0.2(-5) 2.0 0.2 - 129Me 1.9* 1 0.9(-5) <1.1(-6) 4.8 1. 5 - 129Ta * *
- 140Ba 2.5 i 1.0(-6) <1.2(-6) <3(-6)
- l th ola 1.9* 1 0.5(-6) 5.6*1 0.9(-7) 1.1 0.4(-6) 1.0* 1 0.3(-6) i 141C)
- 1 i
B-97 l
t i l TABLE B.21 (cont'd) RADIONUCLIDE CONCENTRATIONS IN RADWASTE EVAPORATOR FEED l 1/9/78; 13:20 1/9/78; 17:33 1/10/78; 09:13 1/11/78; 12:43 Nuclide (uCi/ml) (pCi/ml) (pCi/ml) (uC1/ml) l 131I 4.86 1 0.06( 4) 4.0 2 0.1(-4) 2.32 0.03(-4) 1.10 1 0.02(-4) 4.8 1.1 - 132I <6(-6) <8(-6) <2(-5) 133I 2.1 i 0.1 - 2.0 i 0.1(-5) 1.3 0.2(-5) 1.2 0.1 (-5) l 134I 5.6 2.5 - <1(-5) 2.35 0.09(-4 5.5 0.7(-4) l 135I 1.0 0.2 - 8.3 1.4(-6) 5.0 2.2(-6)) <3(-6) serb 3.9 *
- i 1.6(-5) <2(-4)
' 134Cs 6.6 i 0.1(-5) 6.4 i 0.2(-5 8.8i0.2(-5) 7.8i0.2(- 136Cs <2(-6) 2.8 i 0.6(-6 4.6 1.5(-6) 1.9 0.7(- 137Cs 1.29 0.03(-4) 1.26 0.03( 4) 1.40 0.02 4) 138Cs <6(-5) <1(-5) <6.8(-3) 2.0 0.7 (-5) 8.2 1.9(-4)
** ** ** 2.9 l 3H 0.1(-2)
** ** ** 7.5 i 14C 0.8(-6) l 24Na 8.6 0.6 - 6.7 0.7 - 4.7 i 0.6 - 4.0 0.5(-6)
! 51Cr 2.3 0.7 - 2.3 0.5 - 7.8 i 0.8 - 4.0 i 0.6(-5 , 54Mn 1.3 0.1 - 1.1 0.1 - 3.7 0.2 - 2.9 0.1(-5 55Fe 4.05 1 0.02( 4) 59Fe <2(-6) 4.3* i 1.2(-6) 1.1 0.2(-5) <4(-6) 57Co 5.1 0.8(-6) 3.5 0.5(- ssCo 7.64 i 0.5( 3.0 t) 0.05(-4 ) 1.77 0.02(-3) 1.1 0.1(- l GoCo 2.16 0.03(-4) 6.210.1(-4)) 1.78 ** i 0.02(-4 5.04 0.06(-4) 3.80 1 0.03 a) i l 63Ni 1.66 0.01( 4) , l 65Zn 2.3 0.2(-5) 2.1**i 0.2(-5) 2.4 0.2(-5) 3.3 i 0.2(-5 89Sr 2.76 i 0.08( 6) 90Sr 6.2 0.1(-7) 91Sr <4(-6) <3(- 6) <6(-6) <5(-6)
** ** ** 1.04 91Y 0.02(-6 95Zr 3.6 i 1.1(-6)
- 2.4 0.2(-5) 1.1 1 0.2(-5) )
i 95NL 2.7* i 0.7(-6) 3.2 0.7(-6) 5.3 i 0.4(-5) 2.1* i 0.1(-5) 99Mo * *
<2 -
- 1.5 103Ru 0.1(-5) 5.6* i 1.1(-6) lo6Ru <2 - 5.6 1.2(-5) i 11omAg <2 - *
- 1.2 0.3 -
124Sb 2.8 0.2(-5) 2.4 0.2(.-5) 3.1 0.2(-5) 1.7 i 0.1 - l' 12sSb 1.6 0.3(-5) 1.4* i 0.3(-5) 4.5 0.3(-5) 2.0 0.3 - 129mTe <4(-5) <4(-6) <3(-6) 129Te *- * *
- 14aBa * * *
- 140La 1.0* 1 0.3(-6) 7.6 2.5(-7) <2(-6) <1 (-6) 141Ce * *
- B-98
TABLE B.21 (cont'd) RADIONUCLIDE CONCENTRATIONS IN RADWASTE EVAPORATOR FEED 2/1/78; 16:10 2/5/78; 10:40 2/18/78; 13:30 4/28/78; 10:44 Nuclide (pCi/ml) (pCi/ml) (uCi/ml) (pCi/ml) 131I 2.7 i 0.1(-3) 2.0 0.1(-3) 8.3* i 0.1(-4) 1.9 i 0.1 -
'132I <4(-5) <2(-5) 1.4 0.1 -
133I 5.0 0.8(-5) 2.4 0.5 - 2.64 0.04(-4) 5.1 i 0.1 - 134I <6(-5) 7.6 0.9 - *
- 135I <1(-5) 1.9 0.6 - 3.2 0.1(-4) 3.4 0.1(-5) esRb <1(-4) <2(-5) *
<8(-5) 13'*Cs 6.0 0.1(-4) 7.5 0.1(-4) 1.7 i 0.1 - 1.1 i 0.1(-5) 135Cs <8(-5) <5(-4) 2.3 i 0.1 - 1.0 i 0.4(-6) 137Cs 8.1 0.1(-4) 1.6 0.1(-3) 2.7 0.1 - 1.95 i 0.01(-4) 13eCs <3(-5) 7.2 2.8(-6) 14C ** ** ** **
24Na 3.3 0.8 - 6.9 'O.9 - 7.7 1 0.4 - 7.8 i C.6 - 51Cr 7.1 0.6 - 5.3 0.3 - 7.5 0.2 - 3.6 i 0.3 - 54Mn 4.7 0.1 - 1.9 0.1 - 6.6 2 0.2 - 2.2 i 0.2 - ssFe ** ** ** ** 59Fe 6.7 i 0.9 - 7.9 0.8(-5) 8.3 0.4(-6 2.8 i 0.9 - 57C0 9.6 0.3 - 3.5 0.2 - 2.5 0.1 (-6 2.5 0.3 - 5800 4.0 0.1 - 1.3 0.1 - 8.6 t 0.2(-4 4.1 0.1 - 60C0 9.3 i 0.2 - 3.6 0.1 - 3.410.1(-4 4.2 i 0.1 -
's 3Ni ** ** ** **
65Zn 6.2 1.0(-5) 3.8 0.7(-5) 3.4 0.7(-6) <2(-6) 89Sr ** ** ** ** 90$p ** ** ** ** 91Sr <4(-4) <2(-5) 7.0 0(-5) 3.1 1.5(-6) 3 1 95Zr 1.1 1 0.1 -4) 8.2 i 0.6(-5) 1.5 i 0.1 - 9.2 i 0.7 - 95Nb 1.5 0.1 -4) 1.38 i 0.07(-4) 2.6 2 0.1 - 1.6 0.1 - 99Mo 2.1 1 0.3 -5) 7.4 2.3 -6) 2.6 + 0.1 - 3.5 i 0.2 - 103Ru 1.20 0.06(-4) 8.0 0.4-5) 8.0 0.2(-6) 1.7 0.9 - 106Ru <2(-5) 1.710.4-4) 6.2* i 3.1(-6) 1.2 1 0.4 - 110 mag
- 1.1 i 0.1 -
124Sb 9.1 -i 0.1(-4) 4.1 0.1(-4)- 4.2 0.1(-5) 1.0 i 0.1 - 125Sb 1.0 1 0.1(-3) 4.6 0.1(-4) 5.6* i 0.1(-5) 1.7 i 0.1 - 129mTe <2(-5)- <2(-5) <2(-5) 12sTe 4.2 t 1.3(-4) * *
<3(-5) 140Ba <2(-5) <2(-6) _ 1.9 i 0.1 - 7.4 1.3(-6) 14cLa 7.7 i 1.9(-6) 3.7 0.9(-6) 5.3 0.8 - ~2.9 i 0.2(-6) l 141Ce <8(-6) 1.5 0.4(-5) 8.4 1 2.3 - <8(-7) )
- Radionuclide not detected.
** Radionuclide not measured.
B-99
i l ! TABLE B.22 RADIONUCLIDE' CONCENTRATIONS IN RADWASTE EVAPORATOR DISTILLATE 1/5/78; 13:15 1/6/78; 16:31 1/9/78; 09:46 1/9/78; 12:35 Nuclide (uCi/ml) (uCi/ml) (uC1/ml) _ (uC1/ml) l 131I 5.4* i 0.1(-5) 2.6 0.1(-5) 2.2* i 0.1(-5) 2.1* 2 0.1(-5)
- 132I <3(-7) l 1331 *
<1 (-7) 8.6* i 4.1(-8) 1.2* 1 0.4(-7) l 134I 8.8 1 1.7(-7) 2.9 0.7(-6) .
l las! <1(-7) <1(-7) <2(-7) <1(-7) i serb <2(-7) <1(-6) i 134Cs 6.1 1 0.5 - 1.7 1 0.6(-7) <2(-7) j 136Cs 1.8 0.4 - <1 (-7) <2(-7) <2(-7) l 137Cs 9.7 0.8 - <2(-7) 9.0 5.6(-8) 6.3 5.1(-8) l 13eCs <6(-7) <4(-7) 3H l l ** ** ** ** 14C l 24Na <1(-7) <3(-7)
- 51Cr <1(-7) <2(-7) 54 6.310.6(-7) 3.5,;_0.5(-7) 2.2g0.5(-7) 3.0j 0.5(-7) 59Fe <2(-7) <1 (-7) <2(-7)
<2(-7) l s7Co <1(-7) 4.3 2.0(-8) 8.4 3.4(-8) l
! seCo 2.310.1(-5) 1.87 0.06(-5 1.30 i 0.04(-5) 1.11 i 0.05(-5) 60Co 5.0 0.2(-6) 4.9 0.3(-6)) 2.5 0.2(-6) -2.5 0.2(-6) 63Ni ss n
<1(,7) <1(,7) <2(-7,) <2(g) 903p ** ** ** **
91$p *
.<2(.7) <2(-6) <2(-7) l 91y 95Zr 1.910.6(-7) <1(-7) <2 -
95Nb 1.9*1 0.4(-7) <1(-7) <2(-7) <2 - 103Ru 6.1 2.6(-8) <2(-7) <2 - 106Ru 110 mag _ 1.4 1 0.4 - * *
- l 124Sb 2.0 0.4 - <6(-8) <1(-7) <1(-7) 2.3 0.8 - *
- 12sSb- <2(-7) 129Me *
<1(-7) 3.6 1.5(-6) <3(-7)
* <5 - <2(-7) 129Te <4(-7)
- 140Ba 1.910.9(-7) * <2 -
140La *
<8(-8) <1 - <1(-7) l l
l B-100
4 TABLE B.22 (cont'd) MDIONUCLIDE' CONCENTRATIONS IN RADWASTE EVAPORATOR DISTILLATE 1/9/78; 13:20 1/9/78; 17:33- 1/10/78; 09:13 1/11/78; 12:43 Nuclide (uC1/ml) _ _ (uCi/ml) (vC1/ml) (uCi/ml) 131I 2.07 i 0.04(-5) 2. '. i 0.1(-5) 1.8 0.1(-5) 1.14 0.02(-5) 233I- <2 - < <2 - <8(-7) 133I <1 - <2 - <3 - 8.6i7.0(-8) 134I <3 - <2 - <8 - <1 -5 135I <2 - <1 - <3 - <2 -7
<3 - <6(-5) f'8Rb d Cs <2 - * <2(-7) 1.4 0.5(-7)
* <3(-7) <2 1 136Cs <9 -
137Cs 1.2 0.7(-7) 1.1 0.7(-7) 6.2* i 6.0(-8) <2 - h 138Cs <2(-7)
* <1 - J 3H
** ** ** 3.2 0.1(-2) 14C ** ** ** 5.810.6(-7)
* <4 (-7) 34Na <1 (-7)
SICr <8(-7)
* <3(-7) <2(-7) 54Mn 3.6 i 0.7(-7) 2.8 0.6(-7) 5.0 1.1(-7) 5.313.6(-8)
** ** ** <2(-7) ssFe
- 59Fe <2(-7)
* <2(-7) 57Co <1 (-7)
- 1.7 0.4 -
- seco 1.60 1 0.03(-5) 1.06 0.04(-5 2.5 0.1 - 4.8
- 0.3(-6)
SOCo 3.7 0.2(-6) 2.9 0.3(-6)) 5.2 0.3 - 9.8 i 1.3(-7)
<7(_g) 63pf **
65Zn <2(-7) <2(-8) <2(-7) <2(-7) 89Sr ** ** ** 6 1 2(-8) 90Sr ** ** ** 2.5 0.2(-8)
* <4(-7) <
92Sr <6(-7) ** .2({-7) 91y ** ** <6 ssZr <2 - <2(-7)
<3 -7
<3 -7
<2I
<21 l8) 7
-7 9sNb <8 -
103Ru <2(,7) <3 -7 <2I 7 3, <1 , 110 mag <1 - * <1(-7) 124Sb <1 -
<1(-7) <1 - <1(-7) 12sSb <3 -
* <2 -
129mTe <2{ - * <2 - 129Te <2I - r 1408a * * *
<l(-7) 140La <5(-8)
-* <2(-7) <1(-7)
B-101
i l l l i i TABLE B.22 (cont'd) RADIONUCLIDE CONCENTRATIONS IN RADWASTE EVAPORATOR DISTILLATE l 2/1/78; 16:10 2/5/78; 10:40 2/18/78; 13:30 4/28/78; 10:44 ( Nuclide (uCi/ml) (uC1/ml) (uCi/ml) (uCi/ml) 131I 3.010.1(-5) 5.5 0.1 (-5) 3.8 0.1(-5) 2.710.1(-5) 132I
<41;- *
<3(-7)
- 183I <31 -
<3(-7) 2.2*1 0.2(-6) 2.4 0.2(-6) 134I <1 ll- 5.011.8(-7) <6(-7) 135I * *
<1 (- 2.3 1 0.2(-7) serb <1 - * *
<7(-6) 134Cs <2 - <4(-7) 1.5 0.3 - 9.6 0.7(-8) 136Cs <2 - <3(-7) 1.1 1 0.3 - <9(-8) l 137Cs <3 -
- 5.5 1 0.4 - 9.2 i 7.7(-8) 13eCs <4 - <9(-7) *
<4(-7) 3H ** ** ** **
14C ** ** ** *+ l 24Na <8 - * *
<1(-7) 51Cr <3 - <3(-7) 2.0 1 0.4(-6) <9(-7)
, 54Mn <2 - 9.7 1 3.6(-8) 9.2 1 0.5(-7) 3.6 1.6(-7) i sspe ** ** ** ** l 59Fe <5(-7)
- 2.7 1 0.8 - <2(-7) i s7Co <3(-7)
- 1.5 0.3 - <9(-8) seCo 1.0 0.6(-6) 1.9 t C.2(-6) 3.2 1 0.1 - 1.1 0.1(-6) 60Co 6.3 4.6(-7) 4.2**1 0.5(-7) 1.3 0.1 - 2.3**1 0.2(-6) 63Ni ** **
893p * ** ** ** 903p ** ** ** ** 91 r l <9(-7) <3(-7) <6(-6) <5 (-7) 95Zr <4 - <2(-7) 4.9 0.8 - 4.2i0.9(-8) 95Nb <2 - <2(-7) 9.5 i 1.4 - 1.1 0.3(-7) 10 3Ru <2 -
- 2.1 0.4 - <9 (-8) 106Ru <2 - *
<2(-7) <9(-7) 110n%g <3 -
- 7.5 0.9 -7) 7.2 1.0(-8 124Sb <9 -
- 2.6 0.4 -7) 4.0 1.0(-8 ,
12sSb <5 -
- 3.6 1.1 -7) 1.4* i 0.2(-7 j 129mTe * * * '
129Te * * *
- l 140Ba <7 -7 * *
<3(-7) 140La <5 -7
- 1.1 0.5(-7) <7(-8)
- Radionuclide not detected.
** Radionuclide not measured.
B-102
TABl.E B.23 RADIONUCLIDE CONCENTRATIONS IN RADWASTE EVAPORATOR BOTTOMS 1/5/78; 13:15 1/6/78; 16:31 1/9/78; 09:46 1/9/78; 12:35 Nuclide_ , (uCi/ml) (uCi/ml) (uC1/ml) (uC1/ml) 6.4,10.1(-3) 8.310.1(-3) 8.5, i 0.2(-3) 9,4,?0.1(-3) 133I 4.9 1 1.4(-5) <6(-5) 8.1 0.7(-5) 1.2 1 0.1 - 134I 7.6' 1.8(-5) 8.4i1.7(-5) 4.9 1 1.7(-5) 5.811.3-lasI
- 5.5 1.7(-5) <1(-5) 5.3 1.5 -
esRb <4(-5) <2(-5) <2(-5) 134Cs 1.9 0.1 - 2.7 0.1 - 2.18 1 0.02 - 2.39 i 0.02 - 136Cs 1. 7 i O .1 - 2.010.1- 1.24 0.05 - 1.3110.06-137Cs 3.8 i 0.1 - 5.1 i 0.1 - 4.01 0.06 - 4.37 0.04 - 138Cs 7.3 1.4 - 9.1 i 1.6 - 3.7 0.7(-5)
- 3H ** ** ** **
14C v* ** ** ** 24Na 2.2 1 0.3(-5) 1.2 0.3 -5 2.5 0.2 -5 2.9 i 0.3 -5 31Cr 5.5 1 0.7(-4) 4.2 0.7 -4 3.4 1 0.6 -4 3.4 i 0.8 -4 5 6.6j0.1(-4) 9.8, 0.2 -4 5.7,i 0.1 -4 6.13 0.1 -4 59pe <4(-5) <5(-5) <1(-5) <2(-5) 57Co 5.8 i 0.6 -5 8.2 0.4(-5 6.2 0.5(- ) 58C0 60 2.6 i 0.1 -2 4.1 0.1 -2 6.1 2.24 1 i 0.4(- 0.03 -2) ) 2.3910.03-2) 6.6J0.2-3 1.1J 0.1 -2 5.98,1 0.07 -3) 6.3 { i 0.04 -3) 65 n 2.7j0.4(-4) 2.9 j 0.2(-4) 3.2j 0.1(-4) 2.4j 0.2(-4) 90$r ** ** ** ** r -<4(,5) <4(,5) <2(,5) <2(j) 95Zr - *
<4(-5) <2(-5) <2(-5) 95Nb 5.111.2(-5) 8.8 1.3(-5) 5.0*1 0.6(-5) 3.6* i 0.6(-5) 99g *
- 103Ru 4.5 1.0(-5) 3.710.9(-5) 2.2 0.5(-5) <2(-5) 106Ru 110 mag * * *
- 124Sb 6.8 0.3(a) 9.2 0.4(-4) 7.510.2(-4) 7.7 0.3(.4)
-12sSb 3.9 0.3(-4) 4.8 i 0.3(-4) 3.9 i 0.2(-4) 4.9* i 0.2(-4) -129mTe <3.2(-5) <4.6(-5)
<2(-5) 129T2 * *
- 140Ba <4(-5) 14oto <3.6(-5)(-5) 1.710.4 1.7i0.4(-5) 1.0 2.2(-6) 1.2 0.2(-5)
B-103
~
, _- y - - -
l l TABLEB.23(cont'd) l RADIONUCLIDE CONCENTRATIONS IN RADWASTE EVAPORATOR BOTTOMS 1/9/78; 13:20 1/9/78; 17:33 1/10/78; 09:1; 1/11/78; 12:43 Nuclide (uC1/ml) (uci/ml) (vC1/ml) (vC1/ml) i 1311 1321 9.5*1 0.E(-3) 8.610.1(-3) 9.5*1 0.1(-3) 6.6 0.1 (-3)
<3(-5)
- i 133I 1.210.1(-4) 1.3 0.1(-4) 1.610.1(-4) 1.5 1 0.1(-4)
I 134I <6(-5) 9.8 i 1.9(-5) 1351
- 4.2 1 0.3(-3) i <1(-6) <1.6(-5) <1.7(-5)
I 88Rb <4(-3) <8(-5) <8(-5) 2.5 0.1 - 134Cs 2.44 0.02 - 2.17 0.02 - 2.5 1 0.1 - 2.5 i 0.1 - 136Cs 1.27 i 0.07 - 1.07 i 0.07 - 1.1 1 0.1 - 1.0 1 0.1 - 137Cs 4.41 i 0.04 - 4.00 0.05 - 4.6 0.1 - 4.8 0.1 - 7.1
- 13BCs 1. 3(-4) 1.0 0.2 - 2.4 1 0.5 -
3H ** ** ** 3.210.1(-2) 14C ** ** ** 5.9 i 0.6 - 24Na 3.0 t 0.3 - 4.0 0.3 - 4.6 0. 3 - 3.5 i 0.3 -
- 51Cr 4.9 0.8 - 5.0 0.9 - 3.5 i 0.5 - 7.6 1.1 -
l 54Mn 6.2**i 0.1 - 5.5 0.1 - 6.1 0.1 - ti.2 1 0.1 - l sspe ** ** i 59Fe 6.29 i 0.01 3)
<3(-5) <1(-5)
- 6.0 1 1.1 -
l 57Co 9.2i1.2(-5) 6.4 0.6(-5) 6.0 1 0.7 - 8.6 1 0.5 - 58C0 2.41 1 0.03(-2) 2.2310.04(-2) 2.5 i 0.1 - 2.8 0.1 - l 60Co 6.43 0.05(-3) 5.91 1 0.06(-3) 6.7 1 0.1 - 7.6 i 0.1 - 63Ni ** ** ** 65Zn 2.5 5.26 1 0.01(-3) 0.1(-4) 2.1**i 0.2(-4) 3.6*** 0.3(-4) 4.0 i 0.3(-4) 895r 1.48 90Sr ** ** ** 0.05(-5) j 91Sr *
- 4.98 1 0.05(-6) 91y ** **
<2(-5) l ssZr <2(-5)
- 4.8i0.1(-3))
6.90 0.08 6 3.5 4.2
<2(-5) 1.7 i 0.1(-
l 95Nb 0.7(-5) 0.8(-5) 6.610.8(-5) 3.0 1 0.2(- 99Mo *
- I
<1(-5) lo3Ru 2.7i0.8(-5) 3.0
- 0.7(-5) 4.7i1.1(-5) 9.9*0.9(-5) 106Ru <2(-5) <2(-5)
<3(-5) 2.4 0.9(-4) 110 mag <2(-3) .
124Sb 8.010.2(-4) 7.4 0.2(-4) 8.210.2(-4) 8.210.2(-4) 12sSb 4.3 0.2(-4) 4.8 0.2(-4) 4.4 i 0.2(-4) 5.6 i 0.3(-4) 129mTe * *
<2(-5) <2(-5) 129Te * '
14cBa * * *
- 14ela 1.0 0.2(-5) 1.3i0.3(-5) 1.4
- 0.2(-5) 1.6 -0.3(-5)
B-104
TABLE B.23 (cont'd) RADIONUCLIDE CONCENTRATIONS IN RADWASTE EVAPORATOR BOTTOMS 2/1/78; 16:10 2/5/78; 10:40 2/18/78; 13:30 4/28/78; 10:44
'Nuclide (uC1/mi) (uC1/ml)_ (uci/ml) (uCi/'ll 13 2.9 t 0.1(-2) 4.4 0.1(-2) 2.310.1(-2) 2.5, 0.1(-3) 133I 1.5 1 0.2(-4) 3.4 0.5(-4) 1.9 0.1 - 2.9 0.2(-4) 134I <1(-4)
- 2.2 0.5 -
- l 1as! <4(-3) <8(-5) 1.1 i 0.1 - 9.2 2.7(-5) serb * * *
, <1 (-5) i 134Cs 4.7 i 0.1(-3) 1.20 1 0.01(-2) 5.1 i 0.1 - 2.F 0.1(-3) 13sCs <1.7(-4) <5(-4) 5.4 0.2 - 137Cs 8.1 0.1 -3 2.4 0.1 -2 8.9 0.1 - 4.310.1(-3) 13sCs 2.0 0.3 -4 5.3 2 0.6 -3 *
- 3H ** ** ** **
l l 1%C ** ** ** ** i 24Na- 2.4 0.4 - 4.9 i 0.6 - 4.0 0.2 - 4.320.8(-5) i s1Cr 1.3 0.1 - 3.3 i 0.3 - 7.8 1.2 - <2(-4) 54Mn 2.1 0.1 - 3.8 0.1 - 1.5 0.1 - 4.8 0.2(-4) ssp, ** ** ** ** 59Fe 1.2i0.3(-4) .6 0.5 - 1.1 1 0.2 - <4(-5)' i 5700 4.3 2 0.2(-4) 6.3 0.2 - 1.0 t 0.1 - 4.8 0.9 - seCo 1.80 i'O.04(-1) 2.5 i 0.1 - 3.5 i 0.1 - 9.2 2 0.1 - t 60Co 3.7 0.1 (-2) 6.6**i 0.1 - 1.4 0.1 - 6.6 0.1 - 63Ni ** ** 65 n 4.6 i 0.7(-4) 1.1 30 1(-3) 7.3 0(-5) 3 1 <4(-5) 953p ** ** ** ** S1 r <7(-5) 3.2
<1(-j) 3 0 6(-4) <1(-j) 95Zr 1.9 i 0.2 - 4.3 0.5 - 9.4t1.2-) <4(-5) 95Nb 3.1 2 0.2 - J.2 i 0.4 - 1.6 i 0.1 - 6.2i0.9(-5) 09Mo 2.8 0.1 - 1.7 t 0.1 - 4.220.1-) i 4.7 + 0.6(-5) 10 3Ru 3.1 0.2-4) '
5.0 t 0.3 - 6.110.5-) <2(-5) 106Ru <1(-4)
<l(-4)
<4(-5) <2(-4j 110 mag <4(-3)
- i 124Sb 3.7 1 0.1 - 7.7 0.2(-3) 1.2 0.2(-3) 2.3
- 0.2(-4) l 12sSb 3.6 t 0.1 - 7.7*0.2(-3) 1.5 0.1(-3) 4.8 1 0.4(-4) 129mTe 1.5 i 0.6 - <2(-5) <7(-5) <6 -
.12sTo 2.8i1.2(-3) * <2 -
140Ba <7(-5) <8 - 140La
<8(-5) 1.110.2(-4) *
[ <3(-5) <4(-5) 5.1 0.3(-5)
~o Radionuclide not detected.
! ** Radionuclide not measured. I -- l B-105 l i a y,eam m -yy e ---
TABLE B.24 RADIONUCt.IDE CONCENTRATIONS IN RADWASTE EVAPORATOR DISTILLATE AND BOTTOMS FEED. SHUT OFF'ANO BOTTOMS CONCENTRATING DURING SAMPLE PERIOD Bottoms Activity DistillateActivith 1/10/78; 10:15 1/10/78; 10:07 1/10/78; 10:11 1/10/78; 10:io-1/10/78; 10:05 1/10/78; 10:10 (pC1/ml) (pCi/ml) (pCi/ml)- (pC1/ml) (pCi/ml)' (pCi/ml) Nuclide 2.010.1(-5) 1,82 0.02(-2) 2.15 0.02(-2)- 2.48 0.03(-2) 131I. 1.9 i 0.1(-5) 1.8 i 0.1(-5) 4.5 0.8(-4) 4.1 1 0.6(-4) 138I 1.2 1.0.4(-7) <2(-7)
- 3.710.4(-4) 1.7* i 0.6(-7) <4(-7) '4.8 0.1(-3) 5.810.1(-3) 6.4 i 0.1(-3)
' 13*Cs ~9.2*4.5(-8) 2.7 1 0.3(-4) 3.9 2 0.6(-4) 4.3 1 0.4(-4) -
136Cs . <2 - <4(-7) ~ 1.17 t 0.02(-2)
<2 - *
- 8.6 0.1(-3) 1.03 0.01(-2)
- 137Cs
- 138Cs <2 - <1(-5) <2(-5) 2.1 1.6(-2)
*
- 5.6 1.4(- <1(-4) 9.8 i 1.9 -5) 24Na- <3(-7)' * <3 - 7.2
- 2.2(- 1.2 0.3(-3) 3.1 1 0.6 -
? 51Cr <2(-7)
'3- 1.16 0.03 3) 1.52 1 0.05(-3) 1.8 0.1 -
.y 54Mn 6.7't4.5(-8) 5.4 i 5.3(-8) <
1.6 0.3t-4 2.1 0.3 -
- 57Co <1(-7) <2(-7) <4 - 1.4 i 0.1(- 7.4 2 0.1 -
2.3 0.1(-6) 4.1 2 0.2(-6) 4.8 0.1(- 5.9 i 0.1(-2 i 58Co 5.6 i 0.3(-6) 1.32 0.02 2) 1.62 0.02( 2) 2.08 0.02(-2) ; soCo .1.2 1 0.2(-6) 1.6 0.2(-6) 8.5 i 1.6(-7) 7.2 1.1(-4) 1.0 1 0.1(-3)
'*
- 5.3 0.7(-4) 65Zn <2 - * * <1(-4) <2(-4) <2(-4) l 95Zr <2 - 1.2 0.2(-4) 2.2 0.6(-4) 4.4 0.4(-4) 95Nb <2 - * <3(-7) 8.5 3.4(-5) 2.2 0.6(-4)
* * * <l(-4) 2.0 0.1(-3)
- lo3Ru *
- 1.6 1.9 0.1 - r
<2f- 8.2 0.1(-3} 1.3- 0.1(-3) i 1245b' *
- e 1.3(-4, 1.210.1-12sSb <2f' - <2 - (2(-7) <1(-4) 2.6 i 1.3 - <2(-4) i 140Ba <2 -
- 2.3 0.9(-5) <6(-5) 6.6 1.3(-5) i 14cLa <1 - <1 -
- Radionuclide not detected. ,
_E_._.. _ . _ . _ - . _ _ _ _ . _ _ ____, . - . _ _ . _ . . . _ - - . -_ _ _ _ __ i
TABLE B.25 RADIONUCLIDE CONCENTRATIONS IN INLET AND OUTLET FOR RADWASTE CONDENSATE D 1/6/78 1/10/78 (pCi/ml) 1/11/78 Nuclide (pCi/ml) (pCi/ml) Inlet Outlet Inlet Outlet Inlet Outlet 131I 2.6* 1 0.1(-5) 133g ' 4.3* 1 0.1(-6) 1.810.1(-5) 4.9* i 0.1(-6) 1.14 i 0.02(-5
- 2.4 i 0.1(-6) 1841 2.9 i 0.7(-6) * *
- 8.6* 1 7.0(-8) )<2(-7)
- 134Cs 1.710.6(-7) 7.5 1 3.8(-7) <2(-7) 5.2 i 0.5(-7) 1.4 1 0.5(-7) 137Cs <2(-7) 1.3 i 1.6(-6) 6.2 6.0(-8) 6.8 1 0.7(-7) 1.03 i 0.10(-6) <2(-7) 1.0 0.1(-6) 3H ** ** ** **
to 3.210.1(-2) 3.110.1(-2) 2, 14C ** ** ** ** f3 54Mn 3.510.5(-7) <1(-7) 5.0**i 1.1(-7) 5.8i0.6(-7) 5.6 1 0.6(-7) sspe ** ,* <1(-7)
** 5.3**i 3.6(-8) <2(-7) 1.22 0.04(-5) 57Co 4.312.0(-8) <8(-8) 1.7 i 0.4(-7 <8(-8) *
- seCo 1.87 i 0.06(-5) 5.2 1 3.1(-8) 2.510.1(-5 2.3 i 2.5(-8) 4.8 0.3(-6) <2(-7) 6OCo 4.9**i 0.3(-6) <1(-6)
** 5.2**1 0.3(-6 <1(-7) 9.8 1.3(-7) <2(-7) 6 3Ni **
893p ** ** ** ** <7(-8) 5 1(-7) 90Sr ** ** ** ** 6 2(.8) <](.8) sly ** ** ** ** 2.510.2(-8) 1.1 i 0.2(-8) lo3Ru <6(-8) 1.36 i 0.07(-8) 5.1 1 2.6(-8) <1 (-7) * * *
- 4 I
TABLE B.25 (cont'd) RADIONUCLIDE CONCENTRATIONS IN INLET AND OUTLET FOR RADWASTE CONDENSATE DEMINERALIZER t t 2/1/78 4/28/78 (uCi/ml) (uCi/ml) Nuclide c Inlet Outlet Inlet Outlet 1311 3.0* 1 0.1(-5) 1.80 i 0.03(-6) 2.7 1 0.1(-5) 1.09 0.03(-5) 113I
- 2.4 1 0.2(-6) 9.5 i 1.0(-7) 1 sI *
- 2.3 0.2(-7) <5(-8) 134Cs <2(-7) 5.8 0.2(-7) 9.6 0.7(-8) 4.211.3(-7) 137Cs <3(-7) 1.3010.04(-6) 9.2 1 7.7(-8) 8.4 1 1.0(-7) 54Mn <2(-7) 6.8 i 3.5(-9) 3.6 i 1.6 - 1.9 0.6(-7) 5.6 i 0.6(-8) 1.1 1 0.1 - 5.7 0.7(-8) seCo 1.0 0.6(-6) 1.0 0.2(-7) m 60Co 6.3 4.6(-7) 5.3 1.0(-8) 2.3 0.2 -
1 95Zr *
- 4.2 0.9 - <1(-8) 8 110 mag *
- 7.2 1.0 - <2(-8)
*
- 4.0 .0- 1.2 0.5-8) 124Sb 5.0 0.1 -8) 12sSb
*
- 1.4 0.2 -
- Radionuclide not detected.
** Radionuclide not measured.
t Includes data obtained us'ng resin concentration techniques.
TABLE B 26 RADIONUCLIDE CONCENTRATIONS IN WASTE HOLDUP TANK NO. 2 4/25/78; 11:30 4/26/78; 14:34 4/29/78; 11:43 5/2/78; 14:24 5/10/78; 10:44 Nuclide (uti/ml) (uC1/ml) (uCi/ml) (uCi/ml) (uCi/ml) 131I 1.85 1 0.01(-4) 2.79 i 0.04(-4) 1321
- 1.92 1 0.07.(-4) 4.C i 0.1(- 7.3 0.4(-5)
<1 (-5) 6.2 0.8(- <2(-6) 133I 1 34I -
1.2420.03(-5) 5.2 1 0.6(-6) 5.1310.06(-5 ) 1.8 0.2(-5) 1.3110.02 4) 6.7 1 0.3(-5 1.5810.06(-5)
- 4.5 1.7(-6 1351 2.6 i 0.1(-5) 4.0
<8(-6) 0.1(-5)' 1.2 2 0.1(-5 134Cs- 9.15 0.09(-5 1.96 1 0.02(-4) 1.74 0.03(-4) 8.8 0.2(-5) 1.28 1 0.06(-4) 136Cs 137 2.5 i 0.3(-6) ) 3.8 i 0.4(- <5(-6) 9.3 1 4.2(-
1.60 1 0.02(-4) 3.63 i 0.05 4) 3.10 1 0.03(-4) 1.4 1.62 i 0.6(-6) 0.04(-4 ) 1.94 1 0.09 4) 139cs .* * * *
- 24Na 1.110.1(-6) 8.9 i 0.4 - 3.3 1 0.4(-6) 1.3 0.1-5) 8.5 i 0.5 -
7 7.8 i 0.3 - 51Cr 2.1 0.4(-5) 1.3 i 0.1(-4) 2.3 i 0.3 -5) 1.1 1 0.3 - 54Mn 1.84 0.05(-5 3.2 i 0.1 - 1.01 t 0.02( 4) 2.6 i 0.2 -5 2.0 1 0.1 - 8 59Fe 2.6 0.6(-6)) 8.2 0.8 - 3.1 0.4(-5 4.0 0.7 -6 <2(-6) 57Co 2.7 0. 3 - 9.1 1.0(-6 1.5 1 0.2(-6 8.4i2.5(-7) saco 8.3 1i 0.01(-4 2.57 1.2(-6) ) 5.2 1 0.1 - 1.35 0.01( 3) 2.4 0.4(-6 2.4 i 0.1(-4) 4 60C0 2.40 0.02(-4) 3.87t0.09(-4) 1.20 1 0.08(-3) 2.7i0.1(-4) 2.46 1 0.08(-4) 65Zn <2(-6) *
- 1.7 0.7(-6) <2(-6) 95Zr 4.6 1 0.5 - 1.4 1 0.1 - 4.6 i 0.3 -5 6.0 0.6(-6) 2.6 0.5 -
9sNb 8.1 1 0.4 - 2.3 0.2 - 7.7 1 0.2 -5 1.00 0.04(-5 6.6 0.5 - 99Mo 1.5 1 0.2 - 7.9 0.2 - 2.1 0.6 -6 3.4 0.2(-5) ) 1.7 0.2 - 103Ru 1.110.3-6) 5.0 1 0.4 - 6.1 1 1.1 - 1.9 i 0.4(-6) <9(-7) lo6Ru <1(-5) 1.5 0.4 - 4.7 i 2.0 - <6(-6) <1(-5) i IlomAg 6.8 1 0.4(-6)
- 3.4 0.2 - 5.9 i 0.5(-6 1.7 i 0.6(-6) 124Sb 1.2310.05(-5) 1.89 i 0.07(-5) 2.8 i 0.2 - 1.4 0.1(-5 5.0 0.5(-6) 12sSb 2.2 0.1 - 3.1 0.1(-5) 6.1
- 0.7 - 3.0 0.1 - 1.5 0.1 -
140Ba
- 8.3 1.2 - 4.2 1.7 -
8.3 i 1.0 - <2(-5) 1%oLa 4.3 0.2 - 5.1 i 0.3(-6) 5.6i0.6(-6) 2.2 0.3 - 2.0 1 0.2 -
3: TABLE B.26 (cont'd). RADIONUCLIDE CONCENTRATIONS IN WASTE HOLDUP TANK NO. 2 c 5/16/78;-16:11 5/16/78; 17:55 5/20/78; 11:10 ^ 5/23/78; 12:50 5/25/78; 08:58' Nuclide' -(vC1/ml) ' (uci/ml) (vC1/ml) -(uC1/ml) (uC1/ml) 7.6 2.14 1 0.02(-4)~ -1 .31 i 0 .06(-4) 1311 1.59 i O.03(-4) 1.46' 0.03(-4) 0.5(-5) 1321 ' 8.1 0.7(-6 -6.6 1 0.4(-6) <1.5(-6) <1.8(-6) . <5.6(-6) 133 .1.41 i O.01()4)
' i 0.02(-4)
-1.45 8.8 0.6(-5) 3.73 1 0.08(-5)_ 1.5 i 0.1(-5) :
- 1351 6.2.t0l(-5) 6.38. 0.09(-5) 8.3 i 1.1(-6) 7.6 i 1.4(-6) <4.2(-6) ;
I 134Cs .9.6 i 0.2(-5) -9.1 ' O.2(-5) 1.09' O.04(-4) 1.98 i 0.02(-4) 1.50 0.03(-4) 136C$ 3.9 0.7(-6) 2.6 i 0.4(-6) <7.6(-7) 4.5 i 0.9(-6) 7.2 2 0.05(-4 0.7(-6) ) 2.6 1 0.1 - ) 137Cs 1.58 i 0.03(-4) 1.48 i 0.02(-4) 1.78 i 0.08 -4) 3.10 + 13eCs- <2.5(-6) <1.4(-6) <2.3(-6) <3.2(-) <5(-6) 139Cs <2.8(-3) <2.4(-4) <3.0(-3) <4.9(- ) <9(-3) 24Na .1.54 1 0.06(-5) 1.61 i 0.04 5) 6.9_1 0.5 - 3.4 i 0.4( _1.4 i 0.3 - 51Cr 1.7 i 0.3(-5) 1.5 2 0.1 - 1.6 i 0.3 - 2.0 0.6(-5 6.0 0.8 - 54Mn 3.4 1 0.3(-5) 3.6 0.3 - 2.4 i 0.2 - 2.13 0.09(5) 7.3 i 0.3 - 1 '7.4 1 3.6 - 2.6 1 0.9 - <1.9(-6)- 6.0 i 2.3 -
;; 59Fe <1.3(-6): 7.7 i 0.9 -
57C0 1.8 0.2(-6 1.36 1 0.09 - '2.0 0.3 - , 58C0 2.21 i 0.03 4) 2.02 0.03 - 2.69 0.08 4) 1.64 i 0.07 1.3 0.3(-6)
-4 ) 7.8 0.3 60Co 3.4 0.1 I'- 3.62 1 0.08 2.94 0.07 -4) 3.16 2 0.03 4) 1.6 1 0.06 )
652n 1.9 0.7f- 2.4 i 0.5(- <2.1(-6) 2.0 1 1.0 - 1.3 i 0.3q ssZr 4.910.5(- 5.8 0.4(- 5.3 1 0.9 - 3.5 1.0 - 4.0 i 0.3I - ;
- j. 1.06 1 0.06 5) 8.2 i 0.5 - 5.7 1 0.5 - 8.6 0.2J,-
95Nb 1.11
- 0.04( 5) 99Mo- 2.4 1.0.2(-6)- 2.8 1 0.1(-6 1.4 i 0.3 - 1.36i0.03(5) 5.9 0.5(- j 103Ru 1.5 i 0.3 - 1.5 i 0.2 - <7.6(-7) <9.1(-7) 4.7 1 0.9Il- ;
1.2 0.3 - 9.7 1 2.1 - <7.1(-6) 1.5 - 106Ru 110nng 4.311.2- 3.6 0.3 - 3.1 0.9-6) <1.2(-5)(-6) 2.6 1 0.8 9.6 0.2(I,- 0.3 _124Sb 4.7 0.5-6) 7.510.6(-6) 1.39 0.09(-5) , 12sSb 5.6 1.55 ii 0.4 0.08(-5 6) ) 5.6 1.51 1 0.2(-6) ) 1.3 i 0.2 -5) i 0.05(-5 1.5 0.2(-5) 5.2 t 0.3(-5) 139Ba <4.9(-6)
<1.8(-6)(-6) <3.6(-6)(-6) 3.4 i 1.4 2.7 1 0.3(-5) 1.8
.<2.8(-6)(-6' 1
140Ba 6.2 i 1.C 5.1 0.9 , 140La 2.8 0.2(-6 2.6 1 0.1(-6) 2.0 0.3(-6) 1.71 1 0.06(-5) 1.31 'O.4(~-5)) 0.05(-5
- Radionuclide rot detected.
w
TABLE B.27 RADIONUCLIDE CONCENTRATIONS IN RADWASTE BUILDING MONITOR TANK A 4/25/78; 10:30 4/26/78; 13:30 4/29/78; 11:28 5/20/78; 11:10 5/23/78; 13:55 5/25/78; 15:45 Nuclide (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) (uCi/ml)
~
1 * -
-7 *
<3.7(- <3 -7
"'I <"-"
- 2{3 :2{2 $$ :1:t$ $$
134Cs 1.1 0.5(-7) 2.3 0.9(-7) 6.3 1.3(-7) 1.6 i 0.6(-7) 1.9 0.8(-7) 136Cs <2(-7
<1(-7) <2(-7) <2(-6) <2(-7) <l.2(-7) <9(-8 137Cs 7.0 3.8(-8) 7.1 1.0(-7) 1.0 0.1(-6) 9.8 4.5(-8) 1.5 1.0(-7) <3 -
13Ts * *
<3 -7 *
<1.8 -6 4-139Cs * *
<1 -6 *
<6.0 -3 <3 -
m t 54Mn 23 4.6 0.6(-7)
<5 -
7.8 1 0.8(-7)
$$ 0.1(-7) 2.9
- !:S{m 3.2 1 0.7(-7)
- 8t*
4.9 1.5(-7) 3 59Fe <2(-7) <4 - <4(-7) *
<3.0(-7) <2(-7) a 57Co 3.8 1.8(-8) <3 - 6.2 2.9(-8) <1(-7) <9.9(-8) <6. 3 (-8) seCo 8.2 0.6(-7) 3.4 0.9(-7) 1.1 0.1(-6) 4.3 0.5(-7) 5.8 i 1.0(-7) 9.511.0(-7) 60Co <2(-7) 1.9 0.2(-6) 3.9 0.2(-6) 1.5 0.1(-6) 2.4 0 2(-6) 2.6 0.1(-6) ss2n * * *
<4 - <4.2 - <3 -
ssZr <2(-7) <4 -7) <3 - <2(-7) <1.6 - <2 - 95Nb 1.6 0.3(-7) <4 -7) <2 - *
<9.3 - <1 -
99Mo <6(-7) <2 -7 <1 - *
<9.2(-8 <9 -
103Ru <1(-7) <2(-7 <2 - *
<9.4 - <8(-8 106Ru * <2 -
<4(-6
<l(-7)
<8. 3 - 1.2 0.3(-6) 110mA9 1.1 0.4(-7) <3- <l.5 - <1(-7) 124Sb <1(-7) <4(-7) <4 - <l(-7) <l.5 - 1.9i0.4(-7) 12sSb <4(-5)
<8(-7) 6.312.4(-7) 5.4 0.8(-7) 3.1113(-7) 6.8 1.3(-7) 13983 * <6 - <2-7) <l.2 - <l -
* <8 -
140Ba <3 -7) <1 -7) <3.3 - <3 - 140La <6 -7) <2(-7) <1 - <1 -7) <8.0 - <5 -
- Radionuclide not detected.
TABLE B.28 RADIONUCLIDE CONCEtlTRATIONS IN RADWASTE BUILDING MONITOR TANKS B AND C Tank B Tank C 5/2/78; 14:14 5/8/78; 17:43 5/16/78; 21:35 5/7/78; 15:13 5/10/78; 15:44- 5/25/78; 10:20 (uCi/ml) (uCi/ml) (uC1/ml) (pCi/ml) (uCf/ml) Nuclide. (uCi/ml) l'!! 8 3: 't*o2(-7) :!: ${:81 2{3l 1331 134I 3{:<7-8l *
<5 -
<1 -
<5 -
<3 -
<1 -
<5 -
<6(-7)
<2 -7
<'7 <2 - <3 - <1(-7) <5 (-7) 13sI <3(-7) l':!: 3{ $ <e'- :4:281 :W.H 3: 33 137CS 2.7 i 0.7(-.7) <8 - /) " <1(-7)
<' 2(-6 )
5.2 3.6(-8) <7 - <4 -7
<2 -8
- : 2:J 20 3-i 54Mn 3 :!!:*)
4.2 1 0.4(-7)
$ :"]
7.4 i 0.4 (-7) :f:"))
<1 -7
$ :"1 6.6 0.4 (-7)
- "{:"1 6.8 0.3 (-7) 7.9
$ 1.3 (-7)
- $:F $$ 3l2 $$ $$ jf2
- !: i:18:if$ *::8:?{$ i::?:sf$ ":8: 8:!!$ :8:8:'{$
- ':i:!!$
<4(-7 65Zn <2 - <2(-7 <2(-7) 1.3 .0.6(-7)
<1(-7) <2 -7) <1(-7) <3(-7 ssZr <1 - <8(-8 <2 -
<6,- <5(-8 <1(-) <2 - <1(-7) 95Nb <2 -
4.9 2.1(-8) <6 - <6(-8) 99Mo <6(
<4.4 -8) )
<5(- <? - <71 -8) <2 -
103Ru <61'--
<4 - 2.4 0.9(-6)
<6 <4 - <4.1 -7) <8(-7 <8d-8) 106Ru
<7 - 3.3 0.5(-7) <2(-7) 7.1 2.3(-8) 8.8 i 1.7(-7) 110 mag <9 - 1.4 0.3(-7) <1(-7) 12*Sb <2 - .1. 3 0.5(-7) 4.410.7(-8) 9.0 2.4(-7) 4.6 1.2(-7) <1.2(-7) 3.2 i 1.1 (-7) 7.8 i 0.7(-8) 5.4 0.5(-7) 12sSb <2 - <5 - <5(-6) <2 -
i 139Ba <2- <1 - <1'.0 - .
<6 -
<2 - <2 - <1.6 - <3 - 1.4 0.8(-7) 140Ba
<5.7 - <6 - <8(-8) <8 -
140La <6 - <4 -
- Radionuclide not detected. -
.1ABLE B 29 RADIONUCLIDE CONCENTRATIONS IN AUXILIARY BUILDING MONITOR TANKS Tank A Tank 8 4/29/78; 11:24 5/16/78; 14:48 5/18/78; 11:16 5/02/78; 11:17 5/8/78; 18:13 5/10/78; 14:00 Nuclide- (uci/ml) (uti/ml) (uCi/ml) (uci/ml) (uci/ml) (uci/ml) 181I- 1.70 0.02(-5) '1.1 0.3(-7) <9.7(-8) 5.0 0.1(-5) 133I 5.9 1 0.1(-7) 1.5 0.3(-7) 1.4 1 0.6(-7) <8(-8) <1.0(-7) <5(-7) <2(-8) <1(-7) 134Cs. 5.7 i 5.0(-8) ' 9.8 1 0.5(-9) '1.3 1 0.3(-8) 8.1 '4.5(-8) <5(-8) <1(-7) 136Cs <7(.8)' .
<6(-8) <9(-8) <6(-8) <2(-8) <1(-7) 137CS 2.3 0.5(-7) <1(-6) 1.5i1.7(-7) <1(-7) <4(-8) <1(-7) 5 1Cr~
<8(-7) <5(-7) <6(-7) <4(-6) 7.8 0.8(-7) <1(-7)'
54Mn <6(-8). 8.2 1 4.1(-8 3.8 1.4 -9) <7(-8) 2.1't 0.2(-7) <1(-7) 56Co 1.0 0.4(-7) ' 9.4 1 0.6(-7 7.0 0.5-8) 1.6 1 0.5(-7) 1.62 1 0.07(-6) 9.1 1 4.2(-8)
. -60C0 <2(-7 8.9 i 1.2(-7 8.8 1 0.8 -8) 2.5 1.3(-7) 1.8 i 0.1(-6 ' <3(-7)
.' 65Zn <2 - <2.8(-7) <1 9(-7) <1 - <7(-8)
- g' 95Zr <2- 1.9i0.7(-7) <1 - <9 - 2.0 1 0.2 - <1(-7) 95Nb <8 - 2.3 1 0.7(-7) <9 - <6 - 3.2 1 0.2 - <1(-7) 99Mo <1 - <6(-8 <1 - <4 - 3.0 1.2 - 8.7i3.7(-8) 103Ru <7 - <6(-8 <8 - <5 - 2.8 1.1 - <1(-7)'
106Ru <8 - <7(-7 <7 - <4 - 3.4 1.6 - <1(-7) 110 mag 1.4 i 0.3(-7) 1.2 0.4(-7) <1 - <1 - 6.0 1 1.0 -
- 12sSb <2(-7) 6.2 1 1.3(-7) <2 - 1.5 0.7(-7) 1.8 1 0.4 - <1(-7)
- Radionuclide not detected.
TABLE B.30 CONCENTRATIONS OF BETA-CNLY-EMITTING RADIONUCLIDES IN HASTE HOLDUP AND MONITOR TANKS Waste Holdup Tank #1 Waste Holdup Tank #2 Nucli de - 16:30,12/5/77 12:29,1/11/78 14:35,4/26/78 11:40,4/29/78 16:11,5/16/78 3H 6.85 1 0.03 - 2.9 1 0.1(-2) 6.3 0.2 - 4.6 1 0.1 - 3.2 1 0.1(-2 14C 3.69 1 0.04 - 1.5*** 0.8(-6) 3.5 1 0.4 - 1.5 1 0.2 - 1.2 i 0.1(-5 32P 2.17 i 0.07 - 1.9 1 0.1 - 2.7 1 0.4 - 1.20 1 0.05 ssFe 5.49 0.02 - 4.06 0.02-4) 7.51 0.02(-4) 5.22 1 0.02( 4 9.53 0.02 - 63Ni 7.34 0.03-5) 1.66 1 0.01 4) 1.24 1 0.01 - 2.29 i 0.01 - 9.96 1 0.07 - 89Sr 1.22 2 0.06 -5) 2.76 1 0.08 -6) 3.76 1 0.04 - 1.80 i 0.02 - 4.7 0.4(- ) 90Sr. 3.3 0.9(-7) 3.22 0.08 - 2.41 i 0.03 - 5.58i0.08(-7) 91Y 4.9 i 0.6(-7) 6.2 1.04 i 0.1(-7) 0.02(-6 ) 6.4 i 0.1(-7) 7.4 0.1(-7) 2.6
- 0.6(-7)
T' Monitor Tank B in Aux. Bldg. Monitor Tank A in Auxiliary Building g 11:25, 4/_29/78 1A:48.5/16/78 11:16,_5/18/78 14:00.5/10/78 3H 1.00 1 0.03(-1) 1.59 i 0.05(-2) 8.6 1 0.3(-2) 7.5 0.2(-3) li.C 2.4 1 0.2(-7) 4.1 t 'J.4(-6) 4.3 0.4(-6) 5.1 1 0.5(-7) 32P <9(-7) 4 1 2(-7) <4(-7) <6(-7) 55Fe 2.1 1 0.1(-6) 1.11 0.01(-5) 2.4 i 0.2 -6) 1.9 0.1(-6) 63Ni 3.7 1 0.4(-7) 7.8 0.7(-7) 6.5 0.6 - 9.1 0.7(-7) 89Sr <3(-9) <4(-9) 2.7 i 0.5 - <4(-9) 90Sr 2.8i0.9(-9) 8 i 2 -9) 1.2 0.2 - 1.3 0.2(-8) 91Y 2.7 0.2(-8) 8 1 1 -9) 2.2 1 0.9 - 2.4 i 0.2(-8)
** Radionuclide not measured.
TABLE B 30 (cont'd) CONCENTP.ATIONS OF BETA-0NLY-EMITTING MDI0NUCLIDES IN WASTE HOLDUP AND MONITOR Monitor Tank A in Radwaste Building Monitor Tank B in Radwaste Building Nuclide 13:30,4/25/78 10:32,4/25/78 _11:28,4/29/78 14:14,5/2/78 21:35,5/16/78 3H 4.5 0.1 -2) 4.0 0.1 -2) 4.9 14C 1.1 1 0.1 -6) 2.9 0.2(-2) 4.2 i 0.l(-2) 3.3 0.l(-2) 0.3 -7) 2.810.3(-7) 8.2 0.8(-7) 6.310.6(-6) 32P 3.1 0.4 -7 1.1 1 0.3 -7) <1.5(-6) 55Fe 6.5 0.8(-7 6.0 0.2(-6) <9(-7) <4(-7) 63Ni 9.7 1 0.3(-6) 4.4 0.2(-6) 1.22 1 0.03(-5) 1.8 1 0.4(-7 5.8 0.2(-6) 1.76 0.08(-6 8.5 0.5(-7) 89Sr <4(-9) 1.03 1 0.02(-5) 90Sr 8 1 2(-9)
<9(-9) 1.5 3.810.6(-8)) <6(-9) <7(-9) 91Y 0.3(-8) 3.6 1 0.2(-8) 2.8_0.2(-8) 3.610.4(-8)
<2(-9) <3(-9) 5 1 2(-9) 412(-9) 2.1 0.2(-9)
Monitor Tank C in Radwaste Building 15:13,5/7/78 15:44,5/10/78 3H 2.98 0.09(-3) 4.0 14C 0.l(-2) 5.5 1 0.6(-7) 8.4 1 0.8(-7) 32P <9(-7) sspe <5(-7) 1,14 1 0.03(-5) 1.28 i 0.03(-5) 63Ni 1.94 0.08(-6) 89Sr
.2.6 i 0.l(-6) 90Sr 1.310.6(-8) <5(-9) 3.9 0.3(-8) 3.8 i 0.2 -8) 91Y 4 1(-9) 1.1 0.2 -8)
)
TABLE B.31 UNIT #3 SPENT FUEL PIT DEMINERALIZER INLET AND OUTLET RADIONUCLIDE CONCENTRATIONS Inlet II3 Outlet [2] InletIll Outlet [2] Inlet [3] Outiet I#3 Nuclide (vC1/ml) (pCi/ml) (pCi/ml) (pCi/ml) (pC1/ml) (pCi/ml) 131I 5.3 0.7(-8) 2.5 1 0.l(-8) 5.3 1.0(-8) 1.8i0.4(-8) 4.7 0.3(-6) 1.22 1 0.09(-10) 134Cs 1.1 1 0.l(-6) 3.7 t 0.9(-9) 1.3 1 0.l(-6) 7.0 1.4(-9) 5.2 0.2(-5) 6.9 i 0.2(-10) 137Cs 2.5 0.2(-6) 8.6 1 0.9(-9) 2.2 0.l(-6) 1.8 0.2(-P) 7.0 0.3(-5) 9.9 0.2(-10) 51Cr 1.29 i 0.05(-5) 7.6 0.2 - 3.3 0.1 - 7.7 0.3 - 2.9 0.3 - 3.6
- 0.1(-9) 54Mn 2.1 0.2(-6) 3.6 0.2 - 2.1 1 0.1 - 4.1 1 0.3 - 2.8 0.1 - 2.04 i 0.03(-10 59 Fe 1.7 0.3(-7) 2.5 0.2 - 5.5
- 1.5 - 3.8 0.9 - 5.5 0.7 - 1.66 0.07(-10 57Co 2.15 0.04( 6) 1.13 0.06 - 2.11 2 0.03(-6) 1.5 i 0.2 - 6.5 0.4 - 1.62 1 0.05(-10 5800 5.1 0.1 -4 3.14 0.04 - 5.8 0.2 - 4.410.1-$ 1.6 0.1 - 8.8 i 0.1(-9)
?' soCo 5.5 1 0.1 -4 1.05 1 0.01 - 5.7 0.2 - 1.60 0.03(-6) 9.7 0.4 - 2.41 1 0.07(-9) d 6 5Zn 3.6 1 0.6 -6 1.2 0.3(-8) 1.5 0.4 - *
- 3.7 0.3 -11) m ssZr 1.2 i 0.1(-7 7.7 i 0.2(-8) 8.2 2.3 - 7.9 0.6(-8) 4.310.1-10) 95Nb 9.3 0.9(-7 1.33 i 0.04(-7) 8.0 0.6 - 1.38 0.06(-7) 1.32 10.8(-6) 7.1 0.07(-5 ) 3.9 0.1 -10) 103Ru 1.08 0.01( 6) 5.2 0.1(-8) 9.5 0.6 - 5.5 1 0.4(-8) 6.4 0.8(-6) 5.11 1 0.03(-10) 106Ru 1.2 0.1(-6) 6.4 1.0(-8) 1.310.1- 7.3 1.7(-8)
- 8.3 1.3(-10) 110 mag S.1 0.1(-7) 9.1 0.9(-9) 6. 7 i 1. 0 -7', 1.6 0.3(-8) 6.2 0.6(-6) 7.1 0.2(-11) 124Sb 5.3*1 0.2(-6) 1.64 0.05(-6) 5.4 0.1 -6) 1.76 0.03 - 3.3 0.3(-5) 6.2 0.2(-10) 12sSb
- 3.16 0.04(-6) 1.19 0.05 - 6.0 0.3(-5) 2.79
- 0.04(-9) 129mie 3.3 0.2(-6) 1.66 0.04(-6) 3.4 0.5(-6) 1.98 0.04 - *
- 9.8 1.03
- 1.5 *
- 129Te 1.8(-6 0.10(-6 0.6 -6) 141Ce 5.2 0.5(-7 3.5 0.l(-8) ) 5.8i0.7(-7) 2.5 0.2 -8) 1.8 0.5(-6)
- 144Ce 3.7 0.5(-6 1.47 0.05(-7) 1.9 0.3(-6) 9.3 0.5-8) 7.3 1 2.7(-6) 1.1 0.3(-10)
- Radionuclide not detected
[1] 450 ml sample [2] 100 liter sample collec%d on ion exchange resin l [3] Average value of dt,Aicate samples of 450 m1 which were collected at the beginning, middle and end of 24 sampling period. [4] 215 liter, sample collected on ion exchange resin
TABLE B.31 (cont'd) UNIT #3 SPENT FUEL PIT DEMINERALIZER INLET AND OUTLET RADIONUCLIDE CONCENTRATIONS 4/15/78 4/16/78 Inlet EU Outlet [5] Inlet [5] Outlet [5] Nuclide (uCi/ml) (uC1/ml) (uC1/ml) (uCi/ml) 1311 * * *
- 134Cs 1.10 0.08(-5) 1.16 1 0.02(-5 1.07 i 0.05 - 1.22i0.07-5) 137Cs 1.34 0.08(-5) 1.57 i 0.04(-5 1.30 1 0.04 - 1.65 1 0.04 -5) 51Cr * * *
- 54Mn 1.5* i 0.3(-6) 4.4 0.8(-7) 1.9 0.3(-6) 3.8 0.6(-7) 59Fe * *
- 57Co * *
- 58C0 1.810.2(-
2.74 t 0.04 -4) ) 5.8 0.2(-6) 2.49 i 0.03(-41 5.810.3{-6)
=
,L 60C0 3.86 1 0.09 -4) 6.6 1 0.2(-6) 3.55 0.09(-4} 6.1* t 0.4 t-6) ssZn O 95Zr * * *
- 95Nb 1.9 0.3(-6) 5.9*1 1.8(-7) 9.6 3.3(-7) 8.3 1.1(-7) 103Ru * *
- 106Ru * * *
- 110 mag * * *
- 124Sb 7.810.5(-6) 7.0 0.4(-6) 8.0 0.4(-6) 7.5 1 0.4(-6) 12sSb 9.1* i 0.7(-6) 8.9* i 0.5(-6) 1.1*1 0.1(-5) 8.7* - 0.3(-6) 129mTe 129Te * * *
- 141Ce * * *
- 144Ce * * * *
- Radionuclide not detected
[5] 450 ml grab sample. a
TABLE. B.32 - BETA-ONLY-EMITTING RADIONUCLIDES FOR SPENT FUEL PIT AND ASSOCIATED WATEPS (vCi/ml) Radionuclide Sample Date 3H l C 91Y , 89Sr Inlet SFP Demin. 11/21/77 2.29 t 0.07(-3) 1.6 0.32(-8) 3.2 1 0.2(-7) 1.86 0.05(-5) j. Outlet SFP Demin 11/21/77 2.34 0.07'-3) 6.4 z 0.8(-8) 1 1 1(-8) <2(-8) / Unit #3 RWST ' 11/21/77 4.3 0.1(-3) 9.1 1 0.9(-7) 1.1 0.2(-7) <7(-8) v Unit #3 SFP 12/30/77 2.98 0.09(-3) 3.6 0.5(-8) 1.8 0.1(-7) 1.2 0.1(-6) and Transfer danal (Derr.in. Inlet) Unit #3 SFP 12/30/77 3.06 0.09(-3) 4.4 1 0.7(-8) <8(-9) <1(-8) and Transfer Canal (Demin. Outlet) Unit #3 RWST 12/30/77 4.4 0.1(-3) 9.3 0 4(-7) 1.6 0.8(-8) <2(-81 Unit #3 SFP 1/3/78 2.77 i 0.08(-3) 2.2 0.4(-8) 1.70 1 0.09(-74 1.2 0.1(-6)
$ (Dip Sample) co Unit #3 RWST 1/3/78 3.7 0.1(-3) 1.0 0.1(-6) <7(-9) <2(-8)
Unit #3 Reactor 1/3/78 5.1 1 0.2(-3) 2.7 1 0.3(-7) 2.06 0.02(-5) 8.1 1 0.2(-6) Cavity (Dip Sample) Unit #3 SFP 1/11/78 3.2i0.1(-3) 7.7 i 0.r;(-8) 3.47 0.05(-6) 2.3i0.2(-6) (Demin Inlet) Unit #3 RWST 1/11/78 3.7 0.1 (-3) 9.1 0.9(-7) 1.1 0.6(-8) <2(-8) Unit #3 Reactor 1/11/78 4.410.1(-3) 8.5 0.9(-8) 9.9 0.1(-6) 1.2610.02(-5) Cavity (Dip Sample) 3.3 t 0.1(-3) 5.320.8(-3) 2.62 0.03(-6) 8.3 0.2(-6) Unit #3 SFP 1/25/78 (Demin Inlet) Unit #3 SFP 1/25/78 3.16 0.09(-3) 5.5 0.8(-8) 2.0 0.1(-7) <1(-8) (Demin Outlet)
. . _ . _ _ __ __ . ~. -,
TABLE B.32 (cont'd) , BETA-ONLY-EMITTING RADIONUCLIDES FOR SPENT FUEL PIT AND ASSOCIATED WATERS (pCi/ml) Sample Da te 90Sr sspe c agg Inlet SFP Demin. 11/21/77 2.92 0.03(-6) 5.20 i 0.07(-5) 9.9 i 0.1(-4) ,. 3 Outlet SFP Demin. 11/21/77 9 2(-9) 4.2 0.6(-6) Unit #3.RWST - 11/21/77 8.5 1 0.9(-8) 1.04
<6(-7) )\ w 0.01(-4) 1.0 0.9(-61 < %,
Unit #3 SFP 12/30/77 1.02 i 0.02(-6) and Transfer Canal 4.24 1 0.07(-5) 2.76 0.08( '4),[' 4 4,2 (Demin. Inlet)- *".. Unit #3 SFP b 12/30/77 1.0 0.2(-8) 8.320.3(-6) <4(-8) and Transfer Canal (Demin. Outlet) { . Unit #3 RWST 12/30/77 2.620.2(-8) 1.09 0.01(-4) 2.1 0.6(-6)
$ Unit #3 SFP 1/3/78 1.82 0.02(-6) 2.4310.08(-5)
(Dip Sample) 4.5i0.1(-4) Unit #3 RWST 1/3/78 6.5 0.4(-8) 1.5510.02(-4) 1 1(-6) Unit #3 Reactor 1/3/78 8.0t0.1(-7) 4.67i0.01(-3) 1.200 Cavity (Dip Sample) 0.003(-3) Unit #3 SFP 1/11/78 2.15 0.03(-6)-
' 9.1 1 0.1(-5) 5.1 1 0.1(-4)
(Demin . Inlet) Unit #3 RWST 1/11/78 4.8i0.4(-8) 9.9 0.1(-5) 9 3(-7) Unit #3 Reactor 1/11/78 2.53 0.03(-6) 5.21 0.03(-4) 7.410.3(-4) Cavity (Dip Sample) Unit #3 SFP 1/25/78 2.3320.02(-6) 2.96 2 0.04(-5) 7.63 0.02(-4)
'(DeminInlet)
Unit #3 SFF 1/25/78 4.0 0.3(-8) 2.18 0.09(-5) 6 2(-6) (Demin Outlet) 4 9 4
i l l-TABLE B.33 UNIT #3 SPENT FLEL PIT AND ASSOCIATED WATER SAMPLES I i RWST [2] SFP [3] SFP [4] [1] 12/30/77; 11:30 12/30/77; 11:50 12/30/77; 13:20 12/30/77; 13:15 Nuclide (uCi/ml) (uCi/pl) (uCi/ml) (uci/ml) l i 131I 0.68 0.14(-5) 0.41 0.08(-5) 134Cs 0.37 1 0.06 -4) 0.62 0.04(-6) 0.93 0.2 (-5)
- 0.55 *
- 136Cs 0.15 -5)
- 137Cs 0.4110.02-4) 0.14 0.06(-5) 0.15 0.03(-4) 51Cr 0.5710.08(-3 l s4Mn 0.24 1 0.03(-4 0.55 0.04(-6) 0.3110.10(-5) 0.48 1 0.19(-7) 59Fe 0.17 1 0.04(-4 *
- 57Co 0.74 1 0.23(-5 0.31 0.05(-51 seCo 0.536 0.005(-2) 0.39 i 0.10(-5) 0.50 0.07(-4) 0.14 0.02(-6) i 60Co 0.73 0.08 - 0.72 0.03(-5) 0.25 0.03(-2) 0.174 0.006(-5) '
65Zn 0.65 1 0.22 - *
- 95Zr 0.23 0.041 -
0.13 0.05(-5)
*
- 0.55 ssNb 0.43 0.06I -
0.20(-7) 103Ru 0.35 0.030 - *
- 0.43 0.21(-7) 110 mag 0.52 0.06I -
l 124Sb 0.55 1 0.07I l,- 0.21 0.02(-5)
- 0.11 i 0.04(-6) 12sSb 0.23 1 0.08I -
0.1310.02(-5) l 129mTe 0.16 1 0.04i,-, 0.25 i 0.10(-6)
* - Radionuclide not detected.
[1] Reactor coolant-inlet to letdown demineralizer. [2] Refueling Water Storage Tank [3] Spent fuel pit demineralizer inlet. [4] Spent fuel pit demineralizer effluent. i 1 I B-120 l
j TABLE B.33 (cont'd) UNIT #3 SPENT FUEL PIT AND ASSOCIATED WATER SAMPLES RWST [3] SFP [4] ohant[1] vt [2] 1/3/78; 10:50 1/3/78; 10:30 1/3/78; 11:10 1/3/78; 14:40 Nuclide (pCi/ml) (pCi/ml) (pCi/ml) (uCi/ml) 131I 0.49 0.02(-4) 0.62 i 0.01 (-4) 0.479 0.008(-4) 134Cs *
- 0.76 1 0.07(-6) 0.15 0.03(-4) 136Cs 0.73 i 0.20(-5) 137Cs *
- 0.167 0.007(-5) 0.27 i 0.01(-4) 51Cr 0.21 0.01(-3) 0.25 0.01(-3) 0.58 0.5 (-5) 54Mn 0.20 0.01(-4) 0.62 1 0.01(-4) 0.57 0.60(-6) 0.44 i 0.21(-5)
SOF e 0.56 i 0.19 -5 0.20 1 0.01(-4 0.26 0.2 (-5) 57Co 0.75 0.09-5)
- 0.61 2 0.05(-5) 58Co 0.43 0.01 -2) 0.427 i 0.006 2) 0.123 0.002( 4) 0.77 0.02(-4) 60Co 0.61 0.02 -3 0.91 0.02 - 0.59 0.02(-5 0.396 0.003(-2) 95Zr 0.13 1 0.03 -4 0.53 i 0.02 - 0.2710.13-95Nb 0.27 1 0.08 - 0.87 1 0.02 - 0.64 0.16 -
103Ru 0.1510.04- 0.53 1 0.01 - 0.70 0.10 - 0.07 -
- 110 mag 0.80 0.03 - 0.21 1 0.01 - 0.45 184Sb 0.79 0.05 - 0.75 1 0.03 - 0.136 0.003( 4 12sSb 0.33 1 0.05 - 0.38 2 0.02 -
0.76 0.02(-5))
- 140Ba 0.14 0.06 -
- 141Ce 0.11 1 0.02 -4) 0.137 0.008(-4)
O - Radionuclide not detected. [1] Reactor coolant-inlet to letdown demineralizer. [2] Dip sample from reactor cavity [3] Refueling Water Storage Tank [4] Spent fuel pit dip sample B-121 t
I TABLE B.33 (cont'd) UNIT 13 SPENT FUEL PIT AND ASSOCIATED WATER SAMPLES i a [1] av [2 SFP [3] RWST [4] l 1/11/78; 11:36 1/11/78;17:}.5 _1/11/78; 18:53 1/11/78; 11:34 M 11de (uCi/ml) (uCi/ml) (uCi/ml) (uCi/ml) l 131I 2.710.2(-5) 1.6 i 0.1(-5) 1.9 0.9(-6) 1.3 0.1 (-6) 134Cs 5.310.3(-5) 7.1* i 0.1(-5) 2.3*1 0.2(-5) 5.3 1.5 - 136Cs 2.0 0.5 - 137Cs 6.0 t 0.3 '-5) 8.4 1 0.1(-5) 3.4 0.1(-5) 1.5 0.2 - 51Cr 4.4 0.2(-4 3.74 1 0.08( 4) 8.4 1.5(-5)
- s4Mn 3.9 0.2(-5, 3.1 i 0.2 - 8.5* i 1.4(-6)
- 59Fe
- l.6 1 0.2 -
- 57Co 1.710.2(-5) 1.6 1 0.1 - 9.5 i 1.5(- )
- seCo 7 . 31 0.04(-3) 7.51 1 0.07(-3) 1.46 0.08-3) 2.0 0.2(-6)
SoCo 1.56 1 0.01(-3) 1.89 0.03(-3) 3.99 0.03 -3) .8 0.4(-6) l 95Zr 3.3 0.4-5) 2.9 1 0.2 - *
- l 95Nb 5.210.3-5) 5.4 0.1 - 1.3 1 0.1 - 5.3 1.3(-7) 103Ru 2.4 1 0.4 -5) 2.2 0.1 - 5.9 0.9 - *
-110 mag 8.6 1 0.4 -5) 4.3 0.2 - 2.4 1 0.2 -
- l 124Sb 1.1210.04(-4) 1.20 0.02(-4 2.4 0.2(-5) 8.8*1 0.2(-7) 12sSb 5.5 1 0.6 - 5.210.3(-5) ) 1.3* 1 0.3(-5) 14cLa 2.7 1 0.8 - *
- 141Ce 2.1 0.3 -
- 5.1 0.8(-6)
- 144Ce 3.3 0.9 - 5.0 0.7(-5) *
- l
* - Radionuclide not detected. l
[1] Reactor coolant-letdown demineralizer inlet. [2] Reactor cavity dip sanple. [3] Spent fuel pit dip sample.
-[4] Refueling water storage tank.
l t B-122
TABLE B.33 (cont'd) UNIT #3 SPENT FUEL PIT AND ASSOCIATED WATER SAMPLES SFP SFP SFP SFP Filter & Dip Ffiter Filtrate Filtrate Sample 1/11/78; 18:53 1/11/78; 18:53 1/11/78; 18:53 1/11/78; 18:53 Nuclide (pCi/ml) (pCf/ml) , (pC1/ml) __ (pCi/ml) 131I 6.9 1 0.4(-7) *' 6.9 i 0.4(-7) 1.9 0.9(-6) 134Cs 1.9 i 0.1(-5) 1.9 0.1(-5) 2.3 1 0.2(-5) 137Cs 1.5 2 0.4(-7) 3.01 0.07(-5) 3.02 1 0.07(-5) 3.4 i 0.1(-5) siCr 8.10 0.09(-5) 2.6 0.3(-5) 1.07i0.03(-4) 8.4 i 1.5(-5) 54Mn 2.98 0.08(-6) 7.711.3(-6) 1.07 0.13(-5) 8.5 1.4(-6) 59Fe 2.6 0.1(-6)
- 2.6 i 0.1(-6)
- s7Co 1.510.1(-7)
- 1.5 1 0.1(-7) 9.5 i 1.5(-6) seCo 4.44 1 0.04(-5) 1.32 i 0.01(-3) 1 .36 i 0.01(-3) 1.46 0.08(-3) 60C0 3.20 0.03(-5) 3.76 0.09(-3) 3.79 J.09(-3) 3.99 i 0.03(-3) 957p * *
- ssNb *
- 1.3 0.1(-5) 2.79
- 103Ru 0.05(-6 2.79 0.05(-6) 5.9* i 0.9(-6)'
106RuD 1.1* 1 0.8(-6) )
- 1.1 1 0.8(-6) 110 mag 9 1(-6) 9 1(-6) 2.4 1 0.2 -
124Sb 7.5 1 0.6 - 2.31 0.07(-5) 2.?3 0.07(-5) 2.4 0.2 - 12sSb 7.3 1.2 -
- 7.3 i 1.2 - 1.320.3-141C 5.5 0.6 - 3. s 0.5 - 4.0 t 0.5 - 5.1 0.8 - .
144C3 1.710.2- 2. 2 0.3 - 2.3 1 0.3 - * '
- Radionucl.ide not detected, f
i i B-123 I
TABLE B.33 (cont'd) UNIT #3 SPENT FUEL PIT AND ASSOCIATED WATER SAMPLES SFP SFP SFP Demin. Demin. Demin. Inlet Outlet [1] Outlet [2] 2/4/78; 10:10 2/4/78; 10:20 2/4/78; 10:20 Nuclide (uCi/ml) (uCi/ml) (uCi/ml) 1311 1.3 1 0.1(-8) 134Cs- 2.110.8(-6)
- 137Cs 1.510.5(-6) 1.30 1 0.09(-8)
! 51Cr 1.5 0.4(-5) 7.8 1 0.2 -7) 4.9 54Mn 3.5 1 0.2 -8) 4.6 0.6-8')l 0.7 -9 59Fe
- 2.8 1 0.3 -8 4.5 2.1 -9 57Co 1.1 0.2(-6) 5.4 1 0.3 -9 6.5 i 0.5 -9 seCo 3.05 0.08(-4) 1.89 0.04 6) 2.57 i 0.07( 6) 60Co 3.7* i 0.1(-4) 4.5 0.1(- 1.06 0.04(-6)
< 95Zr 9.2 0.?(-
95Nb
- 1.6 0.4 -
- 10 3Ru 1.6 0.6(-6) 5.110.2- 3.6i0.9(-9) j 106RuD
- 5.6 1 0.9 - 4.8 0.5(-8) -
110 mag
- 9.1 i 1.7 - *
, 124Sb 3.3 1 0.4(-6) 7.1 1 0.2 -
- 12sSb 3.1 0.9(-6) 4.3 1 0.9 -
- 141Ce *
- 7.4 1 0.4 - l 144Ce *
- 3.1 0.1 -
1s4Eu *
- 3.0 i 1.4 -
187W
- 3.2 0.6 -
4
* - Radionuclide not detected.
[1] Anion resin [2] Cation resin B-124 4
TABLE B.34 RADIONUCLIDE CONCENTRATIONS IN SECONDARY WATERS UNIT #3. 11/8-9/77 Steam Generator Blowdown 3A l 11/8/77; 12:30
, Nuclide . (uci/ml) 82Br 7.1 i 2.6 1311 9.910.1-6 1321 *
[- 1331 3.5 *i 0.1(-5) 134I 13sI 3.4 i 0.2(-5) 8%b
- 89Rb
- 134Cs 1.4 0.1 -
13sCs 1.5 0.3 - I 137Cs 2.5 0.1 - 138Cs
- 24Na 1.5 1 0.1(-6) 41Ar
- 51Cr <5.9(-8 54Mn <l.7 -
56Mn <7.2 - - 59Fe <9.0 - l 57Co <1.9 - ssCo <6.1 -8) ! 60C0 <1.2 -7) , 65Zn <5.2-8) 91$p <2,6 _8)
- 93Y <5.2 -
j_ 95Zr <2.4 - ! 95Nb <l.2 - 99Mo <5.1 J- I lo 3Ru -
) i lo6RuD <8.4(I-
<8.2
- 110 mag <9.4 -
124Sb <2.1 - 12sSb <4.2 - 129Te
- 1 129 Me <6.0(-8 I i
131Me <2.4 - 132Te <4.6 - 13988 <l . 7 - 140Ba <6.0 - 14ala 1.12 3(-7) 141Ce- <3.4 -
. 143Ce <3.9 - l 144Ce <3.4 - l 152Eu <5.6 -
i 154Eu <7.4 - 187W <6,4 I 239Np <3.9(-8 l-
- Radicauclide not detected.
I B-125
TABLEB.34(cont'd) RADIONUCLIDE CONCENTRATIONS IN SECONDARY WATERS
- Unit #3,11/8-9/77 Steam Generator Steam Generator Steam Generator Main' Steam Main Steam Blowdown 3A Blowdown 3B Blowdown 3C 3A 3B 11/9/77; 16:51 11/9/77; 18:06 11/9/77; 16:55 11/9/77; 19:29 11/9/77; 20:41 Nuclide (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) (pCi/ml) 82Br *
<1.0(-8) <1.3(-6) <4. 5 (-7) <9.4 (-8) 131I 2.8 i 0.2(-6 2.9 i 0.2(-6) 1.8 0.1 - <3.1 (-7) 8.4 3.L(-8) 132I 1.3 0.2(-6 6.1 1.7(-7) 1.8 0.1 - <1.9 -8) 133I 2.8 i 0.3 -6 2.4 0.2 3.4 0.1 - <1.3(-9 134I 4.2 i 1.9 -7 <6.0(-8) (-6) 7.2 t 0.3 - <4. 2 (-8 <2.4-6-7)1
<1.1 135I 1,4 i o,4 -6) 8.1 2.9(-7) 2.2 0.1 - <7.6(-7 <2.5-8) 88Rb <1.1(-6) <1.9(-8) <6.24(-6) <4.2(-7 89Rb <2.8(-7) <3.0(-7) <1.78(-7) <1.3(-7 <4.1 -
to
' 134Cs 3.111.1(-7) 3.8 i 0.9(-7) <5.5 - <1.9 -
136Cs <4.5(-8) <1.8(-8) 8.0 1.06 1 0.24(-6 0.4(-6) ) <4.6 - <2.3 -
.$ 137Cs 6.1 1.l(-7) 7.0 1 0.8 1.5 1 0.1(-5) <3.2 - <1.3 -
13sCs -8.13 2.47(-7) 8.612.7(-7) t-7) 4.64 i 0.32(-5) <2.1(-7) <3.6(-7) 24Na <4.0 -
- 7.69 i 0.98(-6) *
<8.1 -7) 41Ar <7.0 - <4.0 - * <8.8 - <2.6 -
51Cr <1.4 - <2.1 - <1.62(-7) <3.5 - <1.4 - 54Mn * <1.3 - <1.40(-7) <1.2 - <1.6 - 56Mn <1.6(-7 <3.2 - <2.00(-6 <1.4 -7 <1.3 - 59Fe <2.5(-8 <5.5 - <2.85(-7 <6.3 -7 <1.4 - 57Co <1.6(-8 <1.8 - <6.53(-7 <1.5 -7 <8.6 - seCo <1.4(-7 <1.3 - <4.40(-7 <1.7 -7 1.6 0.7(-7) 60Co <3.1 - <3.6 - <3.7( 7 <2.4(-7 <2.4 -7)
<1.8 - <4.06 -
- ssZn <2.6 - <2.6 -7) 91Sr <2.0 - <2.7 - <9.44 - <2.0 -7) <1.1 -7) .
93Y * <4.0 - <4.32 - <1.1 -7) 1.3 0.8 95Zr <2.8 - <1.4 - <2.50 -
<4.3 -7)
<1.1(-7)(-6) 95Nb <2.3 - <4.8 - <2.1(-8) 99Mo <2.3 - <3.6(-8) <9.12(-7) <1.5(-7) <3.5(-8)
TABLE B.34 (cont'd) RADIONLO.IDE CONCENTRAT1GNS IN SECONDARY WATERS UNIT #3, 11/8-9/77
-Steam Generator Steam Generator Steam Generator. Main Steam Main-Steam Blowdown 3A Blowdown 3B Blowdown 3C 3A 38 11/9/77; 16:51 11/9/77; 18:06 11/9/77; 16:55 11/9/77; 19:29 11/9/77; 20:41 Nuclide~ (uCf/ml) _ (uCi/ml) (uci/ml) _
(uct/ml) (uci/ml) 103Ru <2.6(-7) <1.5(-7) <3.36(-6) <1.6(-8) <1.5(-7) 106RuD <4.2(-8) 1.2 0 4(-6) 8.52 i 1.06(-5) <5.1(-8) 110 mag *
- 124Sb <8.1(-8)
<4.9 -
<2.1(-7)(-7) 1.9i10 <1.6(-7)
<5.8(-8 *
<6.5-9) 12sSb <3.1(-8) <2.4(-9 <1.1 - <1.4 -
* <5.4 -9) 129Te <1.4 -5 <1.2 - <1.3 - <1.9 -5 129mTe <2.5(-8 <1.8 -7 <3.2 - <4.8 - <1.4 -7 131mTe <1.8(-7 <3.1 -8) <3.5 - <2.1-7) <4.3 -
ca 132Te <9.7(-9 <6.9 -
.L
<1.2 -7) <5.1 -9) <1.2 -
13988 <l.2 - 6.1 3.6-6) <1.8 -10 <2.2 - M 140Ba <1.6 - <l .2(-7) 5.8 i 2.2 (-7) 5.4 0.8 -6) <2.4-7)) <1.2 - 140La <2.6 - <1.4(-7) 8.0 i .0-6) *
<1.3 -
141Ce <6.6(-9 <8.2(-7) <7.0 - <l.4(-8 143Ce <2.3(-7 <3.3(-8) 2.7 1.1 (-7) <1.8(-6) <1.5 - <2.4(-8 144Ce <7.9 - <1.2(-7) <4.5 - <3.4 - <9.2 -8
- . 152Eu <1.4 - 1.5106(-7) <6.4 - <2.6 - <4.4 -
154Eu <7.7 - <3.1 - <4.8 - <7.3 - <3.0 - 187W <9.4 - <1.3 - <1.1 - * <1.5 - 239Np <1.9(-7 <l.1 - <2.8 - <2.3(-7) *
- Radionuclide not detected.
TABLE B.34 (cont'd) RADIONUCLIDE CONCENTRATIONS IN SECONDARY WATERS UNIT #3, 11/8-9/77 Main Steam Unit 3 Blowdown Unit 3 Main Unit 3 Main 3C Flash Tank Feed Condensate 11/9/77; 20:14 11/9/77; 14:55 11/9/77; 18:00 11/9/77; 19:23
.Nuclide (pCi/ml) (pCi/ml) (pC1/ml) (pC1/ml) 82Br <l.1(-8) <6.l(-7) <2.5 -7) <7.2 -
131I 2.9 i 0.2 - 8.6 0.3 - <9.9 -8 <8.0 - 132I 6.1 i 1.7 - 8.1 0.2 - <5.6 -7 <l.3 - 133I 2.4 0.2 - 1.7 t 0.1 - <2.9(-8 <3.5 - 1 34I <6.0 - 3.0 0.2 -5 <5.1 -8 <3.1 - 135I <4.5 - 1.2 0.1 -4 <l.3 -7 <6 . 4 - serb <9.4 - 7.2* i E.6(-6) 1.1 0.8(-6) <l.6 - 89Rb <9.9 - <5.4(-7) <3.0 - 134Cs <2.2 - 4.9 0.2(-6) <2.2(-8) <2.0 - 136Cs <4.7(-9) <5.5 - <6.6(-8) L 1.3 i 1.0(-7) g 137Cs <l.8(-7) <l.l(-7)(-6) 9.1 0.4 <l.1 -
<5.5 - <7.2(-8) 13eCs 1.4 0.30(-6) 1.7 1 0.2(-5) 24Na <l,1(-7) 4.6 0.8(-6) <9.9-8) <3.0 -
41Ar <1.8(-7)
* <4.9 -7) <l.4 -
51Cr <9.5 -8) <4.8(-7) <9.8-8) <6.5 - 54Mn <3.4 -7)
- 1.0 0.3(-7) 1.2 0.6(-7) 56Mn <3.4 - * <4.0-7) <l.3(-7) 59Fe <3.1 - <l.9(-8 <l.7-7) 57Co <1.9 - <4.0(-8 <7.6 -8) <l.4(-7)(-7) 1.1 1 0 5 seCo <4.1(-9) <3.4 -7 <6.2 - <2.3(-7 60Co 2.4 1 6(-7) <2.9 -7 <2.1 - <3.2(-7 65Zn <2.5(-7 <7.8-7) <1.2 - <4.2(-8 91Sr <2.5(-7 <1.1 -6) <3.2 - <l.6(-7
<3.9 - <9.0(-8 <8.8 -
'93Y <l.1 -
<8.8 -
95Zr <2.1 - <3.5 - <l.8(-7 9sNb <5.0 - <4.9 - <l.6(-8 <l.8 -
<5.1 - <3.4(-8 <6.5 -
99Mo <5.9 -7)
TABLE B.34'(cont'd) RADIONUCLIDE CONCENTRATIONS IN SECONDARY WATERS UNIT #3.11/8-9/77 l Main Steam ~ Unit 3 Blowdown Unit 3 Main Unit 3 Main 3C Flash Tank Feed Condensate 11/9/77; 20:14 11/9/77; 14:55 11/9/77; 18:06 11/9/77; 19:23 Nuclide- (uCi/ml) (uci/ml) (uti/ml) (uci/ml) . ! 103Ru <2.5(-7) <2.2(-6) <3.8(-8 <1.5(-7 106RuD 5.3107(-5)
~
<1.1(-8 <1.9(-7 110 mag .
<1.3 - <7.0(-8 124Sb <3.2(-8) <1.5 - <1.4 - <9.0(-8 l 12sSb <2.0(-7) <2.6 - <3.2 - <5.4(-8 129Te * <3.5 - <7.2 -
- 129fqe <l.9(-7 <2.0 - <4.9-8) <9.9(-8) 131mTe <1.9(-7 *
<4.6 -8) <l.7(-7) , 132Te <4.9 - <5.2(-7) <9.7(-8) <1.0-) .L 139Ba <7.6 - <l.5(-7) <9.7 - p @ 140Ba <1.6 - <1.4(-7)(-6) 2.9 0.3 2.6
- 1 2(-7) <6.7 - g 140La <5.8 - 5.8105(-6) <4. 3 - <4.0 -
141Ce <4.2 - <1.2 - <1.5 - <3.1 - 143Ce <1.0 - <1.2 - <4.1 - <1.7 -7 144Ce <2.0 -7 <6.5 - <2.4 - <1.2 -7 152Eu <4.7 -8) <1.3(-7 <9.8-8) <2.4-7) is4Eu <2.7 -7) <1.5(-7 <1.5 -7) <1.2 -7) i 187W <2.5(-6) <4.4(-7) <3.2(-8) <5.8(-7) 239Np <2.1(-7) <3.2(-6) <8.2(-7) <7.7(-7)
- Radionuclide not detected.
l l l
' 1
~
- 0 4
- ) ) )
m 6 7 6 7 a 1) - - ( e ;m l ( ( 5 t 4 2 SC7/ ) ) )) ) )) ) )) _ 37i 70 0 5 780 8 88 7 87 n /C - - - - - - - - - - ( - - - - - (( - - i a 4 p ( 1 ( 5 *
* (18.
( 0 ( 5 ((* 36 2220954952 _ M / 1( 1 27 3 . 8 l 18 . 4 15 155291383l 1 <5 2 < < <1 < << < < < < < < < < < < 7 3 )
- ) )
m aL 1) 6 7 7 7 e l ( ( ( m )
- 4. )
_ t ; 3 4 SB7/ ) ) ) )) 37i 80 1 6 6 0 1 )8 8807 n /c - - - - - - - - - - - - - - - - - S R i a 4 p 1( ( ( 1 *
- 6 0 0.*
( 1 ((* 43 (( 16.2 ( 2962552 E T M / 1 3. 7 . 5 9 7 646 1 34 53 2211531 A 1 <1 6 < < < < < < < < <1 < < < < < < < W Y R _ A 2 D 3 N : ) ) O m 6 7 7 C a 1) - - E e l ( ( S7 t ; m 3 8 7 SA7/ ) ) ))) ) )) ))) N/ 37i 80 1 6 797 7 98 877 5 Id n /c - - - - - - - - - - - - - - - - - - (( ((( - 3 1 i 4 u ( ( ((( ( S/ a 1( 1* 0 *
- 2 7 0.
- 8 3 6.* 8539421587 B N1 O1 M /
1 17 6 . 6 8 749 7 24 1471663135 E I T , 1 < 1 5 < < << < < < < < < < < < < < < < L B A3 A R# T T NT r 1 _ EI o 3 _ CN tB: )) ) ) ) NU a31 75 5 7 7 O r 1) - - - - - _ C en l (( ( ( ( nw;m 81 2 ) _ E D eo7/ 5. )~9 ) )) )))))0) _ Gd7i 00 0 080 6 88 7865817 I w/C - - - - - - - - ((( - - - - - - L mo4 p 1 i( ( (( C U al1( 1* * * *
- 1. 8
- 5 7 5.
- 3727686615 N
O t S eB/ 1 1 55 21 6 4 76 3 <7
. 3 44 4651661192 I
D A R r 4 o 3
) ) ) ) )
tA: 5 7 a31 6 5 7 r 1) - - -
- (
en nw;m l ( ( ( ( 6 eo7/ ) 1 1
. 9. ) 6 ) )) )))
8887866 Gd7i 70 0 2 080 6 88 w/c - - - - - - - - - - - - - - - mo4 u (
- i
( ( (( ((( al1( eB/ t 1~ 12 2 . 3 8
- 9. 9.
79 3 2 67.*5968125814 14 4l81543114 S 1 <9 3 6 2<3 < << < < < < < < < <<< d e i r bhssss a rnne00onr rbo l BIy I RpCCCC CMMFC0CZSYZNM c 2 1 23 s s9 46 78 1 4 6 970 0 5 1 3 5 5 9 u 83 a3 a e8 33 3 3 2 55 5 55 5 6 699999 N 1 l 1 l 1 1 1 1 ys$
. TABLE B.35 (cont'd)
RADIONUCLIDE CONCENTRATIONS IN SECONDARY WATERS'
. UNIT #3,'11/14/77-Steam Generator Steam Generator Main Steam Main Steam Main Steam Blowdown'3A Blowdown 3B . 3A .38 ~3C 11/14/77; 11:34 11/14/77; 11:31 11/14/77; 16:32 11/14/77; 16:37 11/14/77; 16:40
'Nuclide (uti/ml) (uCi/ml) (uCi/ml) (uCi/ml) (uCi/ml) lo3Ru <5.3(-8 <1.7-8) <5.5 - <8.8 - <4.8 -
106RuD .<7.9(-9 .<5.3-10) <3.3 - <3.8 - <2.7 -
- l. 11014g <3.1(-7 <4.2 -7) <1.3 - <1.2 -
l 124Sb
<1.0 -
<1.3(-8 <2.5 -8)' .< 3.6 - <3.7 - <1.9 -
b <5.0(-8 . 8.5 -8)
<4.1,- <d.1-8) <3.0-9) 129mie .<3.3 - .<4.0 - ) 8.8- 2(-7) .<1.2 -}}