NG-17-0111, Duane Arnold Energy Center, Revision 24 to Updated Final Safety Analysis Report, Chapter 11, Radioactive Waste Management
Text
UFSAR/DAEC-1 Chapter 11: RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS 11-i Revision 22 - 5/13 Section Title Page
11.1 SOURCE TERMS ....................................................................................... 11.1-1
11.1.1 Basic Data for Source-Term Calculations ................................................ 11.1-1 11.1.2 Primary Radioactive Gas Sources ............................................................. 11.1-6 11.1.2.1 Process Offgas (Steam Jet Air Ejector) ................................................. 11.1-6 11.1.2.2 Mechanical Vacuum Pump Offgas ........................................................ 11.1-7 11.1.2.3 Drywell Ventilation Gas ........................................................................ 11.1-7 11.1.2.4 Gland Seal Condenser Offgas ................................................................ 11.1-7 11.1.2.5 Other Potentially Radioactive Gases ..................................................... 11.1-7
11.2 LIQUID WASTE MANAGEMENT SYSTEM .......................................... 11.2-1
11.2.1 Design Bases ............................................................................................. 11.2-1 11.2.1.1 Power Generation Objectives ................................................................ 11.2-1 11.2.1.2 Power Generation Design Bases ............................................................ 11.2-2 11.2.1.3 Safety Design Basis ............................................................................... 11.2-2 11.2.1.4 Decontamination Factors ....................................................................... 11.2-2 11.2.1.5 Codes and Standards .............................................................................. 11.2-3 11.2.2 System Description ................................................................................... 11.2-3 11.2.2.1 High-Purity Wastes ................................................................................ 11.2-6 11.2.2.2 Low-Purity Wastes................................................................................. 11.2-7 11.2.2.3 Chemical Wastes .................................................................................... 11.2-8 11.2.2.4 Detergent Wastes ................................................................................... 11.2-9 11.2.2.5 Sludges ................................................................................................... 11.2-9 11.2.2.6 Spent Resins ........................................................................................... 11.2-9
11.2.2.7 Means for Keeping Radioactive Discharges as Low as Reasonably Achievable ........................................................... 11.2-10 11.2.2.7.1 System Features ............................................................................. 11.2-10 11.2.2.7.2 Procedural Controls .......................................................................... 11.2-11 11.2.2.8 Power Generation Evaluation .............................................................. 11.2-13 11.2.2.9 Safety Evaluation ............................................................................... 11.2-13 11.2.2.10 Inspection and Testing ....................................................................... 11.2-14 11.2.2.11 Instrumentation and Control .............................................................. 11.2-14 11.2.2.12 Design Pressures, Temperatures, and Material .................................. 11.2-14 11.2.3 Radioactive Releases .............................................................................. 11.2-19 11.2.3.1 Principal Radionuclides ....................................................................... 11.2-19 UFSAR/DAEC-1 Chapter 11: RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS (Continued) 11-ii Revision 22 - 5/13 Section Title Page
11.2.3.2 Effluent Concentration ......................................................................... 11.2-20 11.2.3.3 Effect on the Environment ................................................................... 11.2-20 11.2.3.4 Tritium ................................................................................................. 11.2-20 11.2.3.5 Release of Accumulated Gaseous and Liquid Radwastes ................... 11.2-21 REFERENCES FOR SECTION 11.2 ................................................................ 11.2-23
11.3 GASEOUS WASTE MANAGEMENT SYSTEM ...................................... 11.3-1 11.3.1 Design Bases ............................................................................................. 11.3-1 11.3.1.1 Power Generation Objectives ................................................................ 11.3-1 11.3.1.2 Power Generation Design Basis ............................................................ 11.3-1 11.3.1.3 Safety Design Basis ............................................................................... 11.3-2 11.3.2 System Description ................................................................................... 11.3-3 11.3.2.1 Process Description ................................................................................ 11.3-3 11.3.2.2 Equipment Description .......................................................................... 11.3-4 11.3.2.3 Instrumentation and Control .................................................................. 11.3-8 11.3.2.4 Safety Evaluation ................................................................................... 11.3-8 11.3.2.5 Inspection and Testing ......................................................................... 11.3-10 11.3.3 Radioactive Releases .............................................................................. 11.3-11
11.4 SOLID WASTE MANAGEMENT SYSTEM ............................................ 11.4-1 11.4.1 Design Bases ............................................................................................. 11.4-1 11.4.1.1 Power Generation Objectives ................................................................ 11.4-1 11.4.1.2 Power Generation Design Bases ............................................................ 11.4-1 11.4.1.3 Safety Design Basis ............................................................................... 11.4-2 11.4.2 System Description ................................................................................... 11.4-2 11.4.2.1 General ................................................................................................... 11.4-2 11.4.2.2 Wet Wastes ............................................................................................ 11.4-3 11.4.2.3 Dry Wastes ............................................................................................. 11.4-4 11.4.2.4 Storage Facilities .................................................................................... 11.4-5 11.4.2.5 Safety Evaluation ................................................................................... 11.4-6 11.4.2.6 Inspection and Testing ........................................................................... 11.4-6
11.5 PROCESS AND EFFLUENT RADIATION MONITORING AND SAMPLING SYSTEMS ............................................................................... 11.5-1 11.5.1 Main Steam Line Radiation Monitoring System ...................................... 11.5-1 11.5.1.1 Safety Objective ..................................................................................... 11.5-1 11.5.1.2 Design Bases .......................................................................................... 11.5-1 UFSAR/DAEC-1 Chapter 11: RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS (Continued) 11-iii Revision 22 - 5/13 Section Title Page
11.5.1.3 System Description ................................................................................ 11.5-2 11.5.1.4 Safety Evaluation ................................................................................... 11.5-2 11.5.1.5 Inspection and Testing ........................................................................... 11.5-3 11.5.2 Air Ejector Offgas Radiation Monitoring System .................................... 11.5-3 11.5.2.1 Power Generation Objectives ................................................................ 11.5-3 11.5.2.2 Power Generation Design Bases ............................................................ 11.5-3 11.5.2.3 System Description ................................................................................ 11.5-3 11.5.2.4 Power Generation Evaluation ................................................................ 11.5-6 11.5.2.5 Inspection and Testing ........................................................................... 11.5-6 11.5.3 Offgas Stack Radiation Monitoring System ............................................. 11.5-6 11.5.3.1 Power Generation Objectives ................................................................ 11.5-6 11.5.3.2 Power Generation Design Bases ............................................................ 11.5-6 11.5.3.3 System Description ................................................................................ 11.5-7 11.5.3.4 Power Generation Evaluation ................................................................ 11.5-8 11.5.3.5 Inspection and Testing ........................................................................... 11.5-8 11.5.4 Liquid Process Radiation Monitors .......................................................... 11.5-8 11.5.4.1 Power Generation Objectives ................................................................ 11.5-8 11.5.4.2 Power Generation Design Bases ............................................................ 11.5-8 11.5.4.3 System Description ................................................................................ 11.5-9 11.5.4.4 Power Generation Evaluation .............................................................. 11.5-12 11.5.4.5 Inspection and Testing ......................................................................... 11.5-12 11.5.5 Reactor Building Ventilation Radiation Monitoring System ................. 11.5-12 11.5.5.1 Power Generation and Safety Objectives ............................................ 11.5-12 11.5.5.2 Design Bases ........................................................................................ 11.5-12 11.5.5.3 System Description .............................................................................. 11.5-13 11.5.5.4 Safety Evaluation ................................................................................. 11.5-14 11.5.5.5 Inspection and Testing ......................................................................... 11.5-14 11.5.6 Site Environs Radiation Monitors ........................................................... 11.5-14 11.5.6.1 Power Generation Objectives .............................................................. 11.5-14 11.5.6.2 System Description .............................................................................. 11.5-15 11.5.7 Preoperational and Postoperational Environmental Radioactivity Monitoring Programs ....................................................... 11.5-15 11.5.7.1 General ................................................................................................. 11.5-15 11.5.7.2 Technical Discussion ........................................................................... 11.5-16 11.5.7.3 Aquatic Environment ........................................................................... 11.5-16 11.5.7.4 Terrestrial Environment ...................................................................... 11.5-18 UFSAR/DAEC-1 Chapter 11: RADIOACTIVE WASTE MANAGEMENT TABLE OF CONTENTS (Continued) 11-iv Revision 22 - 5/13 Section Title Page
11.5.7.5 Atmospheric Environment ................................................................... 11.5-19
11.5.8 Postaccident Radiological Monitoring and Sampling Systems ................................................................................... 11.5-19
11.5.9 Extended Range Airbor ne Effluent Radiation Monitoring System ................................................................................. 11.5-19 11.5.9.1 System Description .............................................................................. 11.5-19 11.5.9.2 Monitor Characteristics ........................................................................ 11.5-21 REFERENCES FOR SECTION 11.5 ................................................................. 11.5-22
Appendix 11A - GASEOUS RELEASE RATE LIMIT CALCULATIONS ....... 11A-1 UFSAR/DAEC-1 Chapter 11: RADIOACTIVE WASTE MANAGEMENT LIST OF TABLES 11-v Revision 22 - 5/13 Tables Title Page
11.2-1 Estimated Annual Radioactivity Release in Liquid Effluents ................ T11.2-1 11.2-2 Estimated Activity Concentrations in Wastes Before Introduction into the Effluent Line ............................................. T11.2-4
11.2-3 Release Concentrations of Individual Isotopes During Discharge in Effluent Line ......................................................... T11.2-5 11.2-4 Estimated Radioactive Liquid Effluent Quantities and Dose Rates for Various Offgas Release Rates ........................................ T11.2-6 11.2-5 Quantities of Principal Radioisotopes Estimated in Liquid Wastes Subject to Release .......................................................... T11.2-7 11.2-6 Isotopic Activity Inventory ..................................................................... T11.2-8 11.3-1 Estimated Isotopic Release Rates from the Steam Jet Air Ejector System ........................................................................................ T11.3-1 11.3-2 Noble Gas Release Rate to Environs from Turbine Gland Seal Exhauster System ............................................................................ T11.3-2
11.3-3 Radioactive Particulate Daughter Buildup from Gland Seal Noble Gas Effluent Ratio X i/MPC i Versus Distance in Worst Sector............................................................................................ T11.3-3 11.3-4 Equipment Malfunction Analysis ........................................................... T11.3-4 11.3-5 Process Instrument Alarms ..................................................................... T11.3-5 11.3-6 Steam Jet Air Ejector Isotopic Release Rate ........................................ T11.3-10 11.3-7 Turbine Gland Seal Isotopic Release Rate ........................................... T11.3-11
11.3-8 HPCI Gland Seal Isotopic Release Rate ............................................... T11.3-12 11.3-9 Reactor Building and Turbine Building Combined Ventilation Release Rate ....................................................................... T11.3-13
11.3-10 Drywell Purge Isotopic Release Corresponding to 25,000
µCi/Sec at 30-Min Decay Offgas Rate ................................... T11.3-14 11.4-1 Maximum Isotopic Activity Inventories in Solid Wastes ...................... T11.4-1
11.5-1 Characteristics of Process Radiation Monitoring Systems ..................... T11.5-1
11.5-2 Environmental and Power Supply Design Conditions for Process Radiation Monitoring System .................................................... T11.5-2
11.5-3 Characteristics of Extende d Range Airborne Effluent Monitor System ....................................................................................... T11.5-3
UFSAR/DAEC-1 Chapter 11: RADIOACTIVE WASTE MANAGEMENT LIST OF FIGURES 11-vi Revision 22 - 5/13 Figures Title
11.2-1 Deleted 11.2-2 Liquid Radwaste Sump System - P&ID 11.2-3 Liquid Radwaste System Equipment - P&ID 11.2-4 Floor Drain Radwaste System - P&ID 11.2-5 Radwaste Solids Handling System - P&ID 11.2-6 Service Water System - Pump House P&ID - Depicting Radwaste Dilution Line 11.2-7 Circulating Water System - P&ID - Depicting Radwaste Dilution
Structure 11.2-8 Site Plan Depicting Liquid Radwaste Discharge 11.2-9 Dilution Structure Details 11.2-10 Grading, Drainage and Utilities Area "R" 11.3-1 Offgas System Process Diagram 11.3-2 Offgas System - P&ID 11.3-3 Recombiner System - P&ID 11.3-4 Charcoal Adsorber Vessel Assembly 11.3-5 Distributor Ring 11.3-6 Site Plan - Gaseous Effluents 11.4-1 Solid Radwaste Storage Details 11.5-1 Process Radiation Monitoring System - Schematic and Connection Diagram 11.5-2 Air Ejector Offgas Radiation Monitoring System FCD 11.5-3 Radwaste Sample Station 11.5-4 Reactor Building Ventilation Radiation Monitoring
UFSAR/DAEC - 1 11.1-1 Revision 17 - 10/03 Chapter 11 RADIOACTIVE WASTE MANAGEMENT
The radioactive waste systems are designed to collect, process, and dispose of
potentially radioactive wastes produced during the operation of the plant. These wastes
are grouped as liquid, gaseous, or solid.
The liquid radwaste system is designed to process and return the collected liquid waste to the condensate storage tanks to the maximum extent practicable. The gaseous wastes are processed through a recombiner/charcoal delay system, monitored, and released to the atmosphere via the offgas stack. Solid wastes are packaged in suitable containers for offsite shipment and burial.
The liquid and gaseous effluents from the treatment systems are continuously monitored, and the discharges are terminated if the effluents exceed preset radioactivity
levels.
The radioactive waste treatment system design discussed in this chapter will limit the radioactive releases to the environment form the DAEC to the as low as is reasonably
achievable (ALARA) level.
11.1 SOURCE TERMS
11.1.1 BASIC DATA FOR SOURCE-TERM CALCULATIONS
The basic data for source-term calculations are as follows:
- 1. Operating power at which impact is to be analyzed is 1912 Mwt.
- 2. Weight of uranium loaded (first loading and equilibrium cycle): first core ~152,000-lb uranium. Approximately the same total weight will be maintained through reloading cycles.
- 3. Isotopic ratio in fresh fuel (first loading and equilibrium cycle): first core will contain 1.90% of uranium-235. Reload fuel contains up to 4.4% uranium-235.
- 4. The DAEC Design Basis offgas rate is 100,000 µCi/sec after 30 min delay, the typical rate is 5000 to 25,000 µCi/sec after 30-min delay.
- 5. The curie decontamination factor (DF) used is 61 and the krypton and xenon offgas discharge is approximately 410 µCi/sec. (See Table 11.3-6 for a detailed isotopic release rate.)
UFSAR/DAEC - 1 11.1-2 Revision 17 - 10/03
- 6. Mass of primary coolant in system
- a. Mass of primary coolant in reactor is 289,540-lb water, 9460-lb steam.
- b. Mass of primary coolant in recirculating system is 29,649 lb.
- 7. Steam conditions at turbine
Temperature 542.5°F Pressure 983 psia Flow 8,223,041 lb/hr (maximum)
- 8. Normal recirculation flow rate is 22,400,000 lb/hr.
- 9. The normal cleanup system flow rate is 70,000 lb/hr. The filter- demineralizer units are pressure precoat-type filters that use resinous fibers (Solka-floc) and finely ground mi xed ion-exchange resins (Epifloc). The fiber-to-resin ratio can vary from 1:1 up to 5:1. Specific decontamination factors are not available for particular isotopes.
However, an average decontamination factor of 10 is expected for particulate filtration and 100 for ion demineralization.
- 10. The expected performance of the expanded offgas system is described in Section 11.3.2. The main condenser design air inleakage is 18.5 cfm at
130°F. The condenser air ejector is a two-stage air ejector discharging to the offgas system. There are two condenser shells. The ejector discharge
will pass through 37 tons of charcoal at 77°F.
- 11. The expected leak rate of primary coolant to the drywell is 0.5-gpm steam and 0.5-gpm unidentified reactor water. The drywell is planned to be purged once a year; however, the impact has been analyzed using four
purges per year. The purge is passed through high-efficiency particulate
absorber (HEPA) and deep-bed charcoal filters in the standby gas treatment system.
- 12. The expected leak rate of primary coolant to the reactor building is 1.0-gpm reactor water. The ventilation air flow through the reactor building is 69,500 cfm. There is an additional 3000 cfm from infiltration.
The air flow is discharged through three reactor building discharge stacks.
The air is not filtered or otherwise treated before discharge; however, it is monitored.
UFSAR/DAEC - 1 11.1-3 Revision 13 - 5/97
- 13. The expected leak rate of steam to the turbine building is 5.0 gpm. The ventilation air flow through the turbine building is 37,500 cfm in the winter and 109,500 cfm in the summer. There is an additional 3500 cfm from infiltration. The air is discharged through the reactor building stacks
in the winter and through the reactor building stacks as well as through eight turbine building roof vents during the summer. The air is not filtered or treated before discharge; however, it is monitored.
- 14. Primary steam is the source of steam used in the turbine gland seals.
Effluent steam from the gland seals is discharged to the gland steam condenser. The condensate drains to the main condenser and the noncondensibles are discharged to a short delay in and then to the 100-m stack. Approximately a 2-min delay time exists between steam leaving the reactor vessel and subsequent release to the environment.
- 15. The estimated average gallons per day and microcuries per cubic centimeter are listed below for these categories of liquid waste: high-level wastes (e.g. "clean" or low-conductivity waste and equipment drains):
"dirty" wastes (e.g., floor drain wastes, high-conductivity wastes, and laboratory wastes; chemical wastes; and laundry, decontamination, and
washdown wastes:
Liquid Discharges Waste Gal/day µCi/cm 3 Equipment drains 0 --
Floor drains 2040 8x10
-6 Chemical wastes 500 4x10
-5 Detergent wastes 300 1x10
-5 For the above-listed wastes, the following are provided:
- a. Number and capacity of collector tanks.
- b. Fraction of water to be recycled or factors controlling decision.
- c. Treatment steps, including number, capacity, and process decontamination factor for each principal nuclide for each step. If
step is optional, factors controlling decision are stated.
UFSAR/DAEC - 1 11.1-4 Revision 17 - 10/03
- d. Decay time from primary loop to discharge.
- e. How waste concentrate (filter cake, demineralizer resin, evaporator bottoms) is handled. Total volume or weight and curies
per day or year is given.
Equipment Drains
- a. Waste collector tank (one), 10,000 gal. Radwste surge tanks, 40,000 gal, and 70,000 gal.
- b. One-hundred percent to be recycled.
- c. Filtration and demineralization. The overall decontamination factor for the filter an demineralizer is approximately 100. The decontamination factor for individual radionuclides is unkown.
The decontamination factor will vary in actual operation since it is
a function of inlet concentration of the soluble and insoluble
species present.
- d. Twelve hours.
- e. Waste concentrate is dewatered or solidified. The estimated weight and volume are 63,000 lb/yr and 2200 ft 3/yr, respectively, for all sludges and resins for the plant. The total isotopic inventory
of these solids is expected to be about 1000 Ci/yr.
Floor Drains
- a. Floor drain collector tank (one), 10,000 gal.
- b. Seventy percent recycled. The impact of discharges has been made assuming 30% floor drain, 100% chemical waste, and 100%
detergent drains discharged.
- c. Filtration and demineralization. The overall decontamination factor for the filter and demineralizer is approximately 100. The decontamination factor for individual radionculides is unknown.
This factor will vary in actual operation since it is a function of
inlet concentration of the soluble and insoluble species present.
- d. Twelve hours
UFSAR/DAEC - 1 11.1-5 Revision 13 - 5/97
- e. Waste concentrate is dewatered or solidified. The estimated weight and volume are 63,000 lb/yr and 2200 ft 3/yr, respectively, for all sludges and resins for the plant. The total isotopic inventory
of these solids is expected to be about 1000 Ci/yr.
Chemical Wastes
- a. Chemical waste tank (one), 4000 gal.
- b. None recycled. The impact of discharges has been made assuming 30% floor drain, 100% chemical waste, and 100% detergent drains
discharged.
- c. Normal method of processing is through the floor drain system. However, high-conductivity chemical wastes are received by the chemical waste tank and are processed by filtration.
- d. Twelve hours.
- e. The water concentrate is processed with spent resin in high integrity containers (HICs) or steel liners. Weight, volume, and isotopic inventory included in total cited for equipment and floor
drains.
Detergent Wastes
- a. Detergent drain tanks (two), 1000 gal each.
- b. None recycled. The impact of discharges has been made assuming 30% floor drain, 100% chemical waste, and 100% detergent drains
discharged.
- c. Detergent wastes will be treated in the same manner as chemical wastes to the maximum extent practicable.
- d. Twelve hours.
- e. Same as chemical wastes above.
UFSAR/DAEC - 1 11.1-6 Revision 20 - 8/09 16. For the condensate demineralizers, the flow rate, type of resin used, and expected backwash and regeneration frequency, and expected decontamination factor for each principal nuclide are as follows:
Flow rate 3370 gpm (For 5 demineralizers in service) 4213 gpm (For 4 demineralizers in service)
Type of resin Powdered Resin Expected backwash frequency As needed, to maintain system differential pressure and effluent water chemistry Regeneration frequency No regeneration Expected decontamination Specific decontamination factor for each nuclide factors are not available for particular isotopes. However, an average decontamination
factor of 10 is expected for particulate
filtration and a df of 100 for ion
exchange.
- 17. The normal dilution flow rate for liquid effluents is 6000 to 24,000 gpm as necessary and 3.16 x 10 9 to 1.2 x 10 10 gal/yr as necessary.
11.1.2 PRIMARY RADIOACTIVE GAS SOURCES
The four primary potentially radioactive gas sources are described below.
11.1.2.1 Process Offgas (Steam Jet Air Ejector)
Noncondensible radioactive offgas is continuously removed from the main condenser by the air ejector during plant operation. This is the major source and is larger than all other sources combined. The air ejector offgas will normally contain activation
gases, principally Nitrogen-16, Oxygen-19, and Nitrogen-13. The Nitrogen-16 and Oxygen-19 isotopes have short half-lives and quickly decay. The 10-min half-life Nitrogen-13 isotope is present in small amounts, which are further reduced by decay.
The air ejector offgas also contains the radioactive noble gas parents of biologically significant Strontium-89, Strontium-90, Barium-140, and Cesium-137 isotopes. The concentration of these noble gases depends on the usually extremely small amount of tramp uranium in the coolant and on the cladding surfaces, and the number and size of
fuel cladding leaks.
UFSAR/DAEC - 1 11.1-7 Revision 19 - 9/07 Radioactive-particulate daughters are retained on the HEPA filters and on the
charcoal. The offgas is discharged to the environs by the plant stack. The activity of the gas entering and leaving the offgas treatment system is continuously monitored. Thus, the system performance is known to the operator at all times.
11.1.2.2 Mechanical Vacuum Pump Offgas
During unit startup, air is removed from the main condenser by a mechanical vacuum pump. The mechanical vacuum pump exhaust is discharged to the off gas stack.
The mechanical vacuum pump will normally be in service only during startup and shutdown when little or no radioactive gas is present. It will be manually isolated on the receipt of a high-radiation signal from the offgas stack radiation monitors.
11.1.2.3 Drywell Ventilation Gas
The drywell air is exposed to neutron fluxes around the reactor vessel, which results in some activation products. Activity is also introduced into the drywell atmosphere by the drywell sumps and by the primary system relief valves when they vent to the suppression chamber. The drywell forms a closed system that may be purged with normal reactor building air, if necessary, when access is required. The drywell can also be vented during the plant startup to accommodate the expansion of air with increasing temperature. This air is discharged through the standby gas treatment system.
11.1.2.4 Gland Seal Condenser Offgas
The gland seal condenser exhauster discharges into a separate holdup piping system. There is a holdup of approximately 2 min to permit the decay of the short-lived
radioactive gases present. These are principally Nitrogen-13, Nitrogen-16, Nitrogen-17, and Oxygen-19. The release rate of radioactive gas is less than 0.1% of that from the air ejector offgas system. The gland seal exhaus t gas flows past the elevated release point radiation monitors before release so that its contribution to the release rate is included in the measured total.
11.1.2.5 Other Potentially Radioactive Gases
At times, it is desirable to vent certain tanks and discharge gases from specific laboratories and building service areas to the roof vent after allowing suitable delay time for radionuclide decay. These additions have low-activity levels and add small increments to the total quantity of radioactive gas requiring treatment. In all cases, maximum activities from these sources are examined to ensure that the vent discharge is safely below the established limits.
UFSAR/DAEC-1 T11.2-1 Revision 13 - 5/97 Table 11.2-1 Page 1 of 3 ESTIMATED ANNUAL RADIOACTIVITY RELEASE IN LIQUID EFFLUENTS a,b Isotope c Release Rate d (Ci/yr) 10-Year Inventory e (Ci) Sr-89 f 3.1 x 10-3 6.3 x 10-4 Sr-90 f 2.4 x 10-4 2.1 x 10-3 Sr-91 3.0 x 10
-2 4.8 x 10-5 Mo-99 f 2.0 x 10-2 2.2 x 10-4 I-131 1.4 x 10
-2 4.4 x 10-4 I-133 6.4 x 10
-2 2.2 x 10-4 I-135 4.1 x 10
-2 4.5 x 10-5 Cs-134 1.7 x 10
-4 4.7 x 10-4 Cs-137 2.5 x 10
-4 2.2 x 10-3 Ba-140 f 8.9 x 10-3 4.5 x 10-4 Ce-144 f 3.6 x 10-5 4.0 x 10-5 Np-239 2.2 x 10
-1 2.0 x 10-3 Co-58 2.7 x 10
-3 7.5 x 10-4 Co-60 2.7 x 10
-4 1.5 x 10-3 Total g 4 x 10-1 1.1 x 10-2
See notes on the following page.
UFSAR/DAEC-1 T11.2-2 Revision 13 - 5/97 Table 11.2-1 Page 2 of 3 ESTIMATED ANNUAL RADIOACTIVITY RELEASE IN LIQUID EFFLUENTS a,b a The radioactivity release does not include tritium.
b Based on an offgas rate of 100,000
µCi/sec annual release rate after 30-min decay. c Isotopes having a half-life less than 2.3 hr were excluded because the holdup in the plant generally would be sufficient to result in negligible concentrations in released wastes. Other isotopes of the elements listed were considered. The
radionuclides Zr-95, Nb-95, Ru-103, Ru-106, Te-129m, Te-132, Nd-147, Na-24, S-35, P-32, Cr-51, Mn-54, Mn-56, Fe-55, Fe-59, Cu-64, Ni-65, Zn-65, Zn-69m, Ag-110m, Ta-182, W-187 were also considered, but if present will be negligible
relative to those isotopes listed.
d Although two significant numbers are used to express release rates as a convenience for making further calculations, only one significant figure is
warranted by source data.
e Quantity present in the environment at the end of 10 year as a result of continous discharge at release rate shown.
f Daughter isotopes yttrium, technetium, lanthanum, and praseodymium may be observed in waste samples in equilibrium with, or approaching equilibrum with their parent depending on sample and analysis timing and procedure.
g Based on 30% floor drain, 100% detergent drain, and 100% chemical waste to discharge.
Additional Notes:
The following assumptions were used in the preparation of the isotopes list and
the selection of listed isotopes:
- 1. Radioisotopes with half-lives < 2.3 hr were excluded.
- 2. Radioisotopes observed in BWR water ha ving half-lives > 2.3 hrs were further evaluated.
- 3. An initial list was prepared taking into account fission product noble gas release rates, reactor water fractional cleanup rates, in-plant decay, and in-plant decontamination.
An isotope was included in the final list if three or four of the following criteria were
applicable:
- a. Half-life 24 hr b. Maximum permissible concentration of isotope 2 x 10-5 µCi/cm 3 c. Percent of maximum permissible concentration 0.01% d. Percent of total activity 0.5%
UFSAR/DAEC-1 T11.2-3 Revision 13 - 5/97 Table 11.2-1 Page 3 of 3 ESTIMATED ANNUAL RADIOACTIVITY RELEASE IN LIQUID EFFLUENTS a,b
- 4. Exceptions
- a. Sr-91 with only two criteria applicable was included because its daughter Y-91 would have three of the criteria applicable.
- b. Co-60 with only one of the criteria applicable was included because it would be expected to be of greater environmental significance than Co-58, which has three
of the criteria applicable.
UFSAR/DAEC-1 T11.2-4 Revision 13 - 5/97 Table 11.2-2 ESTIMATED ACTIVITY CONCENTRATIONS IN WASTES BEFORE INTRODUCTION INTO THE EFFLUENT LINE
Source of Waste
Tank Volume (gal) Daily Discharge Volume (avg. gpd)
Expected Annual Average Concentration a (µCi/cm 3) Concentration With Design-Basis Fuel Leak b (µCi/cm 3) Floor drain 10,000 2040 c 8 x 10-6 1 x 10-5 Detergent 1,000 300 1 x 10
-5 1 x 10-5 Chemical waste 4,000 500 4 x 10
-5 2 x 10-4 a Based on a fuel leakage rate that results in a an annual average stack release rate of 25,000 µCi/sec referenced to 30-min decay.
b Based on design-basis fuel leak that results in an annual average stack release rate of 1.0 x 10 5 µCi/sec of a noble gas diffusion mixture referenced to 30-min decay.
c Assuming 70% recycled and 30% discharged.
UFSAR/DAEC-1 T11.2-5 Revision 13 - 5/97 Table 11.2-3 RELEASE CONCENTRATIONS OF INDIVIDUAL ISOTOPES DURING DISCHARGE IN EFFLUENT LINE a Isotope Discharge Concentration Diluted to 1 x 10
-7 µCi/cm 3 Sr-89 7.8 x 10-10 Sr-90 6.0 x 10-11 Sr-91 7.5 x 10-9 Mo-99 5.1 x 10-9 I-131 3.4 x 10-9 I-133 1.6 x 10-8 I-135 1.0 x 10-8 Cs-134 4.1 x 10-11 Cs-137 6.2 x 10-11 Ba-140 2.2 x 10-9 Ce-144 9.0 x 10-12 Np-239 5.4 x 10-8 Co-58 6.7 x 10-10 Co-60 6.8 x 10-11 Total 1.0 x 10-7 a After mixture has been diluted to 10
-7 µCi/cm 3.
UFSAR/DAEC-1 T11.2-6 Revision 13 - 5/97 Table 11.2-4 ESTIMATE RADIOACTIVE LIQUID EFFLUENT QUANTITIES AND DOSE RATES FOR VARIOUS OFFGAS RELEASE RATES
Operational Release Basis Activity Discharge (Ci/yr) Annual Whole-Body Dose Rate a (mrem/yr) Offgas release rate of 250,000
µCi/sec at 30 min 2.0 0.2 Offgas release rate of 100,000
µCi/sec at 30 min 0.4 0.03 Offgas release rate of 25,000
µCi/sec at 30 min 0.1 0.006
a Based on annual average river flow of 3000 cfs.
UFSAR/DAEC-1 T11.2-7 Revision 13 - 5/97 Table 11.2-5 QUANTITIES OF PRINCIPAL RADIOSOTOPES ESTIMATED IN LIQUID WASTES SUBJECT TO RELEASE a (Offgas rate of 25,000
µCi/sec aged to 30 min)
Isotopes Daily Discharge
(µCi/day) Sr-89 1.0
Sr-90 0.1
Sr-91 10.0
Mo-99 7.0
I-131 6.0
I-133 30.0
I-135 20.0
Cs-134 0.07
Cs-137 0.1
Ba-140 4.0
Ce-144 0.01
Np-239 70.0
Co-58 7.0
Co-60 0.7
Total 151.98 (round to 150)
a Based on 30% floor drain, 100% detergent drain, and 100% chemical waste to discharge.
UFSAR/DAEC-1 T11.2-8 Revision 17 - 10/03 Table 11.2-6 Page 1 of 2 ISOTOPIC ACTIVITY INVENTORY (µCi) Isotope
Detergent Drain Tank (each tank)
Waste Sludge Tank (each tank) Waste Sample Tank (each tank)
Floor Drain Sample Tank Waste Surge Tank (40,000 Gal)
WasteSurge Tank (70,000 Gal)
Br 19.5 71.8 0.35 195,000 341,250Br 36.0 6.4 x 10
-4 3.1 x 10-6 360,000 630,000Br 21.0 - - 210,000 367,500I-131 - 18.0 1,965 9.8 180,000 315,000I-132 - 165 491 2.5 1,650,000 2,887,500I-133 - 120 9,072 45.4 1,200,000 2,100,000I-134 - 315 2.6 0.013 3,150,000 5,512,500I-135 - 180 5,670 29.1 1,800,000 3,150,000Sr 3.9 454 2.19 39,000 68,250Sr 0.3 33 0.17 3,000 5,250Sr 90 4,158 21.5 900,000 1,575,000Sr 142.5 756 3.7 1,425,000 2,493,750Zr 0.05 5.67 0.03 525 919Zr 0.04 0.29 0.014 405 709Nb 0.05 0.06 0.03 525 919Mo 28.3 2,873 14.4 285,000 498,750Tc-99m - 360 10,584 52.9 3,600,000 6,300,000Tc-101 - 180 1.06 x 10-11 5.3 x 10-14 1,800,000 3,150,000Ru-103 - 0.026 2.8 0.014 255 446Ru-106 - 0.003 37.8 0.002 33 58Te-129m - 0.05 0.57 0.029 510 893Te-132 - 63 6,426 32.1 630,000 1,102,500Cs-134 7.56 0.21 23.4 0.117 2,100 3,675Cs-136 - 0.135 15.1 0.076 1,350 2,363Cs-137 22.68 0.315 35.5 0.178 3,150 5,513Cs-138 - 240 .005 2.53 x 10
-5 2,400,000 4,200,000Ba-139 - 210 56.7 0.29 2,100,000 3,675,000Ba-140 - 11.55 1,285 6.4 115,500 202,125Ba-141 - 225 3.8 x 10
-8 1.89 x 10-10 2,250,000 3,937,500Ba-142 - 225 1.36 x 10-16 6.8 x 10-19 2,250,000 3,937,500Ce-141 - 0.05 5.67 0.028 510 893Ce-143 - 0.045 4.2 0.02 450 788Ce-144 - 0.045 5.3 0.026 450 788Pr-143 - 0.05 5.3 0.027 495 866
UFSAR/DAEC-1 T11.2-9 Revision 17 - 10/03 Table 11.2-6 Page 2 of 2 ISOTOPIC ACTIVITY INVENTORY (µcI) Isotope
Detergent
Drain Tank (each tank)
Waste Sludge Tank (each tank)
Waste Sample Tank (each tank)
Floor Drain Sample Tank
Waste Surge Tank (40,000 Gal)
Waste Surge Tank (70,000 Gal)
Nd-147 0.018 1.97 0.0098 180 315Np-239 - 315 31,000 155 3,150,000 5,512,500Na 1.35 90.7 0.45 13,500 23,625P 0.014 1.5 0.0076 135 236Cr 0.35 38 0.19 3,450 6,038Mn 0.027 3.1 0.155 270 473Mn 34.5 151 0.76 345,000 603,750Co-58 3.78 3.45 378 1.93 34,500 60,375Co-60 3.78 0.345 38 0.193 3,450 6,038Fe 0.054 6.04 0.03 540 945Ni 2.1 9.1 0.045 21,000 36,750Zn 6.8 x 10
-4 0.076 3.8 x 10
-4 6.75 11.81Zn-69m - 0.021 1.29 0.006 210 368Ag-110m - 0.04 4.5 0.023 405 709W-187 - 2.1 174 0.87 21,000 36,750
UFSAR/DAEC-1 11.3-1 Revision 13 - 5/97 11.3 GASEOUS WASTE MANAGEMENT SYSTEM
11.3.1 DESIGN BASES
11.3.1.1 Power Generation Objectives
The power generation objectives of the gaseous radioactive waste system are to process and control the release of noble radiogases to the environs so as to limit the annual average exposure to an individual at any point on the site boundary to a maximum of 10 mRad/yr to the
total body. This design objective for the DAEC is as low as reasonably achievable.
The design objective for radioactive halogens and particulates with half-lives longer than
8 days is stated in Section 11A.3.3.2.
The corresponding release rates are as stated in the Technical Specifications.
11.3.1.2 Power Generation Design Basis
The gaseous radioactive waste system is designed to limit offsite doses from routine plant releases to the lowest practicable level. The offgas system is designed to provide adequate time for corrective action to limit the activity release rates should they approach established limits.
The design basis for this system is a noble gas input equivalent to an annual average offgas rate (based on 30-min decay) of 25,000 µCi/sec with a diffusion mixture. Table 11.3-1 indicates the noble gas activity from the steam jet air ejector system referenced to 30 min after exiting from the reactor for an offgas rate of 100,000 µCi/sec.
Normally air inleakage is expected to be approximately 7 cfm (at 130°F, 1 atm) per condenser shell. Leakage from two condenser shells corrected to standard conditions gives 12.3 scfm. However, the design leakage of the plant is conservatively taken as 18.5 scfm.
Additonally, air or oxygen is injected to the offgas system upstream of the catalytic recombiner to completely recombine excess hydrogen from the Hydrogen Water Chemistry System.
UFSAR/DAEC-1 11.3-2 Revision 13 - 5/97 The effect of air inleakage and injection on system performance is shown below:
Condenser Curie Reduction Factor Relative
Air Inleakage & Injection to 100,000 µCi/Sec at 30 Min (scfm)
(12 Beds) 18.5 61
20 50
30 22
40 14
50 10.5
60 8.5
There is an oil-free air supply that bleeds into the system during startup. Its flow rate is 56.7 lb/hr, which is stopped after the recombiner comes up to temperature.
The isotopic release from the gland seal exhauster to the environment is presented in Table 11.3-2. The values in the table are based on an effective decay time of 1.75 min and take into consideration that an average of 0.1% of the total steam flow is routed to the gland seals.
The values in the table are also based on the assumption that the offgas release rate from the main steam turbine condenser is equivalent to a release rate of 100,000 µCi/sec after an effective holdup time of 30 min. The calculated annual dose rate from the gland seal noble gas effluent corresponding to an offgas rate of 100,000 µCi/sec (reference 30-min holdup) is 0.42 mrem/yr at
the site boundary.
The gland seal noble gas particulate daughters produced as a consequence of radioactive
decay are presented in Table 11.3-3. Because of di fferent half-lives of the parents and daughters, the location of maximum daughter concentration will vary for each isotope. The radiological effects of the daughter products are expressed in terms of X i/MPC i for each isotope. As noted in Table 11.3-3, there is no daughter product that exceeds 1.9 x 10
-5 of the allowable MPC air. Therefore, it is unnecessary to provide additional holdup or filtration of the gland seal noble gas
effluent.
Shielding is provided as necessary for process piping and equipment.
11.3.1.3 Safety Design Basis
The gaseous radwaste system is designed so that any quantities of gaseous radwastes inadvertently released result in radiation levels within annual exposure limits of 10 CFR 20.
UFSAR/DAEC-1 11.3-3 Revision 15 - 5/00 11.3.2 SYSTEM DESCRIPTION
11.3.2.1 Process Description
The offgas treatment system shown in Figure 11.3-1 uses a high-temperature catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen from the air ejector system. After chilling to strip the condensibles and reduce the volume, the remaining noncondensibles (principally kryptons, xenons, and air) are delayed in a 30-min holdup system
cooled to a dewpoint of 45°F with a chilled glyc ol cooler, passed through a de-entrainer, heated to 74°F (relative humidity of 35%), and passed through a HEPA filter before reaching the adsorption bed. The charcoal adsorption bed, operating in a constant temperature vault, selectively adsorbs and delays the xenons and kryptons from the bulk carrier gas (principally air). This delay on the charcoal permits the xenon and krypton to decay in place. This system results in a reduction of the offgas activity released by a factor of approximately 61 relative to a 30-min holdup system and based on a diffusion mixture.
The design of the offgas system incorporates an automatic loop seal isolation system that monitors system pressure at the 37-sec holdup volume ahead of the recombiner and in the 30-min holdup volume downstream of the recombiner. These isolation setpoints are set at 4.0 psig and 4.5 psig, respectively, with associated control room indication.
In the event of a loop seal liquid loss, leaking gas mixtures would be alarmed in the control room through the reactor building ventilation stack gaseous monitors, reactor building ventilation radiation monitors, ventilation shaft radiation monitors, offgas post-treatment radiation monitors, and offgas stack radiation monitors.
On the isolation of the offgas loop seals, the normal operating procedure is to keep loop
seals isolated until they naturally fill through condensation buildup.
In the event of an inadvertent offgas system explosion, the site preparedness plan and implementing procedures provide adequate guidance for proper response.
During the review of off-normal offgas system operation, one condition was identified that could lead to the accumulation of an explosive mixture. During normal operation, offgas is diluted in jet compressor . In the event of the loss of dilution steam, flow valveis automatically shut, isolating the jet compressor. This causes a pressure transient in the system that actuates several pressure switches, which in turn automatically close the isolation valves on the system loop seals.
UFSAR/DAEC-1 11.3-4 Revision 14 - 11/98 Two loop seals are located upstream of . In the event that insufficient water is available in one of the loop seals upstream of , the loop seal may blow before sufficient pressure to actuate the system pressure switches develops. Under these conditions, undiluted offgas would be discharged to the turbine building equipment drain sump or the offgas retention building equipment drain sump. Considering normal ventilation flow, the average hydrogen concentration in each area would be 16.7% and 13.0% by volume, respectively.
No other conditions have been identified that could result in the accumulation of sufficient hydrogen to form an explosive mixture. The review considered loss of ventilation flow, loss of dilution steam, lost loop seals, blow n rupture disks, and leakage of offgas into isolated portions of the system. Bypassing the recombiners was not considered as this is not
applicable to the DAEC design.
To ensure that the conditions identified above do not occur, the system has been designed in the following manner:
- 1. Water fill lines are installed to all loop seals that do not presently have fill capability.
- 2. System logics are designed to automatically isolate the two loop seals upstream of on the closure of .
The adsorption of noble gases on charcoal depends on gas flow rate, holdup time, mass of charcoal, and a gas-unique coefficient known as the dynamic adsorption coefficient. The parametric interrelationships and governing equations are well proven from 3 yr of operation of a similar unit at .
The basis for these coefficients and supporting experimental data are discussed in a proprietary document submitted with Amendment 1, May 1972, in response to an AEC question.
11.3.2.2 Equipment Description
The design of the DAEC gaseous radwaste system incorporates a catalytic recombiner and a 12-bed charcoal adsorber system. The design fully satisfies the requirements that releases
of radioactivity be reduced to the lowest practicable level.
UFSAR/DAEC-1 11.3-5 Revision 12 - 10/95 The piping and instrumentation drawings for the gaseous waste systems are included as Figures 11.3-2 and 11.3-3. The design codes for piping in these systems are provided by the line classification numbers given in the drawings. The line classification numbers are defined in Section 3.2.2.2.
System components and equipment that serve as pressure boundaries for the offgas system are fabricated in accordance with ASME Code,Section III,
Class 3.
The description of offgas system major equipment items is as follows:
- 1. Offgas Preheaters
Construction: Stainless steel tubes and carbon steel shell.
Design pressure: 350-psig shell, 1000-psig tube.
Design temperature: 400°F shell, 575°F tube.
- 2. Catalytic Recombiners
Construction: Stainless steel cartridge, lo w-alloy steel shell. Catalyst cartridge containing a precious metal catalyst on nichrome strips of porous, nondusting ceramic.
Catalyst cartridge to be replaceable without removing vessel. Design pressure: 350 psig.
Design temperature: 900°F.
- 3. Offgas Condenser
Construction: Low-alloy steel shell, stainless steel tubes.
Design pressure: 350-psig shell, 250-psig tube.
Design temperature: 900°F.
- 4. Water Separator
Construction: Carbon steel shell, stainless steel wire mesh.
Design pressure: 350 psig.
Design temperature: 250°F.
UFSAR/DAEC-1 11.3-6 Revision 12 - 10/95 5. Cooler Condenser
Construction: Stainless steel shell, stainless steel tubes.
Design pressure: 100-psig tube, 350-psig shell.
Design temperature: 50°F tube, 150°F shell.
- 6. Moisture Separators (Downstream of Cooler Condenser)
Construction: Carbon steel shell, stainless steel wire mesh.
Design pressure: 350 psig.
Design temperature: 150°F.
- 7. Offgas Reheater
Construction: Carbon steel pipe. Electrically heated.
- 8. Glycol Storage Tank
Construction: Carbon steel. Capacity is 3000 gal.
Water-filled hydrostatic design pressure.
Design temperature: 0°F.
- 9. Glycol Solution Refrigerators and Motor Drives
Construction: Conventional refrigeration units.
Glycol exit solution temperature: 35°F.
- 10. Glycol Pumps and Motor Drives
Construction: Cast iron, 3-in. connections, 50 ft.
Design temperature: 0°F.
- 11. Prefilters and After Filters
Construction: Carbon steel shell. High-efficiency, moisture-resistant filter element.
Flanged shell. Design pressure: 350 psig.
Design temperature: 130°F.
UFSAR/DAEC-1 11.3-7 Revision 12 - 10/95 12. Carbon Bed Adsorbers (12 Beds)
Construction: Carbon steel. Four ft outside diameter x 21 ft vessels, each with a 19-ft packed section containing ~3 tons of 8-14 mesh carbon (~200 ft 3 of charcoal) Columbia G or equivalent. Design pressure: 350 psig.
Design temperature: 130°F.
Flow channeling and bed settling are minimized in the charcoal adsorber vessels by the following design considerations:
- a. Charcoal adsorber beds are installed for vertical flow of the process gas stream.
- b. The first three charcoal beds in each parallel pathway have piping arrangements that cause up-flow from vessel bottom to vessel top of the process gas.
- c. As shown in Figures 11.3-4 and 5, toroidal-shaped flow-distribution rings are positioned at the bottom and top ends of each charcoal vessel to enhance the
process gas flow pattern. For the first three charcoal vessels in the process stream, the gas enters the distribution ring nozzle at the bottom of the vessel and
flows through a distribution torus and out through 251 one-in. holes and 3 layers of screen on the bottom of the torus; the gas then flows upward around both
outside walls of the distribution torus and through the charcoal to the upper torus
region. In the upper vessel region, the charcoal-filtered gas enters the distribution
torus through 3 layers of screen and 251 holes on the top of this torus and
proceeds out of the distribution ring nozzle to the next charcoal vessel in the
process flow path.
Process gas enters the top distribution ring and exits out the lower ring nozzle for charcoal vessels other than the first three in each parallel path.
- 13. Offgas Jet Compressor
Construction: Carbon steel body.
Design pressure: 2150 psig.
Design temperature: 400°F.
Flow rate: 4624 lb/hr.
UFSAR/DAEC-1 11.3-8 Revision 15 - 5/00 The ventilation system for each DAEC building that can be expected to contain radioactive materials is described in Section 9.4.
The main condenser gas removal system is described in Section 10.4.2. The main steam line isolation valve leakage treatment path is described in Section 6.7.
11.3.2.3 Instrumentation and Control
The radiation levels at the air ejector offgas discharge line and after the offgas treatment system are continuously monitored by pairs of detectors. This system is also monitored by flow and temperature instrumentation and hydrogen analyzers to ensure proper operation and control and to ensure that hydrogen concentration is maintained below the flammable limit. Process radiation instrumentation is described in Section 11.5. Table 11.3-4 lists process instrument alarms.
11.3.2.4 Safety Evaluation
The decay time provided by the 30-min holdup pipe and the long-delay charcoal
adsorbers is established to provide for radioactive decay of the activation gases and fission gases in the main condenser offgas. The adsorbers provide a 15-day xenon and a 19-hr krypton holdup. The daughter products that are solids are removed by filtration following the 30-min
holdup and/or are retained on the charcoal. Final filtration of the charcoal adsorber effluent precludes the escape of charcoal fines that contain radioactive materials. Thus, there is virtually
no particulate activity release.
Iodine input into the offgas system is small because of its retention in the reactor water and condensate. The charcoal effectively removes the iodine entering the system by adsorption
and prevents its release.
Radiation monitors at the recombiner outlet continuously monitor radioactivity releases from the reactor and, therefore, continuously monitor the degrees of fuel leakage and input to the charcoal adsorbers. Radiation monitors are used to provide an alarm on high radiation in the offgas. Two radiation monitors are provided at the outlet of the charcoal adsorbers to continuously monitor the release rate from the adsorber beds. The radiation monitors are further described in Section 11.5.
UFSAR/DAEC-1 11.3-9 Revision 12 - 10/95 Shielding is provided for offgas system equipment to maintain safe radiation exposure levels for plant personnel. The equipment is principally operated from the control room.
The charcoal adsorbers operate at essentially room temperature so that upon system shutdown, radioactive gases in the adsorbers are subject to the same holdup time as during normal operation, even in the presence of continued air flow. Therefore, the radioactive materials are not subject to an accidental release evaluation. The charcoal adsorbers are designed to limit the temperature of the charcoal to well below the charcoal ignition temperature, precluding overheating or fire and consequent escape of radioactive materials. The adsorbers are located in a shielded room and maintained at a constant temperature by an air conditioning system, which removes the decay heat generated in the adsorbers. The failure of the air conditioning system will cause an alarm in the control room. In addition, a radiation monitor is provided to monitor the radiation level in the charcoal bed vault. High radiation causes an alarm in the control room.
The hydrogen concentration of the gases from the air ejector is maintained below the flammable limit by providing adequate steam flow for dilution at all times. This steam flow rate is monitored and alarmed. The preheaters are heated by steam rather than electricity to eliminate
the presence of potential ignition sources and to limit the temperature of the gases in the event of cessation of gas flow. The recombiner temperatures are monitored and alarmed to indicate any deterioration of performance. A hydrogen analyzer downstream of the recombiners provides an additional check.
In addition, this system is designed to be explosion resistant in the unlikely event a combustible mixture exists.
The gaseous radwaste system piping and equipment is designed to be explosion resistant by employing design methods for circular-section steel systems to contain explosions of near stoichiometric mixtures of gaseous hydrogen and oxygen.
The design method used a static analysis with dynamic materials properties. More exact rigorous dynamic analyses were conducted on selected designs with the results confirming that the static method was sufficiently conservative to use for offgas system design. Ratios of maximum pressure to initial pressure (before an explosion) ranging from 17 to 170 were used to determine the maximum peak pressure (Pm) in the component under analysis. Wall thicknesses for the particular component were then computed using Pm as the pressure load.
UFSAR/DAEC-1 11.3-10 Revision 12 - 10/95 An equivalent detonation-containing-static-pressure was then derived for which the component could be "rated" on the basis of the wall thickness calculated per the above
procedure.
This equivalent pressure was computed from one of various code equations on the basis of the purchase order date of the given equipment and the applicable codes in effect at that time.
The air ejector offgas system operates at a pressure of about 5 psig or less so that the differential pressure available to cause leakage of radioactive gases is small. To reduce the possibility of leakage of radioactive gases, the system is welded wherever possible and bellows seal valve stems or equivalent are used.
Operational control is maintained by the use of radiation monitors to ensure that the release rate is within the established limits. Provision is also made for sampling and periodic
analysis of the influent and effluent gases for purposes of determining their composition. This information is used in the calibration of the monitors and in relating the release to the environmental dose. The operator is in full control of the system at all times.
Table 11.3-5 contains a detailed malfunction analysis indicating consequences of the failure of various components of the system and design precautions taken to prevent such failures.
11.3.2.5 Inspection and Testing
The gaseous waste disposal systems are used on a routine basis and do not require specific testing to ensure operability. Calibration and maintenance of monitoring equipment is done on a specific schedule and on indication of malfunction.
The particulate filters are tested when they are changed using a dioctylpthalate (DOP) smoke test or equivalent.
Experience with BWRs has shown that the calibration of the offgas and effluent monitors
changes with isotopic content. Isotopic content can change depending on the presence or
absence of fuel cladding leaks in the reactor and the nature of the leaks. Because of this, the monitors are calibrated using grab samples periodically and at any time there appears to be a
significant change.
UFSAR/DAEC-1 11.3-11 Revision 12 - 10/95 11.3.3 RADIOACTIVE RELEASES
See Appendix 11A for gaseous release rate limit calculations.
All potential sources of radioactivity released to the environment because of gaseous
waste and ventilation will be via the offgas stack, These release point s are shown in Figures 1.2-5, 6, 7, 8, 12, and 14, and Figure 11.3-6.
Isotopic release rates for expected normal operating conditions and design-basis conditions have been estimated for air ejector o ffgas, gland seal, high-pressure coolant injection (HPCI) testing, miscellaneous plant vent losses (including turbine building steam leakage and reactor building primary coolant leakage), and primary containment purge. The assumed fuel failure during normal operation is associated with a 25,000-µCi/sec offgas release rate at 30-min decay. Design-basis estimates are based on a 250,000-µCi/sec offgas rate at 30-min decay. On the basis of operating BWR fuel performance, maximum expected release rates would be less than the design basis of 250,000 µCi/sec annual average (reference 30-min decay).
The total amount of radioactive noble gases estimated to be stored in the first charcoal
bed in the filter train is 5.2 x 10 8 µCi after 10 yr of operation. This estimate is based on residence time assumptions of 1.62 hr for krypton, 1.22 days for xenon, and 4.42 x 10 2 sec for N 13 and an offgas average release rate of 25,000 µCi/sec at 30-min holdup time.
Steam Jet Air Ejector Offgas
The steam jet air ejector offgas exhausts through the catalytic recombiner and 12-bed charcoal adsorption system before exiting from the main stack.
Table 11.3-1 gives the estimated inventory of radioactivity in the main condenser offgas stream corresponding to 100,000 µCi/sec from the 30-min holdup line. As the expected normal release is approximately 25,000 µCi/sec from the 30-min holdup line, the inventory will be approximately 25% of the values given in Table 11.3-1.
Table 11.3-6 lists the noble gas radioisotopes discharged from the air ejector offgas system during normal operation. Design-basis estimates would be a factor of 10 greater.
UFSAR/DAEC-1 11.3-12 Revision 12 - 10/95 The maximum calculated dose to a point on the site boundary for the isotopic release shown in Table 11.3-6, assuming a 365-day continuous exposure, is 0.24 mrem/yr, which occurs
at a distance of . This result is based on a condenser air inleakage rate of 18.5 scfm, discharge from the 100-m stack, and DAEC onsite meteorology data taken at the 156-ft level. The gamma dose model used is described in Section 11A.1.
Radioiodine was assumed to be completely captured in the charcoal adsorption system.
Gland Seal
Process steam is supplied to the shaft seals of the main turbine for the purpose of providing a positive seal on this system. The exhaust steam from the seal system is routed to the gland seal condenser with the noncondensibles being discharged to the environment via the
gland seal exhauster and a short delay line.
Hypothetical radiological effects for this system are based on the following assumptions.
Table 11.3-7 lists the radioisotopes discharged from the gland seal system during normal operation. Design-basis estimates would be a factor of 10 greater.
The maximum calculated dose to a point on the site boundary for the isotopic release shown in Table 11.3-7 is 0.12 mrem/yr, which occurs at a distance of
. This result is based on a steam flow to the gland seal system of 0.1% of the process steam flow, an approximately 2-min delay time between steam leaving the reactor vessel and subsequent release to the environment via the 100-m stack, and DAEC onsite meteorology data taken at the 156-ft level. The gamma dose model used is described in Section
11A.1.
Radioiodine carryover is assumed to be approximately 1%, of which approximately 2%
enters the gland seal system and releases to the environment.
HPCI Testing
Periodic testing of the HPCI system may also result in minor releases of radioactivity to the environment. The only activity that would be released to the environment would be that associated with the HPCI gland seal steam. The process steam used to drive the high-pressure coolant injection is routed to the pressure suppression pool and is considered in the containment purge calculation. The HPCI gland seal steam is discharged to the standby gas treatment system (SGTS) and then to the environment via the plant stack.
UFSAR/DAEC-1 11.3-13 Revision 12 - 10/95 Table 11.3-8 lists the noble gas radioisotopes discharged from the HPCI gland seal system during HPCI operation with expected normal operation fuel defects. Design-basis estimates would be a factor of 10 greater.
The maximum calculated dose to a point on the site boundary for the isotopic release shown in Table 11.3-8 assuming an HPCI test once every 3 months of 1-hr duration is 2.3 x 10
-5 mrem per release period. Steam flow to the HPCI turbine gland seals was assumed to be 500 lb/hr. The SGTS filter efficiency for noble gases was assumed to be 0% with an approximately 2-min delay time between steam leaving the reactor vessel and subsequent release to the environment. The meteorology at the time of release was conservatively assumed to be unstable with a 2 m/sec wind speed. The exposure is a maximum in the at a distance of .
Radioiodine release rate would be insignificant because of the 99.9% SGTS filter
efficiency.
Building Ventilation
Reactor building fission product releases to ventilation air are based on a continuous unidentified 1.0-gpm reactor water leak, for which a halogen decontamination factor of 10 3 is assumed. The escaped fission products are assumed to pass directly to the building vent without filtration. Meteorological assumptions for building ventilation releases were treated as described in Section 11A.3.2.1.1 to determine fission product dispersion to the site boundary.
Turbine building fission product releases to ventilation air are based on a continuous 5-gpm steam leak. An iodine decontamination factor of 80 is assumed for internal steam generation, while an iodine decontamination factor of 1.0 is assumed for the turbine building.
The escaped fission products are assumed to pass directly to the building vent without filtration.
Meteorological assumptions for building ventilation releases were treated as described in Section 11A.3.2.1.1 to determine fission product dispersion to the site boundary.
Table 11.3-9 gives the combined reactor building and turbine building isotope release rates corresponding to normal operation. It is assumed that the release rate from the LLRPSF is negligible in comparison. Design-basis estimates would be a factor of 10 greater. (Reactor building primary water leakage contributes negligible noble gas losses to the overall noble gas
release and less than 10% to the total iodine release.) The resulting annual average exposure at the worst site boundary location is 0.11 mrem/yr.
UFSAR/DAEC-1 11.3-14 Revision 12 - 10/95 Fission product releases from primary containment purge are based on an assumed primary system leakage of 0.5-gpm steam and 0.5-gpm unidentified reactor water. Radiological equilibrium (except for Kr-85) is assumed to occur in containment. An iodine decontamination factor of 80 is assumed for internal steam generation, and an iodine decontamination factor of 10 is used for reactor water leakage. Four controlled purges per year are assumed to occur, with the exhaust passing through the standby gas treatment system (deep-bed charcoal filter 99.9%
efficient for iodine) before being released from the 100-m-high plant main stack. A conservative stack c/Q of 2.3 x 10
-7 sec/m 3 (at the site boundary) for short-term releases was determined for the 90% probability level. Table 11.3-10 gives the total noble gas and iodine releases, in microcuries, for four containment purges corresponding to normal operation fuel defects. Design-basis estimates would be a factor of 10 greater. The calculated site boundary
whole-body exposure for the isotopic release shown in Table 11.3-10 is less than 1 x 10
-3 mrem/yr.
UFSAR/DAEC-1 T11.3-1 Revision 13 - 5/97 Table 11.3-1 ESTIMATED ISOTOPIC RELEASE RATES a FROM THE STEAM JET AIR EJECTOR SYSTEM (Corresponding to 100,000
µCi/sec release from 30-min holdup line)
Isotope Discharge from 30-Min Holdup Line
(µCi/sec) Discharge from Charcoal Adsorbers (12 beds) (µCi/sec) Kr-83m 2,900 2.0 Kr-85m 5,600 266
Kr-85 10-20 b 10-20 b Kr-87 15,000 0.04 Kr-88 18,000 144 Kr-89 180 0 Xe-131m 15 6.5 Xe-133m 280 3.0 Xe-133 8,200 1190 Xe-135m 6,900 0
Xe-135 22,000 0 Xe-137 670 0
Xe-138 21,000 0 Halides - - Insignificant Total (approx.) 100,000 1630
a Based on diffusion mixture.
b Estimated from experimental observations.
UFSAR/DAEC-1 T11.3-2 Revision 13 - 5/97 Table 11.3-2 NOBLE GAS RELEASE RATE TO ENVIRONS FROM TURBINE GLAND SEAL EXHAUSTER SYSTEM a Isotope Release Rate to Environment
(µCi/sec) Kr-83m 3.4 Kr-85m 6.1
Kr-85 0.02
Kr-87 19.0
Kr-88 20.0
Kr-89 87.0
Kr-90 30.0
Kr-91 0.07
Xe-131 0.015 Xe-133m 0.28
Xe-133 8.2 Xe-135m 24.0
Xe-135 22.0
Xe-137 111.0
Xe-138 83.0
Xe-139 46.0
Xe-140 17.0
a Based on 100,000
µCi/sec at 30-min decay diffusion mixture, 0.1% to gland seal system.
UFSAR/DAEC-1 T11.3-3 Revision 13 - 5/97 Table 11.3-3 RADIOACTIVE PARTICULATE DAUGHTER BUILDUP FROM GLAND SEAL NOBLE GAS EFFLUENT a RATIO OF X i/MPC i (Annual average) VERSUS DISTANCE IN WORST SECTOR Distance b (m) Isotope 100 300 800 1200 3000 5000 8000 12000 Rb-87 6.0-30 3.5-22 1.1-20 2.3-30 2.6-20 2.2-20 2.1-20 1.9-20 Rb-88 5.3-15 2.7-7 6.9-6 1.2-5 8.1-6 4.8-6 2.6-6 1.5-6 Rb-89 1.5-14 5.9-7 9.4-6 1.5-5 8.8-6 3.5-6 1.1-6 3.6-7 Rb-90 2.6-15 2.9-7 2.2-6 3.1-6 5.8-7 7.9-8 8.4-9 9.9-10 Rb-91 1.1-18 3.0-10 1.7-9 1.8-9 1.6-10 1.6-11 1.1-12 7.8-14 Sr-89 5.2-17 4.4-9 5.1-8 2.2-7 1.9-7 1.3-7 1.1-7 9.9-8 Sr-90 4.0-15 1.3-7 3.3-6 5.4-6 3.8-6 2.1-6 1.4-6 1.1-6 Sr-91 3.7-20 1.4-12 2.8-11 4.4-11 2.8-11 1.4-11 7.6-12 3.8-12 Y-91m 2.6-23 2.1-15 6.3-14 1.4-13 1.4-13 1.0-13 8.2-14 6.1-14 Y-91 8.8-24 7.7-16 2.6-14 6.2-14 1.0-13 1.1-13 1.5-13 1.8-14 Cs-135 1.7-25 1.0-17 3.3-16 7.7-16 1.1-15 1.2-15 1.6-15 1.9-16 Cs-137 3.5-19 1.3-11 3.1-10 5.0-10 3.5-10 2.0-10 1.4-10 1.1-10 Cs-138 1.2-14 5.3-7 1.2-5 1.9-5 1.2-5 6.8-6 3.4-6 1.8-6 Cs-139 3.3-15 2.1-7 2.1-6 3.3-6 1.3-6 3.8-7 9.3-8 2.2-8 Cs-140 2.5-17 1.4-8 6.7-8 6.0-8 2.9-9 2.0-10 8.3-12 2.7-13 Ba-137m 1.2-20 6.2-13 7.1-12 2.4-11 1.8-11 1.2-11 8.6-12 7.1-12 Ba-139 1.4-16 6.7-9 1.5-7 2.5-7 1.7-7 9.7-8 4.8-8 2.6-8 Ba-140 2.5-19 1.0-11 1.9-10 3.0-10 1.8-10 9.4-11 6.2-11 5.0-11 La-140 2.1-22 1.7-14 5.6-13 7.5-13 2.3-12 2.5-12 3.6-12 4.7-12
a Based on annual average meterology from DAEC site, January 8, 1971 to January 8, 1972. b Second number in each column indicates exponent (e.g., 2.9-17 = 2.9 x 10-17).
UFSAR/DAEC-1 T11.3-4 Revision 13 - 5/97 Table 11.3-4 PROCESS INSTRUMENT ALARMS a
Main Control Room Parameter Indicated Recorded Alarms Preheater discharge temperature - low
x x Offgas line pressure - high x x Recombiner catalyst temperature - high/low x x Offgas condenser drain well (dual) level - high/low x Offgas condenser gas discharge temperature - high x Hydrogen analyzer (Condenser discharge) (dual) -
high x x Gas flow (offgas condenser discharge) - high/low x x Cooler condenser discharge temperature - high/low x x Glycol solution temperature - high/low x x x Prefilter P - high
x x Carbon bed temperature - high x x Carbon vault temperature - high/low x x x Postfilter P - high x x a Instrumentation elements Temperature: thermocouple Level: differential pressure diaphragm Hydrogen: thermal conductivity
Gas flow: flow orifice Differential pressure: differential pressure diaphragm UFSAR/DAEC-1 T11.3-5 Revision 13 - 5/97 Table 11.3-5 Page 1 of 5 EQUIPMENT MALFUNCTION ANALYSIS Equipment Items Malfunction Consequences Design Precautions Preheaters 1. Steam leak Would further dilute process offgas.
Steam compumption would increase.
Space preheated. 2. Low-pressure steam supply Recombiner performance would fall off at low-power level and hydrogen content of recombiner gas discharge
would increase, eventually to a combustible mixture. Low-temperature alarms on preheater exit and recombiner inlet.
Recombiner hydrogen analyzer. Recombiner 1. Catalyst gradually deactivates Temperature profile changes through catalyst. Eventually excess hydrogen would be detected by meter.
Eventually, the gas could be become combustible. Temperature probes in recombiner and hydrogen analyzer recombiner. 2. Catalyst gets wet at startup Hydrogen conversion falls off and hydrogen is detected by downstream
analyzers. Eventually the gas could become combustible. Condensate drains, temperature probes in recombiner. Air bleed system at startup. Recombiner thermal blanket. Spare recombiner
and heater. Hydrogen analyer. Recombiner
condenser Cooling water leak The coolant (service water) would leak to the process gas (shell) side.
This would be detected if drain well
liquid level increases. Moderate leakage would be of no concern from
a process standpoint.
None.
UFSAR/DAEC-1 T11.3-6 Revision 13 - 5/97 Table 11.3-5 Page 2 of 5 EQUIPMENT MALFUNCTION ANALYSIS Equipment Items Malfunction Consequences Design Precautions Drain Well Liquid level instruments fail If both drain valves fail to open, water will build up in the condenser and
pressure drop will increase. Two separate drain systems each provided with high- and low- level alarms. The high p, if not detected by instrumentation, could cause pressure buildup in the main condenser and
eventually a reactor trip.
If a drain valve fails to close, gas will recycle to the main condenser increasing the load on the steam jet air ejector causing back-pressure on the main
condenser, eventually causing a reactor
trip. Water separator Corrosion of wire mesh element Higher quanity of water collected in 30-min holdup line androuted to radwaste. Stainless steel mesh specified. 30-min holdup line Corrosion of line Leakage to soil of gaseous and liquid fission products.
Outside of pipe dipped and wrapped. Cooler condenser 1. Corrosion of bare tubes Glycol-water solution would leak into process (shell) side and be discharged to
clean radwaste. If not detected at
radwaste, the glycol solution would
discharge to the reactor condensate system. Stainless steel bare tube specified.
The inventory of glycol-water can be
observed in tank A002. Spare cooler
provided.
UFSAR/DAEC-1 T11.3-7 Revision 13 - 5/97 Table 11.3-5 Page 3 of 5 EQUIPMENT MALFUNCTION ANALYSIS Equipment Items Malfunction Consequences Design Precautions
- 2. Icing-up of bare tube Shell side of cooler could plug with ice, gradually building up pressure
drop. If this happens, the spare unit could be activated. Complete blockage of both units would increase main
condenser pressure leading to a reactor
trip. Design glycol-water solution temperature of 34-40
°F. Spare unit provided. Redundant temperature indication and alarm system.
Moisture separators Corrosion of wire mesh element Increased moisture would be retained in process gas routed to charcoal adsorbers. Over a long period of time, the charcoal performance could deteriorate owing to moisture pickup. Stainless steal mesh specified. Spare unit provided. Prefilters Hole in filter media More radioactivity would deposit on the charcoal in the first adsorber vessel
of the train. This would increase the
radiation level in the charcoal vault, making maintenance more difficult. P instrumentation provided. Spare unit provided.
Charcoal adsorbers Charcoal gets wet Charcoal performance will deteriorate gradually as charcoal gets wet. Holdup times for krypton and xenon will decrease and plant emissions will
increase. Highly instrumented mechanically simple gas dehumidification system with redundant equipment.
UFSAR/DAEC-1 T11.3-8 Revision 13 - 5/97 Table 11.3-5 Page 4 of 5 EQUIPMENT MALFUNCTION ANALYSIS Equipment Items Malfunction Consequences Design Precautions Vault air
conditioning
units Mechanical failure If ambient temperature exceeds about 80°F, increased emission could occur.
Spare refrigerator unit provided. If ambient temperature is below about 60°F, charcoal could pick up additional moisture. Vault temperature alarms provided. After filters Hole in filter media Probably of no real consequence. The charcoal media themselves should be a
good filter at the low air velocity. P instrumentation provided. Spare unit provided. Glycol refrig.
machines Mechanical failure If spare unit fails to operate, the glycol solution temperature will rise and the dehumidfication system performance
will deteriorate. This will cause gradual buildup of moisture on the
charcoal, with increased plant emissions.
Spare refrigerator provided. Glycol solution temperature alarms provided.
UFSAR/DAEC-1 T11.3-9 Revision 13 - 5/97 Table 11.3-5 Page 5 of 5 EQUIPMENT MALFUNCTION ANALYSIS Equipment Items Malfunction Consequences Design Precautions Steam jet
ejectors 1. Low flow of motive high-pressure steam When the hydrogen and oxygen concentrations exceed 4 and 6 volume percent, respectively, the process gas becomes combustible. Alarms provided for low steam flow and low steam pressure. Inadequate steam flow will cause overheating and deterioration of the
catalyst. Steam flow to be held at constant maximum flow regardless of plant power level. 2. Wear of steam supply nozzle of ejector Increased steam flow to recombiner.
This could reduce degree of recombination at low power levels.
UFSAR/DAEC-1 T11.3-10 Revision 13 - 5/97 Table 11.3-6 STEAM JET AIR EJECTOR ISOTOPIC RELEASE RATE (Corresponding to 25,000
µCi/sec at 30-min decay offgas rate)
Isotope Release Rate
(µCi/sec) Kr-83m 0.54 Kr-85m 66.5
Kr-85 5.95
Kr-87 0.092
Kr-88 36 Kr -
Kr -
Kr -
Xe-131m 1.615 Xe-133m 0.77
Xe-133 294.5 Xe-135m - -
Xe-135 - -
Xe-137 - -
Xe-138 - -
Xe-139 - -
Xe-140 - -
UFSAR/DAEC-1 T11.3-11 Revision 13 - 5/97 Table 11.3-7 TURBINE GLAND SEAL ISOTOPIC RELEASE RATE (Corresponding to 25,000
µCi/sec at 30-min decay offgas rate)
Isotope Release Rate
(µCi/sec) Kr-83m 0.86 Kr-85m 1.54
Kr-85 0.006
Kr-87 4.85
Kr-88 5.
Kr-89 21.8
Kr-90 7.9
Kr-91 0.0175 Xe-131m 0.0038 Xe-133m 0.0695
Xe-133 2.045 Xe-135m 6.1
Xe-135 5.55
Xe-137 27.
Xe-138 20.65
Xe-139 11.35
Xe-140 0.435
I-131 0.00135
I-132 0.012
I-133 0.009
I-134 0.0235
I-135 0.013 UFSAR/DAEC-1 T11.3-12 Revision 13 - 5/97 Table 11.3-8 HPCI GLAND ISOTOPIC RELEASE RATE (Corresponding to 25,000
µCi/sec at 30-min decay offgas rate)
Isotope Release Rate a (µCi/sec) Kr-83m 0.059 Kr-85m 0.106 Kr -
Kr-87 0.334
Kr-88 0.344
Kr-89 1.5
Kr-90 0.51
Kr-91 0.0014 Xe-131m - -
Xe-133m 0.0048
Xe-133 0.141 Xe-135m 0.42
Xe-135 0.382
Xe-137 1.86
Xe-138 1.41
Xe-139 0.78
Xe-140 0.03 a Release rate during operation of HPCI turbine.
UFSAR/DAEC-1 T11.3-13 Revision 13 - 5/97 Table 11.3-9 REACTOR BUILDING AND TURBINE BUILDING COMBINED VENTILATION RELEASE RATE (Corresponding to 25,000
µCi/sec at 30-min decay offgas rate)
Isotope Release Rate
(µCi/sec) Kr-83m 0.34 Kr-85m 0.6
Kr-85 0.0024
Kr-87 1.9
Kr-88 4.3
Kr-89 2.0 Xe-131m 0.0015 Xe-133m 0.028
Xe-133 0.82 Xe-135m 2.1
ú Xe-135 2.2 Xe-137 6.3
Xe-138 7.1
I-131 0.0025
I-132 0.023
I-133 0.015
I-134 0.048
I-135 0.025 UFSAR/DAEC-1 T11.3-14 Revision 13 - 5/97 Table 11.3-10 DRYWELL PURGE ISOTOPIC RELEASE (Four purge/per year, corresponding to 25,000
µCi/sec at 30-min decay offgas rate)
Isotope Release (µCi/sec) Kr-83m 3.1 x 10 2 Kr-85m 1.3 x 10 3 Kr-85 1.4 x 10 3 Kr-87 1.3 x 10 3 Kr-88 2.8 x 10 3 Xe-131m 2.1 x 10 2 Xe-133m 7.7 x 10 2 Xe-133 5.1 x 10 4 Xe-135m 3.3 x 10 2 Xe-135 9.9 x 10 3 Xe-138 1.3 x 10 3 I-131 1.0 x 10 4 I-132 1.1 x 10 3 I-133 7.5 x 10 3 I-134 8.4 x 10 2 I-135 4.3 x 10 2
- 0****0**DUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTCharcoalAdsorberVesselAssemblyFigure11.3-4
,TTOMSCREEN251-1in,diaHOLESBBARSEQUALLYSPACEDASSHOWN---+I-TOPSCREEN)DLESCREENDUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTDistributorRingFigure11.3-5
UFSAR/DAEC - 1 T11.4-1 Revision 21 - 5/11 Table 11.4-1 MAXIMUM ISOTOPIC ACTIVITY INVENTORIES IN SOLID WASTES Initial Design Average Isotopic Curies of Shipped Waste: 2000-2009
Isotope Activity (Ci/yr) Isotope Activity (Ci/yr) Sr-89 Sr-90 Sr-91 Zr-95 Nb-95 Mo-99 Tc-99m Ru-103 Ru-106 Te-129 Te-132 Cs-134 Cs-136 Cs-137 Ba-140 Ce-141 Ce-144 Np-239 I-131 I-133 I-135 Mn-54 Cr-51 Fe-59 Co-58 Ag-110m Co-60 1.9E+02 1.3E+02 1.4E+01 9.5E+00 3.5E+00 1.3E+02 1.2E+01 2.8E+00 1.2E+00 4.7E+00 3.4E+02 8.3E+01 2.9E+00 1.3E+02 2.6E+02 4.4E+00 2.6E+01 1.2E+03 2.0E+02 8.9E+01 8.0E+00 1.3E+01 2.6E+01 8.4E+00 8.9E+02 1.8E+01 2.0E+02 Co-60 Cs-137 Fe-55 Mn-54 Ni-63 Zn-65 7.2E+02 1.0E+00 9.2E+02 3.5E+01 9.6E+01 2.3E+01 Total 4.0E+03 Total 1.8E+03
Notes: 1. Isotopic curies of initial design were estimated on the basis of plant operation at the stack release rate of 100,000 µCi/sec (after 30-min holdup) annual average throughout the year.
- 2. Decay is applied for the duration of the design storage within the systems but does not account for decay occurring after drumming and before shipment.
- 3. Those isotopes with annual curies less than 1 have been omitted from the list.
- 4. The average isotopic curies of shipped waste for 2000-2009 includes a fuel pool clean up campaign, which greatly increases the average curies shipped in this time period.
µ
µ
T11.5-1 Revision 19 - 9/07 Table 11.5-1 CHARACTERISTICS OF PROCESS RADIATION MONITORING SYSTEMS Monitoring System Instrument Range a Instrument Scale Upscale Trips per Channel Downscale Trips per Channel Inoperative Trips per Channel Main steam line 1.0 to 10 6 mR/hr 6-decade log 1 1 b 1 b 1 Air ejector offgas (before treatment)
1.0 to 10 6 mR/hr 6-decade log 2 b 1 b 1 b Air ejector offgas (after treatment) 10-1 to 10 6 counts per
sec c 7-decade log 2 1 b 1 1 Carbon bed vault area radiation
1.0 to 10 6 mR/hr 6-decade log 1 b 1 b -- Offgas vent pipe 10
-1 to 10 6 counts per
sec c 7-decade log 1 1 b 1 b 1 b Liquid processes d 10-1 to 10 6 counts per
sec c 7-decade log 1 b ,e 1 b 1 b Reactor building vent shaft
0.01 to 10 4 mR/hr 6-decade log 1 1 -- Refuel pool ventilation
exhaust 0.01 to 100 mR/hr 4-decade log 1 b 1 b 1 f a Range of measurement is dependent on items such as the source of geometry, background radiation, shielding, energy levels, and method of sampling.
b Alarms only.
c Readout is dependent on the pulse height discriminator setting.
d Five process steam monitored.
e Upscale trip for Radwaste Effluent monitor only.
f Mode switch out of operate.
T11.5-2 Revision 14 - 11/98 Table 11.5-2 ENVIRONMENTAL AND POWER SUPPLY DESIGN CONDITIONS FOR PROCESS RADIATION MONITORING SYSTEM
Sensor Location Control Room Parameter Design Center Range Design Center Range Temperature
25°C 0°C to +60°C 25°C 5°C to +50°C Relative humidity 50% 20% to 98% 50% 20% to 90% Power, ac 115V 60 Hz
+/-10% +/-5% 115V 60 Hz +/-10% +/-5% Power, dc +24V
-24V +22V to +29V
-22V to -29V
+24V -24V +22V to +29V
-22V to -29V T11.5-3 Revision 14 - 11/98 Table 11.5-3 Sheet 1 of 2 CHARACTERISTICS OF EXTENDED RANGE AIRBORNE EFFLUENT MONITOR SYSTEM Characteristics Particulate Iodine Noble Gas REACTOR BUILDING VENTS 1, 2, AND 3 Measurement type
Collection only Collection only Gross beta Detector type Not applicable Not applicable Beta scintillator Geiger-Mueller counter (2 required)
Dynamic range, µCi/cm 3 Not applicable Not applicable 10
-7 to 10 5 Minimum detectable concentration (MDC), µCi/cm 3 Not applicable Not applicable 2 x 10
-7 Calibration Isotope Not applicable Not applicable Mixed noble gases a Xe 133 Maximum response time at MDC Not applicable Not applicable 1 min. Minimum filter efficiency 99% > 0.3µ 95% Not applicable Filter type HV-LB 5211 SAI Not applicable Analog type
-- -- 2 required Remote Indication
-- -- 1 required TURBINE BUILDING VENT Measurement type
Collection only Collection only Gross beta Detector type Not applicable Not applicable Beta scintillator Geiger-Mueller counter (2 required)
Dynamic range, µCi/cm 3 Not applicable Not applicable 10
-7 to 10 5 MDC, µCi/cm 3 Not applicable Not applicable 2 x 10
-7 Calibration Isotope Not applicable Not applicable Mixed noble gasses a Xe 133 Maximum response time at MDC Not applicable Not applicable 1 min. Minimum filter efficiency 99% > 0.3µ 95% Not applicable Filter type HV-LB 5211 SAI Not applicable Analog outputs
-- -- 2 required Remote Indication -- -- 1 required a Mixed noble gases for beta scintillator and Xe 133 for Geiger-Mueller counter.
T11.5-4 Revision 14 - 11/98 Table 11.5-3 Sheet 2 of 2 CHARACTERISTICS OF EXTENDED RANGE AIRBORNE EFFLUENT MONITOR SYSTEM Characteristics Particulate Iodine Noble Gas OFFGAS STACK Measurement type
Collection only Collection only Gross beta Detector type Not applicable Not applicable Beta scintillator Geiger-Mueller counter (2 required)
Dynamic range, µCi/cm 3 Not applicable Not applicable 10
-7 to 10 5 MDC, µCi/cm 3 Not applicable Not applicable 2 x 10
-7 Calibration Isotope Not applicable Not applicable Mixed noble gases a Xe 133 Maximum response time at MDC Not applicable Not applicable 1 min. Minimum filter efficiency 99% > 0.3
µ 95% Not applicable Filter type HV-LB 5211 SAI Not applicable Analog outputs
-- -- 2 required Remote Indication
-- -- 1 required LLRPSF EXHAUST STACK Measurement Type
Collection only Collection only Gross beta Detector Type Not applicable Not applicable Beta Scintillator Gieger-Mueller counter Dynamic Range, µCi/cm 3 Not applicable Not applicable 10
-7 to 10-1 MDC, µCi/cm 3 Not applicable Not applicable 2 x 10
-7 Calibration Isotope Not applicable Not applicable Xe 133 Maximum response time at MDC Not applicable Not applicable 1 min. Minimum filter efficiency 99% > 0.3
µ 95% NA Filter type HV-LB 5211 SAI NA Analog outputs
-- -- 1 required Remote indication -- -- 1 required
a Mixed noble gases for beta scintillator and Xe 133 for Geiger-Mueller counter.
oreN(Ot,lTL£'-)"WJlQ*."(.lJ)1o(,"MllJlO","Wlll)(,l)I,jTl'l:..TDUANEARNOLDCENTERIOWAELECTRICLIGHT&POWERUPDATEDFINALREPORTAirEjectorOffgasRadiationMonitoringSystem-FCDAPED-N62-7(3)-2Rev1Figure11.5-2
(
UFSAR/DAEC-1 APPENDIX 11A: GASEOUS RELEASE RATE LIMIT CALCULATIONS TABLE OF CONTENTS Section Title 11A-i Revision 13 - 5/97 11A.1 ANALYTICAL MODEL ...............................................................................11A-1
11A.1.1 Meteorological Factors ...............................................................................11A-1 11A.1.2 Radiological Factors ...................................................................................11A-4 11A.1.3 Engineering Factors ....................................................................................11A-6 11A.1.4 Averaging Techniques ................................................................................11A-7 11A.1.5 Average Air Concentration .........................................................................11A-8 11A.1.6 Shielding and Occupancy Factors ...............................................................11A-8
11A.2 VERIFICATION OF ANALYTICAL MODEL ............................................11A-10
11A.2.1 Brookhaven National Laboratory ...............................................................11A.10 11A.2.1.1 Meteorology Data ....................................................................................11A.10 11A.2.1.2 Radiological Data .....................................................................................11A.10 11A.2.1.3 Gamma Dose Calculations .......................................................................11A-11 11A.2.1.3.1 Plume Rise ............................................................................................11A-11 11A.2.1.3.2 Isotopic Data .........................................................................................11A-12 11A.2.1.3.3 Dose Rate Calculations .........................................................................11A-12 11A.2.1.3.4 Conclusions About Gamma Dose Calculations ....................................11A-13 11A.2.1.3.5 Ground-Level Air Concentration Calculation ......................................11A-14 11A.2.2 Nuclear Power Station Unit 1 .......................................................11A-14
11A.3 GASEOUS RELEASE RATE CALCULATIONS FOR DAEC ...................11A-16
11A.3.1 Noble Radiogas Release Composition ........................................................11A-16 11A.3.1.1 Steam Jet Air Ejector ...............................................................................11A-16 11A.3.1.2 Gland Seal Exhaust ..................................................................................11A-16 11A.3.1.3 Building Ventilation Exhaust ...................................................................11A-16 11A.3.2 Whole-Body Dose Calculations ..................................................................11A-16 11A.3.2.1 Gamma Dose ............................................................................................11A-16 11A.3.2.1.1 Meteorology ..........................................................................................11A-16 11A.3.2.1.2 Shielding and Occupancy Considerations ............................................11A-17 11A.3.2.1.3 Steam Jet Air Ejector Doses .................................................................11A-17 11A.3.2.1.4 Gland Seal Exhaust Doses ....................................................................11A-17 11A.3.2.1.5 Building Ventilation Exhaust Doses .....................................................11A-18 11A.3.2.1.6 Dose Variation with Release Rate and Distance From Plant Stack ......11A-18 11A.3.2.1.7 Design Objective Release Rate .............................................................11A-18 11A.3.2.1.7.1 Main Stack .........................................................................................11A-18 11A.3.2.1.7.2 Building Ventilation ...........................................................................11A-18 UFSAR/DAEC-1 APPENDIX 11A: GASEOUS RELEASE RATE LIMIT CALCULATIONS TABLE OF CONTENTS Section Title 11A-ii Revision 13 - 5/97
11A.3.2.1.7.3 Correlation of Ventilation and Stack Effluent Releases ....................11A-19 11A.3.2.1.8 Population Dose Calculation .................................................................11A-19 11A.3.2.2 Beta Dose .................................................................................................11A-20 11A.3.3 Internal Dose Calculations ..........................................................................11A-21 11A.3.3.1 Introduction ..............................................................................................11A-21 11A.3.3.2 Design Objectives for Radioiodine Releases ...........................................11A-21 11A.3.3.3 Inventory of Dairy Farms Surrounding DAEC Site.................................11A-22 11A.3.3.4 Calculation of Design Objective I-131 Release Rate from Main Stack ..11A.22 11A.3.3.5 Calculation of Design Objective I-131 Release Rate from Building Ventilation Exhaust ..................................................................................11A-23 11A.3.3.6 Design Objective Release Rate for Combined Ventilation and Stack Effluent Releases .....................................................................................11A-23 11A.4 SIGNIFICANCE OF RADIATION EXPOSURE .........................................11A-25 11A.4.1 Natural Radiation Background ...................................................................11A-25 11A.4.1.1 Cosmic Radiation .....................................................................................11A-25 11A.4.1.2 Radiation from Ground ............................................................................11A-25 11A.4.1.3 Radiation from Air ...................................................................................11A-26 11A.4.1.4 Radiation from Structures ........................................................................11A-26 11A.4.1.5 Radiation from Food and Water ...............................................................11A-26 11A.4.1.6 Total Radiation from Nature ....................................................................11A-27 11A.4.2 Manmade Radiation ....................................................................................11A-27 11A.4.3 Radiation in Perspective .............................................................................11A-27 11A.5
SUMMARY
...................................................................................................11A-29
REFERENCES FOR APPENDIX 11A .....................................................................11A-30 UFSAR/DAEC-1 APPENDIX 11A:
GASEOUS RELEASE RATE LIMIT CALCULATIONS LIST OF TABLES Table Title Page 11A-iii 11A.1-1 Deposition Velocity Coefficients ...........................................................T11A-1 11A.1-2 Diffusion Coefficients ............................................................................T11A-2 11A.1-3 Noble Gas Isotopes Constituting Mixture ..............................................T11A-3 11A.1-4 Particulate Daughter Products ................................................................T11A-6 11A.1-5 Dose Reduction Factors for Shielding....................................................T11A-7 11A.2-1 Frequency of Occurrence, in Percent, of Various Wind Speeds And Directions Observed at the Brookhaven National Laboratory Site Under Very Stable Atmospheric Conditions ..........................................T11A-8 11A.2-2 Frequency of Occurrence, in Percent, of Various Wind Speeds And Directions Observed at the Brookhaven National Laboratory Site Under Moderately Stable Atmospheric Conditions ...............................T11A-9 11A.2-3 Frequency of Occurrence, in Percent, of Various Wind Speeds And Directions Observed at the Brookhaven National Laboratory Site Under Neutral Atmospheric Conditions .................................................T11A-10 11A.2-4 Frequency of Occurrence, in Percent, of Various Wind Speeds And Directions Observed at the Brookhaven National Laboratory Site Under Unstable Atmospheric Conditions ...............................................T11A-11 11A.2-5 Frequency of Occurrence, in Percent, of All Wind Speeds And Directions Observed at the Brookhaven National Laboratory Site Under All Atmospheric Stability Conditions .........................................T11A-12 11A.2-6 Frequency of Occurrence, in Percent, of Various Wind Speeds And Directions Observed at the Brookhaven National Laboratory Site Under All Atmospheric Stability Conditions .........................................T11A-13 11A.2-7 1963 Monthly Average Gamma Dose for Monitoring Stations Around Brookhaven Graphite Research Reactor Site .........................................T11A-14 UFSAR/DAEC-1
LIST OF TABLES (Continued)
Table Title Page 11A-iv 11A.2-8 1963 Average Annual Measured and Calculated Gamma Dose Rates for Monitoring Stations Around Brookhaven Graphite Research Reactor Site ............................................................................................T11A-15 11A.3-1 Noble Radiogas Release Rate to Environs from 12-Bed Recombiner/ Charcoal System .....................................................................................T11A-16 11A.3-2 Noble Radiogas Release Rate to Environs from Gland Seal System .....T11A-17 11A.3-3 Annual Average Gamma Detector Dose at Ground Level from Recombiner/Charcoal System ................................................................T11A-18 11A.3-4 Annual Average Gamma Detector Dose at Ground Level from Gland Seal System ............................................................................................T11A-19 11A.3-5 Annual Average Gamma Detector Dose at Ground Level from Building Ventilation System .................................................................................T11A-20 11A.3-6 Annual Average Gamma Detector Dose Versus Distance for Maximum Direction from Combined Contributions of Gland Seal and Air Ejector Offgas Systems .......................................................................................T11A-21 11A.3-7 Annual Average Gamma Dose from Combined Contributions of Gland Seal and Air Ejector Offgas Systems at Point on Site Boundary Furthest from Stack for Different Levels of Fuel-Defect Operation ....................T11A-22 11A.3-8 Average Whole-Body Doses for Population Distribution from Combined Contributions of Steam Jet Air ejector and Gland Seal Systems ...........T11A-23 11A.3-9 Information on Dairy Farms Surrounding the DAEC Site in 1971, by Direction Sector .................................................................................T11A-24 11A.3-10 Monthly Production of Individual Farms Versus Monthly Production Distributor, June 1971 ............................................................................T11A-26 11A.3-11 Grazing-Season Integrated Ground-Level Air Concentrations from Main Stack Release ..........................................................................................T11A-27 UFSAR/DAEC-1
LIST OF TABLES (Continued)
Table Title Page 11A-v Revision 13 - 5/97 11A.3-12 Grazing-Season Integrated Ground-Level Air Concentrations from Building Ventilation Release ..................................................................T11A-28
UFSAR/DAEC-1 11A-6 Revision 13 - 5/97 11A.1-1 Vertical Cloud Width Versus Distance - Very Stable 11A.1-2 Vertical Cloud Width Versus Distance - Moderately Stable 11A.1-3 Vertical Cloud Width Versus Distance - Neutral 11A.1-4 Vertical Cloud Width Versus Distance - Unstable 11A.1-5 Gamma Radiation Absorption Coeffici ents and Buildup Constants for Air, STP 11A.2-1 Gamma Dose Rates for Various Wind Speeds and Stabilities for BGRR Stack (Release Rate 0.127 Ci/sec)
11A.2-2 Gamma Dose Rate in Air for Various Stability Conditions 11A.2-3 Dose Rate in Each Sector 11A.2-4 Nomenclature of Sector Used for Averaging 11A.2-5 Whole Body Gamma Dose (mR/yr) Pattern around Stack 11A.2-6 Whole Body Gamma Dose (mR/yr) Pattern around Unit 1 Stack 11A.3-1 Site Boundary Dose
UFSAR/DAEC-1 11A-7 Revision 13 - 5/97 APPENDIX 11A GASEOUS RELEASE RATE LIMIT CALCULATIONS 11A.1 ANALYTICAL MODEL
The analytical model is primarily concerned with calculating the annual gamma dose at ground level resulting from a continuous release of radioactive materials. As a direct consequence, a method is also obtained for calculating the annual average
concentration at ground level.
In essence, the gamma dose model considers the integrated dose rate from a
continuously distributed gaseous source (the plume). The source distribution is treated by a standard dispersion model that relates the dispersion of airborne particles to downwind distances and to the meteorological conditions that exist during the release intervals. The annual gamma dose is obtained by weighting the gamma dose rate associated with a given meteorological condition by the frequency of occurrence of that
condition.
11A.1.1 METEOROLOGICAL FACTORS
The air concentration per unit amount released at a point (x, y, z) in the cloud at
any instant is given by Equation 11A.1-1, which is Sutton's equation corrected by Cramer 1 for depletion by ground deposition and radioactive decay.
()()X Q o yz u h y y z z Q Q o t=+2 2 2 2 2 2 2expexp (11A.1-1) where
()X = average air concentration, Ci/m 3 or µCi/cm 3 Q o = release rate, Ci/sec u h = average wind speed at height of release, m/sec zy = standard deviation of cloud width in vertical and horizontal direction, respectively, m t = time after release, sec = radioactive decay constant, sec
-1 UFSAR/DAEC-1 11A-8 Revision 13 - 5/97 The factor Q Q o is the correction for cloud depletion due to deposition and is equal to the fraction of the initial amount released, which is present at a downwind distance x. According to Watson and Gamertsfelder 2 , Q Q o is given by Q Q o V d u o u h u o o t u z h z z dt=
expexp2 2 2 2 (11A.1-3)
Values of the deposition velocity, V d, are obtained from Table 11A.1-1.
It is a reasonable approximation to assume that throughout the year all the plumes
that travel anywhere within a given sector direction do not have a skewed frequency distribution within the sector. Then, the average cloud concentration in the sector is
found by integrating Equation 11A.1-1 in th e crosswind direction and dividing by the sector width.
()X ave Xdy X= (11A.1-3) where X = sector width
Equation 11A.1-3 cannot be integrated because the interrelationships among the
variables y , z and u h with respect to their average values is not generally known. However, for any specific combination of wind speed and stability at a given downwind distance, all these variables are known and can be treated as constants, and the integration can then be performed. Thus, the average concentration in the sector for all
occurrences of any specific condition is given by
()[]X ave ij Q o Q Q o x z u h z z=2 2 2 2expexp (11A.1-4) where
UFSAR/DAEC-1 11A-9 Revision 13 - 5/97 = sector angle (/8 or 22.5 degrees is used in this appendix) x = downwind distance and is equal to u h t z = a function of stability, wind speed, and downwind distance, x Thus, the average cloud is seen to have a uniform concentration distribution vertically, which is of the Gaussian form.
The standard deviation in the vertical direction is described by Watson and Gamertsfelder 2 as ()zaktbt 2 1 22=+exp stable condition (11A.1-5)
()z C z x n 2 2 2 2= neutral, unstable conditions (11A.1-6)
The expression for z , in Equation 11A.1-6 is easily recognized as the standard Sutton equation. The expression for z in Equation 11A.1-5 was derived from Hanford field measurements of the vertical concentration taken at several downwind locations under stable conditions. The constants for Equation 11A.1-5 and 11A.1-6 were evaluated from the Hanford measurements for a source height of 200 ft and correlated with vertical temperature gradients at the point of emission.
Since the concentration measurements were averaged over 30- to 60-min
intervals, the constants used to evaluate z are considered to be more appropriate for long-term releases rather than the shorter term or "puff" releases. Figures 11A.1-1 through 11A.1-4 show vertical cloud width, z as a function of distance for each stability category.
The following stability classification is used along with vertical temperature lapse
rates for each:
Very stable T > 1.5°C/100 m Moderately stable
-0.5 < T < 1.5 Neutral -1.5 < T <-0.5 Unstable -1.5 < T Table 11A.1-1 shows the deposition velocity coefficients for each stability
category. Table 11A.1-2 shows the appropriate values of a, b, k 2 , C z , and n used with each stability condition and wind speed. Such values are used to calculate the vertical dimensions of the plume, z, and as stated earlier, were constants derived from the Hanford field measurements.
UFSAR/DAEC-1 11A-10 Revision 13 - 5/97 The conventional reflection factor of 2 usually applied for releases is not included. For the passing cloud that is primarily a gamma dose, the entire plume volume is integrated as an infinite number of point sources to plus and minus infinity in the z direction. This ignores the interception by the ground so that the entire cloud volume is
included.
Inhalation doses are a function of concentration at the ground and subject to reflection effects if they exist. Since the materials of interest in inhalation effects deposit
on the ground, it is doubtful that "perfect" reflection will occur, but rather that the cloud will expand distorting the Gaussian mass distribution of the cloud resulting in, at most, a small increase in concentration. In addition, no account was taken of the better diffusion at the ground (effective on the portion of the cloud near the ground) compared to the stack exit elevation used. Meteorology and Atomic Energy (AECU 3066) shows that compared to an elevation of 200 m, ground-level diffusion coefficients are larger by about a factor of 2 plus proportionally increasi ng dispersion. In any event, an increase by a factor of slightly more than 1.0 but less than 2 would account for this reflection effect.
11A.1.2 RADIOLOGICAL FACTORS
The ground-level gamma dose rate from an elevated plume of radioactive materials having a spatial distribution as given in Equation 11A.1-3 may be considered as the sum of the dose rates from all, the points in the plume. The source strength of each
point is (X)dV and the total source is
()SXdV= (11A.1-7) where
dV = dxdydz
The flux from a point source, considering buildup in the air, is given by
Glasstone 3 as SBe R Rµ4 2 photons per m 2/sec (11A.1-8)
UFSAR/DAEC-1 11A-11 Revision 13 - 5/97 where S = source strength B = buildup factor =
1+KRµ (Figure 11A.1-5)
K =µµµa a µ = total (linear) attenuation coefficient, m
-1 µa = energy absorption coefficient, m
-1 R = distance from source equal to
()xyz222 12++/ xyz111 ,, = coordinates of dose point at ground level relative to the incremental volume, dV
The gamma dose rate from a flux of a given energy (E) from Glasstone is
()DRE aµ=x510 3 (units of mR/hr) (11A.1-9) so that the total dose rate from the plume at any point is found by combining Equations 11A.1-7, 11A.1-8 and 11A.1-9. Hence, the gamma dose rate
()()DRE a X ave Be R R dVµµ=x510 3 4 2 (mR/hr) (11A.1-10)
As Equation 11A.1-10 is written, it assumes a monoenergetic source. For a mixture of isotopes, it is proper to perform the calculation for each gamma energy present
considering its abundance. Since
µ and µa are energy dependent and appear in an exponential term, care must be exercised if an average energy is to be used. A listing of each of the noble gas isotopes and significant particulate daughter products is shown in Tables 11A.1-3 and 11A.1-4. Also shown are the gamma energies, total attenuation, and
linear absorption coefficients.
In general, Equation 11A.1-10 cannot be solved analytically and must be solved numerically. While integration to infinity is indicated, in practice finite bounds are
placed on the cloud. Integrating Equation 11A.1-10 to
+/-3z includes more than 99.97% of the entire matter per unit length; hence, the dose contributions from points in the cloud when vertical displacement is more than three standard deviations from the plume center line can be ignored. Likewise, because of the geometric and material attenuation shown in Equation 11A.1-8, one can usually ignore the dose contribution from source points that are more than 400 to 500 m downwind or upwind of the receptor point without
significant error. The integration proceeds by reducing the distributed source (the plume) into a large array of point sources. This is done by dividing the cloud into cubical UFSAR/DAEC-1 11A-12 Revision 13 - 5/97 volume elements. The assumption is made that the concentration at the center of the cube is average for the volume sheets.
The total source strength is preserved by multiplying the concentration at the center (µCi/cm 3) by the volume of the element (cm 3). The dose rate from each point source is calculated by Equation 11A.1-10 and summed over all points.
Equation 11A.1-10 then becomes a finite series.
Mathematically the numerical integration can be expressed as
()()DR ij PG ijIPP I P=,; (11A.1-11) where ()G ijIPP ,; is the dose rate contribution from isotope I to point P from a source at P as described by Equation 11A.1-10.
Equations11A.1-10 and 11A.1-11 give the average dose rate for the (ij) th meteorological condition for a point P, which may be immersed in the cloud or at some point outside the cloud. This is a significant item since the gamma dose at ground level from a stack plume is not merely existent when the receptor is immersed in the plume.
Dose is also received when the plume is traveling in some other sector than the one in which the receptor point is located. The effect is particularly important at points close to the stack where the receptor remains at a nearly constant distance from the plume
regardless of angular separation.
11A.1.3 ENGINEERING FACTORS
From Equations 11A.1-4 and 11A.1-10, it is evident that the dose rate is
significantly affected by the height of the plume above ground level. This height is made up of the physical stack height plus plume rise due to exit velocity and buoyancy. Many formulas are available to calculate plume rise. The method used here is the Holland formula as modified by Moses and others.
4 ()H KV s dQ h u h=+x15410 5. (11A.1-12) where V s = exit velocity, m/sec d = stack diameter, m Q h = heat emission of effluent, cal/sec u h = wind speed at stack exit, m/sec UFSAR/DAEC-1 11A-13 Revision 13 - 5/97 K = correction factor for stack diameter (4) (StÜmke regression coefficient)
In proposing the correction factor K in the plume rise formula, Moses used data from an experimental stack at Argonne with a diameter of about 0.46 m and from a stack at Duisburg, Germany, which has a diameter of 3.5 m. His conclusions are that a value
of 3 for the correction factor is proper for large stacks with appreciable buoyancy whereas a factor of 2 is recommended for small stacks with modest buoyancy. In applying the Moses correction to individual situations, a linear interpretation is made from the actual stack diameter compared to those from which data were obtained.
The AEC documents 5 points out similar results in Section 5.2 discussed by Gary A. Briggs. He states that "both the St
Ümke formula and Holland formula times a factor of 3 seem to give good agreement (calculated versus observed plume rise) for the moderate-sized sources (heat emissions of about 10 6 cal/sec), but grossly underestimate rise in the case of the large Colbert plant (heat emissions of about 7 x 10 6 cal/sec)."
11A.1.4 AVERAGING TECHNIQUES
One is usually interested in the cumulative dose over some appropriate time interval, such as a year. To compute the annual gamma dose, the gamma dose rate for a given meteorological condition must be we ighted by the frequency distribution F ijk. F ijk describes the frequency of the i th stability condition with j th wind speed occurring in direction sector k. The average annual gamma dose rate in section k is given
()()DR ijPCDRijk P k k F ijk=; (11A.1-13) where ()DRijkk P; = the gamma dose rate to a point P in sector k from a plume traveling in sector k C = 8760 hr/yr
Equation 11A.1-13 indicates a finite summation over the variables of stability, wind speed, and direction. For stability and direction it has already been indicated how these variables can be grouped into 4 stability classes and 16 directions. The spectrum of
wind speeds can also be grouped into representative ranges. One such grouping that has proven useful, especially when using U.S. Weather Bureau summaries, is as follows:
UFSAR/DAEC-1 11A-14 Revision 13 - 5/97 Wind Speed Range (mph) Average Wind Speed (m/sec) 0-3 1 4-7 2 8-12 5 13-18 7 19-24 10 >24 >13 Also included above is the average wind speed that is representative of each
speed range.
11A.1.5 AVERAGE AIR CONCENTRATION
For doses other than the whole-body gamma dose, the annual average concentration at ground level is of interest. This is easily obtained from the preceding material presented by substituting plume height for z. The air concentration during any meteorological condition has been described by Equation 11A.1-4. However, for materials other than noble gases, the depletion factor Q Q o is not 1 and must be accounted for. For the calculations made in this appendix, the deposition rates shown in Table 11A.1-1 were used.
Using the joint frequency distribution F ijk defined previously, computations of the annual average concentration at the ground can be made from:
()()X gr k X gr ij F ijk ij= (11A.1-14) 11A.1.6 SHIELDING AND OCCUPANCY FACTORS
Radiation doses are usually calculated for certain distances from the point of
release and often are calculated for locations where no actual dose would be received by a human receptor. In fact, it is not too uncommon to see radiation doses from the passing
cloud calculated as if the dose receptors were outdoors day and night. This is certainly possible, but it does not lead to particularly accurate dose estimates for most people. For
this reason, occupancy by individuals should be considered in arriving at reasonable dose estimates. Credit for this is allowed by 10 CFR 20.
In addition, it seems rather incongruous to assume that a person would stay in one place all of the time without being inside some type of shelter. For this reason, the
shielding effect for various types of structures was evaluated. The shielding value 6 of such typical structures is shown in Table 11A.1-5.
UFSAR/DAEC-1 11A-15 Revision 13 - 5/97 It is easily seen that the error introduced by omitting this effect can be a factor of 2 or more. Where larger urban complexes are concerned, such an error may be far
greater.
UFSAR/DAEC-1 11A-16 Revision 13 - 5/97 11A.2 VERIFICATION OF ANALYTICAL MODEL The methods described previously that are used for calculating offgas limits use standard, available numerical techniques.
Certain portions of the calculations, however, involve assumptions and/or extrapolation of existing data. Consequently, while the calculations are conservative, experimental verification of the results has been undertaken. Two pieces of information have been used in this verification effort. One quantity of data comes from a monitoring program at the Brookhaven National Laboratory; the other was accumulated at the site monitoring the effluent from Unit 1. In the case of Brookhaven, the research reactor emitted argon-41 (A-41) in small but easily measurable quantities and in a rather uniform pattern. Meteorological and offgas data taken from this location were used to predict environmental doses 7 and compared to measured data.
11A.2.1 BROOKHAVEN NATIONAL LABORATORY
11A.2.1.1 Meteorology Data
Micrometeorological data for 1963 were obtained from Brookhaven National Laboratory. The data were in the form of computer input cards containing hourly
observations of average wind speed and directi on at levels of 37, 150, and 355 ft and the air temperatures at levels of 37, 75, 150, 300, and 410 ft. The measurements at the 355-ft level were summarized in terms of frequency of occurrence according to wind speed and direction and atmospheric stability. The stability was determined according to the method described in Section 11A.1 by using the temperature gradient measured between
the 410- and 37-ft levels.
The summaries are presented in Tables 11A.2-1 through 11A.2-6. The frequency of occurrence was based on 6464 hr of good observations. Of the missing 2296 hr of 1963, August and September account for 1464 missing hr, the rest being scattered
throughout the year. A total of 12 hr was observed to have a wind speed less than 0.5 mph. These "calm" conditions were included in the wind speed category (0 to 3 mph).
11A.2.1.2 Radiological Data
As is discussed by Hull 8, the radiation dose was measured at several stations around the Brookhaven Graphite Research Reactor (BGRR) in 1963 using 6-liter, atmospheric ion chambers. The dose rate from the release of A-41 was determined from the total dose measurement by subtracting from it the contribution from natural
background and operation of the forest ecology station. The resultant dose rate is shown
in Table 11A.2-7.
It was necessary to adjust the measured values of annual gamma dose to account for the absence of meteorological data during August and September. The average dose rate (millirads per week) was averaged over the 10 months for which meteorological data were available and multiplied by 52 to get annual dose (millirads per year). The UFSAR/DAEC-1 11A-17 Revision 13 - 5/97 exception to this is station E-2, which was moved in December. For this station, a 9-month period was used to determine the annual dose. These normalized values are shown in comparison with calculated values in Table 11A.2-8.
11A.2.1.3 Gamma Dose Calculations
The methods described in Section 11A.1 were used to analyze the effects of the
BGRR stack effluent in the Brookhaven envir ons. The following is a discussin of the calculaions leading to the gamma dose rate matrix, DRijkk,. 11A.2.1.3.1 Plume Rise
The BGRR has a 350-ft stack (107 m) with an exit velocity of 6 m/sec and an effluent temperature difference of 50
°C above ambient. For use in Equation 11A.1-12, these values correspond to
Q h = 1.62 x 10 6 cal/sec, heat rate d = 5.18, stack exit diameter K = 3.47, correction factor in equation
Using these values in Equation 11A.1-12 the plume rise formula becomes
H u h m=387 Using the six standard wind speed groups described earlier, the effective stack heights were computed and are shown below:
BGRR Plume Rises for Various Wind Speeds Wind Speed Range (mph) Average Speed (m/sec) Plume Rise (m) Effective Height (m) 0-3 1 387 494 4-7 2 194 301 8-12 5 77 184 13-18 7 55 162 19-24 10 39 146 >24 >13 30 137
UFSAR/DAEC-1 11A-18 Revision 13 - 5/97 11A.2.1.3.2 Isotopic Data
The BGRR during full-power operation rel eased about 12,960 Ci of A-41 per day (0.15 Ci/sec). However, the actual average release rate during 1963 was 0.127 Ci/sec as determined from communications with the BGRR staff, which represents an 85%
operation factor.
Pertinent radiological properties of A-41 are as follows:
E = 1.29 MeV, gamma energy
µ = 6.93 x 10
-3 M-1, total attenuation coefficient
µa = 3.3 x 10
-3 M-l , energy absorption coefficient = 1.1 x 10
-4 sec-1 , decay constant
11A.2.1.3.3 Dose Rate Calculations
From the above information, the gamma dose rate as given by Equations 11A.1-10 and 11A.1-11 was evaluated using a digital computer program to evaluate Equation
11A.1-10. The dose rate was evaluated for downwind distances of 10, 100, 400, 1400, 2400, 3200, and 6400 meters, using the six wind speeds shown earlier and all four
stability conditions. The results are shown in Figures 11A.2-1-through 11A.2-3. The dose rates for the very stable and moderately stable conditions are essentially identical, because for the distances used here the vertical spread of the plume is small in each case.
Hence, the difference in cloud dimensions between the two stable conditions is not great compared to the attenuation distances involved.
Another important feature is that there is very little variation in dose rate between any of the stability classes for the plume height considered here. Figure 11A.2-2 illustrates this point more clearly by showing the dose rate for a 5-m/sec wind speed for
each of the stability conditions. The variation of dose rate between stability conditions is very small for downwind distances less than 400 m, and is less than a factor of 2 even to a distance of 6 miles. From the shape of the dose rate curves, the maximum usually occurs within 1000 m and decreases rapidly thereafter.
The dose rates shown in Figures 11A.2-1 through 11A.2-3 are for points on the ground directly below the centerline of the sector-averaged plume. As previously mentioned, significant dose contributions can also occur in sections other than the one in which the plume is traveling. Because of symmetry, there are only nine unique sectors for which dose rate calculations can be made.
If the sector in which the plume is traveling is designated as sector 1 (Figure 11A.2-4), then the dose to sector 16 from the plume is equal to the dose to sector 2; the dose to sector 15 is the same as the dose to sector 3, and so on. In terms of the dose rate matrix, the following equalities can be listed:
UFSAR/DAEC-1 11A-19 Revision 13 - 5/97 DR ij DR ij1111 ,,= DR ij DR ij16121 ,,= DR ij DR ij15131 ,,= . . . .
. .
DR ij DR ij9191 ,,= However, for distances greater than 100 m, the dose rate to adjacent sectors is very small because of the large separation distances. This is illustrated by Figure 11A.2-
3, which shows the sector variation of dose rate with distance for one particular meteorological condition. In practice, the dose rate to a point in sector k is not calculated if the dose rate is less than 0.1% of the dose rate to a point at the same downwind
distance in sector 1.
Figures 11A.2-1 through 11A.2-3 indicate how the dose rate matrix DRijkk, is constructed. One must then find the joint frequency distribution F ijk to calculate the annual dose rate.
11A.2.1.3.4 Conclusions About Gamma Dose Calculations
From the data presented in Figure 11A.2-5, it is concluded that the analytical model provides a fairly precise correlation between stack release rate and ground-level gamma radiation dose. The maximum dose is at the closest point to the stack. This is expected because at the base of the stack, for example, the dose rate is continuous and independent of plume direction travel. The dose rate curves presented in Figures 11A.2-1 through 11A.2-3 indicate this.
Further examination of Figures 11A.2-1 through 11A.2-3, which show dose rates during each meteorological condition, leads to additional conclusions. The dose rate does not seem to be very sensitive to the atmospheric stability condition. This is markedly in contrast to the air concentration differences at ground level during the various stability regimes. It is widely known that, during very stable conditions near zero, air concentration exists at ground level from an elevated plume since it remains
very narrow and highly concentrated aloft. On the other hand, unstable conditions promote rapid effluent growth and dispersion and highest ground-level air concentrations.
UFSAR/DAEC-1 11A-20 Revision 13 - 5/97 It appears that, while the gamma dose rate is quite insensitive to atmospheric stability, it is quite dependent on plume height and wind speed. This is to be expected intuitively from Equation 11A.1-4 in which the average concentration that is used to
obtain the dose rate is inversely proportional to wind speed and the attenuation distances increase with plume height. In practice, buoyant effluents are typical (although not universal), so that effluent buoyancy enters the calculations (i.e, plume height is made up of stack height plus plume rise due to buoyancy). The latter is greatest for slowest wind
speeds. Thus, the slowest wind speed conditions do not, a priori, yield the largest dose rates. Experience with calculations using this analytical model verifies this.
Calculations have also shown that most of the dose over a long period of time comes from the conditions where the wind speed is about at the average speed of 4 to 7 m/sec (9 to 16 mph), which most locations are observed to have. The calculation for
Brookhaven is no exception. This can partially be explained by the fact that for elevations considered here (300 to 400 ft), low wind speeds, for example, are rather infrequent accounting for about 3% of the time.
A final conclusion drawn from the comparison of calculated and measured doses
refers to the dose pattern depicted in Figure 11A.2-5. It is observed that for distances out to about 0.5 mile (typical large reactor site) the iso-dose contours exhibit a smooth rather than a peaked pattern. This is quite different from the wind direction distribution (wind rose, see Table 11A.2-5) where total direction frequency is indicated. However, the smooth gamma dose pattern, as indicated in Figure 11A.2-5, is attributed to the fact that the total dose at each point is made up of the dose from plumes traveling in all directions.
At distances of 2 miles and beyond, the gamma dose contours exhibit a peaked pattern similar to the wind rose. At these distances, only plumes traveling in the direction of a dose point contribute significantly to the gamma dose at the point.
11A.2.1.3.5 Ground-Level Air Concentration Calculation
For some kinds of radiation doses, only the ground-level air concentration is of interest. Examples of these are dose from inhalation, external beta dose, and deposition.
In each of these, concentration at the dose point determines the dose regardless of the concentration at other points in the plume. This method of calculating the correlation between stack emission rate and ground-level air concentration is also of interest in assessing environmental effects of a stack effluent.
Some limited air concentration measurements are also made at Brookhaven.
8 These are measurements of small quantities of iodine released from the BGRR. Three monitoring stations were operated in 1963, although since then the scope of this program has been augmented. The release of I-131 from the BGRR was about 0.1
µCi/sec continuously.
11A.2.2 NUCLEAR POWER STATION UNIT 1
UFSAR/DAEC-1 11A-21 Revision 13 - 5/97 In the case of the , data were used for a period when offgas release was measurable in the environment. The values used in these determinations were obtained from paired 10-mR gamma dosimeters in each of the environmental stations at . All chambers were read weekly with results averaged on a monthly basis. The calculated iso-dose contours relating doses in the environment to Unit 1 stack emission were compared to the field data. A 7-month period when the release rate averaged about 51,500
µCi/sec (September 1964 to March 1965) was selected as the test period. Since the environmental background was known to be decreasing during the period of interest, comparable 7-month periods the year before and the year following were averaged to minimize seasonal meteorological variables. The background obtained was subtracted from the gross reading during the test period. Results were compared to predicted values obtained from iso-dose curves at a release rate of 700,000
µCi/sec (license limit)
- related to 51,500
µCi/sec. Statistical analyses have shown that the ion chambers lowest level of detection at the 95% confidence level was approximately 0.4 to 0.5 mR/week above background. Results of these comparisons are shown in Figure 11A.2-6. In this case, the measurements are quite near the sensitivity of the instruments so that more uncertainty exists in the data compared with the Brookhaven
data.
However, agreement between predicted and observed is still quite good. In the direction of predicted maximum dose, predicted is greater than observed. However, in
one direction, predicted is less than observed.
On the basis of the above experimental data supporting the analytical methods, it is concluded that the proposed offgas limits are reasonably realistic, conservative values.
- Unit 1 license limit was recently changed by the NRC to 500,000
µCi/sec.
UFSAR/DAEC-1 11A-22 Revision 13 - 5/97 11A.3 GASEOUS RELEASE RATE CALCULATIONS FOR DAEC
The analytical model described in Sections 11A.1 and 11A.2 was used in the following calculations for the DAEC plant site.
11A.3.1 NOBLE RADIOGAS RELEASE COMPOSITION
11A.3.1.1 Steam Jet Air Ejector
The offgas treatment system used in the DAEC employs a catalytic recombiner
and a 12-bed charcoal adsorption unit as described in Section 11.3.2.1. Table 11A.3-1 lists the radiogases and their respective release rates as they leave the recombiner/
charcoal system to the environment for an assumed fuel defect operational condition corresponding to 100,000
µCi/sec (referenced to 30-min decay) input activity and a condenser inleakage of 18.5 scfm.
11A.3.1.2 Gland Seal Exhaust
An additional source of radioactive noble gases released to the environment is the gland seal steam condenser. Approximately 0.1% of the total steam flow is routed through the gland seal system to the gland seal condenser. The result is a release of approximately 0.1% of the total noble gases through the gland seal system. These gases are routed to the stack and then released to the atmosphere. Based on a total noble gas generation of 100,000
µCi/sec at 30-min decay, release from the gland seal system will be 100 µCi/sec at 30-min decay. However, these gases are approximately 1.75-min old when released from the stack. As a result, 464
µCi/sec is released to the environment. The isotopic composition of this release is shown in Table 11A.3-2.
11A.3.1.3 Building Ventilation Exhaust
Ventilation exhaust from plant structures is an additional source of noble gas
activity that has previously been considered negligible but takes on greater significance when compared to the offgas discharge from a plant employing a recombiner/charcoal system such as the DAEC. The noble gases contained in the ventilation exhaust result primarily from small steam leaks from reactor coolant pressure boundary system piping external to primary containment.
11A.3.2 WHOLE-BODY DOSE CALCULATIONS
11A.3.2.1 Gamma Dose
11A.3.2.1.1 Meteorology
The procedure for calculating annual gamma dose consists of calculating the dose
rate at various points during each different meteorological condition, weighting the dose UFSAR/DAEC-1 11A-23 Revision 13 - 5/97 rate by the frequency of occurrence, and summing over the year to determine the total dose.
Gamma dose rate calculations were done for a number of downwind dose points.
Terrain height of these dose points was considered in the calculation. These dose rates
were weighted by the frequency of occurrence of wind speed and direction and atmospheric stability in accordance with the DAEC onsite meteorological data and summed to give a total "air dose" for the year. Onsite meteorological data were collected from January 1971 through January 1972 at both 156-ft and 33-ft elevations as described in Section 2.3 and summarized into appropriate distributions to allow the calculation of
both stack and building ventilation releases.
For the building ventilation releases, aerodynamic downwash of the effluent plume was assumed to occur whenever wind speeds were greater than 3 m/sec. A
building dilution factor of 1.4 was applied for downwash conditions for distances out to 1500 m. For wind speeds less than 3 m/sec, a release height of 40 m was assumed with
no downwash or building wake effect.
The characteristics of the DAEC stack offgas exhaust system design are such that vertical momentum and buoyancy effects result in a plume rise that is negligible when compared with the stack height. Accordingly, the release elevation used in the calculations is 100 m for the offgas and gland seal discharge.
11A.3.2.1.2 Shielding and Occupancy Considerations
All doses calculated in this section are doses a detector would receive. However, doses to persons would be lower for two reasons. First, at least part of the time persons would be inside structures, such as their homes, which provide shielding. Second, most people would not stay at a given location all the time. While away from this location, they would not be exposed to plant releases. Thus their annual exposure would be lower.
Conservatively, all doses could be reduced by a factor of 2 to give a dose to people. In many cases, the dose would be even more, anywhere from a factor of 3 to a factor of 8 lower.
11A.3.2.1.3 Steam Jet Air Ejector Doses
Table 11A.3-3 shows offsite gamma doses for the air ejector offgas system corresponding to the isotopic composition and release rate shown in Table 11A.3-1. The maximum calculated dose of 0.97 mrem/yr occurs in the north sector. Figure 11A.3-1
shows the calculated exposures for each direction sector on the site boundary.
11A.3.2.1.4 Gland Seal Exhaust Doses
Gamma dose corresponding to the gland seal release indicated in Table 11A.3-2 was calculated in the same manner as the air ejector offgas doses. For an annual average UFSAR/DAEC-1 11A-24 Revision 13 - 5/97 release of 464
µCi/sec, the dose at the site boundary in the north sector is 0.47 mrem/yr. Table 11A.3-4 lists gamma doses calculated for the gland seal stack release.
11A.3.2.1.5 Building Ventilation Exhaust Doses
Gamma doses were calculated at site boundary locations in each sector as indicated in Table 11A.3-5 using an arbitrarily assumed annual average release from ventilation exhausts of 278
µCi/sec using a 5-min decay mix. Table 11A.3-5 indicates that the maximum dose of 1.07 mrem/yr occurs in the south-southeast sector.
11A.3.2.1.6 Dose Variation with Release Rate and Distance From Plant Stack
Table 11A.3-6 presents the variation of calculated annual average gamma dose as a function of distance from the plant stack from the combined gland seal exhaust and air ejector offgas contribution of 2104
µCi/sec assuming an operational condition corresponding to 100,000
µCi/sec at 30-min decay.
Table 11A.3-7 presents the calculated annual average gamma dose at the maximum site boundary location from the combined air ejector offgas and gland seal
exhaust contribution for several different levels of fuel defect operation.
11A.3.2.1.7 Design Objective Release Rate
11A.3.2.1.7.1 Main Stack
As indicated in Section 11.3.1.1, the design objective of the gaseous radioactive waste system is to limit the annual average exposure to a point on the site boundary to a maximum of 10 mrem/yr. This objective for the DAEC meets the intent of as low as reasonably achieveable as defined by 10 CFR 20.
The annual average release rate corresponding to the design objective site boundary exposure of 10 mrem/yr-is 11,800
µCi/sec-release of air ejector offgas from the recombiner/charcoal system, which contributes 7.0 mrem/yr, and a 3,300
µCi/sec-release of 1.75-min-old gland seal exhaust, which contributes 3.0 mrem/yr.
Thus, the 10 mrem/yr design objective release rate for the main stack is 15,100
µCi/sec on an annual average basis.
11A.3.2.1.7.2 Building Ventilation
As indicated in Section 11A.3.2.1.5, the maximum dose from the ventilation exhaust was calculated to be 1.07 mrem/yr for an arbitrarily assumed annual average release rate of 278
µCi/sec and occurs in the south-southeast sector. If this source alone is considered, the design objective site boundary dose of 10 mrem/yr results from an annual average ventilation exhaust release rate of 2,608
µCi/sec.
UFSAR/DAEC-1 11A-25 Revision 13 - 5/97 11A.3.2.1.7.3 Correlation of Ventilation and Stack Effluent Releases
Since the overall objective is to maintain the total site boundary dose at less than 10 mrem/yr, a relationship must be used to relate stack and ventilation releases such that the sum of the exposure contributions is within the 10 mrem/yr design objective. This relationship is as follows:
Q vent Q stack260815000 10 ,,. where Q vent = annual average ventilation release rate, µCi/sec. Qstack = annual average stack release rate, µCi/sec. This combination technique is conservative since the stack and building vent
release rates are predicated on exposures affecting two different sectors.
11A.3.2.1.8 Population Dose Calculation
In addition to calculating the doses to offsite locations, calculations have also been performed to determine the general population dose. The results are shown in Table
11A.3-8.
The population dose is in units of man-rem per year. The assumed annual average release considered corresponds to 100,000
µCi/sec at 30-min decay delayed to the holdup time appropriate for the 12-bed recombiner/charcoal system used for the DAEC giving a net release rate of 2104
µCi/sec. The population distribution used was the estimated population around the DAEC site in the year 2010.
The exposure calculated is 10.2 man-rem/yr. Natural background radiation would result in 139,000 man-rem/yr. Thus, the population exposure from this source would be
0.007% of natural background per year. Even at the design objective release rate of 15,100 µCi/sec, the plant increment would only be about 0.12% of background.
The background assumption is 150 mrem/yr. Assuming that natural background varied only 1 mrem/yr, this would change the natural population dose around the DAEC by about 925 man-rem/yr. This is more than the plant contribution by nearly a factor of 100 when the release averages 2104
µCi/sec. Actually, natural background would vary more than 1 mrem/yr (more like 10) so that the contribution from the plant would be
virtually undetectable.
11A.3.2.2 Beta Dose UFSAR/DAEC-1 11A-26 Revision 13 - 5/97 The range of beta particles in air is only a few meters. Hence, for beta calculations, a cloud of material released via a stack and which expands to large dimensions at downwind distances where the cloud has reached ground level is
frequently considered an "infinite" cloud. In such a cloud, the air dose rate is calculated assuming that the rate of energy release per unit volume in the cloud is equal to the rate of absorption in that volume (no buildup). The body is considered a small volume within
the flux of the cloud and causes no perturbation in the flux.
Beta flux incident on the human body comes from one direction only, so that the
air dose rate at the surface of the body is only one-h alf of that in the air. In addition, the cloud is not infinite since the ground represents a boundary to the cloud, such that at the ground the cloud is a hemisphere of infinite radius. It approaches the infinite cloud at some height above the ground equal to the range of the betas in air. There will be a variation in the dose rate from the head to the foot of an individual, with the highest dose rate at the head. This factor varies from 0.
5 at the ground to 1 at heights greater than the range of betas in air. Taylor 9 has computed this effect to show that the average dose to the body of a person 1.8 m tall is about 0.64 times the semi-infinite cloud dose. This factor applies for mixed fission products with maximum energies of about 1 to 2 MeV.
The following beta dose equation 10 is used and modified:
D B EX=0457. (11A.3-1)
This equation is multiplied by 0.5 for the beta flux factor discussed above and by
0.64 to account for the average dose to the body. Converting Equation 11A.3-1 into a
dose rate yields the equation used in this analysis:
()()DR B EX=x05310 6. (11A.3-2)
Substituting X avg i for (X) gives the average beta dose rate avg for the i th meteorological condition. Since the range of betas in air is quite short, the annual total beta dose in a given direction is the sum of the dose rates (in millirad per hour) during
each i th condition accompanied by wind blowing in that direction weighted by the annual frequency (in hours) of occurrence. The conversion of this dose into a dose delivered to an individual requires adjustments to take th e shielding effect of clothing into account.
In addition, for the beta dose to be truly additive with the whole-body gamma dose, the betas must penetrate the skin. Several discussions are found in the literature.
Even radiation protection regulations diffe rentiate between whole-body doses and the skin of the whole-body dose. Hendrickson 11 indicates that about a factor of 70 difference exists between whole-body and skin doses. Dunster l2 also indicated that care should be exercised when discussing doses by properl y identifying whether one is considering external dose to the whole body from penetrating radiation, skin dose from beta radiation, UFSAR/DAEC-1 11A-27 Revision 13 - 5/97 and dose to the thyroid from internal radiation from inhaling or ingesting I-131. In discussing the inadequacy of some references to dose from Kr-85, Dunster indicated that the whole-body dose and gonad dose from immersion in Kr-85 (a beta emitter) is only about 1% of the skin dose. Some references seem to add or equate skin dose with the
whole-body or gonad dose. Bond l3 basically supports the position of Hendrickson and Dunster but arrives at about a factor of 1000 difference between skin dose and whole-body dose.
Since the beta dose to the skin typically contributes a small fraction of the whole-body dose, it was not added to the whole-body dose. Whole-body dose is far more limiting than external skin dose.
11A.3.3 INTERNAL DOSE CALCULATIONS
11A.3.3.1 Introduction
During normal plant operation, small quantities of radioiodine may be discharged from the gland seal exhaust system via the main stack and from the building ventilation system. Radioiodine discharged from the main stack comes primarily from the gland seal exhaust. Although radioiodine may be present in gases leaving the steam jet air ejector, it will undergo essentially complete adsorption on the charcoal delay beds before being discharged from the main stack. Small leaks that could occur in process piping and equipment may result in discharges of radioiodine from the building ventilation system.
Inhalation or ingestion of radioiodine will result in ultimate deposition of the material in the thyroid gland because of natural metabolic processes. This, of course, presumes that such exposure pathways actually exist. Ingestion, rather than inhalation, is
the controlling path for thyroid exposure from radioiodine at the DAEC site because of concentration.effects unique to the air-vegetation-cow-milk-infant food chain. This concentration effect is accounted for by reducing the annual average maximum permissible air concentrations of I-131 at grazing locations by a factor of 700.
11A.3.3.2 Design Objectives for Radioiodine Releases
To maintain levels of radioiodine in gaseous effluents released from the DAEC as low as reasonably achievable, DAEC pr oposes the following design objectives:
- 1. Annual average concentrations at any location on the boundary of the site or in the offsite environment of radioactive iodines or radioactive material in particulate form with a half-life-greater than 8 days shall be less than the concentrations in air specified in Appendix B, Table 2, Column I of 10 CFR 20 divided by 100.
- 2. Where there are grazing animals providing fresh milk for human consumption, the 10 CFR 20, Appendix B, Table 2, Column I concentration of I-131 will be divided by 7000. (The factor of 7000 results from a design objective of 10% of the Appendix B, Table 2 value divided by 700 to account for concentration effects in the milk food UFSAR/DAEC-1 11A-28 Revision 13 - 5/97 chain.) However, this number may be reduced by considerations such as the following:
- a. The fraction of the year during which grazing is impossible.
- b. The fraction of dilution provided by pooling at a central dairy.
11A.3.3.3 Inventory of Dairy Farms Surrounding DAEC Site
In order to establish a design objective release rate for I-131 from the DAEC, the
grazing season average air concentration at th e nearest pasture supporting a dairy herd in each meteorological sector was required. Accordingly, an inventory was made to determine the location of the nearest dairy farm in each sector, and the results of this inventory are summarized in Table 11A.3-9. This table includes such information as name and location of dairy farm, distance from plant stack, number of dairy cows, months on pasture, and name and location of distributor. Table 11A.3-9 shows that for most of the farms interviewed, the grazing season was identified as the 6-month period, May through October. However, for conservatism the calculations of this section assumed a 7-month grazing period extending from April 15 to November 15. Table 11A.3-10 shows the monthly production of dairy farms versus the monthly production of their distributors for a typical month (June 1971) felt to be representative for milk
production and distribution. This table also delineates Class B dairies. Class B dairies are those whose milk product is not distributed for direct liquid consumption but rather for cheese and other dairy products that w ould not have the potential of radioiodine ingestion.
11A.3.3.4 Calculation of Design Objective I-131 Release Rate from Main Stack The average ground-level air concentration of I-131 was determined for each farm listed in Table 11A.3-9 for the April 15 through November 15 grazing season using onsite data from the 156-ft level of the DAEC meteorological tower. The I-131 release rate from the main stack was arbitrarily assumed to be 0.01
µCi/sec. Table 11A.3-11 presents the results of these calculations. The highest calculated concentration is 2.69 x 10-16 µCi/cml at the farm located in the west-northwest sector. Since the average annual I-131 concentration that would be equal to MPCa/7000 is 1.43 x 10-14 µCi/cm 3 , the design objective release rate from the stack would be 0.53
µCi/sec. However, as discussed previously the farm is used for grazing only 7 months of the year. The consideration of this factor would establish a design objective main stack release rate for I-131 of 0.91
µCi/sec on an annual average basis.
As Table 11A.3-9 shows, the farm located in the west-northwest sector is listed as a Class B dairy, meaning that the milk product is not commercially distributed for direct liquid consumption.
This would appear to indicate that the farm in the south-southwest sector, which is a Class A dairy, should be used as the basis for release rate calculations, with a UFSAR/DAEC-1 11A-29 Revision 13 - 5/97 resultant increase in the design objective release rate. However, for conservatism it has been assumed that family consumption of liquid milk at the Class B farm is possible. In addition, infant consumption of the liquid milk is conservatively assumed.
Further reference to Table 11A.3-9 indicates that the population exposure from Class B dairy products is essentially zero, and the dilution afforded by the pooling of
Class A liquid products at the central dairy is such that population exposure from this source is totally negligible. The monthly production of Class A dairy farms surrounding the DAEC site is on the order of 0.1% of the total product distributed by the central dairy.
11A.3.3.5 Calculation of Design Objective I-131 Release Rate from Building Ventilation Exhaust The meteorological assumptions used in this calculation varied slightly from those used for noble gas ventilation releases in that aerodynamic downwash occurs for wind speeds greater than 3 m/sec, but a release height of 0 m was conservatively assumed instead of using a building dilution factor. For wind speeds equal to or less than 3 m/sec, a release height of 40 m was used.
An annual average release rate of 0.001
µCi/sec of I-131 was arbitrarily assumed as the source term. The resultant grazing season average concentration at the dairy farms
surrounding the DAEC is in Table 11A.3-12. As before, the highest concentration occurs at the farm in the west-northwest sector. Normalizing the ground-level concentration to MPCa/7000 (1.43 x 10-14 ) results in a release rate of 0.044
µCi/sec. However, since the farm is used for grazing only 7 months of the year, the consideration of this factor would establish a design objective annual average ventilation release rate for I-131 of 0.075
µCi/sec.
11A.3.3.6 Design Objective Release Rate for Combined Ventilation and Stack Effluent Releases The correlation for ventilation and stack effluent releases for I-131 is similar to
that developed in Section 11A.3.2.1.7.3 for noble gases.
The design objective release rate is
Q vent Q stack0075091 10... where Q vent = annual average ventilation release rate, µCi/sec Qstack = annual average stack release rate, µCi/sec UFSAR/DAEC-1 11A-30 Revision 13 - 5/97 11A.4 SIGNIFICANCE OF RADIATION EXPOSURE
11A.4.1 NATURAL RADIATION BACKGROUND
Every day people receive radiation from the sky, the ground, the air, and food.
The magnitude of this radiation level is st rongly influenced by where people live, what they do, and even in what kind of house they live. For most locations around the United States, this natural radiation level averages about 140 mrem/yr. This typical value can be discussed in its various component contributions.
11A.4.1.1 Cosmic Radiation
Cosmic radiation is one of the more significant sources of natural radiation. This radiation is to some extent dependent on latitude and to a large extent dependent on
altitude.
In the mid-latitudes, where most people live, the cosmic radiation varies from about 50 mrem/yr at sea level to about 3800 mrem/yr at altitudes that jet aircraft fly (35,000 ft). This does not mean that all commercial jet-airliner crews receive 3800 mrem/yr, since this would assume that the crews were continuously airborne. Assume, for instance, that these crews stay aloft a tenth of the year; thus their occupational radiation exposure due to cosmic radiation alone would be in the range of 300 to 400 mrem/yr. Even one transcontinental round trip would give the business man or vacationer about 4 mrem/yr.
The average cosmic radiation of 50 mrem/yr will increase to about 150 mrem/yr for some mile-high locations such as Denver and Salt Lake City. Even with this, 50 mrem/yr seems to be a good average.
11A.4.1.2 Radiation from Ground
Another source of radiation in nature is the ground itself because it contains many radioactive minerals, particularly the uranium and thorium series, together with the important uranium decay product, radium. A nother significant radioisotope in the ground is potassium-40, the naturally radioactive isotope of the element potassium. This incidence of radioactive material in the ground causes the earth to act as a large plane
radiation source. This produces an average radiation exposure in the continental United States of about 45 mrem/yr. Assuming that the average person spends about one-fourth of the time walking on the ground outside of buildings, this 45 mrem/yr would reduce to 15 mrem/yr.
There are a number of locations in the world where the radiation exposure from the ground is actually much higher. In various locations in Brazil, India, and in the French mountains, the exposure may range from 180 to as high as 1600 mrem/yr. This is
largely because of the presence of deposits of thorium near the surface of the ground.
There also have been reports of exposures higher than these.
UFSAR/DAEC-1 11A-31 Revision 13 - 5/97 11A.4.1.3 Radiation from Air
The radioisotopes in the ground give rise to a secondary source of radiation, since the natural decay of the uranium and thorium series each contains a natural radiogas.
These radiogases evolve from the ground at a fairly constant rate and thus cause
concentrations of natural radiogases in the air.
The principal constituent of this source of exposure of radiation in nature is the radiogas radon, which has a 3.8-day radioactive half-life. This element, together with its daughter decay products, causes a world average of about 5 mrem/yr full-body, external
radiation exposure. Actually, the inhalation of these radiogases and the deposition of their radioactive daughters in the lung may cause a lung dose of about 200 mrem/yr.
11A.4.1.4 Radiation from Structures
Since man uses materials from the ground for buildings, natural radioisotopes from the ground are transferred to these stru ctures. A significant variation will result from the use of different building materials. A wooden structure may give a radiation dose rate of about 50 mrem/yr, while concrete may give 70, and brick as high as 100.
Even these may vary within a particular material on the basis of where the material
originated.
11A.4.1.5 Radiation from Food and Water
Another source of radiation in nature is food and water, both of which contain naturally radioactive materials. The general average radiation exposure from food and water is about 25 mrem/yr from the deposition and retention of these radioactive materials within the body. In a typical case, about 20 mrem/yr of this exposure comes from the natural radioisotope potassium-40, which is found particularly in protein-type foods.
UFSAR/DAEC-1 11A-32 Revision 19 - 9/07 11A.4.1.6 Total Radiation from Nature The following summarizes the various contributions in arriving at the average natural background radiation of 140 mrem/yr.
Source Radiation (mrem/yr)
Cosmic rays 50 Ground (1/4 time) 15 Buildings (3/4 time) 45 Air 5 Food and water 25 Total 140
11A.4.2 MANMADE RADIATION
Man has added to the radiation exposure from nature in a number of ways. The largest contribution has been from use of medical and dental X-rays. Typically, an average of 55 mrem/yr is received by the average person. Recent reports indicate that 100 mrem is applicable. In addition, radiation from luminous watch dials (2 mrem/yr) and television viewing (1 to 10 mrem/yr) also contribute to manmade exposure. The result is that about 50 to 100 mrem/yr is added to the natural background radiation exposure. Therefore, 200 to 250 mrem/yr is the exposure received by the average U.S.
resident.
11A.4.3 RADIATION IN PERSPECTIVE
Reference 14 gives a better perspective on the significance to health of various radiation doses and the experience forming the basis for permissible doses.
Most of the permissible exposure limits for persons have originated from the
accidental radiation exposure that occurred in the 1920s and 1930s, largely in the state of New Jersey, to radium-dial painters. In this industry at that time, workers had the habit of shaping their paint brushes with their tongue when painting radium on instrument dials. A considerable quantity of radium was ingested by a number of women. Although many of them have died as a result of this radiation exposure, many have shown no
significant radiation effects in the subsequent 30 to 40 yr. The dividing line seems to be that if the deposition of radium was less than about 0.5
µCi fixed in the bone marrow, there has been no significant subsequent radiation effect.
The radium deposition of 0.5
µCi in the bone will give an annual radiation exposure of about 150,000 mrem/yr.
UFSAR/DAEC-1 11A-33 Revision 19 - 9/07 The basic permissible doses have been determined largely from this radium exposure experience. Because there has been relatively little effect from 1
µCi of radium fixed in the bone and none from about 0.5
µCi, the permissible dose to bone has been established at 0.1
µCi deposited, which will result in a radiation exposure of 29,000 mrem/yr. Thus, there is a built-in safety factor of 5 to 10 included in any permissible limits based on this consideration as a result of this arbitrary reduction in the permissible
dose to bone.
The permissible occupational exposure of 5000 mrem/yr is determined largely from the permissible dose to bone, but there is an additional reduction of about 6 between the permissible dose to bone and the permissible full-body occupational dose.
When exposure to the neighbors of nuclear power plants is considered, the permissible occupational dose of 5000 mrem/yr c ontains an additional reduction factor of 10 resulting in the permissible general public dose of 500 mrem/yr. Therefore, it can be seen that between a dose per year that might cause injury, and the permissible dose to the
general public, there is an overall safety factor of several hundred. This is not to suggest that the permissible limits are too low, but merely to point out the substantial safety factors included in the current permissible radiation exposure limits. Such safety factors are perhaps greater than will be found in any other limits based on industrial hygiene or
public health considerations.
These are the permissible limits that apply to licensed facilities such as nuclear
power plants. There is, however, no expectation that the actual dose to the general public from the operation of DAEC will come anywhere near the permissible dose of 500 mrem/yr. In fact, the design objective for the DAEC is that the effects on neighbors will be far below any permissible radiation dose consideration and will be as low as reasonably achievable.
The actual expectancy is that the typi cal neighbor in the vicinity of the DAEC will receive a whole-body exposure that may be on the order of 1 mrem/yr from typical waste disposal operations, as averaged over the operating life of the plant. The further away from the plant, the estimated dose would be even lower because whole-body exposures from elevated stacks decrease with distance.
Radiation exposure to man from any one source must be viewed in proper perspective. One essential perspective is provided by considering the full spectrum of
other sources of ionizing radiation that peopl e are exposed to every day. Radioactivity released from nuclear power plants makes only a small contribution to this broad spectrum of natural background radiation.
The National Academy of Sciences has stated that radiation is by far the best understood environmental hazard. This scientific understanding has been translated into permissible levels of radiation exposure to man. These in turn have been incorporated into the limits set by the NRC, limits set in the interest of public health and safety.
UFSAR/DAEC-1 11A-34 Revision 13 - 5/97 11A.5
SUMMARY
The method of calculating a stack release limit is given along with partial verification of the method using data from Brookhaven National Laboratory. The whole-body gamma dose calculations are quite close to that observed at Brookhaven. The
ground-level integrated air concentration calculations give an order-of-magnitude type of verification because of the lack of sensitive field measurements.
Release rate calculations have been performed for the DAEC site using the onsite meteorological data.
Design objective release rates have been established for noble gases and
radioiodines and particulates having half-lives greater than 8 days.
It is expected that it should generally be feasible to keep average annual releases
of radioactivity in airborne effluents within their design objective levels.
At the same time, DAEC is permitted the flexibility of operation, compatible with
considerations of health and safety, to ensure that the public is provided a dependable source of power even under unusual operating conditions that may temporarily result in releases in excess of the design objective but still within the limits specified in 10 CFR
- 20. These limiting conditions of operation are presented in the Technical Specifications.
It is recognized that a precise determination of dose from a certain emission from the stack is only possible by direct measurement. Such information will be provided by the environmental monitoring program conducted at and around the site. If the stack emission ever reaches a level such that it is measurable in the environment, such measurements will provide a basis for adjusting the proposed stack limit long before the effect in the environment is of any safety concern.
UFSAR/DAEC-1 11A-35 Revision 13 - 5/97 REFERENCES FOR APPENDIX 11A
- 1. H. E. Cramer, "A Brief Survey of the Meteorological Aspects of Atmospheric Pollution," Bulletin of the American Meteor ological Society, Vol. 40, Issue 4, pp.
165-171.
- 2. H. C. Watson and C. C. Gamertsfelder, Environmental Radioactive Contamination as a Factor in Nuclear Plant Siting Criteria, HW-SA2808, 1964.
- 3. S. Glasstone and A. Sesonske, Nuclear Reactor Engineering, D. Van Nostrand Co., 1963.
- 4. H. Moses, G. H. Strom, and J. E.
Carson, Effects of Meteorological and Engineering Factors on Stack Plume Rise, Nu clear Safety, Vol. 6, Issue 1, 1964.
- 5. D. H. Slade (editor), Meteoroloqy and Atomic Energy 1968, TID 24190, pp. 189-198, 1968.
- 6. The Effects of Nuclear Weapons, Revised Edition, 1962.
- 7. M. J. May and I. F. Stuart, Comparison of Calculated and Measured Long-Term Gamma Dose from a Stack Effluent of Ra dioactive Gases, Paper presented at the Health Physic Mid-Year Symposium, Augusta, Georgia, 1968.
- 8. A. P. Hull, 1963 Environmental Radiation Levels at Brookhaven National Laboratory, BNL-915, 1964.
- 9. D. H. Slade (editor), Meteorology and Atomic Energy, AECU 3066, p. 100, 1955.
- 10. Meteorology and Atomic Energy 1968, TID 24190, p. 328, 1968.
- 11. M. M. Hendrickson, The Dose from Kr-85 Released to the Earth's Atmosphere, BNWL-SA-3233A or IAEA-SM-146/12, 1970.
- 12. H. J. Dunster, "Skin Deep Genetics," Health Physics, Vol. 17, Issue 6, p. 836, December, 1969.
- 13. V. P. Bond, "The Public and Radiation from Nuclear Power Plant," Atomic Industrial Forum, Collected Papers and Reference Sources Workshop on Radiation and Man's Environment, 1970.
- 14. National Academy of Science, Health Effects of Exposure to Low Levels of Ionizing Radiation (BEIRV), Washington D.C., National Academy Press, 1990.
UFSAR/DAEC-1 T11A-1 Revision 13 - 5/97 Table 11A.1-1 DEPOSITION VELOCITY COEFFICIENTS a
v d u o Stability Condition Particulates Halogens Very stable 1.5 x 10
-4 2.4 x 10-3 Moderately stable 2.2 x 10
-4 3.4 x 10-3 Neutral 3.0 x 10
-4 4.6 x 10-3 Unstable 6.0 x 10
-4 8.0 x 10-3
a To obtain the deposition velocity, multiply this ratio of deposition velocity to surface wind speed by the surface speed
()u o.
UFSAR/DAEC-1 T11A-2 Revision 13 - 5/97 Table 11A.1-2 DIFFUSION COEFFICIENTS
Stability Condition Constants Very Stable Moderately Stable Neutral Unstable a(m 2) 34 97 - - - -
b(m 2/sec) 0.025 0.33 - - - -
K 2 (sec-2) 8.8 x 10
-4 2.5 x 10 - - -
C z(u=1m/sec) - - - - 0.15 0.30 C z(u=5m/sec) - - - - 0.12 0.26 C z(u=10m/sec) - - - - 0.11 0.24 - - - - 0.25 0.20 UFSAR/DAEC-1 T11A-3 Revision 13 - 5/97 Table 11A.1-3 Page 1 of 3 NOBEL GAS ISOTOPES CONSTITUTING MIXTURE
Isotope Half-life
()EMeV ()NNosdis.'/ µper Meter µa per Meter Kr-83m 1.86 hr 1.3-2 a 1.6-1 1.3-1 1.3-1 Kr-85m 4.4 hr 1.3-2 5.2-2 1.3-1 1.3-1 1.5-1 7.7-1 1.8-1 3.3-3 3.1-1 1.4-1 1.4-1 3.7-3 Kr-85 10.76 yr 5.1-1 4.4-3 1.1-2 3.9-3 Kr-87 76 min 4.0-1 5.9-1 1.2-2 3.8-3 6.7-1 2.5-2 9.8-3 3.8-3 8.4-1 8.0-3 88.9-3 3.7-3 8.5-1 8.1-2 8.8-3 3.6-3 1.2+0 1.4-2 7.6-3 3.5-3 1.3+0 7.5-3 7.0-3 3.4-3 1.4+0 5.5-3 6.9-3 3.4-3 1.4+0 4.6-2 5.7-3 3.2-3 2.6+0 1.5-2 4.4-3 2.9-3 3.1+0 1.0-2 4.5-3 2.6-3 Kr-88 2.8 hr 1.7-1 6.9-2 1.7-2 3.4-3 2.0-1 3.8-1 1.6-2 3.5-3 3.6-1 3.0-2 1.3-2 3.8-3 4.3-1 1.2-2 1.2-2 3.9-3 8.3-1 1.3-1 9.0-3 3.7-3 1.1+0 3.4-2 8.0-3 3.5-3 1.5+0 1.5-1 6.8-3 3.3-3 2.0+0 9.6-2 5.6-3 3.0-3 2.3+0 5.7-1 5.4-3 2.9-3 Kr-89 3.2 min 2.2-1 2.8-1 1.6-2 3.5-3 3.2-1 1.4-1 1.3-2 3.8-3 4.7-1 1.5-1 1.2-2 3.9-3 5.9-1 3.2-1 1.2-2 3.9-3 7.3-1 7.5-2 9.6-3 3.8-3 9.1-1 3.0-1 8.6-3 3.6-3 1.8+0 5.5-1 6.1-3 3.2-3 3.4+0 1.1-1 4.6-3 2.6-3 4.4+0 1.3-2 3.7-3 2.4-3 2.9+0 4.7-2 4.7-3 2.7-3
a 1.3-2 = 1.3 x 10
-2 UFSAR/DAEC-1 T11A-4 Revision 13 - 5/97 Table 11A.1-3 Page 2 of 3 NOBEL GAS ISOTOPES CONSTITUTING MIXTURE
Isotope Half-life
()EMeV ()NNosdis.'/ µper Meter µa per Meter Kr-90 33 sec 1.1-1 1.2-1 2.6-2 3.0-3 1.2-1 5.4-1 1.9-2 3.1-3 1.7-1 4.1-2 1.7-2 3.4-3 2.4-1 1.2-1 1.6-2 3.6-3 4.3-1 2.1-1 1.2-2 3.8-3 5.9-1 5.6-1 1.1-2 3.9-3 1.1+0 5.3-1 7.6-3 3.5-3 1.7+0 3.2-1 6.2-3 3.2-3 2.3+0 7.3-2 5.4-3 2.9-3 2.9+0 8.9-2 4.7-3 2.6-3 Kr-91 10 sec 1.1-1 7.5-1 2.0-2 3.1-3 5.1-1 1.5-1 1.1-2 3.8-3 Kr-93 2 sec 2.0-1 5.0-1 1.6-2 3.6-3 Xe-131m 11.7 days 3.0-2 5.9-1 4.6-2 2.0-2 1.6-1 2.3-2 1.7-2 3.4-3 Xe-133m 2.3 days 3.0-2 1.4-1 4.0-2 2.0-2 3.4-2 3.2-2 3.9-2 1.6-2 2.3-1 8.0-2 1.5-2 3.6-3 Xe-133 5.27 days 3.1-2 3.8-1 4.0-2 1.6-2 3.5-1 8.6-2 3.8-2 1.4-2 8.0-2 6.0-3 2.1-2 3.8-3 8.1-2 3.7-1 2.0-2 4.5-3 1.6-1 6.4-4 1.7-2 3.4-3 2.2-1 2.4-6 1.6-2 3.6-3 3.0-1 5.1-5 1.4-2 3.7-3 3.8-1 2.3-5 1.3-2 3.8-3 Xe-135m 15.6 min 3.0-2 1.4-1 4.0-2 1.6-2 5.3-1 8.2-1 1.1-2 3.8-3 Xe-135 9.2 hr 3.1-3 4.5-2 4.0-2 1.6-2 1.6-1 2.1-3 1.7-2 3.3-3 2.0-1 2.6-4 1.6-2 3.5-3 2.5-1 9.2-1 1.5-2 3.6-3 3.6-1 2.2-3 1.4-2 3.8-3 3.7-1 1.1-4 1.3-2 3.8-3 4.1-1 3.1-1 1.2-2 3.8-3 5.7-1 5.0-5 1.1-2 3.8-3 6.1-1 2.4-2 1.1-2 3.8-3 7.6-1 1.3-3 4.4-3 3.7-3 UFSAR/DAEC-1 T11A-5 Revision 13 - 5/97 Table 11A.1-3 Page 3 of 3 NOBEL GAS ISOTOPES CONSTITUTING MIXTURE
Isotope Half-life
()EMeV ()NNosdis.'/ µper Meter µa per Meter Xe-137 3.6-2 7.0-2 1.8-2 3.3-3 1.5-1 5.6-2 1.3-2 3.8-3 3.7-1 1.7-3 1.3-2 3.8-3 4.6-1 3.0-1 1.2-2 3.9-3 6.6-1 4.0-4 1.1-2 3.9-3 8.5-1 6.5-3 8.9-3 3.7-3 1.0+0 4.4-3 8.1-3 3.6-3 1.2+0 6.0-3 7.4-3 3.5-3 1.7+0 8.9-3 6.2-3 3.3-3 2.5+0 4.7-3 5.0-3 2.9-3 Xe-138 14 min 3.0-2 3.0-2 4.0-2 2.0-2 1.6-1 7.8-2 1.8-2 3.3-3 2.4-1 3.6-2 1.5-2 3.6-3 2.6-1 3.7-1 1.4-2 3.7-3 4.0-1 7.4-1 1.3-2 3.8-3 4.1-1 2.8-2 1.2-2 3.8-3 4.3-1 2.3-1 1.2-2 3.9-3 1.8-1 2.0-1 6.1-3 3.2-3 2.0-1 1.6-1 5.7-3 3.1-3 Xe-139 41 sec 1.2-1 6.0-3 2.0-2 3.1-3 1.8-1 2.4-1 1.7-2 3.4-3 2.2-1 7.2-1 1.6-2 3.5-3 2.3-1 2.8-2 1.5-2 3.6-3 2.4-1 1.6-1 1.4-2 3.7-3 3.5-1 1.1-2 1.3-2 3.8-3 4.1-1 1.1-1 1.3-2 3.8-3 5.8-1 8.9-2 1.1-2 3.9-3 Xe-140 13.7 sec 8.3-2 1.5-1 2.1-2 3.7-3 1.2-1 1.7-1 1.9-2 3.1-3 2.1-1 8.0-2 1.6-2 3.6-3 2.9-1 9.0-2 1.4-2 3.7-3 4.2-1 1.3-1 1.2-2 3.8-3 4.8-1 7.8-2 1.2-2 3.9-3 5.9-1 3.4-1 1.1-2 3.8-3 7.0-1 2.3-1 9.8-3 3.8-3 8.6-1 4.9-1 8.8-3 3.7-3 1.4+0 4.4-1 7.1-3 3.4-3 UFSAR/DAEC-1 T11A-6 Revision 13 - 5/97 Table 11A.1-4 PARTICULATE DUAGHTER PRODUCTS Isotope Half-life E (MeV) µper Meter
µa per Meter Rb-88 18 min 9.91 0.0085 0.0037 1.28 0.0072 0.0034 1.85 0.0060 0.0032 2.18 0.0050 0.0030 4.2 0.0038 0.0024
CS-138 32.2 min 0.14 0.018 0.0033 0.19 0.016 0.0035 0.23 0.015 0.0037 0.41 0.0122 0.0037 0.46 0.0116 0.0038 0.55 0.0108 0.0038 0.87 0.0088 0.0037 1.01 0.0082 0.0036 1.43 0.0068 0.0034 2.21 0.0055 0.003 2.63 0.0050 0.0039 3.34 0.0043 0.0026
UFSAR/DAEC-1 T11A-7 Revision 13 - 5/97 Table 11A.1-5 DOSE REDUCTION FACTORS FOR SHIELDING Structure Location Dose Reduction Factor One-story frame First floor, middle 0.4
Basement, middle 0.07
One-story brick veneer First floor, middle 0.3
Basement, middle 0.06
Two-story brick veneer First floor, middle 0.23
Basement, middle 0.03
Multistory reinforced
concrete Upper floors (excluding top floor) 0.02 Basement 0.001
UFSAR/DAEC-1 T11A-8 Revision 13 - 5/97 Table 11A.2-1 FREQUENCY OF OCCURRENCE, IN PERCENT, OF VARIOUS WIND SPEEDS AND DIRECTIONS OBSERVED AT THE BROOKHAVEN NATIONAL LABORATORY SITE UNDER VERY STABLE ATMOSPHERIC CONDITIONS a,b Wind Wind Speed c (mph) All Wind Direction 0-3 4-7 8-12 13-18 19-24 >24 Speeds N 0.0619 0.108 0.433 0.449 0 0 1.05 NNE 0.0619 0.139 0.155 0.124 0 0 0.48 NE 0.0309 0.0464 0.155 0.0309 0 0 0.26 ENE 0 0.0928 0.124 0.0309 0 0 0.25 E 0.0464 0.0774 0.155 0.0774 0.0619 0 0.42 ESE 0.108 0.201 0.263 0.0155 0 0 0.59 SE 0.0464 0.0309 0.0464 0.0155 0.0155 0 0.15 SSE 0.0619 0.139 0.201 0.124 0 0 0.53 S 0.0464 0.201 0.248 0.356 0.232 0.0309 1.11 SSW 0.0619 0.294 0.665 1.13 1.01 0.0928 3.25 SW 0.108 0.170 0.433 1.22 0.897 0 2.83 WSW 0.0928 0.124 0.340 1.11 0.557 0.0155 2.24 W 0.0464 0.186 0.804 0.712 0.541 0.309 2.32 WNW 0.0774 0.186 0.572 0.433 0.139 0 1.41 NW 0.0309 0.186 0.433 0.603 0.186 0 1.44 NNW 0.0155 0.124 0.433 0.789 0.0774 0 1.44 All 0.90 2.31 5.46 7.22 3.71 0.17 19.77 a Frequency of occurrence is based on 6464 hr of good observation during 1963.
b Stability-based temperature difference measured at the 410-ft and 37-ft levels.
c Wind speed measured at the 355-ft level.
UFSAR/DAEC-1 T11A-9 Revision 13 - 5/97 Table 11A.2-2 FREQUENCY OF OCCURRENCE, IN PERCENT, OF VARIOUS WIND SPEEDS AND DIRECTIONS OBSERVED AT THE BROOKHAVEN NATIONAL LABORATORY SITE UNDER MODERATELY STABLE ATMOSPHERIC CONDITIONS a,b Wind Wind Speed c (mph) All Wind Direction 0-3 4-7 8-12 13-18 19-24 >24 Speeds N 0 0.186 0.433 0.433 0.232 0.0464 1.33 NNE 0.0464 0.278 0.557 0.464 0.186 0.0309 1.56 NE 0.0774 0.232 0.557 0.124 0.0928 0 1.08 ENE 0.0309 0.139 0.139 0.201 0.155 0 0.66 E 0.0309 0.186 0.263 0.278 0.232 0.0619 1.05 ESE 0.0928 0.139 0.371 0.278 0.186 0.0619 1.13 SE 0.0309 0.0774 0.155 0.387 0.0619 0 0.71 SSE 0.0309 0.0619 0.495 0.696 0.418 0.449 2.15 S 0.0928 0.108 0.402 1.08 0.804 0.201 2.69 SSW 0.139 0.232 0.913 1.90 1.36 0.108 4.66 SW 0.0464 0.232 0.495 1.53 0.572 0.0619 2.94 WSW 0.0464 0.108 0.449 1.44 0.480 0.0774 2.60 W 0 0.139 0.387 1.42 1.01 0.170 3.12 WNW 0.0464 0.139 0.371 0.727 1.07 0.0464 2.40 NW 0.0464 0.139 0.371 0.743 0.727 0.0309 2.06 NNW 0.0464 0.155 0.655 1.25 0.309 0.0619 2.49 All 0.80 2.55 7.02 12.96 7.89 1.41 32.63
a Frequency of occurrence is based on 6464 hr of good observation during 1963.
b Stability-based temperature difference measured at the 410-ft and 37-ft levels.
c Wind speed measured at the 355-ft level.
UFSAR/DAEC-1 T11A-10 Revision 13 - 5/97 Table 11A.2-3 FREQUENCY OF OCCURRENCE, IN PERCENT, OF VARIOUS WIND SPEEDS AND DIRECTIONS OBSERVED AT THE BROOKHAVEN NATIONAL LABORATORY SITE UNDER NEUTRAL ATMOSPHERIC CONDITIONS a,b Wind Wind Speed c (mph) All Wind Direction 0-3 4-7 8-12 13-18 19-24 >24 Speeds N 0.0619 0.217 0.526 0.325 0.170 0.0309 1.331 NNE 0.0928 0.340 0.495 0.774 0.464 0.0155 1.764 NE 0.0774 0.480 0.526 0.217 0.0155 0 1.316 ENE 0.0928 0.186 0.371 0.325 0.155 0.0155 1.145 E 0.0464 0.0928 0.263 0.155 0 0 0.557 ESE 0.139 0.449 0.743 0.278 0.0774 0.0155 1.702 SE 0.0309 0.217 0.464 0.124 0.0309 0.0155 0.882 SSE 0.0309 0.248 1.45 0.665 0.232 0.201 2.827 S 0.155 0.402 1.58 1.42 0.603 0.0774 4.237 SSW 0.186 0.172 1.67 0.93 0.774 0.0619 5.334 SW 0.139 0.294 0.712 0.804 0.433 0.139 2.521 WSW 0.124 0.263 0.882 1.18 0.990 0.325 3.764 W 0.155 0.248 0.789 1.39 1.73 0.975 5.287 WNW 0.124 0.248 0.851 1.01 1.30 0.619 4.152 NW 0.0464 0.294 0.511 0.851 0.866 0.263 2.831 NNW 0.0619 0.464 0.619 0.990 0.402 0.928 2.630 All 1.56 5.15 12.45 12.44 7.83 2.85 42.28
a Frequency of occurrence is based on 6464 hr of good observation during 1963.
b Stability-based temperature difference measured at the 410-ft and 37-ft levels.
c Wind speed measured at the 355-ft level.
UFSAR/DAEC-1 T11A-11 Revision 13 - 5/97 Table 11A.2-4 FREQUENCY OF OCCURRENCE, IN PERCENT, OF VARIOUS WIND SPEEDS AND DIRECTIONS OBSERVED AT THE BROOKHAVEN NATIONAL LABORATORY SITE UNDER UNSTABLE ATMOSPHERIC CONDITIONS a,b Wind Wind Speed c (mph) All Wind Direction 0-3 4-7 8-12 13-18 19-24 >24 Speeds N 0 0 0.0928 0.0464 0 0 0.1392 NNE 0 0 0.0464 0.0155 0 0 0.0619 NE 0 0 0.0619 0 0 0 0.0619 ENE 0 0 0 0 0.0619 0.0155 0.0774 E 0 0 0 0.0155 0.0155 0 0.031 ESE 0 0 0.0155 0 0 0 0.0155 SE 0 0 0 0 0 0 0 SSE 0 0 0.0155 0.0309 0 0 0.0464 S 0 0 0.0928 0.248 0.232 0.0464 0.6192 SSW 0 0 0.0155 0.217 0.0774 0 0.3099 SW 0 0 0.0619 0.0619 0 0.0155 0.1393 WSW 0 0 0.0309 0.155 0.170 0.0928 0.4487 W 0 0 0.0619 0.402 0.541 0186 0.1909 WNW 0 0.0309 0.0619 0.495 0.402 0.201 1.1908 NW 0 0 0.186 0.309 0.139 0.0155 0.6495 NNW 0 0.0155 0.0928 0.155 0.0619 0 0.3252 All 0.0 0.05 0.84 2.15 1.70 0.57 5.32
a Frequency of occurrence is based on 6464 hr of good observation during 1963.
b Stability-based temperature difference measured at the 410-ft and 37-ft levels.
c Wind speed measured at the 355-ft level.
UFSAR/DAEC-1 T11A-12 Revision 13 - 5/97 Table 11A.2-5 FREQUENCY OF OCCURRENCE, IN PERCENT, OF ALL WIND SPEEDS AND DIRECTIONS OBSERVED AT THE BROOKHAVEN NATIONAL LABORATORY SITE UNDER ALL ATMOSPHERIC CONDITIONS a,b Wind Wind Speed c,d (mph) All Stability Direction VS MS N U Conditions N 1.05 1.33 1.32 0.14 3.85 NNE 0.48 1.56 1.76 0.06 3.86 NE 0.26 1.08 1.31 0.06 2.72 ENE 0.25 0.66 1.15 0.08 2.14 E 0.42 1.05 0.56 0.03 2.06 ESE 0.59 1.13 1.70 0.02 3.44 SE 0.15 0.71 0.88 0.00 1.74 SSE 0.53 2.15 2.83 0.05 5.56 S 1.11 2.69 4.24 0.62 8.66 SSW 3.25 4.66 5.33 0.31 13.55 SW 2.83 2.94 2.52 0.14 8.43 WSW 2.24 2.60 3.76 0.45 9.05 W 2.32 3.12 5.29 1.19 11.92 WNW 1.41 2.40 4.15 1.19 9.15 NW 1.44 2.06 2.83 0.65 6.98 NNW 1.44 2.49 2.63 0.33 6.89 All 19.77 32.63 42.28 5.32 100.00
a Frequency of occurrence is based on 6464 hr of good observation during 1963; 12 hr of calm (wind speed less than 0.5 mph).
b Stability-based temperature difference measured at the 410-ft and 37-ft levels.
c Wind speed measured at the 355-ft level.
d Key: VS = very stable; MS = moderately stable; N = neutral; U = unstable.
UFSAR/DAEC-1 T11A-13 Revision 13 - 5/97 Table 11A.2-6 FREQUENCY OF OCCURRENCE, IN PERCENT, OF VARIOUS WIND SPEEDS AND DIRECTIONS OBSERVED AT THE BROOKHAVEN NATIONAL LABORATORY SITE UNDER ALL ATMOSPHERIC CONDITIONS a,b Wind Wind Speed c (mph) All Wind Direction 0-3 4-7 8-12 13-18 19-24 >24 Speeds N 0.124 0.510 1.48 1.25 0.402 0.077 3.85 NNE 0.201 0.758 1.25 1.38 0.232 0.046 3.87 NE 0.186 0.758 1.30 0.371 0.108 0 2.72 ENE 0.124 0.418 0.634 0.557 0.371 0.031 2.13 E 0.124 0.356 0.681 0.526 0.303 0.062 2.06 ESE 0.340 0.789 1.39 0.572 0.263 0.077 3.43 SE 0.108 0.325 0.665 0.526 0.108 0.015 1.75 SSE 0.124 0.449 2.16 1.52 0.650 0.650 5.55 S 0.294 0.712 2.32 3.11 1.87 0.356 8.66 SSW 0.387 1.24 3.26 5.18 3.22 0.263 13.55 SW 0.294 0.696 1.70 3.62 1.90 0.216 8.43 WSW 0.263 0.495 1.70 3.98 2.20 0.510 9.05 W 0.201 0.572 2.04 3.93 3.82 1.36 11.93 WNW 0.247 0.603 1.86 2.66 2.91 0.866 9.14 NW 0.124 0.619 1.50 2.51 1.92 0.309 6.98 NNW 0.124 0.758 1.81 3.19 0.851 0.155 6.88 All 3.42 10.05 25.77 34.78 21.13 5.00 100.00
a Frequency of occurrence is based on 6464 hr of good observation during 1963.
b Stability-based temperature difference measured at the 410-ft and 37-ft levels.
c Wind speed measured at the 355-ft level.
UFSAR/DAEC-1 T11A-14 Revision 13 - 5/97 Table 11A.2-7 1963MONTHLY AVERAGE GAMMA DOSE FOR MONITORING STATIONS AROUND THE BROOKHAVEN GRAPHITE RESEARCH REACTOR SITE a,b (in mR/wk of AR-41)
Onsite Stations Perimeter Stations Off-site Station Month E-10 E-11 E-12 E-2 E-4 E-7 E-9 0-6 January 1.46 2.08 2.59 0.45 0.26 0.28 0.76 0 February 0.06 2.22 2.92 0.06 0.11 0.76 0.76 0.03 March 0.68 2.58 2.25 0.58 0.05 0.57 0.57 0.03 April 0.78 1.94 2.59 0.14 0.19 1.08 0.74 0.01 May 0.44 6.55 5.19 0.43 0.24 0.41 1.86 0.01 June 0.85 2.31 2.43 0.82 0.32 0.57 0.74 0.02 July 0.35 2.56 4.30 0.47 0.25 0.42 1.49 0.03 August 0.64 3.18 5.02 0.17 0.01 0.48 1.02 0 September 1.63 3.07 3.83 0.21 0.70 0.27 0.55 0.03 October 1.51 2.68 3.46 0.41 0.57 0.53 0.80 0.02 November 0.90 2.16 3.40 0.31 0.39 0.45 0.58 0.04 December 0.58 1.60 1.17 0.19 0.25 0.39 0.35 0.04 Average 0.82 2.74 3.26 0.35 0.28 0.52 0.85 0.02
Peak weekly average 3.23 12.91 7.57 1.97 1.94 1.63 2.29 1.08 a From Brookhaven National Laboratory publication BNL 915.
b Estimated error at 90% confidence level, 0.25 mR/wk.
UFSAR/DAEC-1 T11A-15 Revision 13 - 5/97 Table 11A.2-8 1963 AVERAGE ANNUAL MEASURED AND CALCULATED GAMMA DOSE RATES FOR MONITORING STATIONS AROUND THE BROOKHAVEN GRAPHITE RESEARCH REACTOR SITE Direction Distance from Dose (mR/yr)
Station Sector Stack (m)
Measured a Calculated b
E-2 NW 1100 21 c 20 E-4 WSW 2200 14 13 E-7 SE 2500 28 30 E-9 NE 2750 45 34 E-10 W 520 40 42 E-11 S 420 140 122 E-12 NNW 460 158 156
a Based on a 10-month average.
b Based on an 85% operation factor giving a release rate of 0.127 Ci/sec.
c Based on a 9-month average.
UFSAR/DAEC-1 T11A-16 Revision 13 - 5/97 Table 11A.3-1 NOBLE RADIOGAS RELEASE RATE TO ENVIRONS FROM 12-BED RECOMBINER/CHARCOAL SYSTEM a Noble Radiogas Release Rate (µCi/sec) Kr-83m 2.16 Kr-85m 266.0
Kr-85 23.8
Kr-87 0.366
Kr-88 144.0 Xe-131m 6.46 Xe-133m 3.08
Xe-133 1190.0
All radiogases 1640.0 a Based on 100,000
µCi/sec at 30-minute-decay diffusion mixture and a condenser inleakage of 18.5 scfm.
UFSAR/DAEC-1 T11A-17 Revision 13 - 5/97 Table 11A.3-2 NOBLE RADIOGAS RELEASE RATE TO ENVIRONS FROM GLAND SEAL SYSTEM a Noble Radiogas Release Rate (µCi/sec) Kr-83m 3.44 Kr-85m 6.14
Kr-85 0.0238
Kr-87 19.4
Kr-88 20.0
Kr-89 87.0
Kr-90 29.6
Kr-91 0.0698 Xe-131m 0.0151 Xe-133m 0.278
Xe-133 8.18 Xe-135m 24.4
Xe-135 22.2
Xe-137 108.0
Xe-138 82.6
Xe-139 45.4
Xe-140 1.74
All radiogases 464.0 a Based on 100,000
µCi/sec at 30-minute-decay diffusion mixture; 0.1% to gland seal system.
UFSAR/DAEC-1 T11A-18 Revision 13 - 5/97 Table 11A.3-3 ANNUAL AVERAGE GAMMA DETECTOR DOSE AT GROUND LEVEL FROM RECOMBINER/CHARCOAL SYSTEM a (Based on a continuous release rate of 1640
µCi/sec)
Site Boundary Dose (mrem/yr) as Function of Distance from Stack (km)
Direction Sector Distance From Stack Dose (mrem/yr) 3.218 6.436 12.067 40.225 72.405 N 1176 0.967 0.420 0.255 0.089 0.008 0.001 NNE 1203 0.429 0.226 0.137 0.046 0.004 0.001 NE 695 0.434 0.191 0.112 0.037 0.003 0.001 ENE 642 0.325 0.164 0.097 0.033 0.003 0.001 E 535 0.474 0.270 0.154 0.050 0.004 0.001 ESE 455 0.607 0.342 0.193 0.063 0.006 0.001 SE 588 0.601 0.353 0.207 0.067 0.005 0.001 SSE 481 0.843 0.296 0.171 0.054 0.004 0.001 S 455 0.647 0.237 0.140 0.046 0.004 0.001 SSW 535 0.475 0.169 0.094 0.029 0.002 b SW 668 0.361 0.138 0.079 0.025 0.002 b WSW 749 0.317 0.158 0.091 0.028 0.002 b W 668 0.343 0.136 0.081 0.027 0.002 b WNW 722 0.657 0.270 0.156 0.049 0.004 0.001 NW 936 0.626 0.209 0.125 0.041 0.003 0.001 NNW 1257 0.687 0.247 0.150 0.051 0.004 0.001
a Thirty-minute-decay diffusion mixt ure of noble radiogases would be 100,00
µCi/sec. b Negligible dose.
UFSAR/DAEC-1 T11A-19 Revision 13 - 5/97 Table 11A.3-4 ANNUAL AVERAGE GAMMA DETECTOR DOSE AT GROUND LEVEL FROM GLAND SYSTEM a (Based on a continuous release rate of 464
µCi/sec) Direction Sector Distance from Stack (m) Dose (mrem/yr) N 1176 0.425 NNE 1203 0.194 NE 695 0.218 ENE 642 0.161 E 535 0.240 ESE 455 0.305 SE 588 0.331 SSE 481 0.472 S 455 0.383 SSW 535 0.221 SW 668 0.159 WSW 749 0.145 W 668 0.178 WNW 722 0.306 NW 936 0.285 NNW 1257 0.290
a Thirty-minute-decay diffusion mixture of noble gases would be 100
µCi/sec.
UFSAR/DAEC-1 T11A-20 Revision 13 - 5/97 Table 11A.3-5 ANNUAL AVERAGE GAMMA DETECTOR DOSE AT GROUND LEVEL FROM BUILDING VENTILATION SYSTEM a (Based on a continuous release rate of 278
µCi/sec) Direction Sector Distance from Stack (m) Dose (mrem/yr) N 1176 0.881 NNE 1203 0.502 NE 695 0.764 ENE 642 0.518 E 535 0.624 ESE 455 0.772 SE 588 0.799 SSE 481 1.066 S 455 1.061 SSW 535 1.005 SW 668 0.900 WSW 749 0.783 W 668 0.842 WNW 722 0.818 NW 936 0.471 NNW 1257 0.545 a Five-minute-old mix, thirty-minute-decay diffusion mixture of noble gases would be 100 µCi/sec.
UFSAR/DAEC-1 T11A-21 Revision 13 - 5/97 Table 11A.3-6 ANNUAL AVERAGE GAMMA DETECTOR DOSE VERSUS DISTANCE FOR MAXIMUM DIRECTION FROM COMBINED CONTRIBUTIONS OF GLAND SEAL AND AIR EJECTOR OFFGAS SYSTEMS a
Distance from Stack Dose km miles (mrem/yr) 1.2 0.73 1.39 3.2 2 0.537 6.4 4 0.308 12.1 7.5 0.104 40.2 25 0.009 72.4 45 0.002
a 2104 µCi/sec release from stack.
UFSAR/DAEC-1 T11A-22 Revision 13 - 5/97 Table 11A.3-7 ANNUAL AVERAGE GAMMA DOSE FROM COMBINED CONTRIBUTIONS OF GLAND SEAL AND AIR EJECTOR OFFGAS SYSTEMS AT POINT ON SITE BOUNDARY FURTHEST FROM STACK FOR DIFFERENT LEVELS OF FUEL-DEFECT OPERATION Annual Average 30-Min-Decay Release Rate (µCi/sec) Total Fixed-Point or Diffusion Mixture
(µCi/sec) Recombiner/ Charcoal System 1.75-Min Gland Seal System Detector Dose (mrem/yr) 25,000 a 410 116 0.35 100,00 b 1,640 464 1.39 718,400 c 11,800 3,300 10.00
a Typical release rate corresponding to normal operation.
b Release rate that is expected only infrequently.
c Design-objective release rate.
UFSAR/DAEC-1 T11A-23 Revision 13 - 5/97 Table 11A.3-8 AVERAGE WHOLE-BODY DOSES FOR POPULATION DISTRIBUTION FROM COMBINED CONTRIBUTIONS OF STEAM JET AIR EJECTOR AND GLAND SEAL SYSTEMS a,b (Anuual Doses in man-rem; based on a continuous release rate of 2104
µCi/sec)) Distance Distance from Stack (miles)
Sector SB-3 c 3-5 5-10 10-20 20-30 30-40 40-50 SB-50 c N 1.6-2 d 2.6-2 1.4-1 3.0-2 6.4-3 6.9-3 7.6-3 2.3-1 NNE 1.1-2 1.1-2 1.2-2 3.5-2 5.1-3 5.1-3 3.4-3 8.3-2 NE 9.0-3 8.7-3 9.6-3 2.7-2 3.9-3 4.0-3 2.7-3 6.5-2 ENE 8.1-3 9.2-3 2.0-2 3.8-2 8.9-3 3.4-3 2.1-3 8.9-2 E 2.2-2 1.8-2 2.5-2 5.9-2 1.4-2 5.4-3 3.4-3 1.5-1 ESE 2.2-2 3.7-2 1.2+0 9.3-1 2.2-2 3.7-2 4.7-3 2.2+0 SE 6.2-2 6.8-2 3.0+0 9.9-1 2.2-2 3.5-2 4.4-3 4.2+0 SSE 5.5-2 3.9-2 2.2+0 5.2-2 1.4-2 8.1-3 2.9-3 2.3+0 S 9.0-3 7.1-3 1.7-1 4.6-2 1.2-2 7.2-3 2.5-3 2.6-1 SSW 4.5-2 1.9-2 2.4-2 9.7-3 5.1-3 1.4-3 7.2-4 1.0-1 SW 5.0-3 3.7-3 6.9-3 8.4-3 4.3-3 1.2-3 6.0-4 3.0-2 WSW 4.7-3 1.1-2 7.2-3 1.6-2 2.6-3 1.0-3 1.2-3 4.5-2 W 7.9-3 3.5-2 1.1-2 1.7-2 2.8-3 1.1-3 1.4-3 7.6-2 WNW 1.2-2 1.2-2 1.3-2 2.0-2 6.6-3 1.7-2 3.5-2 1.1-1 NW 1.1-2 5.6-3 1.1-2 1.7-2 5.9-3 1.5-2 3.0-2 9.6-2 NNW 4.0-3 6.3-3 4.2-2 1.7-2 3.7-3 4.0-3 4.3-3 8.1-2 Total: 10.2 Natural background total: 139,000
a Whole-body doses are doses to individuals. A combined occupancy and shielding factor of 2 has been assumed.
b Population estimate for year 2010.
c SB = site boundary.
d 1.6-2 = 1.6 x 10
-2.
UFSAR/DAEC-1 T11A-24 Revision 13 - 5/97 Table 11A.3-9 Page 1 of 2 INFORMATION ON DAIRY FARMS SURROUNDING THE DAEC SITE IN 1971, BY DIRECTION SECTOR Direction Sector Distance from Stack (miles) Azimuth (from true north)
Farm Location Number of Dairy Cows Months on Pasture Distributor Location of Distributor N 3.5 010° August Holub Section 27, Washington
- Township, Linn Co.,
Center Point, IA 20 May-October Ryan Cooperative Ryan, IA NNE 5.0 029° Melvin Cress Section 13, Creek,
- Township, Linn Co.,
Center Point, IA 50 May-October Mid-America Dairymen Inc. Marion, IA NE 6.7 048° Bernice Haehlen Section 17, Creek
- Township, Linn Co.,
Alburnett, IA 22 May-October Maquoketa Valley Cooperative Ryan, IA ENE a E a ESE a SE a SSE a a No dairy farms in this sector within 10 miles of DAEC site.
UFSAR/DAEC-1 T11A-25 Revision 13 - 5/97 Table 11A.3-9 Page 2 of 2 INFORMATION ON DAIRY FARMS SURROUNDING THE DAEC SITE IN 1971, BY DIRECTION SECTOR Direction Sector Distance from Stack (miles) Azimuth (from true north)
Farm Location Number of Dairy Cows Months on Pasture Distributor Location of Distributor S 6.0 185° George Young Section 9, Palo, IA 53 b Mid-America Dairymen, Inc. Marion, IA SSW 2.0 216° M. Van Note Section 20, Palo, IA 40 May - October Mid- America Dairymen, Inc. Marion, IA SW 3.4 219° Ronald Beatty Section 19, Palo, IA 40 May - October Mid- America Dairymen, Inc. Marion, IA WSW a W 2.6 272° Rose Myers Section 7, Shellsburg, IA 5 April - November Walker Creamery Walker, IA WNW 1.6 298° W. W. Andrews Section 5, 32 April - November Ryan Cooperative Ryan, IA NW 7.5 317° Raymond Lerch Section 14, Vinton, IA 30 May - October Wapsie Vally Cooperative Maquoketa, IA NNW 4.6 346° Cliff Gott Section 20, Center Point, IA 12 May - October Maquoketa Valley Cooperative Ryan, IA W 7.2 260° Donald Beatty Section 17, Shellsburg, IA 75 4 months Mid- America Dairymen, Inc. Marion, IA b Not applicable; feed local chopped alfalfa to confined animals.
a No dairy farms in this sector within 10 miles of DAEC site.
UFSAR/DAEC-1 T11A-26 Revision 13 - 5/97 Table 11A.3-10 MONTHLY PRODUCTION OF INDIVI DUAL FARMS VERSUS MONTHLY PRODUCTION OF DISTRIBUTOR, JUNE 1971 Distributor Farm Name Production (lb) Name Production (lb)
Maquoketa Valley
Cooperative 8,623,969 A. Holub 17,696 a B. Haehlen 9,726 a C. Gott 6,968 a W. Andrews 28,360 a
Walker Creamery 3,366 R. Myers 54 b
Wapsie Valley
Cooperative 8,882,187 R. Lerch 22,710 a Mid-America Dairymen, Inc. 13,008,275 M. Cress 16,312 G. Young 29,867 M. Van Note 53,221 R. Beatty 40,237 D. Beatty 46,716 c
a Class B dairy milk production is not commercially distributed for direct consumption but rather is used for cheese, dried milk, etc.
b Product is butterfat.
c Product is distributed in Des Moines ar ea. Monthly production of distributor is 14,453,485 lb.
UFSAR/DAEC-1 T11A-27 Revision 13 - 5/97 Table 11A.3-11 GRAZING-SEASON INTEGRATED GROUND-LEVEL AIR CONCENTRATIONS FROM MAIN STACK RELEASE a,b Direction Sector Distance from Stack (m)
Concentration (µCi/cm 3) N 5,631 1.15 x10-16 NNE 8,040 5.08 x 10-17 NE 10,780 3.23 x 10-17 ENE c - - - -
E c - - - - ESE c - - - -
SE c - - - - SSE c - - - - S 9,650 3.54 x 10-17 SSW 3,220 1.13 x 10-16 SW 5,470 4.28 x 10-17 WSW c - - - - W 4,180 7.66 x 10-17 WNW 2,575 2.69 x 10-16 NW 12,067 2.36 x 10-17 NNW 7,401 5.16 x 10-17 a Grazing season is April 15 through November 15.
b Assumed I-131 release rate of 10
-2 µCi/sec. c No dairy farms in this sector within 10 miles of DAEC site.
UFSAR/DAEC-1 T11A-28 Revision 13 - 5/97 Table 11A.3-12 GRAZING-SEASON INTEGRATED-GROUND LEVEL AIR CONCENTRATIONS FROM BUILDING VENTILATION RELEASE a,b Direction Sector Distance from Stack (m)
Concentration (µCi/cm 3) N 5,631 1.28 x 10-16 NNE 8,040 6.06 x 10-17 NE 10,780 2.77 x 10-17 ENE c - - - -
E c - - - -
ESE c - - - -
SE c - - - -
SSE c - - - - S 7,650 1.06 x 10-17 SSW 3,220 5.22 x 10-17 SW 5,470 3.27 x 10-17 WSW c - - - - W 4,180 4.61 x 10-17 WNW 2,575 3.24 x 10-16 NW 12,067 1.36 x 10-17 NNW 7,401 4.46 x 10-17 a Grazing season is April 15 through November 15.
b Assumed I-131 release rate of 10
-3 µCi/sec. c No dairy farms in this sector within 10 miles of DAEC site.
GIFFORO-PASQUILLTYPEF/"/'"/'",/'././OISTANCEFROMSOURCE(meters)..DUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTVerticalCloudWidthversusDistance-VeryStableFigure GIFFORO-PASQUILL//'.//'/'/'100____102103104105DISTANCEFROMSOURCE(meters)DUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTVerticalCloudWidthversusDistance-ModeratelyStableFigure
///////-GIFFORO-PASQUILL,/TYPEC,/"1msec/'5m"sec/'10m'sec102100__102103104105OISTANCEFROMSOURCE(meterslDUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTVerticalCloudWidthversusDistance-NeutralFigure
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- 10-2Ezuw...w...ouzQo:ocodOIJ-a=ENERGYABSORPTIONIJ-=TOTALABSORPTIONIJ--IJ-k=aIJ-aI-zol.0u=><::>=>co10-3__--.l.__0.10.11.010GAMMAENERGY.MevDUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTGammaRadiationAbsorptionCoefficientsandBuildupConstantsforAir,STPFigure DOSERATEIMRHRIMODERATELYSTABLE&VERYSTA8lE10RATE,!'IIRi1RlDOSERATEIMRHRIUNSTABLE1001000OOWNWINOOISTANCE,METERS,DUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTGammaDoseRateforVariousWindSpeedsandStabilitiesforBGRRStack(ReleaseRateCi/sec)Figure WINOSPEED5Ill/secPLUMEHEIGHT182mRELEASERATE0,127Ci/secMODERATELYSTABLEVERYSTABLE10,00010010-3'-_-'-..-JI.-JL.-L..J-L..l..J-'-_--'-_-'--'-........J-l..J..J..__.l-...1--L...J-J...J..J...U10DOWNWINDDISTANCE(meters)DUANEARNOLDENERGYCENTER"IOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTGammaDoseRateinAirforVariousStabilityConditionsFigure r:--------------------------.,.NEUTRALSTABILITY5MSWINDSPEEDH=181METERSQC"secDOSERATEIMRHRI0.100.011001000DOWNWINDDISTANCE(METERS)SECTORI10.000DUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYFINALSAFETYANALYSISREPORTDoseRateinEachSectorFigure n131415SECTOR1STACK93457DUANEARNOLDENERGYCENTERIOWAELECTRICLIGHT&POWERCOMPANYUPDATEDFINALSAFETYANALYSISREPORTNomenclatureofSectorUsedforAveragingFigure