ML20216C101
| ML20216C101 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 03/26/1998 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| PROC-980326, NUDOCS 9804140307 | |
| Download: ML20216C101 (95) | |
Text
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4 March 26, 1998 3 ~
X244 ALL Document Control Desk Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Station PI 137 Washington,DC 20555 Attached is a revision to the Offsite Dose Calculation Manual, Braidwood Annex, Chapters 10 and
- 12. Please update your manual as follows:
Bsmas Braidwood Chapter 10, Revision 1.8 Braidwood Chapter 12, Revision 2 Imstt:
Braidwood Chapter 10, Revision 1.9 Braidwood Chapter 12, Revision 3 m
j Please sign below indicating your manual has been updated and that your controlled copy number is correct.
Name Date Return to :
Comed Central Files 1300 Opus Place,4th Floor Downe s Grove,IL 60515
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O Chapter 10 Change Summary i l- () ODCM Revision 1.9, January 1998 I
Pave Channe Descrintion I
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10-i Updated revision number. Deleted page revision inde: since individual pages are no longer revised. The revision number for the index is assigned to the chapter. Updated file designator.
10-ii, iii, Updated section titles and/or page numbers to reflect changes m text.
iv, y 10-1 Changed reference to UFSAR from "Section" to " Chapter" to be consistent with the wording used in the UFSAR. Removed reference to specific section of the
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UFSAR Chapter. Deleted information concerning low, mid, and high noble gas channels from Section 10.1.2.1 since that information applied to the WRGM monitors, not the 1/2RE-PR028 monitors.
10-3 Updated the description for setpoint determinations for the Auxiliary Building Vent EfIluent Monitors. Deleted discussion on Component Cooling Water Monitors since it is not considered to be an effluent pathway. Therefore, the setpoint calculations for this monitor are not applicable.
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10 4 Added definitions for each term used in equations 10-1 and 10-2 and deleted reference to Appendix A for definitions of the parameters. Updated the value for j
the limiting noble gas release rate based on current data. Updated the value for i tne procedural limit for each release path. Deleted reference to the fraction of the permissible station release hmit. 1 10-5 Changed reference to UFSAR from "Section" to " Chapter" to be consistent with the wording used in the UFS AR. Removed reference to specific section of the UFSAR Chapter.
10-6 Corrected typographical error in designator for monitor 1RE-PR0d2.
10-7,8,9,10 Changed order of Sections 10.2.3.1.1 through 10.2.4 to make reading easics.
l 10-7 Changed description of maximum permitted discharge flow rate in Equation 10-3 for the Station Blowdown Monitor to require use of the more restrictive of the radiological or chemistry limits.
10-7,8 Changed the name for F6 in Equation 10-4 to F', a, Maximum Radiological Permitted Discharee Flow Rate to reflect the radiological basis for the flow rate.
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F Braidwood Station Chapter 10 Change Summary Revision 1.9, continued Pace Chance Description 10-8 Corrected reference to EPA regulation for public drinking water limit from 40CFR190 to 40CFR141. Clarified the bases for the ' cooling pond tritium administrative action level and the appropriate LLD used for monitoring this pathway.
10-10 Deleted reference to Cs-137 in the section for Conversion Factors. The conversions are monitor specific.
10-11 Placed ; rid around table data to make it easier for the reader to identify the percent of annual release by nuclide.
10-12 Reorganized list of radionuclides by atomic weight to make it easier for the reader to find a specific nuclide.
10-15 Corrected typographical error in Figure 10-2.
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1 BRAIDWOOD Rsvision 1.9 January 1998 CHAPTER 10 BRAIDWOOD ANNEX INDEX l
! Revision 1.9 l
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BRAIDWOOD Rrvision 1.9 January 1998 CHAPTER 10 RADIOACTIVE EFFLUENT TREATMENT AND MONITORING TABLE OF CONTENTS SECTION PAGE 10.1 AIRBORNE RELEASES., . . . . . . .. . . . . . . . .1
- 1. System Desenption . .. .. .. . . . . ... . 1
- 1. Waste Gas Holdup System.. . . . .. ... . . .. . . . . . . . . . . 1
- 2. Ventilation Exhaust Treatment System.. .. . . . . . . . .. .. .. 1
- 2. Radiation Monitors . . . . . . .. . .. . . . . . . . . .. 1
- 1. Auxiliary Building Vent Effluent Monitors. .. . . . . . . . . . . . 1
- 2. Containment Purge Effluent Monitors.. . . . . . . . ...2
- 3. Waste Gas Decay Tank Monitors.. . ... . . .2
- 4. Gland Steam and Condenser Air Ejector Monitors.. . . . .. . .........2
- 5. Radwaste Building Ventilation Monitor. . . . . . .. 2
- 6. Component Cooling Water Monitor . . . . . . ..... . . . . . . .2
- 7. Miscellaneous Ventilation Monitors . .. . . . . ..... . . 3
- 3. Alarm and Trip Setpoints.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 3
- 1. Setpoint Calculations. . . . . . . . . . . . . . . . ... . . .. . . .3
- 1. Auxiliary Building Vent Effluent Monitors. .. . . . . .. ... . .. . 3
- 2. Containment Purge Effluent Monitors.. . . . .....3
- 3. Waste Gas Decay Tank Effluent Monitors.. . . . . .....3
- 2. Release Limits. . . . . . . . . . . . . .. .. . . . . .3
- 3. Release Mixture.. . . . .. .. . . . . . . .5
- 4. Conversion Factors . . . . . . . . 5
- 5. HVAC Flow Rates.. . . . . . . . . . .. . . ...5 4 Allocation of Effluents from Common Release Points.. . . . 5
- 5. Dose Projections for Batch Releases.. . . . . . . . . . .. .. .5 O
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. BRAIDWOOD Revision 1.9 Janutry 1998
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C CHAPTER 10 RADIOACTIVE EFFLUENT TREATMENT AND MONITORING TABLE OF CONTENTS
- SECTION PAGE 1
, 10.2 LIQUID RELEASES.. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .5 1
- 1. System Description . . . . . . ... . .. . . . . . . . 5
- 1. Release Tanks.. . . . . . . ... . .. . . . . .5
- 2. Radiation Monitors . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . 6
- 1. Liquid Radwaste Emuent Monitors.. . . . ... . . . . ... . . . . . .. 6
- 2. Station Blowdown Monitor.. ...... . . . . . . . . . . . .. . . . . . . 6
- 3. Reactor Containment Fan Cooler (RCFC) and Essential Service Water (ESSW) l Outlet Line Monitors.. . . . .. . . . . . . . . .. . . . . ..6 l
- 4. Turbine Building Fire and Oil Sump Monitor.. . . . . . . . . .. 6
- 5. Condensate Polisher Sump Monitor.. ... .. .6 I
- 3. Alarm and Trip Setpoints... . . . . . . . . . . . . . . . . . .. . .. 7 l
- 1. Setpoint Calculation.. .. . . .. . . . . . . . . . . . ... . . . . .7 I
- 1. Station Blowdown Monitor... . . . . . .. . . . . . . . . . . . 7
- 2. Liquid Radwaste Emuent Monitor.... . . . . . . . . . . . . . . . . . . 7
- 1. Release Tank Discharge Flow Rate. . . . . . . . . . .. 7
- 2. Release Limits . .. ... .... . . .. . . . . .... . .. . . . . .8 i
- 3. Release Mixture.. . . . . . . . . . . ..... . . . . . . .8
- 4. Liquid Dilution Flow Rates.. . . . . . . . . . . . . . .. . . . . . .9
- 5. Projected Concentrations for Release.. . .. . .......9 l
- 3. Other Liquid Emuent Monitors. . .. . . . . .. .9 l
- 4. Conversion Factors. . . . . . . .. .. . ... . . 10 l l
- 4. Allocation of Emuents from Common Release Points. . .. . . . .. .. 10 10.3 SOLIDIFICATION OF WASTE / PROCESS CONTROL PROGRAM. . . . . . 10 nworddata\odem\bw10rt ,9 doc 10-iii
BRAIDWOOD R: vision 1.9 Jrnu ry 1998 CHAPTER 10 LIST OF TABLES NUMBER PAGE 10-1 Assumed Composition of the Braidwood Station Noble Gas Effluent 10-11 10-2 Assumed Composition of the Braidwood Station Liquid Effluent 10-12 O
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- BRAIDWOOD Rsvision 1.9 January 1998 CHAPTER 10 LIST OF FIGURES NUMBER PAGE 10-1 Simplified HVAC and Gaseous Effluent Flow Diagram 10-13 10-2 Simplified Liquid Radwaste Processing Diagram 10 15 l
10-3 Simplified Liquid Effluent Flow Diagram 10 16 l
10-4 Simplified Solid Radwaste Processing Diagram 10-17 l
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, BRAIDWOOD Revision 1.9 Januiry 1998
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t CHAPTER 10 !
4 RADIOACTIVE EFFLUENT TREATMENT AND MONITORING 10.1 AIRBORNE RELEASES 10.1.1 System Description A simplified HVAC and gaseous effluent flow diagram is provided in Figure 10-1. The -
principal release points for potentially radioactive airborne effluents are the two auxiliary building vent stacks (designated Unit i Vent Stack and Unit 2 Vent Stack in Figure 10-1), in the classification scheme of Section 4.1.4, each is classified as a vent release I point (see Table A-1 of Appendix A).
l 10.1.1.1 Waste Gas Holdup System The waste gas holdup system is designed and installed to reduce radioactive gaseous effluents by collecting reactor coolant system off-gases from the reactor coolant system
{
and providing for delay or holdup to reduce the total radioactivity by radiodecay prior to i release to the environment. The system is described in Chapter 11 of the l
Byron /Braidwood UFSAR. !
10.1.1.2 Ventilation Exhaust Treatment System d 1
V) Ventilation exhaust treatment systems are designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in gaseous effluents by passing ventilation or vent exhaust gases through HEPA filters (and charcoal adsorbe.a when required to mitigate potential iodine releases) prior to release to the environment. Such a system is not considered to have any effect on noble gas effluents. The ventilation exhaust treatment systems are shown in Figure 10-1.
Engineered safety features atmospheric cleanup systems are not considered to be ventilation exhaust treatment system components.
10.1.2 Radiation Monitors 10.1.2.1 Auxiliary Building Vent Effluent Monitors Monitors 1RE-PR028 (Unit 1) and 2RE-PR028 (Unit 2) continuously monitor the final effluent from the auxiliary building vent stacks.
Both vent stack monitors feature automatic isokinetic sampling, grab sampling, and tritium sampling.
No automatic isolation or control functions are performed by these monitors. Pertinent information on these monitors is provided in Byron /Braidwood UFSAR Table 11.5-1.
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1 BRAIDWOOD R: vision 1.9 January 1998 10.1.2.2 Containment Purge Effluent Monitors l Monitors 1RE-PR001 (Unit 1) and 2RE-PR001 (Unit 2) continuously monitor the effluent I
from the Unit 1 and Unit 2 containments, respectively. When airborne radioactivity in the containment purge effluent stream exceeds a specified level, station personnel will follow established procedures to terminate the release by manually activating the containment purge valves. Additionally, the auxiliary building vent effluent monitors provide an independent, redundant means of monitoring the containment purge effluent.
No automatic isolation or control functions are performed by these monitors.
Pertinent information on these monitors is provided in Byron /Braidwood UFSAR Table 11.5-1.
Area Radiation Monitors 1(2) RE-ARO11 and 1(2) RE-ARO12 monitor the containment atmosphere. On high alarm during a containment purge, these monitors will automatically terminate the purge.
10.1.2.3 Waste Gas Decay Tank Monitors Monitors ORE-PR002A/B continuously monitor the noble gas activity released from the gas decay tanks.
On high alarm, the morsitors automatically initiate closure of the valve OGWO14 thus terminating the release.
Pertinent information on these monitors and associated control devices is provided in Byron /Braidwood UFSAR Table 11.5-1.
10.1.2.4 Gland Steam and Condenser Air Ejector Monitors Monitors 1RE-PR027 and 2RE-PR027 continuously monitor the condenser air ejector gas from Units 1 and 2, respectively. On high alarm 1(2)RE-PR027 initiates startup of the offgas treatment system.
Pertinent information on these monitors is provided in Byron /Braidwood UFSAR Table 11.5-1.
10.1.2.6 Radwaste Building Ventilation Monitor Monitor ORE-PR026 continuously monitors radioactivity in the radwaste building ventilation system. No control device is initiated by this channel.
Pertinent information on this monitor is provided in Byron /Braidwood UFSAR Table 11.5-1.
10.1.2.6 Component Cooling Water Monitor Monitor ORE-PR009 (common),1RE-PR009 (Unit 1), and 2RE-PR009 (Unit 2) continuously monitor the component cooling water heat exchanger outlets. On high alarm, ORE-PR009 initiates closure of both component cooling water surge tank (CCWST) vents,1RE-PR009 initiates closure of the Unit 1 CCWST vent, and 2RE-PROC 3 initiates closure of the Unit 2 CCWST vent.
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- BRAIDWOOD Rsvision 1.9 Jtnurry 1998 10.1.2.7 Miscellaneous Ventilation Monitors Monitor ORE-PR003 continuously monitors radioact vity in the ventilation exhaust from the laboratory fume hoods. No control device is initiated by this channel.
Pertinent information on this monitor and associated devices is provided in Byron /Braidwood UFSAR Table 11.5-1.
10.1.3 Alarm and Trip Setpoints 10.1.3.1 Setpoint Calculations 10.1.3.1.1 Auxiliary Building Vent Effluent Monitors The High Alarm setpoint for the High Range Noble Gas Channel (1/2PR028D) is established at the maximum release rate for the station as calculated in 10.1.3.2. The Alert Alarm setpoint for the High Range Gas Channelis established at a fraction of the maximum release rate for the station.
The High Alarm setpoint for the Low Range Noble Gas Channel (1/2PR0288) is established at less than or equal to 50% of the maximum release rate for the station as calct'ated in 10.1.3.2. The Alert Alarm setpoint for the Low Range Gas Channelis estabuhed at a fraction of the High Alarm setpoint for the Low Range Noble Gas Channel.
10.1.3.1.2 Containment Purge Effluent Monitors The setpoints are established at 1.50 times the analyzed containment noble gas activity during purge, plus the background reading of the monitor prior to purge.
10.1.3.1.3 Waste Gas Decay Tank Effluent Monitors The setpoints are established at 1.50 times the analyzed waste gas tank activity during release.
10.1.3.2 Release Limits Alarm and trip setpoints of gaseous effluent monitors are established to ensure that the release rate limits of RETS are not exceeded. The release limits are found by solving Equations 10-1 and 10-2 for the total allowed release rate of vent releases, Ow.
(1.11) qui (V if,} s 500 mrem /yr (10 1)
Qu E {(f) i [Li (X/Q), exp (-Ai R/3600 uy )* (10-2)
+ 1.11V,]} < 3000 mrem /yr O
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l BRAIDWOOD Rsvision 1.9 ;
January 1998 l 1
The summations are over noole gas radionuclides i.
1.11 Conversion Constant (mrem / mrad) f, Fractional Radionuclide Composition The release rate of noble gas radionuclide i divided by the total release rate of all noble gas radionuclides.
L, Beta Skin Dose Factor (mrem /yr)/( Ci/m')
Beta skin dose rate per unit of radioactivity concentration for radionuclide i.
2 Attenuation of beta radiation during passage through 7 mg/cm of dead skin is accounted for.
Q. Total Allowed Release Rate, Vent Release [ Ci/ soc)
The total allowed release rate of all noble gas radionuclides released as vent releases.
exp (-A R/3600u,) is set equal to 1.0 for setpoint calculations.
V, Gamma Whole Body Dose Factor (mradlyr)/(pCi/sec)
Gamma whole body dose rate at a specified location per unit of radioactivity release rate for radionuclide i released from a nnt. The attenuation of gamma )
radiation due to passage through 1 cm of body tissue of 1 g/cm' density is taken into account in calculating this quantity.
(X/Q), Relative Concentration Factor (sec/m')
Radioactivity concentration at a specified location per unit of radioactivity release rate for a vent release.
Equation 10-1 is based on Equation A-8 of Appendix A and the RETS restriction on whole body dose rate (500 mrem /yr) due to noble gases released in gaseous effluents (see Section A.1.3.1 of Appendix A). Equation 10-2 is based on Equation A-9 of Appendix A and the RETS restriction on skin dose rate (3000 mrem /yr) due to noble gases released in gaseous effluents (see Section A.1.3.2 of Appendix A).
Since the solution to Equation 10-2 is more conservative than the solution to Equation 10-1, the value of Equation 10-2 (7.02 x 10' pCi/sec) is used as the limiting noble gas release rate. During evolutions involving releases from the containment or waste gas decay tanks, the release rate from each release path is procedurally limited to 1 x 10' '
pCi/sec.
Calibration methods and surveillance frequency for the monitors will be conducted as specified in the RETS. J t
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I BRAIDWOOD Rsvision 1.9 Jrnuiry 1998 O 10.1.3.3 Release Mixture In the determination of alarm and trip setpoints, the radioactivity mixture in exhaust air is assumed to have the radionuclide composition of Table 10-1.
10.1.3.4 Conversion Factors The response curves used to determine the monitor count rates are based on the sensitivity to Xe-133 for conservatism.
10.1.3.5 HVAC Flow Rates The plant vent stack flow rates are obtained from 1/2 PR28J. However, if the readout indicates "0" flow, the following minimum rated fan flow values are currently used:
Unit 1 - 6.15 x 10' cc/sec Unit 2 - 4.55 x 10'cc/sec 10.1.4 Allocation of Effluents from Common Release Points Radioactive gaseous effluents released from the auxiliary building, miscellaneous ventilation systems and the gas decay tanks are comprised of contributions from both units. Consequently, allocation is made evenly between units.
10.1.5 Dose Projections for Batch Releases O Dose projections are not made prior to release. Doses are calculated after purging the containment or venting the waste gas decay tanks. Per procedure, representative samples are obtained and analyzed, and the doses calculated on a monthly basis to verify compliance with 10CFR50.
10.2 LIQUID RELEASES 10.2.1 System Description A simplified liquid effluent flow diagram is provided in Figure 10-3. A simplified liquid waste processing diagram is provided in Figure 10 2.
The liquid radwaste treatment system is designed and installed to reduce radioactive liquid effluents by collecting the liquids, providing for retention or holdup, and providing for treatment by domineralizer or a concentrator for the purpose of reducing the total radioactivity prior to release to the environment. The system is described in Chapter 11 of the Byron /Braidwood UFSAR.
10.2.1.1 Release Tanks There are two radwaste release tanks (0WX01T - 33,100 gallon capacity, and OWX26T
- 33750 gallon capacity) which receive liquid waste before discharge to the Kankakee river.
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BRAIDWOOD R: vision 1.9 Jrnurry 1998 10.2.2 Radiation Monitors 10.2.2.1 Liquid Radwaste Effluent Monitors Monitor ORE-PR001 is used to monitor all releases from the release tanks. On high alarm, the monitor automatically initiates closure of valves 0WX-353 and OWX-896 to terminate the release.
Pertinent information on the monitor arm associated control devices is provided in Byron /Braidwood UFSAR Table 11.5-2.
10.2.2.2 Station Blowdown Monitor Monitor ORE-PRO 10 continuously monitors the circulating water blowdown. No control device is initiated by this channel.
Pertinent information on this monitor is provided in Byron /Braidwood UFSAR Table 11.5-2.
10.2.2.3 Reactor Containment Fan Cooler (RCFC) and Essential Service Water (ESSW) Outlet Line Monitors Monitors 1RE-PR002,2RE-PR002,1RE-PR003, and 2RE-PR003 continuously monitor the RCFC and ESSW outlet lines.
No control device is initiated by these channels.
Pertinent information on these monitors is provided in Byron /Braidwood UFSAR Table 11.5-2.
10.2.2.4 Turbine Building Fire and Oil Sump Monitor Monitor ORE-PR005 continuously monitors the fire and oil sump discharge. On high alarm the monitor automatically initiates an interlock to trip the discharge pumps, close valve 00D030, and terminate the release.
Pertinent information on this monitor is provided in Byron /Braidwood UFSAR Table 11.5-2.
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10.2.2.5 Condensate Polisher Sump Monitor Monitor ORE-PR041 continuously monitors the condensate polisher sump discharge.
On high alarm the monitor w!;matically initiates an interlock to trip the discharge pumps and terminak 'J.o release.
l Pertinent information on this monitor is provided in Byron /Braidwood UFSAR Table 1
11.5-2.
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r BRAIDWOOD R8 vision 1.9 January 1998 10.2.3 Alarm and Trip Setpoints 10.2.3.1 Setpoint Calculations Alarm and trip setpoints of liquid effluent monitors at the principal release points are established to ensure that the limits of RETS and 10CFR20 are not exceeded in the unrestricted area.
10.2.3.1.1 Station Blowdown Monitor The monitor setpoint is found by solving equation 10-3.
P s C wc + (1.25 x C T) x [(F' x / (Few , prma)] (10-3)
P Release Setpoint [pCl/ml]
l 1.25 Factor to account for minor fluctuations in count rate.
C'" Concentration of activity in the circulating water blowdown [pC1/ml]
at the time of discharge. (" Background reading")
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C' Analyzed activity in the release tank [pCl/ml]
i F"' Circulating Water Blowdown Rate [gpm)
Fm'. Maximum Release Tank Discharge Flow Rate [gpm)
The flow rate from the radwaste discharge tbnk based on the more restrictive ;
of the maximum chemistry permitted flow rate or the Maximum Radiological I
- l. Permitted Discharge Flow Rate.
10.2.3.1.2 Liquid Radwaste Effluent Monitor l During release the setpoint is established at 1.5 times the analyzed tank activity plus the background reading, 10.2.3.1.2.1 Release Tank Discharge Flow Rate Prior to each batch release, a grab sample is obtained.
The results of the analysis of the waste sample determine the discharge rate of each batch as follows:
F'au = 0.5(F",ct/E(C',/10
- DWC)) (10 4)
The summation is over radionuclides 1.
0.5 Factor for conservatism O
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t BRAIDWOOD R: vision 1.9 Januiry 1998 l
l F',.a Maximum Radiological Permitted Discharge Flow Rate (gpm] !
The maximum permitted flow rate from the radwaste discharge tank based 1 on radiological limits (not chemistry limits which may be more restrictive)
Fi Circulating Water Blowdown Rate (gpm)
C[ Concentration of Radionuclide i in (pCl/ml]
the Release Tank The concentration of radioactivity in the radwaste discharge tank based on measurements of a sample drawn from the tank.
DWC, Denved Water Concentration [ Ci/ml]
of Radionuclidei The concentration of radionuclide i given in Appendix B, Table 2, Column 2 to 10CFR20.100120.2402.
10 Multiplier 10.2.3.1.2.2 Release Limits Release limits are detennined from RETS. Discharge rates and setpoints are adjusted to ensure that 50% of applicable RETS are not exceeded. (See Section 10.2.3.1.2.1.)
in addition to the limits identified within the RETS, an administrative action level for tritium has been established for the Braidwood cooling pond. This limit, based on drinking water pathways, has been established as a control mechanism to ensure this pathway does not become a significant contributor to public dose. Because the public has access to the Braidwood cooling pond for fishing and/or boating, an administlative limit for discharges to the cooling pond is prudent to ensure dose to the public f rom this path remains well below limits.
The controls for this pathway will be established by limiting the quantity (Curies) discharged to the Braidwood cooling pond. The administrative action level will be established at 4 Ci/ year. During times when tritium discharged to the cooling pond is in excess of the 4 C1/ year administrative action level, cooling pond tritium samples should be collected and analyzed (tritium LLD as defined in ODCM Chapter 12 Table 12.3-1) in order to assess actual tritium cooling pond tritium concentrations. Effluent pathways to the cooling pond are analyzed for tritium in accordance with ODCM Chapter 12, Table 12.3-1.
The administrative action level was chosen based on an equilibrium concentration of 200 pCi/l in the cooling pond water (1% of the public drinking water limit as specified in 40CFR141.) Information regarding calculation and assumptions can be found in Braidwood Health Physics Technical Document 98-001, " Cooling pond tritium issues".
10.2.3.1.2.3 Release Mixture For monitors ORE-PR001 and ORE-PRO 10 the release mixture used for the setpoint determination is the radionuclide mix identified in the grab sample isotopic analysis or the mix in Table 10-2.
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f f . BRAIDWOOD Rcvision 1.9 J nury 1998 10.2.3.1.2.4 Liquid Dilution Flow Rates l Dilution flow rates are obtained from circulating water blowdown transmitter loop OFT-CWO32.
10.2.3.1.2.5 Projected Concentrations for Releases After determining F;.from Equation 10-4, RETS compliance is venfied using Equations 10-5 and 10-6.
Cf '= CT[FL /(F;. + Fi,)] (10-5)
E{ Cf /10
- DWCo } S 0.5 (10-6) l The summation is over radionuclides i. ,
1 Cl Concentration of Radionuclide iin the Unrestricted Area [pCi/ml]
The calculated concentration of radionuclide iin the unrestricted area as determined by Equation 10-5.
l C[ Concentration of Radionuclide iin the Release Tank [pCi/ml]
p i
The concentration of radioactivity in the radwaste discharge tank based on measurements of a sample drawn from the tank.
DWC, Derived Water Conecatration of Radionuclide i [ Ci/ml]
of Radionuclide i The concentration of radicauclide i given in Appendix E, Table 2 Column 2 to 10CFR20.1001-20.2402.
, 10 Multiplier F;. Maximum Release Tank Discharge a Flow Rate [gpm)
Fi, Circulating Water Blowdown Rate [gpm) 0.5 Factor for conservatism 10.2.3.1.3 Other Liquid Effluent Monitors For all other liquid effluent monitors, including ORE PR001 and ORE-PR010 when not batch releasing, setpoints are determined such that the concentration limits do not
! exceed 10 times the DWC value given in Appendix B, Table 2. Column 2 to l 10CFR20.1001 - 20.2402 in the unrestricted area. Release mixtures are based on a representative isotopic mixture of the waste stream or inputs to the waste stream, or defaulted to the mix listed in Table 10-2.
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BRAIDWOOD RIvision 1.9 Januiry 1998 10.2.3.1.4 Conversion Factors The readouts for the liquid effluent monitors are in pCl/ml. The epm to Ci/ml conversion is determined for each monitor.
10.2.4 Allocation of Effluents from Common Release Points Radioactive liquid effluents released from either release tank (0WX01T or OWX26T) are comprised of contributions from both units. Under normal operating conditions, it is difficult to apportion the radioactivity between the units. Consequently, allocation is made evenly between units.
10.3 SOLIDIFICATION OF WASTE / PROCESS CONTROL PROGRAM The process control program (PCP) contains the sampling, analysis, and formulation determination by which solidification of radioactive wastes from liquid systems is ensured.
Figure 10-4 is a simplified diagram of solid radwaste processing system.
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. BRAIDWOOD R vision 1.9 JInuiry 1998 Table 10-1 Assumed Composition of the Braidwood Station Noble Gas Effluent Isotope Percent of Total Annual R6 leases l
l Ar-41 00.89 Kr-85m 00.18 Kr-85 24.90 Kr-87 00.04 Kr-88 00.28 O Xe-131m 01.40 Xe-133m 00.57 Xe 133 71.10 Xe-135 00.53 l
Xe-138 00.04 O
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BRAIDWOOD R: vision 1.9
- Jrnuary 1998 Table 10 2 Assumed Composition of the Braidwood Station Liquid Effluent Isotope Concentration isotope Concentration
( Ci/ml) (pCi/ml)
Mn-54 1.00E 05 l-132 8.00E - 07 Co-58 9.00E - 06 l-133 1.00E - 07 Fe-59 5.00E - 06 Cs-134 9.00E - 07 Co-60 3.00E - 06 l-135 4.00E - 07 Rb-86 2.00E - 06 Cs-136 9.00E - 06 Nb-95 1.00E - 05 Cs-137 2.00E - 06 Zr-95 6.00E - 06 Ce-144 1.00E - 06 Mo-99 4.00E - 06 Np-239 1.00E - 05 Ru-103 8.00E - 06 Ag-110m 3.00E - 06 q Te-127 2.00E - 05 Te-129m 2.00E - OS l-130 3.00E - 07 l-131 3.00E - 08 Te-131m 4.00E - 06 Te 132 2.00E - 06 l
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( Chapter 12 Change Summary l\ ODCM Revision 3, January 1998 Pane Channe Description l I 12-i Updated the revision number, date and file desiytator for this revision.
i
[ 12-ii Updated the revision number, j
! 12 13 Corrected reference to Annual Radioactive Effluent Release Report. This is no longer a semiannual report.
l 12-25 Corrected Decay equation. Last tenn of equation was missing a parenthesis.
- l. 12-33 Corrected title of first column from " Liquid" to "G; ..ous Release Type."
- Identified minimum analysis frequency for noble gas monitor as N.A. No l
composite noble gas sampling is performed. j 12-35 Corrected Decay equation. Section in first bracket was missing division line and several terms.
l Corrected typographical error, changing " accrue" to " accurate."
12-41 Swapped the position of"and" and "or" in the second surveillance requirement to k reflect the actual regelatory requirement.
l I 12-45 Corrected the spacing on the last line of the page.
i 12-56 Changed the LLD for tritium from 200 to 2000 pCi/1. An LLD of 2,000 pCi/l is l the LLD defined in NUREG-1301 by the NRC.
l 12-58 Added footnote 7 to describe that the vendors performing off-site analyses of environmental samples will still be required to use the 200 pCi/l LLD for tritium.
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' BRAIDWOOD Rcvision 3 Januzry 1998 O CHAPTER 12.0 SPECIAL NOTE The transfer of the Radiological Emuent Technical Specificatkas to the ODCM by Technical Specif,cs. tivi, Amendment 35, dated April 13,1992, was approved by the Nuclear Regulatory Commission.
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h:Wbrad12r.,3. doc 12-i
BRAIDWOOD Revision 3 January 1998 CHAPTER 12 ANNEXINDEX Revision 3 O
O h:bdcmtradi12r_3. doc 12-il
, BRAIDWOOD Rtvision 3 Jtnutry 1998 CHAPTER 12 RADIOACTIVE EFFLUENT TECHNICAL STANDARDS (RETS)
TABLE OF CONTENTS PAGE 12.0 RADIOLOGICAL EFFLUENT TECHNICAL STANDARDS 12-1 12.1 DEFINITIONS 12-4 12.2 INSTRUMENTATION 12-8
- 1. Radioactive Liquid Effluent Monitori..g Iristrumentation 12-8
- 2. Radioactive Gaseous Effluent Monitoring Instrumentation '12-13 12.3 LIQUID EFFLUENTS 12-20
- 1. Coi :entration 12-20
- 2. Dose 12-27
- 3. Liquid Radwaste Treatment System 12-29 12.4 GASEOUS EFFLUEN TS 12 31 O' 1. Dose Rate 12-31
- 2. Dose - Noble Gases 12-37
- 3. Dose - lodine-131 and 133, Tritium, and Radioactive Materialin Particulate Form 12-39
- 4. Gaseous Radwaste Treatment System 12-41
- 5. Total Dose 12-43
- 6. Dose Limits for Members of the Public 12-45 12.5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 12-46
- 1. Monitoring Program 12-46
- 2. Land Use Census 12 59
- 3. Intertaboratory Companson Program 12-60 12.6 REPORTING REQUIREMENTS 12-61
- 3. Offsite Dose Calculation Manual (ODCM) 12-64
- 4. Major Changes to Liquid and Gaseous Radwaste Treatment Systems 12 65 O
h:bdcm\ brad \12r_3. doc 12-lii
BRAIDWOOD R; vision 3 January 1998 CHAPTER 12 RADIOACTIVE EFFLUENT TECHNICAL STANDARDS (RETS)
LIST OF TABLES PAGE 12.0-1 Emuent Compliance Matrix 12-2 12.0-2 REMP Compliance Matrix 12-3 12.1-1 Frequency Notations 12-7 12.2-1 Radioactive liquid Emuent Monitoring Instrumentation 12-9 12.2-2 Radioactive Liquid Effluent Monitoring instrumentation Surveillance Requirements 12-11 12.2-3 Radioactive Gaseous Effluent Monitoring Instrumentation 12-14 12.2-4 Radioactive Gaseous Effluent Monitoring instrumentation Surveillance Requirements 12-17 12.3-1 Radioactive Liquid Waste Sampling and Analysis Program 12-22 12.4-1 Radioactive Gaseous Waste Sampling and Analysis Program 12-33 12.5-1 Radiological Environmental Monitoring Program 12-49 12.5-2 Reporting Level:s for Radioactivity Concentrations in Environmental Samples 12-55 12.5-3 Detection Capabilities for Environmental Sample Analysis 12 56 l
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, BRAIDWOOD Rsvision 3 January 1998 12.0 RADIOLOGICAL EFFLUENT TECHNICAL STANDARDS Chapter 12 of the Braidwood Station ODCM is a compliance of the various regulatory requirements, surveillance and bases, commitments and/or components of the radiological effluent and environmental monitoring programs for Braidwood Station. To assist in the understanding of the relationship between effluent regulations, ODCM equations, RETS (Chapter 12 section) and related Technical Specification requirements, Table 12.0-1 is a matrix which relates these various components. The Radiological Environmental Monitoring Program fundamental requirements are contained within this chapter with Braidwood specific information in Chapter 11 and with a supplemental matnx in Table 12.0-2. I I
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BRAIDWOOD R: vision 3 January 1998 I Table 12.0-1 EFFLUENT COMPLIANCE MATRIX Regulation Dose Component Limit ODCM RETS Technical Equation Specification 10 CFR 50 1. Gamma air dose and beta air dose due A-1 12.4.2 6.8.4.e.8 Appendix I to airborne radioactivity in effluent A-2 plume.
- a. Whole body and skin dose due to A-6 N/A N/A airborne radioactivity in effluent A-7 plume are reported only if certain gamma and beta air dose criteria are exceeded.
- 2. CDE for all organs and all four age A-13 12.4.3 6 8.4.e.9 groups due to iodines and particulates in effluent plume. All pathways are considered.
- 3. CDE for all organs and all four age A-29 12.3.2 6.8.4.e.4 groups due to radioactivity in liquid effluents.
10 CFR 20 1. TEDE, totaling all deep dose equivalent A-38 12.4.6 6.8.4.e.3 components (direct, ground and plume shine) and committed effective dose equivalents (all path /1, both airbome and liquid-borne). CDE evalu'ation is made for adult only using FGP 11 data base.
40 CFR 190 1. Who's body dose (DDE) due to direct A-35 12.4.5 6.8.4.e.10 (now by dose, ground and plume shine from all reference, sources at a station.
also part of l
l 10 CFR 20) 2. Organ doses (CDE) to an adult due to all A-13 l pathways.
Technical A-8
. 1. " Instantaneous" whole body (DDE), skin 12.4.1 6.8.4.e.7 Specifications (SDE), and organ (CDE) dose rates to A-9 an adult due to radioactivity in airborne A-28
. effluents. For the organ dose, only l
inhalation is considered.
- 2. " Instantaneous" concentration limits for A-32 12.3.1 6.8.4.e.2 fiquid effluents.
Technical NA 12.6.2 6.9.1.7 Specifications 1. Radiological Effluent Release Report htdemvarad12r 3. doc 12-2
7
,_ BRAIDWOOD RIvision 3 JInuiry 1998 I
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Table 12.0-2 i
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REMP COMPLIANCE MATRIX 1
Regulation Dose Component Limit RETS Technical Specification 10 CFR 50 Implement environmental monitonng 12.5.1 6.8.4.f Appendix i program.
Section IV.B.2
~ Technical Land Use Census 12.5.2 6.8.4.f.2 -
Specifications Technical interlaboratory Companson Program 12.5.3- 6.8.4 f.3 Specifications Technical Radiological Environmental Operating Report 12.6.1 6.9.1.6 Specifications O
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12-3
l CRAIDWOOD Ruvision 3 January 1998 12.0 RADIOLOGICAL EFFLUENT TECHNICAL STANDARDS 12.1 DEFINITIONS 12.1.1 AC'Jon shall be that which presenbes remedial measures required under designated conditions. l 12.1.2 Analoa Channel Ooerational Test shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shallinclude adjustments, as necessary, of the alarm interlock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.
12.1.3 Channel Calibration shall be the adjustment, as necessary, of the channel such that it responds within the required range anc. accuracy to known values of input. The CHANNEL CAllBRATION shall encompass the entire channelincluding the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
12.1.4 Channel Check shall be the qualitative assessment of channel behavior during operation by observation. This determination shallinclude, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
12.1.5 Diaital Channel Ooerational Test shall consist of exercising the digital computer hardware using data base manipulation and injecting simulated process data to verify OPERABILITY of alarm and/or trip functions.
12.1.6 Dose Eauivalent I-131 shall be that connection of I 131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132,1-133,1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table lli of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites".
12.1.7 Frecuenev - Table 12.1-1 provides the definitions of various frequencies for which surveillance's, sampling, etc. are performed unless defined otherwise. The 25%
variance shall not be applied to Operability Action Statements. The bases to Technical Specification 4.0.2 provide clanfications to this requirement.
- 12.1.8 Member (s) of the Public means any individual except when thet individualis receiving an occupational dose.
12.1.9 QCcuoational Dose means the dose received by an individualin the course of employment in which the individual's assigned duties involve exposure to radiation and/or radioactive material from licensed and unlicensed sources of radiation, whether in the possession of the licensee or other person. Occupational dose does not include dose from background radiation, as a patient from medical practices, from voluntary participation in medical research programs, or as a member of the public.
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, BRAIDWOOD Rsvision 3 Janurry 1998 12.0 RADIOLOGICAL EFFLUENT TECHNICAL STANDARDS (Contl 12.1.10 Ooerable/Ocerability a system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).
12.1.11 Ooerational Mode (i.e. Mode) shall correspond to any one inclusive combination of core
)
reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2 of the Technical Specifications.
12.1.12 Process Control Proaram (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20,61,71 and State regulations, burial ground requirements, and other requirements governing the disposal of radioactive wastes.
12.1.13 Purce/Puraina shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to punfy the confinement.
p 12.1.14 Rated Thermal Power shall be a total core heat transfer rate to the reactor coolant of c,d' 3411 MWt.
12.1.15 Site Boundarv shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. l 12.1.16 Solidification shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.
l 12.1.17 Source Check shall be the qualitative assessment of channel response when the l channel sensor is exposed to a source of increased radioactivity. l I
12.1.18 Thermal Power shall be the total core heat transfer rate to the reactor coolant.
12.1.19 Ullmstricted Area means an area, access to which is neither limited nor controlled by the licensee.
12.1.20 Ventilation Exhaust Treatment System shall be any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Features Atmospheric Cleanup Systems are not ccasidered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
h3odemibraid\127.3 doc 12-5
BRAIDWOOD R vision 3 January 1998 12.0 RADIOLOr;JCAL EFFLUENT TECHNICAL STANDARDS (Cont.)
12.1.21 Venting shall be any controlled process of discharging air or gas from a confinement to maintain temperature. pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
12.1.22 Waste Gas Holduo System shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off-gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
12.1.23 Definitions Peculiar to Estimating Dose to Members of the Public using tb ODCM Computer Program.
- a. ACTUAL - ACTUAL refers to using known release data to project the dose to members of the public for the previous time period. This data is stored in the database and used to demonstrate compliance with the reporting requirements of Chapter 12.
- b. PROJECTED - PROJECTED refers to using known release data from the previous time period or estimated release data to forecast a future dose to members of the public. This data is not incorporated into the database.
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, BRAIDWOOD Rsvision 3 Janu:ry 1998 O TABLE 12.1 1 1 REQUENCY NOTATIONS
- NOTATION FREQUENCY S - Shiftly At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D - Daily At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W - Weekly At least once per 7 days.
M - Monthly At least once per 31 days.
Q - Quarterly At least once per 92 days.
SA - Semiannually At least once per 184 days.
A- Annually At least once per 366 days.
R - Refueling cycle At least once per 18 months (550 days).
S/U - Startup Prior to each reactor startup.
P - Prior Prlor to each radioactive release.
N.A. Not applicable.
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Each frequency requirement shall be performed within the specified time interval with the maximum allowable extension not to exceed 25% of the frequency interval. The 25% variance shall not be applied to Operability Action Statements. The bases to Technicai Specification 4.0.2 provide clarifications to this requirement. These frequency notations do not apply to the Radiological Environmental Monitoring Program (REMP) as desenbed in Section 12.5.
O h:V4cmerad\12r_3. doc 12-7
ERAIDWOOD R; vision 3 January 1998 12.2 INSTRUMENTATION 12 2.1 Radioactive Liquid Effluent Monitoring Instrumentation Ooerability Reauirements 12.2.1.A The radioactive liquid effluent monitoring instrumentation channels shown in Table 12.2-1 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of 12.3.1.A are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.
Anoticabilitv: At all times Action
- 1. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.
- 2. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 12.2-1. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in l the next Radioactive Effluent Release Report pursuant to Section 12.6 why this inoperability was not corrected within the time specified.
Surveillance Reauirements 12.2.1.B Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION and DIGITAL and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 12.2-2.
Bases 12.2.1.C The radioactive liquid effluent instrumentation is provided to monitor and control, l as applicable, the releases of radioactive materials in liquid effluents during actual l or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of RETS. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60,63, and 64 of Appendix A to 10 CFR Part 50.
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h:bdcm\ brad 12r 3 doc 12-8
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. BRAIDWOOD Ravision 3 January 1998 l TABLE 12.21 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l MINIMUM l
CHANNELS INSTRUMENT OPERABLE ACTION
- 1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release
- a. Liquid Radwaste Effluent Line (ORE-PR001) 1 31
- 2. Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release i a. Essential Service Water l
l b. Station Blowdown Line (ORE PRO 10) 1 32 l l
- 3. Flow Rate Measurement Devices l
l a. Liquid Radwaste Effluent Line (Loop-WX001) 1 33 i b. Liquid Radwaste Effluent Low Flow Line 1
(Loop-WX630) 1 33
- c. Station Blowdown Line (Loop-CWO32) 1 33 l
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h:bdcmtrad\12r 3. doc 12-9
BRAIDWOOD R; vision 3 January 1998 TABLE 12.2-1 (Continued)
RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION ACTION STATEMENTS ACTION 31 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided that prior to initiating a release:
- a. At least two independent samples are analyzed in accordance with Section 12.3 and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 32 - With the number of channels OPER/ GLE less than required by the Minimum Channels OPERABLE requiremen'., effluent releases via this pathway may continue for up to 30 days provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for principal gamma emitters and 1-131 at a lower limit of detection as specified in Table 12.3-1.
ACTION 33 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves generated in place may be used to l
estimate flow.
l ACTION 34 - With the number of channels OPERABLE lese An required by the Minimum l
Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for principal gamma emitters and 1131 at a lower limit of detection as specified in Table 12.31:
- a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT l 131, or
- b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT l-131.
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ERAIDWOOD R; vision 3 January 1998 TABLE 12.2-2 (Continued)
RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATIONS l
I i
(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or
- b. Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor loss of power), or
- c. Detector check source test failure, or
- d. Detector channel out-of-service, or
- e. Monitor loss of sample flow. This is only applicable for ORE-PR001 and ORE-PR005.
Monitor ORE-PR041 will not trip on loss of sample flow.
(2) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument Indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor loss of power), or
- c. Detector check source test failure, or
- d. Detector channel out-of-service, or l
- e. Monitor loss of sample flow.
(3) The initial CHANNEL CAllBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that f have been related to the initial calibration shall be used.
l (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release.
CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.
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O h:bdcmtragf\12r_3. doc l
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, BRAIDWOOD Rcvision 3 l
January 1998 l 1
12.2.2 Radioactive Gaseous Effluent Monitonna Instrumentation L Ooerability Reauirements 1
12.2.2.A The radioactive gaseous effluent monitoring instrumentation channels shown in Table 12.2-3 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure !
that the limits of Section 12.4 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the )
I methodology and parameters in the ODCM.
]
l Acolicability: As shown in Table 12.2-3 l
6C.tiOD:
- 1. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above section, immediately suspMJ the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.
l
- 2. With less than the minimum number of radioactive gaseous effluent '
monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 12.2-3. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in tne next Annual Radioactive Effluent Release Report pursuant to Section 12.6 why this inoperability was not corrected within the time specified, p Surveillance Reauirements J
V 12.2.2.8 l Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION and DIGITAL and CHANNEL OPERATIONAL TEST at the frequencies shown in Table 12.2-4.
Bases 12.2.2.C The racioact' lous effluent instrumentation is provided to monitor and control, as appli.,able, the releases of radioactive materials in gaseoas effluents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the f methodology and parameters in the ODCM to ensure that the alarm / trip will occur l prior to exceeding the limits of RETS. The OPERABILITY and use of this j
instrumentation is consistent with the requirements of General Design Criteria 60, ;
63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas i activity monitor used to show compliance with the gaseous effluent release requirements of Section 12.4 shall be such that cu%6r,tranns as low as 1x104 uCi/cc are measurable. I i
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BRAIDWOOD R; vision 3 January 1998 TABLE 12.2-3 (Continued)
RADIOACTIVE GAS,50US EFFLUENT MONITORING INSTRUMENTATION TABLE NOTATIONS
- At all times.
ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to initiating the release:
- a. At least two independent samples of the tank's contents are analyzed, and
- b. At least two technically qualified members of the facility staffindependently verify the release rate calculations and dischargo valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway. Releases may continue via this pathway for up to 7 days provided real time monitoring of radioactive effluents released via this pathway is established.
ACTION 38 - Not used.
ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for principle gamma emitters at an LLD as specified in Table 12.41.
ACTION 40 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 12.4-1.
ACTION 41 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, liquid grab samples are collected and analyzed for radioactivity at a lower limit of detection as specified in Table 12.3-1.
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- BRAIDWOOD R::visien 3 l
January 1998 i
TABLE 12.2-4 (Continued) g.
k_, RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION l SURVEILLANCE REQUIREMENTS TABLE NOTATIONS At al. times.
(1) The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or
- b. Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or monitor loss of power), or l
- c. Detector check source test failure, or l
- d. Detector channel out-of-service, or l
- e. Monitor loss of sample flow. Monitoring ORE-PR002A and 2B will not trip on loss of sample flow. This is only applicable for functional unit 6, ORE-PR009 and RE-009.
(2) The DIGITAL CHANNEL OPERATIONAL TEST shat? also demonstrate that control room alarm
,q annunciation occurs if any of the following conditions exists:
ih a. Instrument indicates measured levels above the Alarm Setpoint, or I
- b. Circuit failure (monitor loss of communications - alarm only, detector loss of counts, or i monitor loss of power), or l
l
- c. Detector check source test failure, or t
- d. Detector channel out-of-service, or i
- e. Monitor loss of sample flow.
t (3) The initial CHANNEL CAllBRATION shall be performed using one or more of the reference j standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range ,
of energy and measurement range. For subsequent CHANNEL CAllBRATION, sources that have been related to the initial calibration shall be used.
I I
}nl G
h%demtrad\12r_3 doc 12-19
BRAIDWOOD Revision 3
nuary 1998 ,
l 12.3 LIQUID EFFLUENTS 12.3.1 Concentration Ooerability Reauirements 12.3.1.A The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Braidwood Station ODCM Annex, Appendix F, Figure F-1) shall be limited to 10 times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2x10" microcurie /mi total activity.
Apolicability: At all times Action
- 1. With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately i restore the concentration to within the above limita.
l Surveillance Reauirements l
l 12.3.1.1.B Radioactive liquid wastes shall be sampled and analyzed according to the l sampling and analysis program of Table 12.3-1, 12.3.1.2.B The results of the radioactivity analysis shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of 12.3.1.A.
I Bases j 12.3.1.C This section is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than 10 times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section ll.A design objectives of Accendix 1.10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.1301.
This section applies to the release of radioactive materials in liquid effluents from all units at the site.
O h:bdcmtrad\12r_,3. doc 12 20
,. BRAIDWOOD Rtvision 3 Jtnutry 1998 12.3 LIQUID EFFLUENTS (Continued) l j
[
Bases The required detection capabilities for radioactive materiais in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD. and other detection limits can be found in HASL Procedures Manual. HASL-300 (revised annually), Cume, L.A., Limits for Qualitative Detection and Quantitative Determination - Application to -
Radiochemistry," Anal. Chem. 40.586-93 (1968), and Hartwell, J.K.," Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH SA-215 (June 1975).
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CRAIDWOOD R;vi: ion 3 January 1998 TABLE 12.31 RADIOACTIVE LlOUID WASTE SAMPLING AND ANA. LYSIS PROGRAM LIQUlG RELEASE SAMPLING MINIMUM ANALYSIS TYPE OF ACTIVITY LOWER LIMIT OF TYPE FREQUENCY FREQUENCY ANALYSIS DETECTION (LLD)"8 (pCVml)
- 1. Batch Release P P Principal Gamma 5x10
Tanks
- C.3 Batch Each Batch Emitters
- l-131 1x10*
P M Dissolved and Entrained 1x10
One Batch /M Gases (Gamma Emitters) m l P M H-3 1x10
Each Batch Composite
- Gross Alpha ix10
St-89, Sr-90 4 P Q 5x10 Each Batch Composite
- 4 Fe-55 1x10
- 2. Continuous W al Gamma Princip#
5x10
Releases "' Composite
Eme Continuous
l-131 1x10 4 O
- a. Circulating Water M M Dissolved and Entrained 1x10
Blowdown Grab Sample Gases (Gamma Emitters)*
- b. Waste Water M H-3 1x10.s
" Continuous
- Composite'8 D em Circulating Water Discharge Gross Aloha 1x10
- c. Condensate Continuous
- Q St-89, Sr-90 5x10
Polisher Sump Composite
- Discharge 4
Fe-55 1x10 O
l h:bdcmtraid\12r.,3 doc 12-22
, BRAIDWOOD Ravision 3 Janu:ry 1998 TABLE 12.3-1 (Continued)
RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LIQUID RELEASE SAMPLING MINIMUM ANALYSIS TYPE OF ACTIVITY LOWER LIMIT TYPE FREQUENCY FREQUENCY ANALYSIS OF DETECTION (LLD)")(pCi/ml) 3.Continuots W') W') Principal Gamma 5x10
Release ( Grab Sample Emitters
- Essential Service Water Reactor Containment Fan Cooler (RCFC) Outlet Line I131 4 1x10 H3 1x10 4
M ) Dissolved and 1x10 4 Entrained Gases (Gamma Emitters)*
O 4. Continuous Surge Tank None None Principal Gamma Emitters
- 5x10
1 Vent-Component Cooling Water Line (*)
Dissolved and 4 1x10 Entrained Gases (Gamma Emitters)*
l-131 4 1x10 l
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h:bdcmtraid\12r_3. doc 12-23 i
BRAIDWOOD R:; vision 3 January 1998 TABLE 12 3-1 (Continued)
RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS (1) The LLD is defined, for purposes of these sections, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separations:
LLD = 4.86s3 E
- V
- 2,22 x 10' Y exp (-AAt)
Where:
LLD = the lower limit of detection (microCuries per unit mass or volume),
s, = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),
E = the counting efficiency (counts per disintegration),
V = the sample size (units of mass or volume),
2.22 x 10' = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, 1 A = the radioactive decay constant for the particular racionuclide (sec "), and at = the elapsed time between the midpoint of sample collection and the time of counting (sec).
Typical values of E, V, Y, and At should be used in the calculation.
Alternative LLD Methodoloav An alternative methodology for LLD determination follows and is similar to the above LLD equation:
(2.71 + 4.65VB)* Decay LLD =
E q t) Y t(2.22E06)
O h:bdcmtrad\12r_3. doc 12 24 l
CRAIDWOOD R vision 3 I
TABLE 12.3-1 (Continued) fQ RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS Where:
B = background sum (counts)
E = counting efficiency, (counts detected / disintegration's) q z sample quantity, (mass or volume) b = abundance,(if applicable)
Y = fractional radiochemical yield or collection efficiency, (if applicable) t = count time (minutes) 2.22E06 = number of disintegration's per minute per microcurie 2.71 + 4.65VB = k2+ (2k V 2 V B), and k = 1.645.
(k=value of the t statistic from the single-tailed t distribution at a significance level of 0.95 and infinite degrees of freedom. This means that the LLD result represents a 95% detection probability with a 5% probability of falsely concluding that the nuclide present when it is t,ot or that the nuclide is not present when it is.)
Decay = e* [ ART /(1-e#)]lAT,/(1-e#d)], (if applicable)
A = radioactive decay constant, (units consistent with At, RT and T.)
At = " delta t", or the elapsed time betweOn sample collection or the midpoint of sample collection and the time the count is started, depending on the type of sample, (units consistent with A)
RT= elapsed real time, or the duration of the sample count, (units consistent with A)
To = sample deposition time , or the duration of analyte collection onto the sarnple media, (units consistent with A)
The LLD may be determined using installed radicanalytical software, if available. In addition to determining the correct number of channels over which to total the background sum, utilizing the software's ability to perform decay corrections (i.e. during sample collection, from sample collection to start of analysis and during counting), this alternate method will result in a more accurate determination of the LLD.
It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement.
(O h%demerad\12r_3. doc 12-25
CRAIDWOOD RLvision 3 January 1998 TABLE 12 3-1 (Continued)
RADIOACTIVE LIQUlO WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS (2) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.
(3) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.
(4) A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.
(5) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously whenever the effluent stream is flowing. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
(6) Not required unless the Essential Service Water RCFC Outlet Radiation Monitors RE-PR002 and 4
RE-PR003 indicates measured levels greater than 1x10 Ci/mi above background at any time during the week.
(7) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for dissolved and entrained for principal gamma emitters. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report 1, June 1974.
(8) A continuous release is the discharge of dissolved and entrained gaseous waste form a nondiscrete liquid volume.
O h%demtrad\12r.,3 doc
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12-26
, CRAIDWOOD Rtvision 3 JInurry 1998 12.3.2 Rose l (O / Ooerability Reauirements 12.3.2.A The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see l Braidwood Station ODCM Annex, Appendix F, Figure F-1) shall be limited:
- 1. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 mrems to any organ, and
- 2. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.
Anolicabilitv: At all times.
Action
- 1. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the t Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special l Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
Surveillance Reauirements i I i il 12.3.2.B Cumulative dose contributions from liquid effluents for the current calendar quarter and !
the current calendar year shall be determined in accordance with the methodology and l parameters in the ODCM at least once per 31 days.
Bases 12.3.2.C This section is provided to implement the requirements of Sections ll.A, Ill.A and IV.A of !
Appendix 1,10 CFR Part 50. The Operability Requirements implement the guides set forth in Section ll.A of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix l to assure that the releases of radioactive material in liquid effluents to i UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dose calculation methodology and parameters in the ODCM implement the requirements in Section ill.A of Appendix 1 that conformance with the guides of Appendix l be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.
The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109," Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents For the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I" Revision 1, October 1977 and Regulatory Guide 1.113, Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of implementing Appendix I," April 1977.
fO hMdem\ brad \12r_3 doc 12-27
l BRAIDWOOD RLvision 3 January 1998 12.3.2 QQsg (Continued)
Bases This section applies to the release of radioactive materials in liquid effluents from each reactor at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be asenbed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to Operability Requirements, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.
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- BRAIDWOOD R; vision 3 1 January 1998 12.3.3 Liauid Radwaste Treatment System l 0 Q Ooerability Reauirements 12.3.3.A The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Braidwood Station ODCM Annex, Appendix F Figure F-1)would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period.
Acolicability: At all times.
Action
- 1. With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System not in I operation, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:
- a. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- b. Action (s) taken to restore the inoperable equipment to OPERABLE j
status, and 1
( c. Summary description of action (s) taken to prevent a recurrence.
Surveillance Reauirements 12.3.3.1.8 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when the Liquid Radwaste Treatment System is not being fully utilized.
12.3.3.2.8 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Sections 12.3.1.A and 12.3.2.A.
Bases 12.3.3.C The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropnate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable". This section implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in SectWi ll.D of Appendix 1 to 10 CFR Part 50.
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l h bdcmtraid\12r_3 doc 12-29
BPAIDWOOD R. vision 3 January 1998 12.3.3 Liouid Radwaste Treatment Svstem (Continued)
Bases The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section ll.A of Appendix 1,10 CFR Part 50, for liquid effluents.
This section applies to the release of radioactive ma'.eriais in liquid effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be rnade of the contnbutions from each unit based on input conditions, e g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be ailocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to Operability Requirements, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.
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9 hMdem\ braid \12r_3 doc 12-30
. BRAIDWOOD R; vision 3 J"nu;ry 1998 12.4 GASEOUS EFFLUENTS !
f^s l b 12.4.1 Dose Rate l
Ooerability Reauirements l
l 12.4.1. A The dose rate due to radioactive materials released in gaseous effluents from the site l to areas at and beyond the SITE BOUNDARY (see Braidwood Station ODCM Annex, i Appendix F, Figure F-1) shall be limited to the following:
- 1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the whole body and less than or equal to a dose rate of 3000 mrems/yr to the skin, and
- 2. For lodine 131 and 133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.
Acolicabilitv: At all times.
Action
- 1. With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s). {
Surveillance Reautrements i 1
\
- 'j 12.4.1.1.B The dose rate due to noble gases in gaseous effluents shall be determined to be within l
the above limits in accordance with the methodology and parameters in the ODCM. !
12.4.1.2.B The dose rate due to lodine 131 and 133, tntium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance witli the methodology and parameters in the I ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 12.4-1.
i Bases
)
12.4.1.C This section is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual I dose limits of 10CFR20. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC exceeding the limits specified in 10CFR20.1301.
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Y h:bdcmtrad12r.3. doc 12-31
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BRAIDWOOD Rr, vision 3 January 1998 ,
12.4 GASEOUS EFFLUENTS (Continued)
Bases For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, ;
the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to i compensate for any increase in the atmospheric diffusion factor above that for the (
SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, i with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 mrems/ year to the skin. These release rate limits also restrict, at all times the corresponding thyroid dose rate above background via the inhalation pathway to less than or equal to 1500 mrems/ year.
This section applies to the release of radioactive materials in gaseous effluents from all units at the site.
The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40. 586-93 (1968), and Hartwell, J.K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
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hMdemtradi12r_3 doc 12 32
, BRAIDWOOD Rtvision 3 TABLE 12.41 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM GASEOUS SAMPLING MINIMUM TYPE OF ACTIVITY LOWER LIMIT OF RELEASE TYPE FREQUENCY ANALYSIS ANALYSIS DETECTION l
FREQUENCY (LLD)")(pCi/cc)
- 1. Waste Gas P P Principal Gamma 1x10" Decay Tank Each Tank Each Tank Emitters (2)
- 2. Principal Gamma 4 Containment P P 1x10 t
Purge Each Purge ') Each Purge (*) Emitters (2)
Grab Sample H-3 1x10
- 3. Auxiliary Bldg. M "") M Principal Gamma 1x10 4
Vent Stack Grab Sample Emitters (2)
(Unit 1 and 2)
H-3 1x10
Continuous (*) W ") 1-131 1 x10
Charcoal Sample iO 1-133 1 x10
Continuous (') W ") Principal Gamma 1 x10'"
Particulate Emitters <23 Sample Continuous (*) Q Gross Alpha 1 x10'"
Composite Particulate Sample Continuous (') Q Sr-89, Sr-90 1 x10'"
Composite Particulate Sample Continuous N.A. Noble Gases, Gross 4 1x10 Beta or Gamma Phonitor l
O h:bdcmtrad12r 3. doc 12-33
BRAIDWOOD R:: vision 3 January 1998 TABLE 12.4-1 (Contingeg1}
FADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS (1) The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separation:
LLD = 4.66S 5 V
- 2.22 x 10'
- Y = exp (-AAt)
Where:
LLD = the lower limit of detection (microCuries per unit mass or volume),
s, a the standard deviat. ion of the background counting rate or of the counting rate of a blank s. ample as appropriate (ccunts per minuts),
E = the counting efficiency (counts per disintegration),
V = the sample size (units of mass or volume),
2.22 x 10' = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, l A= the radioactive decay constant for the particular radionuclide (sec "), and l At = the elapsed time between the midpoint of sample collection and the time of counting (sec).
i l Typical values of E, V, Y, and At should be used in the calculation.
l Altemate LLD Methodology An alternate methodology for LLD determination follows and is similar to the above LLD equation:
(2.71 + 4.65VB) Decay LLD =
E q b Y t (2.22E06)
O h.bdcmtradu2r.3 doc 12-34
GRAIDWOOD R vision 3 Januiry 1998 TABLE 12.4-1 (Continued)
IMDIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS Where:
l B = background sum (counts) '
E = counting efficiency, (counts detected / disintegrations) q = sample quantity, (mass or volume) i b = abundance,(if applicable)
Y = fractional radiochemical yield or collection efficiency, (if applicable) t = count time (minutes) 2.22E06 = number of disintegrations per minute per microcurie (2.71 + 4.65VB) = k2 + (2k V 2 V B), and k = 1.645.
f (k=value of the t statistic from the single-tailed t distnbution at a significance level i
of 0.95% and infinite degrees of freedom. This means that the LLD result l
'v represents a 95% detection probability with a 5% probability of falsely concluding that the nuclide present when it is not or that the nuclide is not present when it is.)
( Decay = e* [ ART /(1-e")) [AT /(1-e'"d)], (if applicable) l l A = radioactive decay constant, (units consistent with At, RT and T )
l At = " delta t*, or the elapsed time between sample collection or the m;dpoint of sample collection and the time the count is started, depending on the type of sample, (units consistent with A)
RT = elapsed real time, or the duration of the sample count, (units consistent with A)
T = sample deposition time, or the duration of analyte collection onto the sample media, (unit consistent with 1)
I The LLD may be determined using installed radioanalytical software, if available. In addition to determining the correct number of channels over which to total the background sum, utilizing the software's ability to perform decay corrections (i.e. during sample collection, from sample collection to start of analysis and during counting), this altemate method will result in a more accurate determination of the LLD.
It should be recognized that the LLD is defined as a before the fact limit and not as an after the fact limit for a particular measurement.
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hnodemtrad12r_3 doc 12-35
BRAIDWOOD R:; vision 3 January 1998 O
TABLE 12.4-1 (Continued)
RACIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS (2) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99,1-131, Cs-134. Cs-137, Ce-141, and Ce-144 in particulate releases. This list does not mean that only these nuclides are to be considered.
Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to Section 12.6.2, in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.
(3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour penod.
(4) Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.
(5) Tritium grab samples shall be taken at least once per 7 days from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.
(6) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time oeriod covered by each dose or dose rate calculation made in accordance with Sections 12.4.1.A, 12.4.2.A and 12.4.3.A.
(7) Samples shall be changed at least once per 7 days and analyses shall be completed within a timeframe necessary to meet the applicable lower limits of detection, but not to exceed 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br />.
Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period and analyses shall be completed within a timeframe necessary to meet the applicable lower limits of detection, but not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if: (1) analysis shows that the DOSE EQUlVALENT l-131 concentration in the reactor coolant has not increased more than a factor of 3, and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
O h:bdcmibrad\12r_3 coc 12-36
BRAIDWOOD RIvision 3 Januiry 1998 12.4.2 Dose - Noble Gases Ooerability Reautrements 12.4.2.A The air dose due to noble gases released in gaseous effluents, from each unit, to areas I at and beyond the SITE BOUNDARY (see Braidwood Station ODCM Annex, Appendix F, Figure F-1) shall be limited to the following: 3 1
- 1. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and {
less than or equal to 10 mrads for beta radiation, and )
l
- 2. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
Acolicability: At all times.
Action
- 1. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
Surveillance Reauirements l 12.4.2.B Cumulative dose contributions for the curre' t calendar quarter and the current calendar
% year for noble gases shall be determined in accordance with the methodology and i parameters in the ODCM at least once per 3! days. )
Bases 12.4.2.C This section is provided to implement the requirements of Sections ll.B Ill.A and IV.A of Appendix 1,10 CFR Part 50. The Operability Requirements implement the guides set forth in Section ll.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix ! to assure that the releases of radioactive material in gaseous effluents to area at or beyond the SITE BOUNDARY will be kept "as low as is reasonable achievable." The Surveillance Requirements implement the requirements in Section Ill.A of Appendix l that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.
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h%dem\ brad \12r.,3 doc 12-37
BRAIDWOOD Rrvision 3 January 1998 12.4.2 Dose - Noble Gases (Continued)
Bases The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive materials in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109,
" Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents For the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I" Revision 1, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors, Revision 1," July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.
This section applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed br shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each un t based on input conditions,
( e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated j effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to
(
Operability Requirements, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attnbuted to each unit to obtain the total releases per unit.
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BRAIDWOOD Rsvision 3
, January 1998 l
12.4.3 Dose - todine I 131 and 133. Tritium. and Radioactive Material in Particulate Form j
Ooerability Reauirements l
l 12.4.3.A The dose to a MEMBER OF THE PUBLIC from lodine 131 and 133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY (see Braidwood Station ODCM Annex, Appendix F, Figure F-1) shall be limited to the following:
- 1. During any calendar quarter: Less than or equal to 7.5 mrems to any organ, and
- 2. During any calendar year: Less than or equal to 15 mrems to any organ.
Anolicabilitv: At all times.
Action l 1. With the calculated dose from the release of lodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce !
the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
I Surveillance Reauirements O 12.4.3.B Cumulative dose contributions for the current calendar quarter and the current calendar year for lodine-131 and 133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at i least once per 31 days.
I Bases l 12.4.3.C This section is provided to implement the requirements of Sections ll.C, Ill.A and IV.A of Appendix I 1,10 CFR Part 50. The Operability Requirements are the guides set forth in Section ll.C of l )
Appendix 1. The ACTION statements provide the required operating flexibility and at the same l time implement the guides set forth in Section IV.A of Appendix i to assure that the releases of l radioactive material in gaseous effluents to areas at or beyond the SITE BOUNDARY will be kept "as low as is reasonable achievable." The ODCM calculational methods specified in the l Surveillance Requirements implement the requirements in Section Ill.A of Appendix 1 that conformance with the guides of Appendix 1 be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through I appropriate pathways is unlikely to be substantially underestimated.
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hMdcm\ brad \12r ,3 doc 12 39
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. 1 BRAIDWOOD Revision 3 January 1998 12.4.3 Ogie(Continued)
Bases The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109," Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents For the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I" Revision 1. October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1 July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions The release rate specifications for lodine-131 and 133, tritium, and radionuclides in p articulate form with half-lives greater than 8 days are dependent upon the existing radior iclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathy,3ys that were examined in the development of these calculations were: ( y ir.a:Vidual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animal's graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure to man.
This section applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to l Operability Requirements, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attnbuted to each unit to obtain the total f releases per unit.
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, CRAIDWOOD Rtvision 3 Jinu:ry 1998 12.4.4 Ganeous Radwaste Treatment Sy1!Bm O Ooerability Reautrements 12.4.4. A The VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous rIfluent releases, from each unit, to areas at and beyond the SITE BOUNDARY &ce Braidwood Station ODCM Annex, Appendix F, Figure F-1) would exceed:
- 1. 0.2 mrad to air from gamma radiation, or
- 2. 0.4 mrad to air from beta radiation, or
- 3. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC. l 1
Anolit-shilitv: At all times.
l Action
- 1. With radioactive gaseous waste being discharged without treatment and in i excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Technical 3pecification 6.9.2, a Special Report that includes the following information:
p a. Identification of any inoperable equipment or subsystems, and the reason Q for the inoperability,
- b. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- c. Summary description of action (s) taken to prevent a recurrence.
Surveillance Reauirements 12.4.4.1.8 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized.
12.4.4.2.B The installed VENTILATION EXHAUST TREATMENT SYSTEM and WASTE GAS HOLDUP SYSTEM shall be considered OPERABLE by meeting Section 12.4.1 or 12.4.2 and 12.4.3.
Bases ,
12.4.4.C The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment.
O h%dctnerad\12r,3 doc 12-41 1
BRAIDWOOD Revision 3 January 1998 12.4.4 Gaseous Radwaste Treatment System (Continued)
Ba3es The requirement that the appropriate portions of this system be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This section implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section ll.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Gaseous Radwaste Treatment System were specified as a 2% fraction of the dose design objectives set forth in Section ll.B and ll.C of Appendix 1,10 CFR Part 50, for gaseous effluents.
This section applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contnbutions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not praCUcable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to Operability Requirements, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.
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hWem\ brad \12r_3 doc 12-42
, BRAIDWOOD Revision 3 Janusry 1998 12.4.5 Total Dose p
V Ooerability Reauirements 12.4.5.A The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the ,
thyroid, which shall be limited to less than or equal to 75 mrems.
{
Applicab. titty At all times.
Action
{
- 1. With the calculated doses from the release of radioactive materials in liquid or l gaseous effluents exceeding twice the limits of Sections 12.3.2,12.4.2, or 12.4.3, l calculations should be made including direct radiation contributions from the units I and from outside storage tanks to determine whether the above limits of Section 12.4.5.A have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Technical SpecificaJ,n 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and dire.::t radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and (3) concentration of radioactive materialinvolved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the s'~'
release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shallinclude a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granteti 1til staff action on the request is i complete. '
Surveillance Reauirements 12.4.5.1.A Cumulative dose contnbutions from liquid end gaseous effluents shall be determined in accordance with Sections 12.3.2,12.4.2, and 12.4.3, and in accordance with the methodology and parameters in the ODCM.
12.4.5.2.8 Cumulative dose contnbutions from direct radiation from the units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in ACTION 1 of Section 12.4.5.A.
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E BRAIDWOOD Revision 3 January 1998 12.4.5 Iotal Dose (Continued)
Bases 12.4.5.C This section is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The section requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix 1, and if direct radiation doses from the reactor units and j outside storage tanks are kept small. The Special Report will describe a course of I action that should result in the limitation of the annual dose to a MEMBER OF THE l PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose
! contributions from other nuclear fuel cycle facilities at the same site or within a radius i
of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 l
CFR 190.11 and 10 CFR 20.2203, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance l
only relates to the limits of 40 CFR Part 190, and does not apply in any way to the '
other requirements for dose limitation of 10 CFR Part 20, as addressed in Sections 12.3.1 and 12.4.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
O h:bdcmtrad\12r_3 doc 12-44
, liRAIDWOOD Rgvision 3 January 1998 12.4.6 Dose Limits for Members of the Public Ooerability Reauirements 12.4.6. A The licensee shall conduct operations such that the TEDE to individual MEMBERS OF THE PUBLIC does not exceed 100 mrem in a year. In addition, the dose in any unrestricted area from external sources does not exceed 2 mrem in any one hour. The Effluents Program shall implement monitoring, sampling and analysis of radioactive liquid and gaseous etfluents in accordance with 10CFR20.1302 and with the methodology and parameters in the ODCM.
Anolicability: At all times.
d&l!QD'
- 1. If the calculated dose from the release or exposure of radiation meets or exceeds j the 100 mrem / year limit for the MEMBER OF THE PUBLIC, prepare and submit l a report the Commission in accordance with 10CFR20.2203.
- 2. If the dose in any unrestricted area from external sources of radiation meets or exceeds the in any one hour limit for the MEMBER OF THE PUBLIC, prepare and submit a report to the Commission in accordance with 10CFR20.2203.
Surveillance Reauirements i !
12.4.6.B Calculate the TEDE to individual MEMBERS OF THE PUBLIC annually to determine l compliance with the 100 mrem / year limit in accordance with the ODCM. In addition, evaluate and/or determine if direct radiation exposures exceed 2 mrem in any hour O Bases in unrestricted areas.
12.4.6.C This section applies to direct exposure of radioactive materials as well as radioactive materials released in gaseous and liquid effluents.10CFR20.1301 sets forth the 100 mrem / year dose limit to members of the public; 2 mrem in any one hour limit in the unrestricted area; and reiterates that the licensee is also required to meet the i 40CFR190 standards.10CFR20.1302 provides options to determine compliance to 10CFP20.1301. Compliance to the above operability requirement is based on 10CFR20,40CFR190 and Braidwood Station Technical Specifications.
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h:bdomeraidu2r_3. doc 12-45
CRAIDWOOD R; vision 3 January 1998 12.5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 12.5.1 Monitorina Procram Ooerabilitv Reauirements 12.5.1. A The Radiological Environmental Monitoring Program shall be conducted as specified in Table 12.51.
Aeolicabiktv: At all times.
Action
- 1. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 12.5-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Section 12.6.1, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
Deviations are permitted from the required samphng schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of sampling equipment, if a person / business who participates in this program goes out of business or no longer can provide sample, or contractor omission which is corrected as soon as discovered. If the equipment malfunctions, corrective actions shall be completed as soon as practical. If a person / business supplying samples goes out of business, a replacement supplier shall be found as soon as possible. All deviations from the sampling schedule will be described in the Annual Radiological Environmental Operating Report.
j
- 2. With the level of radioactivity as the result of plant effluents in an environmental i sampling medium at a specified location exceeding the reporting levels of Table 12.5-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Specia!
Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose
- to a MEMBER OF THE PUBLIC is less than the calendar year limits of Section 12.3.2,12.4.2, or 12.4.3. When more than one of the radionuclides in Table 12.5.2 are detected in the sampling medium, this report shall be submitted I
if:
1 concentration (1) . concentration (2) + 21.0 reporting level (1) reporting level (2)
When radionuclides other than those in Table 12.5-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
- to A MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than l the calendar year limits of Section 12.3.2,12.4.2, or 12.4.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Section 12.6.1.
'The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
h%demtrad\12r.,3 doc 12-46
I l . BRAIDWOOD R; vision 3 J nu;ry 1998 l 12.5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (Continued) i 1 (3 V 3. If the sample type or sampling location (s) as required by Table 12.5-1 become(s) permanently unavailable, identify suitable alternative sampling media for the pathway of interest and/or specific sampling locations for obtaining replacement samples and add them to the Radiological Environmental Monitoring Program as ;
soon as practicable. The specific locations from which samples were unavailable may then be deleted from the monitoring program. i l
Prepare and submit a controlled version of the ODCM within 180 days including a revised figure (s) and table reflecting the new location (s) with supporting i information identifying the cause of the unavailability of samples and justifying the selection of new location (s) for obtaining samples.
Surveillance Reauirements 12.5.1.B The radiological environmental monitoring program samples shall be collected pursuant to Table 12.5-1 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 12.5-1 and the detection capabilities required by Table 12.5-3.
Bases 12.5.1.C The Radiological Environmental Monitoring Program required by this section provides representative nieasurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential e- radiation exposures of MEMBERS OF THE PUBLIC resulting from the station (s
)
oporation. This monitoring program implementsSection IV.B.2 of Appendix ! to 10 i CFR Part 50 and thereby supplements the radiological effluent rr.onitoring program by i venfying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and {
I the modeling of the environmental exposure pathways. Guidance for this monitoring I program is provided by the Radiological Assessment Branch Technical Position on I Environmental Monitonng. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this penod, program l changes may be initiated based on operational experience. l The required detection capabilities for environmental sample analyses are tabulated l in terms of the lower limits of detection (LLDs). The LLDs required by Table 12.5-3 l are considered optimum for routine environmental measurements in industrial I laboratories. It should be recognized that the LLD is defined as a before the fact limit l representing the capability of a measurement system and not as an after the fact limit l for a particular measurement.
Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, LA., " Limits fe. Qualitative Detection and Quantitative Determination - Application to Radiochemistry," AQak Chem 40. 586-93 (1968), and Hartwell, J.K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
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Modem \ brad \12rJdoc 12-47
BRAIDWOOD Rsvision 3 January 1998 12.5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (Continued)
Interoretations 12.5.1. D Table 12.5-1 requires "one sample of each community drinking water supply downstream of the plant within 10 kilometers." Drinking water supply is defined as water taKen from rivers, lakes, or reservoirs (not well water) which is used for drinking.
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BRAIDWOOD Revi: ion 3 Jrnu ry 19'98 TABLE 12.5-1 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE NOTATIONS (1) Specific parameters of distance and direction from the centerline of the midpoint of the two units and additional description where pertinent, shall be provided for each and every sample location in Table 12.5-1 of the ODCM Station Annexes. Refer to NUREG-0133,
" Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,"
October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979.
(2) Far field samples are analyzed when the respective near field sample results are inconsistent with previous measurements and radioactivity is confirmed as having its origin in airbome effluents from the station, or at the discretion of the Radiation Protection Director.
(3) Airbome particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
(4) Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attnbutable to the effluents from the station.
(5) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. Film badges shall not be used as dosimeters for measuring duect radiation.
The 40 locations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographicallimitations; e.g., If a station is ad.iacent to a lake, some sectors may be over water thereby reducing the number of dosimeters which could be placed at the indicated distances. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.
(6) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.
(7) The " downstream" sample shall be taken in an area beyond but near the mixing zone.
The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. Upstream samples in an estuary must be taken far enough upstream to be beyond the station influence.
(8) If milking animals are not found in the designated indicator locations, or if the owners decline to participate in the REMP, all milk sampling may be discontinued.
(9) Biweekly refers to every two weeks.
(10) 1-131 analysis means the analytical separation and counting procedure are specific for this radionuclide.
(11) One sample shall consist of a volume / weight of sample large enough to fill contractor specified container, i
htdemibrad12r_3 doc 12-54
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, .. CRAIDWOOD Revision 3 Janutry 1998 TABLE 12.5 3 (Continued) l Oi DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS TABLE NOTATIONS t
(1) The nuclides on this list are not the only nuclides intended to be considered. Other peaks that are identifiable.
l together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological
! Environmental Operating Report.
(2) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.
l (3) The Lower Limit of Detection (LLD) is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will l be detected with 95% probability with only 5% probability of falsely conduding that a blank observation represents a "real" signal.
For a particular measurement systb,n. which may include radiochemical separation, the LLD is defined as follows:
4.66 S, + 3/t, LLD =
(E) (V) (2.22) (Y) (exp (-Mt))
4.66 S.
LLD -
(E)(V) (2.22) (Y) (exp (-Mt))
O l 1
Where: 4.66 So n 3/t, j i
LLD = the "a priori" Minimum Detectable Concentration (picoCuries per unit mass or volume), l so = the standard deviation of the background counting rate or of the counting rate of a blank sample, as I appropriate (counts per minute), !
, 4 Total Counts is E = the counting efficiency (counts per disintegration), i V = the sample size (units of mass or volume),
2.22 = the number of disintegratione per minute per picocurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec"),
h:bdcmtrann12r.3. doc 12 57
BRAIDWOOD Rrvision 3 January 1998 TABLE 12.5-3 (Continued)
DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS TABLE NOTATIONS to = counting time of the background or blank (minutes), and at = the elapsed time between sample collection, or end of the sample collection period, and the time of counting (sec).
Typical values of E, V, Y, and at should be used in the calculation.
It should be recognized that the LLD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for a particular measurement.
Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.
! Occasionally, background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and desenbed in the Annual Radiological Environmental Operating Report.
(4)lf no drinking water pathway exists, the value of 15 pCill may be used.
(5) A value of 0.5 pCill shall be used when the animals are on pasture (May through October) and a value of 5 pCl/l shall be used at all other times (November through April).
(6) This LLD applies orly when the analytical separation and counting procedure are specific for this radionuclide l (7) This LLD is the minimum allowable, however, vendors performing environmental sample analyses off-site will be required to meet an LLD of 200 pCi/l.
I l
h%demibrad12rJdoc l
l 12-58
, .. .' CRAIDWOOD Rivision 3 Janurry 1998 12.5.2 Land Use census Qgggbilitv Raouirements 12.5.2.A. A Land Use Census shall be conducted and shall identify within a distance of 10 km (6.2 miles) the location in each of the 16 meteorological sectors
- of the nearest milk animal, the nearest residence", and an enumeration of livestock. For dose calculation, a garden will be assumed at the nearest residence.
Apphcabihty' At all times.
Action:
- 1. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment, via the same exposure pathway 20% greater than at a location from which samples are currently being obtained in accordance with Section 12.5.1, add the new location (s) within 30 days to the Radiological Environmental Monitoring Program given in Chapter 11. The sampling location (s), excluding the controllocation, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Submit in the next Annual Radiological Environmental Operating Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new location (s) with information supporting the change in sampling locations.
- This requirement may be reduced according to geographicallimitations; e.g. at a lake site l
where some sector's will be over water. I "The nearest industrial facility shall also be documented if closer than the nearest residence.
1 Surveillance Reauirements
{
}
'12.5.2.8 The Land Use Census shall be conducted during the growing season, between June 1 and October 1, at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report.
p....
12.5.2.C This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by the results of this census.
This census satisfies the requirements of Section IV.B.3 of Appendix ! to 10 CFR Part 50. An annual garden census will not be required since the licensee will assume that there is a garden at the nearest residence in each sector for dose calculations.
l O l h%demtrad12r.,3 doc l
12 59
1 e
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CRAIDWOOD R: vision 3 January 1998 12.5.3 Intertaboratorv Comoarison Program Ooerability Reauirements 12.5.3.A Analyses shall be performed on radioactive materials supplied as part of an interlaboratory Comparison Program that correspond to samples required by Table 12.5-1.
Aeolicabilitv: At all times.
Action 1.With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.
Surveillance Reauirements 12.5.3.B A summary of the results obtained as part of the above required interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.
Bases 12.5.3.C The requirement for particirsation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements .
of radioactive material in environmental samples matrices are performed cs part of the quality Ossurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix 1 to 10 CFR Part 50.
O h:bdcmtrad\12r 3 doc 12-60
i
, .* IRAIDWOOD Rsvision 3 i Janurry 1998 12.6 REPORTING REQUIREMENTS
/
3 b
12.6.1 Annual Radiological Envirorgental Oceratina Reoort*
Routine Annual Radiological Environmental Operating Report covering the operation of the Unit (s) during the previous calendar year shall be submitted prior to May 1 of each year.
The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.
i The Annual Radiological Environmental Operating Report shall include the results of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the tables and figures in l Chapter 11 of the ODCM, as well ac summanzed and tabulated results of these analyses l and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following: a summary description of the Radiological fm Environmental Monitoring Program; legible maps covering all sampling locations keyed to
('~,) a table giving distances and directions from the midpoint between the two units; reasons for not conducting the Radiological Environmental Monitoring Program as required by l Section 12.5.1, a Table of Missed Samples and a Table of Sample Anomalies for all deviations from the sampling schedule of Table 11.1-1; discussion of environmental sample measurements that exceed the reporting levels of Table 12.5-2 but are not the result of plant effluents, discussion of all analyses in which the LLD required by Table 12.5-3 was not achievable; result of the Land Use Census required by Section 12.5.2; and the results of the licensee participation in an Interlaboratory Comparison Program and the corrective actions being taken if the specified program is not being performed as required j by Section 12.5.3.
'A single submittal may be made for a multiple unit station.
A V) i Modemtraid\12r_3. doc 1241
e r,
GRAIDWOOD RQvision 3 January 1998 12.6 REPORTING REQUIREMENTS (Cont'd) 12.6.1 &lnual Radioloalgal Environmental Ooeratina Reoort (Cont'd)
The Annual Radiological Environmental Operating Report shall also include an annual summary of hourly meteorological data collected over the applicable year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. In lieu of submission with the Annual Radiological Environmental Operating Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
The Annual Radiological Environmental Operating Report shall also include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the Unit or Station during the previous calendar year. This report shall also include an assessment of the radiation doses to the most likely exposed MEMBER OF THE PUBLIC from reactor releases and other near-by uranium fuel cycle sources including doses from primary effluent pathways and direct radiation, for the previous calendar year. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the ODCM, and in compliance with 10CFR20 and 40 CFR Part 190," Environmental Radiation Protection Standards for Nuclear Power Operation."
O l
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1 O h:bdontrad12r.3 doc l
12-62 l
l BRAIDWOOD Rsvision 3 JInuIry 1998 12.6 REPORTING REQUIREMENTS (Continued) 12.6.2 Annual Radioactive Effluent Release Reoort" Routine Annual Radioactive Effluent Release Reports covering the operation of the unit during the previous calendar year operation shall be submitted prior to May 1 of the following year.
The Annual Radioactive Effluent Release Reports shallinclude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, " Measuring Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.
For solid wastes, the format for Table 3 in Appendix B shall be supplemented with )
three additional categodes: class of solid wastes (as defined by 10 CFR Part 61), type of j container (e.g., LSA, Type A Type B, Large Quantity), and SOLIDIFICATION agent or '
absorbent (e.g., cement, urea formaldehyde).
The Annual Radioactive Effluent Release Reports shallinclude a list and description of unplanned releases from the site to areas beyond the site boundary of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Annual Radioactive Effluent Release Reports shall include any changes made during !
the reporting period to the PCP, as well as any major changes to Liquid, Gaseous or Solid )
Radweste Treatment Systems, pursuant to Section 12.6.3.
The Annual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the tim'; s.mcified in Section 12.2.1 or 12.2.2, l respectively; and description of the events leading to ilquid holdup tanks or gas storage tanks exceeding the limits of Technical Specification 3.11.1.4 or 3.11.2.6, respectively.
I "A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
O hMdemtraidi12r_3. doc 12-63
e-
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CRAIDWOOD R: vision 3 January 1998 12.6 REPORTING REQUIREMENTS (Continued) 12.6.3 Offsite Dose Calculation Manual (ODCM) 12.6.3.1 The ODCM shall be approved by the Commission prior to initialimplementation.
12.6.3.2 Licensee-initiated changes to the ODCM:
- a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.2. This documentation shall contain:
- 1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes (s); and
- 2. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20,160,40 CFR Part 190,10 CFR 50.36a, and Appendix 1 to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- b. Shall become effective after review and acceptance by the Onsite Review and Investigative Function and the approval of the Plant Manager on the date specified by the Onsite Review and investigative Function.
- c. Shall be subrnitted to the Commission in the form of a complete legible copy of the entire ODCM or updated pages if the Commission retains a controlled copy.
If an entire copy of the ODCM is submitted, it shall be submitted as part of or '
concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM wes made effective. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shallindicate the date (e.g., month / year) the change was implemented.
O h:bdcmtraid\12r_3 doc 12-64
O
,g BRAIDWOOD RIvision 3 Janu:ry 1998 12.6 REPORTING REQUIREMENTS (Continued) 12.6.4 Maior Chanaes to Liauid and Gaseous Radwaste Treatment Systems
- Licensee-initiated major changes to the Radwaste Treatment Systems (liquid and gaseous):
- a. Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for l the period in which the evaluation was reviewed by the Onsite Review and Investigative i Function. The discussion of each change shall contain:
- 1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;
- 2) Sufficient detailed information to totally support the reason for the change without benefit of additionaland supplementalinformation;
- 3) A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems.
l
- 4) An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto; i
- 5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC and to the general population that differ from those previously estimated in the License application and amendments thereto;
- 6) A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;
- 7) An estimate of the exposure to plant operating personnel as a result of the change; and
- 8) Documentation of the fact that the change was reviewed and found acceptable by the Onsite Review and investigative Function.
- b. Shall becume effective upon review and acceptance by the Onsite Review and investigative Function.
- Licensees may choose to submit the information called for in this section as part of the annual FSAR update.
b d
h:bdcmibrad\17r.,3. doc 12-65