ML20137B922
| ML20137B922 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 02/28/1997 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| PROC-970228, NUDOCS 9703240102 | |
| Download: ML20137B922 (106) | |
Text
... --
X244 ALL i
Document Control Desk Director of Nuclear Reactor Regulation
)
i U.S. Nuclear Regulatory Commission February 27,1997 Mail station PI-137 Washington,DC 20555 j
bi Attached is a revision to the Offsite Dose Calculation Manual, Byron Annex, Chapters 10 through
- 12. Please update your manual as follows:
l Remove:
Byron Chapter 10, Revision 1.2 Byron Chapter i1, Revision 1.3 Byron Chapter 12, Revision 1.3 Insert:
Byron Chapter 10, Revision 1.3 i
Byron Chapter 11, R vision 1.4 Byron Chapter 12, Revision 1.4 i
1 Please sign below indicating your manual has been updated and that your controlled copy number is correct.
i Name Date i
Return to:
h Comed
[ Ng Central Files 1400 Opus Place,4th Floor Downers Grove,IL 60515
- or -
Central Files 4th Floor p(j Dosmers Grove 21004.0 9703240102 970228 PDR ADOCK 05000454 gggg P
PDR g,
b
i l
Byron ODCM O
Chapter 10 J
Revision 1.3 Change Summary Page 10-i Removed page revision index.
No longer are individual pages revised.
The revision number for the index is assigned to the chapter.
j Page 10-1i Updated page numbers.
Page 10-lii Updated page numbers.
Corrected system des 4.gnator for Essential Service Water.
Page 10-v Updated page numbers.
Page 10-1 Removed specific UFSAR Chapter numbers due to a UFSAR revision in_ progress that affects Chapter i
- 11. 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(2)RE-PR028 monitors.
Added j
that the 28's provide noble gas monitoring.
Changed reference to UFSAR from "Section" to
" Chapter" to be consistent with the wording used j
\\
in the UFSAR.
i Page 10-2 Removed specific UFSAR_ Chapter numbers due to a UFSAR revision in progress that affects Chapter 11.
Updated the high alarm function for monitors 1 (2 ) RE-PR027.
i Page 10-3 Removed specific UFSAR Chapter numbers due to a UFSAR revision in progress that affects Chapter 11.
Updated the high alarm function for monitor ORE-PR026. Removed statement concerning a procedural limit on the release rate from section 10.1.3.1.2; this is covered in Section 10.1.3.2.
Changed the 1/4% in section 10.1.3.1.1 to 0.25%
to make reading the value easier.
Page 10-4 Removed statement concerning a procedural limit on the release rate from section 10.1.3.1.3; this is l
covered in Section 10.1.3.2.
Page 10-5 Removed from 10.1.3.2 the specific release rate j
limit indicated in station procedures.
This limit is not specifically used in all procedures; i
however, the limit used still ensures the station maximum release rate is not exceeded when releasing from multiple sources.
Removed default O
l l
=
fan flows listed in section 10.1.3.5.
Fan flows are estimated from operating fan combinations; standard defaults are not normally used.
Page 10-6 Removed specific UFSAR Chapter numbers due to a UFSAR revision in progress that affects Chapter 11.
Changed reference to UFSAR from "Section" to
" Chapter" to be consistent with the wording used in the UFSAR.
Corrected system designator for Essential Service Water.
Page 10-7 Removed specific UFSAR Chapter numbers due to a UFSAR revision in progress that affects Chapter 11.
Changed reference to UFSAR from "Section" to
" Chapter" to be consistent with the wording used in the UFSAR.
Clarified wording in section 10.2.3.1 to give a generic overview of the purpose of setpoint calculations.
Pages 10-7,8,9 Regrouped information as it applies to specific liquid monitors.
Added clarifying information.
Page 10-9 Revised wording for determining liquid dilution flow rates to agree with Admin Tech Requirements.
Page 10-10 Clarified information under "other monitors" to include release monitors during non-release times and updated section to include the current process to determine isotopic mixtures for setpoint calculations.
Removed specific reference to Cs-137 for the use in determining conversion factors. The source used is dependent on the monitor.
Deleted reference to Figure 10-4.
Page 10-11 Corrected the percent for Kr-87. Originally typed as 00.4 instead of 0.04.
1 Page 10-14 Inserted dotted lines indicating an occasional flowpath from the gland steam condensers due to a modification being conducted to the system.
Page 10-15 Inserted dotted lines indicating an occasional flowpath through the evaporators.
Corrected spelling error of " evaporators".
Page 10-17 Deleted Figure 10-4, Simplified Solid Radwaste Processing Diagram.
1 0
1 BYRON Revision 1.3 g
February 1997 BYRON ANNEX INDEX a
i 1
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CHAPTER 10 i
REVISION 1.3 l
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10-1
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BYRON Rivision 1.3 Ftbrutry 1997 CHAPTER 10.
4 m
RADIOACTIVE EFFLUENT TREATMENT AND MONITOP.ING TABLE OFCONTENTS f
NUMBER PAGE
.i 10 1 AIRBORNE RELEASES.
10-1 1.
System Description...
10 1 4
1.
Waste Gas Holdup System...
10-1 2.
Ventilation Exhaust Treatment System..
10-1
,I 2.
Radiation Monitors..
10-1 1.
Auxiliary Building Vent Effluent Monitors..
10-1 2.
Containment Purge Effluent Monitors..
10-2 i
3.
Waste Gas Decay Tank Monitors...
10-2 4.
Gland Steam and Condenser Air Ejector Monitors........
10-2 i
5.
Radwaste Building Ventilation Monitor..........
10-3 6.
Component Cooling Water Monitor..........
10-3 4
7.
Misce!!aneous Ventilation Monitors..
10-3 i
3.
Alarm and Trip Setpoints.
. 10-3 1.
Setpoint Calculations.......
10-3 1.
Auxiliary Building Vent Emuent Monitors.....
10-3 l
2.
Containment Purge Effluent Monitors....
10-3 j
3.
Waste Gas Decay Tank Emuent Monitors..
10-4 2.
Release Limits.
10-4 3.
Release Mixture....................
10-5 i
4.
Conversion Factors........
10-5 5.
HVAC Dilution Flow Rates....
10-5 4.
Allocation of Effluents from Common Release Points.........
10-5 5.
Dose Projections for Batch Releases.................
10-5 i
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t 1
4 4
10-ii
BYRON Rzvision 1.3 February 1997 CHAPTER 10 RADIOACTIVE EFFLUENT TREATMENT AND MONITORING TABLE OF CONTENTS (Cont'd)
InLMSf.8 PAGE 10.2 LIQUID RELEASES.
10-5 1.
System Description..
10-5 1.
Release Tanks..
10-6 2.
Turbine Building Fire and Oil Sump..
10-6 3.
Condensate Polisher Sump.
10-6 2.
Radiation Monitors..
10-6 1.
Liouid Radweste Effluent Monitors..
10-6 2.
Station Biowdown Monitor..
10-6 3.
Reactor Containment Fan Cooler (RCFC) and Essential Service Water (SX) Outlet Line Monitors..
10-6 4.
Turbine Building Fire and Oil Sump Monitor.
10-7 5.
Condensate Polisher Sump Monitor...
.10-7 3.
Alarm and Trip Setpoints.
. 10-7 1.
Setpoint Calculations.....
. 10-7 1.
Station Blowdown Monitor..
10-7 1.
Release Mixture...
.10-8 2.
Liquid Radwaste Effluent Monitor...
10-8 1.
Release Tank Discharge Flow Rate 10-8 2.
Release Mixture 10-9 3.
Liquid Dilution Flow Rates.
10-9 4.
Projected Concentrations for Releases 10-9 3.
Other Liquid Effluent Monitors 10-10 4.
Conversion Factors......
10-10 4.
Allocation c: Effluents from Common Release Points..
10-10 10.3 SOLIDIFICATION OF WASTE / PROCESS CONTROL PROGRAM..
10-10 0
10-iii
- - ~ _ _ _ _ _... _ _ _ _.. _ _. _..
BYRON Revision 1.3 Februrry 1997 CHAPTER 10 LIST OF TABLES NUMBER PAGE 10-1 Assumed Composition of the Byron Station Noble Gas Emuent 10-11 10-2 Assumed Composition of the Byron Station Liquid Emuent 10-12 i
J i
l i
)
1 10-iv
BYRON R*. vision 1.3 7'997 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 10 3 Liquid Release Flowpath 10-16
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O 10-v
i BYRON Rsvision 1.3 i
Febru ry 1997 CHAPTER 10
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RADIOACTIVE EFFLUENT TREATMENT AND MONITORING 10 1 AIRBORNE RELEASES 10 1.1
System Description
i l
A simplified HVAC and gaseous effluent flow diagram is provided in Figure 10-1. The j
principal release points for potentially radioactive airbome effluents are the two auxiliary l
buil ding vent stacks (designated Vent Stack 1 and Vent Stack 2 in Figure 10-1). In the g
classification scheme of Section 4.1.4, each is classified as a vent release point (see 1
L Table A-1 of Appendix A).
10 1.1.1 Waste Gas Holdup System The waste gas holdup system is designed and installed to reduce radioactive gaseous i
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
{
release to the environment. The system is described in Chapter 11 of the Byron /Braidwood UFSAR.
i 10 1.1.2 Ventilation Exhaust Treatment System l
l 5
Ventilation exhaust treatment systems are designed and installed to reduce gaseous j
radioiodine or radioactive material in particulate form in gaseous effluents by passing i
ventilation or vent exhaust gases througn charcoal adsorbers and/or HEPA filters 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 atmosphenc cleanup systems are not considered to be
- ventilation exhaust treatment system cc,irpc,r,6iits.
10 1.2 Radiation Monitors 10.1.2.1 Auxiliary Building Vent Effluent Monitors Monitors 1 RE-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, noble gas monitoring, grab sampling, iodine and particulate sampling and tritium sampling.
No automatic isolation or control functions are performed by these monitors. Pertinent informabon on these monitors is provided in Byron /Braidwood UFSAR Chapter 11.
I i
l O
10-1 l
l
]
BYRON RIvision 1.3 Fcbruary 1997 10 1.2.2 Containment Purge Effluent Monitors Monitors 1RE-PR001 (Unit 1) and 2RE-PR001 (Unit 2) continuously monitor the effluent from the Unit 1 and Unit 2 containments, respectively. When airbome 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 Chapter
{
11.
Monitors 1(2)RE-AR011 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 monitoru automatically initiate closure of the valve OGW104 thus terminating the release.
Pertinent information on these monitors and associated control devices is provided in Byron /Braidwood UFSAR Chapter 11.
10.1.2.4 Gland Steam and Condenser Air Ejector Monitors Monitors 1RE-PR027 and 2RE-PR027 continuously monitor the condenser air ejector gas l
from Units 1 and 2, respectively.
Pertinent information on these monitors is provided in Byron /Braidwood UFSAR Chapter l
11.
i i
O 10-2
-..... ~.... -. -.... -..
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BYRON Rsvision 1.3 l
Ftbruary 1997 10 1.2.5 Radweste Building Ventilation Monitor 1
Monitor ORE-PR026 continuously monitors radioactivity in the radweste building ventilation system. On high alarm, ORE-PR026 initiates isolation of the radwaste building ventilation system.
i
\\
Pertinent information on this monitor is provided in Byron /Braidwood UFSAR Chapter 11.
1 10 1.2.6 Component Cooling Water Monitor Monitors 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 Mnk (CCWST) vents, 1RE-PR009 initiates closure of the Unit 1 CCWST vent, and 2RE-PR009 initiates closure i
of the Unit 2 CCWST vent.
i Pertinent information on this' monitor is provided in Byron /Braidwood UFSAR Chapter 11.
10 1.2.7 Miscellaneous Ventilation Monitors Monitor ORE-PR003 continuously' monitors radioactivity in the ventilation exhaust from the laboratory fume hoods. No control device is initiated by this channel.
l Pertinent information on this monitor and associated devices is provided in l
1 Byron /Braidwood UFSAR Chapter 11.
j 10 1.3 Alarm and Trip Setpoints 10.1.3.1 Setpoint Calculations 10.1.3.1.1 Auxiliary Building Vent Effluent Monitors i
The setpoints for the low range noble gas channel are conservatively established at 2.5%
of the maximum permissible release rate for the high alarm and 0.25% of the maximum i
release rate for the alert alarm.
1 The setpoints for the high range noble gas channel are conservatively established at 50%
4 j
of the maximum permissible release rate for the high alarm and 5% of the maximum release rate for the Cert alarm.
10.1.3.1.2 Containment Purge Emuent Monitors The setpoints are established at 1.25 times the containment noble gas activity during purge.
10-3
BYRON RIvision 1.3 F bruary 1997 I
10 1.3.1.3 Waste Gas Decay Tank Effluent Monitors The setpoints are established at 1.25 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 Section 12.4 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, Cw.
(1.11)Qel(VjF } s 500 mrem /yr (10-1) i Qel((f ) [L (X/Q)yexp(-X,gR/3600 yv)
(10-2) i i
+ 1.11 V )}
3000 mrem /yr i
The summations are over noble gas radionuclides i.
f.
Fractional Radionuclide Composition The release rate of noble gas radionuclide i divided by the total release rate of all noble gas radionuclides.
Q.
Total Allowed Release Rate, Vent Release
[ C!/sec]
The total allowed release rate of all noble gas radionuclides released as vent releases.
The remaining parameters in Equation 10-1 have the same definitions as in Equation A-8 of Appendix A. The remaining parameters in Equation 10-2 have the same definition as in Equation A-9 of Appendix A.
Equation 10-1 is based on Equation A-8 of Appendix A and the RETS restriction on whole 4
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 und 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).
O 10-4
4 BYRON Rsvision 1.3 Fr$bruary 1997
]
Since the solution to Equation 10-2 is more conservative than the solution to Equation 10-j 1, the value of Equation 10-2 (1.02 x 10 pCi/sec) is used as the limiting noble gas
[
7 t
release rate. During evolutions involving releases from the containment or waste gas
_i
[
decay tanks, the total station release rate is procedurally limited such that the maximum i
permissible release rate is not exceeded.
i i
10 1.3.3 Release Mixture G
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.
l 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 consesvatism.
f 10 1.3.5 HVAC Dilution Flow Rates I
The plant vent stack flow rates are obtained from the RM-11 console in the control room.
?
If the values cannot be obtained from RM-11, flow rates can be estimated from the l
operating fan combinations.
s 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 contribubons from both i
units. Consequently, allocabon is made evenly between units.
i l
10 1.5 Dose PisjecG0iis for Batch Releases 4
)
The 10CFR20 dose limits have been converted into a stabon administrative release rate limit using the methodology in the ODCM. Compliance is venfled prior to each release.
5 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 i
calculated on a monthly basis to verify compliance with 10CFR50.'
10 2 LIQUID RELEASES J
i 10 2.1
System Description
A simplified liquid release flowpath diagram is provided in Figure 10-3. A simplified liquid i
radwaste processing diagram is provided in Figure 10-2.
. The liquid radweste treatment system is designed and installed to reduce radioachve liquid effluents by collecting the liquids, providing for retention or holdup, and providing i
l 3
- O 10-5 a
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.e,
i BYRON Rrsion 1.3 fib'ruaiy 1997 for treatment by demineralizer 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 Updated Final Safety Analysis Report.
i 10 2.1.1 Release Tanks f
There are two radwaste release tanks (OWXOIT and OWX26T 30.000-gallon capacity each) which receive liquid waste before discharge to the Rock river.
10 2.1.2 Turbine Building Fire and Oil Sump l
The turbine building fire and oil sump receives water from selected turbine building sumps, the tendon tunnel sumps, and the diesel fuel oil storage sumps, all of which are normally non-radioactive but potentially contaminated. The effluent from this sump is i
monitored, and if radioactive contamination exceeds a predetermined level pump operation is automatically terminated. The water may then be sent to the liquid radwaste l
treatment system.
10.2.1.3 Condensate Polisher Sump The condensate polisher sump receives waste water from the condensate polisher system which is normally non-radioactive but potentially conwaminated. The effluent from this sump is monitored and if radioactive contamination exceeds a predetermined level sump discharge is terminated and major condensate polisher inputs to the sump are automatically isolated. The water may then be sent to the liquid radwaste treatment system.
10.2.2 Radiation Monitors J
10 2.2.1 Liquid Radwaste Effluent Monitors Monitor ORE-PR001 is used to monitor all releases from the release tanks. On high 1
alarm, the monitor automatically initiates closure of valves OWX-353 and OWX-869 to terminate the release.
l Pertinent information on the monitor and associated control devices is provided in Byron /Braidwood UFSAR Chapter 11.
l j
10 2.2.2 Station Blowdown Monitor Monitor ORE-PRO 10 continuously monitors the recirculating water blowdown. No control device is initiated by this channel.
Pertinent information on this monitor is provided in Byron /Braidwood UFSAR Chapter 11.
I 10 2.2.3 Reactor Containment Fan Cooler (RCFC) and Essential Service Water (SX) Outlet Line l
Monitors
' Monitors 1RE-PR002,2RE-PR002,1RE-PR003, and 2RE-PR003 continuously monitor
~ the RCFC and SX outlet lines.
I J
O 10-6
-. ~.
BYRON R3 vision 1.3 FsbruIry 1997 No control device is initiated by these channels.
Pertinent information on these monitors is provided in Byron /Braidwood UFSAR Chapter l.
11.
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 Chapter 11.
g 10 2.2.5 Condensate Polisher Sump Monitor Monitor ORE-PR041 continuously monitors the condensate polisher sump discharge. On high alarm the monitor automatically inibates an interlock to trip the discharge pumps and terminate the release. Pertinent informabon on this monitor is provided in Byron /Braidwood UFSAR Chapter 11.
l 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 are not exceeded in tne unrestncted area.
Setpoint calculations normally consist of identified release mixtures, dilution factors, conversion factors (detector sensitivity), and conservatism factors i
f%
10 2.3.1.1 Station Blowdown Monitor I
During release, the monitor setpoint is found by solving equation 10-3.
P s C" + (1. 50 x C) x (f,/ (f + Y,) )
(10-3)
P Release Setpoint (pCi/ml]
L50 Factor to account for minor fluctuations in count rate 8
Concentration of activity in the circulating water blowdown (pCi/ml]
at the time of discharge
(" Background reading")
C Analyzed activity in the release tank
[ Cl/ml) excluding tritium O
10-7 w
BYRON R2 vision 1.3 F(bruary 1997
)
N Circulating Water Blowdown Rate
[gpm]
F',,,,,
Maximum Release Tank Discharge Flow Rate
[gpm]
The flow rate from the radwaste discharge tank.
10 2.3.1.1.1 Release Mixture The release mixture used for the setpoint determination is the radionuclide mix identified in the release tank grab sample isotopic analysis.
10 2,3.1.2 Liquid Radwaste Effluent Monitor During release the setpoint is established at 1.5 times the analyzed tank activity plus the background reading.
However, per procedure, the maximum discharge flow rate is limited to a value that will result in less than 50% of 10*DWC at the discharge point. (See Section 10.2.3.1.2.1) 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:
L = 0.5(FLa / E(C, /10
- DWCi))
zto.4y The summation is over radionuclides i.
5 Maximum Permitted Discharge Flow Rate The maximum permitted flow rate from the radwaste discharge tank based on radiological limits (not chemistry limits which may be more restrictive)
[gpm]
N='
Circulating Water Blowdown Rate
[gpm]
C, Concentration of Radionuclide iin the Release Tank
[ Ci/ml]
The concentration of radioactivity in the radwaste discharge tank based on measurements of a sample drawn from the tank.
0 10-8
BYRON Revision 1.3 February 1997 l
- DWC, Derived Water Concentration
[ Ci/ml]
The concentration of radionuclide i given in Appendix B Table 2, Column 2
~
to 10CFR20.1001 - 20.2402.
10 Multiplier 10.2.3.1.2.2 Release Mixture The release mixture used for the setpoint determination is the radionuclide mix identified in the release tank grab sample isotopic analysis.
10.2.3.1.2.3 Liquid Dilution Flow Rates Dilution flow rates are obtained from the main control board in the control room. If this information is unavailable, releases may continue for up to 30 days provided the dilution flow rates are estimated every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during the release, in accordance with Administrative Technical Requirements Table 3.3-12.
10.2.3.1.2.4 Projected Concentrations for Releases After determining Em= from Equation 10-4, RETS compliance is verified using Equations 10-5 and 10-6.
C = C![F%!(FL + Fa' )]
z30.sy El C / /0
- DWC,151 (10 6)
The summation is over radionuclides i.
C Concentration of Radionuclide iin the Unrestricted Area
[ Ci/mL]
The calculated concentration of radionuclide i in the unrestricted area as determined by Equation 10-5.
CT Concentration of Radionuclide iin the
[ Ci/mL]
Release Tank The concentration of radioactivity in the radweste discharge tank based on measurements of a sample drawn from the tank.
- DWC, Derived Water Concentration
[ Ci/ml]
The concentration of radionuclide i given in Appendix B, Table 2, Column 2 to 10CFR20.1001 - 20.2402.
10 Multiplier FL Maximum Release Tank Discharge
[gpm]
Flow Rate Ne Circulating Water Blowdown Rate
[gpm]
O 10-9
BYRON Rsvision 1.3 Fcbruary 1997 10.2.3.1.3 Other Liquid Effluent Monitors For all other liquid effluent monitors, including ORE-PR001 and ORE-PRO 10 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 10CFR20.1001 20.2402 in the unrestricted area. Release mixtures are based on a I
representative isotopic mixture of the waste stream or inputs to the waste stream, or defaulted to the mix listed in Table 10-2.
10.2.3.1.4 Conversion Factors The readouts for the liquid effluent monitors are in uCi/ml. Conversion factors are determined for each monitor (CPM /uCi/ml).
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.
O O
I 10-10
.. -.. ~. - - - _ - -. - -
1
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BYRON Revision 1.3 l
February 1997 Tabla 10-1
}
Assumed Cimposition of the Byron Station lloble Gas Effluent
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t i
i Percent of Isotope Effluent Ar-41 00.89 1
a
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Kr-85m 00.18 Ii l
Kr-85 24.9 1
l Kr-87 0.04
}
l Kr-88 00.28 Xe-131m 01.4 i
i Xe-133m 00.57 i
Xe-133 71.1 1
i Xe-135 00.53 1
4 i
Xe-138 00.04 s
i l
1 4-i i
k i
10-11
BYRON Rsvision 1.3 Fcbruary 1997 Table 10-2 Assumed Composition of the Byron Station Liquid Effluent O1 Isotope Concentration Isotope Concentration
( Ci/ml)
( Cl/ml)
Ru-103 8.00E - 06 Mn-54 1.00E - 05 Ag-110m 3.00E - 06 Fe-59 5.00E - 06 Te-127 2.00E - 05 Co-58 9.00E - 06 Te-129m 2.00E - 06 Co-60 3.00E - 06 Te-131m 4.00E - 06 Rb-86 2.00E - 06 Te-132 2.00E - 06 Zr-95 6.00E - 06 l-130 3.00E - 07 Nb-95 1.00E - 05 l-131 3.00E - 08 Mo-99 4.00E - 06 l-132 8.00E - 07 l-133 1.00E - 07 l-135 4.00E - 07 Cs-134 9.00E - 07 Cs-136 9.00E - 06 Cs-137 2.00E - 06 Ce-144 1.00E - 06 Np-239 1.00E - 05 i
1 10-12 i
BYRON Revision 13 February 1997 i
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Chapter 11 Revision 1.4 Change Summary i
Page 11-1 Removed page revision index.
No longer are individual pages revised.
The revision number for i
.the index is assigned to the chapter.
Page 11-2 Revised radiciodine canister collection and i
analysis frequency to biweekly.
Page 11-6 Added well water location BY-32, due to an ANI auditor request to have a well water location between the site and the Rock River.
1 Page 11-7 Corrected spelling error of " quadrant".
Page 11-9 Corrected title of Health Physics Support Director to Radiation Protection Director.
Page 11-13 Added well-water location BY-32 to the map.
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(
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CHAPTER 11 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM TABLE OF CONTENTS i
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PAGE 1
'v 110 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 11-1 i.
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BYRON Revision 1.4 February 1997 CHAPTER 11 LIST OF TABLES MUMBER TITLE PAGE 11-1 Radiological Environmental Monitoring Program 11-2 O
11-iii
. ~. -
BYRON Ravision 1.4 February 1997 CHAPTER 11
}
LIST OF FIGURES i
tE'MBER IIILE EAGE 4-t l'
j 11 1 Onsite Air Sampling Locations 11-10 i
2 11 2 Offsite Air Sampling Locations 11-11 i
t 11 3 inner Ring and Outer Ring TLD Locations 11-12 f
i 11 4 Ingestion and Waterbome Exposure Pathway Sample Locations 11-13 1
4 i
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11-iv
BYRON Ravilion 1.4 February 1997 i
j CHAPTER 11 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM l
The radiological environmental monitoring pogram fo' the environs around Byron Station is given in Table 11-1.
1 Figures 11-1 through 11-4 show sampling and monitoring locations, l
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Exposure Pathway Samphng or Type of Frequency and/or Samole Samphng or Mondonng Locahons Collechon Freauency of Analysis
- 1. Arborne Radiosodme and
- a. Indmators-Near Field Continuous @ operabon R*R-%se Canester Partculales with parbculate sample I-131 analysis bmeeldy BY-21, Byron Nearsite N.
collechon weeldy, or more on near field and cmtrol 0.26 mi N (0.42 km A) frequently if required by dust samples.'
I BY-22, Byron Nearsde ESE, loading, and radioiodine canister 0.30 mi ESE (0.48 km F) callechon twweeldy Particulate Samoler BY-23, Byron Nearsde S, O.60 mi S (0.97 km J)
Gross beta analysis BY-24, Byron Neanute SW, weeldy filter
(
0.65 mi SW(1.05 km L) change andgamma isotopec malysis*
quarterly a composite filters by locabon on near field and control samples.
- b. Indcators-Far Field BY-1, Byron,3.5 mi N (5.6 km A)
BY-4, Paynes Pt.,4.5 mi SE (7.2 km G)
BY-6, Oregon,4.6 mi SSW (7.4 km K) t t
11-2 5
BYRON Remcon 14 February 1997 Tabte 11-1 (Cont.)
Radiological Environmental Monitoring Program Exposure Pathway Sampling or Type of Frequency and/or Samole Samolina or Monitorina Locations Collection Frecuency of Analysis
- 1. Airbome (Cont'd)
- c. Controis BY-8, Leaf River,7.0 mi NW(11.3 km Q)
- 2. Direct Radiation a.
Indicators-Inner Ring Quarterly Gamma dose quarterly BY-101-1,0.26 mi N (0.42 km A)
BY-101-2,0.26 mi N (0.42 km A)
BY-102-1,1.0 mi NNE (1.6 km B)
BY-102-2,1.0 mi NNE (1.6 km B)
BY-103-1,1.7 mi NE (2.7 km C)
BY-103-2,1.7 mi NE (2.7 km C)
BY-104-1,1.4 mi EME (2.2 km D)
BY-104-2,1.4 mi ENE (2.2 km D)
BY-105-1,1.3 mi E (2.1 km E)
BY-105-2,1.3 mi E (2.1 km E)
BY-106-1,1.4 mi ESE (2.2 km F)
BY-106-2,1.4 mi ESE (2.2 km F)
PY-107-1,1.4 mi SE (2.2 km G)
BY-107-2,1.4 mi SE (2.2 km G)
BY-108-1,0.6 mi SSE (1.0 km H)
BY-108-2,0.6 mi SSE (1.0 km H)
BY-109-1, 0.6 mi S (1.0 km J)
BY-109-2,0.6 mi S (1.0 km J)
BY-110-1,0.6 mi SSW (1.0 km K)
BY-110-2,0.6 mi SSW (1.3 km K)
BY-111-3,0.8 mi SW (1.2 km L)
BY-111-4,0.8 mi SW (1.2 km L)
BY-112-3,0.8 miWSW (1.3 km M)
BY-112-4,0.8 mi WSW (1.3 km M)
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BYRON Revn> son 1.4 February 1997 Table 11-1 (Cont.)
Radiological Environmental RSonitoring Program j
Exposure Pathway Samping or Type of Frequency and/or Samme Sampimg or Mondonng Locahons Collechon Frequency of Analysis f
- 2. Direct Radiabon BY-113-1,0.7 miW (1.1 km N)
(Cont'd)
BY-113-2,0.7 miW (1.1 km N)
BY-114-1,0.8 miWNW(1.3 km P)
BY-114-2,0.8 miWNW(1.3 km P)
BY-115-1,1.0 mi NW (1.6 km Q) -
BY-115-2,1.0 mi NW (1.6 km Q)
BY-116-1,1.4 mi NNW(2.2 km R)
BY-116-2,1.4 mi NNW(2.2 km R)
}
- b. Indicakxs-Outer Ring BY-201-3,4.5 mi N (7.2 km A) f BY-201-4,4.5 mi N (7.2 km A) i-BY-202-1,4.5 mi NNE (7.2 km B)
BY-202-2,4.5 mi NNE (7.2 km B) 7 BY-203-1,5.1 mi NE (8.2 km C)
BY-203-2,5.1 mi NE (8.2 km C)
BY-204-1,4.2 mi ENE (6.8 km D)
BY-204-2,4.2 mi ENE (6.8 km D)
BY-205-1,3.9 mi E (6.3 km E)
BY-205-2,3.9 mi E (6.3 km E)
BY-206-1,4.2 mi ESE (6.8 km F)
[
BY-206-2,4.2 mi ESE (6.8 km F)
[
BY-207-1,4.2 mi SE (6.8 km G)
BY-207-2,4.2 mi SE (6.8 km G) a I
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Radiological Environmental Monitoring Program Exposure Pathway Sampling or Type of Frequency and/or Samole Samolina or Monitorina Locations Collection Freauency of Analysis
- 2. Direct Radiation BY-208-1,4.1 mi SSE (6.6 km H)
(Cont'd)
BY-208-2,4.1 mi SSE (6.6 km H)
BY-209-1,3.8 mi S (6.1 km J)
BY-209-4,3.6 mi S (5.8 km J)
BY-210-3,4.75 mi SSW (7.6 km K)
BY-210-4,4.75 mi SSW (7.6 km K)
BY-211-1, 5.2 mi WSW (8.4 km L)
BY-211-4,4.9 miWSW(7.9 km L}
BY-212-1,4.9 mi SW (7.9 km M)
BY-212-4,4.9 mi WSW (7.8 km M)
BY-213-1,5.0 mi W (8.0 km N)
BY-213-4,5.0 mi W (8.0 km N)
BY-214-1,4.8 mi WNW (7.7 km P)
BY-214-4,4.8 mi WNW (7.7 km P)
BY-215-1,5.2 mi NW (8.4 km Q)
BY-215-4,5.2 mi NW (8.4 km Q)
BY-216-1,4.8 mi NNW(7.7 km R)
BY-216-2,4.8 mi NNW(7.7 km R)
- c. Other Indicators One at each airbome location given in part 1.a and 1.b.
- d. Control One at each airbome control location given in part 1.c.
11-5
m.
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O O
O BYRON Revidson 1.4 February 1997 Yable 11-1 (Cont.)
Radiological Environmental Monitoring Program Exposure Pathway Sampling or Type of Frequency and/or Samole Samolino or Monitonna Locations C% Freauency of Analysis
- 3. Waterborne
- a. Ground /Well
- a. Indicators Quarterly Gamma isotopic
- and tritium analysis BY-14, Comed Offsite Well quarterly.
0.3 mi ESE (0.5 km F)
BY-18, McCoy Farmstead 1.0 mi SW(1.6 km L)
BY-32, Wolford Well 1.0 rai W (1.6 km N)
- b. Drinkina There is no dnnlang water pathway wimin 6.2 mi downstream of the station.
- c. Surface BY-12, Oregon Pool of Rock River, Weeldy grab samples.
Gross beta and gamma Downstream of Descharge, isotopicanalysis on 4.5 mi SSW(7.2 km K) -
monthly composite; tritium analysis on quarterly composite.
- d. Control BY-29, Byron, Upstream ofIntake Weekly grab samples.
Gross beta and gamma 3.5 mi N (5.6 km A) isotopicanalysis on monthly composite; tritium analysis on quarterly composite.
- e. Sediment BY-12, Oregon Pool of Rock River, Sem annually Gamma isotopic' Downstream of Discharge, analysis semaannual!y.
4.5 mi SSW(7.2 km K) 11-6
BYRON Revision 1.4 February 1997 Table 11-1 (Cont.)
Radiological Environmental Monitoring Program Exposure Pathway Sampling or Type of Frequency and/or Samole Samplina or Monitorina Locations Collection Freauency of Analysis
- 4. Incestion
- a. Milk
- a. Indicators Biweekly: May through Gamma isotopic' and October; monthly:
1-131 analysis
- on each BY-20, K. Reeverts Dairy Farm, November thmugh April.
sample.
2.1 mi NE (3.4 km C)
BY-27, Kenneth Druien Dairy Farm, 5.8 mi WSW (9.3 km M)
BY-30, Don Roos Dairy, 5.13 mi SE (8.2 km G)
- b. Controls BY-26, Glen Gazzard's Dairy, 13.5 mi N (21.6 km A)
- b. Fish
- a. Indicator BY-31, Rock River in vicinity of Discharge, Two times annually Gamma isotopic' 2.6 mi WNW (4.2 km P) analysis on edible portions.
- b. Control BY-29, Byron, Upstream of Intake 3.5 mi N (5.6 km A)
- c. Food Products
- a. Indicators Annually Gamma isotopic' analysis on each Two samples from each of the four sample.
l major quadrants within 6.2 miles of the station.
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i BYRON Revision 1.4 February 1997 Table 11-1 (Cont.)
Radiological Environmental Monitonng Program Exposure Pathway Sampling or Type of Frequency and/or Samole.
Samplino or Monitorina Locatxms Collection Freauency of Analvsis Sample locations for food products may vary based on availability and therefore are not required to be identified here but shall be taken.
- b. Control Two samples within 9.3 to 18.6 miles of the station.
11-8
BYRON Revision 1.4 February 1997 TABLE 11-1 (Cont'd)
Radiological Environmental Monitoring Program
'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.
2Airbome 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 ten times the yearty mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
' Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the station.
'l-131 analysis means the analytical separation and counting procedure are specific for this radionuclide.
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Revision 1.4 Change Summary Page 12-ii Removed page revision index.
No longer are individual pages revised.
The revision number for the index is assigned to the chapter.
Page 12-5 Corrected a typographical error in the title for the REMP program.
Page 12-15 Corrected the column header alignment.
Page 12-16 Corrected the column header alignment.
Page 12-20 Corrected spelling error of " Emitters".
Page 12-29 Corrected spelling error of " Emitters".
Page 12-30 Corrected spelling error of " elapsed".
Page 12-40 Added a paragraph to explain the requirements behind program deviations.
This information was located in Chapter 9 (UREMP) and has been relocated into this chapter. Chapter 9 is to be 7_
I 1
deleted.
O Page 12-41 Revised wording of 12.5.1.A.3.
The wording was taken from Chapter 9 and relocated into this chapter.
Page 12-43 Revised radiciodine collection and analysis frequency to biweekly.
Page 12-44 Added and revised ranges for inner ring and outer ring TLDs.
Original ranges did not encompass all existing locations.
Corrected spelling error
" access".
Page 12-45 Changed the number of required samples to agree with the addition of a well water sampling location in Chapter 11.
Page 12-46 Revised control range for milk sampling.
Original ranges did not incorporate cows between 6.2 and 9.8 miles.
Page 12-48 Changed title in footnote 2.
Revised footnote 8 to indicate that milk sampling may be discontinued but not required to be discontinued.
,O N.)
i
Page 12-50 Added footnote 6 to I-131 Page 12-52 Added footnote 6 to indicate I-131 LLD applies l
to specific I-131 analyses. This was at the request of the contractor laboratory to remove any confusion when I-131 may be identified on a gamma spec when it is not specifically being analyzed for.
Page 12-56 Corrected spelling error of " calendar".
Page 12-58 Corrected a typographical error.
There was a footnote notation in the last paragraph but no footnote e.xisted or was required.
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4 4
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l SPECIAL NOTE 1
The transfer of the Byron Radiological Effluent Technical Specifications to the ODCM was approved by the Nuclear Regulatory Commission in Technical Specification Amendment 46, dated April i
13,1992.
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CHAPTER 12 i
RADIOACTIVE EFFLUENT TECHNICAL 5TANDARDS j
(RETS) l TABLE OF CONTENTS i
PAGE 12.1 DEFINITIONS 12-2 i
12.2 INSTRUMENTATION 12-6 4
l 1.
Radioactive Liquid Effluent Monitoring Instrumentation 12-6 2.
Radioactive Gaseous Effluent Monitoring Instrumentation 12-11 8
12.3 LIQUID EFFLUENTS 12 18 1.
Concentration 12 2.
Dose 12-23 3.
Liquid Radweste Treatment System 12-25 12.4 GASEOUS EFFLUENTS 12-27 1.
Dose Rate 12-27 2.
Dose - Noble Gases 12 32 i
3.
Dose - lodine-131, lodine.133, Trttium, and Radioactive Material in Particulate Form 12-34 4.
Gaseous Radweste Treatment System 12-36 5.
Total Dose 12-38 12.5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 12-40 i
1, Monitoring Program 12-40 2.
Land Use Census 12-53 i
3.
Interiaboratory Comparison Program 12-54 12.6 REPORTING REQUIREMENTS 12-55 1.
Annual Radiological Environmental Operating Report 12-55 2.
Annual Radioactive Effluent Release Report 12-57 3.
Offsite Dose Calculation Manual (ODCM) 12-58 4.
Major Changes to Liquid and Gaseous Radweste Treatment Systems 12-59 g wnweemwmaneroney12ri-4.aoc O
12-iii
l BYRON Rrvision 1.4 Fcbruary 1997 CHAPTER 12 RADIOACTIVE EFFLUENT TECHNICAL STANDARDS (RETS)
LIST OF TABLES TABLE TITLE Pffag 12.0-1 Compliance Matrix 12-1 12.1-1 Frequency Notations 12 5 12.2 1 Radioactive Liquid Emuent Monitoring Instrumentation 12 7 12.2-2 Radioactive Liquid Emuent Monitoring Instrumentation Surveillance Requirements 12-9 12.2 3 Radioactive Gaseous Emuent Monitoring instrumentation 12 12 12.2-4 Radioactive Gaseous Emuent Monitoring Instrumentation Surveillance Requirements 12-15 12.3-1 Radioactive Liquid Waste Sampling and Analysis Program 12 19 12.4-1 Radioactive Gaseous Waste Sampling and Analysis Program 12-29 12.5-1 Radiological Environmental Monitoring Program 12-43 12.5-2 Reporting Levels for Radioadivity Concentrations in Environmental Samples 12 49 12.5-3 Detection Capabilities for Environmental Sample Analysis 12 50 i
12-iv
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BYRON Revision 1.4 Fsoruary 1997 BYRON STATION j
Table 12.01 g
i COMPLIANCE MATRIX l
j Regulation Dose Component Limit ODCM RETS Technical Equation Specifica-tion i
- 1. Gamma air dose and beta air dose due to A1 12.4.2 6.8.4.e.8 Appendix I airbome radioactivity in effluent plume.
A-2 s
L
)
- a. Whole body and skin dose due to airbome A4 N/A N/A radioactivity in effluent plume are reported A-7 only if certain gamma and beta air dose l
criteria are exceeded A-13 12.4.3 6.8.4.e.9 i
- 2. CDE for all organs and all four age groups due to l
lodines and particulates in effluent plume. All A-29 12.3.2 6.8.4.e.4 l
pathways are consadored i
- 3. CDE for all organs and all four a0e groups due to radioactivity in liquid effluents.
- 1. TEDE, totaling all deep dose equivalent A-38 6.8.4.e.3
~
j components (direct, ground and plume shine) i and committed effective dose equivalents (all pathways, both aistWme and liquid-bome). CDE evaluation is made for adult only usin0 FGR 11 data base.
40 CFR 190
- 1. Whole body dose (DDE) due to direct dose, A-35 12.4.5 6.8.4.e.10 (now by ground and plume shine from asl sources at a i
reference, also station.
j part of A-13 10 CFR 20)
- 2. Organ doses (CDE) to an adult due to all 3
pathways.
l Technical
A-8 12.4.1 6.8.4.e.7 Specifications and organ (CDE) dose rates to an adult due to A-9 l
radioactivity in airbome effluents. For the organ A-28 i
dose, only inhalation is considered.
A-32 12.3.1 6.8.4.e.2
{
- 2. " Instantaneous" concentration limits for liquid j
effluents.
i a
1 1
g%matemennemeranwy12ri-4 doc 1
{
12 1 l
i i
i 1
~
BYRON R: vision 1.4 Fcbruary 1997 12.1 DEFINITIONS 12.1.1 ACTION shall be that which prescribes remedial measures required under designated conditions.
12.1.2 ANALOG CHANNEL OPERATIONAL 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 shall include 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 and accuracy to known values of input. The CHANNEL CAllBRATION shall encompass the entire channel including the sensors and alarm, interiock and/or trip functions and may be performed by any series of sequential, overiapping, 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 shall indude, where possible, comparison of the channel indica; ion and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
12.1.5 DIGITAL CHANNEL OPERATIONAL 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 EQUIVALENT l-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 thymid dose conversion factors used for this calculation shall be those listed in Table 111 of TID-14844,
" Calculation of Distance Factors for Power and Test Reactor Sites".
12.1.7 FREQUENCY - Table 12.1 1 provides the definitions of various frequencies for which surveillances, sampling, etc., are performed unless defined otherwise. The 25% variance shall not be applied to Operability Action statements. The bases to Technical Spedfication 4.0.2 provide clartfications to this requirement.
12.1.8 MEMBERfS) OF THE PUBLIC means any individual except when that individual is receiving an occupational dose.
12.1.9 OCCUPATIONAL DOSE means the dose received by an individual in the course of employment in which the individual's assigned duties involve exposure to radiation and/or to radioadive 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.
9 W2rt-toac 12-2
~
1 BYRON Rsvision 1.4 I
Febru ry 1997 i
l 12.1.10 OPERABLE / OPERABILITY a system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified I
h function (s), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required i
for the system, subsystem, train, component, or device to perform its function (s) are j
also capable of performing their related support function (s).
12.1.11 OPERATIONAL MODE (i.e. Mode) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature a
specdied in Table 1.2 of the Technical Specifications.
l-12.1.12 PROCESS CONTROL PROGRAM (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 l
actual or simulated wet solid wastes will be accomplished in such a way as to assure I
j compliance with 10 CFR Parts 20,61,71 and State regulations, burial ground requirements, and other requirements goveming the disposal of radioactive wastes.
i i
12.1.13 PURGE / PURGING shall be any controlled process of discharging air or gas from a l.
confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify i
the confinement.
1 4
12.1.14 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor i
coolant of 3411 MWm.
l
{
12.1.15 SITE BO!!lDelLY shall be that line beyond which the land or property is not owned, j
leased, or r Aferwise controlled by the licensee.
!p 12.1.16 SOLIDIFl0ATION shall be the conversion of wet wastes into a fonn that meets
- - (
shipping and burial ground requirements 1
12.1.17 SOURCE CHECK shall be the qualitative assessment of channel response when the i
channel sensor is exposed to a source of increased radioactivity.
I i
12.1.18 THERMAL POWER shall be the total core heat transfer rate to the reactor coolant.
12.1.19 UNRESTRICTED AREA means an area, access to which is neither limited nor j
controlled by the licensee i
i 12.1.20 VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and
{
installed to reduce gaseous radioiodine or radioedive material in particulate form in
{
effluents by possing ventilation or vent exhaust gases through chemal adsorbers i
and/or HEPA filters for the purpose of removing lodines or particulates from the
~
gaseous exhaust stream prior to the release to the environment. Such a system is j
not consulered to have any efferA on noble gas effluents Engineered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
4 a
l 4
j gwnedomwmemeyroneyt2rt-4 doc 4
12-3 i
i e
-ae
+
vp t-m -
"e er P
BYRON R:: vision 1.4 FIbrusry 1997 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 HOLDUP 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 the ODCM Computer Program.
a.
ACTUAL - ACTUAL refer? t: udng 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 Mblic. This data is not incorporated into the database.
awoocm nn-wonyawooc 12-4
BYRON Rsvision 1.4 l
Fcbru:ry 1997
.i i
i l
TABLE 12.1-1 1
EREQUENCY NOTATIONS *
)
i Notation Frecuency y
4 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 /> i
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 /> l
W - Weekly At least once per 7 days
]
M - Monthly At least once per 31 days 5
Q - Quarterly At least once per 92 days 4
i SA - Semiannually At least once per 184 days j
A-Annually At least once per 366 days R RefuelCycle At least once per 18 months l
S/U - Startup Prior to each reactor startup N.A.
Not apphcable i
P - Prior Prior to each radioactive release
- Each frequency requirement shall be performed within the specified time interval with the maximum alloweble extension not to exceed 25% of the frequency interval. The 25% variance shall not be applied to Operability Action statements The bases to Technical Specification 4.0.2 provide clartfications to this requirement These frequency notations do not apply to the Radiological Environmental Monitoring Program as described in Section 12.5.
l 12-5
BYRON Revision 1.4 February 1997 12.2 INSTRUMENTATION 12.2.1 Radioactive Liquid Effluent Monitoring Instrumentation Qoerability 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.
Aooficability: 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 the next Radioactive Effluent Release Report pursuant to Section 12.6 why this inoperability was not corrected within the time specified Surveillance Recuirements 12.2.1.B Each radioactive liquid emuent monitoring instrumentation chann91 shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION and DIGITAL snd 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, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of RETS.
The OPERABILIT( 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.
o wnedemwwwmty12ri-4 doc
__-.m._._
m O
O O
BYRON Revision 1.4 February 1997
[
TABLE 12.2-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTAT;ON MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1.
Radioactivity Monitors Providing Alarm and i
Automatic Terminett..,f Release t
a.
Liquid Radweste Emuent Line (ORE-PR001) 1 31 b.
Fire and Oil Sump (ORE-PR005) 1 34 l
c.
Condensate Pohsher Sump Discharge l
(ORE-PR041) 1 34 i
2.
Radioactivity Monitors Providing Alarm Dut Not Providing Automatic Termination of Release a.
Essential Service Water 1)
Unit 1 a)
RCFC 1 A and 1C Outlet (1RE-PR002) 1 32 b)
RCFC 18 and 1D Outlet (1RE-PR003) 1 32 2)
Unit 2 t
a)
RCFC 2A and 2C Outlet (2RE-PR002) 1 32 b)
RCFC 28 and 2D Outlet (2RE-PR003) 1 32 b.
Station Blowdown Line (ORE-PRO 10) 1 32 3.
Flow Rate Measurement Devices j
a.
Liquid Radwaste Emuent Line (Loop-WXOO1) 1 33 b.
Liquid Radwaste Effluent Low Flow Line (Loop-WX630) 1 33 c.
Station Blowdown Line (Loop-CWO32) 1 33
[
g wmbdtmwmex'byronty12rt-4 doc 12-7
{
BYRON Ravision 1.4 F1bruary 1997 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 - Wrth 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 />, grab samples are collected and analyzed for radioactivity 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 estimate flow.
ACTION 34 - 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 analyzed for radioactivity at a lower limit of detection as specified in Table 12.3-1:
a.
At least once por 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 microCurle/ gram DOSE EQUlVALENT 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.
gremedemwmexeyroney12r1-4 doc
. _.... _. _.- ~ _ _ -. _ _.....--.. _. _ _._.... _. _ _ _. _.. _ _. _...
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BYRON Revision 1.4 Fetxuary 1997 TABLE 12.2-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS DIGITAL ANALOG CHANNEL-CHANNEL-CHANNEL SOURCE CHANNEL
. OPERATIONAL OPERATIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST TEST 1.
Radioadivity Monitors Providmo Alarm and Automatic Termination of Release
- a. Liquid Radweste Efiluent Line (0RE-PR001)
D P
R(3)
Q(1)
N.A.
- b. Fire and Oil Sump Discharge (ORE-PR005)
D M
R(3)
Q(1)
N.A.
- c. Condensate Polisher Sump Discharge (ORE-PR041)
D M
R(3)
Q(1)
N.A.
2.
Radioactivity Monitors Providin0 Alarm But Not Providmg Automatic Termmation of Release a.
Essential Service Water 1)
Unit 1 a)
RCFC 1A and 1C Outlet (1RE-PR002) D M
R(3)
Q(2)
N.A.
b)
RCFC 1B and 1D Outlet (1RE-PR003) D M
R(3)
Q(2)
NA j
2)
Unit 2 a)
RCFC 2A and 2C Outlet (2RE-PR002) D M
R(3)
Q(2)
NA l
b)
RCFC 28 and 2D Outlet (2RE-PR003) D M
R(3)
Q(2)
N.A.
b.
Station Blowdown Line (ORE-PRO 10)
D M
R(3)
Q(2)
N.A.
3.
Flow Rate Measurement Devices a.
Liquid Radweste Effluent Line (Loop WX001)
D(4)
N.A.
R N.A.
O b.
Liquid Radweste Effluent Low Flow Line D(4)
N.A.
R N.A.
O f
i (Loop WX630) c.
Station Blowdown Line (Loop-CWO32)
D(4)
N.A.
R N.A.
O i
i g Wmwdctn' annex %ron%12rt-4 doc f
12-9 i
BYRON R: vision 1.4 Fzbruary 1997 TABLE 12 2-2 (Continued)
RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVElLLANCE REQUIREMENTS TABLE NOTATIONS (1)
The OlGITAL CHAR 4NEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
i a.
Instrument indicates measured levels above the Alarm / Trip Setpoint, or i
b.
Circuit failure (monitor loss of communications - alarm only, detectm loss of counts, or monitorloss of power), or c.
Detector check source test failure, or d.
Detector channel out-of-service, or e.
Monitorloss 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 fsilure (monitor loss of communications - alarm only, detector loss of counts, or monitorloss of power), or c.
Detector check source test failure, or d.
Detector channel out-of-service, or e.
Monitorloss of sample flow.
1 (3)
The initial CHANNEL CAllBRATION shall be performed using one or more of the reference standards cettified 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.
(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 we made.
gwnew. ce rrr.ey12r1-4 doc r
12-10
. ~ _
BYRON Revision 1.4 i
February 1997 1
4 12.2.2 Radioactive Gaseous Effluent Monitorina Instrumentation Ooerability Recuirements 4
12.2.2.A The radioactive gaseous effluent monitoring instrumentation channels shown in Table i
12.2-3 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of 3-Section 12.4 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the l
ODCM.
AoolicatMisty As shown in Table 12.2-3 i
1.
With a radioactive gaseous effluent monitoring instrumentation channel Alarrr/ Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous emuents monitored by the affected channel, or declare the channel inoperable.
2.
With less than the minimum number of radioactive gaseous effluent mondoring 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 the next Radioactive Emuent Release Report pursuant to Section 12.6 why this inoperability was not corrected within the time specified Surveillance Reauirements 12.2.2.8 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and DIGITAL and CHANNEL OPERATIONAL TEST at the frequencies shown in Table 12.2-4.
Bases 12.2.2.C The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous emuents during actual or potential releases of gaseous effluents The Alarm /Tnp 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 instrumentat)on s' c includes provisions for monitoring (and controlling) the a
concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is ou Jtent wth the requirements of General Design Criteria 60,63, and 64 of Appendix A to RETS. The sensitivity of any noble gas activity monitor used to show compliance with the gaseous emuent release requirements of Section 12.4 shall be such that conoontrations as low as 1x10* uCi/cc are measurable.
s wnwsomwmemeymney12r1-4 doc 12-11 l
4
l l
l BYRON Revision 1.4 February 1997 TABLE 12.2 'J RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION l
MINIMUM CHANNELS i
INSTRUMENT OPERABLE APPLICABILITY ACTION 3
1.
Plant Vent Monitoring System - Unit 1 f
a.
Noble Gas Activity Monitor-Providing Alarm
- 1) High Range (1RE-PR028D) 1 39
- 2) Low Range (1RE-PR0288) 1 39 l
b.
lodine Sampler (1RE-PR028C) 1 40 c.
Particulate :bmpler (1RE-PR028A) 1 40 i
d.
Effluent System Flow Rate Measuring Device (LOOP-VA019) 1 36 e.
Sampler Flow Rate Measuring Device (1FT-PR165) 1 36 i
2.
Plant Vent Monitoring System - Unit 2 a.
Noble Gas Activity Monitor-Providing Alarm
- 1) High Range (2RE-PR028D) 1 39
- 2) Low Range (2RE-PR0288) 1 39 b.
lodine Sampler (2RE-PR028C) 1 40 c.
Particulate Sampler (2RE-PR028A) 1 40 d.
Effluent System Flow Rate f
Measuring Device (LOOP-VA020) 1 36 e.
Sampler Flow Rate Measurin0 Device (2FT-PR165) 1 36 g wmedermamexeyroney12rt-4 doc 12-12
l BYRON Revision 1.4 February 1997 TABLE 12.2-3 (Continued)
RAOlOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (CONTD)
MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION 3.
Not Used.
4.
Gas Decay Tank System a.
Noble Gas Activity Monitor-Providing Alarm and Automatic Termination of Release (ORE-PR002A and 2B) 2 35 5.
Containment Purge System a.
Noble Gas Activity Monitor-Provid'ag Alarm (RE-PR001B) 1 37 b.
lodine Sampler (RE-PR001C) 1 40 c.
Particulate Sampler (RE-PR001A) 1 40 6.
Radioactivity Monitors Providing Alarm and Automatic Closure of Surge Tank Vent-Component Cooling Water Line (ORE-PR009 and RE-PR009) 2 41 o m w== w oneytaidcoc 12-13
~
BYRON R vision 1.4 F;bruary 1997 TABLE 12 2-3 (Continued)
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION (CONT'D)
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 staff independently verify the release rate calculations and discharge valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway.
1 l
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 i
established.
ACTION 38 -
Not used.
ACTION 39 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, emuent 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 12.4-1.
ACTION 40 -
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affeded 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 />, gaseous grab samples are collected and analyzed for radioactivity at a lower limit of detection as specified in Table 12.4-1.
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i BYRON Revision 1.4 February 1997 TABLE 12.2-4 j
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS l
DIGITAL CHANNEL MODES FOR WHICH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE i
l FUNCTIONAL UNIT CHECK CHECK CALIBRATION TEST IS REQUIRED 1.
Plant Vent Monitoring System - UnK 1 a.
Noble Gas Activity Monitor-Providing I
Alarm
- 1) High Range (1RE-PR0280)
D M
R(3)
O(2) i
- 2) Low Range (1RE-PR0288)
D M
R(3)
Q(2) i b.
lodine Sampler (1RE-PR028C)
D M
R(3)
Q(2) i c.
Particulate Sampler (1RE-PR028A)
D M
R(3)
Q(2) j L
d.
Effluent System Flow Rate Measuring D
N.A.
R Q
Device (LOOP-VA019) i i
e.
Sampler Flow Rate Measunng Device D
N.A.
R Q
[
{
- 2. Plant Vent Monitoring System - Unit 2 f
t a.
Noble Gas Activity Monitor-Providing Alarm i
I
- 1) High Range (2RE-PR0280)
D M
R(3)
Q(2) t
- 2) Low Range (2RE-PR0288)
D M
R(3)
Q(2)
I I
b.
lodine Sampler (2RE-PR028C)
D M
R(3)
O(2) j i
i g wmwk.mwnnenwron%t2rt-4 doc 12-15 l
t
l' BYRON Pevision 1.4 Febiuary 1997 TABLE 12.2-4 (Continued)
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVElLLANCE REQUIREMENTS DIGITAL CHANNEL MODES FOR WH'.CH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE l
FUNCTIONAL UNIT CHECK CHECK CALIBRATION TEST IS REQUIRED 2.
Plant Vent Monitonng System - Unit Two (Continued) c.
Particulate Sampler (2RE-PR028C)
D M
R(3)
O(2) d.
Effluent System Flow Rate Measuring D
N.A.
R O
Device (LOOP-VA020) e.
Sampler Flow Rate Measuring Device D
N.A.
R Q
- 3. Not Used
- 4. Gas Decay Tank System a.
Noble Gas Actuity Monitor P
P R(3)
O(1)
Providing Alarm and Automatic Termination of Release (ORE-PR002A and 28)
- 5. Containment Purge System a.
Noble Gas Activity Monitor-Providing Alarm (RE-PR0018)
D P
R(3)
O(2) b.
lodine Sampler (RE-PR001C)
P P
R(3)
N A.
c.
Particulate Sampler (RE-PR001A)
P P
R(3)
N A.
- 6. Radioactivity Monitors Providing Alarm and Automatic Closure of Surge Tank Vent-Component Coelmo Water Line (ORE-PR009 and RE-rR009)
D M
R(3)
O(1) cm\\annembyronty12rt-4 doc
5 BYRON Rsvision 1.4
{
February 1997 l
I (
TABLE 12 2-4 (Continued)
(
RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEllLANCE REQUIREMENTS l
TABLE NOTATIONS j
- At all times.
i i
(1)
The DIGITAL CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation i
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 i.
monitorloss of power), or a
c.
Detector chJck source test failure, or i
d.
Detector channel out-of-service, or i
j e.
Monitorloss of sample flow.
i l
(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 i
i b.
Cintuit failure (monitor loss of communications - alarm only, detector loss of counts, or l
monitorloss of power), or j
c.
Detector check source test failure, or d.
Detector channel out-of-service, or l
e.
Monitorloss of sample flow.
(3)
The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the Nabonal Institute of Standards and Technology (NIST) or using a'
standants 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.
g%miodernanneweyranty12ri-4 doc 12 17
BYRON Revision 14 Februtry 1997 12.3 LIQUID EFFLUENTS 12.3 1 Concentration Ooerabsfitv Recurrements
- 12. 3.1.A The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Byron Station ODCM Annex, Appendix F, Figure F-1) conforming to 10 times the concentration values in Appendix B, Table 2, Column 2 to 10CFR20.1001-20.2402. For dissolved or entrained noble gases, the concentration shall be limited to d
2x10 microcurie /ml total activity.
Aeolicabiirtv: At all times Action 1.
With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits. immediately restore the concentration to within the above limits.
Surveillance Recuirements 12.3.1.1.B Radioactive liquid wastes shall be sampled and analyzed according to the 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 with the limits of 12.3.1.A.
Bases 12.3.1.C This specification 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 Part 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 Appendix 1,10 CFR Part 50, to a MEMBER OF l
THE PUBLIC, and (2) the limits of 10 CFR Part 20.1301 to the population.
l This specification applies to the release of radioactive materials in liquid 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).
O g wmwicmunnexeyroney12ri-4 doc 12-18
BYRON Revision 14 February 1997 J
TABLE 12 3-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)'"
TYPE FREQUENCY FREQUENCY ANALYSIS (uCi/ml)
- 1. Batch P
P Pnncipal Gamma 5x10" Release Each Batch Each Batch Emitters )
Tanks)
1-131 1x10' P
M Dissolved and 1x10
One Batch /M Entrained Gases (Gamma Emitters)
P M
H-3 1x10' Each Batch Composite")
Gross Alpha 1x10" P
Q Sr-89, Sr-90 5x10*
Each Batch Composite")
Fe-55 1x10'
- 2. Continuous Continuous
- W Pnncipal Gamma 5x10"
. Releases *)
Composite *>
Emitters (*)
1-131 1x10'
- a. Circulating M
M Dissolved and 1x10
Water Grab Sample Entrained Gases Blowdcwn (Gamma Emitters)
- b. Waste Water Continuous
- M H-3 1x10'*
d Treatment Composite )
Discharge to Circulating Water Discharge Gross Alpha 1x10"
- c. Condensate Contineous*
Q Sr-89, St-90 5x10' d
Polisher Sump Composite )
Discharge Fe-55 1x10' g WmWemunnexeyronty12r14 doc 12-19
BYRON Rsvision 1.4 February 1997 TABLE 12.3-1 (Continued)
RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)")
TYPE FREQUENCY FREQUENCY ANALYSIS
( Cl/ml)
M M
- 3. Continuous W
W Principal Gamma 5x10 Release *'
Emitters )0 Grab Sample Essential Service Water, Reactor i
Containment i
Fan Cooler (RCFC)
Outlet Line l-131 1F10*
Dissolved and Entrained Gases (Gamma Emitters) 1x105 H-3 1x10*
O g wnwsemwwmbymney12ri-4 doc 12-20 i
i J
f BYRON Revision 1.4 February 1997 l
TABLE 12.3-1 (Continued)
RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM l
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 back0round, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
j For a particular measurement system, which may include radiochemical separation:
i LLD =
4.66 s.
E
- V + 2.22 x 10% Y exp ( AM) i Where i
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 rata of l
i e
a blank sample as appropriate (counts per minute),
i 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, A=
the radioactive decay constant for the particular radionuclide (sec "), and M=
the elapsed time between the midpoint of sample collection and the time of counting (sec).
Typical values of E V, Y, at and 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.
(2)
A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shaK be isolated, and then thoroughly mixed by a method described in the ODCM to assure representative sampling.
(3)
The principal gamma emitters for which the LLD specification applies include the following radionuclides Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Co-141, and Ce-144. This list does not mean that only these nuclides are to be considered Other gamma peaksihat 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.
swnesemanneerenty12ri-4 doc O
v 12-21
BYRON Rsvision 1.4 FcbruIry 1997 l
TABLE 12.3-1 (Continued)
RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS (4)
A composite sample is one in which the quantity of liquid sampied 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.
(5)
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.
(6)
To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. 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.
(7)
Not required unless the Essential Service Water RCFC Outlet Radiation Monitors RE-PR002 and j
RE-PR003 indicates measured levels greater than 1x10 Ci/mi above background at any time j
during the week.
1 l
O o w w 2 m * = v m or12ri-4.o =
.. ~ _ = _.
BYRON.
Revision 14 L
February 1997 12.3 2 D.Qat
(
Ooerability Reauirements 12.3 2.A The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive matenals in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Byron Station ODCM Annex, Appendix F, Figure F-1) shall be limited:
I Dunng any calendar quarter to less than or equal to 1.5 mrems to the whole body 1.
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.
4 t
j Acolicabilitv: 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 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 Raouirements 12.3.2. B Cumulative dose contnbutions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and 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 I,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 1 to assure that the releases of radioactive matenal in liquid emuents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dose calculation methodology and parameters in the ODCM implement the requirements in Secbon Ill.A of Appendix 1 that conformance with the guides of Appendix l be shown by 1
calculational procedures based on models and data, such that the actual exposure of a I
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 1," April 1977.
g Wm\\odem\\annexeyroniby12r1-4 doc 12-23
BYRON Rsvision 1.4 Fcbruary 1997 12.3.2 Qgjig(Continued)
Bases This section applies to the release of radioactive materials in liquid effluents from each reactor at the site. When shared Radweste 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 Radweste Treatnaent System. For determining conformance to Operability Requirements, these allocations from shared Radweste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the totsl releases per unit.
O 12-24
.. _ _. ~. _ _ - _ _ - _ _
BYRON Rsvision 1.4 Fabruary 1997 i
t i
i 12.3.3 Liould Radweste Treatment System Operability Reauirements j
i 12.3.3.A The Liquid Radweste Treatment System shall be OPERABLE and appropriate portions i
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 Byron Station 4
ODCM Annex, Appendix F, Figure F 1) would exceed 0.06 mrem to the whole body or j
0.2 mrem to any organ in a 31 day pesiod.
I Acohcability At all times.
l Achon 1.
With radioactive liquid waste being discharged without treatment and in excess i
of the above limits and any portion of the Liquid Radweste Treatment System j
not in operation, propere and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that includes the following information:
i a.
Explanation of why liquid redweste was being discharged without L
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 status, and c.
Summary description of action (s) taken to prevent a recurrence
)
Surveillance Reauirements 12.3.3.1.B 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 Radweste Treatment System is not being fully utilized.
12.3.3.2.B The installed Liquid Radweste Treatment System shall be considered OPERABLE by meeting Sections 12.3.1.A and 12.3.2.A.
Bases 12.3.3.B The OPERABILITY of the Liquid Radweste Treatment System ensures that this system will be available for uss whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provules assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably adtievable". This section impleme@ the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objedive given in Section II.D of Appendix 1 to 10 CFR Part 50.
12-25
l BYRON Revision 1.4 Fcbruary 1997 12.3.3 Liouid Radwaste Treatment System (Continued)
Bases The specified limits Goveming 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.
l This section applies to the release of radioactive materials in liquid effluents from each l
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 detem11ning conformance to Operability Requirements, these allocations from shared Radweste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.
O g
s_ _
12-26
-..~..-.-.-. -....-.-.-..
BYRON Revision 1.4 Ftbruary 1997 12.4 GASEOUS EFFLUENTS 12.4.1-Dose Rate Ooerability Reauirements 12.4.1.A The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Byron Station ODCM Annex, Appendix F, Figute F-1) shall be limited to the following:
1.
For noble gases: less than or equal to a dose rate of 500 mrem /yr to the whole body and less than or equal to a dose rate of 3000 mrem /yr to the skin, and 2.
For lodine-131, lodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: less than or equal to a dose rate of 1500 mrem /yr to any organ.
Acolocability At all times.
Action 1.
With the dose rate (s) exceeding the above limits, immediately restore the i
release rate to within the above limit (s).
Surveillance Reauirements j
12.4.1.1.B The dose rate due to noble gases in gaseous effluents shall be determined to be within 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. lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous affluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaming representative samples and performing analyses in accordance with the sampling and analysis program spectfled in Table 12.4-1.
Bases 12.4.1.C This section is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY fmm gaseous effluents from all units on the site will be within the annual dose limits of RETS. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY specified in 10 CFR 20.1301.
BYRON Rrvision 1.4 Ftbruary 1997 12.4 GASEOUS EFFLUENTS Eases For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion facter above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, 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 a dose rate of 500 mrem / year to the whole body or to less than or equal to a dose rate of 3000 mrem / 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 a dose rate of 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. 41586-93 (1968), and Hartwell, J.K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
Interoretations 12.4.1.0 This Technical Standard requires sampling and analysis following a power change exceeding 15% of Rated Thermal Power within a 1 - hour petiod The interpretation of this requirement for power changes is as follows:
a)
Samp6es are required to be pulled within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the power transient.
b)
If there are several power transients that exceed 15% RATED THERMAL POWER per hour, sampling need only be performed after the lag transient but within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the first transient that exceed 15% of RATED THERMAL POWER.
In all cases, sample analysis shall be compieted within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the start of the initial transient.
12 28
o O
BYRON Revision 1.4 February 1997 TABLE 12.4-1 i
RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM i
t MINIMUM TYPE OF LOWER LIMIT OF SAMPLING ANALYSIS ACTIVITY ANALYSIS DETECTION (LLD)"'
l GASEOUS RELEASE TYPE FREQUENCY FREQUENCY
{ Ci/cc)
P P
f
- 1. Waste Gas Decay Each Tank Each Tank Principal Gamma Emitters" 1x10' Tank Grab Sample
- 2. Containment Purge P
P l
Each Purge
- Each Purge
- Principal Gamma Emitters" 1x10*
i Grab Sample H-3 1x10 '
- 3. Auxiliary Bid 0 M"
M Principal Gamma Emitters"'
1x10*
Vent Stack Grab Sample (Unit 1 and 2)
H-3 1x10' Continuous" W
l-131 1x10 "
Charcoal Sample l-133 1x10 "
l Continuous" W
Principal Gamma Emittem"3 Particulate Sample 1x10'"
[
Continuous" Q
Gross Alpha Composite 1x10 "
t Particulate Sample Continuous
- Q St-89 Sr-90 f
Composite 1x10 "
Particulate Sample Continuous N.A.
Noble Gases: Gross Beta or Noble Gas Monitor Gamma 1x10' g wnui..e.-xbyronty12rt-4 coc 12-29 e
m m
BYRON R: vision 1.4 February 1997 TABLE 12.4-1 (Continued)
RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS (1)
The LLD is defined, for purposes of these specifications, as the smallest er acentration of radioactive material in a sample that will yield a net count, above system cackground, that will be detected with 95% probability with only 5% probability of falsely conclucing that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separation:
LLD =
4 66 s.,
E V 2.22 x 10*
- Y exp (-1.s)
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 o
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),
8 2.22 x 10 = the number of disintegrations per minute per microcurie, Y=
the fractional radiochemical yield, when applicable, A=
the radioactive decay constant for the particular radionuclide (sec "), and 1
.st =
the elapsed time between the midpoint of sample collection and the time l
of counting (sec).
Typical values of E. V, Y, and.st 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.
(2)
The principal gamma emitters for which the LLD specification applies include the following radionuchdes: 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 iodine and 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.
gwnedermannertyroney12ri-4 doc 12-30
1 BYRON Rsvision 1.4 February 1997 TABLE 12.4-1 (Continued)
A RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM TABLE NOTATIONS i
(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.
1 (5)
Tntium 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 period 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.
l (7)
Samplas shall be changed at least once per 7 days and analyses shall be completed within 48 j
hours after changing, or after removal from sampler. 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 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. 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:
4 (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 I
E J
W O
12-31
1 BYRON Rsvision 1.4 Fcbruary 1997 12.4.2 Dose - Noble Gases Operability Recuirements 12.4.2.A The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Byron Station ODCM Annex, Appendix F, Figure F-1) shall be limited to the following:
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 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 1
adions to be taken to assure that subsequent releases will be in compliance with the above limits.
Surveillance Recuirements 12.4.2.B Cumulative dose contributions for the curTent calendar quarter and the current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
)
Bases 12.4.2.C This section is providad 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 11.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 1 to assure that the releases of radioactive material in gaseous effluents at or beyond the Site Boundary will be kept "as low as is reasonable achievable.' The f orveillance Requirements implement the requirements in Section Ill.A of Appendix l that conformance with the guides of Appendix I be shown by calculational procedures bened on models and data such that the actual exposure of a MEMBER OF THE PUBLIC throu0h appropriate pathways is unlikely to be substantially underestimated.
i 12-32
_. _ _ _ _ _ _ _ _ _ _ _ _. _. _.. ~. _ _ _ _ ~ _.. _..
BYRON Revision 1.4 FebnJary 1997 I.
.12.4.2 Dose - Noble Gases (Continued)
Et.ises i
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 j
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 i
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 4
Transpott and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors, Revision 1," July 1977. The ODCM equations provided for j
determining the air doses at and beyond the SITE BOUNDARY are based upon the 1
historical average atmospheric conditions, 1
i This section applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radweste Treatment Systems are used by more than i
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 1
estimate should be made of the contributions from each unit based on input conditions, i
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 Redweste Treatment System. For determining conformance to operability Requirements, these allocations from shared Radweste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per i
unit.
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BYRON Revision 1.4 Fcbruary 1997 12.4.3 Dose - todine I-131. fodine-133. Tritium. and Radioactive Material in Particulate Form Ooerability Reauirements 12.4.3.A The dose to a MEMBER OF THE PUBLIC from lodine-131, lodine-133, tntium, 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 Byron Station ODCM Annex, Appendix F, Figure F 1) shall be limited to the fo! lowing:
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.
Aoolicabdity At all times.
Action 1.
With the calculated dose fmm 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.
Surveillance Reauirements 12.4.3.B Cumulative dose contributions for the current calendar quarter and the current calendar year for lodine-131 and 133, tritium, and radionudides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
Bases 12.4.3.C This section is provided to implement the requirements of Sections ll.C. Ill.A and IV.A of Appendix 1,10 CFR Part 50. The Operability Requirements are the guides set forth in Sedion ll.C of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Sedion IV.A of Appendix 1 to assure that the releases of radioadtve material in gaseous effluents at or beyond the Site Boundary will be kept "as low as is reasonable achievable." The ODCM calculational methods spedfied in the Surveillance Requirements 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.
g wnw:rnwinemymney12ri-4 doc
BYRON Rsvision 1.4 February 1997 12.4.3 E8st(Continued) i i
Bases The ODCM calculational methodology and parameters for calculating the doses due to j
the actual release rates of the subject materials are consistent with the methodology j
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 l
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, lodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) individual inhalation of airbome radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposMion onto grassy areas where milk animals and meat pmducing animal's graze with consumption of the milk and mest by man, and (4) deposition on the ground with subsequent exposure to man.
This sodion applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radweste 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 radioedive weste producing units sharing the Radweste Treatment System. For determining conformance to Operability Requirements, these allocations fmm shared Radweste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.
g wnwiemwnereyroner12ri 4 doc O
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BYRON Rsvision 1.4 February 1997 12.4.4 Gaseous Radwaste Treatment System Ooerability Reauirements 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 effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Byron Station ODCM Annex, Appendix F, Figure F-1) wouk1 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, Acolicatxlity At all times.
Action 1.
With radioactive gaseous waste being discharged without treatment and in excess of the above limits, 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.
Identification of any inoperable equipment or subsystems, and the reason 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 protected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Gaseous Radweste 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 and 12.4.2 or 12.4.3.
Bases 12.4.4.1.C The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenevef gaseous effluents require treatment prior to release to the environment.
gsemedemwmerdyroney12ri-4 doc 12-36
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BYRON Revision 1.4 l
February 1997
.i 12.4.4 Gaseous Radwaste Treatment System (Continued)
(
Bases 3
The requirement that the appropriate portions of this system be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous i
effluents will be kept "as low as is reasonably achievable". This section implements the j
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 1 to 10 CFR Part 50.
4 4
The specified limits goveming the use of appropriate portions of the Gaseous Radweste j
Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section 11.B and II.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 Radweste Treatment Systems are used by more than 4
one unit on a site, the wastes from all units are mixed for shared treatment; by such 1
mixing, the effluent releases cannot accurately be ascribed to a specafic 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 tmated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radweste Treatment System. For determining conformance to Operability Requirements, these allocations from shared Radweste Treatment Systems are to be j
added to the releases specifically attributed to each unit to obtain the total releases per unit.
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gwwlcmwmextymey12ri-4 doc 12-37 l
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BYRON Rrvision 1.4 Fzbruary 1997 1
12.4.5 Total Dose Qperability Reouirements 12.4.5.A The mual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC de 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.
Acolicability: At all times.
Action 1.
With the calculated doses from the release of radioactive materials in liquid er gaseous effluents exceeding twice the limits of Sections 12.3.2,12.4.2, or 12.4.3, calculations should be made induding direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of Section 12.4.5 have been exceeded if such is the case, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce i
subsequent releases to prevent recurrence of exceedin0 the above limits and indudes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.2203, shall indude an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cyde sources, induding all effluent pathways and direct radiation, for the calendar year that indudes the release (s) covered by this report. It shall also describe levels of radiation and concentration of radioactive material involved; andIN :ause of the exposure levels or Dncentrations. If the estimated donN ixceeds the above limits, and if the release condition resulting in vicen of 40 CFR Part 190 has not already been corrected, the Special Report shall include a requed 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 granted until staf'." tion on the request is complete.
Surveillance Rmuiremeres i
12.4.5.1.B Cumulative dose contributions from liquid and 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.B Cumulative dose contributions from direct radiation from the units and from radwaste storage tanirs shaft 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.
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12-38
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i BYRON Revision 1.4 l
February 1997 1
4-12.4.5 Total Dose (Cordinued)~
Bases 1
1 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 1
to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the j
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 reador units and outside storage tanks
}
are kept smail. The Special Report will describe a course of action that should result in j
the limitation of the annual dose to a MEMBER t.,. THE 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, wth the exception that dose contributions from other nuclear fuel cycle facilities at the same site orwithin a radius of 8 km must be considered if the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 4
CFR Part 190, the Special Report with a request for a variance (provided the release j
conditions resuiting in violation of 40 CFR Part 190 have not already been corrected), in j
accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.2203, is considered to i
be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff j
adion is completed. The variance 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 i
20, as addressed in Sections 12.3.1 and 12.4.1. An individual is not consulered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nucisar fuel cycle.
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BYRON Revision 1.4 Fzbruary 1997 12.5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 12.5.1 Monitorino Procram Ooerability Recuirements 12.5.1. A The Radiological Environmental Monitoring Program shall be conducted as specified in Table 12.5-1.
Apolicability: 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 Enviro' mental Operating Report required by Section 12.6.1, a description of the reaso' e not conducting the prograrn as required and the plans for preventing a rect.. 6nce.
Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal availability, malfunction of sampling equipment, if a person / business who participates in the 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 deviabons from the sampling schedule will be descnbod in the Annual Radiological Environmental Operating Report.
2.
With the level of radioactivity as the result of plant effluents in an environmental 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.g.2, a Special 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 ana detected in the sampling medium, this report shall be submitted if; concentration (1) concentration (2) +. 2,1.0 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 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 Radiolonical 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.
gwedemannertyoney12r14 doc e
l BYRON Revision 1.4 Fcbruary 1997 l~
12.5 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM (Continued) 3.
If the sample type or sampling location (s) as required by Table 12.5-1 become(s) permanently unavailable, identify suitable altemative sampling media for the pathway of interest and/or specific locations for obtaining replacement samples and add them to the Radiological Environmental Monitoring Program as soon as l
practicable. The specific locations from which samples were unavailable may I
i then be deleted from the monitoring program.
Prepare and submit controlled version of the ODCM within 180 & neluding a j
revised figure (s) and table reflecting the new location (s) with suppu% a information identifying the cause of the unavailability of samples ano justifying t
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, 4
{
and shall be analyzed pursuant to the requirements of Table 12.51 and tne detection capabilities required by Table 12.5-3.
Bases f
12.5.1.C The Radiological Environmental Monitoring Program required by this section provides representative measurements of radiation and of radioactive materials in ihnee exposure pathways and for those radionucledes that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting fmm the station operation. This monitonng program implementsSection IV.B.2 of Appendix l to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring pmgram by verifying that the measurable concentratums of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the envimnmental exposure pathways Guidance for this monitoring program is provided by -
the Radiolog! cal Assessment Branch Technical Position on Environmental Monitoring.
The instially specified monitoring program will be effective for at least the first 3 yeam of commercial operation. Following this period, program changes may be initiated based on operational experience.
The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 12.5-3 are considered optimum for routine environmental measurements in industrial laboratories.
It should be recognized that the LLD is defined as a before the fact limit representing the capatulity of a measurement system and not as an after the fact limit for a particular measurement.
Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedtres Manual, HASL-300 (revesed annually), Currie, LA., "Umits for Qualitative Detodion and Quantitative Determination - Applicellon to Radiochemistry,* Anal. Chem.
40, 586-93 (1988), and Hartwell, J.K., "Detodion Limits for Radioenelytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
12-41
BYRON Revision 1.4 F(bruary 1997 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 drinkirig.
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BYRON Revision 1.4 February 1997 TABLE 12.5-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND COLLECTION TYPE AND l
ANDIOR SAMPLE AND SAMPLE LOCATION 8"'
FREQUENCY FREQUENCY OF ANALYSIS
- 1. Airtnome Samples from a total of eight locations:
Continuous particulate sampler Radmodme Canister-Radoodine and operation with sample collection 1-131 analysis bmeekly Particulates
- a. Indicator-Near Field weekly, or more frequently if on near field samples y
required due to dust loading, and control.m i
Four samples from locations within 4 km (2.5 mi) and radiosodme canister in different sectors.
collection bmeekly.
Particulate Sampler-Gross beta analysis
- b. Indicator-Far Field followingweekly filter change and gamma Three additional locations within 4 to 10 km (2.5 isotopic analysis"'
to 6.2 mi) in defferent sedors, quarterly on composite t
filters by location on near field sam I
control.*ples and
- c. Con'.rol One sample from a control location within 10 to 30 km (6.2 to 18.6 mi).
4 h
f g Wnbdcmwinestyranty12r14 doc 12-43 r
BYRON Revison 14 February 1997 TABLE 12.5-1 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSUP.E PATHWAY NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND COLLECTION TYPE AND ANDIOR SAMPLE AND SAMPLE LOCATIONS"3 FREQUENCY FREQUENCY OF ANALYSIS
- 2. Direct Forty routine monitonng statens either with a Quarterly Gamma dose on each Radiaton*'
thermoluminescent dosimeter (TLD) or with one TLD quarterly.
instrument for measuring dose rate continuously, placed as follows:
- a. Indicator-Inner Ring (100 Series TLD)
One in each meteorological sector, in the general area of the SITE BOUNDARY l
(0.1 to 2 miles);
- b. Indicator-Outer Ring (200 Senes TLD)
One in each meteorological sector, within 3.2 to 10 km (2 to 6.2 mi); and l
- c. Other One at each Airbome locaton given in part 1.a. and 1.b.
The balance of the TLDs to be placed at special interest locatons beyond the Restricted Area where either a MEMBER OF THE PUBLIC or Commonwealth Edison employees have routine l
access.
(300 Senes TLD) 9 9 wmbdem\\annentyronty12rt-4 doc 12-44
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U BYRON Rsvision 14 February 1997 TABLE 12.5-1 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NURABER OF REPRESENTATIVE SARAPLES SAMPLING AND COLLECTION TYPE AND ANDI OR SARAPLE AND SAMPLE LOCATIONS"3 FREQUENCY FREQUENCY OF ANALYSIS
- 2. Direct
- d. Control Quarterly Gamma dose on each Radiation *8 (Cont'd)
TLD quarterly.
One at each Airbome control location given in part 1.c
- 3. Waterborne
- a. Indcator Quarterly Gamma isotopc and
- a. Ground / Well intrum analysis quarterty.
Samples from three sources only if hkely to be affected.88
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isotopic anaryses on One Sample from each community drinking monthly composite, water supply that could be affected by the tntium analysis on station discharge within 10 km (6.2 mi) quarterly composite.
_d_owystreampf,disch_a_rge_ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- c..%rface if no community water supply (Dnnking Water)
Weekly grab samples.
Gross beta and gamraa Water'78 exists within 10 km downstream of discharge isotopc analyses on then surface water samphng shall be performed monthly composite, intium analysis on
- a. Indcator quarterly composite.
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- d. Control
- a. Control Weekly grab samples.
Gross beta and gamma Sample
- isotope analyses on One surface sample upstream of discharge.
monthly composite, tntium analysis on quarterly composite g Wmbdcmunnentyranty12rl-4 doc 12-45
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BYRON Revision 14 February 1997 TABLE 12.5-1 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND COLLECTION TYPE AND ANDIOR SAMPLE AND SAMPLE LOCATIONS"3 FREQUENCY FREQUENCY OF ANALYSIS
- e. Sediment
- a. Indicator Semiannually.
Gamma tsotope At least one sample from downstream (#8 area
within 10 km (6 2 mi).------------------------
pasture (May through October),
1-131" ' analysis on each
- a. Milk
- Samples from milking animals from a monthly at other times sample.
maximum of three locations within 10 km (Novemberthrough Apni).
(6 2 mi) distance.
- b. Control One sample from milking an;mals at a control l
Jocatpn_withiq10_to,30,hmf _2_to 18,6 rni).____
6
- a. Indicator b Fish Two times annually.
Gamma isotopc analysis
- on edible Representative samples of commercially and portions recreationally important species in discharge area.
- b. Control Representative samples of commercially and recreationahy important species in control locations upstream of discharge.
O g Wmbdcmunnembyronty12rt-4 doc 12-46
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O BYRON Revision 14 '
February 1997 TABLE 12.5-1 (Continued)
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NURABER OF REPRESENTATIVE SARAPLES SAMPLING AND COLLECTION TYPE AND ANDI OR SARAPLE AND SARAPLE LOCATIONS"3 FREQUENCY FREQUENCY OF ANALYSIS
- c. Food Products
- a. Indcator Annually Gamma isotopic"'
analysis on each sample.
Two representative samples from the principal food pathways grown in each of four major quadrants within 10 km (6.2 mi):
1 At least one root vegetable sample"'8 At least one broad leaf vegetable (or vegetation)"'8 l
- b. Control Two representative samples similar to indcator samples grown within 15 to 30 km (9.3 to 18.6 mi).
i i
g kmbdcrnkannextyronty12r14 doc 12-47
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BYRON Revision 14 February 1997 TABLE 12.51 (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 desenption where pertinent, shall be provided for each and every sample location in Table 1.1-1 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 ongin in airborne effluents from the station, or at the discretion of the Radiation Protection l
Director.
(3)
Airborne 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 attributable to the effluents from the station.
(5)
One or more instruments, such as a pressunzed ion chamber, for measunng 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 direct radiation.
The 40 locations is not an absolute number. The number of direct radiation monitonng stations may be reduced according to geographical limitations; e g., if a station is adjacent to a lake, some sectors may be over water thereby reducing the number of dosimeters wnich could be placed at the indicated distances. The frequency of analysis or readout for TLD systems will depend upon the charactenstics 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 drinkir:g or irngation 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 l
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.
9!
g WmbdCm\\annexbyrornby12r14 doc 12-48
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1 BYRON Revision 1.4 -
February 1997 TABLE 12.5-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS ANALYSIS (pCi/I)
OR GASES (pCi/m')
(pCi/kO, wel)
(pCi/l)
(pCi/kg, wet)
H-3 20,000"3 t
Mn-54 1,000 30,000 i
Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-85 300 20,000 Zr-Nb-95 400 1-131 2<23 0.9 3
100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 300 (1)
For drinking water samples. Tnis is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/l may be used.
(2)
If no drinking water pathway exists, a value of 20 pCill may be used.
{
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g wnddem'annemeyranty12rl-4 doc 12-49
BYRON Revision 1.4 February 1997 TABLE 12.5-3 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSISm LOWER LIMIT OF DETECTION (LLD)**'
WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS SEDIMENT ANALYSIS (pCi/I)
OR GASES (pCi/m')
(pCi/kg, wet)
(pCi/I)
(pCi/kg, wet)
(pCukg, dry)
Gmss Beta 4
0.01 1000 H-3 200 Mn-54 15 130 Fe-59 30 260 Co-58,60 15 130 Zn-65 30 260 Zr-Nb-95 15 l-131*
1/15 0.07 100 0.5/5*
60 W
Cs-134 15 0.01 100 15 60 150 Cs-137 18 0.01 100 18 80 180 Ba-La-140 15 15 g wnwk:m'annextyroney12r1-4 doc 9
92-9
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BYRON Revision 1.4 Fsbruary 1997 TABLE 12.5-3 (Continued) l (q DETECTION CAPA81LITIEF, FOR ENVIRONMENTAL SAMPLE ANALYSIS
)
- TABLE NOTATIONS I
(1)
The nuclides on this list are not the only nuclides intended to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological j
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.
4 (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 not 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" a
signal.
t 9
For a particular measurement system, which may include radiochemical separation, the LLD is defined as follows:
I i
l 4.66 Se + 3/to
{
i LLD
=
l (E) (V)(2.22) (Y) (exp (-L4))
1 4.66 So LLD (E) (V) (2.22) (Y) (exp (-bt))
1 4
Where: 4.66 Se >> 3/t.
i LLD
=
i hl the "a pnon" Minimum Detectable Concentration (picoCuries per unit mass or volume),
so the standard deviation of the back0round counting rate or of the counting rate of a blank sample,
=
as appropriate (counts per minute),
}
4TotalCounts to a
j E
=
the counting efficiency (counts per disintegration),
V
=
the sample size (units of mass or volume),
2.22
=
the number of disintegrations per minute per picocurie, Y
=
the framonal radiochemical yield, when applicable, A.
the radioactive decay constant for the particular radionuclide
=
(sec"),
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12-51
4 BYRON Remsion 1.4 FGbruary 1997 TABLE 12.5-3 (Continued)
DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS TABLE NOTATIONS l
to counting time of the background or blank (minutes), and
=
the elapsed time between sample collection, or end of the sample collection period, and the time
.st
=
of counting (sec).
Typical values of E V, Y, and.st should be used in the calculation.
It should be recognized that the LLD is defined as a before the fad 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 fluduations, 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 described in the Annual Radiological Environmental Operating Report.
(4)
If no drinking water pathway exists, the value of 15 pCill may be used.
(5)
A value of 0.5 pCl/l shall be used when the animals are on pastule (May through October) and a value of 5 pCi/l shall be used at all other times (November through April).
(6)
This LLD applies only when the analytical separation and counting procedure are specific for this radionuclide.
)
O'
BYRON Revision 14 F6bruary 1997 12.5.2 Land Use census Ooerability Reauirements h - 12.5.2.A.
. A Land Use Ce' sus shall be conducted and shall identify within a distance of 10 km (6.2 miles) the location in
)
each of the 16 cieteorological 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
. Aeolicabilitv: At all times.
Action 1,
With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment. Via the l
sary exposure pathway 20% greater than at a location from which samples are currently being obtained in accc. dance 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 j
monitonng 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 j
change in sampling locations.
"This requirement may be reduced according to geographicallimitations; e.g at a lake site where some sector's will be over water.
r "The nearest industrial facility shall also be documented if closer than the nearest residence.
Surveillance Raouirements O
The Land Use Census shall be conducted during the growing season, between June 1 and October 1, at least 12.5.2.B once per 12 months using that information that will provide the best results, such as by a door-to4oor survey,
'V
. 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.
Bases 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 I 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.
g wmedemwnnexeyroney12r14 doc 12-53
BYRON R3 vision 14 Fcbruary 1997 12.5.3 latertaboratory Comoarison Proaram Operability Reauirements 12.5.3. A Analyses shall be performed on r&dioactive materials supplied as part of an Interlaboratory Companson Program that correspond to samples required by Table 12.5-1.
Acclicability: 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 Interiaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.
Bases 12.5.3.C The requirement for participation 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 as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix l to 10 CFR Part 50.
O gwnwerniennemyron'ey12ri-4 doc 12-54 I
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BYRON Revision 1.4 February 1997 12.6 REPORTING REQUIREMENTS 12.6.1 Annual Radioloaical Environmental Operatino Reoort*
Routine Annual Radiological Environmental Operating Report covering the operation of the Unit (s) dunng 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, j
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.
The Annual Radiological Environmental Operating Report shall include the results of all radiological environmental samples and of all environmental radiation measurements taken during the pedod pursuant to the locations specified in the tables and figures in Chapter 11 of the ODCM, as well as summarized and tabulated results of these analyses 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 Environmental Monitoring
)
i Program; legible maps covedng 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 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 a
reporting levels of Table 12.5-2 but are not the result of plant effluents, discussion of all analyses in which the i
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 corredive actions being taken if the specified program is not being performed as required by Sedion 12.5.3.
i N
- A single submittal may be made for a multiple unit station.
u 4
i g%medemwmertyroney12ri-4 doc 12-55
j BYRON RQuisson 1.4 February 1997 12.6 REPORTING REQUIREMENTS (Cont'd) l
~
12.6.1 Annual Radiolooical Environmental Ooeratino Report (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 stabiltty, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stabilrty.
In lieu of submission with the Annual Radiological Environmental Operating Repert, the licensee has the option of retalning this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
1 The Annual Radiological Environmental Operating Report shall also include an assessment of the radiation doses 1
due to theiradioactive liquid and gaseous effluents released from the Unit or Station durtng the previous calendar 4
p year. Thisreport 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 l
doses Ghall be performed in accordance with the mettdxiology and parameters in the ODCM, and in compliance with 10CFR20 and 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation."
l i
I O
I g:Wntocmwinextpfronty12r1-4 doc 12-56
BYRON Revision 1.4 12.6 REPORTING REQUIREMENTS (Cont'd)
I i
12.6.2-Annual Radioactive Effluent Release Reoort" i
Routine Annual Radioactive Effluent Release Reports covenng the operation of the unit dunng the previous j
calendar year of operation shall be submitted prior to May 1 of the following year.
The Annual Radioactive Emuent Release Reports shall include a summary of the quantrties of radioactive liquid 1'
and gaseous emuents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Measunng Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and l
Gaseous Emuents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, vnth data summartzed on a quarterly basis following the format of Appendix B thereof.
i For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories:
j class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A. Type B, Large Quantity), and SOLIDIFICATION agent or absorbent (e.g., cement, urea formaldehyde).
' The Annual Radioactive Emuent Release Reports shallinclude a list and description of unplanned releases from j
the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid emuents made during the j
reporting period.
The Annual Radioactive Emuent Release Reports shall include any changes made during the reporting period to the PCP as well as any mavx changes to Liquid, Gaseous or Solid Radweste Treatment Systems, pursuant to 4
Section 12.6.3.
l The Annual fladioactive Emuent Release Reports shall also include the following: an explanation as to why the inoperatsty of liquid or gaseous emuent monitoting instrumentation was not corrected within the time specifici in i
Section 12.2.1 or 12.2.2, respectively; and description of the events leading to liquid hoidup tanks or gas storage tanks exceedir.g 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 redweste systems, the submittal shall specify the releases of radioactive material from each unit.
I i
I i
4 h'.M M M
\\
12-57 l
l
l BYRON Remsion 1.4 FGbruary 1997 12.6.3 Offsite Dose Calculation Manual (ODCM) 12.6.3.1 The ODCM shall be approved by the Commission pnor to implementation.
12.6.3.2 Licensee-initiated changes to the ODCM:
Shall be documented and records of reviews performed shall be retained as required by Specification a.
6.10.2. This documentation shai: contain:
1.
Sufficient inform tion 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 20,106,40 CFR Part 190,10 CFR 50.36a, and Appendix i 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 submitted to the Commission in the form of the complete legible copy of the entire ODCM, or updated pages if the Commission retains a controlled copy. If an en it shall be submitted as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was 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 shall indicate the date (e.g., month / year) the change was implemented.
4 O'
t 12-58 w
BYRON Rsvision 1.4 l
February 1997 5
12.6.4 Maior Chanoes to Liouid and' Ge==aus Radweste Treatment Systems" i
Licensee-initiated major changes to the Radweste Treatment Systems (liquid and gaseous):
Shall be reported to the Commission in the Annual Radioactive Effluent Release Report for the a.
period in which the evaluation was reviewed by the Onsite Review and investigative Function. The j
l discussion of each change shall contain:
J i
l 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 l
of additional and supplementalinformation; I
3)
A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems.
4)
An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents that differ from those previously predicted in the License application and amendments thereto; S)
An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;
?
6)
A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents, 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 become effective upon review and soceptance by the Onsite Rev6ew and investigative Function.
" Licensees may choose to submit the information called for in this standard as part of the annual FSAR update.
g:wmencmunnartyoney12r14 doc 12-59
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