ML20210U195
| ML20210U195 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Catawba, 05000000 |
| Issue date: | 09/19/1986 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML15244A163 | List: |
| References | |
| PROC-860919, NUDOCS 8610090447 | |
| Download: ML20210U195 (11) | |
Text
3 September 0, 1985 V
SUBJECT:
Offsite Dose Calculation Manual Revision 12 The General Office Radwaste Engineering staff is transmitting to you this date, Revision 12 of the Offsite Dose Calculation Manual.
As this revision only affects Catawba Nuclear Station, the approval of other station managers is not necessary.
Please update your copy No.
/
and discard affected pages.
REMOVE THESE PAGES INSERT THESE PAGES C-1 Rev. 4 C-1 Rev. 12 C-2 Rev. 4 C-2 Rev. 12 C-4 Rev. 4 C-4 Rev. 12 C-5 Rev. 4 C-5 Rev. 12 C-6 Rev. 4 C-6 Rev. 12 C-7 Rev. 4 C-7 Rev. 12 C-9 Rev. 4 C-9 Rev. 12 C-9a Rev. 12 C-10 Rev. 6 C-10 Rev. 12 NOTE:
As this letter contains "1OEP" information, please insert this in front of the Sept 2mber 8, 1986 letter.
9f 9 !flo Approval Date:
f- / 7-hd Approval Date:
i a Effective Date:
9/19/86 Effective Date:
9/19/86 U
ary L. Birch J. W.
H mpt n, Manager System Radwaste Engineer Catawba Nuclear Station If you have any questions conce ing Revision 12, please call Jim Stewart at (704) 373-5444.
d-James M.
Stewart, Jr.
Associate Health Physicist Radwaste Engineering JMS/pj a. 020 Enclosures G.
GOO 90447861002 p
ADOCK 05000269 PDR
(V JUSTlFICATIONS FOR REVISION 12 Section C1.1 Typo errors corrected spelling of " effluents" (Page C-1)
Section C1.2 Updated description of Gaseous Radwaste System.
(Page C-2)
Section C2.1.1 Updated example with more explicit information.
(Page C-4)
Section C2.1.2a Upda.ted description to more accurately describe (Page C-5) actual station operation.
Sections C2.1.2c Reversed for clarity purposes.
& C2.1.2d (Page C-5)
Sections C2.1.2b Added option to continue releases provided
& C2.2.1d administrative are implemented to assure that (Page C-5) release limits are not exceeded.
Pages C-6 & C-7 Pagination only, no change in information.
Page C-9 Added setpoint guidance to section's C3.1.3 and C3.1.4 Moved description of Turbine Building Sump Discharge Line to page C-9a Section C3.1.5 New page. Added typical radiation monitor setpoint (Page C-9a) calculation to desciption.
l Section C3.2 Changed the word "most" to "the" for clarity purposes, (Page C-10) no change in meaning.
l l
l l
l O
C1.0 CATAWBA N'UCLEAR STATION HADWASTE SYSTEMS
'N C1.1 LIQUID RADWASTE PROCESSING The liquid radwaste system at Catawba Nuclear Station (CNS) is used to collect and treat fluid chemical and radiochemical by products of unit operation.
The system produces effluents which can be reused in the plant or discharged in l
small, dilute quantities to the environment. The means of treatment vary with waste type and desired product in the various systems:
A)
Filtration.- All waste sources are filtered during processing.
In some cases, such as the Floor Drain Tank (FDT) Subsystem of the Liquid Waste (WL) System, filtration may be the only treatment required.
B)
Adsorption - Adsorption of halides and organic chemicals by activated charcoal (Carbon Filter) is used primarily in treating waste in the Laundry and Hot Shower Tank (LHST) Subsystem of the WL System.
FDT waste may also be treated by this method.
C)
Ion Exchange - Ion exchange is used to remove radioactive cations from solution, as in the case of either LHST or FDT waste in the WL System after removal of organics by carbon filtration (adsorption).
Ion exchange is also used in removing both cations (cobalt, manganese) and anions (chloride, fluoride) from evaporator distillates in order to purify the distillates for reuse as makeup water.
Distillate from the g
Waste Evaporator in the WL System and the Boron Recycle Evaporator in the
(
Boron Recycle System (NB) can be treated by this method, as well as FDT, LHST waste, and letdown.
D)
Gas Stripping - Removal of gaseous radioactive fission products is accomplished in both the WL Evaporator and the NB Evaporator.
E)
Distillation - Production of pure water from the waste by boiling it away from the contaminated solution which originally contained it is accomplished by both evaporators.
Proper control of the process will yield water which can be reused for makeup. Polishing of this product
.can be achieved by ion exchange as pointed out above.
F)
Concentration - In both the WL and NB Evaporators, dissolved chemicals are concentrated in the lower shell as water is boiled away. In the case of the WL Evaporator, the volume of water containing waste chemicals and radioactive cations is reduced so that the waste may be more easily and cheaply solidified and shipped for burial.
In the NB Evaporator, the Jilute boron is concentrated to 4% so that it may be reue.ed for makeup to the reactor coolant system.
Figure C1.0-1 is a schematic representation of the liquid radwaste system at Catawba.
5 C-1 Rev. 12 9/19/86
i f) us.s unauvua anunnans anonsua V
The gaseous waste disposal system for Catawba is designed with the capability of processing the fission product gases from contaminated reactor coolant fluids resulting from operation. The system shown schematically in Fig. C1.0-2 is designed to allow for the retention and subsequent decay of the gaseous fission products generated from the reactor coolant system via the chemical and volume control system and/or the boron recycle system in order to limit the need for intentional discharge of high level radioactive gases from the waste gas holdup tanks.
Sources of low-level radioactive gaseous discharge to the environment include periodic purging operations of the containment, the auxiliary building ventilation system, the secondary system air ejector and decayed WG Tanks. With respect to purging operations, the potential contamination is expected to arise from uncollectable reactor coolant leakage.
With respect to the air ejector, the potential source of contamination will be from leakage of the reactor coolant to the secondary system through defects in steam generator tubes. The gaseous waste disposal system includes two waste gas compressors, two catalytic hydrogen recombiners, six gas decay storage tanks for use during normal power generation, and two gas decay storage tanks for use during shutdown and startup operations.
C1.2.1 Gas Collection System The gas collection system combines the waste hydrogen and fission gases from the volume control ta'nks and that from the boron recycle gas stripper evaporator produced during normal operation with the gas collected during the shutdown f
degasification (high percentage of nitrogen) and will cycle it through the catalytic recombiners to convert all the hydrogen to water. After the water vapor is removed, the resulting gas stream will be transferred from the recom-biner into the gas decay tanks, where the accumulated activity may be contained in six approximately equal parts. ~From the decay tanks the gas will flow back to the compressor suction to complete the loop circuit.
C1.2.2 Containment and Auxiliary Building Ventilation Nonrecyclable reactor coolant leakage occurring either inside the containment or inside the auxiliary building will generate gaseous activity.
Gases result-ing from leakage inside the containment will be contained until the containment air is released through the VQ or VP system. The containment atmosphere will be discharged through a charcoal adsorber and a particulate filter prior to release to the atmosphere.
Gases resulting from leakage inside the auxiliary building are released, with-out further decay, to the atmosphere via the auxiliary building ventilation system. The ventilation exhaust from potentially contaminated areas in the auxiliary building is normally unfiltered. However, on a radiation monitor alarm, the exhaust is passed through charcoal adsorbers to reduce releases to the atmosphere.
C1.2.3 Secondary Systems Normally, condensate flow and steam generator blowdown will go parallel C-2 Rev. 12 9/19/86
1 s
O C2.0 RELEASE RATE CALCULATION
\\vl Generic release rate calculations are presented in Section 1.0; these calcula-tions will be used to calculate release rates for Catawba Nuclear Station.
C2.1 LIQUID RELEASE RATE CALCULATIONS There are two potential release points at Catawba. They are as follows:
1.
Liquid Waste Effluent Discharge Line (WL) 2.
Conventional Waste Water Treatment System Effluent Line (WC)
C2.1.1 Liquid Waste Effluent Discharge Line (WL)
There are three low-pressure service water pumps with a minimum flow rate of 16,500 gpm each and four nuclear service water pumps with a minimum flow rate of 9,000 gpm each which provide the required dilution water needed for a telease. The LPSW system flow rate monitor has a variable setpoint which term-inates the release by closing the isolation valve (1 WL124) should the dilution flow fall below the setpoint. The following is a typical equation which can be used to calculate a discharge flow, in gpm.
n f5F
[a I
i
] [ 0.9 ]
RL i=1 MPC.
1
/~'
where:
N-h I
f = the undiluted effluent flow, in gpm.
F
= actual I W Pressure service water flowrate, in gpm.
RL a = the recirculation factor at equilibrim (dimensionless), 1.027.
a=1+O--
= 1 + 4400 cfs s = 1.027 R
OH where:
Q = average dilution flow (120 cfs)
R Q
avera8e fl W Past Wylie Dam (4400 cfs)
H C.
= the concentration of radionuclide, i, in undiluted effluent as determined by laboratory analyses, in pCi/ml, MPC = the concentration of radionuclide, i, from 10CFR20, Appendix B, g
Table II, Column 2.
.If radionuclide, i, is a dissolved noble gas, the MPC. = 2.0E-04 pCi/ml.
1 0.9
= factor used to reduce the WL flowrate (f) to allow the WC system to simultaneously make 10% of the stations releases.
O C-4 Rev. 12 9/19/86
C2.1.2 Conventivaal Waste Watcr Trcat=cnt Sys:c= Effluent Line (WC)
The conventional waste water treatment system effluent is normally considered nonradioactive; that is, it is unlikely the effluent will contain measurable activity above background.
It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements and by periodic analyses of the composite sample collected on that line.
The water sources listed below that are normally discharged via the conventional waste water treatment system and/or the Turbine Building Sump will be diverted if they will cause the WC dischange to exceed administrative limits designed to ensure that station release limits will not be exceeded.
a.
Containment Ventilation Unit Condensate Effluent Line Normally the containment ventilation unit condensate ef fluent line would discharge into the Turbine Building sump, but if radiation is detected above background, the discharge will be terminated and an alarm actuated.
The containment ventilation unit condensate tank will then be pumped to the RHT or WMT, recirculated, sampled, processed thru the WL system if necessary, and then discharged through the liquid waste effluent line and monitored.
b.
Auxiliary Feedwater Sump Pumps and Floor Drain Sump Pump Line Normally the discharge line coming from these sumps will discharge into f.
the Turbine Building sump, but if radiation is detected above background,
(
)
the discharge flow will automatically be routed to the floor drain tank for processing and later be discharged through the liquid waste effluent line. Subsequent radioactive releases may be allowed to discharge into the TBS if administrative 1y controlled to assure that release limits are not exceeded.
c.
Steam Generator Blowdown Line Normally the discharge from the Steam Generator Blowdown will be pumped to the Turbine Building Sump, but if radiation is detected above background, each blowdown flow control valve, the atmospheric vent, and the valve to the Turbine Building Sump will close, thus terminating the discharge.
Blowdown can only be continued by venting the steam to "D" heater and pumping the liquid to the condensate system.
d.
Turbine Building Sump Discharge Line Normally the discharge from the Turbine Building sump will go into the conventional waste water treatment system, but if radiation is detected above background, the sump pumps A, B, and C will stop and an alarm actuated. The Turbine Building sump discharge line can then either be routed to the floor drain tank for processing, routed directly to the liquid waste effluent discharge line, or allowed to continue being dis-charged via the circuit with proper administrative controls implemented to assure that release limits are not exceeded.
I a
i)
C-5 Rev. 12 9/19/86
dak C2.2 GASEOUS RELEASE RATE CALCULATIONS W
The elit vent is the release point for waste gas decay tanks, containment air releases, the condenser air ejector, and auxiliary building ventilation. The condenser air ejector effluent is normally considered nonradioactive; that is, it is unlikely the effluent will contain measurable activity above background.
It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements and/or by analyses of periodic samples collected on that line.
Radiation monitoring alarm / trip setpoints in con-junction with administrative controls assure that release limits are not exceeded; see section C.3.0 on radiation monitoring setpoints.
The following calculations, when solved for flowrate, are the release rates for noble gases and for radioiodines, particulates and other radionuclides with half-lives greater than 8 days; the most conservative of release rates calculated in C2.2.1 and C2.2.2 shall control the release rate for a single release point.
C2.2.1 Noble Gases IK
[(X/Q)Q ]
< 500 mrem /yr, and g
1 I (L. + 1.1 M.)
[(X/Q)Q ]
< 3000 mrem /yr 1
1 f
where the terms are defined below.
C2.2.2 Radioiodines, Particulates, and Other Radionuclides With T 1/2 > 8 Days s
IPg [W Q ]
< 1500 mrem /yr g
1 where:
K.
= The total body dose factor due to gamma emissions for each identified 3
noble gas radionuclide, in mrem /yr per pCi/m from Table 1.2-1.
L.
= The skin dose factor due to beta emissions for each identified noble 3
gas radionuclide, in mrem /yr per pCi/m from Table 1.2-1.
M.
= The air dose factor due to gamma emissions for each identified noble 3
l gas radionuclide, in mead /yr per pCi/m from Table 1.2-1 (unit conver-sion constant of 1.1 mrem / mrad converts air dose to skin dose).
P
= The dose parameter for radionuclides other than noble gases for the 3
inhalation pathway, in mrem /yr per pCi/m and for the food and ground 2
plane pathways in m.(mrem /yr) per pCi/sec from Table 1.2-2.
The dose factors are based on the critical individual organ and most restrictive age group (child or infant).
s
= The release rate of radionuclides, i, in gaseous effluent from all O
Q*.
i release points at the site, in pCi/sec.
C-6 Rev. 12 9/19/86
..v 4llP (X/Q) = 3.10E-05 sec/m. The highest calculated annual average relative concen-3 tration for any area at or beyond the unrestricted area boundary. The location is the NNE sector @ 0.5 miles.
W
= The highest calculated annual average dispersion parameter for estimating the dose to an individual at the controlling location:
8 W = 3.1E-05 sec/m ', for the inhalation pathway. The location is the unrestricted area in the NNE sector @ 0.5 miles.
W = 1.1E-07 meter 2, for the food and ground plane pathways. The location is the unrestricted area boundary in the NE/NNE sector
@ 0.5 miles (nearest residence, and vegetable garden).
Q. = k C.f + k2 = 4.72E+2C.f i
1 1
1 where:
C.
= the concentration of radionuclide, i, in undiluted gaseous effluent, in pCi/ml.
4 f
= the undiluted effluent flow, in efm 3
kg
= conversion factor, 2.83E4 ml/ft
\\O' k
= conversion factor, 6El sec/ min 2
C-7 Rev. 12 4
9/19/86
Adhk MPC = 1.0E-07 pCi/ml, the MPC for an unidantified mixture W
o = 1.027 (See Section C2.1.1)
F = the dilution flow may vary as described in section C2.1.1, but is conservatively estimated at 25,500 gpm, the minimum flow available.
C3.1.2 Containment Ventilation Unit Condensate Effluent Line - EMF 44 As described in Section C2.1.2 on release rate calculations for the containment ventilation unit condensate effluent, it is possible but unlikely that the effluent will contain measurable activity above background.
It is assumed that no activity is present in the effluent until indicated by radiation monitoring.
Since the tank contents are discharged automatically, the radiation monitor setpoint will be set at 1.0E-06 pCi/ml (the monitor's lowest level of detection) plus background to assure that release limits are not exceeded.
C3.1.3 Auxiliary Feedwater Sump Pumps and Floor Drain Sump Pump - EMF 52 As described in Section C2.1.2 on release rate calculations for the auxiliary feedwater sump pumps and floor drain sump pump effluents, it is possible that the effluent will contain measurable activity above background.
It is assumed that no activity is present in the effluent until indicated by radiation monitoring.
Since the, sumps are discharged automatically, the radiation monitor setpoint will initially be set at 1.0E-06 pCi/ml (the monitor's lowest level of detection) plus background to assure that no activity is unknowingly
-m
,7
)
discharged into the Turbine Building sump. The setpoint may be changed after initial detection to allow positive control of effluent releases using the guidance given in Section C3.1.5.
C3.1.4 Steam Generator Blowdown Line - EMF 34 As described in Section C2.1.2 on Release Rate Calculations for the Steam Generator Blowdown, it is possible but unlikely that the effluent will contain measurable activity above background.
It is assumed that no activity is present in the effluent until indicated by radiation monitoring.
Since the Steam Generator Blowdown line is discharged automatically, the radiation monitor setpoint will be initially set at 1.0E-06 pCi/ml (the monitor's lowest level of detection) plus background to assure no activity is unknowingly discharged into the Turbine Building sump. The setpoint may be changed after detection to allow positive control of effluent releases using the guidance given in Section C3.1.5.
/m V
C-9 Rcv 12 9/19/86
C3.1.5 Turbine Building Sump Discharge Line - EMF 31 s
As described in Section C2.1.2 on re" lease rate calculations for the turbine building sumps, it is possible that the effluent will contain measurable activity above background. Since the sump contents are discharged automatical-ly, the radiation monitor setpoint will be initially set at 1.0E-06 ~pCi/ml (the monitor's lowest level of detection) plus background to assure that no activity is unknowingly discharged into the WC system.
Should radioactive effluent releases need to be made from the TBS via the WC system a typical monitor setpoint may be calculated as follows:
M xF c$
$ 1.42E-06 pCi/ml g
where:
c = the gross activity in undiluted effluent, in pCi/ml f = the undiluted effluent flow in gpm; for this example the flow is from the Turbine Building Sumps and is assumed to be 250,000 gallons / day or E175 gpm.
MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture o =,;.027 (See Section C2.1.1)
F = the dilution flow, in gpm, available to dilute the undiluted effluent O'
discharge flow (f); for this example it is assumed that 2550 gpm (10% of the stations RL minimum flow) will be used to dilute the discharge of the WC system. This flowrate will allow the WC system to discharge 10% of the stations MPC and dose limits.
4
\\_ /
C-9a Rev. 12 9/19/86
m fs C3.2 GAS MONITORS
\\
The following equation shall be used to calculate noble gas radiation monitor setpoints based on Xe-133 (Historical data shows that Xe-133 is the predominant 8
isotope):
K(X/Q)D.<500 (see Section C2.2.1) 1 Q. = 4.72E+02 C.f (see Section C2.2.2)'
1 1
C. < 116/f 1
where:
C.
= the gross activity in undiluted effluent, in pCi/ml 1
f
= the flow from the tank or building sources, in cfm 3
K
= from Table 1.2-1 for Xe-133, 2.94E+2 mrem /yr per pCi/m X/Q
= 3.1E-05, as defined in Section C.2.2.2 As stated in Section C2.2, the unit vent is the release point for the contain-me,nt purge ventilation system, the containment air release and addition system, the condenser air ejector, and auxiliary building ventilation.
D
(
For releases from the containment purge ventilation system, a typical radiation monitor setpoint may be calculated as follows:
C. < 116/f = 6.5E-04 1
where:
f = 151,000 cfm (auxiliary building ventilation) + 28,000 cfm (containment purge) = 179,000 cfm For release from the containment air release and addition system, the waste gas decay tanks, the condenser air ejectors, and the auxiliary building ventilation, a typical radiation monitor setpoint may be calculated as follows:
C. < 116/f = 7.7E-04 I
where:
f = 151,000 cfm (auxiliary building ventilation) v4
[
N._
C-10 Rev. 12 9/19/86
--