ML20127M496

From kanterella
Jump to navigation Jump to search
Rev 7 to Offsite Dose Calculation Manual
ML20127M496
Person / Time
Site: Mcguire, McGuire, 05000000
Issue date: 01/01/1985
From: Birch M, Mcintosh M, Stewart J
DUKE POWER CO.
To:
Shared Package
ML20127M449 List:
References
PROC-850101-01, NUDOCS 8507010255
Download: ML20127M496 (37)
v • d • e


Text

{{#Wiki_filter:, O Decembcr 12, 1984

SUBJECT:

Offsite Dose Calculation Manual Revision 7 The General Office Radwaste Engineering staff is transmitting to you this date, Revision 7 of the Offsite Dose Calculation Manual. As this revision only affects McGuire Nuclear

Station, the approval of other station managers is not necessary.

Please update your copy No. I , and discard affected pages. REMOVE THESE PAGES INSERT THESE PAGES Remove all but Insert this package Figure B5.0-1 located behind the tab labeled behind the tab labeled "McGuire" and in front "McGuire" and in front of of Figure B5.0-1. the tab labeled " Catawba". O V 74 ,zs-Approval Date: 11/? /A $- Approval Date: 11/19/A4 Effective Date: 01/1/85 Effective Date: 01/01/M Mary L. Birch M. D. McIntosh, Manager System Radwaste Engineer McGuire Nuclear Station If you have any questions concerning Revision 7, please call Jim Stewart at (704) 373-5444. ?$ @rt&' James M. Stewart, Jr. Associate Health Physicist Radwaste Engineering JMS/nem Enclosures 8507010255 850624 DR ADOCK 0500 9

s s 't 3 JUSTIFICATIONS FOR REVISION 7 Figure Bl.0-1 Turbine Building sump information was added to the figure to illustrate information added s k in Sections B2.1, B2.1.4, and B3.1.4 of-this manual. Table Bl.0-1 Abbreviations that did not appear on Figure Bl.0-1 were deleted; abbreviations that were on Figure Bl.0-1 but not on Table were added. Section Bl.2.2 Wording change to reflect actual station design and operation. .Section Bl.2.3 Wording change to reflect actual station design and operation Figure Bl.0-2 EMF information added to figure. Flow rate information changed to reflect actual station operation. Section B2.1 Turbine Building Sump information added to 7-~ system modifications added per NSM 1106. Section B2.1.1 Change in wording only, no change in meaning. Section B2.1.1 Typo error (2 places): changed upper case "C" to lower case "c" in "cfs". Section B2.1.2 Removed the word "unlikely" since the C Containment Ventilation ~ Unit Condensate Effluent Line~usually contains measurable activity. Section B2.1.4 Section added to reflect system modifications added per NSM 1106. .Section B2.2 Added the words " condenser air ejector" for clarity. e Section B2.2.2 Added the words " dispersion parameter" for clarification purposes since X/Q is referred l to as a dispersion parameter. Section.B2.2.2 Added the words "or deposition" since "W" can be either X/Q or D/Q. Section B3.0 Added/ changed words for clarification purposes, no change in meaning. Section B3.1.1 Changed "100" to "120" to reflect actual station operation.

2 O Section B3.1.2 Removed the word "unlikely" since the Containment Ventilation Unit Condensate Effluent Line usually contains measurable activity. Changed the word "are" to "can be" for the above reason. Changed "100" to "120" to reflect actual station operation. Section B3.1.3 Changed "5.4E-7" to "5.36E-07" to be consistent with the format used in the rest of the section. Added the words "past Cowan's Ford Dam" for clarification of flow's origination. Section B3.1.4 Section added to reflect system modifications added per NSM 1106. Section B.3.2 Added/ changed words/ flow rates to reflect actual station operation. O' Section B4.2.1 Typo error; Cs-137 listed twice; one should have b'een Cs-134. Section B4.3.1 Typo error; 5.73E+03 should have been 5.63E+03. Added definition of "m" for clarification (2 places). Added the words "The time period during which all release (s) are made" for clarification (2 places). Section B4.3.2.1 Changed lower case "w" to upper case "W" for clarification. (2 places). Section B4.4 Added Fuel Cycle section since Catawba Nuclear station can impact on McGuire's 40CFR190 calculations. Table B4.0-1 Added units (sec/m ) for clarification. Table B4.0-2 Added units (m-2) for clarification. Table B4.0-3 Added units (MREM /hr per uCi/ml) for (3 pages) clarification. i. O

3 Section B5.0 Deleted portion of 1st paragraph; special low level I-131 analyses are performed on bi-weekly drinking water and surface water composite samples in order to meet the 1 pCi/ liter LLD for water samples. food products will be collected monthly, during harvest season, from a garden within a 5-mile radius of the station if a garden which is irrigated with water in which liquid plant effluents have been discharged is identified during the annual land use census. Tables B5.0-2 Revised Tables B5.0-2 and B5.0-3 are in and B5.0-3 response to changes in requirements in the McGuire Nuclear station Environmental Radiological Monitoring Program Technical Specifications. O O

i I n l l APPENDIX B i MCGUIRE NUCLEAR STATION SITE SPECIFIC INFORMATION l I i

APPENDIX B - TABLE OF CONTENTS O m Bl.0 MCGUIRE NUCLEAR STATION RADWASTE SYSTEMS B-1 B2.0 RELEASE RATE CALCULATION B-4 B3.0 RADIATION MONITOR SETPOINTS B-8 - B4.0' DOSE CALCULATIONS-B-12 B5.0~ RADIOLOGICAL ENVIRONMENTAL MONITORING B-18 4 4 i i i O 4 f e.,--- .,,-y .-~-,, - -,.,,,..., -.. .,,en. ,,,mw,--.w-------.,..- , -..e r w-w-~ +-.n.c-.

B1.0 MCGUIRE NUCLEAR STATION RADWASTE SYSTEMS . {. -^ Bl.1 ' LIQUID RADWASTE PROCESSING The liquid radwaste system at McGuire Nuclear Station (MNS) 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 small', dilute quantities to the environment. The means of treatment vary with. waste type and desired product in the various systems: A): ' Filtration.- 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) may be used in treating waste in the Laundry and Hot Shower Tank.(LHST). The carbon filter is designed to remove organ-ophosphates and free. chlorine. Activated charcoal need not be used when .these chemicals are not present (e.g., phosphate detergents are not used at the station). Ion exchange resin or other media may be used in the carbon filter vessel as desired. 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 ,O purify the distillates for reuse as makeup water. Distillate from the 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 reactor bleed. D)_ Gas Stripping - Removal aof 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 controllof 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 dilute boron is normally concentrated to 4% so that it may be reused for

4 makeup to the reactor coolant system.

Figure Bl.0-1 is a schematic representation of the liquid radwaste system at i .McGuire. 2: ( ,s B-1

pa UNIT 1 LETDOWN UNIT 2 LETDOWN 'r ir 'v NCOT*

  • ONE WDT k

GAL UNIT

GAL, 350 PER 5.000 NCDT WDT PUMPS PUMPS n

1 I I I W <s 5 7 [l C 9 NCDT l HX l N8 EVAP. pt FEED I DEMIN. I l 1 A A { Q_ I NB EV AP. I h hFEED h h g gF NB WL ERS 1 EVAP EVAP. p t ) ( ) a j g _ A AWEFT RHT RHT NB + WL A B COND. COND. 117,000 117,000 DEMIN. DEMIN. GAL GAL NBEVAP. h",ecoNaQ gnb con FEED gp p PU APS AWEFT 4 g g PUMP [ i d% ' L L ,r EMF 47 I I RMT I I RMT A B 5,@ 5,M gNBm TO I F R GAL R GAL BA TANKS RMT A PUMP TO s i WS ( lI RMWST

P if lf = LHST a FDT VUCDT*

  • ONE 10.000 10,000 e 4,000 PE R GAL CAL GAL UNIT et GAL "8

gp PUMP VUCDT PUMPS AWE FT s/ \\ p" h EFT f s M 'N j l EMF 44 l e [ MST u AL ) PUMP S fs T h1 FILTH 0 EFT I h20 FILTH SLUDG FILTER I Wi& I m EVAP. JERS I TURBINE fg s BUILDING y ( SU MP* / l EMF-31l

  • ONE WMT U T DEMIN.

D ^ rd WM FILTER m CATAWBA RIVER -j T1 = l p APERTUlth eu"" CMD m WMT ( A O 5.000 GAL u l EMF 49 l WMT A i PUMP CCW l FIGURE B1.0-1 McGUIRE NUCLEAR STATION LIQU:D RADWASTE SYSTEM REVISION 6 gg {g 7dr70/d2 SJ/

P TABLE Bl.0-1 fg ABBREVIATIONS k~s/ Systems: CM - Condensate Cooling l KC - Component Cooling NB - Boron Recycle NC - Reactor Coolant WC - Conventional Waste Water Treatment WG - Waste Gas WL - Liquid Waste WM - Liquid Waste Monitor and Disposal Terms: BA - Boric Acid Tank CCW - Condenser Cooling Water CDT - Chemical Drain Tank ECST - Evaporator Concentrates Storage Tank /~'h FDT - Floor Drain Tank O FWST - Fueling' Water Storage Tank (formerly Refueling Water Storage Tank) LHST - Laundry and Hot Shower Tank MST - Mixing and Settling Tank NCDT - Reactor Coolant Drain Tank RBT - Resin Batching Tank RHT - Recycle Holdup Tank RMT - Recycle Monitor Tank RMWST - Reactor Makeup Water Storage Tank SRST - Spent Resin Storage Tank VUCDT - Ventilation Unit Condensate Drain Tank WDT - Waste Drain Tank WEFT - Waste Evaporator Feed Tank WMT - Waste Monitor Tank TABLE Bl.0-1 Revision 7 1/1/85

Bl.2 GASEOUS RADWASTE SYSTEMS U The gaseous waste disposal system for McGuire is designed with the capability of processing the fission product gases from contaminated reactor coolant fluids resulting from operation. The design base for the system shown schema-tically in Fig. Bl.0-2 is the retention, through the plant lifetime, of all the gaseous fission products to be discharged from the reactor coolant system to the chemical and volume control system and other plant systems to eliminate the need for intentional discharge of radioactive gases from the waste gas holdup tanks. Actual system operation is aimed at maximizing storage time for decay prior to infrequent releases. Unavoidable sources of low-level radio-active gaseous discharge to the environment will be from periodic purging operations of the containment, from the auxiliary building ventilation system, and through the secondary system air ejector. With respect to the former, the potential contamination is expected to arise from non-recyclable 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. Bl.2.1 Gas Collection System The gas collection system combines the waste hydrogen and fission gases from the volume control tanks, the boron recycle and liquid waste gas stripper evaporators, and other sources produced during normal operation or the gas (Vg) collected during'the shutdown degasification (high percentage of nitrogen) and cycles it through the catalytic recombiners to convert the hydrogen to water. Af ter the water vapor is removed, the resulting gas stream is transferred from the recombiner into the gas decay tanks, where the accumulated activity may be contained in six approximately equal parts. From the decay tanks, the gas flows back to the compressor suction to complete the loop circuit. Bl.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 is purged. The containment atmosphere will be circulated 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 passed through charcoal adsorbers to reduce releases to the atmosphere upon a radiation monitor alarm. l O v B-2 Revision 7 1/1/85

Bl.2.3 Secondary Systems c When activity is in the secondary system, steam generator blowdown will be recycled through demineralizers to remove activity. The gases removed from the secondary system by the air ejectors are discharged to the unit vent. Gland leak-off steam, which represents a minor source of activity, is routed to the gland condenser. The non-condensable gases are passed through a vent stack to the roof; the condensables are condensed and drained to the condensate storage tank. Figure Bl.0-2 is a schematic representation of the gaseous radwaste system at McGuire. O ~ k B-3 Revision 7 1/1/85

g ROOF VENTS 800,200 af 142 FT. ,_f ~ - ~ ~ ~ ~ ~ ~ 1_ -{"*g CUTSIDE GROUND OUTSIDE AIR %,2.' A'" TURBINE l!". l A I ( (") UNIT PRIMARY l i VENT COOLANT CONDENSE R } I ( ml EMF 35 1 I abefm ST E AM l EMF 38 4 CONDEN5ER E M F.37 OENERATOR L_ [--' STE AM JET AIM EJECTOR 5 l EMF 34 M EMF 33 l HEATERS BLOWOOWN RECYCLE TANK m SECONDARY SYSTEM (2 UNITS) PROM RE ACTOR COOLANT SYSTEM _,,,,,,, (2) HYDROGEN m O I l' R E COM BIN E RS f h D WASTE G AS I ---gu COMPR E SSO RS 3 OXYGEN WATE] I pH2 \\eH [ 3 + TO 1 2 VOLUME // n\\ / n\\ SHUTDOWN VOLUME I CONTROL' CONTROLTANK / % S / # '5 STORAGE UNIT 1 TANKS l TANK l VOLUME i .i.,1 l ( CONTROL TANK 4.,,,,,____,,,,,_,,,,,,.I TO TO UNIT 2 REACTOR R E A CTO R + 4 BORON RECYCLE STORAGE 800 FT.3 SHIM 8LEED TANKS 100 pois EVAPORATOR E b O AS STRIPPE R i l EMF 50 l WASTE CAS SYSTEM EMF 38 E NOf MALLY CLOSED UPPE R VO LUME 18,000 efm CONTAINMENT PURGE (VP) m j ,it g s P A C _,_ @E i W LOWER VOLUME PURGE INLET 9 CONTAINMENT AIR RELEASE [ ~~ 10,000 cf W -m & ADDITION SYSTEM (VO) P A C 300 cfm CONTAINMENT VENTILATION SYSTEMS (2) INTE RN ATL. R ECO RCU. T R AINS (2) l EMF 42 l 8.000 cfm E ACH + 35,000 cfm EJ Pl AlC h 'l F UE L HANDLING ] PER UNIT l EMF 41 l AREA POTENTIALLY $4,000 c fm EJ 5 CONTAMINATED PlAlC OUTSIDE PER UNIT y AREAS 4 OTHER ARE AS AUXILI ARY PER UNIT SUILOlNG SUPPLY LEGENO: AUXILIARY BUILDING SYSTEM (SHARED) P PREFILTER A HIGH. EFFICIENCY PARTICULATE FILTER C CHARCOAL ABSOR8ER

  • FUEL HANDLING AREA IS NORMALLY UNFILTERED. UPON A RADIATION ALARM BY EMF 42.THE EXHAUST WILL BE DIVERTED TO THE FILTERED MODE.

) POTENTIALLY CONTAMINATED AREAS OF THE AUXILIARY BUILDING ARE NORMALLY UNFILTERED. UPON s RADIATION ALARM BY EMF 41 THE EXHAUST WILL BE DIVERTED TO THE FILTERED MODE. FIGURE B1.0-2 McGUIRE NUCLEAR STATION REVISION 6 G ASEOUS RADWASTE SYSTEM JAll 1 G35 1

B2.0 RELEASE RATE CALCULATION ) Generic release rate calculations are presented in Section 1.0; these calcu-lations will be used to calculate release rates for McGuire Nuclear Station. B2.1 LIQUID RELEASE RATE CALCULATIONS There are three potential release points at McGuire. Two of these release points, the waste liquid effluent line and the containment ventilation unit condensate effluent line, discharge into the condenser cooling water flow; the third release point, the Turbine Building sump, can either be discharged into the condenser cooling water flow or into the conventional waste water treat-ment system. B2.1.1 Waste Liquid Effluent Line To simplify calculations for the waste liquid effluent line, it is assumed that no activity above background is present in the containment ventilation unit condensate effluent. This assumption shall be confirmed by radiation monitoring measurements and by periodic analysis of the composite sample collected on that line. For releases made via the waste liquid effluent line, the following calculation shall be performed to determine discharge flow, in gpm: n C.* f < F + (o I ) ~ i=1 MPC. I ,e s f ) where: f = the undiluted effluent flow, in gpm. F = the dilution flow available depending on the number of condenser cooling water pumps in service, in gpm. where: Ft = 2.50E+05 gpm Fh=5.00E+05gpm F3 = 7.50E+05 gpm F. = 1.00E+06 gpm o = The recirculation factor at equilibrim (dimensionless), 2.4 o=1+ = 7,3720 = 2.4 R Q 2670 H where: QR= ver ge luti n I w (3720 cfs) p B-4 Revision 7 1/1/85

Q average fl W Past Cowans Ford Dam (2670 cfs) l ,s H t \\ 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. = 2.00E-04 pCi/ml. 1 B2.1.2 Containment Ventilation Unit Condensate Effluent Line As described in the preceding section, it is possible that the containment l ventilation unit condensate effluent line 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 analysis of the composite sample collected on that line. If measurable activity is present.in the effluent, administrative controls shall be implemented to assure that release limits are not exceeded; see section on radiation monitoring alarm / trip setpoints. B2.1.3 Conventional Waste Water Treatment System Effluent Line 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 issumed that no activity is present in the effluent until indicated by radiation monitoring measurements and by periodic ps analyses of the composite sample collected on that line. Radiation monitoring \\'- alarm / trip setpoints assure that release limits are not exceeded; see section on radiation monitoring alarm / trip setpoints. - B2.1.4 Turbine Building Sump Discharge Line Normally the discharge from the Turbine Building sump is considered non-radioactive; that is, it is unlikely the effluent will contain measurable 4 . activity above background, and will flow into the conventional waste water treatment system. It is assumed that no activity is present in the effluent until indicated by radiation monitoring measurements and by periodic analysis of the composite sample collected on that line. If measurable activity is present in the effluent, sump discharge will be terminated and an alarm activated. At this time the discharge may be routed to the floor drain tank for processing or routed directly to the condenser cooling water (CCW) flow; rather than the conventional waste water treatment system; administrative controls shall be implemented to assure that release limits are not exceeded; see section on radiation monitoring alarm /setpoints. b-U B-5 Revision 7 1/1/85

r= B2.2 GASEOUS RELEASE RATE CALCULATIONS .The unit vent is the release point for waste' gas decay tanks, containment building purges,, 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 back-ground. It is assumed that no activity is present in the condenser air ejector ef fluent until indicated by radiation monitoring measurements and by analyses of periodic samples collected on that line. Radiation monitoring alarm / trip setpoints in conjunction with administrative controls assure that 1 release limits are not exceeded; see section B3.0 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 B2.2.1 and B2.2.2 shall control the release rate for a single release point. B2.2.1 Noble Gases s I'(Kg [(X/Q)Q ] < 500 mrem /yr, and g i s I (Lg + 1.1 M ) [(X/Q)Q ] < 3000 mrem /yr g g -i \\ where the terms are defined below. B2.2.2 Radiolodines, Particulates, and Others s .IPg [W Q ] < 1500 mrem /yr g 1 where: K = The total body dose factor due to gamma emissions for each identified g 3 from Table 1.2-1. noble gas radionuclide, in mrem /yr per pCi/m L = The skin dose factor due to beta emissions for each identified noble I 8 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 I 3 gas radionuclide, in mrad /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 g 3 inhalation pathway, in mremiyr 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). ( B-6 Revision 7 1/1/85 4

Q = The release rate of radionuclides, i, in gaseous effluent from all V release points at the site, in pCi/sec. 3 X/Q = 7.2E-5 sec/m. The highest calculated annual average relative concentration (dispersion parameter) for any area at or beyond the unrestricted area boundary. W =Thehighestcalculatedannualaveragedispersionordepositionparameterl for estimating the dose to an individual at the controlling location: 3 W = 1.2E-5 sec/m, for the inhalation pathway. The location is the unrestricted area in the ESE sector (nearest residence). W = 3.0E-8 meter 2, for the food and ground plane pathways. The location is the' unrestricted area boundary in the E sector (nearest vegetable garden). s Q =kCf+k2 = 4.72E+2C f g ig g where: C = the' concentration of radionuclide, i, in undiluted gaseous effluent, i in pCi/ml. f = the undiluted effluent flow, in efm 3 kg = conversion factor, 2.83E4 ml/ft k = conversion factor, 6El sec/ min 2 O B-7 Revision 7 1/1/85 \\ i

4 = ..a w a + + ?l. /N - B3.0 RADIATION MONITOR SETPOINTS Using the generic calculations presented in Section 2.0, final effinent radi-l ation monitoring setpoints are calculated for monitoring as required by the Technical Specifications. All radiation monitors for McGuire are off-line except EMF-50 (Waste Gas System) which is in-line. These monitors alarm on low flow; the minimum flow L alarm level for both the liquid monitors and the gas monitors is based on the manufacturer's recommendations. These monitors measure the activity in the liquid or gas volume exposed to the ' detector and are independent of flow rate if a minimum flow rate is assured. Radiation monitoring setpoints calculated in the following sections are expressed in activity concentrations; in reality the monitor readout is in counts per minute. Station radiation monitor setpoint procedures which cor-relate concentration and counts per minute shall be based on the following relationship: c= r 6 2.22 x 10 e V where: c = the gross activity, in pCi/ml r = the count rate, in cpm 2.22 x 108= the disintegration per minute per pCi O e = the counting efficiency, cpm /dpm V = the volume of fluid exposed to the detector, in ml. B3.1 LIQUID RADIATION MONITORS B3.1.1 Waste Liquid Effluent Line As described in Section B2.1.1 on release rate calculations for the waste liquid effluent, the release is controlled by limiting the flow rate of t effluent from the station. Although the release rate is flow rate controlled, the radiation monitor setpoint shall be set to terminate the release if the effluent activity should exceed that determined by laboratory analysis and that used to calculate the release rate. When releases are not being made, a radiation monitor setpoint shall be calculated to assure that release limits are not exceeded. A typical setpoint is calculated as follows: c < MPC x F 1.04E-4 pCi/ml = where: c = the gross activity in undiluted effluent, in pCi/ml f = the flow from the tank may vary from 0-120 gpm but, for this calcu-l lation, is assumed to be 100 gpm (the actual flow is set by flow i selection on the manually loaded throttle valve, a Kerotest globe valve) B-8 Revision 7 1/1/85

(N., MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture \\ / a = 2.4 (See Section B2.1.1) F = the dilution flow is based on having only one condenser cooling water in service or 2.5E+5 gpm. Should the number of pumps in service increase, the setpoint may be recalculated. B3.1.2 Containment Ventilation Unit Condensate Effluent Line As described in Section B2.1.2 on release rate calculations for the contain-ment ventilation unit condensate effluent, it is possible that the effluent l will contain measurable activity above background. It is assumed that no activity is present in the effluent until indicated by radiation monitoring and by routine analyses of the composite sample collected on that line. Since the tank contents can be discharged automatically, a maximum tank concentr-l ation, which also is the radiation monitor setpoint, is calculated to assure that release limits are not exceeded. A typical monitor setpoint and maximum tank concentration is calculated as follows: HPC x F c1 of = 1.04E-4 pCi/ml where: c = the gross activity in undiluted effluent, in pCi/ml (h l ( ) f. = the flow from the tank may vary from 0-120 gpm but, for this calculation,,is assumed to be 100 gpm MPC = 1.0E-07 pCi/mi, the MPC for an unidentified mixture o = 2.4 (See Section B2.1.1) F = the dilution flow is based on having only one condenser cooling pump in service or 2.5E+5 gpm. Should the number of pumps in service increase, the setpoint may be recalculated. The above calculation will determine the maximum setpoint for this release point; releases and/or setpoints may be administratively controlled to assure that release limits are not exceeded since more than one release source may be released to the condenser cooling water. B3.1.3 Conventional Waste Water Treatment System Discharge Line As described in Section B2.1.3 on release rate calculations for the conventional waste water treatment system effluent, the effluent is normally considered non-radioactive; that is, it is unlikely the effluent will contain measurable activity above background. It is assumed that no activity is present in the ef fluent until indicated by radiation monitoring and t> routine analysis of the composite sample collected on that line. Since the system discharges auto-matically, the maximum system concentration, which also is the radiation moni- [~'} tor setpoint, is calculated so that release limits are not exceeded. A typical (_f monitor setpoint and maximum effluent concentration is calculated as follows: B-9 l Revision 7 l 1/1/85 l

/N MPC x F CI 9g,) of = 5.36E-07 pCi/ml where: .c = the gross activity in undiluted effluent, in pCi/ml f = _ the flow rate of undiluted effluent which may vary from 0-6700 gpe, but is assumed to be 6700 gpm MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture F = the flow past. Cowan's Ford Das may vary from 80 to 50,000 cfs, but { is conservatively estimated at 80 cfs (3.59E+04 gpm), the minimum flow available o = 1 [The Conventional Waste Water System discharge line is located down-stream of Cowan's Ford Dam and, therefore, has no reconcentration (recirculation) factor associated with it.] B3.1.4 Turbine Building Sump Discharge Line to the Condenser Cooling Water (CCW) As described in Section B2.1.4 on release rate calculations for the Turbine Building sump effluent, it is possible that the effluent will contain measur-able activity above background. it is assumed that no activity is present in the effluent until indicated by radiation monitoring and by routine analyses of the composite sample collected on'that.line. Since the sump contents can y, be discharged automatically to the CCW, a maximum sump concentration, which also is the radiation monitor setpoint, is calculated to assure that release limits are not exceeded. A typical monitor setpoint and maximum sump concent-ration is calculated as follows: MPC x F

  • I of

= 5.21E-06 pCi/ml where: c = the gross activity in undiluted effluent, in pCi/ml f = the flow rate of undiluted effluent which may vary from 0-2000 spa, but is sssumed to be 2000 spm MPC = 1.0E-07 pCi/m1, the MPC for an unidentified mixture o = 2.4 (See Section B2.1.1) F = the dilution flow is based on having only one condenser cooling pump in service or 2.5E+5 gpm. Should the number of pumps in service increase, the setpoint may be recalculated. The above calculation will determine the maximum setpoint for this release (' point; releases and/or setpoints may be administrative 1y controlled to assure \\, that release limits are not exceeded since more than one release source may be released to the condenser cooling water. B-10 Revision 7 1/1/85

s jq _ B3.2 GAS MONITORS The following equation shall be used to calculate final effluent noble gas l radiation monitor setpoints based on Xe-133: K(X/Q)D < 500 (See section B2.2.1) g s g = 4.72E+2 C f (See Section B2.2.2) Q g Cg < 5.00E+01/f where: C = the gross activity in undiluted effluent, in pCi/ml g f = the flow from the tank or building and varies for various release sources, in cfm 3 K = from Table 1.2-1 for.Xe-133, 2.94E+2 mrem /yr per pCi/m X3 = 7.2E-5 sec/m, as defined in Section B2.2.2. 3 As stated in Section B2.2, the unit vent is the release point for the waste gas systesi, containment purge ventilation system, the containment air release and addition system, the condenser air ejector, and auxiliary building ventilation. Since all of these releases are through the unit vent, the radiation monitor on /m the unit vent may be used to assure that station release limits are not exceeded. For release from the containment air release and addition system and the containment purge ventilation system, a typical radiation monitor setpoint may be calculated as follows: Cg < 5.00E+01/f = 1.79E-03 pCi/ml where: f = 28,000 cfm (containment purge) For release from the containment air release and addition system, the waste gas decay tanks, the condensec air ejectors, and the auxiliary building ventilation system, a typical radiation monitor setpoint may be calculated as follows: Cg < 5.00E+01/f = 3.51E-04 pCi/ml where: f = 142,500 cfm (auxiliary ventilation systems) O) iv B-11 Revision 7 1/1/85

TN B4.0-DOSE CALCULATIONS \\ B4.1 FREQUENCY OF CALCULATIONS Dose contributions to the maximum exposed individual shall be calculated every 31 days, quarterly, semiannually, and annually (as required by Technical Spec-ifications) using the methodology in the generic information sections. This methodology shall.also be used for any special reports. Dose projections shall be performed using simplified estimates. Fuel cycle dose calculations shall be performed annually or as required by special reports. Dose contributions may be calculated using the methodology in the appropriate generic information sections. B4.2 DOSE MODELS FOR MAXIMUM EXPOSED INDIVIDUAL B4.2.1 Liquid Effluents For dose contributions from liquid radioactive releases, one of the two following cases will apply: 1. If the radionuclides Co-58 and/or Co-60 have been detected and Cs-134 and/or Cs-137 have not been detected (i.e., plants without l any fuel failure) dose calculations indicate that the maximum exposed individual would be a child who consumed fish caught in the discharge canal and who drank water from the nearest " downstream" potable , water intake. The dose from these two radionuclides has been cal-culated to be 27% of that individual's total body dose. tx 2. If the radionuclides Cs-134 and/or Cs-137 have been detected, dose calculations indicate that the maximum exposed individual would be an adult who consumed fish caught in the discharge canal and who drank water from the nearest " downstream" potable water intake. The dose from these two radionuclides has been calculated to be 90% of that individual's total body dose. B4.2.2 Gaseous Effluents B4.2.2.1 Noble Gases For dose contributions from exposure to beta and gamma radiation from noble gases, it is assumed that the maximum exposed individual is an adult on the site boundary in each meteorological sectors. B4.2.2.2 Radioiodines, Particulates, and Other Radionuclides T 1/2 > 8 days For dose contributions frc.m radioiodines, particulates and other radionuclides; it is assumed that the maximum exposed individual is an infant who breathes l the air and consumes milk from the nearest goat or cow in each meteorological sector. l B-12 l Revision 7 l 1/1/85 l

/~N . B4.3 - SIMPLIFIED DOSE ESTIMATE .s. B4.3.1 Liquid Effluents For dose estimates, two simplified calculations using the assumptions presented in Section B4.2.1 and source terms presented in the FSAR are presented. Once operational source term data is available, this information shall be used to revise these calculations, if necessary. Case 1 - No Cs-134 or Cs-137 present in effluent. DWB = 5.73E+03 I (F )(T ) (CCo-60 + 0.35 CCo-58) g g f=1 where: BF ) DF,g (3.70) 5.63E+03 = 1.14E+05 (U,,/D,+ U,g g where: 3 1.14E+05 = 10s Ci/pci x 10 ml/kg + 8760 hr/yr p U,, = 510 kg/yr, child water consumption D = 1, dilution factor from the near field area to the nearest potable water intake, Huntersville Water Intake O tj U,g = 6.9 kg/yr, child fish consumption BFg = 5.0E+01, bioaccumulation factor for Cobalt (Table 3.1-1) DF,gg = 1.56E-05, child, total body, ingestion dose factor for Co-60 (Table 3.1-4) 3.70 = factor derived from assumption that 27% of dose is from Co-58 and Co-60 or 100% + 27% = 3.70 m = number of releases And where: I Fg= o F+f where: f = liquid radwaste flow, in gpm o = recirculation factor at equilibrium, 2.4 F = dilution flow, in gpm O B-13 Revision 7 1/1/85

fN And'where: b T = The length'of time, (Thetimeperiodduringwhihfi5$1rhie!!e,s(m)Fkre in hours, over which C ,~C and are averaged. made) C = the average concentration of Co-58 in undiluted effluent, in -58 pCi/ml, during the time period considered. the average concentration of Co-60 in undiluted effluent, in pCi/ml, CCo-60 = during the time period considered. 0.35' = The ratio of the child total body ingestion dose factors for Co-58 and Co-60 or 5.51E-06 + 1.56E Table 3.1-4. Case 2 - Cs-134 and/or Cs-137 present in effluent. D = 6.48E+05 (F )(T ) (CCs-134 + 0.59 CCs-137) WB g g 2=1 where: 6.48E+05 = 1.14E+05 (U,,/D,+ U,g BF ) DF,g (1,10) g where: s 3 1.14E+05 = lo pci/pci x 10 ml/kg + 8760 hr/yr U,, = 730 kg/yr, adult water consumption D = 1, dilution factor frc:a the near field area to the nearest potable water intake, Hunterville Water Intake U,g = 21 kg/yr, adult fish consumption BF = 2.00E+03, bioaccumulation factor for Cesium (Table 3.1-1) g DF,gz = 1.21E-04, adult, total body, ingestion dose factor for Cs-134 (Table 3.1-2) 1.10 = factor derived from the assumption that 90% of dose is from Cs-134 and Cs-137 or 100% + 90% = 1.10 m = number of releases l And where: I FA=F+f where: f = liquid radwaste flow, in gpm f O I ( a = recirculation f actor at equilibrium, 2.4 l' B-14 l-Revision 7 1/1/85 L.

',r' F = dilution flow, in gpm \\ And where: i = The length of time, in hours, over which C C an (Thetimeperiodduringwhib$~afk,rekeadhd,(m)dFarh are averaged. 'made) CCs-134 = the average concentration of Cs-134 in undiluted effluent, in pCi/m1, during the time period considered. he average concentration of Cs-137 in undiluted enluent, in pCi/ml, CCs-137 = during the time period considered. 0.59 : The ratio of the adult total body ingestion dose factors for Cs-134 and Cs-137 or 7.14E-05 + 1.21E-04 = 0.59 B4.3.2 Gaseous Effluents Meteorological data is provided in Tables B4.0-1 and B4.0-2. B4.3.2.1 Noble Gases For dose estimates, simplified dose estimates using the assumptions in B4.2.2.1 and source terms in the FSAR are presented below. Once operational source term data is available, this information shall be used to revise these calculations, if necessary.. These calculations further asseme that the annual average O, dispersion parameter is used and that Xenon-133 contributes 45% of the dose, s D = 8.06E-10 [Q]Xe-133 (

  • p = 2.40E-09 [Q]Xe-133 (
  • D where:

3 -X/Q = 7.2E-05 sec/m, as defined in Section B2.2.2 8.06E-10 = (3.17E-8)(353) (X3 ), derived from equation presented in Section 3.1.2.1. 2.40E-09 = (3.17E-08) (1050) (XTQ),' derived from equation presented in Section 3.1.2.1. s [Q]Xe-133 = the total Xenon-133 activity released in pCi 2.22 = factor derived from the assumption that 45% of the dose is contributed l by Xe-133. B4.3.2.2 Radioiodines, Particulates, and Other Radionuclides with T 1/2 > 8 days + For dose estimates, simplified dose estimates using the assumptions in B4.2.2.2 O and source terms in the FSAR are presented below. Once operational source term l B-15 Revision 7 1/1/85

data is available, this information shall be used to revise these calculations, p) if necessary. These calculations further assume that the annual average i, V dispersion / deposition parameter is used and that 95% of the dose is from Iodine-131 concentrated in goat's milk. The simplified dose estimate to the thyroid of an infant is: D = 1.84E+04 W (Q)I-131 (1.05) where: 2 from Table B4.0-2 W = 2.3E-09 = D/Q for food and ground plane pathway, in m for the location of the nearest real goat (SSE sector at 1.5 miles). r (Q)I-131 = the total Iodine-131 activity released in pCi. =(3.17E-08)(Rf[D3])withtheappropriatesubstitutonsjor 1.84E+04 goat's milk In the grass-cow-milk pathway factor, R [D/Q) for Iodine-131. See Section 3.1.2.2. 1.05 = factor derived from the assumption that 95% of the dose is contributed by I-131. B4.4 FUEL CYCLE CALCULATIONS As discussed in Section 3.3.5, more than one nuclear power station site may contribute to the doses to be considered in making fuel cycle dose assessments lq in accordance with 40CFR190. The fuel cycle dose assessments for McGuire bj Nuclear Station must include dose contributions from Catawba Nuclear Station, which, is located approximately thirty miles SSW of McGuire. For this dose assessment, the maximum exposed individual is conservatively assumed to live 5 miles NNE of Catawba and 5 miles SSW of McGuire. Doses from Catawba's liquid effluents will not affect McGuire since Catawba is downstream of McGuire. The dose contributions resulting from gaseous effluents are calculated using the methodology in Section 3.1.2: D (g) < 0.47 D (g) + 0.55 D (8) f 3 C where: D (g) = Total airborne dose contribution from nuclear power plant operation g to the fuel cycle dose assessment. 3 D (g) = dose contribution from McGuire claculated using X/Q = 1.5E-07 sec/m g and D/Q = 3.8 E-10 m.a. The location is 5 miles SSW of McGuire. 0.47 = fraction of time the wind direction is out of NNE. 3 D (g) = dose g ntribution from Catawba calculated using X/Q = 3.3E-07 sec/m C and D/Q = 5.8E-10 m.a. The location la 5 miles NNE of Catawba. l v B-16 l Revision 7 1/1/85 L

4 i ti i d 2 r f-g 0.55 = fraction of time the wind direction is out of SSW. b Using the methodology above and the assumption that each-station releases their maximum Technical Specification dose limit, the gaseous effluent contri- _bution to the fuel cycle calculation is but a small fraction (< 1/100) of the j allowable dose. Therefore, fuel cycle calculations will not normally be per-formed unless either station exceeds their gaseous effluent Technical Specifi-l cations by a factor of 10. I t In summary, Technical Specification 3.11.4 will be the deciding criteria for i. fuel cycle calculations since it is more restrictive than the case outlined above. t I I ~ q !i-1 i 1 4 r 4 i i 4 J i r 1 F i I ) I i t i 1 1 1 l I L i I f B-17 f-Revision 7 ( 1/1/85 i j. i

\\ TABLE B4.0-1 (1 of 1) MCGUIRE NUCLEAR STATION DISPERSION PARAMETER (X/Q) FOR IDNG TERM RELEASES > 500 HR/YR OR > 125' HR/QTR (sec/m ) l 3 Distance to the control location,,in miles Sector 0.5 1.0 -1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 : S 1.3 E-5 3.4 E-6 1.4 E-6 7.4 E-7 .4.7 E-7 3.3 E-7 2.4 E-7 1.9 E-7 1.5 E-7 1.3 E-7 SSW 1.6 E-5 4.2 E-6 1.7 E-6 9.0 E-7 5.7 E-7 4.0 E-7 3.0 E-7 2.3 E-7 1.9 E-7 1.5 E-7 SW 1.6 E-5 4.2 E-6 1.6 E-6 8.8 E-7 5.6 E-7 3.9 E-7 2.9 E-7 2.2 E-7 1.8 E-7 1.5 E-7 WSW 1.0 E-5 2.8 E-6 1.1 E-6 6.0 E-7 3.7 E-7 2.6 E-7 1.9 E-7 1.5 E-7 1.2 E-7 9.9 E-8 W 4.5 E-6 1.2 E-6 4.7 E-7 2.6 E-7 1.6 E-7 1 1 E-7 8.5 E-8 6.7 E-8 5.4 E-8 4.5 E-8 WNW 4.0 E-6 1.1 E-6 4.2 E-7 2.3 E-7 1.4 E-7 1.0 E-7 7.4 E-8 5.8 E-8 4.6 E-8 3.8 E-8 NW 9.0 E-6 2.4 E-6 9.7 E-7 5.3 E-7 3.3 E-7 2.3 E-7 1.7 E-7 1.4 E 1.1 E-7 9.2 E-8 NNW 1.4 E-5 3.6 E-6 1.5 E-6 8.1 E-7 5.2 E-7 3.7 E-7 2.8 E-7 2.2 E-7 1.8 E-7 1.5 E-7 N 5.8 E-5 1.5 E-5 6.1 E-6 -3.4 E-6 2.2 E-6 1.6 E-6 1.2 E-6 9.6 E-7 7.8 E-7 6.6 E-7 NNE 7.2 E-5 1.8 E-5 7.5 E-6 4.2 E-6 2.8 E-6 2.0 E-6 1.5 E-6 1.2 E-6 9.7 E-7 8.1 E-7 NE 4.0 E-5 1.0 E-5 4.2 E-6 2.3 E-6 1.5 E-6 1.1 E-6 8.2 E-7 6.5 E-7 5'.3 E-7 4.5 E-7 ENE 2.3 E-5 5.9 E-6 2.5 E-6 1.4 E-6 9.0 E-7 6.4 E-7 4.8 E-7 3.8 E-7 3.1 E-7 2.i6 E-7 E 1.8 E-5 4.6 E-6 1.9 E-6 1.1 E-6 6.9 E-7 4.9 E-7 3.7 E-7 2.9 E-7 2.4 E-7 2.0 E-7 ESE 1.2 E-5 3.2 E-6 1.3 E-6 7.4 E-7 4.8 E-7 3.4 E-7 2.5 E-7 2.0 E-7 1.6 E-7 1.4 E-7' SE 1.1 E-5 2.9 E-6 1.2 E-6 6.6 E-7 4.2 E-7 3.0 E-7 2.2 E-7 1.8 E-7 1.4 E 1.2 E-7 SSE 7.7 E-6 2.1 E-6 8.5 E-7 4.6 E-7 ~3.0,E-7 2.1_E-7 1.5 E-7 1.2 E-7 9.9 E-8 8.2 E-8 Revision 7 1/1/85 + t g s -

,J3 + (- L i ~ 'iABLE B4.0-2 ) ~' x ') (l 'of 1) g MCGUIRE NUCLEAR STATION 0 DEPOSITION PARAMETER (D/Q) FOR LDNG TERN RELEASES > 500 HR/YR OR > 125 HR/QTR (m 2) l Distance to the control location, in miles Sector 0.5 1.0 t 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 S 4.9 E-8 1.2 E-8 4.3 E-9 2.1 E-9 1.3 E-9 8.2 E-10 5.8 E-10 4.3 E-10 3.3 E-10 2.6 E-10 SSW 7.1 E-8 1.7 E-8 6.2 E-9 3.1 E-9 1.8 E-9 1.2 E-9 8.3 E-10 6.2 E-10 4.8 E-10 3.8 E-10 SW 9.4 E-8 2.3 E-8 8.2 E-9 4.1 E-9 2.4 E-9 1.6 E-9 1.1 E-9 8.2 E-10 6.3 E-10 5.0 E-10 WSW 5.3 E-8 1.3 E-8 4.7 E-9 2.3 E-9 1.4 E-9 8.9 E-10 6.3 E-10 4.7 E-10 3.6 E-10 2.9 E-10 W 1.3 E-8 3.1 E-9 1.1 E-9 5.6 E-10 3.3 E-10 2.1 E-10 1.5 E-10 1.1 E-10 8.6 E-11 6.9 E-11 WNW 1.1 E-8 2.7 E-9 9.8 E-10 4.9 E-10 2.9 E-10 1.9 E-10 1.3 E-10 9.8 E-11 7.6 E-11 6.0 E-11 NW 1.9 E-8 4.7 E-9 1.7 E-9 8.4 E-10 5.0 E-10 3.2 E-10 2.3 E-10 1.7 E-10 1.3 E-10 1.0 E-10 NNW 2.3 E-8 5.7 E-9 2.1 E-9 1.0 E-9 6.0 E-10 3.9 E-10' 2.8 E-10 2.0 E-10 1.6 E-10 1.3 E-10 N 9.3 E-8 2.3 E-8 8.1 E-9 4.0 E-9 2.4 E-9 1.6 E-9 1.1 E-9 8.1 E-10 6.3 E-10 5.0 E-10 NNE 1.3 E-7 3.2 E-8 1.1 E-8 5.7 E-9 3.3 E-9 2.2 E-9 1.5 E-9 1.1 E-9 8.8 E-10 7.0 E-10 NE 7.1 E-8 1.7 E-8 6.2 E-9 3.1 E-9 1.8 E-9 1.2 E-9 8.3 E-10 6.2 E-10 4.8 E-10 3.8 E-10 ENE 3.8 E-8 9.3 E-9 3.3 E-9 1.7 E-9 9.8'E-10 6.4 E-10 4.5 E-10 3.3 E-10 2.6 E-10 2.0 E-10 E 3.0 E-8 7.3 E-9 2.6 E-9 1.3 E-9 7.6 E-10 5.0 E-10 3.5 E-10 2.6 E-10 2.0 E-10 1.6 E-10 ESE 3.0 E-8 7.4 E-9 2.7 E-9 1.3 E-9 7.8 E-10 5.1 E-10 3.6 E-10 2.6 E-10 2.0 E-10 1.6 E-10 SE 3.1 E-8 7.6 E-9 2.7 E-9 1.3 E-9 7.9 E-10 5.2 E-10 3.7 E-10 2.7 E-10 2.1 E-10 1.7 E-10 SSE 2.7 E-8 6.5 E-9 .2.3 E-9 1.2 E-9 6.8 E-10 4.5 E-10 3.1 E-10 2.3 E-10 1.8 E-10 1.4 E-10 Revision 7 1/1/85

.o l /'~} TABLE B4.0-3* l \\s_/ } (1 of 3) i l MCGUIRE NUCLEAR STATION ADULT A,1 DOSE PARAMETERS f (mrem /hr per pCi/ml) l NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LII l H 3 0.0 8.96E+00 8.96E+00 8.96E+00 8.96E+00 8.96E+00 8.96E+00 { NA 24 5.48E+02 5.48E+02 5.48E+02 5.48E+02 5.48E+02 5.48E+02 5.48E+02 l CR 51 0.0 0.0 1.49E+00 8.94E-01 3.29E-01 1.98E+00 3.76E+02 MN 54 0.0 4.76E+03 9.08E+02 0.0 1.42E+03 0.0 1.46E+04 t l MN 56 0.0 1.20E+02 2.12E+01 0.0 1.52E+02 0.0 3.82E+03 = FE 55 8.87E+02 6.13E+02 1.43E+02 0.0 0.0 3.42E+02 3.52E+02 FE 59 1.40E+03 3.29E+03 1.26E+03 0.0 0.0 9.19E+02 1.10E+04 1 L C0 58 0.0 1.51E+02 3.39E+02 0.0 0.0 0.0 3.06E+03 CO 60 0.0 4.34E+02 9.58E+02 0.0 0.0 0.0 8.16E+03 NI 63 4.19E+04 2.91E+03 1.41E+03 0.0 0.0 0.0 6.07E+02 NI 65 1.70E+02 2.21E+01 1.01E+01 0.0 0.0 0.0 5.61E+02 ((N_,) ZN 65 2.36E+04 7.50E+04 3.39E+04 0.0 5.02E+04 0.0 4.73E+04 CU 64 0.0 1.69E+01 7.93E+00 0.0 4.26E+01 0.0 1.44E+03 ZN 69 5.02E+01 9.60E+01 6.67E+00 0.0 6.24E+01 0.0 1.44E+01 BR 83 0.0 0.0 4.38E+01 0.0 0.0 0.0 6.30E+01 BR 84 0.0 0.0 5.67E+01 0.0 0.0 0.0 4.45E-04 BR 85 0.0 0.0 2.33E+00 0.0 0.0 0.0 0.0 RB 86 0.0 1.03E+05 4.79E+04 0.0 0.0 0.0 2.03E+04 RB 88 0.0 2.95E+02 1.56E+02 0.0 0.0 0.0 4.07E-09 RB 89 0.0 1.95+02 1.37E+02 0.0 0.0 0.0 1.13E-11 SR 89 4.78Et04 0.0 1.37E+03 0.0 0.0 0.0 7.66E+03 SR 90 5.95E+05 0.0 1.60E+05 0.0 0.0 0.0 3.40E+04 SR 91 8.79E+02 0.0 3.55E+01 0.0 0.0 0.0 4.19E+03 SR 92 3.33E+02 0.0 1.44E+01 0.0 0.0 0.0 6.60E+03 Y 90 1.38E+00 0.0 3.69E-02 0.0 0.0 0.0 1.46E+04 Y 91M 1.30E-02 0.0 5.04E-04 0.0 0.0 0.0 3.82E-02 Y 91 2.02E+01 0.0 5.39E-01 0.0 0.0 0.0 1.11E+04 Y 92 1.21E-01 0.0 3.53E-03 0.0 0.0 0.0 2.12E+03

  • Table provided by:

M. E. Wangler, RAB:NRR:NRC on 2/24/83. TABLE B4.0-3 (1 of 3) Revision 7 1/1/85

TABLE B4.0-3 (2 of 3) MCGUIRE NUCLEAR STATION ADULT A,1 DOSE PARAMETERS (mrem /hr per pCi/ml) l NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LII l Y 93 3.83E-01 0.0 1.06E-02 0.0 0.0 0.0 1.22E+04 ZR 95 2.77E+00 8.88E-01 6.01E-01 0.0 1.39E+00 0.0 2.82E+03 ZR 97 1.53E-01 3.09E-02 1.41E-02 0.0 4.67E-02 0.0 9.57E+03 NB 95 4.47E+02 2.49E+02 1.34E+02 0.0 2.46E+02 0.0 1.51E+06 l MO 99 0.0 4.62E+02 8.79E+01 0.0 1.05E+03 0.0 1.07E+03 TC 99M 2.94E-02 8.32E-02 1.06E+00 0.0 1.26E+00 4.07E-02 4.92E+01 l l TC 101 3.03E-02 4.36E-02 4.28E-01 0.0 7.85E-01 2.23E-02 1.31E-13 i RU 103 1.98E+01 0.0 8.54E+00 0.0 7.57E+01 0.0 2.31E+03 RU 105 1.65E+00 0.0 6.52E-01 0.0 2.13E+01 0.0 1.01E+03 RU 106 2.95E+02 0.0 3.73E+01 0.0 5.69E+02 0.0 1.91E+04 AG 110M 1.42E+01 1.31E+01 7.80E+00 0.0 2.58E+01 0.0 5.36E+03 TE 125M 2.79E+03 1.01E+03 3.74E+02 8.39E+02 1.13E+04 0.0 1.11E+04 ~~ 4 () TE 127M 7.05E+03 2.52E+03 8.59E+02 1.80E+03 2.86E+04 0.0 2.36E+04 TE 127 1.14E+02 4.11E+01 2.48E+01 8.48E+01 4.66E+02 0.0 9.03E+03 TE 129M 1.20E+04 4.47E+03 1.89E+03 4.11E+03 5.00E+04 0.0 6.03E+04 TE 129 3.27E+01 1.23E+01 7.96E+00 2.51E+01 1.37E+02 0.0 2.47E+01 TE 131M 1.80E+03 8.81E+02 7.34E+02 1.39E+03 8.92E+03 0.0 8.74E+04 TE 131 2.05E+01 8.57E+00 6.47E+00 1.69E+01 8.98E+01 0.0 2.90E+00 TE 132 2.62E+03 1.70E+03 1.59E+03 1.87E+03 1.63E+04 0.0 8.02E+04 I 130 9.01E+01 2.66E+02 1.05E+02 2.25E+04 4.15E+02 0.0 2.29E+02 I 131 4.96E+02 7.09E+02 4.06E+02 2.32E+05 1.22E+03 0.0 1.87E+02 I 132 2.42E+01 6.47E+01 2.26E+01 2.26E+03 1.03E+02 0.0 1.22E+01 I 133 1.69E+02 2.94E+02 8.97E+01 4.32E+04 5.13E+02 0.0 2.64E+02 I 134 1.26E+01 3.43E+01 1.23E+01 5.94E+02 5.46E+01 0.0 2.99E-02 I 135 5.28E+01 1.38E+02 5.10E+01 9.11E+03 2.22E+02 0.0 1.56E+02 CS 134 3.03E+05 7.21E+05 5.89E+05 0.0 2.33E+05 7.75E+04 1.26E+04 CS 136 3.17E+04 1.25E+05 9.01E+04 0.0 6.97E+04 9.55E+03 1.42E+04 CS 137 3.88E+05 5.31E+05 3.48E+05 0.0 1.80E+05 5.99E+04 1.03E+04 CS 138 2.69E+02 5.31E+02 2.63E+02 0.0 3.90E+02 3.85E+01 2.27E-03 BA 139 9.00E+00 6.41E-03 2.64E-01 0.0 5.99E-03 3.64E-03 1.60E+01 OV TABLE B4.0-3 (2 of 3) Revision 7 1/1/85

TABLE B4.0-3 ,-s (3 of 3) MCGUIRE NUCLEAR STATION ADULT A DOSE PARAMETERS g (mrem /hr per pCi/ml) l NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LII BA 140 1.88E+03 2.37E+00 1.23E+02 0.0 8.05E-01 1.35E+00 3.88E+03 BA 141 4.37E+00 3.30E-03 1.48E-01 0.0 3.07E-03 1.87E-03 2.06E-09 BA 142 1.98E+00 2.03E-03 1.24E-01 0.0 1.72E-03 1.15E-03 2.78E-18 LA 140 3.58E-01 1.80E-01 4.76E-02 0.0 0.0 0.0 1.32E+04 LA 142 1.83E-02 8.33E-03 2.07E-03 0.0 0.0 0.0 6.08E+01 CE 141 8.01E-01 5.42E-01 6.15E-02 0.0 2.52E-01 0.0 2.07E+03 CE 143 1.41E-01 1.04E+02 1.16E-02 0.0 4.60E-02 0.0 3.90E+03 CE 144 4.18E+01 1.75E+01 2.24E+00 0.0 1.04E+01 0.0 1.41E+04 PR 143 1.32E+00 5.28E-01 6.52E-02 0.0 3.05E-01 0.0 5.77E+03 PR 144 4.31E-03 1.79E-03 2.19E-04 0.0 1.01E-03 0.0 6.19E-10 ND 147 9.00E-01 1.04E+00 6.22E-02 0.0 6.08E-01 0.0 4.99E+03 W-187 3.04E+02 2.55E+02 8.90E+01 0.0 0.0 0.0 8.34E+04 ,,_s NP 239 1.1.8E-01 1.25E-02 6.91E-03 0.0 3.91E-02 0.0 2.57E+03 ss l N TABLE B4.0-3 (3 of 3) Revision 7 1/1/85

B5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING The radiological environmental monitoring program shall be conducted in accor-dance with Technical Specification 3/4.12. The monitoring program locations and analyses are given in Tables B5.0-1 through B5.0-3 and Figure B5.0-1 Site specific characteristics make groundwater sampling unnecessary. Groundwater recharge is from Lake Norman and local precipitation. The groundwater gradient flows directly to the Catawba River; therefore, contamination of groundwater from liquid effluents is highly improbable. Additionally, two site boundary TLD locations in the NW and NNW sectors do not exist since the required loc-ations are over water. The laboratory performing the radiological environmental analyses shall parti-cipate in an interlaboratory comparison program which has been approved by the NRC. This program is the Environmental Protection Agency's (EPA's) Environmental Radioactivity Laboratory Intercomparison Studies (Crosscheck) Program, our participation code is CP. O B-18 Revision 7 1/1/85

1 %.) ~G TABLE B5.0-1 (1 of 1) MCGUIRE RADIOLOGICAL MONITORING PROGRAM SAMPLING LOCATIONS (TLD LOCATIONS) i SAMPLING LOCATION DESCRIPTION

  • SAMPLING LOCATION DESCRIPTION
  • 143 SITE BOUNDARY (0.5 MILES NW) 163 4-5 MILE RADIUS (5.0 MILES SE) 144 SITE BOUNDARY (0.6 MILES NNE) 164 4-5 MILE RADIUS (4.5 MILES SSE) 145 SITE BOUNDARY (0.5 MILES NE) 165 4-5 MILE RADIUS (5.0 MILES S) 146 SITE BOUNDARY (0.5 MILES ENE) 166 4-5 MILE RADIUS (5.2 MILES SSW) 147 SITE BOUNDARY (0.5 MILES E) 167 4-5 MILE RADIUS (4.9 MILES SW) 148 SITE BOUNDARY (0.5 MILES ESE) 168 4-5 MILE RADIUS (4.7 MILES WSW) 149 SITE BOUNDARY (0.7 MILES SE) 169 4-5 MILE RADIUS (4.4 MILES W) 150 SITE BOUNDARY (0.5 MILES SSE) 170 4-5 MILE RADIUS (4.4 HILES WNW) 151 SITE BOUNDARY (0.5 MILES S) 171 4-5 MILE RADIUS (4.5 MILES NW) 152 SITE BOUNDARY (0.5 MILES SSW) 172 4-5 MILE RADIUS (5.2 MILES NNW) 153 SITE BOUNDARY (0.5 MILES SW) 173 SPECIAL INTEREST (8.5 MILES NNW) 154 SITE BOUNDARY (0.7 MILES WSW) 174 SPECIAL INTEREST (8.7 MILES WNW) 155 SITE BOUNDARY (0.7 MILES W) 175 SPECIAL INTEREST (12.7 MILES WNW) 156 SITE BOUNDARY (0.5 MILES WNW) 176 SPECIAL INTEREST

-(11.0 MILES SW) 157 4-5 MILE RADIUS (4.8 MILES N) 177 SPECIAL INTEREST (8.6 MILES S) 158 4-5 MILE RADIUS (4.4 MILES NNE) 178 SPECIAL INTEREST (9.2 MILES SE) 159 4-5 MILE RAL IUS (5.0 MILES NE) 179 SPECIAL INTEREST (10.4 MILES ESE) i 160 4-5 MILE RADIUS (4.9 MILES ENE) 180 SPECIAL INTEREST (11.5 MILES NNE) 161 4-5 MILE RADIUS (4.7 MILES E) 181 SPECIAL INTEREST (6.7 MILES NE) 162 4-5 MILE RADIUS (4.6 MILES ESE) 182 SPECIAL INTEREST (6.0 MILES NE) 183 CONTROL (5.5 MILES S)

  • All TLD samples are collected quarterly l

1

75 8 n/ goH B5 >hQem M M M M M o1 i/ s1 \\)v i M M M M v xdx S S S S e R ( A A =5k S S g mhAh A8k M A A A $h ggW8m S S S e*E%A W W W W B B B B S ) N e O c I a T l A p C e O h3 $ $m W W W r L B B B l G l N i I w L P 0 M 2 A) h Hk* eli! 1 E / SS / N 0mhSg d4 W W W W W n MO AI o RT i GA t 2 OC m 5 c 0 O d l oO X X X X X a RO c PL o

0. )1 l

5 GG ) E BfNN s oI I N E RL E 5 L1 OP ) ) 8 4 B(TM i E E / A I A m N ) N 1 T NS E E N / O 6 S ) 1 i S W i ) MRE 0 m N m E d LH ( i W N e AT 0 m 5 t CO n i ) i e I( a ) 3 2 m N 2 ) m l G m W ( 1 E e O r S 1 7 i ( S 5 d L ) o y 1 m S O N E N l l ( 8 y 7 e I O N i p ( 0 l i ( b D y I E e m p y p) m A y l T k u l l 2 p N ) y o R l l P i a 6 S p o 1 u E 8 l T hya I m L p o ( S i S p ( E tl u R 0 r u h m E 1 ) ) p ) R nrn C 4 o ( e S c l r ( E W u ) E I oen S ) t t S a e 0 i S S E N U mt a E E ) 0 m a r n t m s S i N N n G iri D N E )W( e a W e h a a 2 t m r o C mam N N S c e t g C W 1 a i e i i s s 8 o m 3 t m m n a M eue N e n r l a i ( SQS O i i i g a t a W ) H e l 1 G a e e (7 W 0 4 d s I m m m d r s i E k a a ( 8 r T i t n p l N r a p e e A 7 5 5 r n w i a o t i r s c 2 l 5 4 a t MQA C B E o c p i i n c A w n ( s a ( ( G s S S O 0 0 0 D n i m n I i o e w p e L ((( l l u c u n s C d s o i s s v a a e H i 2 J l u s i w C c u u r G y y y n n g n l M e y s o i i i s a N r r r a a d e u 6 n a c r e C y n d d u h I a a a C C i l M ( l h e c i R r u a a i L d d d r l o s l A a y i M R R d g P n n n e e B i e s c r l D l r a a n y M u u u g g v t u n a i e l i D n e e R i l y A o o o r r 3 s t i i M v l d e a o l l r ykl S B B B a a 7 r o l L s c r w D r s i i e u l eh h h e l e t e a a o e d M M l d t r n t n r n b H d v i i k et e e e c c ewn t t t s s y n a r s a o n b d e v 5 5 M ) eio i i i i i w u h o a l o i u C i e a a WBM S S S D D H H C C E P M P H R K K D 4 4 5 ( /. E D m O WWM k C B 1 2 8 9 0 0 1 5 8 9 0 1 2 3 4 5 6 7 8 9 0 2 2 2 2 2_3 3 3 3 3 3 3 3 3 3 4 4 4 5 5 6 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1_1

TABLE B5.0-3 '(1 of 1) MCGUIRE RADIOLOGICAL MONITORING PROGRAM ANALYSES ANALYSES SAMPLE MEDIUM ANALYSIS SCHEDULE GAMMA ISOTOPIC TRITIUM LOW LEVEL I-131 GROSS BETA TLD 1. Air Radioiodine and X Particulates Weekly X X 2. Direct Radiation Quarterly X 3. Surface Water Biweekly X Monthly Composite X X Quarterly Composite X 1 i 4. Drinking Water Biweekly X Monthly Composite X X Quarterly Composite X 5. Shoreline Sediment Semiannually X 4 i 6. Milk Semimonthly X X i l } 7. Fish Semiannually X 8. Broadleaf Vegetation Monthly X s Monthly (* I 9. Food Products X l (a) during harvest season Revision 7 1/1/85 m

J}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20127M496&oldid=4264469"