ML20012A377

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Rev 26 to, Offsite Dose Calculation Manual McGuire Nuclear Station.
ML20012A377
Person / Time
Site: Mcguire, McGuire, 05000000
Issue date: 01/01/1990
From: Birch M, Mcconnell T
DUKE POWER CO.
To:
Shared Package
ML15217A104 List:
References
PROC-900101-01, NUDOCS 9003090348
Download: ML20012A377 (41)
v • d • e


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, )[ DAcemberi27, 1989

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Subject:

McCulre-Nuclear Station Offsite 90se Calculation Manual ,

, RevisionR26 The General Office-Radiation Protection Staff is transmitting to you this date- Revision 26-of the Offsite Dose Calculation Manual. As this revision only'affects-McGuire Nuclear Station, the approval of-other station .

manegers'is_ pot required. Please update your copy No. 8D , and discard l the affected pages.

HEMOVE THESE PAGES INSERT THESE PAGES -

Figure 81.0-1 Rev. 06 Figure B1.0-1 Rev. 26 Figure B1.0-2 Rev. 06 Figure B1.0-2 Rev. 26 (1 of 3; Figuie B1.0-2 Rev. 26 (2 of 3) .

Figure'B1.0-2 Rev. 06 Figure B1.0-2 Rev. 26 (2 of 2) (3 of 3) >

-B-6 Rev. 22 B-6 Rev. 26 B-7 Rev. 22 B-7 Rev. 26 B-11a Rev. 16 B-11a Rev. 26 B-12 Rev. 22 B-12 Rev. 26 d . (gg . B-13 Rev. 22 B-13 Rev. 26

\e) B-14 Rev. 22 B-14 Rev. 26 B-15 Rev. 22 B-15 Rev, 26 B-16 Rev. 22 B-16 Rev. 26 D-17 Rev. 22 B-17 Rev. 26 '

B-18 Rev. 22 'B-18 Rev. 26 B-19 Rev. 22 B-19 Rev. 26 8--20 Rev. 22 B-20 Rev. 26 Table 85.0-1 Rev. 07 Table 85.0-1 Rev. 26L NOTE: As this letter, witn it's attachments, contains "LOEP." information,

.please insert this letter in front of the December 26, 1989 letter.

Approva1.Date: 12/21/89 Approvai Date: 12/21/89 Effective Date: 1/t/90 Effective Date: 1/1/90 t

. (AAy Mary L. Bi7ch

} A' T.L. McConnell, Manager

  • Radiation Protection. Manager McGuire Nuclear Station-

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'If you have-any questions concerning Revision 26, please call Jim Stewart at.(704)'373-5444 ,

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James'M. Stewart, Jr.

, , Scientist-

~Radiat' ion-Protection 9003090348 900220 1

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JUSTIFICATIONS FOR REVISION 26-Figure B1'.0-1 Replaced f i g u r e ., w i t h CAD drawnng, l No changes made. 4 Figure B1.0-2 Replaced. figures with CAD: drawings.  ;

Split single station d r a w i n g i s.n t o = i separate unit drawings for c l a r e t y '. +

and added information on: changes ti n; flow-rates per NSM-1202  ;

4 Provided additional filter and-flow i rate information on drawings for 'I clarity purposes.

Pages B-6 and B-7~ Updated sections using dose  :

calculations based'on 1989^ Effluent' l Release Data-(first nine months) and ,

the 1989 Land Use Census Data. 4 Page B-11a Changed name from Contaminated Parts i Warehause (EMF-53) to Waste Handling-Area (EMF-53) 'and changed flow. rate to~

s agree-changes madenper7 NSM-1202;

?'N . Pages B-12 thru B-20 Updated acctions using dose

( ). calculations based on 1989: Effluent Release Data (f6 rat nine months) and the 1989: Land'Use Census' Data. '

Page B-21  ; Updated-the dates the latest ,

Land;Use Census was preformed'. i e

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Table'B5.0-1' Changed per attached January 27, 1.989-letter from J W Foster. '

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3, O M.L.: Birch -

Is Radiation Protection Manager =

r ' Nuclear Production-

. . Duke Power. Company .

'I Wachovia Center.

' Charlotte, N.C. ,

Attention: Jim Stewart "

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Subject:

McGuire Nuclear Station

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ODCM Changes f I YM, During the McGuire Nuclear Station REMP Review,Senvironmental TLD

, locations were examined. The control location (#183) for TLD's-

(direct radiation) is not,inlthe.most: desirable location. McGuire LIlealth Physics-Sections requests that location #175 be designated las.the' control:. location effective for 1989 measurements. ' Location

.#175_has!a much. lower D/Q value and-is located at a'more appropriate distance from McGuire-NuclearLStation.: Location #183

-will be designated'as a special-interest location so as'to i L mairitain the ~ total number of ' sites being' monitored at' forty. .,

!e Changes will also be necessary in' selected ASC procedures and reports. ._By copy of this letter, ASC personnel are requested to y make-these changes. J f .. .

Q ,The necessary 0DCM changes are' marked on the attached sheet. t 1

[ cfM J. W. Foster Radiation Protection Manager McGuire Nuclear Station j'N . 'RJS/plb a Q,)';. .

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cc: P '. m Wingo J

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APPENDIX B a

MCGUIRE NUCLEAR STATION' SITE SPECIFIC INFORMATION l r

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N ,'~'k APPENDIX B -.TIBLE OF CONTENTS.

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Int Bl~0'MCGUIRE NUCLEAR ~ STATION RADWASTE SYSTEMS B-1 n .,.

B2.0 RELEASE RATE CALCULATION B-4 ,

k ~B3.0' RADIATION MONITOR SETPOI!frS B-8, iB4.0 DOSE CALCULATIONS ~ .B-12'

=B5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING B-21 s

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, Bl.1 MCGUIRE NUCLEAR STATION RADWASTE SYSTEMS'

. LIQUID RADWASTE PROCESSING J IThe liquid radwaste system at McGuire Nuclear Station (MNS) is used to collect j 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: .,

3. '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 andLfree 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 L solution, as in.the case of either LHST or FDT' waste in the WL System Ion exchange lM. _

after removal of organics-by carbon filtration (adsorption).

y is also used'in removing both cations (cobalt, manganese) and anions l G'~

- (chlorida, fluoride) from evaporator distillates in order to 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 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 l 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

p .F.igure Bl.0-1 is a schematic representation of the liquid radwaste system at McGuire.

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, -\ J  ; ABBREVIATIONS:

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Systems:-  !

CM'- Condensat'e Cooling I w

>KCE- ' Component Cooling. .

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! . NC' . Reactor Coolant  ;

  • WC - Conventional Waste Water-Treatment i WG - Wasta Gas:  :

i WL'- Liquid Waste'.  ;

I 'WM .' Liquid Waste Monitor.and Disposal l Terms: j BA - Boric Acid Tank e RC'- Condenser Cooling Water CDT ' Chemical Drain Tank ~

ECST - Evaporator Concentrates Storage Tank L '['

1 FDT --Floor Drain Tank

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. .FWST - Fueling Water Storage Tank (formerly Refueling Water Storage Tank) 1 LHST - Laundry and Hot Shower Tank- j u MST - Mixing and Settling Tank

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s ,NCDT -~ Reactor Coolant' Drain Tank RBT - Resin Batching Tank r,

,RHT --Recycle Holdup Tank 4 - '-w RPrr Recycle Monitor Tank- >

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,RMWST -; Reactor Makeup Water Storage Tank

[' SRST.- Spent Rosin-Storage Tank j .?

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VUCDT'- Ventilation Unit Condensate Drain Tank b WDT Waste Drain Tank l WEFT . Waste Evaporator Feed Tank'

$ WMT - Waste Monitor Tank p: , ,

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./ \t Bl.2 : GASEOUS'RADWASTE SYSTEMS d

4' The gaseous waste disposal system for-McGuirn is designed with the capability

  • of processing the fission-product gases from contaminated reactor coolant J

, ' fluids resulting' from operation. The design base for the system shown schema-l n tically in Fig.,Bl.0-2-is the retention, through the plant lifetime, of all the l p gaseous fission products to be discharged from the reactor coolant syntes to

- the chemical and volume control system and other plant systems to eliminate the

,, 7 need for intentional discharge of radioactive gases from the waste gas holdup ,

tanks. . : Actual system operation'is aimed at. maximizing storage time for decay l

prior to infrequent releases. Unavoidable sources of low-level radioactive 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 defecto 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 11he gas collection system combines the waste hydrogen and fission gases fros' f'% Ethe' volume' control tanks,-the boron recycle.and liquid waste-gas stripper.

1 evaporators, and other sources produced during normal operation or the gas collected during the: shutdown degasification (high percentage of nitrogen) and cycles it through the catalytic recombiners:to. convert the hydrogen to water.

After the water vapor-is removed, the.resulting gas stream is transferred from the recombinerzinto the gas decay tanks, where the accumulated activity may be r

' 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 s Nonrecyclable reactor coolant leakage occurring either inside the containment or insido'the auxiliary building will. generate gaseous activity. Gases result-ing _from leakage inside the containment will be contained until the containment Jis purged. The containment atmosphere will be circulated-throughia charcoal

.adsorber and a. particulate filter' prior to release to the atmosphere.

Gases resulting from-leakage inside the auxiliary building are released, with- '"

H 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.

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. Bl.2.3 Secondary Systems

The gases removed from'the secondary system by the air ejectors are dischargsd=

' to the unit vent, If the secondary system contains activity, the steam generator blowdown may be'either. discharged directly"to-the RC system or-through 'demineralizers to reduce activity levels.

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 radweste system at McGuire, n.

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B-3 Revision 14 1/1/87

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THE FILMR MODE.

    • P01DillALLY CONTAlmMA1ED AREAS OF THE AUXIJARY BURAING ARE N0hWALLY UNF1.10tED. UPON A RA0lA110N ALARW SY DIF-41. THE EXHAUST WILL BE OlVUt1ED TO THE FIL10tED MODE.

REVISION 26 nouRE si.0-2 1/1/go McGulRE NUCLEAR STAfl0N oAmous RA0was1E SYrn PACE 1 or s N89121 A .

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'**POTENDALLY CONTAMINA1ED AREAS OF THE AU11UARY SUILDiNO ARE NORWALLY UNFILTUtED. UPON A RADIAfl0N ALMtM SY OdF-41, THE EXHAUST veu. OE Dist1ED TO THE FIL1UtED WOOL REVISION 26 e nouRE ei.0-2 wrouwit NUCLEAR STATION 1/1/90 c?

CASE 0uS RA0 waste SYSTEM

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.. GASEOUS WASTE SYSTEM REVISION 26 McGUIRE NUCLEAR. STATION ~ 1/1/90

. FIGURE 81.0-2 PAGE 3 OF 3 ~ N891210 -

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'D B2.0 RELEASE RATE CALCULA, TION

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Generic release rate calculations are pre.sented in'Section 1.0; these calcu---

Llations 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 system; '

the third-release point, tho' Turbine Building s. ump, can either be discharged

into the condenser cooling water system or into the conventional waste water treatment system.

7

.B2.1.1 Waste Liould Effluent =Line d i For releases made via the waste liquid effluent line, the following calculation-

  • shall-.be performed to determine discharge flow, in spe:

n f<F+(o I C 1

) .:

MPC ,

1

- where:

'g f = the undiluted effluent flow, in spa.

1 F = the dilution flow available depending on the number (1-8) of con '

denser cooling water (RC) pumps in service, in gpm.

where

F = (2.50E+05 gpm/ pump)(Number of: RC pumps in service) o = The' recirculation factor at equilibrium (dimensionless), 2.4 R ,1, a=1+

60 = 2.4 .,

OH where:

Q = average dilution flow (3720 cfs)

R Q

H.

" average fl W Past Cowans Ford Dam (2670 cfs)

Cg = the concentration of radionuclide, "i", in undiluted effluent as determined by laboratory analyses, in pC1/ml.

JMPC the concentration of-radionuclide, "i",..from 10CFR20, Appendix B, p= Table II, Column 2. If radionuclide, "i", is a dissolved noble 1 gas, the MPC = 2.00E-04 pCi/ml.

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,yN 1B2.1.2 Containment Ventilation Unit Condensate Effluent-Line l 1

t V ' d -The containment. ventilation. unit.condens' ate effluent line normally contains measurable activity above background and administrative controls.have been- .I

' 6 implemented-tor assureithat release ' limits are not exceeded; ..see1section 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 l 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. Radiation monitoring ,

Lalara/ trip.setpoints assure that release limits are not exceeded; see section  :

on radiation monitoring alarm / trip setpoints. ]

B2.1.4 Turbine Buildina Sumo Discharme Line

' Normally the discharge' from the Turbine Building sump is considered non- '

radioactive; that is, it is unlikely the effluent will contain measurable activity :above background, and will flow. into~the-conventional waste water-treatment system. It is assumed that no activity is present in the: effluent Lunti1Jindicated by radiation monitoring measurements. If measurable. activity.is  ;

present in the effluent, sump discharge-will be-terminated and an alarm activated.

At this time tho' discharge may be routed to the floor drain tank-for. processing i p (

[. 1 or routed ~.directly to the condenser cooling water (RC) 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.

B2.2 GASEOUS RELEASE RATE CALCULATIONS

~

The unit vent 1s the release point for waste gas decay tanks, containment 0; -building purges, the: condenser air ejector, and auxiliary building r

j. ventilation.' The condenser air ejector effluent is normally considered l

-nonradioactive; that is, it is unlikely the effluent will contain measurable  ;

activity above background. It is assumed that no activity is present in the j? condenser air ejector effluent until indicated by radiation monitoring  ;

l l - measurements and by analyses of periodic . samples collected on -that line.

F Radiation. monitoring alarm / trip setpoints in conjunction with administrative controls; assure-that 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 limit the release rate for a single r l  ;

release point.  !

1 B-5 Revision 14 1/1/87

). - ~ _ _ . - . . _ - - - _ - - - - - - - - - - - - , w

~

@M; g 77c. 4 17 c  ?* -

, I t i 1 ,

ll } ( ,

a l:m p j P p M 3 2. 2.~1 J Noble Gases 1 NA <-

L y , I (K 1:[(X/Q)Qil- < 500 mrem /yr, and

, -i [

I (L g -+.1.1 . Mg ).-[(X/Q) ) -< 3000 mrem /yr"  ;

i .;

y  :

where the terms are defined below.

' '~

B2.2.2. Radioiodines. Particulates, and Others s

9 h ' I:' P g. [W Qg-] > < 1500 mren/yr  !

3 j; 1_ p

,4 , .

  • 4 : ~

w here: i

,' 'l

> K'. = The total body dose factor'due to gamma emissions for each identified noble gas radionuclide,~in mres/yr per pCi/mfrom Table!1.2-1. ]

= The' skin dose ~factoridue to beta emissions for each identified noble.

.Lj F',; gas'radionuclide,-in;mres/yr per.pCi/m'cfrom Table 1.2-1.- ]

c r H-I  :=,The.-air dose. factor due to gamma omissions-for each identified-noble r U

'" (f .

[i}- gas radionuclide,-'in mrad /yr per UCi/m'.from Table 1.2-1 (unit conver-sion constant-of;1.1 mrem / mrad converts air dose to skin dose).m i t

m

~ P g^ = The' dose parameter for-radionuclides other than-noble. gases;for.the s r ' inhalation pathway,--in aree/yr per pCi/m8 'and for the-food and ground-

planeipathways, In m2 -(mres/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).

4 9

= The release rate of radionuclides, i, in gaseous effluent from all

Q g ,

release pointslat the site, in pCi/sec. i a

h E

'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. The location is the NNE' sector @ 0.5 j r miles.

! , ~ l. '

B 7, W' = Thefannual. average dispersion or deposition parameter for estimating-H  : the-dose to an individual at a controlling location in the b unrestricted area where the total inhalation, food and ground plane

[- - pathway. dose'is' determined to be a maximum based on operational source term--data, land use surveys,-and NUREG-0133 guidance:

3 W = 1,9E-6'sec/m , for the-inhalation pathway. The location is h[.

v(  :

the 'E sector @ 1.5 miles, B-6 Revision 26

.?.

1 v 1/1/90 i ,

,~

- i, f

,;I ' &

h W =-2.6E-9 meter ~8,-for the food and gound plane pathways. ~The

-i location is the E sector @ 1.5 miles.

e v

Qg.- -=.kiC gf + kg = 4,72E+2Cg f where:

C.g- =-the concentration of radionuclide, 1, in undiluted gaseous effluent, in pCi/al.

f = the undiluted effluent flow,.in cfm i

k i = conversion factor, 2.83E4 ml/ft' '

s kg = conversion .f actor, 6El sec/ min i

. O e

1 I

g 5

i i

1 l

A V-B-7 Revision 26 1/1/90 q l

1

, i 1

~

5/fi B3.0' RADIATION MONITOR SETPOINTS' a c Ys. E .

i a 'Using the . generic. calculations: presented'in Section 2.0, final ef fluent radi-  ;

n'

, .ation monitoring setpoints. are calculated for monitoring as . required by the _

l Technical' Specifications.

All radiation monitors for McGuire are off-line except ENF-50.(Waste das System) which is~in-line. These' monitors alarm on low' flow; the minimum flow alarm' level'for both the liquid. monitors and the gas monitors is based on the i manufacturer's~ recommendations.. These monitors measure the activity in.the 'q liquid or gas volume exposed to the detector and are independent of flow rate if a minimum flow rate is assured.

4 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: y c= r 2.22 x 10's V l

-l where: ,

4 e = the gross activity, in pCi/ml-g r = the count rate. in cpm j  ; '- 2.22 x:10' = the disintegration per minute per 401

^~- 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 Liould Effluent Line As described in-Section B2.l.1 on release rate calculations for the waste .

liquid effluent, the release is controlled by limiting the flow rate of 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 5that used to calculate the release rare. 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<" f

= 1.04E-4 pCi/mi where:

c = the gross activity in undiluted effluent, in EC1/ml jN f = the flow from the tank may vary from 0-120 gpm but, for this calcu-j ,):_

lation, is assumed to be 100 gpm.

q l

B-8 Revision 14 i 1/1/87 l 1

h y -- _. ,-,__,_,,,,,.__,,_,,_-------,---__.-_-a -

,m >

\:

/ Ql MPC = 1.0E-07-UCi/ml; 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 water

, pump in service or 2.5E+5 gpe.- 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 containment ventilation unit condensate effluent, it is probable that the effluent will contain measurable activity above background. Since the tank contents can be discharged automatically, a maximum tank concentration, which j '

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:

c <.

g

= 1.04E-4 pCi/ml where:

l c = the gross activity in undiluted effluent, in uCi/ml I

= the flow from the tank may vary from 0-120 gpm but, for this 7% f calculation, is assumed to be-100 gpm MPC = 1.0E-07 pCi/ml, the MPC for an unidentified mixture l}

a = 2.4 (See Section B2.1.1) l F = the dilution flow is based on having only one condenser cooling ,

pump-in service or 2.5E+5 spm. Should the number of pumps in i service increase, the setpoint may be recalculated. l 1

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.  ;

B3.1.3 Conventional Waste Water Treatment System Discharme Line  ;

As described in Section B2,1.3 on release rate calculations for the conven- I tional waste water treatment system offluent, the offluent is normally con--

sidered-non-radioactive; that is, it is unlikely the effluent will contain j measurable activity above background. It is-assumed that no activity is j present in the effluent until indicated by radiation monitoring and by routine

-analysis of.the composite sample collected on that.line. Since the system discharges automatically, the maximum system concentration, which also is the radiation. monitor setpoint, is calculated so that release limits are not

/"3 exceeded. A typical monitor setpoint and maximum effluent concentration in V calculated as follows:

B-9 Revision 14 1/1/87

g. ,

fl' i

-i

.Q J g[ c5 MPC x F = 5.36E-07 pCi/ml

- where:

c = the gross activity in undiluted effluent, in pC1/mi 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/m1, the MPC for an unidentified mixture F.

= the flow past Cowan's Ford Dam may vary from 80 to 50,000 cfs, but j is conservatively estimated at-80 cfs (3.59E+04 gpm), the minimua j flow available i t

o = 1 [The Conventional Waste Water System discharge line is located, downstream of Cowan's Ford Dam and, therefore, has no reconcentra-t-

tion (recirculation) factor associated with it.]

-B3.1.4s Turbine Buildina Sump Discharme Line to the Condenser Cooling-Water (RC)

As described in Section B2.1.4 on release-rate calculations for-the Turbine-

= Building sump effluent,.it is'possible that the affluent will contain measur -

Lp - able activity above background.; Since the sump contents can be discharged

[ aj l

automatically to the RC,-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 concentration is calculated as follows:

MPC cI  !=5.21E-06pCi/ml-E where:

J -

l- c = the gross activity in undiluted effluent, in pCi/mi u

f = the flow rate of undiluted effluent which may vary from 0-2000 '

l~ gpm, but is assumed to be 2000 gpm n

. MPC = 1.0E-07 pCi/ml, 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 (1-8) of pumps l in service increase, the setpoint may be recalculated. '

l 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.

l B-10 Revision 17 1/1/88 L

l i i w 1

., ; r

>9 -

' r 't i v ,

  • . ' f, ,

.Tl 5 ,

a

, ,g .m. ,

e e A q y;9  ;*

w m' o B3.2- ' GAS MONITORS h +

~.m .

?The fol' lowing equationishall be used to calculate final. effluent noble gas.

~

radiation monitor setpoints based.on Xea133: l y K(X/Q)Qg ;<-500 (See section B'2'.2.1) og= 4.72E+2 Cgf (See Section B2.2.2). Il 6 ,

lCg < 5.00E+01/f l where : --

l C = the gross activity in undiluted effluent, in pCi/ml f l i

, f = the flow from the tank or building and varies for various release  ;

sources, in efm K- = from Table 1 2-1 for'Xe-133,-2.94E+2 mres/yr per UCi/m

X/Q =.7.2E-5 sec/m 8 , as defined in Section B2.2.2.

'B3.2.1 ' Unit, Vent I 1,G % Asistated'in Section'B2.2,.the unit vent is-the release point for the waste gasc

, j., b " system, containment purge ventilation system, the-containment. air release and u

' ' addition system, thescondenser air ejector and. auxiliary building ventilation. _l

<  : Since all'of.these releases are through:the unit: vent.--the radiation monitor on j"

the unit vent =may be used to assure that station release limits:are not.

i' exceeded.

For release from the containment air release and-addition system and=the 4

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)

LFor release from the containment air release and addition. system..the waste

< gas decay: tanks, the condenser air ejectors, and the auxiliary building

~ *

-ventilation system, a typical radiation monitor setpoint may be calculated as follows:

lC ~ 3.0CD 01/f = 3.51E-04 pC1/ml 1

where:

L/] . f = 142,500 cfm (auxiliary ventilation systems)

&.f B-11 Revision 16 5/15/87 d 66 --

{y , " e

~

s

?

Y-1)

IV' - .

((%, Ji c.

'B3;2.2.2 . Waste Management Facility (EMF-52)

Ventilation exhaust =from the Waste Management Facility is not released through

the unit'ventland41s considered'a separate release. point. .This. exhaust is normally-cons'idered non-radioactive; thatLis, it is possible but unlikely that

-the' effluent will contain measurable activity above background. Since the exhaust is continuous, a maximum concentration.of gases in the. exhaust, which

~is also the radiation monitor setpoint, is calculated to assure compliance-

with release limits. A typical radiation monitor setpoint may be calculated as follower I

C < 5.00E+01/f = 5.95E-03 where: .

f = 8400'cfm B3.2.3 Waste Handling Area (EHF-53) j

-Ventilation exhaust from the Waste Handling Area is not released through the j~

unit vent and is considered a separate release point.- This exhaust is-

,normally considered non-radioactive; that is,.it is,possible but unlikely that  !

the' effluent will contain measurablefactivity above. background. Since the-  !

4r . exhaust ris continuous,V a maximum concentration. of gases "in - the exhaust. . which is'also the radiation' monitor setpoint, is calculated to assure ~ compliance Nt with release' limits. A typical. radiation monitor setpoint may be calculated ~ ,

(_,) . as-follows:

}

l C < S.00E+01/f = 6.51E-04 i 1

where:

f = 76,800 cfm a B3.2.4- Unit.2 Staging Building (2 EMF-59)

Ventilation-exhaust from the Unit 2 Staging Duilding is not released through the unit-vent and is considered a separate release point. This exhaust is normally considered non-radioactive; that is, it is possible.but.unlikely that.

the effluent will contain measurable activity above background. Since the exhaust is continuous,Ja maximum concentration of gases in the exhaust, which L is also the radiation monitor setpoint, is calculated to assure compliance

,with release limits. A typical radiation monitor setpoint may-be calculated 7 as follows:

a.

L C < 5.00E+01/f = 7.35E-03 ,

L.

where:

l f = 6800 cfm L

1

< B-11a Revision 26

[ 1/1/90

^ ~~ ^ -~

y , ,

l p l16 j

_p -

i y 7 Vm ..

DOSE CALCULATIONS ,

n i .'1 0 1B4.0 ,

M; c

-B4.1; h FREQUENCY OF CALCULATIONS'  ;

t Dose-contributions ~to the maximum exposed individual shall be calculated at '

'least every 31 days, quarterly, semiannually, and annually (or as required by Technical Specifications).using=the methodology in the. generic'information sections. '. This- methodology shall also be used for any special reports. Dose (

calculations that are required for individual pre-release calculations, and/or

' abnormal releases shall not be calculated by using the simplified dose calcula-tions. Station dose projections for these types and others that are known to vary from the station historical averages shall be calculated'by using the  !

t  ! methodology in the generic information sections. STATION Dose projections may.

be performed using simplified dose estimates.

" Fuel cycle dose calculations shall be performed annually or as required by special reports. Dose contributions shall be calculated using the methodology in the appropriate generic information sections. .j B4.2 DOSE MODELS FOR MAXIMUM EXPOSED' INDIVIDUAL

-B4.2.1 Liquid Effluents

? For-dose contributions from liquid radioactive-releases,. dose calculations l i based on operational. source. term data and NUREG-0133 guidance indicate that

  • TN the maximum exposed individual would be an adult who consumed fish caught in L'( j the discharge canal and who-drank water from the nearest " downstream" potable water intake. The dose from Cs-134 and Cs-137 is calculated to be 87*. of that l

G

~

. individual's totalewhole body dose.

.j

-B4.2.2 Gaseous Effluents g l

B4.2.2.1 Noble Gases' .I 1

i For dose contributions from exposure to beta and gamma radiation from noble j gases, it is assumed that the maximum exposed individual is an adult at a j

' controlling- location in the unrestricted area where the total noble gas dose is l determined to be 'a maximum. _I

~ B4'.' 2 f 2. 2 Radioiodines, Particulates, and Other Radionuclides T 1/2 >8 days For dose contributions from radiolodines, particulates and other radionuclides; q it'is' assumed that the maximum exposed individual'is a child or infant at a l controlling location in the unrestricted area where.the total inhalation, food l aud: ground plane pathway-dose is determined to be a maximum based on opera- i tional source term data, land use surveys, and NUREG-0133 guidance. ~l B4.3 SIMPLIFIED DOSE ESTIMATE B4.3.1 Liould ~Ef fluents tO For dose estimates, a simplified calculation based on the assumptions b presented.in Section B4.2.1 and-operational source term data is presented

.below. Updated = operational source. term data.shall be used to revise these zcalculations as necessary.

B-12 Revision 26

+ . . . .-. . -. _ _ . 1/1/90..

4 ,* , ,

a y 8 , ;!i r

~

~ '

<.6 \ g

'Y l

g. , w ,

, j Vff m I Q' .

~D VB =6.78E+5~I-(F)pg)(CCs-134.+

g 0.59 CCs-137) t=1' .

, , where:  !

q

6.'78E+5".=I1.14E+05(U,,/D,+U,g BFg ) DFah (1.15)'

, where:

~

8

';1.14E+05 =' 10'pC1/pCi .x 10 ml/kg + 8760 hr/yr

]

' U,, = 730.1/yr,Tadult water consumption-f i

=.1, dilution' factor'from the near field area'to the nearest possible

. D" potable water. intake (Hunterville Water Intake). .

c',

U g = 21 kg/yr, adult fish consumption

< -BFf ;= 2.00E+03, bioaccumulation. factor for Cesium (Table 3.1-1)

- DF = 1.21E-04, adult total body ingestion dose factor for Cs-134-ah .

, (Table 3.1-2)  !

1.15'=: factor derived frorn the'; assumption that 87% of, adult,wholeibody; dose-is 4 ,e contributed ~by Cs-134'and Cs-137 via the fish-and drinking water pathway-o W- or .100% + 87% =' 1.15  :

L' j

- m = number of relecses l- -

p: where:

1 7 y to

, t- F+f i

g where:

?' ,

~f~= liquid radwaste flow,
in gpm 1

.o = recirculation factor at equilibrium, 2.4 q t- a F =. dilution ~ flow, in gpm- t L

where:-

Tg = The length of time, in hours, over which CCs-134, CCs-137' "" 2-v' .are averaged. (The time period during which all releases (m) are made)

JC = the average concentration of Cs-134 in undiluted effluent, in Cs-134 pCi/ml, during the time' period considered.

fy C Cs'137-

= the average concentration of Cs-137 in undiluted effluent, in

'Nj pCi/ml, during the time period considered.

B-13 Revision 26 1/1/90

1 4

[~

'O 59 = The ratio of th's adult total body ingestion dose factors for v Cs-134 and'Cs-137 or 7.14E-05 + 1.21E-04 = 0.!)9_

. B4.3.2 Gaseous Effluents Meteorological data is provided.in Tables B4.0-1 and B4.0-2.

-1

- B4.3.2.1 ' Noble Gases For dose estimates, simplified dose calculations based on the assumptions in i

- B4.2.2.1 and operational source term data are presented below. Updated opera-  !'

tional source term data shall be used to revise these calculations as neces-sary. These calculations further assume that the_ annual average dispersion 'i parameter is used and that Xenon-133 cor. tributes 49% of the gamma air dose and i 76% of the beta air dose, s q Dg = 8.06E-10 [Q]Xe-133 ( .04 Dg = 2.40E-09 [Q)Xe-133 ( * }

where:

\

X/Q = 7.2E-05 sec/m*, as defined in Section B2.2.2

- 8.06E-10 = (3.17E-8)(353) (X/Q), derived from equation presented in n Section 3.1.2.1.

] V 2.40E-09 = (3.17E-08) (1050) (X/Q), derived from equation presented in Section 3.1.2.1.

~

[Q]Xe-133 = the total Xenon-133 activity released in 1:C1 2.04 = factor derived from the assumption that 49% of the gamma air _ dose is contributed by Xe-133, 1.32 = factor derived from the assumption that 76% of the beta air dose is contributed by Xe-133.  ;

B4.3.2.2 Radiolodines, Particulates, and Other Radionuclides with T 1/2 '

> 8 days For dose estimates, simplified dose calculations based on the assumptions in B4.2.2.2 and. operational source term data are presented below. Updated opera- .

- tional source term data shall be used to-revise these calculations as neces- l

- sary. These calculations further assume that the annual average dispersion /  ;

deposition parameters are used and that 97% of the dose results from Iodine-131

-ingested by the' maximally exposed individual via the goat milk pathway at the controlling location. The simplified dose estimate to the thyroid of an infant is: l s

D = 1.84E+04 W (Q)I-131 (1.03) l . {pJ.

where:

l 3 W = 2.6E-09-= D/Q for food and ground plane pathway, in sec/m from Table l B4.0-2 for the controlling location (E sector at 1.5 miles).

B-14 Revision 26 j, 1/1/90

M..<+/ A J 4

%v ,

,m s .;

k/ -(Q)1 131_= the total Iodine-131 activity released in pC1.

<1.84E+04 :='_(3.17E-08)(R [D/Q)) with the< appropriate, substitutions for .

infant goat milk pathway factor, R [D/Q] for Iodine-131. See i Section 3.1.2.2.

1.03 = factor derived from the assumption that 97% of the total inhala- ,

tion, food and ground plane pathway dose to the maximally. exposed

' individual is contributed by I-131 via the goat slik pathway.

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 accordance with 40CFR190. The fuel cycle dose assessments for McGuire Nuclear Station must include gaseous "

dose contributions from Catawba Nuclear Station, which is located (epproximately-thirty miles SSW of McGuire. For this dose assessment, the total body and maximum organ-dose contributions to the-maximum exposed

. individual frca McGuire liquid releases and the combined Catawba and McGuire gaseous releases-are estimated using the following calculations:-

DWB(T) = DWB(I m ) + DWBI8)+DWB(8 m c) h '

DM0(T) = DM0( m) + DMOC8)+DM0(8 m c) where:

.DWB(T) =' Total-estimated fuel cycletwhole body' dose. commitment resulting from the combined liquid-and gaseous effluents of Catawba and

  • McGuire during the calendar year of interest,-.in. mrem.

DM0(T);= Total estimated fuel cycle maxmium organ dose commitment resulting from the combined 11guld and gaseous effluents of ,

Catawba and McGuire during the calendar year of interest', in mrem.

B4.4.1 LIQUID EFFLUENTS Liquid pathway' dose estimates are based on values and aseumptions presented in

Section B4.3.1. Station operational source terms shall be used'to update these simplified calculations as necessary.

E B4 '. 4.1.1 McGuire's Liquid Contributions

. Based on operational history, the McGuire fuel cycle whole body dose resulting

from>McGuire+11guid effluent releases (D simplified dose calculation given below:WB(1m ))~is. estimated using the DWB(1,) = (6.78E+05) (Fg ) (T g) (CCs-134 + 0.59 CCs-137)

B-15 Revision 26 1/1/90

i

',\

(

where't I Xf  :

6.78E+5 ='1.14E+05 (U, / D,+ U,g BFt ) (DF,gg) (1.15) where:

1.14E+05 = (1.0E+06 pCi/uci'x 1.0E+03Lal/kg) / (8760 hr/yr)

.U,,= 730 1/yr, Adult water consumption- i D, = 1, dilution factor from the near field area to the nearest possible potable' water intake (Huntersville Water Intake).

'U,g.= 21 kg/yr, Adult fish consumption BF g = 2.00E+03, Bioaccumulation factor for Cesium (Table 3.1-1) t 1.21E-04, Adult total body ingestion dose factor for DF,gg = Cs-134 (Table 3.1-2) 1.15 = Factor derived from the assumption that 87% of the dose.is. derived -

from Cs-134 and Cs-137 or 100%./ 87% = 1.15 where:

A -F = (f) (c) / (F + f) ,

where:

f = McGuire's. average liquid radwaste flow for=the calendar year of interest, in gpm F = McGuire's average dilution flow for the calendar year of interest, in spa 'I o = 2.4, the recirculation factor at equilibrium where:

Tg= 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />, the time period over which CCs-134, CCs-137 and Fg are averaged. -

'C = The average concentration of Cs-134 in McGuire's undiluted Cs-134 effluent, in uCi/m1, during the calendar. year of interest.

C Cs-137

= The average concentration of Cs-137 in McGuire's undiluted-effluent, in uCi/ml, during the calendar year of interest.

. 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 h

'\/.

Based on operational history, the McGuire fuel cycle organ dose (teen liver) resulting from McGuire's liquid effluent releases (DM0(Im )) is estimated using l

the simplified-dose calculation given below:

I l

B-16 Revision 26 l 1/1/90 l

r, Ju . - ,

+%

l'S3 j

= NJ =x ..

, E-j,

.F g

x xq

[  ?

fl.. .[Mj\ DM0(I m ) = (8.16E+5) g(F ) pg NCs-134 +' 0 Cs-137) f

'N

'where: N ~ l

'N '

8.16E+5 = 1.14E+05 (U,,/D ,+ U,g BFg )(DF,gg)(1.11) N's _

where x 1.14E+05 = (1.0E+06 pCi/uCi x 1.0E+03 ml/kg) /-(8760 hr/yr)

U,,= 730 1/yr, Teen water consumption ,

D = 1,-dilution factor from the near field area to the nearest possible-r potable water intake (Huntersville Water Intake).

p U,g = 16 kg/yr, Teen fish consumption BFg = 2.00E+3, Bioaccumulation factor for Cesium (Table 3.1-1).

DF = 1.97E-4 . Teen liver ingestion dose factor for Cesium "It (Table 3.1-3) 1.11 = Factor derived from the assumptica that'90% of the' Teen-liver

-dose'is contributed.by Cs-134 and Cs-137 via the fish and drinking ,

water pathway or 100% / 90% =-1.11'

.:j where:

Fg = (f) (a) / (F + f) where:

f = McGuire's liquid radwaste flow, in gpm-u F = McGuire's dilution flow, in'gpm <

i o = 2.4,.the recirculation factor at equilibrium p where:

Tg = 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br />, the time period of time over which Cs-134, Cs-137 and F g i are averaged h

l C Cs-134 = The average concentration of Cs-134 in McGuire's undiluted effluent, in pC1/ml, during'the calendar year of interest.  !

C Cs-137

= e average concentration of Cs-137 in McGuire's undiluted effluent, in pCi/ml, during the. calendar year of interest.

l:

L .0.76 = The ratio of the Teen-liver ingestion dose factors for Cs-134 L

L and Cs-137 or 1.49E-4/1.97E-4 = .76.

y l

l B-17 Revision ~26 1 1/1/90

r b

+

, ;i p

f(-(f) B4.4.2' GASEOUS EFFLUENTS i

s, "AirborneLeffluent pathway dose estimates are based on the values and assump-

.{ tions presented in Sections B4~.3.2. and-C4.3.2. Operational source term data '

m , .shall^be-used to' update these calculations:as>necessary.-

.B4.4.2.1' McGUIRE'S GASEOUS CONTRIBUTION k

. = l Based on operational history,-the McGuire-fuel cycle maximum whole body; dose *

, resulting from McGuire's; gaseous effluent releases (DWB(8 m )) is estimated

.using the simplified dose calculation-given-below:  ;

t DWB(8 m ) = (9.3 -06)(w)(6Xe-133)(OF)(2.38) where:

w = 7.20E-05 = (X/Q) for the plume immersion pathway,.in

-i sec/m , which corresponds to a location 0.5 miles NNE of the 8

McGuire site (See Table B4.0-1) ,

, ;QXe-133 = The total Xe-133 activity released from McGuire during the. calendar. I year of interest, .in'uCi. ,

gg 9.32E-06 = (3.17E-08) (K g [X/Q]), with appropriate substitutions for whole body <

V) exposure in a semi-infinite cloud of Xe-133. See Section 1.2.1.

Sp = 0.7--=. External radiation shielding factor.for individuals.

2.38 = Theifactor derived.from-the-assumption (based on historical data) that i

, 42% of the whole body dose to the maximally exposed individual is contributed by Xe-133.-

Based-on operational history, the fuel cycle maximum organ dose is the- '

Teen-liver. . For McGuire gaseous releases the organ dose-(DM0(8 "Ill h* i i calculated for the Adult-GI Tract using the simplified dose calEu))lation given below:- _,

i D

M0(8 m ) = (8.27E-05)(w)H d -3} ( ' ')

!where:

w = 1.80E-5, X/Q for the garden pathway, in sec/m' from Table B4.0-1, for the McGuire fuel cycle maximum organ dose controlling location (E @ 0.5 miles).

6-3 H = The total H-3 activity released from McGuire during the calendar year of interest, in uC1.

[Lsy ~8.27E-5 = (3.17E-08)(R V

g (X/Q]) with appropriate substitions for the V

garden pathway, Rf [X/Q] for H-3. See Section 3.1.2.2.

B-18 Revision 26 1/1/90

(W , ,

g} y F>  !

i;,

(M L ! ) 1.59 = The factor derived from=the conservative assumption (based'oni '

L' M historical data) that 63% of the total inhalation,-food and ground 3 plane pathway dose to the maxima 11y' exposed individual is!

e

. contributed by H-3 via the inhalation pathway.

B4.4.2.2-- CATAWBA'S GASEOUS CONTRIBUTION  ;

Based on operational history, the McGuire fuel cycle maximum whole body Jose resulting from Catawba's gaseous effluent releases (DWB(8 c )) is estimated using the-simplified-dose calculation given below: .:

D WB(8c ) = (9.32E-06)(w) Xe d -133)(8F )(1. 0)

, where:. I w = 3.30E-07 -(X/Q) for the plume-immersion pathway.which corresponds to a location 5 ailes NNE of the Catawba site. (See Table C4.0-1)'

Q

~

= The-total Xe-133 activity released from Catawba during the.. calender.  !

Xe-133 year of'. interest, in uCi.

' 9.32E-06 = (3.17E-08)(K g

[X[Q']), with appropriate substitutions- for whole body ,

~

t(( exposure in-a semi-infinite cloud of Xe-133. See-Section 1.2.1.

.Sp = 0.7.= External radiation shielding factor for individuals.

2.56 = The factor' derived fros'the conservative assumption (based on historical data)-that 39% of the whole-body. dose to-the maximally-exposed-indivdual is contributed by Xe-133 via1the plume. immersion- -

, pathway.

. Based on operational history,.the fuel cycle maximum. organ. dose for McGuire is . .

the. Teen-liver. For Catawba gaseous releases the organ dose (Dg(g e )) "'11 calculated for the Teen-liver using the-simplified dose calculation given .

below: i

'l DM0(8 c ) = (8.2 -05)(w)@H-3)(2.38) )

where: ,

1 w = 3.3E-07 = X/Q for the food and ground plane' pathway in sec/m', for'a 'l location 5 miles NNE of the Catawba site (see Table- l P, C4.0-1).  ;

4

,Q e. talsH-3 activity released Jrom Cetawba.during.the calendar H -3'=-

year of interest, in uC1.

j Ol D '8.27E-05 = (3.17E-08)(R [X/Q]) with appropriate substitutions for l the teen-vegetable garden pathway, R [X/Q] for H-3. See Section 1

3.1.2.2. l B-19 Revision 26 1/1/90

~ ~~ ~

' ~

p , , :n , , , [

- + .i ji , j 'i -

e s 1

'll ; '

1 kpX , _ 1

't , , . . . . . . . . . .

" 2.38 = The factor derived:from the assumption'(based on historical data) that:

j

,l '("-)' >

. 42% of the total inhalation, food and ground plane pathway dose to the.

r;

- maximally exposed individual is contributed by-H-3 via the vegetable

+

. garden pathway. ,

i t

ti,'

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, 4 t

l i

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i T

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f i

el i '.;

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-B-20 Revision 26

, _ 1/1/90 ti

.L ; -

ys :s

-fg 7.h #

'( }

.x .

TABLE B4.0-1 ,

-(1 of 1) '

HCGUIRE NUCLEAR'ST5 TION DISPERSION PARAMETER (X/0) FOR LONG TERM RELEASES > 500 HR/YR OR > 125 HR/QTR

.(sec/m') ,

s 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 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 1.4 E-7 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 NNE 7.2 E 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 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 1.4 E-6 '9.0 E 6.4 E-7 4.8 E-7 3.8 E-7 3.1 E-7 2.6 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-71 ESE 1.2 E 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 3.0 E-7 2.2 E-7 1.8 E-7 1.4'E-7 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' 1

Revision 7, 1/1/85 a m

_%--i. - ti y .  %  % , +wh.r9 - f. ,

rgen-..,., W .eg-..yi-- -g ., ._,,.e y e yg e ,i.

.-a "- =w.; , .

3; - _ . '

~~ ~

,n

, ,n 3.; ~

y&

~

~

TABLE B4.0-2^

~(1 of 1) -

MCGUIRE NUCLEAR STATION .

w ' .

DEPOSITION PARAMETER (D/0) FOR LONG 'ITRM RELEASES > 500 HR/YR OR > 125 HR/QTR. -

(m 8) ,

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 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-10f 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-9 819 E-10 6.3 E 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 3.1 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 cl.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 _

N 9.3 E-8 2.3'E-8 8.1 E-9 4.0 E-9 2.4 E-9. 1.6 E 1.1 E-9 8.1 E-10 6.3 E-10 5.0'E 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-3 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 14 1/1/87

d-t g -

1 WF TABLE B4.0-3*

L{ Q .

(1 of 3)

NCGUIRE NUCLEAR STATION ADULT A DOSE M AME M S. )

ait (area /hr per WC1/ml)

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LII 1 1

H 3 0.0 8.96E+00 8.96E+00 8.96E+00 8.96E+00 8.96E+00 8.96E+00 l 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 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+C4 CO 58 0.0 1.51E+02 3.39E+02 0.0 0.0 0.0 3.06E+03. l 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 L

x CU 64 0.0 1.69E+01 7.93E+00 0.0 4.26E+01 0.0 - 1.44E+03 p ZN 65. 2'.36E+04 7.50E+04 3.39E+04 0.0 5.02E+04 0.0 4.73E+04

[.

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 1.95E+02 1.37E+02 0.0 0.0 0,0 1.13E-11 l SR 89 4.78E+04 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 i

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 l

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 11 9/30/86

.~ __ . . . . - - . - - - - -.

!T 10 ,

)

n f ,

[ TABLE.B4.0-3 1

(2 of 3) .

MCGUIRE NUCLEAR STATION ADULT A,gt MSE PARAME M S j (arna/hr per pCi/ml) i 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 J

ZR 95 2.77E+00 8.88E-01 6.01E-01 0.0 1.39E+00 0.0 2.8EE+03 i g 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 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 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 i 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 q / . TE 125M 2.79E+03 1.01E+03 3.74E+02 8.39E+02 1.13E+04 0.0 1.11E+04 TE 127M 7.05E+03 2.52E+03 8.59E+02 1.80E+03 2.86E+04 0.0 2.36E+04 0.0

~

TE 127 1.14E+02 -4.11E+01 '2.48E+01 8.48E+01 4.66E+02 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 1.37K+02 0.0 2.47E+01 TE 131H 1.80E+03 -8.81E+02 7.34E+02 1.39E+03 8.92E+03 0.0 8.74E+04 MDE 131. .2.05E+01. 8.57E+00 6.47E+00 1.69E+01 8.98E+01- 0.0 2.90E+00 12 132 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 3.03E405 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 4- 3.88E+05 5.31E+05 3.48E+05 CS 137 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 TABLE.B4.0-3 (2 of 3)

Revision 7 1/1/85 er

I

- i

" 'I y ,-: -

/ ) TABIE B4.0-3 _

"\s,/ i J -= (3 of 3)

MCGUIRE NUCLEAR STATION i

ADULT A,g DOSE PARAMETERS (aram/hr per UC1/ml)

NUCLIDE BONE . LIVER -T. BODY THYROID KIDNEY LUNG GI-LII s 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-~31 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 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 15.77E+03

.PR 144 4.31E-03 1.79E-03 2.19E-04 0.0 1.01E-03 0.0 6.19E-10

- )'*N _KD 147- 9.00E-01 1.04E+00 6.22E-02 0.0 6.08E-01 0.0 4.99E+03-t t W 187 3.04E+021 2.55E+02 8.90E+01 0.0 0.0 0.0 8.34E+04' NP 239 1.28E-01 1.25E-02 6.91E-03 .0. 0 3.91E-02 0.0 2.57E+03 4

1 1

. l I

l l

i l

l

-l l

1

/(_/ TABLE B4.0-3 l (3 of 3)

Revision 7 1/1/85

7

\

i O

),J 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 j specific charac.teristica make groundwater sampling unnecessary. Groundwater ,

  • recharge is from Lake Norman and local precipitation. The groundwater gradient I 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.

The dates of the land-use census that was used to identify the controlling receptor locations was 05/15/89 - 06/29/89.

I O

e I

A U +

B-21 Revision 26 1/1/90 i

O O O TAB E B5.0-1 (1 of 1)

MCGUIRE RADIOIDGICAL MONI1tMtING PROGRAM SAMPLING LOCATIONS (TLD IDC&TIONS)

I i

SAMPLING IDCATION DESCRIPTION

  • SAMPLING IDCATION DESCRIPTION
  • i 143 SITE BOUNDARY (0.5 MIES Nii) 163 4-5 MI E RADIUS (5.0 MIIES SE) 144 SITE BOUpOARY (0.6 MIIES DBE) 164 4-5 MILE RADIUS (4.5 MILES SSE)-

145 SITE BOUpmARY (0.5 MIES IE) 165 4-5 MIIE RADIUS (5.0 MIIAS 5) i 146 SITE BOUpWARY . (0.5 MII25.EIE) 166 4-5 MII2 RADIUS (5.2 MIIES SSW) j 147 SITE BOUpWARY (0.5 MIIES E) 167 4-5 MIIE RADIUS (4.9 MIIXS SW) .

! 148 SI*IE BOUNDARY (0.5 MIIES ESE) 168 4-5 MIII RADIUS (4.7 MIES USW) 4 149 SITE BOUISARY (O.7 MIIES SE) 169 4-5 MIM RADIUS (4.4 MIES W) i 153 SI~IE BOUNDARY (0.5 MIIES SSE) 170 4-5 MILE RADIUS (4.4 MILES WORf)

. 151 SITE BOUNDARY (0.5 MIIES S) 171 .4-5 MIE RADIUS (4.5 MIIES IGF) l 152 SITE BOUNDARY (0.5 MIIES SSW) 172 4-5 MI E RADIUS (5.2 MIIES feef) l 153 SITE BOUNDARY (0.5 MIIES SW) 173 SPECIAL IltIEREST (8.5 MII25 leaf)

154 SITE BOUNDARY (0.7 MIIES WSW) 174 SPECIAL IltIEREST (8.7 MIIES WIGf) i 155 SITE BOUNDARY (0.7 MIIES V) 175 CONUiOL (12.7 MIIES Witi) l l 156 SI'IE BOUNDARY (0.5 MIES WNW) 176 SPECIAL INIEREST (11.0 MIIES SW) l 157 4-5 MIII RADIUS (4.8 MIIES N) 177 SPECIAL INIEREST (8.6 MIIAS S) l 158 4-5 MIII RADIUS (4.4 MIIES 791E) 178 SPECIAL INIEREST (9.2 MIIES SE)

! 159 4-5 MIE RADIUS (5.0 MIIES ME) 179 SPECIAL INIEREST (10.4 MILES ESE) 160 4-5 MIE RADIUS (4.9 MIIES ENE) 180 SPECIAL INIEREST (11.5 MIES MIE)

! 161 4-5 MILE RADIUS (4.7 MIIES E) 181 SPECIAL INIEREST (6.7 MI25 NE)

! 162 4-5 MILE RADIUS (4.6 MIIES ESE) 182 SPECIAL INIEREST ' (6.0 MIIES NE) 183 SPECIAL INIEREST (5.5 MIIES 5) l

  • All TID samples are collected quarterly I

' Revision 26 1/1/90

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TABLE B5.0-2 (1 of 1)

MCGUIE RADIOLOGICAL MONITORING PN0 GRAM SAMLING IACATIONS (OTER SAMLING LOCATIONS) ~

c

. # 5

    • E CNE: $ I 3" u Y . &

SM - Semimoothly 11 2 $

W - Weekly 3  %  %

  • j 8 BW - Biweekly Q - Quarterly 3 $3 , 4 M - Monthly SA - Semissemally Q% , I S & E E JE E 5 , 5 um t S  : 1 1 d y

SAMPLING IACATION DESCRIPTION f" 53 $ $  ;". g, C $

120 Site Boundary (0.7 mi NNE) W M 121 Site Boundary (0.5 mi NE) W 125 Site Boundary (0.5 mi SW) W M 128 Discharge Canal Bridge (0.4 mi ER) W

129 Discharge Canal Estrance to Lake Norman (0.6 mi ER) SA SA 130 Hwy. 73 Bridge Downstream (0.6 mit SW) W SA 131 Deleted 01/01/89 l 132 Charlotte Municipal Water Supply (11.2 mi SSE) W 133 Cornelius (6.2 mi NE) W 134 East Lincoln Junior Nigh School (8.7 mi WNW) I W H 135 Plant Marshall Istake Canal (12.0 mi N) I N 136 Mooresville Nsaicipal Water Supply (12.5 mi NR) I W 137 Pinnacle Access Area (12.0 mi N) -

I -

SA SA 138 Henry Cook Dairy (2.75 mi ESE) SM 139 William Cook Deiry (2.25 mi E) SM 140 Kidd Dairy-Cows (2.8 mi SSE) SM 141 Lynch Dairy-Cows (14.8 mi 14Ri) I SM 142 Davidson Municipal Water Supply (7.5 mi NE) NW 158 4-5 Mile Radius (5.0 mi M ) M 159 4-5 Mile Radius (4.4 mi NME)(To be deleted 1/1/85 - location #120 will replace) ,a M 184 5 Mile Radius - Gardens (2.5 mi ENE)

M"' l (a) during barvest season ,

Revision 22 1/1/89 t

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_ d d TABIZ B5.0-3

~(1 of 1)

MCGUIRE RADIGIDGICAL MONITURING PROGRAM ANALYSES Ana A Y_YRE R SAMPE MEDIUM ANALYSIS SCHEDUE Gaffla ISUTOPIC TRTTIUM IaK IZVEL I-131 GROSS BETA TIR

1. Air Radiolodine and X Particulates Weekly X X
2. Direct Radiation Quarterly X
3. Surface Water Biweekly X Monthly Composite X Quarterly Composite X
4. Drinking Water Biweekly X Monthly Composite X X Quarterly Composite X
5. Shoreline Sediment Semiannually X i 6. Milk Semimonthly X X
7. Fish Semiannually X
8. Broadleaf Vegetation Monthly X
9. Food Products Monthly I*) X (a) during harvest season 2

Revision 9 1/1/86

^

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