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Text

Rev 3 to Offsite Dose Calculation Manual.
	ML18093A338
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Site:	Salem
Issue date:	07/30/1987
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SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Revision 3 07/30/87 * * ---------------------.., ( 8709030520 870828 I PDR ADOCK 05000272 R PDR

*

SALEH NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION HANUAL Table of Contents Introduction

. . . . . . . . . . .. 1 1.0 Liquid Effluents 2.0 3.0 1.1 Radiation Honitoring Instrumentation and Controls * * * *

2 1.2 Liquid Effluent Honitor Setpoint Determination

* * * *

3 1.2.1 Liquid Effluent Honitors <Radwaste, Steam Generator Blowdown and Service Water> * * * * * * * * * * *

4 1.2.2 Conservative Default Values *************

5 1.3 Liquid Effluent Concentration Limits -10 CFR 20 5 1.4 Liquid Effluent Dose Calculations

-10 CFR 50 * * *

7 1.4.1 Hember of the Public Dose -Liquid Effluents

7 1.4.2 Simplified Liquid Effluent Dose Calculation 8 1.5 Secondary Side Radioactive Liquid Effluents

-Dose Calculations During Primary to Secondary Leakage *

10 1.6 Liquid Effluent Dose Projection

* * * * * * * * * * * *

12 Gaseous Effluents 2.1 Radiation Honitoring Instrumentation and Controls

2.2 Gaseous Effluent Honitor Setpoint Determination 2.2.1 Containment and Plant Honitor **** 2.2.2 Conservative Default Values 2.3 Gaseous Effluent Instantaneous

*

13 * *

15 * *

15 *

16 Dose Rate Calculations

-10 CFR 20 * * * * * * *

18 2.3.1 Site Boundary Dose Rate -Noble Gases **** 18 2.3.2 Site Boundary Dose Rate -Radioiodine and Particulates

19 2.4 Noble Gas Effluent Dose Calculations

-10 CFR 50 **** 21 2.4.1 UNRESTRICTED AREA Dose -Noble Gases **********

21 2.4.2 Simplified Dose Calculation for Noble Gases 21 2.5 Radioiodine and Particulate Dose Calculations

-10 CFR 50

23 2.5.1 UNRESTRICTED AREA Dose -Radioiodine and Particulates

23 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates

.****************

2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations

2.7 Gaseous Effluent Dose Projection

Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY 3.2 Doses to HEHBERS OF THE PUBLIC -40 CFR 190

3.2.1 Effluent Dose Calculations

- 3.2.2 Direct Exposure Determination

24

25 *

28 29

30 30 * *

31 4.0 Radiological Environ*ental Monitoring Program 4.1 Sampling Progra* *********

* * * *

32

32 4.2 Interlaboratory Comparison Program . * . . . . .

33 )

* * ---Salem ODCH Rev. 3 07/30/87 Table of Contents -Continued Tables . 1-1 1-2 1-3 1-4 2-1 2-2 2-3 2-4 2-5 A-1 A-2 B-1 Para*eters for Liquid Alarm Setpoint Determination

-

1 ** 37 Para*eters for Liquid Alarm Setpoint Determination

-Unit 2 ** 38 Site Related Ingestion Dose Commitment Factors. Aio ****** 39 Bioaccumulation Factors <BFi> ****************

41 Dose Factors for Nob 1 e Gases * * * * * * * * * * * * * * * *

44 Para*eters for Gaseous Alarm Setpoint Determinations

-Unit 1

45 Parameters for Gaseous Alarm Setpoint Determinations

-Unit 2

46 Controlling Locations.

Pathways and Atmospheric Dispersion for Dose Calculations

Path1111ay Dose Parameters

-Atmospheric Re.leases

* *

Calculation of Effective HPC -Unit 1 ***** Calculation of Effective HPC -Unit 2 ******.***

Adult Dose Contribut.ions Fish and Drinking Water Pathways Unit 1 * * * * . * * * * * * * * * * * * * * * * *

49 ** so

A-4

A-5

B-5 Adult Dose Contributions Fish and Drinking Water Pathways Unit 2 * * * * * * * * * * * * * * * * * . * . * *

B-5 C-5 Effective Dose Factors * * * * * * * * * * *

C-5 Appendic ies Appendix A -Evaluation of Conservative.

Default HPC Value for Liquid Effluents

* * * *

A-1 Appendix B Technical Basis for Effective Dose Factors -Liquid Radioactive Effluents

B-1 Appendix C -Technical Bases for Effective Dose Factors Gaseous Radioactive Effluents Appendix D -Radiological Environmental Monitoring Program -Sample Type. Location and Analysis **** ** * *

D-1

*

SALEH NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION HANUAL Salem ODCH Rev. 3 07130187 The Salem Offsite Dose Calculation Hanual COOCH> describes the methodology and parameters used in! 1) the calculation of radioactive liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints:

and 2> the calculation of radioactive liquid and gaseous concentrations, dose rates and cumulative quarterly and yearly doses. The methodology stated in this manual is acceptable for use in demonstrating compliance with 10_CFR 20.106, 10 CFR 501 Appendix I and 40 CFR 190. Hore conservative calculation methods and/or conditions (e.g., location and/or exposure pathways) expected to yield higher computed doses than appropriate for the maximally exposed person may be assumed in the dose evaluations

The ODCH will be maintained at the station for use as a reference guide and training document of accepted methodologies and calculations.

will be made to the ODCH calculation methodologies and parameters as is deemed necessary to ensure reasonable conservatism in keeping with the principles of 10 CFR 50.36a and Appendix I for demonstrating radioactive effluents are ALARA. NOTE! As used throughout this document.

excluding acronyms.

words appearing all capitalized denote the application of definitions as used in the Salem Technical Specifications

1

*

Salem ODCH Rev. 3 07130187 The liquid effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material relea*ses in accordance with the Salem Radiological Effluent Technical Specifications are summarized as follows: 1> R18 <Unit 1> and 2-R18 <Unit 2> provide the alarm and automatic termination of liquid radioactive material releases as required by Technical Specification 3.3.3.8. 1-R19 A181C1and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines. 2-R19 A181C and D provide this function for Unit 2. 2)

-The alarm functions ror the Service Water System are provided by the radiation monitors on the Containment Fan Cooler discharges C1-R 13 A181C1D and E for Unit 1 and 2-R 13 A181and C for Unit 2>. Releases from the secondary system are routed through the Chemical Waste Basin where the effluent is monitored Cwith an alarm function>

by R37 prior to release to the environ*ent.

Liquid radioactive waste flow diagrams with the applicable, associated radiation

onitoring instrumentation and controls are presented ae Figures 1-1 and 1-2 for Units 1 and 2, respectively

2

*

Salem ODCH Rev. 3 07/30/87 Per the require*ents of Technical Specification 3.3.3.81 alarm setpoints shall be established for the liquid effluent monitoring instrumentation to ensure that the release concentration limits of Specification 3.11.1.1 are met (i.e ** the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 21 for radionuclides and 2.0E-04 uCi/al for dissolved or entrained noble gases>. The following equation*

ust be aatisfied to meet the liquid effluent restrictions:

where: c = c = f = C <F+f) (1.1) c { -------' the effluent concentration limit of Technical Specification (3.11.1.1>

implementing the 10 CFR 20 HPC for the site. in uCi/ml the setpoint1 in uCi/ml1 of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release: the setpoint.

represents a value which. if exceeded.

would result in concentrations exceeding the li*its of 10 CFR 20 in the UNRESTRICTED AREA the flow rate at the radiation monitor location.

in volume per unit ti*e* but in the sa*e units as F, below F = the dilution water flow rate as measured prior to the release point. in volu*e per unit ti*e [Note that if no dilution is provided, c{ C. Also. note that when <F> is large co*pared to (f), then <F + f) = F.l

Adapted fro* NUREG-0133 3

*

Salem Rev. 3 01130187 -1. 2.1 f

The setpoints for the liquid effluent monitors at the Sale* Nuclear Generating Station are determined by the following equations:

where: SP = HPCe = HPCe

SEN

CW SP ---------------

+ bkg RR ( 1.2) Ni th: [. Ci HPCe = ------------( 1.3) -( c* J L alar* setpoint corresponding to the maximum allowable release an effective HPC value for the mixture of radionuclides in effluent strea* CuCi/*1> rate (cpm) the Ci = the concentration of radionuclide i in the liquid effluent CuCi/*l>*

HPCi = SEN = cw = RR = bkg = *NOTE The concentration

ix must include the *ost recent co*posite of alpha e**itters1 Sr-89, Sr-901 Fe-55, and H-3 per Technical Specification 3.11.1.1

the HPC value corresponding to radionuclide i fro* 10 CFR 201 Appendix 81 Table 11, Colu*n 2 <uCi/*1> the sensitivity value to which the monitor is calibrated (cp* per uCi/*l> the circulating water flow rate (dilution water flow> at the tiae of release (gal/*in) the liquid effluent release rate (gal/*in) the background of the *onitor Ccp*) The radioactivity monitor setpoint equation <1.2> remains valid during outages when the circulating water dilution is potentially at its lowest value. Reduction of the waste strea* flow <RR> may be necessary during these periods to meet the discharge criteria.

However. in order to maximize the available plant discharge dilution and thereby minimize the potential offsite doses, releases fro* either Unit-1 or Unit-2 may be routed to either the Unit-1 or Unit-2 Circulating Water System discharge.

This routing is possible via interconnections between the Service Water Systems <see Figures 1 and 2>

4

*

Salem ODCH Rev. 3 07/30/87 Procedural restrictions prevent simultaneous releases fro* either a single unit or both units into a single Circulating Water Syste* discharge

1.2.2 Conservative alarm aetpoints may be deter*ined through the use of default para*eters.

Tables 1-1 and 1-2 su*marize all current default valuea in use for Salem Unit-1 and Unit-21 respectively.

They are based upon the following:

a> b) c> d) substitution of the effective HPC value with a default value of lE-05 uCi/*1 for radwaste releases (refer to Appendix A for justification>:

for additional substitution of the 1-131 HPC value 3E-07 uCi/ml for the R19 Stea* Generator blowdown monitors*

Service Water monitor and R37 Che*ical Waste Basin *onitor: substitutions of the operational circulating water flow with the lowest flow. in gal/*in: and, substitutiona of the effluent release rate with the highest allowed rate. in gal/*in. of R13 With pre-established alarm setpoints.

it is possible to control the radwaste release rate <RR> to ensure the inequality of equation <1.2> is *aintained under changing values for HPCe and for differing Circulating Water System dilutions.

Technical Specification 3.11.1.1 li*its the concentration of radioactive material in liquid effluents (after dilution in the Circulating Water System) to less than the concentrations as specified in 10 CFR 201 Appendix B. Table 111 Column 2 for radionuclides other than noble gases. Noble gases are limited to a diluted concentration of 2.0E-04 uCi/ml. Release rates are controlled and

Use of the effective HPC value as derived in Appendix conservative for the R19 Steam Generator blowdown monitors Waste Basin *onitors where 1-131 transfer during primary to may potentially be *ore controlling

5 A may and R37 be Chem i cal secondary leakage

*

Salem ODCM Rev. 3 07130187 radiation monitor alarm setpoints are established as addressed above to ensure that these concentration limits are not exceeded.

However. in the event any liquid release results in an alarm setpoint being exceeded, an evaluation of co*Pliance with the concentration limits of Techn.ical Specification 3.11.1.1 may be performed using the following equation: 1 ( 1.4) where: Ci = actual concentration of radionuclide as measured in the undiluted liquid effluent (uCi/*l> HPCi = the HPC value corresponding to radionuclide from 10 CFR 20, Appendix e, Table lls Colu*n 2 <uCi/ml> = 2E-04 uCi/*l for dissolved or entrained noble gases RR = the actual liquid effluent release rate (gal/*in>

CW = the actual circulating water flow rate (dilution water flow) at the ti*e of the release <gal/min) 6

*

Salem ODCH Rev. 3 07/30/87 1.4.1 Technical Specification -3.11.1.2 limits the dose or dose commitment to HEHBERS OF THE PUBLIC fro* radioactive materials in liquid effluents from each unit of the Sale11 Nuclear Generating Station to: during any calendar quarter: 1. S 11re11 to total body per unit 5.0 mrem to any organ per unit during any calendar year: 3.0 11re11 to total body per unit 10.0 mrem to any organ per unit. Per the surveillance require11ents of Technical Specification 4.11.1.2.

the following calculation 11ethode 11ay be used for deter11ining the dose or dose co1111it11ent due to the liquid radioactive effluents fro* Sale11. where: Do = Aio = Ci = VOL = cw = 1.67E-02

VOL Do = --------------

CW * [ .CCi

Aio) ( 1.5) dose or dose com11it11ent to organ O* including total body (mrem> site-related ingestion dose co*11itment factor to.the total body or any organ o for rad i onuc l i.de i ( 11re11/hr per uC i /111 l > average concentration of radionuclide i* in undiluted liquid effluent representative of the volu11e VOL CuCi/111) voluae of liquid effluent released (gal> average circulating water discharge rate during release period Cga1/11in) 1.67E-02 = conversion factor Chr/11in)

The site-related ingestion dose/dose commitment factors CA > are presented in io Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:

7

*

Salem Rev. j Aio = 1.14E+05 CCUI

Bii> + CUF

Bfi)l DFi ( 1.6) where: Aio = composite dose parameter for the total body or critical organ o 1.14E+05 UI = Bii = UF = BFi = DFi = of an adult for radionuclide i1 for the fish and invertebrate ingestion pathways <*rem/hr per uCi/*l> = conversion factor CpCi/uCi

ml/kg + hr/yr) adult invertebrate consumption CS kg/yr) bioaccu*ulation factor for radionuclide i in invertegrates from Table 1-4 CpCi/kg + pCi/l) adult fish consu11ption

<21 kg/yr) bioaccumulation factor for radionuclide in fish from Table 1-4 CpCi/kg per pCi/1) dose conversion factor for nuclide i for adults in pre-selected organ, 01 from Table E-11 of Regulatory Guide 1.109 <mrem/pCi)

The radionuclides included in the periodic dose assessment per the requirements of Technical Specification 3/4.11.1.2 those as identified by gamma spectral analysis of the liquid waste samples collected and analyzed per the requirements of Technical Specification 3/4.11.1.1, Table 4.11-1. Radionuclides requiring radiochemical analysis (e.g., Sr-89 and Sr-90> will be added to the dose analysis at a frequency consistent with the required minimum analysis frequency of Technical Specification Table 4.11-1. In lieu of the individual radionuclide dose assessment as presented in Section 1.4.1, the following si*plified dose calculational equation may be used for demonstrating compliance with the dose limits of Technical Specification 3.11.1.2.

<Refer to Appendix B for the derivation and justification for this simplified method.> 1.21E+03

VOL Dtb = --------------

cw ( 1. 7) L_ Ci 8

where: Ci = VOL = cw = Dtb -D*ax = 1.21E+D3 = 2.52E+D4 = *

Salem ODCH Rev. 3 07/30/87 2.52E+D4

VOL D*ax = --------------

( 1.8) cw average concentration of radionuclide i* in undiluted liquid effluent representative of the volu*e VOL <uCi/*l> volu*e of liquid effluent released (gal> average circulating water discharge rate during release period (gal/*in) conservatively evaluated total body dose <*re*> conservatively evaluated

axi*u* organ dose <*re*> conversion factor (hr/min) and the conservative total body dose conversion factor*<Fe-59, total body --7.27E+04 *rem/hr per uCi/*l > conversion factor <hr/*in) and the conservative 11axi11tu*

organ dose conversion factor <Nb-95, GI-LLI --1.51E+06 *re*/hr per uC il*l > 9

*

Salem OOCM Rev. 3 07/30/87 1.5

__

During periods of primary to secondary leakage (i.e ** steam generator tube leaks>. radioactive material will be transmitted from the primary system to the secondary system. The potential exists for the release of radioactive material to the off-site environment

<Delaware River*) via secondary system discharges.

Potentially significant radioactive material levels and potential releases are controlled/monitored by the Steam Generator blowdown monitors <R19) and the Chemical Waste Basin monitor <R37>. However to ensure compliance with the regulatory limits on radioactive material releases.

it may be desirable to account for potential releases from the secondary system during periods of primary to secondary leakage. Any potentially significant releases will be via the Cheaical Waste Basin with the major source of activity being the Steam Generator blowdown.

With identified radioactive material levels in the secondary system. appropriate samples should be collected and analyzed for the principal gamma emitting radionuclides.

Based on the identified radioactive material levels and the volume of water discharged, the resulting environmental doses may be calculated based on equation <1.5). Because the release rate from the secondary system is indirect (e.g., SG blowdown is nor*ally routed to condenser where the condensate clean-up syste* will reaove much of the radioactive

aterial>*

samples should be collected from the final releage point Che*ical Waste Basin) for quantifying the radioactive

aterial releases.

However. for conservatism and ease of controlling and quantifying all potential release paths* it is prudent to sample the SG blowdown and to assume all radioactive material is released directly to 10

*

Salem ODCH Rev. 3 07/30/87 the environment via the Chemical Waste Basin. This approach while not exact, is conservative and ensures timely analysis for regulatory compliance.

Accounting for radioactive material retention of the condensate clean-up system ion exchange resins *BY be needed to more accurately account for actual releases

11

*

Salem OOCH Rev. 3 07/30/87 Technical Specification 3.11.1.3 requires that the liquid radioactive waste processing syste* be used to reduce the radioactive material levels in the liquid waste prior to release when the quarterly projected doses exceed: 0.375 *re* to the total body1 or 1.25 mre* to any organ. The applicable liquid waste processing system for maintaining radioactive material releases ALARA is the ion exchange system as delineated in Figure 1-3. Alternately, the waste evaporator as presented in the Salem FSAR has processing capabilities meeting the NRC ALARA design requirements and may be uaed in conjunction or in lieu of the ion exchange syste* for waate processing require-ments in accordance with Technical Specification 3.11.1.3.

Theae processing require*ents are applicable to each unit individually.

Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing require*ents for the other unit. Dose projections are *ade at least once per 31 days by the following equations:

Dtbp = Otb (91 t d) (1.9) Omaxp = Dmax <91

d> (1.10> where: Dtbp = the total body doae projection for current calendar quarter <*re*> Dtb = the total body dose to date for current calendar quarter as determined by equation <1.S> or <1.7> <*re*> Dmaxp = the maximum organ dose projection for current calendar quarter <*re*> Dmax = the maximum organ dose to date for current calendar quarter as deter*ined by equation (1.5) or (1.8) <mre*> d = the number of days to date for current calendar quarter 91 = the number of days in a calendar quarter 12

*

Salem ODCM Rev. 3 07/30/87 / The gaseous effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Radiological Technical Specifications are summarized as follows! 1>

-The vent header gases are collected by the waste*gas holdup system. Gases may recycled to provide cover gas for the eves hold-up tank or held in the waste gas tanks for decay prior to release. Waste gas decay tanks are batch released after sampling and analysis.

The tanks are discharged via the Plant Vent. 1-R41C provides noble gas monitoring and automatic isolation of waste gas decay tank releases for Unit-1: this function is provided by 2-R41C for Unit-2. 2>

-Containment purges and pressure/vacuum reliefs are released to the atmosphere via the respective unit Plant Vent. Noble gas monitoring and auto isolation function *are provided by 1-R41C for Unit-1 and 2-R41C for Unit-2. Additionally, in accordance with Technical Specification 3.3.3.9, Table 1-R12A and 2-R12A may be used to provide the containment.monitoring and automatic isolation function during purge and pressure/vacuum reliefs.*

3>

-The Plant Vent for each respective unit receives discharges from the waste gas hold-up systems condenser evacuation system, containment purge and pressure/vacuum reliefs, and the Auxiliary Building ventilation.

Effluents are monitored by R41C, a flow through gross activity monitor (for noble gas monitoring).

Additionally, in-line gross activity monitors <1-R16 and 2-R16) provide redundant back-up capabilities to the R41C monitors.

Radioiodine and particulate sampling capabilities are provided by charcoal cartridge and filter medium samplers with redundant back-up sampling capabilities provided by R41B and R41A, respectively.

Plant Vent flow rate is measured and as a back-up may be determined empirically as a function of fan operation (fan curves>. Sampler* flow rates are determined by flow rate instrumentation

<e.g., venturi rotometer>.

The R12A monitors also provide safety function of containment isolation in the event of a fuel handling accident during refueling.

During HOOE 6 in accordance with Technical Specification 3/4.3.3, Table 3.3-6, the R12A alarm/trip setpoint shall be established at twice backgrounds providing early indication and containment isolation accompanying unexpected increases in containment airborne radioactive material levels indicative*

of a fuel degradation.

The R41C monitor may provide this function if the. R12A monitor is inoperable during HOOE 6

13

*

ODCM Rev. 3 07/30/87 A gaseous radioactive waste flow diagrams with the applicable.

associated radiation monitoring instrumentation and controls are presented as Figures 2-1 and 2-2 for Units 1 and 2, respectively

14

*

Salem ODCH Rev. 3 07/30/87 2.2.1 Per the requirements of Technical Specification 3.3.3.9, alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of Specification 3.11.2.1, which corresponds to a dose rate at the SITE BOUNDARY of 500 mrem/year to the total body or 3000 mrem/year to the skin. Based on a grab sample analysis of the applicable release (i.e., grab sample of the Containment atmosphere, waste gas decay tank, or Plant Vent), the radiation monitoring alarm setpoints may be established by the following calculational method. The measured radionuclide concentrations and release rate are used to calculate the fraction of the allowable release rate, as li*ited by Specification 3.11.2.1, by the equation:

FRAC = C4.72E+02

X/Q

VF* L CCi

Ki)J + 500 <2.1> FRAC = C4.72E+02

X/Q *VF* [. CCi *<Li+ 1.1 Hi))J + 3000 <2.2> where: FRAC = X/Q = VF = Ci = Ki = Li = Hi = 1.1 = 4.72E+02 500 = 3000 = fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate annual average meteorological dispersion to the controlling site boundary location (sec/m3) , ventilation system flow rate for the applicable release point and monitor (ft3/*in) concentration of noble gas radionuclide as determined by radioanalysis of grab sa*ple (uCi/c*3>

total body dose conversion factor for noble gas radionucl i-de <*re*/yr per uCi/*3, fro* Table 2-1> beta skin dose conversion factor for noble gas radionuclide (nre*/yr per uCi/*3, from Table 2-1> gam*a air dose conversion factor for noble gas radionuclide (mrad/yr per uCi/m3, from Table 2-1) mrem skin dose per mrad gam*a air dose (mrem/mrad)

= conversion factor <c*3/ft3

min/sec> total body dose rate limit (mre*/yr) skin dose rate li*it (*re*/yr) 15

Salem OOCWRev. 3 07/30/87 Based on the more limiting FRAC (i.e ** higher value) as determined above. the alarm setpoints for the applicable monitors <Rl61 R41Cs and/or R12A> may be calculated by the equation:

where: SP = SEN = bkg = AF = SP = CAF * '[Ci

SEN + FRACJ + bkg (2.3> alarm setpoint corresponding to the maximum allowable release rate (Cp*) monitor sensitivity (cpm per uCi/c*3> background of the monitor (Cp*) administrative allocation factor for the specific monitor and type release. which corresponds to the fraction of the total allowable release rate that is administratively allocated to the release. The allocation factor <AF> is an administrative control imposed to ensure that combined releases from Salem Units 1 and 2 and Hope Creek will not exceed the refulatory limits on release rate from the site (i.e., the release rate limits of Technical Specification 3.11.2.1).

Normally, the combined AF value for Salem Units 1 and 2 is 0.5 (0.25 per unit), with the remainder 0.5 allocated to Hope Creek. Any increase in AF above 0.5 for the Salem Nuclear Generating Station will be coordinated with the Hope Creek Generating Station to ensure that the -combined alloiation factors for all units do not exceed 1.0. 2.2.2 A conservative alarm setpoint can be established.

in lieu of the individual radionuclide evaluation based on the grab sample analysis*

to eliminate the potential of periodically having to adjust the setpoint to reflect minor changes in radionuclide distribution and variations in release flow rate. The alarm setpoint may be conservatively determined by the default presented in Table 2-1 and 2-2 for Units 1 and 21 respectively

16

*

Salem ODCH Rev. 3 07/30/87 These values are based upon: the maximu* ventilation Cor purge) flow rate: a radionuclide distribution*

comprised of 95X Xe-133, 2X Xe-135. 11 Xe-133** 11 Kr-88 and 11 Kr-85: and an administrative allocation factor of 0.25 to conservatively ensure that any simultaneous releases from Salem Units 1 and 2 do not exceed the maxi*um allowable release rate. For this radionuclide distribution.

the alarm setpoint based on the total body dose rate is more restrictive than the corresponding setpoint based on the skin dose rate. The resulting conservative.

default setpoints are presented in Tables 2-2 and 2-3 * ------------------------

Adopted from ANSI N237-1976/ANS-18.1, Source Term Table 6 17

* * -Salem ODCH Rev. 3 07/30/87 limits the dose rate at the SITE BOUNDARY due to noble gas releaaes to {500 *re*/yr1 total body and mrem/yr, skin. Radiation monitor alar* aetpoints are established to ensure that these release limits are not exceeded.

In the event any gaseous releases from the station results in an alarm aetpoint being exceeded.

an evaluation of the SITE BOUNDARY doae rate resulting fro* the release may be performed using the following equations:

Otb = X/Q *I:: CKi *Qi) <2.4) and Os = X/Q

l:_ CCLi + 1.1Hi)

Qi) (2.5) where: Otb = total body dose rate <*re*/yr)

Da = skin dose rate <*re*/yr)

X/Q = at*ospheric dispersion to the controlling SITE BOUNDARY location Qi Ki Li c*eec/*3>

= average release rate of radionuclide i over the release period under evaluation CuCi/sec>

= total body dose conversion factor for noble gas radionuclide Cmre*/yr.per uCi/*31 fro* Table 2-1> = beta skin dose conversion factor for noble gas radionuclide Cmre*/yr per uCi/*31 fro* Table 2-1> Hi = gamma air dose conversion factor for noble gas radionuclide Cmrad/yr per uCi/*31 fro* Table 2-1) 1.1 = 11rem skin dose per 111rad gam*a air dose C111re*/mrad)

As appropriate.

simultaneous releases from Salem Units 1 and 2 and Hope Creek will be considered in evaluating compliance with the release rate limits of Specification 3.11.2.1a, following any release exceeding the above prescribed alar* setpointa.

Honitor indications

<readings) may be averaged over a time period not to exceed 15 minutes when determining noble gas release rate based on correlation of the *onitor reading and monitor sensitivity.

The 15 *inute averaging is needed to allow for reasonable monitor response to potentially changing radioactive

aterial concentrations and to exclude potential electronic 18-

*

Salem ODCH Rev. 3 07/30/87 spikes in monitor readings that may be unrelated to radioactive material releases

As identified, any electronic spiking monitor responses

BY be excluded from the analysis.

NOTE: For ad*inistrative purposes.

more conservative alarm setpoints than those as prescribed

'above may be imposed. However. conditions exceeding these more limiting alar* setpoints do not necessarily indicate radioactive material release rates exceeding the li*its of Technical Specification 3.11.2.1a.

Provided actual releases do not result in radiation monitor indications exceeding alarm setpoint values based on the above criteria.

no further analyses are required for de*onstrating compliance with the limits of Specification 3.11.2.1a.

Actual meteorological conditions concurrent with the release period or the default. annual average dispersion parameters as presented in Table 2-4 *BY be used for evaluating the gaseous effluent dose rate. 2.3.2 Technical Specification 3.11.2.1.b limits the dose rate to mrem/yr to any organ for 1-131, tritium and particulates with half-lives greater than 8 days. To demonstrate Co*pliance with this limit. an evaluation is performed at a frequency no greater than that corresponding to the sampling and analysis time period (e.g., nominally once per 7 days). The following equation may be used for the dose rate evaluation:

where: Do = X/Q = Ri = Qi = Oo = X/Q

l:_ CRi

Qi) (2.6) average organ doae rate over the sampling time period (mrem/yr) atmospheric dispersion to the controlling SITE BOUNDARY location for the inhalation.

pathway <sec/*3> dose parameter for radionuclide i1 (mrem/yr per uCi/*3> for the child inhalation pathway fro* Table 2-5 average release rate over the appropriate sampling period and analysis frequency for radionuclide i --I-131, 1-1331 tritiu* or other radionuclide in particulate for* with half-life greater than 8 days <uCi/sec>

19

*

Salem ODCH Rev. 3 07/30/87 By substituting 1500 mrem/yr for Do and solving for Q, an allowable release rate for 1-131 can be determined

Based on the annual average meteorological dispersion

<see Table 2-4> and the most limiting potential pathway, age group 3 and organ (inhalation, child, thyroid --Ri = 1.62E+07 mrem/yr per uCi/m ), the allowable release rate for 1-131 is 42 uCi/sec. Reducing this release rate by a factor of 4 to account for potential dose contributions from other radioactive particulate material and other release points *ce.g., Hope Creek), the corresponding release rate allocated to each of the Salem units is 10.S uCi/sec. For a 7 day period. which is the nominal sampling and analysis frequency for I-131. the cumulative release is 6.3 Ci. Therefore.

as long as the 1-131 releases in any 7 day period do not exceed 6.3 Ci1 no additional analyses are needed for verifying compliance with the Technical Specification 3.11.2.1.b limits on allowable release rate

20

* * -Salem ODCH Rev. 3 07130187 2.4.1 Technical Specification 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate compliance with the quarterly dose limits of mrad, gamma-air and mrad, beta-air and the calendar year limits mrad, gamma-air and mrad, beta-air.

The limits applicable separately to each unit and not combined site limits. The following equations may be used to calculate the gamma-air and beta-air doses: Dg = 3.17E-08

X/Q * [ <Hi

Qi) (2.7) and Db =. 3.17E-08

X/Q

C <Ni *Qi) (2.8) where: Dg = air dose due to gamma emissions for noble gas radionuclides (mrad) Db = air dose due to beta emissions for noble gas radionuclides (mrad) X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location Qi Hi Ni <sec/a3> = cumulative release of noble gas radionuclide i over the period of interest CuCi) = air dose factor due to gamma emissions from noble gas radionuclide i (arad/yr per uCi/m3, from Table 2-1> = air dose factor due to beta emissions from noble gas radionuclide (mrad/yr per uCi/m3, Table 2-1> 3.17E-08 = conversion factor. (yr/sec> In lieu of the individual noble gas radionuclide dose assessment as presented above, the following simplified dose calculational equations may be used for verifying compliance with the dose limits of Technical Specification 3.11.2.2.

<Refer to Appendix C for the derivation and justification for this simplified aethod.> 21

*

where: He ff = Neff = Qi = a.so = Salem ODCH Rev. 3 07/30/87 3.17E-08 Og = --------* X/Q

HeH * 'L Qi (2.9) a.so and 3.17E-08 Ob = --------* X/Q

Neff * 't. Qi a.so <2.10) S.3E+02. effective ga1111a-air dose factor <mrad/yr per uCi/m3> 1.1E+03. effective beta-air dose factor (mrad/yr per uCi/m3> cumulative release for all noble gas radionuclides CuCi) conservatism factor to account for potential variability in the radionuclide distribution Actual ' meteorological conditions concurrent with the release period or the default. annual average dispersion parameters as presented in Table 2-4, may be used for the evaluation of the gamma-air and beta-air doses

22

*

where: He ff = Neff = Qi = o.so = Dg Db Salem ODCM Rev. 3 07/30/87 3.17E-08 = --------* X/Q

Heff

I: Qi a.so and 3.17E-08 = --------* X/Q

Neff

t_ Qi a.so (2.9) <2.10) -s.3E+021 effective gamma-air dose factor (mrad/yr per 1.1E+Q3, effective beta-air dose factor (mrad/yr per uCi/m > cumulative release for all noble gas radionuclidea CuCi> conservatism factor to account for potential variability in the radionuclide distribution Actual meteorological condition*

concurrent with the release period or the default, annual average dispersion parameters aa presented in Table 2-4, may be used for the evaluation of the gamma-air and beta-air doaea

22

*

Salem ODCH Rev. 3 07/30/87 2.5 Radioiodine and Particulate Dose Calculations

-10 CFR 50 2.5.1 UNRESTRICTED AREA Ooae -Radioiodine and Particulates.

In accordance with requirements of Technical Specification 3.11.2.31 a periodic aeaeasment shall be perforaed to evaluate compliance with the quarterly dose limit of i7.5 mrem and calendar year limit i15 mrem to.any organ. The following equation may be used to evaluate the maximum organ dose due to releases of I-1311 tritium and particulates with half-lives greater than 8 days: Daop = 3.17E-08

W

SFp

L. CRi

Qi> <2.11) where: Daop = dose or dose commitment via controlling pathway p and age group a <a* identified in Table 2-4> to organ o. including the total body <111rem) Ri = atmospheric dispersion parameter to the controlling location(e) as identified in Table 2-4 = X/Q = atmospheric dispersion for inhalation

!athway and dose contribution via other pathways Csec/m ) D/Q = atmospheric deposition milk and ground plane exposure pathways Cm 3 2 dose factor for radionuclide i1 <mrem/yr per uCi/m) or (m mrem/yr per uCi/mec> from Table 2-5 for each age group a and the applicable pathway pas identified in Table 2-4. Values for R. were derived in accordance with the methods described in NUREG! 0133. Qi = cumulative releaee over the period of interest for radionuclide i --I-131 or radioactive material in particulate form with half-1 ife greater than 8 days CuCi). SFp = annual *eaeonal correction factor to account for the fraction of the year that the applicable expo*ure pathway does not exist. 1> For *ilk and vegetation exposure pathways:

= A eix month fresh vegetation and grazing season <Hay through October) = o.s 2) For inhalation and ground plane exposure pathways:

= 1.0 For evaluating the 111axi111um exposed individual 1 the infant age group is controlling for the milk pathway and the child age group is controlling for the 23 Salem ODCH Rev. 3 07/30/87 vegetation pathway. Only the controlling age group and pathway aa identified in

Table 2-4 need be evaluated for compliance with Technical Specification

3.11.2.3.

2.5.2 Simplified Dose Calculation for Radioiodinea and Particulatea.

In lieu of the individual radionuclide CI-131 and particulates>

doae aaseaament aa presented above. the following simplified doae calculational equation may be uaed for -verifying compliance with the doae limita of Technical Specification 3.11.2.3 (refer to Appendix D for the derivation and justification of thia simplified method). where: Dmax * = RI-131 = = w = Qi = Dmax = 3.17E-08

W

SFp

RI-131 .-[_Qi <2.12) maximum organ dose <*rem> I-131 dose parameter for the thyroid for the identified controlling pathway 1.05E+121 thyroid doae parameter with the cow-milk pathway controlling (m -mrem/yr per uCi/eec) D/Q for radioiodine1 2.lE-10 1/* cumulative releaae over the period of interest for radionuclide i --I-131 or radioactive material in particulate form with half life greater than 8 daya (uCi) The location of expoeure pathways and the maximum organ do*e calculation may be based on the available pathway* in the surrounding environ*ent of Salem as identified by the annual land-u*e census <Technical Specification 3.12.2>. Otherwise, the doae will be baaed on the 9redetermined controlling pathwaya as identified in Table

24

*

Salem ODCH Rev. 3 07/30/87 2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations During periods of primary to secondary leakage, minor levels of radioactive material may be released via the secondary system to the atmosphere.

Non-condensables

<e.g., noble gases) will be predominately released via the condenser evacuation system and will be monitored and quantified by the routine plant vent Monitoring and sa*pling syetem and procedures (e.g., R15 on condenser evacuation.

R41C on plant vent. and the plant vent particulate and charcoal samplers).

However. if the Steam Generator blowdown is routed directly to the Chemical Waste Basin (via the SG blowdown flash tank) instead of being recycled through the condenser, it may be desirable to account for the potential atmospheric releases of radioiodines and particulates from the flash tank vent Ci.e ** releases due to moisture carry over). Since this pathway is not sampled or monitored.

it is necessary to calculate potential releases

Based on the guidance in NRC NUREG-01331 the releases of the radioiodines and particulates may be calculated by the equation:

Qi = Ci

Rsgb

Fft

C1-SQftv>

<Z.13> where: Qi = the release rate of radionuclide.

i1 from the steaa generator flash Ci = Rsgb = Fft = SQftv = tank vent CuCi/aec>

the concentration of radionuclide.

i1 in the secondary coolant water averaged over not *ore than one week CuCi/ml> the stea* generator blowdown rate to the flash tank (ml/sec> the fraction of blowdown flashed in the tank determined fro* a heat balance taken around the flash tank at the applicable reactor power level the measured steam quality in the flash tank vent: or an assumed value of 0.851 based on NUREG-0017

25

* *** Salem ODCH Rev. 3 07/30/87 Tritium releases via the steam flashing may also be quantified using the above equation with the aseumption of a steam quality <SQftv) equal to O. Since the H-3 will be associated with the water molecules.

it is not necessary to account for the moisture carryover which is the transport media for the radioiodinea and particulates.

Based on the design and operating conditions at Sale** the fraction of blowdown converted to steam <Fft> is approximately 0.48. following:

Qi = 0.072 Ci Rsgb For H-3, the simplified equation is! Qi = 0.48 Ci Rsgb The equation simplifies to the <2.14) (2.15) Also during reactor shutdown operations with a radioactively contaminated secondary system. radioactive material may be released to the atmosphere via the atmospheric reliefs <PORV> and the safety reliefs on the main steam lines and the steam driven auxiliary feed pump exhaust. The evaluation of the radioactive material concentration in the steam relative to that in the steam generator water is based on the guidance of NUREG-00171 Revision 1. The partitioning factors for the radioiodines is 0.01 and is 0.001 for all other particulate radioactive Material.

The resulting equation for quantifying releases via the atmo*pheric stea* releases is: 26

*

where: Qij SFj PF Salem ODCH Rev. 3 07/30/87 Qi = 0.13

L_ CCij

SFj) *PF =release rate of radionuclide i via pathway j <uCi/sec)

= steam flow for release pathway j = 4501000 lb/hr per PORV = 8001000 lb/hr per safety relief valve = 501000 lb/hr for auxiliary feed pump exhaust = partitioning factor, ratio of concentration in steam to that in the water in the steam generator

= 0.01 for radioiodines

= 0.005 for all other particulates

= 1.0 for H-3 (2.16) Any significant releases of noble gases via the atmospheric steam releases can be quantified in accordance with the calculation methods of the Salem Emergency Plan Implementation Procedure.

Alternately, the quantification of the release rate and cumulative releases may be based on actual samples of main steam collected at the R46 sample locations.

The measured radionuclide concentration in the steam may be used for quantifying the noble gases, radioiodine and particulate releases.

Note: The expected mode of operation wouid be to isolate the effected steam generator.

thereby reducing the potential releases during the shutdown/cooldown process. Use of the above calculational methods should consider actual operating conditions and release mechanisms.

The calculated quantities of radioactive materials may be used as inputs to the equation <2.11> or C2.12) to calculate offsite doses for demonstrating compliance with the Radiological Effluent Technical Specifications

27 l

*

Salem ODCH Rev. 3 07/30/87 2.7 Gaaeoua Effluent Doae Projection Technical Specification 3.11.2.4 requires that the GASEOUS RADWASTE TREATHENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce radioactive material levels prior to discharge when projected dosee exceed one-half the annual deaign objective rate in any calendar quarter, i.e ** exceeding:

0.625 mrad/quarter.

gamma air: 1.25 mrad/quarter, beta air: or

re*/quarter.

maximum organ. The applicable gaseous proceeeing ayete*a for *aintaining radioactive material releaaee ALARA are the Auxiliary Building normal ventilation eyatem (filtration syete*a M 112 and 3) and the Waate Gaa Decay Tanke ae delineated in Figures 2-3 and 2-4. Doee projection&

are perfor*ed at least once per 31 days by the following equationa:

D gp = 09 * <91+ d) (2.17> D bp = Db * (91+ d> (2.18) Dmaxp = D*ax * (91+ d> <2.19) where: D gp = Og D bp = Db = Dmaxp = D*ax = d = 91 = = 9am*a air doee projection for current calendar quarter Cmrad) 9a**a air doae to date for current calendar quarter aa determined by equation <2.7> or <2.9) <*rad) beta air doae projection for current calendar quarter <*rad) beta air doee to date for current calendar quarter ae determined by equation <2.8) or <2.10) (mrad> maximu* organ doae projection for current calendar quarter (mrem> maximu* organ doae to date for current calendar quarter aa determined by equation <2.11) or <2.12> (mre*> nu*ber of daya to date in current calendar quarter number of daya in a calendar quarter 28 Salem OOCH Rev. 3 07/30/87 3.0 Special Dose Analyses

3.1 Doses Due To Activities Inside the SITE BOUNDARY In accordance with Technical Specification 6.9.1.11.

the Radioactive Effluent Re)ease Report <RERR> submitted within 60 days after January 1 of each year shall include an aseeesment of radiation doses fro* radioactive liquid and gaseoue effluent*

to HEHBERS OF THE PUBLIC due to their activities ineide the SITE BOUNDARY.

There is one location on Artificial Island that is acceesible to HEHBERS OF THE PUBLIC for unrelated to PSE&G operational and support activitiea.

This location i* the Second Sun <vieitor*s center> located near the contractors gate for the Salem Nuclear Generating Station. The calculation Methods ae presented in Section* 2.4 and 2.5 may be used for determining the maxi*um potential dose to a HEHBER OF THE PUBLIC based on the

para*eters fro* Table 2-4 and 2 houre per visit per year. The default value for the mereorological dispereion data as presented in Table 2-3 *BY be used if current year *eteorology in unavailable at the ti*e of NRC reportin.

However. a follow-up evaluation shall be performed when the data beco*ee available

* . 29

*

Salem ODCH Rev. 3 07/30/87 3.2 Total dose to HEHBERS OF THE PUBLIC -40 CFR 190 The Radioactive Effluent Releaee Report <RERR> submitted within 60 days after January 1 of each year ahall aleo include an aseess*ent of ihe radiation doae to the likely moat exposed HEHBER OF THE PUBLIC for reactor releaeea and other nearby uranium fuel cycle sources (including doae contributions from effluents and direct radiation from on-site sources>.

For the likely most expoeed HEHBER OF THE-PUBLIC in the vicinity of Artificial Ieland1 the source* of exposure need only consider the Sale* Nuclear Generating Station and the Hope Creek Nuclear Generating Station: No other fuel cycle facilities contribute to the HEHBER OF THE PUBLIC doae for the Artificial Ieland vicinity.

The doee contribution from the operation of Hope Creek Nuclear Generating Station will be estimated baaed on the *ethoda as presented in the Hope Creek Offaite Dose Calculation Hanual CHCGS ODCH>

As appropriate for de*onstrating/evaluating compliance with the limits of Technical Specification 3.11.4 <40 CFR 190)1 the result* of the environmental monitoring program *ay be used for providing data on actual measured levels of radioactive material in the actual pathways of exposure.

3.2.1 Effluent Doee Calculation*.

For purpoees of implementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting requirements of 6.9.1.11 CRERR>s doae calculations for the Salem Nuclear Generating Station may be performed using the calculational method* contained within thie ODCH: the conservative controlling pathway* and locations of Table 2-4 or the actual pathway* and locations ae identified by the land use ceneue <Technical Specification 3/4.12.2) may be uaed. Average annual meteorological diapereion 30

*

Salem ODCH Rev. 3 07/30/87 para*eters or *eteorological condition*

concurrent with the release period under evaluation may be used

3.2.2 Direct Exposure Pose Determination.

Any potentially significant direct exposure contribution to off-site individual doees may be evaluated based on the results of the environmental

easurementa

<e.g.. TLD1 Ion cha*ber meaaurements>

and/or by the uee of a radiation traneport and ehielding calculational method. Only during atypical condition*

will there exiet any potential for significant on-site eources at Sale* that would yield potentially significant off-site doees Ci.e., in excess of 1 mrem per year to a HEHBER OF THE PUBLIC>. that would require detailed evaluation for de*onstrating compliance with 40 CFR 190. However. should a situation exist whereby the direct exposure contribution is potentially significant.

on-site measure*ente.

off-site meaeurements and/or calculational techniques will be used for determination of dose for aseesaing 40 CFR 190 compliance

31

Salem ODCH Rev. 3 07/30/87 4.D Radiological Environ*ental Monitoring Program 4.1 Sa*pling Program The operational phaae of the Radiological Environmental Horiitorin9 Progra11 CREHP> ia conducted in accordance with the require11enta of Appendix A Technical Spec if icat ion 3 .12. The object i vee of the pro9ra11 are: -To determine whether *ny ei9nificant increaaea occur in the concentration of radionuclide*

in the critical pathway* of exposure in the vicinity of Artificial Ieland: -To deter11ine if the operation of the Salem Nuclear Generatin9 Stations ha* reeulted in any increase in the inventory of lon9 lived radionuclidea in the To detect any changes in the ambient ga1111a radiation levela: and -To verify that SNGS operation*

have no detrimental effecta on the health and eafety of the pub 1 ic or on the environ*ent.

The aa11plin9 require111enta

<type of sa*ples, collection frequency and analysis) and sa*ple location*

are presented in Appendix E

32

*

Salem OOCH Rev. 3 07/30/87 4.2 Interlaboratorv Comparison Proqra* Technical Specification 3.12.3 requires analyses be perforaed on radioactive material supplied as part of an Interlaboratory Comparison.

Participation in an approved Interlaboratory Comparison Prograa provides a check on the precisness of measurements of radioactive materials in environmental samples. A summary of the Interlaboratory Comparison Program results will be provided in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.9.1.10

33

ro -If.I FUER iii RADIATION MONITOR £::, INSTRUMENT PANEL -#-PNEUMATIC -ELECTiaC.AL

-+--UOUID EFFLUENT -+-GAS EFFLUENT ... .. ... ... ... ... *M l EfR.UENT -+-GAS EfR.UENT

121 <XlHlllNSER CtlNdlHSlff ... llMSlll 11t£11t1 .. ntilS1l1 ........ ... .. '"' -.. 111 -..
Monitoring for the 124 and 125 containment fan coil units is provided by the Rl3A, B or C monitors through interconnects.
121 SUIVIC£
nSUNU *AIOI HEADBI * .,..._, PStiGCO -1.IL. Mlil"'1. .... .. .... .... ..,... BtlC .. _ .. ,. ...,,. .... ....,.. ... ...,,. .... .. ... -AA1Rr ... ... ... -....... .... --... .... ... , ----.. , -------tO -....... ----------... -* --* ...., ....... ----121XN> alC WIR OliDi
FOR INFORMATION ONLY FIGURE 1-2
RADIATION MONITORING
... """" UOUID °"""""'*"'
2UNIT
I I I I I I WU HUT OTHER UNIT Wl.177 1 VVMT Wl.47 Wl 8ti I TO&FROM OTHER UNIT 2VVMT L-----IWi.2611 Wl.115 TOANDFROM 1---ciO-+
OTHER UNIT SERVICE WAJER SW222 Wl..84 WASTEEVAP FEEOPUMP --------------------------------------
FROM LAUNDRY & CHEM TKS Wl68 Wl172 FROM llAEVN' Wl70 I I I 1WHUT "' Wl71 2WHUT Wl.178 Wl.11 **-2UNIT WHUTS Wl71 Wl73 ... I WASTEEVAP:
I I FC EOUIPORS AUX BlllG SU ... TK FLOOR DRS Cl/CS HUT RELIEFS RHRSUMl'S RWPP Wl222 RWST WI.Ill:! CllCSHUT COHTSUMP 'f' 2Wl.174 2Wl.193 Wl.7 RCPHEADTKS RC LOOP ORS RX FLANGE LO ACCUMUUUOR ORS EXCESS LET DOWN RCDT RCPS 3SEALLO OllERllOARO ORWMTS 12 2lWl6 -II WASTE EVAP Wl.183 21 ORC WAJER DISCH I.WI& ILW41 Wl.1711 2WMHUT PUMP Wl.11 Wl.77
FOR INFORMATION ONLY FIGURE 1-3 LIQUID WASTE __ ____j
*
Sale1 ODc"* Rev. 3 07/30/87 Table 1-1 Para1eters for Liquid Alar1 Setpoint Deter1inations Unit 1 Para1eter Actual Default Units Co11ents Value Value ----------------------------------

"Pee calculated 1E-05 1. uCi/ml calculate for each batch to be released KPCI-131 3E-07 N/A uCi/11 I-131 "PC conservatively used for SG blow-down and Service Water 1onitor setpoints Ci measured N/A uCi/11 . taken from ga11a spectral analysis of liquid effluent KP Ci as N/A uCi/11 taken f ro1 10 CFR 20, Appendix 8, Table II, deter1ined Col. 2. SEN 1-R18 as 2.9E+07 CPI per uCi/111 radwaste effluent !Cs-137) deter1ined 1-R19 2.9E+07 Stea1 Generator blowdown !Cs-137! !A,8,C,Dl 1-R13 1.2E+08 Service Water -Contain1ent fan cooling !A,B,C,D,El

!Cs-137! cw as 1.85E+05 gp; Circulating Water Syste1, single CW pu1p deter1ined RR 1-R18 as 120 .. gp1 deter11ined prior to release; release rate deter1ined can be adjusted for Technical Specification CDIPliam 1-R19 80 Stea11 Generator blowdown rate per generator 1-R13 2500 Service Water flow rate for Contain1ent fan coolers SP 1-R18 calculated 4.4E+05!+bkgl CPI Default alar1 setpoints; 1ore conservative values 1ay be used as dee1ed appropriate and 1-R19H calculated 2.0E+04!+bkgl desirable for ensuring regulatory co1pliance 1-R13H calculated and for 1aintaining releases ALARA. 2.6E+03!+bkgl

' Refer tu Appendix A for derivation ff The KPC values of 1-131 IJE-07 uCi/mll has been used for derivation of the R19 Stea1 Generator blowdown and R13 Service Water monitor setpoints as discussed in Section 1.2.2
37
*
Sale1 OOC" Rev. 3 07/30/87 Para1eters for Liquid Alarm Setpoint Deter1inations Unit 2 Para1eter Actual Value Default Value "PCe calculated 1E-05 1 "PCI-131 3E-07 N/A Ci 1easured N/A "PCi as N/A deter1ined SEN 2-R18 as 8.8E+07 2-R19 fA,B,C,Ol 2-R13 fA,B,C,D,El R37 deter1ined 8.8E+07 8.8E+07 8.8E+07 cw as 1.85E+OS deter1ined RR 2-R18 as 2-R19 *2-RtJ R37 deter1ined 120 80 2500 300 Units Co11ents ----------

uCi/11 calculate for each batch to be released uCi/11 I-131 "PC conservatively used for SG down, Service Water and Che1ical Waste Basin 1onitor setpoints uCi/11 taken fro1 ga11a spectral analysis of liquid effluent uCi/11 taken fro1 10 CFR 20, Appendix B, Table II, Col. 2. cp1 per uCi/11 radwaste effluent !Cs-137) gp1 gp1 Stea1 Generator blowdown !Cs-137) Service Water -Contain1ent fan cooling !Cs-137) Che1ical Waste Basin (Cs-137) Circulating Water Syste1, single CW pu1p deter1ined prior to release; release rate can be adjusted for Technical Specification co1pliance Stea1 Generator blowdown rate per generator Service Water flow rate for Contain1ent fan coolers Che1ical Waste Basin discharge SP 2-R18 calculated.
8.0E+OS!+bkg) cp1 Default alar1 setpoints; 1ore conservative values 1ay be used as dee1ed appropriate and
desirable for ensurfng regulatory co1pliance and for 1aintaining releases ALARA. 2-Rt911 calculated 6.1E+04!+bkg) 2-Rt311 calculated 1.9E+OJ!+bkg)
R371111 calculated t.6E+04!+bkgl 1 Refer to Appendix A for derivation 11 Based on Cs-137 response 111 Actual calculated setpoint for 2-R18 (l.3E+06l is greater than the full scale 1onitor indicator, therefore, for conservatis1 the reco11ended setpoint has been reduced to 8.0E+OS cp1 1111 The "PC value of 1-131 l3E-07 uCi/1ll has been used for derivation of the R19 Stea1 generator . blowdovn, R13 Service Water and the R37 Che1ical Waste Basin 1onitor setpoints as discussed in Section 1.2.2 38 Salem ODCH Rev. 3 07/30/87 Table 1-3
Site Related Ingestion Dose Commitment Factors
Aio (mrem/hr per uCi/ml) Nuclide Bone Liver T.Body Thyroid Kidney Lung Gl-LLI --------------------------------------------------------H-3 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-1 C-14 1.45E+4 2.90E+3 2.90E+3 2.90E+3 2.90E+3 2.90E+3 2.90E+3 Na-24 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 P-32 4.69E+6 2.91E+5 1.81E+5 5.27E+5 Cr-51 5.58E+O 3.34E+O 1.23E+O_ 7.40E+O 1.40E+3 Hn-54 7.06E+3 1.35E+3 2.10E+3 2 .16E+4 Hn-56 1.78E+2 3 .15E+1 2.26E+2 5.67E+3
Fe-55 5.11E+4 3.53E+4 8.23E+3 1.97E+4 2.03E+4 Fe-59 8.06E+4 1.90E+5 7.27E+4 5.30E+4 6.32E+5 Co-57 1 * .42E+2 2.36E+2 3.59E+3 Co-58 6.03E+2 1.35E+3 1.22E+4 Co-60 1.73E+3 3.82E+3 3.25E+4 Ni-63 4.96E+4 3.44E+3 1.67E+3 7 .18E+2 Ni-65 2.02E+2 2.62E+1 1. 20E+1 6.65E+2 Cu-64 2.14E+2 1.01E+2 5.40E+2 1.83E+4 Zn-65 1.61E+5 5.13E+5 2.32E+5 3.43E+5 3.23E+5 Zn-69 3.43E+2 6.56E+2 4.56E+1 4.26E+2 9.85E+1 Br-82 4.07E+O 4.67E+O
Br-83 7.25E-2 1.04E-1 Br-84 9.39E-2 7.37E-7 Br-85 3.86E-3 Rb-86 6.24E+2 2.91E+2 1.23E+2 Rb-88 1.79E+O 9.49E-1 2.47E-11 Rb-89 1.19E+O 8.34E-1 6.89E-14 Sr-89 4.99E+3 1.43E+2 8.00E+2 Sr-90 1.23E+5 3.01E+4 3.55E+3 Sr-91 9.18E+1 3.71E+O 4.37E+2 Sr-92 3.48E+1 1.51E+O 6.90E+2 .Y-90 6.06E+O 1.63E-1 6.42E+4 Y-91m 5.73E-2 2.22E-3 1.68E-1 Y-91 8.88E+1 2.37E+O 4.89E+4 Y-92 5.32E-1 1.56E-2 9.32E+3 Y-93 1.69E+O 4.66E-2 5.35E+4 Zr-95 1.59E+1 5.11E+O 3.46E+O 8.02E+O 1.62E+4 Zr-97 8.81E-1 1.78E-1 8 .13E-2 2.68E-1 5.51E+4 Nb-95 4.47E+2 2.49E+2 1.34E+2 2.46E+2 1.51E+6 Nb-97 3.75E+O 9.49E-1 3.46E-1 1.11E+O 3.50E+3 Ho-99 1.28E+2 2.43E+1 2.89E+2 2.96E+2 Tc-99m 1.30E-2 3.66E-2 4.66E-1 5.56E-1 1.79E-2 2.17E+1 Tc-101 l.33E-2 1.92E-2
1.88E-1 3.46E-1 9.81E-3 5.77E-14 39 Sale* OOCH Rev. 3 07/30/87 Table 1-3 <cont'd)
Site Related Ingestion Dose Commitment Factors. Aio Cmrem/hr per uCi/ml) Nuclide Bone Liver T.Body Thyroid Kidney Lung Gl-LLI --------------------------------------------------------Ru-103 1.07E+2 4.60E+1 4.07E+2 1.25E+4 Ru-105 8.89E+O 3.51E+O 1.15E+2 5.44E+3 Ru-106 1.59E+3 2.01E+2 3.06E+3 1.03E+5 Rh-103m Rh-106 Ag-110m 1.56E+3 1.45E+3 8.60E+2 2.85E+3 5.91E+5 Sb-124 2.77E+2 5.23E+O 1.10E+2 6.71E-1 2.15E+2 7.86E+3 Sb-125 1.77E+2 1.98E+O 4.21E+1 1.80E-1 1.36E+2 1.95E+3 Te-125m 2.17E+2 7.86E+1 2.91E+1 6.52E+1 8.82E+2 8.66E+2 Te-127m 5.48E+2 1.96E+2 6.68E+1 1.40E+2 2.23E+3 1.84E+3 Te-127 8.90E+O 3.20E+O 1.93E+O 6.60E+O 3.63E+1 7.03E+2 Te-129m 9.31E+2 3.47E+2 1.47E+2 3.20E+2 3.89E+3 4.69E+3 Te-129 2.54E+O 9.55E-1 6.19E-1 1.95E+O 1.07E+1 1.92E+O Te-131m 1.40E+2 6.85E+1 5.71E+1 1.08E+2 6.94E+2 6.80E+3 Te-131 1.59E+O 6.66E-1 5.03E-1 1.31E+O 6.99E+O 2.26E-1 Te-132 2.04E+2 1.32E+2 1.24E+2 1.46E+2 1.27E+3 6 .-24E+3 1-130 3.96E+1 1.17E+2 4.61E+1 9.91E+3 1.82E+2 1.01E+2 1-131 2.18E+2 3.12E+2 1. 79E+2 1.02E+5 5.35E+2 8.23E+1
1-132 1.06E+1 2.85E+1 9.96E+O 9.96E+2 4.54E+1 5.35E+O 1-133 7.45E+1 1.30E+2 3.95E+1 1.90E+4 2.26E+2 1.16E+2 1-134 5.56E+O 1.51E+1 5.40E+O 2.62E+2 2.40E+1 1.32E-2 1-135 2.32E+1 6.08E+1 2.24E+1 4.01E+3 9.75E+1 6.87E+1 Cs-134 6.84E+3 1.63E+4 1.33E+4 5.27E+3 1.75E+3 2.85E+2 Cs-136 7.16E+2 2.83E+3 2.04E+3
-1.57E+3 2.16E+2 3.21E+2 Cs-137 8.77E+3 1.20E+4 7.8SE+3 4.07E+3 1.35E+3 2.32E+2 Cs-138 6.07E+O 1.20E+1 5.94E+O 8.81E+O 8.70E-1 5.12E-5 Ba-139 7.85E+O 5.59E-3 2.30E-1 5.23E-3 3.17E-3 1.39E+1 Ba-140 1.64E+3 2.06E+O 1.08E+2 7.02E-1 1.18E+O 3.38E+3 Ba-141 3.81E+O 2.88E-3 1. 29E-1 2.68E-3 1.63E-3 1.80E-9 Ba-142 1. 72E+O 1.77E-3 1.-0SE-1 1.50E-3 1.00E-3 2.43E-18 La-140 1.57E+O 7.94E-1 2.10E-1 5.83E+4 La-142 8.06E-2 3.67E-2 . 9.13E-3 2.68E+2 Ce-141 3.43E+O 2.32E+O 2.63E-1 1.08E+O 8.86E+3 Ce-143 6.04E-1 4.46E+2 4.94E-2 1.97E-1 1.67E+4 Ce-144 1. 79E+2 7.47E+1 9.59E+O 4.43E+1 6.04E+4 Pr-143 5.79E+O 2.32E+O 2.87E-1 1.34E+O 2.54E+4 Pr-144 1.90E-2 7.87E-3 9.64E-4 2.73E-9 Nd-147 3.96E+O 4.58E+O 2.74E-1 2.68E+O 2.20E+4 W-187 9.16E+O 7.66E+O *2.68E+O 2.51E+3 Np-239 3.53E-2 3.47E-3 1.91E-3 1.08E-2 7.11E+2
40
*
Salem ODCH Rev. 3 07/30/87 Table 1-4 Bioaccumulation Factors <BFi > (pCi/kg per pCi/liter>*
E.iab H . 9.0E-01 9.3E-01 c 1.8E+03 1.4E+03 Na 6.7E-02
1.9E-01 p 3.0E+03 3.0E+04 Cr 4.0E+02 2.0E+03 Hn. 5.5E+02 4.0E+02 Fe 3.0E+03 2.0E+04 Co 1.0E+02 1.0E+03 Ni 1.0E+02 2.5E+02 Cu 6.7E+02 1. 7E+03 Zn 2.0E+03 5.0E+04 Br 1.5E-02 3.1E+OO Rb -8.3E+OO 1.7E+01 Sr 2.0E+OO 2.0E+01 y 2.5E+01 1.0E+03 Zr 2.0E+02 8.0E+01 N_b 3.0E+04 1.0E+02 Ho 1.0E+01 1.0E+01 Tc 1.0E+01 5.0E+01 Ru 3.0E+OO 1.0E+03 Rh 1.0E+01 2.0E+03 Ag 3.3E+03 3.3E+03 Sb 4.0E+01 5.4E+OO Te 1.0E+01 1.0E+02 I 1.0E+01 5.0E+01 Cs 4.0E+01 2.5E+01 Ba 1.0E+01 1.0E+01 La 2.5E+01 1.0E+03 Ce 1.0E+01 6.0E+02 Pr 2.5E+01 1.0E+03 Nd 2.5E+01 1.0E+03 w 3.0E+01 3.0E+01 Np 1.0E+01 1.0E+01
Values in this table are taken from Regulatory Guide 1.109 except for phosphurus (fish> which is adapted from NUREG/CR-1336 and silver and antimony which are taken from UCRL 50564, Rev. 1, October 1972
41 u
11 .. COAE SlAL TABLE CXlNWHMENt
CONJ-NT ..........

* PU11iE 5'f51EM -r ----:..--=-=-.J t FROM AUX BLDG VENl SUPR.Y FANS QWtNEL PSE&GCO au:ulfR ----1.IL-...JIWll...IQ..

1.IL-lhA RA 4121 AOOOIA ** RA l.14J AIJOOM U RA 4322 R0002A Id AA 4323 RnJlA M RA4l24 RD004A M RA 432S fDD5A MA RA 432' A0001SA .. Y24!.10 R1 RA 4l2P ROD01A M RA 4329 ADOOeA -RA 4111 RD001A 111UA RA <1312 fDQIA .... RAa:N$

IUIA AA 430.1 flDmM. MIA RA 43JO fl'.l012A lllH RA4lll AD038A ... RA433> ....... ... ........ ...... ... .... , -........ -.... RA4l14 -.... RA&JIS . ..... ... ....... .... """"° -... ..., ...... ..... flA 101$1 ... AAIOIM *-___ _J ...........
-.. IMIC .._ --
8'i AREA RADIATION MONITOR -+-GASEOUS EfFWEN1' ----ELECTRICAl.
---PNEUMATIC

<X>>Clf:NSER AIR REl<<ML SYSIDI I -----------'
.. -----i WG41 I
FOR INFORMATION ONLY FIGURE 2-1 RADIATION MONITORING GASEOUS 1UNIT
..QWllL..
... -1111 = .. * ..... .... lll1A .... ... ... .. 11111 ---.... ..... MIC rwu '* I SV ---____J L°":. r-I I
S"f&Jfll
' llA<>>I ..QWllL ... IOCO ---*-I _J ::: = .. --------:::= :fl ::: L...-="""" ...... , ...... .......
AA.Ull ..... "41114 ::::: =--""-:::= =--... -* ... _ .. .... ....... ....... lllAIOIU MIOIM ==-..
JJ ... ..CXMr4( .. te......_GAS
............ , Ž PROCESS RAD!.( iii MONITOR iTICIN AREA RADIATION MONITOR -+-GASEOUS

ELEC1RtCAL EFfUENT --PNElAIATIC

' FOR INFORMATION ONLY FIGURE 2-2 RADIATION 2UNIT Sale1 ODO! Rev. 3 07/30/87 , Table 2-1 Dose Factors for Noble Gases
Total Body 6a11a Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor Radionuclide Ki Li "i Ni (1re1/yr per uCi/13) !1re1/yr per uCi/;3) (1rad/yr per uCi/13) (1rad/yr per uCi/13l ------------

Kr-8311 7.56E-02 1. 93E+01 2.88E+02 Kr-851 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1. 72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 L41E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1. 73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-1311 9 .15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-1331 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-13511 3.12E+03 7 .11E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1. 92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03 * *
*
Sale1 ODC" Rev. 3 07/30/87 Table 2-2 Para1eters for Gaseous Alar1 Setpoint Deter1inations Unit-1 P-aratehr Actual Default Units Co11ents Value Value ----------------------------------

3 X/Q calculated 2.2E-06 USNRC Sale1 Safety Evaluation, Sup. 3 3 VF as 1easured 1.25E+05 ft /min Plant Vent -nor1al operation
!Plant or Ventl fan curves J VF as aeasured J;SE+04 ft /1in Contain1ent purge (Cont. or Purge} fan curves AF coordinated 0.25 unitless Ad1inistrative allocation factor to with HC6S ensure co1bined releases do not exceed release rate li1it for site. J Ci 1easured NIA uCi/c11 J Ki nuclide specific N/A 1re1/yr per uCi/1 Values fro1 Tab I e 2-1 J Li nuclide specific N/A 1re1/yr per uCi/1 Values fro1 Table 2-1 3 ni nuclide specific N/A 1rad/yr per uCi/1 Values fro1 Table 2-1 J SEN 1-R41CI as 1.6E+07 cp1 per uCi/c1 Plant Vent deter1ined 1-R16 3.6E+07 Plant Vent (redundant}
1-R12A 2.1E+06 Contain1ent SP 1-R41C calculated 3.3E+04(+bkgl CPI Default alar1 setpoints; 1ore conservative values 1ay be used as dee1ed appropriate and 1-R16 calculated 7.4E+04(+bkg}
desirable for ensuring regulatory co1pliance and for maintaining releases ALARA. 1-R12A11 calculated 1.SE+04(+bkg}
1 Based on 1ean for calibration with 1ixture of radionuclides 11 Applicable during "ODES 1 through 5. During "ODE 6 (refueling), 1onitor setpoint shall be reduced to 2X background in accordance*vith Tech Spec Table 3.3-6.
*
Sale1 ODC" Rev. J 07/J0/87 Tab 11! 2-J Para1eters for Gaseous Alar1 Setpoint Deter1inations Unit-2 Parameter Actual Default Units Co11ents Value Value ----------------------------------

J I/& calculated 2.2E-06 sec/1 licensing technical specification value J VF as 1easured 1.25E+05 ft /1in Plant Vent -normal operation (Plant or Vent) fan curves J VF as 1easured J.5E+04 ft /1in Contain1ent purge (Cont. or Purge! fan curves AF coordinated 0.25 unitless Ad1inistrative allocation factor to with HCGS ensure co1bined releases do not exceed release rate li1it for site. J Ci 1easured N/A uCi/c1 J Ki nuclide specific N/A 1re1/yr per uCi/1 Values fro1 Table 2-1 J Li nuclide specific N/A 1re1/yr per uCi/1 Values fro1 Table 2-1 J "i nuclide specific N/A 1rad/yr per uCi/1 Values fro1 Table 2-1 J SEN 2-R41C1 as 1.6E+07 cp1 per uCi/c1 Plant Vent deter1ined 2-R16 .J.5E+07 Plant Vent (redundant!
2-R12A 3.3E+07 Containaent SP 2-R41C calculated J.3E+04(+bkg)
CPI Default alar1 setpoints; 1ore conservative values 1ay be used as dee1ed appropriate and 2-R16 calculated 7.2E+04(+bkg) desirable for ensuring regulatory co1pliance 2-R12A11 calculated 2.4E+05(+bkg) and for 1aintaining releases ALARA. ------------------------
1 Based on 1ean for calibration with 1ixture of radionuclides 11 Applicable during "ODES 1 through 5. During "ODE 6 (refueling), 1onitor setpoints shall be reduced to 21 background in accordance with Tech Spec Table 3.J-6. 46
*
Technical Specification
3.11.2.1a 3.11.2.1b 3.11.2.2 3.11.2.3 6.9.1.10 Tab le 2-4 Controlling Locations, Pathways and AtlDsplieric Dispersion for Dose Calcuiations 1 Location Path11ay(sJ Control ling Age Group -------------

site boundary noble gases N/A !0.83 tile, NJ direct exposure site boundary inhalation child (0.83 1ile, NJ site boundary ga11a-air N/A !0.83 1ile, NJ beta-air residence/dairy 1ilk and infant (4.8 1iles, NNEJ ground plane Second sun direct exposure N/A !0.21 1i le/SEJ and inhalation

Sale1 ODC" Rev. 3 07/30/87 1t1ospheric Dispersion

X/Q D/Q (sec/13J (1/12) --------. 2.2E-06 N/A 2.2E-06 NIA 2.2E-06 N/A 5.4E-08 2.1E-10 8.22E-06 N/A I The identified controlling locations, pathways and at1Dspheric dispersion are fro1 the Safety Evaluation Report, Supple1ent No. 3 for the Sale* Nuclear Generating Station, Unit 2 !NUREG-0517, Dece1ber 19781
47 Salem ODCH Rev. 3 07/30/87 Table 2-5 Pathway Dose Factors -Atmospheric Releases
RCio), Inhalation Pathway Dose Factors *-ADULT Cmrem/yr per uCi/m3) Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body --------------------------------------------------------H-3 1.26E+3 1. 26E+3 1.26E+3 1.26E+3 1.26E+3 1.26E+3 C-14 1.82E+4 3.41E+3 3.41E+3 3.41E+3 3.41E+3 3.41E+3 3.41E+3 P-32 1.32E+6 7.71E+4 8.64E+4 5.01E+4 Cr-51 5.95E+1 2.28E+1 1.44E+4 3.32E+3 1.00E+2 Hn-54 3.96E+4 9.84E+3 1.40E+6 7.74E+4 6.30E+3 Fe-55 2.46E+4 1. 70E+4 7.21E+4 6.03E+3 3.94E+3 Fe-59 1.18E+4 2.78E+4 1.02E+6 1.88E+5 1.06E+4 Co-57 6.92E+2 3.70E+5 3.14E+4 6.71E+2 Co-58 1.58E+3 9.28E+5 1.06E+5 2.07E+3 Co-60 1.15E+4 5.97E+6 2.85E+5 1.48E+4 Ni-63 4.32E+5 3.14E+4 1. 78E+5 1.34E+4 1.45E+4 Zn-65 3.24E+4 1.03E+5 6.90E+4 8.64E+5 5.34E+4 4.66E+4 Rb-86 1.35E+5 1.66E+4 5.90E+4 Sr-89 3.04E+5 1.40E+6 3.50E+5 8.72E+3 Sr-90 9.92E+7 9.60E+6 7.22E+5 6.10E+6 Y-91 4.62E+5 1. 70E+6 3.85E+5 1.24E+4
Zr-95 1.07E+5 3.44E+4 S.42E+4 1. 77E+6 1.50E+5 2.33E+4 Nb-95 1.41E+4 7.82E+3 7.74E+3 5.05E+5 1.04E+5 4.21E+3 Ru-103 1.53E+3 5.83E+3 5.05E+5 1.10E+5 6.58E+2 Ru-106 6.91E+4 1.34E+5 9.36E+6 9 .12E+5 8.72E+3 Ag-110m 1.08E+4 1.00E+4 1.97E+4 4.63E+6 3.02E+5 5.94E+3 Sb-124 3.12E+4 5.89E+2 7.55E+1 2.48E+6 4.06E+5 1.24E+4 Sb-125 5.34E+4 5.95E+2 5.40E+1 1. 74E+6 1.01E+5 1.26E+4 Te-125m 3.42E+3 1.58E+3 1.05E+3 1.24E+4 3.14E+5 7.06E+4 4.67E+2 Te-127m 1.26E+4 5.77E+3 3.29E+3 4.58E+4 9.60E+5 1.50E+5 1.57E+3 Te-129m 9.76E+3 4.67E+3 3.44E+3 3.66E+4 1.1oE+6 3.83E+5 1.58E+3 1-131 2.52E+4 3.58E+4 1.19E+7 6.13E+4 6.28E+3 2.05E+4 Cs-134 3.73E+5 8.48E+5 2.87E+5 9.76E+4 1.04E+4 7.28E+5 Cs-136 3.90E+4 1.46E+5 8.56E+4 1.20E+4 1.17E+4 1.10E+5 Cs-137 4.78E+5 6.21E+S 2.22E+5 7.52E+4 8.40E+3 4.28E+5 Ba-140 3.90E+4 4.90E+1 1.67E+1 1.27E+6 2.18E+5 2.57E+3 Ce-141 1.99E+4 1.35E+4 6.26E+3 3.62E+5 1.20E+5 1.53E+3 Ce-144 3.43E+6 1.43E+6 8.48E+5 7.78E+6 8.16E+5 1.84E+5 Pr-143 9.36E+3 3.75E+3 2.16E+3 2.81E+5 2.00E+5 4.64E+2 Nd-147 5.27E+3 6.10E+3 3.56E+3 2.21E+5 1.73E+5 3.65E+2 .pa
48 Salem ODCH Rev. 3 07/30/87
Table 2-5 (cont'd) R<io), Inhalation Pathway Dose Factors -TEENAGER (mrem/yr per uCi/m3) Nuclide Bone liver Thyroid Kidney lung GI-llI T.Body --------------------------------------------------------H-3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 1.27E+3 C-14 2.60E+4 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 P-32 1.89E+6 1.10E+5 9.28E+4 7.16E+4 Cr-51 7.50E+1 3.07E+1 2.10E+4 3.00E+3 1.35E+2 Hn-54 5 .11E+4 1.27E+4 1.98E+6 6.68E+4 8.40E+3 Fe-55 3.34E+4 2.38E+4 1.24E+5 6.39E+3 5.54E+3 Fe-59 1.59E+4 3.70E+4 1.53E+6 1.78E+5 1.43E+4* Co-57 6.92E+2 5.86E+5 3.14E+4 9.20E+2 Co-58. 2.07E+3 1.34E+6 9.52E+4 2.78E+3 Co-60 1.51E+4 8.72E+6 2.59E+5 t.98E+4 Ni-63 5.80E+5 4.34E+4 3.07E+5 1.42E+4 1.98E+4 Zn-65 3.86E+4 1.34E+5 8.64E+4 1.24E+6 4.66E+4 6.24E+4 Rb-86 1.90E+5 1. 77E+4 8.40E+4 Sr-89 4.34E+5 2.42E+6 3.71E+5 1.25E+4 Sr-90 1.08E+8 1.65E+7 6.68E+6 Y-91 6.61E+5 2.94E+6 4.09E+5 1. 77E+4
Zr-95 1.46E+5 4.58E+4 6.74E+4 2.69E+6 1.49E+5 3.15E+4 Nb-95 1.86E+4 1.03E+4 1.00E+4 7.51E+5 9. 68E+4. 5.66E+3 Ru-103 2.10E+3 7.43E+3 7.83E+5 1.09E+5 8.96E+2 Ru-106 9.84E+4 1.90E+5 1.61E+7 9.60E+5 1.24E+4 Ag-110m 1.38E+4 1.31E+4 2.50E+4 6.75E+6 2.73E+5 7.99E+3 Sb-124 4.30E+4 7.94E+2 9.76E+1 3.85E+6 3.98E+5 1.68E+4 Sb-125 7.38E+4 8.08E+2 7.04E+1 2.74E+6 9.92E+4 1. 72E+4 Te-125m 4.88E+3 2.24E+3 1.40E+3 5.36E+5 7.50E+4 6.67E+2 Te-127m 1.80E+4 8.16E+3 4.38E+3 6.54E+4 1.66E+6 1.59E+5 2.18E+3 Te-129m 1.39E+4 6.58E+3 4.58E+3 5.19E+4 1. 98E+6 4.05E+5 2.25E+3 I-131 3.54E+4 4.91E+4 1.46E+7 8.40E+4 6.49E+3 2.64E+4 Cs-134 5.02E+S 1.13E+6 3.75E+S 1.46E+5 9.76E+3 5.49E+S Cs-136 S.1SE+4 1.94E+S 1.10E+S 1. 78E+4 1.09E+4 1.37E+S Cs-137 6.70E+S 8.48E+5 3.04E+S 1.21E+5 8.48E+3
3.11E+S Ba-140 5.47E+4 6.70E+1 2.28E+1 2.03E+6 2.29E+S 3.52E+3 Ce-141 2.84E+4 1.90E+4 8.88E+3 6.14E+5 1.26E+5 2.17E+3 Ce-144 4.89E+6 2.02E+6 1.21E+6 1.34E+7 8.64E+5 2.62E+5 Pr-143 1.34E+4 5.31E+3 3.09E+3 4.83E+5 2.14E+S 6.62E+2 Nd-147 7.86E+3 8.56E+3 5.02E+3 3.72E+5 1.82E+5 5.13E+2
49 L -
Sale* ODCH Rev. 3 07/30/87 Table 2-5 (cont'd) RC io), Inhalation Pathway Dose Factors -CHILD * (mrem/yr per uC i/m3) Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body --------------------------------------------------------H-3 1.12E+3 1.12E+3 1.12E+3 1.12E+3 1.12E+3 1.12E+3 C-14 3.59E+4 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 P-32 2.60E+6 1.14E+5 4.22E+4 9.88E+4 Cr-51 8.55E+1 2.43E+1 1. 70E+4 1.08E+3 1.54E+2 Hn-54 4.29E+4 1.00E+4 1.58E+6 2.29E+4 9.51E+3 Fe-55 4.74E+4 2.52E+4 1.11E+5 2.87E+3 7.77E+3 Fe-59 2.07E+4 3.34E+4 1.27E+6. 7.07E+4 1.67E+4 Co-57 9.03E+2 5.07E+5 1.32E+4 1.07E+3 Co-58 1. 77E+3 1.11E+6 3.44E+4 3.16E+3 Co-60 1.31E+4 7.07E+6 9.62E+4 2.26E+4 Ni-63 8.21E+5 4.63E+4 2.75E+5 6.33E+3 2.80E+4 Zn-65 4.26E+4 1.13E+5 7.14E+4 9.95E+5. 1.63E+4 7.03E+4 Rb-86 1.98E+5 7.99E+3 1.14E+5 Sr-89 5.99E+5 2.16E+6 1.67E+5 1. 72E+4 Sr-90 1.01E+8 1.48E+7 3.43E+5 6.44E+6 Y-91 9.14E+5 2.63E+6 .1.84E+5 2.44E+4 Zr-95 1.90E+5 4 .18E+4 5.96E+4 2.23E+6 6.11E+4 3.70E+4 Nb-95 2.35E+4 9.18E+3 8.62E+3 6.14E+5 3.70E+4 6.55E+3
Ru-103 2.79E+3 7.03E+3 6.62E+5 4.48E+4 1.07E+3 Ru-106 1.36E+5 1.84E+5 1.43E+7 4.29E+5 1.69E+4 Ag-110m 1.69E+4 1.14E+4 2 .12E+4 5.48E+6 1.00E+5 9.14E+3 Sb-124 5.74E+4 7.40E+2 1.26E+2 3.24E+6 1.64E+5 2.00E+4 Sb-125 9.84E+4 7.59E+2 9.10E+1 2.32E+6 4.03E+4 2.07E+4 Te-125m 6.73E+3 2.33E+3 1.92E+3 4.77E+5 3.38E+4 9.14E+2 Te-127m 2.49E+4 8.55E+3 6.07E+3 6.36E+4 1.48E+6 7.14E+4 3.02E+3 Te-129m 1.92E+4 6.85E+3 6.33E+3 5.03E+4 1.76E+6 1.82E+5 3.04E+3 I-131 4.81E+4 4.81E+4 1.62E+7 7.88E+4 2.84E+3 2.73E+4 Cs-134 6 *. 51E+5 1.01E+6 3.30E+5 1.21E+5 3.85E+3 2.25E,+5 Cs-136 6.51E+4 1. 71E+S 9.SSE+4 1.45E+4 4.18E+3 1.16E+S Cs-137 9.07E+S 8.25E+S 2.82E+5 1.04E+S 3.62E+3 1.28E+5 Ba-140 7.40E+4 6.48E+1 2.11E+1 1.74E+6 1.02E+S 4.33E+3 Ce-141 3.92E+4 1.95E+4 8.55E+3 5.44E+5 5.66E+4 2.90E+3 Ce-144 6.77E+6 2.12E+6 1.17E+6 1.20E+7 3.89E+5 3.61E+5 Pr-143 1.85E+4 5.55E+3 3.00E+3 4.33E+5 9.73E+4 9.14E+2 Nd-147 1.08E+4 8.73E+3 4.81E+3 3.28E+5 8.21E+4 6.81E+2
50 Salem ODCH Rev. 3 07/30/87 Table 2-5 <cont'd> ** R< io), Inhalation Pathway Dose Factors -INFANT (mrem/yr per uCi/m3) Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body ---------------------

H-3 6.,47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 C-14 2.65E+4 5.31E+3 5.31E+3 5.31E+3 5.31E+3 5.31E+3 5.31E+3 P-32 2.03E+6 1.12E+5 1.61E+4 7.74E+4 Cr-51 5.75E+1 1.32E+1 1.28E+4 3.57E+2 8.95E+1 Hn-54 2.53E+4 4.98E+3 1.00E+6 7.06E+3 4.98E+3 Fe-55 1.97E+4 1.17E+4 8.69E+4 1.09E+3 3.33E+3 -Fe-59 1.36E+4 2.35E+4* 1.02E+6 2.48E+4 9.48E+3 Co-57 6.51E+2 3.79E+5 4.86E+3 6.41E+2 Co-58 L22E+3 7.77E+5 1.11E+4 1.82E+3 Co-60 8.02E+3 4.51E+6 3.19E+4 1.18E+4 Ni-63 3.39E+5 2.04E+4 2.09E+5 2.42E+3 1.16E+4 Zn-65 1.93E+4 6.26E+4 3.25E+4 6.47E+5 5.14E+4 3 .11E+4 Rb-86 1.90E+5 3.04E+3 8.82E+4 Sr-89 3.98E+5 2.03E+6 6.40E+4 1.14E+4 Sr-90 4.09E+7 1.12E+7 1.31E+5. 2.59E+6 Y-91 5.88E+5 2.45E+6 7.03E+4 1.57E+4

Zr-95 1.15E+5 2.79E+4 3.11E+4 1.75E+6 2.17E+4 2.03E+4 Nb:..95 1.57E+4 6.43E+3 4.72E+3 4.79E+5 1.27E+4 3.78E+3 Ru-103 2.02E+3 4.24E+3 5.52E+5 1.61E+4 6.79E+2
Ru-106 8.68E+4 1.07E+5 1.16E+7 1.64E+5 1.09E+4 Ag-110m 9.98E+3 7.22E+3 1.09E+4 3.67E+6 3.30E+4 5.00E+3 Sb-124 3.79E+4
5.56E+2 1.01E+2 2.65E+6 5.91E+4 1.20E+4 Sb-125 5.17E+4 4.77E+2 6.23E+1 1.64E+6 1.47E+4 1.09E+4 Te-125m 4.76E+3 1.99E-t:3 1.62E+3 4.47E+5 1.29E+4 6.58E+2 Te-127m 1.67E+4 6.90E+3 4.87E+3 3.75E+4 1.31E+6 2.73E+4 2.07E+3 Te-129m 1.41E+4 6.09E+3 S.47E+3 3.18E+4 1.68E+6 6.90E+4 2.23E+3 I-131 3.79E+4 4.44E+4 1.48E+7 5.18E+4 ;.... 1.06E+3 1.96E+4 Cs-134 3.96E+S 7.03E+5 1.90E+S 7.97E+4 1.33E+3 7.4SE+4 Cs-136 4.83E+4 1.35E+5 S.64E+4 1.18E+4 1.43E+3 S.29E+4 Cs-137 5.49E+5 6 .12E+S 1. 72E+S 7.13E+4 1.33E+3 4.55E+4 Ba-140 5.60E+4 S.60E+1 1.34E+1 1.60E+6 3.84E+4 2.90E+3 Ce-141 2.77E+4 1.67E+4 S.2SE+3 5-.17E+5 2.16E+4 1.99E+3 Ce-144 3.19E+6 1.21E+6 5.38E+S 9.S4E+6 1.48E+5 1.76E+5 Pr-143 1.40E+4 S.24E+3 1.97E+3 4.33E+5 3.72E+4 6.99E+2 Nd-147 7.94E+3 8.13E+3 3.15E+3 3.22E+5 3.12E+4 5.00E+2
51 Salem OOCH Rev. 3 07/30/87 Table 2-5 <cont'd)
RC io), Grass-Cow-Hilk Pathway Dose Factors -ADULT Cmrem/yr per uCi/m3) for H-3 and C-14 ( m2
mrem/yr per uCi/sec> for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body ------------------------------------------. --------------H-3 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 C-14 3.63E+5 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 P-32 1. 71E+10 1.06E+9 1.92E+9 6.60E+8 Cr-51 1. 71E+4 6.30E+3 3.80E+4 7.20E+6 2.86E+4 Hn-54 8.40E+6 2.50E+6 2.57E+7 1.60E+6 Fe-SS 2.S1E+7 1. 73E+7 9.67E+6 9.9SE+6 4.04E+6 Fe-59 2.98E+7 7.00E+7 1.9SE+7 2.33E+8 2.68E+7 Co-S7 1.28E+6 3.2SE+7 2.13E+6 Co-S8 4.72E+6 9.S7E+7 1.06E+7 Co-60 1.64E+7 3.08E+8 3.62E+7 Ni-63 6.73E+9 4.66E+8 9.73E+7 2.26E+8 Zn-65 1.37E+9 4.36E+9 2.92E+9 2.7SE+9 1.97E+9 Rb-86 1.35E+5 1.66E+4 S.90E+4 Sr-89 1.4SE+9 2.33E+8 4.16E+7 Sr-90 4.68E+10 1.3SE+9 1.1SE+10 Y-91 8.60E+3 4.73E+6 2.30E+2
Zr-9S 1.07E+S 3.44E+4 5.24E+4 1. 77E+6 1.SOE+S 2.33E+4 Nb-9S 1.41E+4 7.82E+3 7.74E+3 5.0SE+S 1.04E+S 4.21E+3 R-u-103 1.02E+3 3.89E+3 1.19E+S 4.39E+2 Ru-106 2.04E+4 -3.94E+4 1.32E+6 2.58E+3 Ag-110m S.83E+7 S.39E+7 1.06E+8 2.20E+10 3.20E+7 Sb-124 2.S7E+7 4.86E+S 6.24E+4 2.00E+7 7.31E+8 1.02E+7 Sb-12S 2.04E+7 2.28E+5 2.08E+4 1.58E+7 2.2SE+8 4.86E+6 Te-12S111 1.63E+7 S.90E+6 4.90E+6 6.63E+7 6.SOE+7 2.18E+6 Te-127111 4.S8E+7 1.64E+7 -1.17E+7 1.86E+8 1.54E+8 5.58E+6 Te-129m 6.04E+7 2.2SE+7 2.08E+7 2.S2E+8 3.04E+8 9.57E+6 I-131 2.96E+8 4.24E+8 1.39E+11 7.27E+8 1.12E+8 2.43E+8 Cs-134 _S.6SE+9 1.34E+10 4.3SE+9 1.44E+9 2.3SE+8 1.10E+10 Cs-136 2.61E+8 1.03E+9 5.74E+8 7.87E+7 1.17E+8 7.42E+8 Cs-137 7.38E+9 1.01E+10 3.43E+9 1.14E+9 1.9SE+8 6.61E+9 Ba-140 2.69E+7 3.38E+4 1.1SE+4 1.93E+4 S.54E+7 1. 76E+6 Ce-141 4.84_E+3 3.27E+3 1.S2E+3 1.2SE+7 3. 71E+2 Ce-144 3.S8E+5 1.50E+S 8.87E+4 1.21E+8 1.92E+4 Pr-143 1.S9E+2 6.37E+1 3.68E+1 6.96E+5 7.88E+O Nd-147 9.42E+1 1.09E+2 6.37E+1 5.23E+5 6.52E+O
52

Sale11 ODCH Rev. 3 07/30/87 Table 2-5 <cont'd> R< io), Grass-Cow-Hilk Pathway Dose Factors -TEENAGER * (mrem/yr per uCi/m3) for H-3 and C-14 (m2

mrem/yr per uCi/sec> for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body --------------------------------------------------------H-3 9.94E+2 9.94E+2 9_.94E+2 9.94E+2 9.94E+2 9.94E+2 C-14 6.70E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 P-32 3.15E+10 1.95E+9 2.65E+9 1.22E+9 Cr-51 2.78E+4 1.10E+4 7.13E+4 8.40E+6 5.00E+4 Hn-54 1.40E+7 4.17E+6 2.87E+7 2.78E+6 Fe-55 4.45E+7 3.16E+7 2.00E+7 1.37E+7 7.36E+6 Fe-59 5.20E+7 1.21E+8 3.82E+7 2.87E+8 4.68E+7 Co-57 2.25E+6 4.19E+7 3.76E+6 Co-58 7.95E+6 1.10E+8 1.83E+7 Co-60 2.78E+7 3.62E+8 6.26E+7 Ni-63 1.18E+10 8.35E+8 1.33E+8 '4.01E+8 Zn-65 2 .11E+9 7.31E+9 4.68E+9 3.10E+9 3.41E+9 *Rb-86 4.73E+9 7.00E+8 2.22E+9 Sr-89 2.67E+9 3.18E+8 7.66E+7 Sr-90 9.92E+7 9.60E+6 7.22E+5 6.10E+6 Y-91 1.58E+4 6.48E+6 4.24E+2 Zr-95 1.6SE+3 5.22E+2 7.67E+2 1.20E+6 3.59E+2
Nb-95 1.41E+S 7.80E+4 7.57E+4 3.34E+8 4.30E+4 Ru-103 1.81E+3 6.40E+3 1.52E+5 7.75E+2 Ru-106 3.75E+4 7.23E+4 1.80E+6 4.73E+3 Ag-110m 9.63E+7 9.11E+7
1. 74E+8 2.56E+10 5.54E+7 Sb-124 4.59E+7 8.46E+5 1.04E+S 4.01E+7 9.25E+8
1. 79E+7 Sb-125 3.65E+7 3.99E+5 3.49E+4 3.21E+7 2.84E+8 8.54E+6 Te-125m 3.00E+7 1.08E+7 8.39E+6 8.86E+7 4.02E+6 Te-127m 8.44E+7 2.99E+7 2.01E+7 3.42E+8 2.10E+8 1.00E+7 Te-129m 1.11E+8 4.10E+7 3.57E+7 4.62E+8 4.15E+8 1.75E+7 I-131 S.38E+8 7.53E+8 2. 20E+11 1.30E+9 1.49E+8 4.04E+8 Cs-134 9.81E+9 2.31E+10 7.34E+9 2.80E+9 2.87E+8 1.07E+10 Cs-136 4.4SE+8 1. 75E+9 9.53E+8 1.SOE+8 1.41E+8 1.18E+9 Cs-137 1.34E+10 1.78E+10 6.06E+9 2.35E+9 2.53E+8 6.20E+9 Ba-140 4.8SE+7 S.9SE+4 2.02E+4 4.00E+4* 7.49E+7 3.13E+6 Ce-141 1.99E+4 1.3SE+4 6.26E+3 3.62E+S 1.ZOE+S 1.53E+3 Ce-144 6.58E+S 2.72E+S 1.63E+5 1.66E+8 3.54E+4 Pr-143 2.92E+2 1.17E+2 6.77E+1 9.61E+S 1.45E+1 Nd-147 1.81E+2 1.97E+2 1.16E+2 7 .11E+S 1.18E+1
53 Salem ODCH Rev. 3 07/30/87 Table 2-5 <cont'd> RCio), Grass-Cow-Hilk Pathway Dose Factors -CHILD
Cmrem/yr per uCi/m3) for H-3 and C-14 Cm2
mrem/yr per uCi/sec> for others Nuclide Bone Liver Thyroid Kidney Lung Gl-LLI T.Body --------------------------------------------------------H-3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 1.57E+3 C-14 1.65E+6 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+5 P-32 7.77E+10 3.64E+9 2.15E+9 3.00E+9 Cr-51 S.66E+4 1.55E+4 1.03E+5 5.41E+6 1.02E+5 Hn-54 2.09E+7 S.87E+6 1.76E+7 5.58E+6 Fe-55 1.12E+8 5.93E+7 3.35E+7 1.10E+7 1.84E+7 Fe-59 1.20E+8 1.95E+8 S.65E+7 2.03E+8 9.71E+7 Co-57 3.84E+6 3.14E+7 7.77E+6 Co-58 1.21E+7 7.08E+7 3.72E+7 Co-60 4.32E+7 2.39E+8 1.27E+8 Ni-63 2.96E+10 1.59E+9 1.07E+8 1.01E+9 Zn-65 4.13E+9 1.10E+10 6.94E+9 1.93E+9 6.85E+9 Rb-86 8.77E+9 5.64E+8 S.39E+9 Sr-89 6.62E+9 2.56E+8 1.89E+8 Sr-90 1.12E+11 1.51E+9 2.83E+10 Y-91 9.14E+5 2.63E+6 1.84E+5 2.44E+4 Zr-95 3.84E+3 8.45E+2 1.21E+3 8.81E+5 7.52E+2 ** Nb-95 3.18E+5 1.24E+5 1.16E+5 2.29E+8 8.84E+4 Ru-103 4.29E+3 1.08E+4 1.11E+5 1.65E+3 Ru-106 9.24E+4 1.25E+5 1.44E+6 1.15E+4 Ag-110m 2.09E+8 1.41E+8 2.63E+8 1.68E+10 1.13E+8 Sb-124 1.09E+8 1.41E+8 2.40E+5 6.03E+7 6.79E+8 3.81E+7 Sb-125 8.70E+7 1.41E+6 8.06E+4 4.85E+7 2.08E+8 1.82E+7 Te-125m 7.38E+7 2.00E+7 2.07E+7 7.12E+7 9.84E+6 Te-127m 2.08E+8 5.60E+7 4.97E+7 S.93E+8 1.68E+8 2.47E+7 Te-129m 2.72E+8 7.61E+7 8.78E+7 8.00E+8 3.32E+8 4.23E+7 1-131 1.30E+9 1.31E+9 4.34E+11 2.1SE+9 1.17E+8 7.46E+8 Cs-134 2.26E+10 3.71E+10 1.15E+10 4.13E+9 2.00E+8 7.83E+9 Cs-136 1.00E+9 2.76E+9 1.47E+9 2.19E+8 9.70E+7 1. 79E+9 Cs-137 3.22E+10 3.09E+10 1.01E+10 3.62E+9 1.93E+8 4.55E+9 Ba-140 1.17E+8 1.03E+5 3.34E+4 6.12E+4 5.94E+7 6.84E+6 Ce-141 2.19E+4 1.09E+4 4.78E+3 1.36E+7 1.62E+3 Ce-144 1.62E+6 5.09E+5 2.82E+5 1.33E+8 8.66E+4 Pr-143 7.23E+2 2.17E+2 1.17E+2 7.80E+5 3.59E+1 Nd-147 4.45E+2 3.60E+2 1.98E+2 5. 71E+5 2.79E+1
54

Salem ODCH Rev. 3 07/30/87 Table 2-5 (cont'd) RCio), Grass-Cow-Hilk Pathway Dose Factors -INFANT * (mrem/yr per uCi/m3> for H-3 and C-14 (m2

mrem/yr per uCi/sec> for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body --------------------------------------------------------H-3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 C-14 3.23E+6 6.89E+5 6.89E+5 6.89E+5 6.89E+5 6.89E+5 6.89E+5 P-32 1.60E+11 9.42E+9 2 .17E+9 6.21E+9 Cr-51 1.05E+5 2.30E+4 2.05E+5 4. 71E+6 1.61E+5 Hn-54 3.89E+7 8.63E+6 1.43E+7 8.83E+6 Fe-55 1.35E+8 8.72E+7 4.27E+7 1.11E+7 2.33E+7 Fe-59 2.25E+8 3.93E+8 1.16E+8 1.88E+8 1.55E+8 Co-57 8.95E+6 3.05E+7 1.46E+7 Co-58 2.43E+7 6.05E+7 6.06E+7 Co-60 8.81E+7 2 .10E+8 2.08E+8 Ni-63 3.49E+10 2.16E+9 1.07E+8 1.21E+9 Zn-65 5.55E+9 1.90E+10 9.23E+9 1.61E+10 8.78E+9 Rb-86 2.22E+10*
5.69E+8 1.10E+10 Sr-89 1.26E+10 2.59E+8 3.61E+8 Sr-90 1.22E+11 1.52E+9 3.10E+10 Y-91 7.33E+4 5.26E+6 1.95E+3
Zr-95 6.83E+3 1.66E+3 1.79E+3 8.28E+5 1.18E+3 Nb-95 5.93E+5 2.44E+5 1. 75E+5 2.06E+8 1. 41E+5 Ru-103 8.69E+3 1.81E+4 1.06E+S 2.91E+3 Ru-106 1.90E+5 2.25E+5 1.44E+6 2.38E+4 Ag-110m 3.86E+8 2.82E+8 4.03E+8 1.46E+10 1.86E+8 Sb-124 2.09E+8 3.08E+6 5.56E+5 1.31E+8 6.46E+8 6.49E+7 Sb-125 1.49E+8 1.45E+6 1.87E+5 9.38E+7 1.99E+8 3.07E+7 Te-125m 1.51E+8 5.04E+7 5.07E+7
-7 .18E+7 2.04E+7 Te-127m 4.21E+8 1.40E+8 1.22E+8 1.04E+9 1.70E+8 5.10E+7 Te-129m 5.59E+8 1.92E+8 2.15E+8 1.40E+9 3.34E+8 8.62E+7 I-131 2.72E+9 3.21E+9 1.05E+12 3.75E+9 1.15E+8 1.41E+9 Cs-134 3.65E+10 6.80E+10 1.75E+10 7 .18E+9 1.85E+8 6.87E+9 Cs-136 1.96E+9 S.77E+9 2.30E+9 4.70E+8 8.76E+7 2 .15E+9 Cs-137 S.15E+10 6.02E+10 . 1.62E+10 6.55E+9 1.88E+8 4.27E+9 Ba-140 2.41E+8 2.41E+5 5.73E+4 1.48E+5 5.92E+7 1.24E+7 Ce-141 4.33E+4 2.64E+4 8.15E+3 1.37E+7 3.11E+3 Ce-144 2.33E+6 9.52E+5 3.85E+5 1.33E+8 1.30E+5 Pr-143 1.49E+3 . 5.59E+2 2.08E+2 7.89E+S 7.41E+1 Nd-147 8.82E+2 9.06E+2 3.49E+2 5.74E+5 S.55E+1
55 Salem ODCH Rev. 3 07/30/87 Table 2-5 <cont'd) R< io>, Vegetation Pathway Dose Factors -ADULT * <mrem/yr per uCi/m3) for H-3 and C-14 <m2
mrem/yr per uCi/sec> for others Nuclide Bone Liver Thyroid l<i dney Lung GI-LLI T.Body -----------------------------------

H-3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 C-14 8.97E+5 1. 79E+5 1.79E+5 1. 79E+5 1. 79E+5 1. 79E+5 . 1.79E+5 P-32 1.40E+9 8.73E+7 1.58E+8 5.42E+7 Cr-51 2.79E+4 1.03E+4 6.19E+4 1.17E+7 4.66E+4 Hn-54 3.11E+8 9.27E+7 9.54E+8 5.94E+7 Fe-55 2.09E+8 1.45E+8 8.06E+7 8.29E+7 3.37E+7 Fe-59 1.27E+8 2.99E+8 8.35E+7 9.96E+8 1.14E+8 Co-57 1.17E+7 2.97E+8 1.95E+7 Co-58 3.09E+7 6.26E+8 6.92E+7 Co-60 i.67E+8 3.14E+9 3.69E+8 Ni-63 1.04E+10 7.21E+8 1.50E+8 3.49E+8 Zn-65 3.17E+8 1.01E+9 6.75E+8 6.36E+8 4.56E+8 Rb-86 2.19E+8 4.32E+7 1.02E+8 Sr-89 9.96E+9 1.60E+9 2.86E+8 Sr-90 6.05E+11 1.75E+10 1.48E+11 Y-91 5.13E+6 2.82E+9 1.37E+5 Zr-95 1.19E+6 3.81E+5 5.97E+5 1.21E+9 2.58E+5
Nb-95 1.42E+5 7.91E+4 7.81E+4 4.80E+8 4.25E+4 Ru-103 4.80E+6 1.83E+7 5.61E+8 2.07E+6 Ru-106 1.93E+8 3.72E+8 1.25E+10 2.44E+7 Ag-110m 1.06E+7 9.76E+6 1.92E+7 . 3.98E+9 5.80E+6 Sb-124 1.04E+8 1.96E+6 2.52E+5 8.08E+7 2.95E+9 4.11E+7 Sb-125 1.36E+8 1.52E+6 1.39E+5 1.05E+8 1.50E+9 3.25E+7 Te-125m 9.66E+7 3.50E+7 2.90E+7 3.93E+8 3.86E+8 1.29E+7 Te-127m 3.49E+8 1.25E+8 8.92E+7 1.42E+9 1.17E+9 4.26E+7 Te-129m 2.55E+8 9.50E+7 8.75E+7 1.06E+9 1.28E+9 4.03E+7 1-131 8.09E+7 1.16E+8 3.79E+10 1.98E+8 3.05E+7 6.63E+7 Cs-134 4.66E+9 1.11E+10 3.59E+9 1.19E+9 1.94E+8 9.07E+9 Cs-136 4.20E+7 1.66E+8 9.24E+7 1.27E+7 1.89E+7 1.19E+8 Cs-137 6.36E+9 8.70E+9 2.95E+9 9.81E+8 1.68E+8 5.70E+9 Ba-140 1.29E+8 1.62E+5 5.49E+4 9.25E+4 2.65E+8 8.4.3E+6 Ce-141 1.96E+5 1.33E+5 6.17E+4 5.08E+8 1.51E+4 Ce-144 3.29E+7 1.38E+7 8.16E+6 1.11E+10 1. 77E+6 Pr-143 6.34E+4 2.54E+4 1.47E+4 2.78E+8 3.14E+3 Nd-147 3.34E+4 3.86E+4 2.25E+4 1.85E+8 2.31E+3
56 Salem ODCH Rev. 3 07/30/87 Table 2-5 <cont'd>
R< io), Vegetation Pathway Dose Factors -TEENAGER (mrem/yr per uCi/m3) for H-3 and C-14 (m2
mre111/yr per uC i/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body --------------------------------------------------------H-3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 C-14 1.45E+6 2.91E+S 2.91E+S 2.91E+S 2.91E+S 2.91E+S 2.91E+S P-32 1.61E+9 9.96E+7 1.35E+8 6.23E+7 Cr-S1 3.44E+4 1.36E+4 8.8S!:'.+4 1.04E+7 6.20E+4 Hn-S4 4.S2E+8 1.3SE+8 9.27E+s* 8.97E+7 Fe-SS 3.2SE+8 2.31E+8 1.46E+8 9.98E+7 S.38E+7 Fe-S9 1.81E+8 4.22E+8 1.33E+8
9.98E+8 1.63E+8 Co-S7 1. 79E+7 3.34E+8 3.00E+7 Co-S8 4.38E+7 6.04E+8 1.01E+8 Co-60 2.49E+8 3.24E+9 S.60E+8 Ni-63 1.61E+10 1.13E+9 1.81E+8 S.4SE+8 Zn-6S 4.24E+8 1.47E+9 9.41E+8 6.23E+8 6.86E+8 Rb.,.86 . 2. 73E+8 4.0SE+7 1.28E+8 Sr-89 1.S1E+10 1.80E+9 4.33E+8 Sr-90 7 .S1E+11 2.11E+10 1.8SE+11 Y-91 7.87E+6 3.23E+9 2.11E+S
Zr-9S 1.74E+6 S.49E+S 8.07E+S 1.27E+9 3.78E+S Nb-9S 1.92E+S 1.06E+S 1.03E+S 4.SSE+8 S.86E+4 Ru-i.03 6.87E+6 2.42E+7 S.74E+8 2.94E+6 Ru-106 3.09E+8 S.97E+8 1.48E+10 3.90E+7 Ag-110m 1.S2E+7 1.44E+7 2.74E+7 4.04E+9 8.74E+6 Sb-124 1.SSE+8 2.8SE+6 3.SlE+S 3 .11E+9 6.03E+7 Sb-12S 2.34E+6 2.04E+S 1.88E+8 1.66E+9 S.OOE+7 Te-12Sm 1.48E+8 S.34E+7 4.14E+7 4.37E+8 1. 98E+7 Te-127m S.S1E+8 1.96E+8 1.31E+8 2.24E+9 1.37E+9 6.S6E+7 Te-129m 3.67E+8 1.36E+8 1.18E+8_ 1.54E+9 1.38E+9 S.81E+7 1-131 7.70E+7 1.08E+8 3.14E+10 1.8SE+8 2.13E+7 S.79E+7 Cs-134 7.09E+9 1.67E+10 5.30E+9 2.02E+9 2.08E+8 7.74E+9 Cs-136 4.29E+7 1.69E+8 9.19E+7 1.4SE+7 1.36E+7 1.13E+8 Cs-137 1.01E+10 1.3SE+10 4.S9E+9 1. 78E+9 1.92E+8 4.69E+9 Ba-140 1.38E+8 1.69E+S 5.7SE+4 1.14E+S 2 .13E+8 8.91E+6 Ce-141 2.82E+S 1.88E+S 8.86E+4 S.38E+8 2.16E+4 Ce-144 S.27E+7 2.18E+7 1.30E+7 1.33E+10 2.83E+6 Pr-143 7.12E+4 2.84E+4 1.6SE+4 -2.34E+8 3.SSE+3 Nd-147 3.63E+4 3.94E+4 2.32E+4 1.42E+8 2.36E+3
57 Salem ODCH Rev. 3 07130187 Table 2-5 <cont'd)
RC io), Vegetation Pathway Dose Factors -CHILD <mrem/yr per uCi/m3) for H-3 and C-14 <m2
mrem/yr per uCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung Gl-LLI T.Body -------------------------------------------------

H-3 4.01E+3 4.01E+3 4.01E+3 4.01E+3 4.01E+3 4.01E+3 C-14 3.56E+6 7.01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 P-32 3.37E+9 1.58E+8 9.30E+7 1.30E+8 Cr-51 6.54E+4 1.79E+4 1.19E+5 6.25E+6 1.18E+5 Mn-54 6.61E+8 1.85E+8 5.55E+8 1. 76E+8 Fe-55 8.00E+8 4.24E+8 2.40E+8 7.86E+7 1.31E+8 Fe-59 4.01E+8 6.49E+8 1.88E+8 6.76E+8 3.23E+8 Co-57 2.99E+7 2.45E+8 6.04E+7 Co-58 6.47E+7 3.77E+8 1.98E+8 Co-60 3.78E+8 2.10E+9 1.12E+9 Ni-63 3.95E+10 2.11E+9 1.42E+8 1.34E+9 Zn-65 8.12E+8 2.16E+9 1.36E+9 3.80E+8 1.35E+9 Rb-86 4.52E+8 2.91E+7 2.78E+8 Sr-89 3.59E+10 1.39E+9 1.03E+9 Sr-90 1.24E+12 1.67E+10 3.15E+11 Y-91 1.87E+7 2.49E+9 5.01E+5

Zr-95 3.90E+6 8.58E+5 1. 23E+6 8.95E+8 7.64E+5 Nb-95 4.10E+5 1.59E+5 1.50E+5 2.95E+8 1.14E+5 Ru-103 1.55E+7 3.89E+7 3.99E+8 5.94E+6 Ru-106 7.45E+8 1.01E+9 1.16E+10 9.30E+7 Ag-110m 3.22E+7 2.17E+7 4.05E+7 2.58E+9 1.74E+7 Sb-124 3.52E+8 4.57E+6 7.78E+5 1.96E+8 2.20E+9 1.23E+8 Sb-125 4.99E+8 3.85E+6 4.62E+5 2.78E+8 1.19E+9 1.05E+8 Te-125m 3.51E+8 9.50E+7 9.84E+7 3.38E+8 4.67E+7 Te-127m 1.32E+9 3.56E+8 3.16E+8 3.77E+9 1.07E+9 1.57E+8 Te-129m 8.54E+8 2.39E+8 2.75E+8 2.51E+9 1.04E+9 1.33E+8 1-131 1.43E+8 1.44E+8 4.76E+10 2.36E+8 1.28E+7 8.18E+7 Cs-134 1.60E+10 2.63E+10 8.14E+9 2.92E+9 1.42E+8 5.54E+9 Cs-136 8.06E+7 2.22E+8 1.18E+8 1.76E+7 7.79E+6 1.43E+8 Cs-137 2.39E+10 2.29E+10 7.46E+9 2.68E+9 1.43E+8 3.38E+9 Ba-140 2.77E+8 2 .* 43E+5 7.90E+4 1.45E+5 1.40E+8 1.62E+7 Ce-141 1.23E+5 6.14E+4 2.69E+4 7.66E+7 9.12E+3 Ce-144 1.27E+8 3.98E+7 2.21E+7 1.04E+10 6.78E+6 Pr-143 1.48E+5 4.46E+4 2.41E+4 1.60E+8 7.37E+3 Nd-147 7.16E+4 5.80E+4 3.18E+4 9.18E+7 4.49E+3
58
*
Salem OOCH Rev. 3 07/30/87 Table 2-5 (cont'd) RCio), Ground Plane Pathway Dose Factors Cm2
mrem/yr per uCi/sec) Nuclide H-3 C-14 P-32 Cr-51 Hn-54 Fe-55 Fe-59 Co-58 Co-60 Ni-63 Zn-65 Rb-86 Sr-89 Sr-90 Y-91 Zr-95 Nb-95 Ru-103 Ru-106 Ag-110m Te-125* Te-127111 Te-129m I-131 Cs-134 Cs-136 Cs-137 Ba-140 Ce-141 Ce-144 Pr-143 Nd-147 59 Any Organ 4.68E+6 1.34E+9 2.75E+8 3.82E+8 2.16E+10 7.45E+8 8.98E+6 2.16E+4 1.08E+6 2.48E+S 1.36E+8 1.09E+8 4.21E+8 3.47E+9 1.55E+6 9.17E+4 2.00E+7 1. 72E+7 6.75E+9 1.49E+8 1.04E+10 2.05E+7 1.36E+7 6.95E+7 8.40E+6
*
APPENDIX A Evaluation of Default HPC Value for Liquid Effluents A-1 Salem OOCH Rev. 3 07/30/87
*
Appendix A Evaluation of Default HPC Value for Liquid Effluents Salem ODCH Rev. 3 07130187 In accordance with the requirements of Technical Specification
<3.3.3.10) the radioactive liquid effluent monitors shall-be operable with alarm setpoints established to ensure that the concentration of radioactive material at the discharge point does not exceed the HPC value of 10 CFR 201 Appendix B, Table 111 Column 2. The determination of allowable radionuclide concentration and corresponding alarm setpoint is a function of the individual radionuclide distribution and corresponding HPC values. In order to limit the need for routinely having to reestablish the alarm setpoints as a function of changing radionuclide distributions.
a default alarm setpoint can be established.
This default setpoint can be based on an evaluation of the radionuclide distribution of the liquid effluents from Salem and the effective HPC value for this distribution.
The effective HPC value for a radionuclide distribution is calculated by the equation:
where: HPCe Ci HPCi L Ci HPCe = ------------
<A. l > [
= an effective HPC .value for a mixture of radionuclide CuCi/ml) = concentration of radionuclide i in the mixture = the 10 CFR 20, Appendix 81 Table II, Column 2 HPC value for radionuclide i (uCi/ml> Based on the above equation and the radionuclide distribution in the effluents for past years from Salemi an effective HPC value can be determine.
Results are A-2
*
Salem ODCH Rev. ! 07/30/87 presented in Table A-1 and A-2 for Unit 1 and Unit 2, respectively
Considering the average effective MPC value for the years 1981 through 1986, it is reasonable to select an HPCe value of lE-05 uCi/ml as typical of liquid radwaste discharges.
Using this value to calculate the default R18 alarm setpoint value. results in a setpoint that: 1) Will not require frequent re-adjustment due to minor variations in the nuclide distribution which are typical of routine plant operations.
and 2) Will provide for a liquid radwaste discharge rate Cas evaluated for each batch release) that is compatible with plant operations (refer to Tables 1-1 and 1-2)
A-3 t
*
Nuclide "PCt * !uCi/11 l --------Na-24 3E-05 Cl'-51 2E-03 l'ln-54 1E-04 Fe-59 5E-05 Co-58 9E-05 Co-60 3E:.os ll'-95 6E-05 Nb-95 1E-04 Nb-91 9E-04 Tc-991 3E-03 51'-89 3E-06 51'-90 3E-07 "o-99 4E-05 Ag-1101
3E-05 5n-113 8E-05 5b-124 2E-05 5b:.125 1E-04 I-131 3E-07 I-133 1E-06 I-135 4E-06 Cs-134 9E-06 Cs-137 2E-05 Ba-140 2E:.05 La-140 2E-05 Total Ci __ Ci __ "PCi "PCe !uCi/11 l Table A.:.1 Calculation of Effective nPC 5ale1 Unit 1 . Activity Released fCiJ 5ale1 ODC" Rev. 3 07/30/87 1981 1982 1983 1984 1985 1986 ----------

2.4E-02 1.9E-03 5.3E-03 5.6E-03 6.2E-03 9.2E-04 5.9E-D2 1.4E-01 6.2E-02 5.3E-02 3.6E-02 6.0E-02 7.4E-02 2.1E-01 1.6E-01 1.9E-01 8.7E-02 1.9E-01 3.8E-03 8.6E-03 4.2E-02 5.8E-03 1.4E-03 2.4E-03 4.5E-01 1.7 1.8 1.6 6.6E-01 2.22 3.2E-01 . 9.1E-01 7 .1E-01 1.2 6.SE-01 3.1E-01 1.0E-02 1.1E-02 8.0E-03 1.8E-03 3.2E-03 4.3E-03 1.8E-02 4.8E-02 2.2E-02 1. 7E-02 1.3E-03 1.8E-02 N/OH 9.SE-03 3.6E-04 2.0E-02 7.2E-03 1.5E-03 N/D N/D N/D 1.6E-03 N/D N/D N/D N/D 1.2E-03 4.2E-04 1.7E-03 3.5E-07 N/D N/D N/D 2.2E-05 1.7E-04 3.1E-08 2.5E-04 1.0E-03 1.6E-03 1.9E-03 1.0E-04 N/D N/D 4.7E-03 N/D N/D N/D N/D 8.9E-04 2.2E-04 3.8E-04 9.4E-04 N/D 3.5E-04 6.2E-04 8.0E-03 1.4E-02 1. 7E-02 5.7E-03 8.4E-02 3.0E-01 6.8E-03 4.4E-02 4.9E-03 N/D 3.6E-02 1.tE-01 6.5E-02 2.4E-02 4.5E-02

7.9E-02 1.2E-01 8.8E-02 5.SE-03 3.3E-02 2.SE-02 1.9E-02 1.4E-03 N/D 3.SE-04 1.6E-03 1.2E-03 N/D N/D 6.0E-02 4.0E-02 1.8E-02 5.1E-02 1.6E-01 3.4E-01
7.6E-02 5.9E-02 3.0E-02 5.8E-02 2.1E-01 3.6E-01 N/D N/D 1.3E-02 2.1E-03 N/D N/D N/D 7.SE-03 1.3E-02 1.6E-02 1.1E-04 3.SE-04 1.59 3.24 3.00 3.32 1.93 3.75 4.86E+05 2.83E+05 1.66E+05 2.46E+05 3.42E+05 4.99E+05 3.3E-06 1.14E-05 1.80E-05 1.35E-05 5.63E-06 7.51E-06 f l'IPC value fol' unl'estricted al'ea fl'OI 10 CFR 20, Appendix B, Table II, Colu1n 2. ff N/D -not detected A-4 *1 I I
*
Table A-2 Calculation of Effective "PC -Sale1 Unit 2 Activity Released !Ci) -Sale1 ODC" Rev. 3 07/30/87 . --------------------------------------------------------------------------------
Nuclide "PCI 1981 1982 1983 1984 1985 1986 !uCi/111)

Na-24 3E-05 2.0E-01 1.2E-03 9.2E-03 4.4E-03 3.5E-03 3.6E-03 Cr-51 2E-OJ 9.5E-02 1.1E-01 4.6E-02 3.6E-02 3.5E-02 9.5E-02 "n-54 1E-04 4.4E-02 2.0E-01 1.4E-01 1.6E-01 1.1E-01 2.2E-01 Fe-59 5E-05 5.8E-03 5.6E-03 J.1E-02 7.6E-03 1.1E-03 4.0E-03 Co-58 9E-05 8.1E-01 1.7 1. 7 1.3 8.4E-01 3.32 Co-60 3E-05 2.6E-01 8.6E-01 5.7E:.01 9.8E-01 6.3E-01 3.8E-01 Zr-95 6E-05 1.0E-02 . 9.7E-03 5.2E-'03 1.2E-03 4.6E-03 1.1E-02 Nb-95 1E-04 1.5E-02 2.3E-02 1.6E-02 1.4E-02 1.4E-02 2.5E-02 Nb-97 9E-04 N/DH 1.1E-02 1.1E-02 2.1E-02 5.7E-03 2.7E-03 Tc-991 3E-03 N/D N/D N/D 1.4E-03 N/D N/D Sr-89 3E-06 N/D N/D 3.2E-04 J.2E-04 1.5E-03 4.1E-07 Sr-90 3E-07 N/D N/D N/D 4.1E-05 1.0E-04 3.2E-08 "o-99 4E-05 7.4E-05 1.7E-04 J.OE-03 t.4E-03 N/D N/D Ag-1101 3E-05 N/D 3.9E-03 N/D N/D N/D N/D Sn-113 8E-05 2.6E-04 1.6E-04 5.9E-04 t.2E-03 N/D 1.1E-03 Sb-124 2E-05 9.1E-O 1.0E-02 2.0E-02 3.0E-02 1.2E-03 1.2E-01 Sb-125* 1E-04 8.4E-03 1.0E-02 9.6E-02 3.6E-03 N/D 5.4E-02 I-131 JE-07 2.6E-02 1.3E-01 J.6E-02 4.2E-02 8.4E-02 t.2E-01 I-133 1E-06 6.0E-04 6.0E-03 5.4E-02 2.6E-02 1.2E-02 2.6E-03 I-135 4E-06 N/D N/D 1.6E-03 4.4E-04 N/D N/D Cs-134 9E-06 1.8E-02 5.1E-02 2.0E-02 2.bE-02 1.8E-01 3.bE-01 Cs-137 2E-05 2.9E-02 7.6E-02 J.bE-02 4.8E-02 2.3E-01 3.7E-01 Ba-140 2E-05 N/D N/D 9.8E-03 6.6E-03 N/D N/D La-140 *2E-05 N/D 6.7E-03 8.ff-02 N/D 6.9E-04 Total Ci 1.52 3.21 ' 2.89 2.74 2.15 5.09 1.16E+05 5.00E+05 2.26E+05 2.24E+05

3.56E+05 5.20E+05 "PCi "Pee !uCi/11} 1.31E-05 _ 6.42E-06 t.28E-05 1.22E-05 6.04E-06 9.79E-06 -------------------------
1 "PC value for unrestricted area fro1 10 CFR 20, Appendix 8, Table II, Colu1n 2. 11 N/D -not detected A-5
*
Salem ODCH Rev. 3 07/30/87 APPENDIX B Technical Basis for Effective Dose Factors LiQuid Radioactive Effluent B-1
*
Salem OOCH Rev. 3 07/30/87 APPENDIX B Technical Basis for Effective Dose Factors -Liquid Effluent Releases The radioactive liquid effluents for the years 19861 1985, 19841 19831 and 1982 were evaluated to determine the dose contribution of the radionuclide distribution.
This analysis was performed to evaluate the use of a limited dose analysis for determining environmental doses, providing a simplified method of determining compliance with the dose limits of Technical Specification 3.11.1.2.
For the radionuclide distribution of effluents from Salem. the controlling organ is the GI-LLI. The calculated Gl-LLI dose is predominately a function of the Fe-59, C0-581 C0-60 and Nb-95 releases.
The radionuclides1 Co-58 and C0-60 contribute the large majority of the calculated total body dose. The results of the evaluation for 19861 1985, and 1984 are presented in Table B-1 and Table B-2. For purposes of simplifying the details of the dose calculational process* it is conservative to identify a controlling, dose significant radionuclide and limit the calculational process to the use of the dose conversion factor for this nuclide. Multiplication of the total release (i.e ** cumulative activity for all , radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
For the evaluation of the maximum organ dose. it is conservative to use the Nb-95 dose conversion factor Cl.51 E+06 mrem/hr per uCi/ml1 GI-LLI>. By this approach, the maximum organ dose will be overestimated since this nuclide has the highest organ dose factor of all the radionuclides evaluated.
For the total body calculation.
the Fe-59 dose factor C7.27 E+04 mrem/hr per uCi/ml1 total body) is the highest among the identified dominant nuclides
B-2
*
Salem ODCH Rev. 3 07130187 For evaluating compliance with the dose limits of Technical Specification 3.11.1.2, the following simplified equations may be used: where: Dtb A Fe-59,TB VOL Ci CLI 1.67E-02 1.67E-02
VOL Dtb = --------------
A Fe-59,TB * '[, Ci cw = dose to the total body (mrem> <B.1> = 7.27E+04, total body ingestion dose conversion factor for Fe-59 <mrem/hr per uCi/ml) = volume of liquid effluent released (gal) = total concentration of all radionuclides
<uCi/ml> = average circulating water discharge rate during release period (gal/rain)
= conversion factor Chr/min) Substituting the value for the Fe-59 total body dose conversion factor, the equation simplified to: where: Dmax A Nb-95,GI-LLI 1.21E+03
VOL Dtb = --------------
cw 1.67E-02
VOL
A Nb-95,GI-LLI Dmax = ---------------------------
L: Ci cw = maximum organ dose (mrem) <B.2> <B.3> = 1.51E+06, Gi-LLI ingestion dose conversion factor for Nb-95 Cmrem/hr per uCi/*l> Substituting the value for A Nb-95.GI-LLI the equation simplifies to; 2.52E+04
VOL Dmax = --------------
cw B-3 [ Ci <B.4>
*
Salem ODCH Rev. 3 07/30/87 Tritium is not included in the limited analysis dose assessment for liquid releases.
because the potential dose resulting from normal reactor releases is relatively negligible.
The average annual tritium release from each Salem Unit is approximately 350 curies. The calculated total body dose from such a release is 2.4E-03 mrem/yr via the fish and invertebrate ingestion pathways.
This amounts to 0.08X of the design objective dose of 3 mrem/yr. Furthermore.
the release of tritium is a function of operating time and power level and is essentially unrelated to radwaste system operation
B-4 Sale1 ODC" Rev. 3 07/30/87 Table B-1 Adult Dose Contributions
Fish and Invertebrate Pathways Unit 1 1986 1985 1984 -----------------------------------

Radio-Release TB 6I-LLI -Liver Re lease TB 6I-LLI Liver Release TB 6I-LLI Liver nuclide !Cil Dose Dose Dose (Cil Dose Dose Dose (Ci) Dose Dose Dose Frac. Frac. Frac.
Frac. Frac. Frac. Frac. Frac. Frac. ------------------------Fe-59 2.40E-03 0.01 0.02 0.03 1.40E-03 0.01 0.04 0.04 5.83E-03 0.05 0.04 0.18 Co-58 2.22 0.25 0.42 0.10 6.60E-01 0.12 0.36 0.05 1.58 0.25 0.21 0.02 Co-60 3.10E-01 0.10 0.07 0.04 6.50E-01 0.34 0.42 0.15 1.20 0.54 0.42 0.34 Ag-110111 N/D f f f N/D f f f N/D f f f "n-54 1. 90E-01 0.02 0.06 0.10 8.70E-02 0.02 0.08 0.08 1. 93E-01 . 0.03 0.05 0.22 Nb-95 1.80E-02 f 0.42 f 1.30E-03 f 0.09 f 1. 74E-02 f 0.28 f Cs-137 3.60E-01 0.14 0.01 0.32 2. iOE-01 0.22 f 0.33 5.84E-02 0.05 f 0.11 Cs-134 3.40E-01 0.38 f 0.41 1.60E-01 . 0.29 f 0.35 5.06E-02 0.08 f 0.13 Cr-51 6.00E-02 f f f 3.60E-02 f f f 5.30E-02 f f f Total 3.50E+OO 1.80E+OO 3.33E+OO --------------------------
f less than 0.01 N/D = not detected
Table B-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1986 1985 1984 -----------------------------------

Radio-Release TB 61-LLI Liver Release TB 6I-LLI Liver Release TB 61-LLI Liver nuclide {Cil Dose Dose Dose !Cil Dose Dose Dose (Cil Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac. . Frac. Frac. Frac * ------------------------Fe-59 4.00E-03 0.02 0.03 0.05 1.10E-03 0.01 0.02 0.02 7.56E-03 0.08 0.06 0.24 Co-58 3.32 0.32 0.44 0.13 8.40E-01 0.14 0.23 0.06 1.30 0.25 0.21 0.13 Co-60 3.80E-01 0.10 0.06 0.04 6.30E-01 O.JO 0.21 0.13 9.79E-01 0.53 0.41 0.28 Ag-1101 N/D f f f N/D f f f N/D f f f "n-54 2.20E-01 0.02 0.05 0.10 1.10E-01 0.02 0.05 0.09 1.61E-01 O.OJ 0.05 0.18 Nb-95 2.50E-02 f 0.41 f 1.40E-02 f 0.48 f 1.36E-02 f 0.27 f Cs-137 3.70E-01 0.20 f 0.29 2.30E-01 0.2.J f 0.33 4.81E-02 0.05 f 0.10 Cs-134 3.60E-01 0.34 f 0.38 1.80E-01 0.30 f 0.35 2.63E-02 0.05 f 0.07 Cr-51 9.50E-02 f f f 3.50E-02 f f f 3.64E-02 f f f Total 4.77E+OO 2.04E+OO 2.75E+OO --------------------------
f Jess* than 0.01 N/D = not detected
B-5
*
Salem ODCH Rev. 3 07/30/87 APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent C-1
* '.
Salem ODCH Rev. 3 07/30/87 APPENDIX C Technical Bases for Effective Dose Factors -Gaseous Radioactive Effluents The evaluation of doses due to releases of radioactive material to the atmosphere can be simplified by the use of effective dose transfer factors instead of using dose factors which are radionuclide specific.
These eHective factors, which can be based on typical radionuclide distributions of releases, can be applied to the total radioactivity released to approximate the dose in the environment (i.e., instead of having to perform individual radionuclide dose analyses only a single multiplication CK H or N times the total eU eH eH quantity of radioactive material released would be needed). This approach provides a reasonable estimate of the actual dose while eliminating the need for a detailed calculational technique
Effective dose transfer factors are calculated by the following equations:
where: KeH Ki fi where: KeH = L <Ki
fi > ( c .1) = the effective total body dose factor due to gamma emissions from all noble gases released = the total body dose factor due to gamma emissions from each noble gas radionuclide i released = the fractional abundance of noble gas radionuclide relative to the total noble gas activity CL+ 1.1 H>eH = ["((Li + 1.1 Mil
fil <C.21 <L + 1.1 HleH = the effective skin dose factor due to beta and gamma emissions from all no_b le gases rel eased
= the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released C-2
* *-where: tfeff Hi where: Neff Salem ODCH Rev. 3 07/30/87 Heff = L CHi
fi > CC.3) = the effective air dose factor due to gamma emissions from all noble gases released = the air dose factor due to gamma emissions from each noble gas radionuclide i released Neff = l:<Ni
fi) CC.4> = the effective air dose factor due to beta emissions all noble gases released Ni = the air dose factor due to beta emissions from each noble gas radionuclide i released Normally, it would be expected that past radioactive effluent data would be used for the determination of the effective dose factors. However. the noble gas releases from Salem have been maintained to such negligible quantities that the inherent variability in the data makes any meaningful evaluations difficult.
For the past years. the total noble gas have been limited to 1400 Ci for 19821 900 Ci for 19831 21000 Ci for 19841 21800 Ci for 1985. and 21700 for 1986. Therefore.
in order to provide a reasonable basis for the derivation of the effective noble gas dose factors1 the primary coolant source term from ANSI N237-1976/ANS-18.1. "Source Term Specifications*" has been used as representing a typical distribution.
The effective dose factors as derived are presented in Tab le C-1. To provide an additional degree 4f conservatism.
a factor of 0.50 is introduced into the dose calculational process when the effective dose transfer factor is used. This conservatism provides additional assurance that the evaluation of doses by the use of a single effective factor will not significantly underestimate any actual doses in the environment.
C-3
*
Salem ODCH Rev. 3 07/30/87 For evaluating compliance with the dose limits of Technical Specification 3.11.2.2, the following simplified equations may be used: D and D where: 3.17E-08 = --------* X/Q
Heff a.so 3.17E-08
I Qi = --------* X/Q
Neff * "'[_ Qi a.so <C.S> <C.6> D = air dose due to gamma emissions for the cumulative release of all noble gases Cmrad) D = air dose due to beta emissions for the release of all noble gases (mrad) X/Q = atmospheric dispersion to the controlling site boundary He ff = Neff = Qi = 3.17E-08 = a.so = (sec/m3> S.3E+02. effective gamma-air dose factor <mrad/yr per uCi/m3) 1.1E+03. effective beta-air dose factor (mrad/yr per uCi/m3) cumulative release for all noble gas radionuclides (uCi) conversion factor (yr/sec> conservatism factor to account for the variability in the effluent data Combining the constants.
the dose calculational equations simplify to: D = 3.5E-05
X/Q
l:° Qi <C.7> and D = 7.0E-05
X/Q * )__ Qi <C.8> The effective dose factors are used a very limited basis for the purpose of facilitating the timely assessment of radioactive effluent releases.
particularly during periods of computer malfunction where a detailed dose assessment may be unavailable
C-4
*
Noble Gases -Total Body and Skin Radionuclide Kr-85 Kr-88 Xe-133m Xe-133 Xe-135 Total Noble Gases -Air Radionuclide Kr-85 Kr-88 Xe-133m Xe-133 Xe-135 Total 0.01 0.01 0.01 0.95 0.02 0.01 0.01 0.01 0.95 0.02 Salem ODCH Rev. 3 07/30/87 Table C-1 Effective Dose Factors Total Body Effective Dose Factor KeH (mrem/yr per uCi/m3) 1.5E+02 2.5E+OO 3.0E+02 3.6E+01 4.8E+02 Gamma Air Effective Dose Factor HeH (mrad/yr per uCi/m3) 1.5E+02 3.3E+OO 3.4E+02 3.8E+01 5.3E+02 Skin Effective Dose Factor <L+ 1.1 H>eH (mrem/yr per uCi/m3) 1.4E+01 1.9E+02 1.4E+01 6.6E+02 7.9E+01 9.6E+02 Beta Air Effective Dose Factor NeH (mrad/yr per uCi/m3) 2.0E+01 2.9E+01 1.5E+01 1.0E+03 4.9E+01 1.1E+03
Based on Noble gas distribution from ANSI N237-1976/ANSI-18.1, Source Term Specifications." C-5
*
APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent D-1
*
Salem ODCH Rev. 3 07/30/97 APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent Releases The pathway dose factors for the controlling infant,age group were evaluated to determine the controlling pathway, organ and radionuclide.
This was performed to provide a simplified method for determining compliance with Technical Specification 3.11.2.3 For the infant age group, the controlling pathway is the grass-milk-cow (g/m/c) pathway. An infant receives a greater radiation dose from the g/m/c pathway than any other pathway. Of this g/m/c pathway, the maximum exposed organ including the total body, is the thyroid, and the highest dose contributor is radionuclide I-131. The results for this evaluation are presented in Table D-1. For purposes of simplifying the details of the dose calculation process, it is consetvative to identify a controlling, significant and radionuclide and limit the calculation process to the use of the dose factor for the organ and radionuclide.
Multiplication of the total release (i.e. cumulative activity for all radionuclides) by this dose conversion factor provides for a dose method that is simplified while also being conservative.
For the evaluation of the dose commitment via a controlling pathway and age group, it is conservative to use the infant, g/m/c, thyroid, I-131 pathway dose factor <1.675E12 m2 mrem/yr per uCi/sec>.
By this approach, the maximum dose. commitment will be overestimated since I-131 has the highest pathway dose factor of all radionuclides evaluated.
For evaluating compliance with the dose limits of Technical Specification D-2
*
Salem ODCH Rev. 3 07/30/87 3.11.2.3.
the following simplified equation may be used: where: Dmax = 3.17E-8
W
RI-131
t_Qi Dmax = maxi111um organ dose Cmre111>
W = atmospheric dispersion ,parameters to the controlling X/Q = D/Q = Qi = 3.17E-8 = RI-131 = = location(s) as identified in Table 3.2-4. atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways Csec/m3> atmosperic deposition for vegetation1 milk nad ground plane exposure pathways Cm-2> cumulative release over the period of interest for radioiodines and particulates conversion factor (yr/sec> 1-131 dose parameter for the thyroid for the identified controlling pathway 1.675E12 Cm2 mrem/yr per uCi/sec), infant thyroid dose parameter with the cow-milk-grass pathway controlling The ground plane exposure and inhalation pathways need not be considered when the above simplified calculation method is used because fo the overall negligible contribution of these pathways to the total thyroid dose. It is recognized that for some particulate radioiodines
<e.g., Co-60 and Cs-137), the ground exposure pathway may represent a higher dose contribution than either the vegetation or milk pathway. However. use of the I-131 thyroid dose parameter for all radionuclides will maximize the organ dose calculation.
especially considering that no other radionuclide has a higher dose parameter for any organ via any pathway than _I-131 for the thyroid via the milk pathway <see Table D-1>. The location of exposure pathways and the maximum organ soe calculation may be based on the available pathways in the surrounding environment of Salem as identified by the annual land-use census <Technical Specification 3.12.2>. Otherwise.
the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-4
D-3
Iac:gd Qc:gana Total Body Liver Thyroid Kidney Lung GI-LLI *
Salem ODCH Rev. 3 a7/3a/S7 Table D-1 Infant Dose Fraction of TQtal Organ and Body Dose eer1::1wers Gc:aaa=C2w=IH1.k a.a2 a.23 a.59 a.a2 a.a1 a.a2 2£ D2at b f aibl!ID::t fa:t.bwa::t
£ Grass-Cow-Hilk a.92 Ground Plane a.as Inhalation
D-4 YC:QYD.d f hnt a.15 a.14 a.15 a.15 a.a2 a.15
*
Salem ODCH Rev. 3 07/30/87 Appendix E Radiological Environmental Monitoring Program Sample Type. Location and Analysis E-1
PATHWAY STATION CODE I. DIRECT lFl 1G3 # 2S2 2El 2F2 2F5 2F6 3El 3F2 3F3 3Gl # 3Hl # 3H3 # 4D2 5Sl 501 5Fl 6S2 6Fl 7Sl 7F2 9El lOSl 1001 10F2 lOGl # # Control Station M P85 183/15 4-dh
E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION COLLECTION METHOD 5.8 miles N of or1g1n 2 TLD's will be collected 18.5 miles N of origin from each location quarterly 0.4 miles NNE of origin 4.4 miles NNE of origin 8.7 miles NNE of origin 7.4 miles NNE of origin 7.3 miles NNE of origin 4.1 miles NE of origin 5.1 miles NE of origin 8.6 miles NE of origin 16.6 miles NE of origin 32 miles NE of origin 110 miles NE of origin 3.7 miles ENE of origin 1.0 mile E of origin 3.5 miles E of origin 8.0 miles E of origin 0.2 miles ESE of or1g1n 6.4 miles ESE of origin 0.12 miles SE of origin 9.1 miles SE of origin 4.2 s of origin 0.14 miles SSW of origin 3.9 miles SSW of origin 5.8 miles SSW of origin Collected quarterly 11.*6 miles SSW of origin Rev.
Page.I of 7 ANALYSES Gamma dose quarterly Gamma-dose quarterly 3 7/30/87
PATHWAY I. DIRECT (Con't) STATION CODE 10F2 lOGl # llSl 11E2 . llFl 12El 12Fl 13El 13F.4 13F2 13F3 1401 14F2 15F3 16El 16F2 16Gl #
TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 5.8 miles SSW oi or1g1n 11.6 miles SSW of origin 0.07 miles SW of origin
5.0 miles SW of origin 5.2 miles SW of origin 4.4 miles WSW of origin 9.4 miles WSW 6r origin . 4.2 miles NE of origin 9.8 miles W of origin 6.5 miles W of origin 9.3 miles W of origin 3.4 miles WNW of origin 6.6 miles WNW of origin 5.4 miles NW of origin 4.1 miles NNW of origin 8.1 miles NNW-of origin 14.8 miles NNW of origin COLLECTION METHOD Collected quarterly
Page 2 of 7 'ANALYSES Gamma dose quarterly
Control Station (in addition to controls listed, two additional program controls are used for internal studies and are referred to as SITE-CAL and SITE-ZERO).
M P85 183/15 4-dh Rev. 3 7/30/87
EXPOSURE PATHWAY II. AIRBORNE (a) P A R T I c L A T E s STATION CODE 2S2 2F2 3H3 # 1001 16El # Control Station M P85 183/15 4-dh
TABLE E-1 ' ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 0.4 miles NNE of origin 8.7 miles NNE of origin 110 miles NE of origin 3.9 miles SSW of origin 4.1 miles NNW of origin COLLECTION METHOD Continous low volume air sampler. Sample collected every week along with filter change.
Page 3 of 7 TYPE AND FREQUENCY OF ANALYSES beta analysis on each weekly sample. Gamma spectrometry shall be performed if gross beta exceeds ten times the yearly mean of control station value. Gross beta analysis done > 24hr. after sampling to allow for Radon and Theron daughter decay Gamma isotopic analysis on quarterly composite Rev. 3 7 /30/87
EXPOSURE PATHWAY II. AIRBORNE (Con't) ( b) I 0 D I N E STATION CODE 2S2 2F2 3H3 # 16El 1001 # Control Station M P85 183/15 4-dh ** TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 0.4 miles NNE of origin 8.7 miles NNE of origin 110 miles NE of origin 4.1 miles NNW of origin 3.9 miles SSW of origin COLLECTION METHOD A TEDA impregnated charcoal flow-through cartridge connected to.air particulate air sampler and is collected weekly at filter change.
Page 4 of 7 TYPE AND FREQUENCY OF ANALYSES Iodine 131 analyses are performed on each weekly sample. Rev. 3 7/30/87
PATHWAY III. WATER (a) S u R F A c E ( b) G R 0 u N D , ( c) s E D I* M E N T STATION CODE 7El 12Cl # 16Fl 2S3 5Dl 3El 7El 12Cl # 16Fl # Control Station M P85 183/15 4-d.h !:
TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 1 mile W of Mad Horse Creek; 4.5 miles SE of origin West bank opposite Artificial Island; 2.5 miles WSW of origin C&D canal; 6.9 miles NNW of origin Fresh water holding tank; 700 feet NNW of origin Local farm; 3.5 miles E of origin Local farm; 4.1 miles NE of origin 1 miles W of Mad Horse Creek 4.5 miles SE of origin West bank opposite Artificial Island; 2.5 miles WSW of origin C&D Canal; 6.9 miles NNW of origin COLLECTION METHOD Sample to be collected monthly providing winter icing conditions allow sample collection Collected monthly A sediment sample is taken
Page 5 of 7 ANALYSES Gamma isotopic analysis monthly H-3 on quarterly composite Gamma isotopic monthly Tritium analysis monthly Gamma isotopic anaylsis -semi-* annually Rev. 3 7 /30/87
PATHWAY IV. INGESTION
{a) M { b) I L K F I s H STATION CODE 2F7 3Gl # 5F2 13E3 14Fl llAl l2Cl # # Control Station M P85 183/15 4-dh
TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 5.7 miles NNE of origin 16.6 miles NE of origin 7.0 miles E of origin 4.9 miles w of origin 5.5 miles WNW of origin Outfall area; approx. 650 feet SW of origin West bank opposite Artificial Island;* 2.5 miles WSW of origin COLLECTION METHOD Sample of fresh milk is collected for each farm semimonthly when cows are on pasture, monthly at other times. Two batch samples of fish are sealed in plastic bag or jar and frozen semiannually or when in in season
Page 6 of 7 ANALYSES
Gamma isotopic and I-131 analyses on each sample on collection Gamma isotopic analysis of edible portion on collection Rev. 3 7 /30/'d7
PATHWAY IV. INGESTION (Cont'd) ( c) I N v E R T E B R A T E s STATION CODE llAl 12Cl # # Control Station M P85 183/15 4-dh
TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION Outfall area; approx. 650 feet SW of origin West bank opposite Artificial Island; 2.5 miles WSW of origin COLLECTION METHOD Two batch samples of crab are sealed in a plastic bag oi jar and frozen semiannually or when in season.
Page 7 of 7 ANALYSES Gamma Isotopic analysis of edible portion on collection Rev. 3 7/30/87
* * .. * *e e' i! .. ..... ***** ... FIGURE E-1 OFFSITE SAMPLING LOCATIONS ARTIFICIAL ISLAND \ , \ . Rev. 3 7/30/87
* \
FIGURE E-2 ONSITE SAMPLING ARTIFICIAL LOCATIONS
ISLAND \ Rev. 3 7/30/87

@@ Line 32: / Line 32: @@
 * 10 1.6 Liquid Effluent Dose Projection
 * * * * * * * * * * * * *
-* 12 Gaseous Effluents
+* 12 Gaseous Effluents 2.1 Radiation Honitoring Instrumentation and Controls
+* 2.2 Gaseous Effluent Honitor Setpoint Determination 2.2.1 Containment and Plant Honitor **** 2.2.2 Conservative Default Values 2.3 Gaseous Effluent Instantaneous
-===2.1 Radiation===
-Honitoring Instrumentation and Controls
-* 2.2 Gaseous Effluent Honitor Setpoint Determination
-====2.2.1 Containment====
-and Plant Honitor **** 2.2.2 Conservative Default Values 2.3 Gaseous Effluent Instantaneous
 * *
 * 13 * *
@@ Line 55: / Line 47: @@
 * 23 2.5.1 UNRESTRICTED AREA Dose -Radioiodine and Particulates
 * 23 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates
 .****************
+.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations
-===2.6 Secondary===
+************
+.7 Gaseous Effluent Dose Projection
-Side Radioactive Gaseous Effluents and Dose Calculations
-************
-===2.7 Gaseous===
-Effluent Dose Projection
 **********
 Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY 3.2 Doses to HEHBERS OF THE PUBLIC -40 CFR 190
@@ Line 242: / Line 229: @@
 where: SP = SEN = bkg = AF = SP = CAF * '[Ci
 * SEN + FRACJ + bkg (2.3> alarm setpoint corresponding to the maximum allowable release rate (Cp*) monitor sensitivity (cpm per uCi/c*3> background of the monitor (Cp*) administrative allocation factor for the specific monitor and type release. which corresponds to the fraction of the total allowable release rate that is administratively allocated to the release. The allocation factor <AF> is an administrative control imposed to ensure that combined releases from Salem Units 1 and 2 and Hope Creek will not exceed the refulatory limits on release rate from the site (i.e., the release rate limits of Technical Specification 3.11.2.1).
-Normally, the combined AF value for Salem Units 1 and 2 is 0.5 (0.25 per unit), with the remainder
+Normally, the combined AF value for Salem Units 1 and 2 is 0.5 (0.25 per unit), with the remainder 0.5 allocated to Hope Creek. Any increase in AF above 0.5 for the Salem Nuclear Generating Station will be coordinated with the Hope Creek Generating Station to ensure that the -combined alloiation factors for all units do not exceed 1.0. 2.2.2 A conservative alarm setpoint can be established.
-===0.5 allocated===
-to Hope Creek. Any increase in AF above 0.5 for the Salem Nuclear Generating Station will be coordinated with the Hope Creek Generating Station to ensure that the -combined alloiation factors for all units do not exceed 1.0. 2.2.2 A conservative alarm setpoint can be established.
 in lieu of the individual radionuclide evaluation based on the grab sample analysis*
 to eliminate the potential of periodically having to adjust the setpoint to reflect minor changes in radionuclide distribution and variations in release flow rate. The alarm setpoint may be conservatively determined by the default presented in Table 2-1 and 2-2 for Units 1 and 21 respectively
@@ Line 326: / Line 309: @@
 * Table 2-4 need be evaluated for compliance with Technical Specification
 *
 * 3.11.2.3.
+.5.2 Simplified Dose Calculation for Radioiodinea and Particulatea.
-====2.5.2 Simplified====
-Dose Calculation for Radioiodinea and Particulatea.
 In lieu of the individual radionuclide CI-131 and particulates>
 doae aaseaament aa presented above. the following simplified doae calculational equation may be uaed for -verifying compliance with the doae limita of Technical Specification 3.11.2.3 (refer to Appendix D for the derivation and justification of thia simplified method). where: Dmax * = RI-131 = = w = Qi = Dmax = 3.17E-08
@@ Line 393: / Line 373: @@
 For the likely most expoeed HEHBER OF THE-PUBLIC in the vicinity of Artificial Ieland1 the source* of exposure need only consider the Sale* Nuclear Generating Station and the Hope Creek Nuclear Generating Station: No other fuel cycle facilities contribute to the HEHBER OF THE PUBLIC doae for the Artificial Ieland vicinity.
 The doee contribution from the operation of Hope Creek Nuclear Generating Station will be estimated baaed on the *ethoda as presented in the Hope Creek Offaite Dose Calculation Hanual CHCGS ODCH>
 * As appropriate for de*onstrating/evaluating compliance with the limits of Technical Specification 3.11.4 <40 CFR 190)1 the result* of the environmental monitoring program *ay be used for providing data on actual measured levels of radioactive material in the actual pathways of exposure.
+.2.1 Effluent Doee Calculation*.
-====3.2.1 Effluent====
-Doee Calculation*.
 For purpoees of implementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting requirements of 6.9.1.11 CRERR>s doae calculations for the Salem Nuclear Generating Station may be performed using the calculational method* contained within thie ODCH: the conservative controlling pathway* and locations of Table 2-4 or the actual pathway* and locations ae identified by the land use ceneue <Technical Specification 3/4.12.2) may be uaed. Average annual meteorological diapereion 30
 * *
@@ Line 710: / Line 688: @@
 * Salem ODCH Rev. 3 07/30/87 Appendix E Radiological Environmental Monitoring Program Sample Type. Location and Analysis E-1
 * PATHWAY STATION CODE I. DIRECT lFl 1G3 # 2S2 2El 2F2 2F5 2F6 3El 3F2 3F3 3Gl # 3Hl # 3H3 # 4D2 5Sl 501 5Fl 6S2 6Fl 7Sl 7F2 9El lOSl 1001 10F2 lOGl # # Control Station M P85 183/15 4-dh
-* E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION COLLECTION METHOD 5.8 miles N of or1g1n 2 TLD's will be collected 18.5 miles N of origin from each location quarterly
+* E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION COLLECTION METHOD 5.8 miles N of or1g1n 2 TLD's will be collected 18.5 miles N of origin from each location quarterly 0.4 miles NNE of origin 4.4 miles NNE of origin 8.7 miles NNE of origin 7.4 miles NNE of origin 7.3 miles NNE of origin 4.1 miles NE of origin 5.1 miles NE of origin 8.6 miles NE of origin 16.6 miles NE of origin 32 miles NE of origin 110 miles NE of origin 3.7 miles ENE of origin 1.0 mile E of origin 3.5 miles E of origin 8.0 miles E of origin 0.2 miles ESE of or1g1n 6.4 miles ESE of origin 0.12 miles SE of origin 9.1 miles SE of origin 4.2 s of origin 0.14 miles SSW of origin 3.9 miles SSW of origin 5.8 miles SSW of origin Collected quarterly 11.*6 miles SSW of origin Rev.
-===0.4 miles===
-NNE of origin 4.4 miles NNE of origin 8.7 miles NNE of origin 7.4 miles NNE of origin 7.3 miles NNE of origin 4.1 miles NE of origin 5.1 miles NE of origin 8.6 miles NE of origin 16.6 miles NE of origin 32 miles NE of origin 110 miles NE of origin 3.7 miles ENE of origin 1.0 mile E of origin 3.5 miles E of origin 8.0 miles E of origin 0.2 miles ESE of or1g1n 6.4 miles ESE of origin 0.12 miles SE of origin 9.1 miles SE of origin 4.2 s of origin 0.14 miles SSW of origin 3.9 miles SSW of origin 5.8 miles SSW of origin Collected quarterly 11.*6 miles SSW of origin Rev.
 * Page.I of 7 ANALYSES Gamma dose quarterly Gamma-dose quarterly 3 7/30/87
 * PATHWAY I. DIRECT (Con't) STATION CODE 10F2 lOGl # llSl 11E2 . llFl 12El 12Fl 13El 13F.4 13F2 13F3 1401 14F2 15F3 16El 16F2 16Gl #

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