Text

Rev 4 to ODCM
	ML20084U591
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Site:	Davis Besse
Issue date:	02/22/1991
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DAVIS-BESSE OFFSITE DOSE CALCULATION MANUAL Revision 4. 1991 O

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v THE TOLEDO EDISON COMPANY DAVIS-BESSE NUCLEAR POWER STATION OFFSITE DOSE CALCULATIONS MANUAL l

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TABLE OF CONftNTS

1.0 INTRODUCTION

t 2.0 LIQUID EFFLURKr$.................

2 2.1 Radiation Monitoring Instrumentation and Controls..

2 2.1.1 Technical Specification 6.8.4 Requirement 3

2.1.2 Non Technical Specification Monitors........

4 2.2 Sampling and Analysis of Liquid Effluents..........

4 2.2.1 Batch Vaste Release Tanks.............

5 2.2.2 Turbine Building Susp/Stors Sever Drain (TBS /SSD).

5 2.2.3 Condensate Deelneraliser Backvash.........

6 2.3 Liquid Effluent Monitor Setpoints..............

7 2.3.1 Liquid Radvaste Effluent Line Monitor.......

8 2.3.2 Turbine Building Sump /Stora Sever Drain Monitor......................

11 2.3.3 Alara Setpoints for the Hon Technical Specification Radiation Monitors..........

12 2.3.4 Alara Response - Eveluating Actual Release Conditions.....................

12 2.4 Liquid Effluent Dose Calculations - 10 CFR 50........

13 2.4.1 MEMBER OF THE FUSLIC Dose - Liquid If fluents....

13 2.4.2 Simplified Liquid Effluent Dose Calculation....

16 2.4.3 Contaminated T35/SSD System - Dose Calculation...

17 2.5 Liquid Effluent Dose Projections...............

17 31 3.0 CASIOUS EFFLUENTS........................

3.1 Radiation Monitoring Instrumentation and Controls......

31 Station Vent Stack.................

32 3.1.1

. Vaste Gas Decay System...............

32 3.1.2 6

3.1.3 Containannt Purge Rxhaust Filter Monitor......

33 3.1.4 Hydrogen Purge Line................

33 3.1.5 Vaste Gas Synten Orygen Monitor..........

33 3.2 Sampling and Analysis of Gaseous Effluents 33 3.2.1 Station Vent Release................

33 3.2.2 Vaste Gas Decay Tank Reles e and 34 Containment FtRGE (SATCE).....

3.3 Gaseous Effluent Monitor Setpolat Determination.......

34 3.3.1 Station Vent....................

34 3.3.2 Conservative, Generic Alara Setpoints.......

36 3.3.3 Gaseous Itfluent Alarm Response -

Evaluating Actual Release Conditions........

37 3.4 Reletoe Rate tvaluation -

vasta Gas Decay Tank Releases and Containment FURCE.....

38 1

I

o (3.0 CASROUS RFFLUENIS - continued) 3.5 Ousatifying Releases. Noble Cases..............

38 3.5.1 Quantifying Releases Using Station Vent Noble Cas Monitor.................

38 3.5.2 Quantifying Releases with Inoperable Monitors...

39 3.6 SITE BOUNDARY Dose Rate - Radiciodine and Particulates 40 3.6.1 Simplified. Dose Rate Evaluation for Radiotodines and Particulates...........

40 3.7 Noble Gas Effluent Dose Calculations - 10 CFR 50 41 3.7.1 UNRESTRICTED AREA Dose - Noble Cases.....

41 3.7.2 Simplified Dose Calculation for Noble Cassa....

42 3.8 Radiotodine and Particulate Dose Calculations -

10 CFR 50..........................

43 3.8.1 UNRESTRICTED ARIA Dose - Radiolodine and Particulates..................

43 3.8.2 Simplified Dose Calculation for Radiotodines and Particulates..................

45 3.9 Gaseous Effluent Dose Projection 45 4.0 SPECIAL DOSE ANALYSIS.=.....................

75 4.1 Doses To The Public Due To Activities Inside the SITE SOUNDARY 75 4.2 Doses to MEMBERS OF THE FUBLIC - 40 CFR 190.........

75 4.2.1 Effluent Dose Calculations.............

77 4.2.2 Direct Exposure Dose Determination -

Onsite Sources...................

78 4.2.3 Dose Assessment Based on Radiological Environmental Monitoring Data...........

79 4.2.4 Use of Environmental TLD for Assessing Doses Due to Noble cas Releases.............

81

5.0 ASSESSMENT

0F LAND USE CENSUS DATA................

83 5.1 Land Use census as Required by TS 6.8.4...........

83 5.2 Land Use Census to Support Realistic Dose Assessment 85 6.0 RADIOLOGICAL INVTRONMENTAL MONITORIleG FROGRAN 86

6. l' Frogram Description.....................

86 6.2 Reporting 14rels 87 6.3 Interlaboratory Comparison Program 88 7.0 ADMINISTRATIVE CONTROLS 7.1 Annual Radiological Environmental Operating Report 100 7.2 Sesiannual effluent and Weste Disposal Report.

100 7.3 Special Reports.......................

101 7.4 Major Changes to Radioactive Liquid and Caseous Veste Treatment Systems...................

102 7.5 Definitions.........................

103 11 0

APPENDICES APPENDIX A -

Technical Basis for Simplified Dose Calculations, Liquid Effluent Releases.............

A-1 APPENDIX B -

Technical Basis for Effective Dose Factors Caseous Effluent Releases................

B-1 APPENDIX C -

Radiological Environmental Monitoring Program, Sample Location Maps............

C-1 APPENDIX J -

Justifications......................

J-l LIST OF TABLES Table 2 Radioactive Liquid Effluent Monitoring Instrumentation 20 Table 2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements................

22 Table 2 Radioactive Liquid Vaste Sampling and Analysis Program..

24 Table 2 Limiting Radionuclide concentrations in Secondary Side Clean-up Resins for Ad ovable Discharges to Onsite Settling Basin......................

27 Tabis 2 Davis-Besse Site-specific Liquid Ingestion Dose Commitment Factors, A 28 g.

Table 2 Eioaccumulation Factors (Bri) 29 Table 3 Radioactive Gaseous Effluent Monitoring Instrumentation.

48 Table 3 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements................

51 Table 3 Radioactive Gaseous Waste Sampling and Analysis Program.

53 Table 3 Atmospheric Dispersion Parameters............

56 Table 3 Dose Factors for Noble Gases.

57 Table 3 Controlling Incations, Fathwys and Atmospheric Dispersion for Dose calculations.......

58 Table 3 Gaseous Effluent Pathway Dose Commitment Factors.....

59 Table 4 Roccamanded Exposure Rates in Lieu of Site Spscific Data....................

82 Table 6 Radiological Environmental Monitoring Prograa......

89 Table 6 3ampling Iocations 94 111

-l (List of Tables - Cont.)

Table 6 towe r Limits of Detection................

97 Table 6 Reporting levels for Radioactivity Concentrations in Environmental.............

99 Table B Default Noble Gas Radionuclide Distribution of Gaseous Effluents B-4 Table B Effectin Dose ractors - Noble Gas Effluents B-5 LIST Or FIGURES Liquid RadioactiW Effluent Monitoring and rigure 2-1 Processing Diagram 30 rigure 3-1 Gaseous Radioactive Effluent Monitoring and Ventilation Systems Diagram...............

74 O

I l

l I

O DAVIS-BESSE OFFSITE DOSE CALCULATION MANUAL Revision 4, 1991 O

O

DAVIS-BESSE OFFSITE DOSE CALCULATION HANUAL 1.0 IttrRODUCTION The Davis-Besse Offaite Dose Calculation Manual (ODCH) describes the methodology and parameters used in 1) determining the radioactive material release rates and cumulative releases 2) calculating the radioactive liquid and gaseous effluent monitoring instrumentation alare/ trip setpoints and 3) calculating the corresponding dose rates and cumulative quarterly and yearly doses. The Radiological Environmental Honitoring Program is also described. Sampling locations, media and collection frequencies are presented analytical requirements are specified. The methodology provided in this manual is acceptable for use in demonstrating compliance with concentration limits of 10 CFR 20.106 and the cumulative dose criteria of 10 CFR 50, Appendix I and 40 CFR 190, and the Davis-Besse Radiological Effluent Technical Specifications.

The exposure pathway and dose modeling as presented in this ODCM vill, in general, provide estimates (e.g., calculational results) that are conservative (i.e., higher than actual exposures in the environment). This conservatism does not invalidate the modeling since the main purpose of these calculations is for demonstrating "As Lov As is Reasonably Achievable" (ALARA) for radioactive effluents. In using these models for evaluation and controlling actual effluents, further simplification and conservatism may be applied. For purposes of demonstrating compliance with the EPA environmental dose standard for the Uranium Fuel Cycle (40 CFR 190), more realistic dose assessment modeling may be used to provide more accurate assessment of actual radiation exposures resulting from the operation of the Davis-Besse Nuclear Power Station.

The ODCH will be maintained at the station for use as a reference guide and training document of accepted methodologies and calculations. Changes to the ODCH calculational methodologies and parameters vill be made as necessary to ensure reasonable conservatism in keeping with the principles of 10 CFR 50, Appendix I, Section III and IV. Questions about the ODCH should be directed to the Manager - Radiological Control.

NOTE:

Throughout this document words appearing all capitalized denote definitions specified in the Davis-Besse Technical Specifications (TS), in Section 7.5 of this manual, or common acronyms.

Section 2.0 of the ODCH describes equipment for monitoring and controlling liquid effluents, sampling requirements, and dose evaluation methods.

Section 3.0 provides similar information on gaseous effluent controls, sampling, and dose evaluation. Section 4.0 describes special dose analyses required for Regulatory Guide 1.21, Semiannual Effluent Reporting and EPA Environmental Dose Standard of 40 CTR 190. Section 5.0 describes the role of the annual land use census in identifying the controlling pr.thvays and locations of exposure for assessing the potential offsite doses. Section 6.0 describes the Radiological Environmental Honitoring Program. Section 7.0 describes the environmental, effluent and special reporting requirements, procedural requirements for major changes to liquid and gaseous radvaste systems, and definitions.

Davis-Besse ODCH 1

Revision 4, 1991

0 2.0 LIQUID EFFLUENTS This section summarises information on the liquid effluent radiation monitoring instrumentation and controls. More detailed information is provided in the Davis-Besse USAR, Section 11.2. Liquid Vaste Systems and associated design drawings from which this summary was derived. This j

section also describes the sampling and analysis required by Technical Specifications. Hethods for calculating alarm setpoints for the liquid effluent monitors are presentedi methods for evaluating doses from liquid effluents are derived.

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm / trip setpoints j

for these instruments shall be calculated in accordance with methods in i

Section 2.3 to ensure that the alarm / trip vill occur prior to exceeding the limits of 10 CFR Part 20.

The radioactive liquid effluent monitoring instrumentation channels listed j

in Table 2-1 shall be OPERABLE vith their alare/ trip setpoints set to ensure that the limits of ODCH Section 2.2 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters of Section 2.3.

2.1 Radiation Monitoring Instrumentation and Controls This Section summarizes the instrumentation and controls that monitor the liquid offluents. This discussion focuses on the role of this equipment in assuring compliance with the Davis-Besse Technical Specifications and ODCH.

Location and control function of the monitors are displayed in OOCH Figure 2-1.

Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the l

frequencies shown in Table 2-2.

Each of these operations shall be performed within the specified time interval with a maximum allowable extension not to exceed 25 percent of the specified interval.-

l l

NOTE: The monitors indicated in a, b, and c below are inoperable if surveillances are not performed or setpoints are less conservative than required.

Vith a radioactive liquid effluent monitoring instrumentation channel l

alarm / trip setpoint less conservative than required, without delay w m M the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is l

acceptably conservative.

With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the actions described in Table 2-1.

Exert best efforts to return the instruments to GPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Effluent and Vaste Disposal Report (Section 7.2) why the inoperability was not corrected in a timely manner.

Davis-Besse ODCM 2

Revision 4, 1991

2.1.1 Technical Specification Requirement _

-This section prescibes the monitoring required during liquid releases in (s_s~'h order to comply with TS and the backup sampilng required when monitors are 1

inoperable.: The liquid effluent monitoring instrumentation for controlling and monitoring radioactive liquid effluents in accordance with Davis-Besse TS is summarised below.

a)

Alarm (and Automatic Termination) 1.

Clean Radvaste Effluent Monitors (RE-1770 A & B).

Discharges from the Clean Radvaste Monitor Tanks (2) are monitored by redundant radiation monitoring systems (RE-1770 A &

B).

These monitors detect gross gamma activity in.the effluent prior to mixing in the Collection Box. Measurements from each detector read out on the Victoreen panel in the Control Room.

Each monitoring system is capable of initiating an alarm and an automatic isolation of the release by closing valve VC-1771. The method for determining setpoints for the High Alarm, which initiates isolation, is discussed in ODCM Section 2.3.

11. Miscellaneous Radvaste Effluent Monitors (RE-1878 A & B).

Discharges from the Miscellaneous Liquid Vaste Monitor Tank and the Detergent Vaste Drain Tank are monitored by redundant radiation monitoring systems (RE-1878 A & B).

These monitors detect gross gamma activity in the affluent line prior to mixing in the Collection Box. Measurements from each detector read out on the Victoteen panel in the Control Room. -Each monitor is-separately-capable of initiating an alarm and automatic isolation of the release by closing valve VM-1876. Setpoint determination for the High Alarm, which initiates isolation, is discussed in I

ODCM Section 2.3.1.

b)

Alarm'(only)-

1.

Turbine Building Sump Effluent Line (RE 4686).

The purpose of the monitor on-the Turbine Butiding sump effluent line is to detect abnormal radionuclide concentrations in the sump effluent prior to discharge to the onsite: basin Training Center Pond. This monitor is located near the end of the storm sever drain pipe, upstream of_the final discharge point into the Training Center Pond. This stream is commonly referred to as the Turbine Building Sump / Storm Sever Drain (TBS /SSD). The source of any radioactive material =in the sump would be from the secondary steam system. Therefore, activity is expected in the turbine buildsng sump effluent-system if a primary-to-secondary leak has occured. If a primary-to-secondary leak is present, the activity-in the sump effluent system would be ecmprised of those radionuclides found in the secondary system. Evaluation of alarm setpoint for RE-4686 is discussed in ODCM Section 2.3.2.

()

Davis-Besse ODCM 3

Revision 4, 1991

c)

Flow Rate Measuring Devices In order to comply with TS, the release rate of liquid radvante discharges shall be monitored. The following flov indicators and totalizers meet this requirement 1.

Clean Radvaste Effluent Line Flov Indicator (F1) 1700 A & B Flow Totalizer (F01) 1700 A & B

11. Miscellaneous Radvaste Effluent Line Flow Indicator (FI) 1887 A & B Flov Totalizer (FOI) 1887 A & B 111. Dilution Flov to the Collection Boy Computer Foint F201 2.1.2 Non-Tee.hnical Specification Monitors Additional monitors, although not required to satisfy the requirements of the Davis-Besse TS, have been installed to control liquid radioactive The monitors material and reduce the likelihood of unmonitored releases.

are Collection Box outlet to the Lake (RE-84s3) - monitors the final station effluent to the lake.

Component Cooling Vater System (CCVS) (RE-1412 & 1413)-

monitors the CCVS return line. High alarm closes the atmospheric vent valves on the CCVS surge tank.

Service Vater System (SVS) (RE-8432) - single off-line detector monitoring the SVS outlet prior to discharge to the Collection Box.

Intake Forebay (RE-8434) - single detector continuously monitors the station intake water from Lake Erie.

2.2 Sampling and Analysis of Liquid Effluents The program for sampling and analysis of liquid vaste is prescribed in this Section and incorporates the basic requirements outlined in TS.

Radioactive liquid vastes shall be sampled and analysed according to the sampling and analysis program of Table 2-3.

Table 2-3 identifies three potential sources of liquid radioactive effluents for which sampilng and analysis are required to ensure releases are controlled in accordance "l'h the TS limits.

The results of the radioactivity analyses shall be used in accordance with the methodology and parameters of this Section of the ODCH to assure that the concentrations at the point of release are maintained within the following limits.

The concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR Part 20.106 for radionuclides other than dissolved or entrained noble For dissolved or entrained noble gases, the concentration shall be gases.

4 Revision 4, 1991 Davis-Besse ODCH

limited to 2.0 E-04 UC1/ml. -If the concentration of radioactive material pa released in liquid effluents to UNRESTRICTED AREAS exceeding these limits, then without dslay restore the concentrations to within these limits.

t-d The sources of radioactive effluents and the associated sampling and analysis requirements are discussed below.

2.2.1 Batch Vaste Release Tanks BATCH RELEASES are defined as the discharge of liquid vastes of a discrete volume. The releases from the Clean Vaste Monitor Tanks 1-1 an6 1-2, the Miscellaneous Liquid Vaste Monitor Tank, and the Detergent Vasta Drain Tank are classified as BATCH RELEASES. The following sampling and analysis requirements must be met for all releases from these tanks.

Prior to each BATCH RELEASE, analysis of a representative grab sample for principal gamma emitters (including I-131 and other peaks identified by gamma spectroscopy).

Once per month, analysis of one sample from a BATCH RELEASE for dissolved and entrained gases (gamma emitters).

(See note belov.)

Once per month, analysis of a COMPOSITE SAMPLE of all releases that month for tritium and gross alpha activity. The COMPOSITE SAMPLE is required to be representative of the liquids released. Samples contributed to the composite are to be proportional to the quantity of liquid discharged.

Once per quarter, analysis of a COMPOSITE SAMPLE of all releases that

.-A.

quarter for Strontium (Sr}-89, Sr-90, and Iron (Fe)-55.

()

NOTE: Identification of noble gases that are principal gamma-emitting radionuclides are included as a part of the gamma spectral-analysis performed on all liquid radvaste effluents.

Therefore, the Table 2-3 requirement for sampling and analysis-of one batch per month for noble gases need not be performed as a separate program. The gamma spectral analysis on each BATCH RELEASE meets the intent of-this requirement.

2.2.2 Turbine Building Sump / Storm Sever Drain (TBS /SSD)

Releases *'from the TBS /SSD are classified as continuous releases, since these discharges are not controlled on a batch basis.- Table 2-3 requires that a sample shall be collected from the TBS /SSD, if the on-line monitor is out-of-service and the activity level'of the condensate (i.e., hot vell vater) exceeds 1.0 E-05-uci/ mis gross beta / gamma. During this period.)

sample is to be' collected once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and analyzed for principal gamma emitters.

As a back-up to the on-line monitors, grab samples are periodically collected (nominally once per week) from the TBS /SSD and analyzed by gamma spectroscopy. If activity is identified, additional controls are enacted to ensure that the release concentrations are maintained belov HPC as L

required by TS and that the cumulative releases are a small fraction of the dose limit of TS. The following actions vill be considered for controlling any radioactive material releases via the TBS /SSD:

{&

T Davis-Besse ODCM 5

Revision 4, 1991 i

L

~

t As needed for controlling and quantifying releases, the sampling frequency of the TBS /SSD vill be increased to every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months untti the source of the contamination is found and controlled.

Gamma spectral analysis vill be performed on each sample for principal gamma emittera.

The measured radionuclide concentrations from the gamma spectral analysis vill be compared with MPC (equation 2-2) to ensure releases are within the limits.

Based on the measured concentrations, a re-evaluation of the alarm setpoint for the TBS /SSD monitor (RE-4686) vill be performed as specified in ODCH Section 2.3.2.

Each sample vill be considered representative of the releases that have occurred since the previous sample. The volume of liquid 1

released vill be determined based on the Turbine Building Sump pump runtimes and flows.

From the sarple analysis and the calculated volume released, the total radioactive material released vill be determined and considered j

representative of the release period, cumulative doses vill be I

determined in accordance with ODCM Section 2.5.

Discharges from the TBS /SSD are routed to the Training Center Pond with the pond overflov discharging to the Toussaint River. For conservatism, it is assumed that any radioactive material releases from the TBS /SSD to the Training Center Pond are ultimately discharged to the lake environment (unless actions are taken to prevent occurrence).

2.2.3 Condensate Demineralizer Backvash Discharges fror. the Condensate Demineralizer Backvash Receiving Tank (BRT) to the South Settling Basin are controlled as BATCH RELEASES in accordance with Table 2-3.

Samples are collected prior to each release of the resin / vater slurry and separated into the liquid phase (transfer vater) and solid phase (resin). These samples are separately analyzed for principal gamma emitters. Toledo Edison has imposed guidelines on concentrations of radionuclides that may be discharged to the onsite settling basin. These guidelines are presented in Table 2-4.

O Davis-Besse ODCH 6

Revision 4, 1991

D The radioactive material contamination in the condtnsate demineraliser backvash vill be contained on the povdered resing soluble or suspended p

radioactive material associated with the water phase is not expected.

Q However, the resin and the water cre analyzed separately thus allowing for a determination of the amounts retained onsite in the settling basin (i.e.,

the resin) and the amounts released to Lake Erie as an effluent (i.e., the water that is ultimately released to the lake as the decant from the basin).

The BRT receives the spent resin from the Condensate Clean-up System.

Low-level radioactive material contamination of the spent resin is periodically expected due primarily to minor veeps in the steam generators and the leaching of residual activity in the secondary system that was deposited from the steam generator tube leaks.

During primary-to-secondary leakage, activity levels vill be elevated and typically above the limits imposed for acceptable discharge to the basin.

Under these conditions, the powdered resins are retained within the plant and processed as solid radvaste for offsite transport and disposal at a licensed radiositive vaste disposal site. If within the criterion of Table 2-4, the BRT may be discharged to the onsite settling basin.

2.3 Liquid Effluent Monitor Setpoints Technical Specifications require that i.he concentration of radioactive materials released in liquid radioactive effluents from the site to UNRESTRICTED AREAS shall not exceed the UNRESTRICTED AREA HPC at the discharge point to Lake Erie. This limitation provides additional assurance that the levels of radioactive material in bodies of water outside the site should not result in exposures exceeding:

The Section II.A design objective of Appendix I, 10 CFR Part 50, to an individual, and the limits of 10 CFR Part 20.106(e) to the population.

Dissolved or entrained noble gases in liquid effluents are limited to a concentration of 2.0 E-04 uC1/ml, total activity. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its HPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

Radiation monitor setpoints shall be established to alarm and trip prior to exceeding the limits specified above. To meet this requirement, the alarm / trip setpoint for liquid effluent monitors are determined in accordance with the following equation:

CL (DF+RR)

SP<

(2-1)

RR Davis-Besse ODCH 7

Revision 4, 1991

Where the effluent concentration limit implementing 10 CFR Part CL

=

20.106 (i.e., HPC at discharge point) in uC1/ml, defined in equation (2-4).

the setpoint, in uCi/ml, of the monitor measuring the SP

=

radioactivity concentration in the efiluent line prior to dilution. The setpoint represents a value which, if exceeded, vould result in concentrations exceeding the MPC in the UNRESTRICTED AREA.

the liquid effluent release rate as measured at the radiation RR

=

monitor location, in volume per unit time, but in the same units as DF, below.

the dilution vater flov as measured prior to the release point in DF

=

volume per unit time.

At Davis-Besse a minimum required dilution vater flov is established for a given release, and the vaste tank release rate (RR) and monitor setpoint (SP) are set to meet the condition of equation 2-1 for a given effluent concentration limit, CL.

NOTE: If no dilution is provided SP $ CL, Also, when DF is large compared to RR, then (DF + RR) e DF.

2.3.1 Liquid Radvaste Effluent Line Monitor (RE-1770 A & B, RE-1878 A & B)

The Liq..id Radvaste Effluent Line Honitors provide alarm and automatic termination of releases prior to exceeding HPC. As required by Table 2-3 and as discussed in ODCM Section 2.2.1, a sample of the liquid radvaste to be discharged is collected and analyzed by gamma spectroscopy to identify principel gamma emitting radionuclides. From the measured individual radionuclide concentrations, the required dilution flov and the allovable release rate are determined.

The dilution flov and allovable release rate are inversely proportional to the ratio of the radionuclide concentrations to their MPC values. This ratio of measured concentration to HPC values is referred to as the "HPC f raction" and is calculated by the equation' MPCT = E (2-2) i MPC, Davis-Besse CDCM 8

Revision 4, 1991

where MPCT =

f raction of the unrestricted atea MPC for a mixture of O

radionuclides b

C' concentration of each radionuclide (i) measured in tank prior to

=

release (vCi/ml)

MPC* =

unrestricted area MPC for each radionuclide (1) from 10 CrR Part 20, Appendix B, Table II, Column 2.

For dissolved and entrained noble gases an MPC value of 2.0E-04 vci/ml shall be used.

As expressed in equation (2-1), the concentration limit (CL) of a liquid radwaste discharge is the same as the effective MPC for the radionuclide mixture of the discharge. Simply, the CL (or effective MPC) represente the equivalent MPC value for a mixture of radionuclides evaluated collectively.

W e equation for determining CL ist CL =

(2-3)

MPCT Based on the MPCF, the mininum dilution factor (DF) for the conduct of the release is established at 3.33 times larger than actually required. This safety factor (SF) provides conservatism, accounting for variations in monitor response and flow rates and also for the presence of radionuclides that may not be detected by the monitors (i.e., non-gar.ma emitters). The following equation is used for calculating the required minicum dilution factort Dr = MPCr/Sr (2-4) pd where:

minimum required dilution factor Dr

=

0.3 administrative safety factor Sr

=

t e allowable release rate is then calculated by dividing the available dilution flow (ADr) at the Collection Box by Dr as calculated by equation (2-4).

ADF/Dr' (2-5)

MAX RR

=

where:

maximum allowable release rate (gal / min)

MAX RR

=

available dilution flow at the Collection Box as measured by ADr

=

Computer Point r 201 (gal / min) tCIT:

Equations (2-3) and (2-4) are valid only for MPCT >1; for MPCT

<l, the waste tank concentration meets the limits of 10 CFR Part 10 without dilution, and MAX RR may take on any desired value.

Davis-Besse ODCM 9

Revision 4, 1991

If MAX RR as calculated above is greater than the maximum discharge pump capacity, the pump capacity should be used in establishing the actual release rate RR for the radwaste discharge. For releases from the Miscellansous Waste Monitor Tank arv3 Detergent Waste Drain Tank, the discharge pump capacity is 100 gpn (i.e., the design limit for MAX RR is 100 gpn); for the Clean Radwaste Tank, this value is 140 gpn.

Based on the calculated release rate (RR), the dilution factor (DF) and the concentration limit (CL), the alarm setpoint is calculated as prescribed in equation (2-1) by the equation:

SP =

- + Bkg (2-6)

MPCT

RR where setpoint of the radiation monitor (counts per second - cps)

SP

=

C' concentration of radionuclide (1) as measured by gansna

=

I spectroscopy (uci/ml) monitor sensitivity for radionuclide (1) based on calibration SIN

=

curve (eps/(uci/ml))

the required minimum dilution factor as defined in equation (2-4)

Dr

=

(gal / min) actual release rate of the liquid radwaste discharge (gal / min)

RR

=

MPCF =

MPC fraction as determined by equation (2-2) background reading of monitor (cps)

Bkg

=

The Cs-137 sensitivity may be used in lieu of the sensitivity values for individual radionuclides. The Cs-137 sensitivity provides a reasonably conservative nonitor response correlation for radionuclides of interest in reactor effluents. Coupled with the safety factor Sr in equation (2-3),

this assumption simplifies the evaluation without invalidating the overall conservatism of the setpoint determination.

Prior to conducting any batch liquid radwaste release, equations (2-4) and (2-5) are used to determine the minimum required dilution flow and the allowable release rate. Equation (2-6) is then applied to determine the RE-1770 A & B or RE-1878 A & B alarm setpoints.

i Davis-Besse ODCM 10 Revision 4, 1991

2.3.2 'Nrbine Building SunsVStorm Sewer Drain Monitor' (RE-4686)

[]f We setpoint for the MS/SSD radiation monitor RE-4686 shall be established

(

to ensure the radioactive material concentration in the effluent prior to discharge offsite does not exceed hPC, WRESTRICTED AREA (10 CrR 20, Appendix B, Table II, Column 2).

We M S/SSD is not normally radioactively contaminated. Therefore, the setpoint for this monitor has been established at its lowest practical level (i.e., two times the normal background) in order to provide an early indication of any abnormal conditions. If radioactivity is found in this system, then a setpoint may be determined by using the awasured radioactive material concentration from the grab sample coupled with the algorithm of equation (2-8),

mis approach for determining the alarm setpoint is the same as presented in Section 2.3.1 for the Liquid Radweste Effluent Line Monitors. Equation (2-1) remains valid, except that, for the %S/SSD line monitor, the dilution flow previously assumed for diluting the batch liquid radweste effluents is now the release rate as determined from the Turbine Building sung flow. There is no additional dilution prior to discharge to the Training Center Pond. Sus, Equation (2-1) slaplifies to; SP f CL (2-7)

Also, since discharge is to the Training Center Pond, exceeding a setpoint does not necessarily mean cxceeded T5 limit on release concentration to the lake. - Se verification of compliance with the TS limits on concentration should be based on actual samples of the effluent from the pond to the lake environment. (Refer to CDCM Section 2.3.4).

Substituting equation (2-3) for CL, the alarm setpoint can be calculated by the equation:

I (C,

S m,).

(2-8)

SP <- MPCF where -

C

concentration of each radionuclide (i) in the us/SSD effluent

=

(uciAnl)

MPCF =

- MPC fraction as determined by equation (2-2)

Sm' =

monitor sensitivity for radionuclide (1) based on calibration curve (cpavuci/ml)

Again, the Cs-137 sensitivity may be used in lieu of the individual radionuclide evaluation as discussed for equation (2-6).

t Davis-Besse 00CM 11 Revision 4, 1991

2.3.3 Alarm Setpoints for the Non-Technical Specification Radiation Monitors 1.

Collection Box outlet to the Lake (RE-8433). As discussed in ODCM Section 2.1.2, the Radiation Monitor on Collection Box outlet utilizes a single off-line detector to continuously monitor all station liquid effluent discharges to the lake. Although this is the final effluent monitor, it does not serve any control function. Control functions have been placed on the upstream undiluted effluent line that will terminate the release prior to exceeding the MPC for UtGESTRICTED AREAS. This monitor therefore provides the Control Room operator with a final check of the total diluted effluent stream. Since this monitor views the diluted radwaste discharges, its response during routine operations will be minimal (i.e., typical of background levels). Werefore, the alarm setpoint for this monitor should be established as close to background as possible without incurring a spurious alarm due to background fluctuations. The alarm is established in accordance with the Radiation Monitor Setpoint Manual.

11. Component Cooling Water System (CCWS) (RE-1412 & 1413). The monitors RE-1412 and 1413 provide indication of a breach in the CCWS integrity, allowing primary system water to enter and contaminate the system.

Therefore, the alarm setpoint is established as close to background as possible without incurring a spurious alarm due to background fluctuations. The alarm is established in accordance with the Radiation Monitor Setpoint Manual.

iii. Service Water System (SWS) (RE-8432). No radioactive material is expected to be contained within the SWS during normal operations.

Werefore, the alarm setpoint is established as close to background as possible without incurring a spurious alarm due to background fluctuations. The alarm is established in accordance with the Radiation Monitor Setpoint Manual, iv. Intake Forebay Monitor (RE-8434). The alarm setpoint for this monitor should be established as close to background as possible without incurring a spurious alarm due to background flue;.uations. Although a very remote potential, a verified alarm from this system would indicate a possible contamination of the station intake water. The alarm is established in accordance with the Radiation Monitor Setpoint Manual.

2.3.4 Alarm Response - Evaluating Actual Release Conditions Liquid release rates are controlled and alarm setpoints are established to ensure that releases do not exceed the concentration limits of Section 2.3 (i.e., 10 CTR 20 MPC's at the discharge to Lake Erie). However, if any of the monitors (RE-1770 A & B, RE-1878 A & B, or RE-4686) alarm during a i

liquid release, it becomes necessary to re-evaluate the release conditions to determine compliance with the limits, rollowing an alarm, the actual release conditions should be determined. Radioactive material Davis-Besse ODCM 12 Revision 4, 1991

concentrations should be evaluated by sampling the offluent stream (or resampling the vaste tank). Dischg ae flow and dilution water flow should be redetermined. The following actions siw13 be considered:

verify radiation monitor alarm setpoint, ensure consistency with the setpoint evaluation for the release.

re-sample and re-analyze the source of the release (e.g., release tank, TB sump, decant from Training Center Pond to the Toussaint River).

re-define the release conditions on the release rate and the dilutica water flow.

Based on these data, the following equation may be used for evaluating the actual release conditions:

E

<1 (2-9)

~

MPC, DF + RR where measured concentration of radionuclide (1) in the effluent stream C

=

(vCi/ml)

MPC, =

the MPC value for radionuclide (1) from 10 CrR 20, Appendix B, Table II, Column 2 (aci/ml) 2.0E-04 pC1/ml for dissolved or entrained noble gases

=

actual release rate of the liquid effluent at the time of the RR

=

alarm actual dilution water flow at the time of the release alarm Dr 2.4 Liquid Effluent Dose calculation - 10 CrR 50 W e parameters of the liquid release (or estimated parameters, for a pre-release calculation) may be used to calculate the hypothetical dose to a MEMBER Or DIE PUBLIC from the release (or planned release). We dose calculation provides a conservative method for estimating the impact of radioactive effluents released by Davis-Besse and for cesparing the impact against limits set by the NRC in the Davis-Desse TS.

te limits are specified below as quarterly and calendar year limits.

2.4.1 MEMBER OF THE PUBLIC Dose - Liquid Effluents This requirement is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix I, 10 CTR Part 50.

Davis-Besse CDCM 13 Revision 4, 1991

Technical Specification limits the dose or dose ccomitsent to MEMBERS OF

'nIE PUBLIC f rom radioactive materials in liquid ef fluents from Davis-Besse.

ne limits are:

during any calendar quarter:

3 1.5 mrem to total body 3 5.0 mrem to any organ during any calendar years 3 3.0 mrem to total body 3 10.0 mrem to any organ Wese limits implement the guides set forth in Section II.A of Appendix I, 10 CrR, Part 50.

With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Comission within 30 days, pursuant to Section 7.3, a special Peport that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. Wis action provides the required operating flexibility and at the same time implements the guides set forth in Section IV.A of Appendix I, 10 CTR Part 50 to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable."

NOTE:

Tor fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operatiori of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CTR 141. The dose calculations in the 00CM ireplement the requirements of Section III.A of Appendix I of 10 CTR Part 50 that conformance with the guides of Appendix I is to be shown by calculational procedures based on modes and data such that the actual exposure of an individual thorough appropriate pathways is unlikely to be substantially underestimated, ne equations specified in the 00CM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CTR Part 50, Appendix I," Revision 1, october 1977.

l l

TS requires that cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

l l

t Davis-Besse ooCM 14 Revision 4, 1991

We calculation of the potential dor,es to MDEERS OF UtB PUBLIC is a function of the radioactive material releases to the lake, the subsequent y

(

transport and dilution in the exposure pathways, and the resultant individual uptake. At Davis-Besse, the combined fish consunption and drinking water pathway has been modeled to provide a conservative dose assessment for exposures to MDEERS Ol' DIE FUBt!C. For the fish pathway, it has been conservatively assumed that the maximam exposed individual consumes 21 kg per year of fish taken in the innediate vicinity of the Davis-Besse discharge to the lake. For the drinking water pathway, the conservative modeling is based on an individual drinking 730 liters per year of water f rom the beach wells located 966 m to the tM of the site discharge.

(It is inportant to note that because of the high sulfur content, the water from these beach wells is not suitable for consumptions however, for conservatiam this pathway has been included in the dose modeling for the maximum exposed individual.)

The equation for assassing the maximam potential dose to MDmERS Or W E PUBLIC from liquid radweste releases from Davis-Besse is 1.67E-02

YoL

E (C'

A)

(2-10)

D

=

Dr

z where D,

dose or dose connitment to organ (o) including total body (mrem)

=

A,*

site-specific ingestion dose connitment factor to the total body or any organ (o) for.radtonuclide (1) (mrenVhr per vCiAnl)

C, average concentration of radionuclide (i) in undiluted liquid

=

sffluent representative of the the volume VOL (vCiAni) total volume of liquid etfluent released (gal) voL

=

average dilution water flow during release period (galanin)

Dr

=

(minima value is typically 20,000 gpa) 10, near field dilution factor

E

=

1 hr/60 min 1.67E-02

=

represent a n e site-specific ingestion dose / dose connitment factors (Acompositedosefac site-specific dose factor is based on the NRC's generic maximum indl # mi consumption rates. Values of A are presented in Table 2-5.

Wese values were derived in accordance with'I.he guidance of NUREG-0133 using the following equation:

(2-11)

A,, = 1.14E+05 (U, / D, + U,

Br,) Dr, OV Davis-Besse 00CM 15 Revision 4, 1991

I where 21 kg/yr adult fish consumption U,

=

730 liters /yr adult water consumption it,

=

5.7, additional dilution from the near fiold to the beach wells D"

=

(net dilution of 57*)

bioaccumulation factor for radionuclide (1) in fish from Table Br'

=

2-6 (pciAg per aci/1)

Dr' dose conversion factor for nuclide (1) for adults in organ (o) f rom Table E-11 of Regulatory Guide 1.109 (mrenVpC1) 10' (oci/uci)

10' (m1Ag) / 9760 (hr/yr) 1.14E+05 n e radionuclides included in the periodie dose assessment required by TS are those identified by ganna spectral analysis of the liquid waste samples collected and analyzed per the requirements of Table 2-3.

In keeping with the IUPIG-0133 guidance, the adult age group represents the maximum exposed individual age group. Evaluation of doses for other age groups is not required for denonstrating compliance with the done criteria of TS. We dose analysis for radionuclider requiring radiochemical analysis will be perforned af ter receipt of results of the analysis of the conposite sanples. In keeping with the required analytical frequencies of Table 2-3, tritiusi dose analyses will be performed at least sonthly; St-09, St-90 and Fe-55 dose analyses will be performed at least quarterly.

Near field dilution factor and dilution to beach wells are based on a study perforRed by Stone & Webster for Toledo Edison entitled " Aquatic Dilution ractors within 50 Miles of the Davis-Besse Unit i Nucle 4r Power Plant", June 1980.

2.4.2 Sinclified Liquid Effluent Dose calculation In lieu of the individual radionuclide dose assessment presented in Section 2.4.1, the following sinplified dose calculation nay be used for demonstrating coupliance with the dose limits required by TS.

Radionuclides included in this dose calculation should be those measured in the grab sample of the release (principal ganea emittern measured by ganna ripectroscopy). H-3 should not be included in this analysis. Refer to Appendix A for the derivation of this staplified method.

Total uody 9.70E+02

vot

I C, (2-12)

D

=

g Maximum organ 1.19E+03

VOL

I C, (2-13)

D

=

Davis-Besse ODCM 16 Revision 4, 1991

vhere h

C' averare concentration of radionuclide (i) excluding H-3 in

=

V undiluted liquid effluent representative of the volume voL (vC1/ml) volume of liquid effluent released (gal)

VOL

=

average dilution water flow during release period (gal / min)

Dr

=

conservatively evaluated total body dose (mrem)

D

=

g conservatively evaluated maxiren organ dose (mrem)

D

=

9.70t+02 = 0.0167 (hrAnin)

5.81t+05 (mrevhr per act/ml, Cs-134 total body dose factor from Table 2-5) / 10 (near fiald dilution) 1.19t+03 - 0.0167 (hr/ min)

7.11t+05 (mrem /hr per vCi/mi, cs-134 liver dose factor from Table 2-5) / 10 (near field dilution) 2.4.3 Contaminated TBS /SSD System - Dose Calculatio9 If the TBS /SSD system becomes contaminated, any radioactive material releases amist be included in the evaluation of the cumulative dose to a MEMBER Or 2E PUBLIC as required by ODCM Section 2.2.2.

Section 2.2.2 describes the methods for quantifying and controlling releases from the TBS /SSD system.

Although the discharges are via the Training Center Pond to the Toussaint

/

River (instead of directly to Lake trie), the modeling of equation (2-10)

C]

remains reasonably conservative for determining a hypothetical maximum individual dose. The following assunption should be applied for the dose assessment of any radioactive material releases from the as/SSD into the Training Center Pond and subsequently to the Toussaint Rivers If no additional controls are taken, it should be assumed that any radioactive material releated to the Training Centar Pond will ultimately be discharged to the lake environment.

If actions are taken to limit any release, the assessment of dose should be made based on an evaluation of actual releases.

% e dilution flow (DF) should consider any additional dilution of the nS/SSD discharge from other sources into the Training Center Pond prior to release to the river.

2.5 Liquid Effluent Dose Projections 10 CrR 50.36a requires licensees to maintain and operate the radwaste system to ensure releases are maintained ALAPA. This Section implements the requirements of 10 CrR Part 50.360, General Design criterion 60 of Appendix A to 10 CrR Part 50 and design objective Section II.D of Appendix I to 10 CrR Part 50. Based on a cost analysis of treating liquid radwaste, the specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as the dose design objectives as set forth in Section II.A of Appendix I,10 CrR Part 50, for liquid effluents. This requirement is implemented through this ODCM.

Davis-Besse ODCM 17 Revision 4, 1991

We liquid radioactive waste processing system shall be used to reduce the radioactive material levels in the liquid waste prior to release when the projected doses in any 31 day period would exceed:

0.06 mrem to the total body, or 0.20 mre: to any organ.

When the projected doses exceed either of the above limits, the waste must be p.rocessed by the liquid radweste system prior to ralease. 21s dose criteria for processing is established at one quarter (1/4) of the design objective rate (i.e., 1/4 of 3 mrem /yr total body and 10 mre v yr any organ over a 31 day projection).

With radioactive liquid waste being discharged without treatment and in excess of the above limits, in iteu of a Licensee Event Report, prepare and submit to the Consnission within 30 days, pursuant to Section 7.3, a Special Report that includes the following information:

Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equirnent or subsystems, and the reason for the inoperability, Action (s) taken to restore the inoperable equipment to OPERABLE status, and Jumary description of action (s) taken to prevent a recurrence.

W e applicable liquid waste processing system for msintaining radioactive material releases ALARA is the ion exchange system as delineated in rigure 2-1.

TS requires that in any month in which radioactive liquid effluent is being discharged without treatment, doses due to liquid releases to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the 00CM.

We projection of doses is made to evaluate the need for additional radwaste processing to ensure future releases are maintained AIARA. nese projections are required only if the radwaste system has not been used (e.g., demineralizer system bypassed or resin bed exhausted). We following equations may be used for the dose projection calculation:

0

-D (31 / d)

(2-14)

(2-15)

D,,

= D,,, (31 / d) where:

the total body dose projection for current 31 day period (mrem)

D

=

the cumulative total body dose to date for current calendar D

=

quarter including release under consideration as determined by g

equation (2-10) or (2-12) (arem)

Davis-Besse coCM 18 Revision 4, 1991 l

D,

,=

the maximum organ dose projection for current 31 day period (arem)

, b the maximum organ dose to date for current calerdar quarter D

=

including release under consideration as determined by equation (2-10) or (2-13) (mrem) the number of days to date in current calendar quartar d

=

31 the number of days in projection

=

O O'

Davis-Besse ODCM 19 Revision 4, 1991

Table 2-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITT ACTION 1.

Cross Radioactivity Monitors Providing Alarms and Automatic Termination of Release

a.. Liquid Radvaste Effluent Line 1

(1)

A (either Miscellaneous or Clean, but not both simultaneously) 2.

Flow Rate Measurement Devices a.

Liquid Radvaste Effluent Line 1

(1)

B b.

Dilution Flow to Collection Box 1

(1)

B 3.

Gross Beta or Gamma Radioactivity Monitors Providing Alara But Not Providing Automatic Termination of Release a.

Turbine Building / Storm Sever Drain 1

(1)

B,C Davis-Besse 20 Revision 4, 1991 O

O O.

TABLE 2-1 (continued)

TABLE NOTATION (1) During radioactive releases via this pathway ACTION A Vith the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases may be resumed, provided that prior to initiating a release:

1 1.

At least two independent samples are analyzed in accordance with Table 2-3 for analyses performed with each batch 2.

At least two independent verification of the release rate calculations are performed 3.

At least two independent verifications of the discharge valving are performedi

(

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION B Vith the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump curves may be used to estimate flow.

ACTION C Vith the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this i

/

pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , O

grab-samples are collected and analysed for gross radioactivity (beta or gamma) at a lover limit of detection no greater than i

1.0E-07 uct/al.

Davis-Besse ODCM 21 Revision 4,-1991 m

~

TABLE 2-2 RADIOACTIVE LIQUID EFFUTENT MONITORING INSTRUMENTATION SURVEILLA. 2 REQUIREMENTS W

CELMNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST 1.

Gross i. eta or Camma Radioacti.ity Monitors Providing Alars and Automatic Isolation a.

Liquid Radvaste Effluents Line D

F R' ' '

Qvae 2.

Flow Rate Monitors a.

Liquid Radwaste Effluent Line D' * '

N.A.

R Q

b.

Dilution Flow to Collection Box D' *

N.A.

R Q

Davis-Besse ODCM 22 Revision 4, 1991 9

9 9

E l

t TAatz 2-2 (continued)

TABLE NCfrATION (1) During releases via this pathway.

(2) The OIAleNL FUNCTIONhL TEST shall also demonstrate thet automatic 1 solation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the alarsvtrip setpoint.

_ (3) '!he initial owe 4EL CALIBRATION for radioactivity measurement instrumentation shall be performed using one or more of the reference i

standards certified by the National Institute of Standards and Technology or using standards that have been obtained from suppliers that participate-in measurement assurance activities with NIST. These standards should permit calibrating t.he system over its intended range of energy and rate capabilities. For subsequent CHAl@iEL CALIBRATICN, sources that have been related to the initial calibration should be used, at intervals of at least once per eighteen sonths. For high range monitoring instrumentation, where calibration with a radioactive source is impractical, an electronic calibration may be substituted for the radiation source calibration.

i (4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. QW4EL CHECK shall be made at least once daily on any day on which' continuous, periodic, or RA'!Of RELEA8E8 are made.

(D) At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G (P) Prior'to each release.

D

- (R) At least once per 18 month (550 dcys).

(Q) At least once per 92 days.

i l-r L

Davis-Besse CDCM 23 Revision 4, 1991

(

,n

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

a'_.,

.,, - - - - +,,,,..,, -..

...,,..,,-,.-..,,n..~...

.T](@LE 2-3 RADIOACTIVE LIQUID R$TE SAMPLitJ3 AfJD Af1ALYSIS PPOGRAM Minimum type of Lower Limit Liquid Release 'rype Sampling Analysis Activity of Detection Frequency rrequency Analysis (LLD) (vCi/mi)'

P P

Principal A.

Batch Waste Each Batch Each Batch Gamma 5.0E-07*

Release Tanks

Emitters' I-131' 1.0E-06 P

Dissolved One natchai M

and Entrained 1.0E-05 Gases l

P M

Each Batch Conposite' 11-3 1.0E-05 Gross Alpha 1.0E-07 i

P Q

l Each Batch Conposite' sr-89, Sr-90 5.0E-08 te-55 1.0E-06 B.

Turbine Building Principal Sunp/ Storm Continuous S*

Gamma 5.0E-07*

Sever Drain Emitters'

~

1-131f 1.0E-06 P

P Principal C.

Condensate Each Batch Each Batch Gamma 5.0E-07" Demineralizer Emitters' Backwash J.-131' 1.0E-06 Davis-Besse ODCM 24 Revision 4, 1991 0

Table 2-3 (continued)

+

O

~ - "

a.

'me 11D is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

11D =

4.66 s, l

t

v

2.22

Y

exp (-Aot) where LLD is the lower limit of detection as defined above (as pC1 per unit I

m as or volume);

S is the standard deviation of the backgrrund counting rate or of the cbunting eato of a blank sample as appre;st* ate (as counts per minute);

E is the counting efficiency (as counts per transformation):

V is the sample size (in units of recs or volume):

2.22 is the number of transformations per minute per picoeurie; i

Y is the fractional radiochemical yield (when' applicable)#

A is the radioactive decay constant for the particular radionuclide;

.At for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

It should be recognised that the Ltb is defined as an a g,iori (before the E

fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

1 l

i Davis-Besse ODcM 25 Revision 4, 1991

~

Table 2-3 (continued)

TABLE torATION b.

The principal ganna emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, re-59, co-50, co-60, In-65, Mo-99, Cs-134, Cs-137, and Ce-141. For Ce-144, the LLD is 2.0E-06 vCi/ml.

Other peaks which are measured and identified shall also be reported.

Nuclides which are below the LLD for the analysis should not be reported as being present at the LLD level. When unusual circumstances result in LLDs higher than required, the reasons shall be documented tri the semiannual

)

Effluent and Waste Disposal Peport.

c.

A COMPOSITE SAMPLE is one in which the method of sanpling employed results is a specisen which is representative of the liquids released.

d.

A nA101 RELEASE is the discharge of liquid wastes of a discrete volume, e.

When the monitor is out of service, a grab sample shall be taken and analyzed once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if the condensate pump discharge exceeds 1.0E-05 vCi/mi gross beta or gaona.

f.

If an isotopic analysis is unavailable, gross beta or gansna measurement of RA101 RELEASE may be substituted provided the concentration released to the UNRESTRICTED AREA does not exceed 1.0E-07 pCiAnl and a COMPOSITE SAMPLE is analyzed for principal gasma emitters when instrumentation is available.

9 Frequency notation:

P - Prior to each release.

M - At least once per 31 days.

Q - At least once per 92 days.

S - At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (when the monitor is inoperable).

Davis-Best,e ODCM 26 Revision 4, 1991 0

Table 2-4 Limiting Radionuclide Concentrations

In Fecondary-81de

O Clean-Up Resins for Discharges to Onsite Settling Basin Radionuclide Limitingconcen} ration **

(uCi/cm )

Cr-51 3.3E-02 Mn-54 6.2E-05 re-59 5.1E-04 Co-58 3.0E-04 Co-60 5.4E Y-91 2.1E-03 Er-95 4.1E-04 Nb-95 1.0E-03 Mo-99 3.5E-02 Ru-103 1.0E-03 Ru-106 1.6E-05 Ag-110m 1.6E-05 Te-125m 5.4E Te-127m 1.SE-05 Te-129s 6.2E-05 Te-131m 1.1E-02 Te-132 7.4E-03 1-131 1.1E-04 I-133 3.8E-04 1-135 1.5E-03 Cs-134 1.1E-05 Cs-136 2.6E-03 O.

Cs-137 1.0E-05 Ba-140 1.1E-02 La-140 7.4E-03 Ce-141 5.8E-03 Co-144 4.1E-05 Pr-143 1.9E-02

Concentration limits based on the study,-Disposal of tow-Level Radioactively Contaminated Secondary-Side Clean-up Resins in the On-site Settling Basins at the Davis-Besse Nuclear Power Station, J. Stewart Bland, May 1963. M limits represent a hypothetical maximum individual dose of less than 1 mrom per year due to an inadvertent release to the offsite environment. The allowable releases limits as presented in Table 2 of the above reference report have been reduced by a factor of 10 for added conservatism -

representing a hypothetical dose of less than 0.1 mrem.

- With more than one radionuclide identified in_a resin batch, the evaluation for acceptable discharge to the onsite settling basin shall be based on the-l

" sum of the fractions" rule as follows: Determine for each identified radionuclide the ratio between the sensured concentration and the limiting concentrations the sum of these ratios for all radionuclides should be less than one (1) for discharge to the basin.

i Davis-Besse 00CM 27 Revision 4, 1991 O

Table 2-5 Davis-Besse Site-Specific L.iquid Ingestion Dose Constitment Factors, A,*

(mrenuhr per vCi/ml) gW

~e ase u *.,

t. net. te m..ie al t

lese,

.3

a.a.

a.*****

e.)

0. 00e *e l.164 4 1.564 4 4.tM 4 8.144 4

6. t64 *e 1.164 *4 C.64 3.435*4 6.364 4 6.368*) 6,364 4 43684 6.3M

8 4.36tel neoge 4.33t*8

4. J38 4 4.338*3 4.338*8 4.338*8 e.338*3 4.338*8 p.43 4,394 4 0.64t*4 9.37t*4 0.008*e 0.00t*g 3.get *4 4,44 * $

ct4 8 0.00t*e 0.00t*8 4.348*8 1.058 8 3.998+l 3.168*4 3.308 4 me*64 s.00s*4 4.44E + l 4.640*8 0.088 4 1 338*l 0.00s*4

8. 44

e se** H 0.006 4 4 138 3 4.908 4 0.088 4 8.430e8 0.004 0

8. $18 4 re.nl 4.99t*8 4.438 4 1.138*8 0.088*0 0.000*4

f. 694

8 3.?ts*8 Peelt 8.10sel 3.998 4 9.98tal 0.008 4 8 eesee t.344*

0.648 0 coalt

4. 00t *4 3.385 4 3.98888 0.088 4 8.Desee 0.ses*e 9,9684 cease e.00t 4

1. 00t e l 4.3a0*4 0.eetoe e.ges*e e,cogee 3.03:4 ce*ee 0.Desee 8,4184 4.348*8 8.ge8 4

0. Des 4 0 0es*e 6.40e*1 e443

3. 30e *4 8.394 4 4 118+l

9. 0eg g 3,0eges 3.0stee 4.70s 3 elael 4.344*8 41484 f.99t*4 0.088*0 0.004*4 8.get*6 4.488*8 Ct*64 0.008*0 1.438 4 9.364*4

0. 0e8 4 8.43881 0.8et*9 9.le8*4

&a*68 3.338*4 1 set *4 4.348*4

6. ges 4 4.968*4 0.004.e 4, Ma*4 ga*69 4.998 4 0,4M 4 4.648 *9 0.0e84 6.198 4 4.004 0 4.438*l St*43 0.008*4 8.00 4

3. tes

3 0.608+e 0.00 4 e.ses*0 3.9et*3 6t*63 0.00E*0

0. 00t *0 4.408 4 0.008 4

8. Des ee 0.pos ee

5. 9 88 + l 6t*64 0.008 4 0.008*0 9.448*4 0.084 0.008 4 0.00e *0 4 493*4 St*D6 0.008 4

0. 0H *e 3.108 *0 8.008 4 0.0e8*4 0.0e8*e 0.008*0 e>+e6

8. Det *4 1.044*t e. fil e t 0.088*4 4 tosee 4.000*6 3.ged*e 8e.00 0.008 4 3.964 4 L. 644 4

0. deb *4 0.00s 4 g,ses *e 4.0454 es*e9 0.000*4 8.938 4

8. 3M 4

0. ges 4

0. pas *e g.cogee 6.433*ll steet 3.Ma *e 0.000 4 L 644 + 8 0.000 4 0.000 4 0,0egee 4.3784 State 6.8M4 0.004*4 4 4La*$ 0.ees*e e.cee4 0.ses4 1.39e*4 st*98 4.900*I 0.004*e 4.988*l 0.085*e 8.ge84 0.astee 8.3M4 Statl 1.se8*3

0. 006 *e

4. 044 *4 0.ee8 0 p Desee

3. geed 3 tead tete f. tth i 8.tetoe 4.938*4 0.088 4 0.Det*e 9.eeE*e

f. Steel f at te 4.?? cal 0.004*0 3.638 *4 0.085*e 0.ses*e 3.easee 4,9ge*3 V*98 4.958*1 0.088 e 3.814*l S.ses*e e.ses*e g.ast*4

6. f e8 4 tal 6.398 4 0.0e04 8.tesel 0.sesee e.eee 4 0.ses 4 6.lesel tet t 3.000 4 0.088*0 9.868 8 0.08e *4 0.000*4 e.sen et

4. 3 M 4 treet 6.644*l 3.49tel

1. 4M 4 8.eesee 3.4es*4 0.0esee 4.968*4 statt 3 198*3 f. 6 38

3 3.498*3 0.ses*4 4.g63 3 g,segee 3.3644 m*tt 4.41t*8 3.40t*3 4 848*4 0.tes*e 3.4es*8 0.0e8 *4 4.648 4 utatt 3.fl4*e 9.448*n 3.444*l 40004 4.111 4 0.004*e 3.5084 see=M

8. 0e8 4 L. Md = 3 8.844 4 0.088*0 3.te8*3 8.sesee 3.ete*3 teette 4.398 3

3. 8 M

3 f 498 4 0.08e*4 S.30eal 1.734*3 3 0004 ts*198 4.388 3 8.008*8 4.488*l 0.e88*0 3.338 8 9.eled 5.644 64 tue403 1.134 4 0.000 4 8.Ste 4 4.tes e 3.73g*l 0.000 *4 8.339*8 eweapt 1.944 4 0.088 4

3. M8 a l 8.se84 1.6ftee 8.ges4

3. 6 M

3 auqM l.064*3 0.008 0 1.344*l 8.sesee 3.06t*3 e,Desse 6.eM4 thaltje 0,008 4 0.088 4 0.088*e 0.088*4 0.est*e 0.eesee 0.0e84 thelse 8.008*4 0.904 4

0. e884 0.880 4 8.ses 4 4.est'e 0.088 4 Ae440s 3.338 e 3.9854 4.ftt's 0.0884 8.ses *4

8. esE 4 4.348 4 ab434 4.tedel 4.998 8 4.988*n l.lttal 4.ses 4 3.19e*4 1.364 4 en*l39 3.e484 3.4eE*l 1.348*e 8.000*3 0.ses4 3.30tel 8.395 4 fe* Late

3. 4 68 * $ 9.444*3 1.448*8

1. he

8 4 084 4 0.sesee 4.648+e tean3?e 6.908 4 8.3M4 0.0M*3 4.ser*8 3.678 4 8.ses*e 3.348*4 te431 1.07t*3 3.648 4 4.314 *l f.93 ben 4.Me*3 4.0es *4 0 648*8 fo*189e 1.838*4 4.I?lel 4.?78 4 3.ee8e3 4.67t*4 4 0e8*e 5.634*4 feellt 3.etS4 L. nle e t 1.444 4 3.348 4 1.3es*4 0.sesee 4.3esel feel 3te 1 45*8 0.338 4 6.000 4

1. MS* 8 8.338 4 0.seB+4 4.41t*4 7e.634 4.93 del 9.seB 4 4.008 4 L.8 704

4. ME

4 0.088 4 8.918 4 te433 3.4e8*3 1.488 4 4.ees*3

1. tts* 8 1,las *4 p.ses ee f. He *4 tel30 8.638*8 4.L34*8 4.648*l

9. HS

8 l*tes*8 4.eeE*4

9. 7e54 3*431 3.45*3 3.0 48 *8 4.738 *8 9.888*4 S.498*8 6.ge8 4 f.9 Mal 8433 4.4M4 3.144 4 9.M84

9. MS

4

4. 3tt 4 4.ses 4

8. l M *e tell) 1.415 4 1.364*3 3.estet 1.6as *e 3.18s 4 4.0es*e 4.838*8 8.tes 4.3?t4 8434 8.HA4 5.448 4 S.388 4 3.838*3

4. 3 L4

4 0.005.4 3*lat 3.340*l 8.eeten 3.148*l 8.seS*8 9.MG 4 4 6.635*4 Coelle 3.988 4 f.lleet 8.018 4 4.est*4 3.3eset f 648 4 l.344 *4 Cs4 M

4. l M *4 1.3 M 4 4.088*4 0.088*4

6. ett *4

9. e t44 4 448*4 Ce*L31 8.8M4 5.3 34

5 3.4M*4 0.0e8*e 8.70E*l S.914*4 l.018*4 CaelM

3. Me

8 8.3M*8 3.548*8 0.088 4 1.ee4*3

3. easel 3.3 Mal heal 39 3.8tB*e 16784 4.819 3 0.08B 4 4.948 3 9.eeB+4

4. 360 'e te*let 4.918*8 6.164 4 8.333 4 0.e8s4

3. legal 3.8 38 *l L.0140 te*Lal 1.444 4 0 4 t4*4 3.648 3 S.088*e 0.0es*4 4.Osg 4 4.378*L8 teal 43 5.858 4 8.398 4 3.348 4 4.888*e 4.478 4 8.000 4 f.Jltall Leal 44 L.048*l 9.380 3 8.48B*3 0.0e84 8.gesee 4.segee 4.0e64 Leal 43 9.lM4 4.338*8 1.0e84

0. set *e e esgae 0.oes ee 31484 Ca*l43 4.898 4 8.DeE*l 4.338*3 8.geB+4 8.tes*3 0.000 4 4.4L8*3 Cedel 3.seE*3 3.071 4 8.39E*3 0.008 4 9.lM*3 0 086 4 f.fl4*4 Ce444 8.398 4 3.47 tee 4.458 4 0.00E*e 3.ges4 0,0e44 3.es* 3 Pr*lel 6.098 4 3.fl8+l 1.39t*3 0.0e8 +e 4.stgel 4.008 4 3.004 4 Prel64 3 348 8 9.31t*4 4.848*4 0.es5*e 6.384 4 0.0e8 4 1.138 10 ed*441 4.6s54 t,4 684 3.84e*8 9.883 4 3.nes 4 4.sesee 3.6804 o*Let 3.97t*3 3.4eget 6.6e54 S. tee *e e.ges*e e.ses*e 0.444*4 sy* 4 M

4. H5

1 4 568 8 3.t95*3

9. eRe *e 4 4644 e.gend 9.39e*8 Davis-Besse ODCM 28 Revision 4, 1991 0

Table 2-6 O

stoaccum.ilation raetors (eri)

(pC1/ig per pCi/ liter)*

Element rreshwater fish H

9.0E-01 C

4.6t+03 Na 1.0t+02 P

3.0E+03 Cr 2.0t+02 Mn 4.0t+02 re 1.0t+02 Co 5.0t+01 Ni 1.0t+02 Cu 5.0E+01 En 2.0E+03 er 4.2E+02 i

Rb 2.0E+03 Sr 3.0t+01 Y

2.St+01 tr 3.3E+00 Nb 3.0E+04 Mo 1.0t+01 TO 1.5E+01 Ru 1.0E+01 Rh 1.0t+01 A

Ag 2.3E+00 8b 1.0t+00

-(,)

Te 4.0E+02 I

1.St+01 Cs 2.0E+03 Ba-4.0E+00 La 2.5t+01 ce 1.0t+00 Pr 2.5E+01 Nd 2.St+01 W

1.2t+03 Np

-1.0t+01

Values in this Table are taken (som Pegulatory Outde 1.109 except for phosphorus which is adapted from NURIG/CR-1336 and silver and antimony vM-i are taken from UCRL 50564, Rev. 1, October 1972.

l i

~ ' ' ~ ~ ' " "

O

. =. -

. ~....

- -. ~ ~.. _., -

~.,..

rigure 2-1 Liquid Radioactive Effluent Monitoring and Processing Diagram I

I 8

8 I

i l

E il h

~

eeI 1

?

l 3pI l_-l a

W U Wi g

3 gil I I __ _

g el li

-n m,

f I

k h

9

=d I.

i l

1 w

Hi t@-

-=

m He t

e sis u

j -i mw 3

de L

Eu II 3

o t

h WE I

8l &lt H

g k$.

E g

30 Revision 4, 1991

a 3.0 GASEOUS EFFLUENTS The radioactive gaseous effluent instrumentatinn Ia provided to mnnitor and m

control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with methods in the ODCH Section 3.3 to ensure that the alarm / trip vill occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of ODCH Section 3.3 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in Section 3.3.

3.1 Radiation Monitoring Instrumentation and Controls This Section of the ODCM specifies the gaseous effluent monitoring instrumentation required at Davis-Besse for controlling and monitoring radioactive effluents as required by TS.

Location and control function of these monitors are displayed in ODCH Figure 3-1.

Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNF.L CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 3-2.

Each of these operations shall be performed within the specified time interval with a maximum allovable extension not to exceed 25 percent of the specified interval.

NOTE 6 The monitors specified in Table 3-2 are inoperable if surveillances are not performed or setpoints are less conservative than required.

Vith a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptable conservative.

Vith less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the actions shown in Table 3-1.

Exert best efforts to return the instruments to nPERABLE status within in u "v ' -

days and, if unsuccessful, explain in the next Semlannual Effl e ne Disposal Report (Section 7.2) vhy the inoperability vas not corrected in a timely manner.

fn Davis-Besse ODCM 31 Revision 4, 1991

a 1

1.

4 3.1.1 Station Vent Stack (RE 4598 AA, BA)

The Station Vent is the final release point for all gaseous radioactive effluents. The Station Vent stack monitoring system consists of a high and lov range isokinetic samplers. Three separate channels (A, B and C) are provided for each monitoring system. Channel A represents a gross gamma detector vieving a fixed particulate filter campler. Channel B is a gross gamma detector on a cartridge sampler (e.g., charcoal or Ag teolite) and Channel C is the gross gamma detector viewing a fixed air volume measuring for noble gases. Only the Channel C radiation detector is required in order to comply with the TS requirements. Channels A and B detectors provide information on potential radiciodine and particulate releases. However, these type configurations, monitor viewing a fixed filter, experience vide variations in response due in part to the much more abundant noble gases in the effluent stream relative to the particulate or radioiodines being sampled. Therefore, while the Channels A and B provide useful information for identifying potential particulate and radiciodine releases, they are not used for quantifying the release rate es required by TS.

Refer to Section 3.6.

The following sampling / monitoring instrumentation on the Station Vent is required by Table 3-1, Radioactive Gaseous Effluent Monitors -

Instrumentation noble gas activity monitor (Channel C) iodine sampler cartridge (Channel B) particulate sampler filter (Channel A) sampler flow rate measuring device unit vent flow rate measuring device (computer points. FB83 and F885) 3.1.2 Vaste Gas Decay System (RE-1822 A&B)

The radioactive vaste gas discharge line is rontinuously monitored by two off-line detectors, each measuring gross activity. The menitors control function vill isolate the vaste discharge lines prior to exceeding the established alarm setpoint. Table 3-1 requires that the Vaste Gas Decay System contein as a minimum the following instrumentations noble gas activity monitor (RE-1822 A or B) effluent system flow rate measuring device (PT-1821 and 1821 A)

If the noble gas detector is declared inoperable, the contents of the tank may be released provided that prior to the releases at least two independent gas samples are collected and analyred by gamma spectroscopy for principal gamma emitters (noble gases):

at least two independent verifications of the release rate calculations are performed and Davis-Besse ODCH 32 Revision 4, 1991 i

l at least two independent verifications of the discharge valve line-up are performed.

If the flow rate device is inoperable, offluent releases may continue provided that the flow rate is estimated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Flov rates may be estimated based on fan curves or discharge valve header positioning.

L 3.1.3 Conteinment Porte Exhaust Filter Monitor (RE-5052 A,B&C)

This detector monitors the containment atmosphere for radioactivity during Containment VENT or PURGE. The noble gas activity monitor (Channel C) is required by Table 3-1, providing an automatic isolation of the release prior to exceeding the limits of Section 3.3 and 75. Although not required in order to comply with TS, the particulate and iodine detectors (Channel A&B, respective) provide indications of increasing levels of particulate and radiciodine releases.

3.1.4 Hydrozen Purge Line The hydrogen purge line serves as a Containment pressure relief roote through the Station Vent. A separate ru i.sion monitor on this line is not required. Any release vill be monitored by the Station Vent monitor RE-4598.

3.1.5 Veste cas system oxygen Monitor The Vaste Gas System in provided with an oxygen monitor (with an alarm function) as required by TS 3.3.3.10 to alert operators in the unlikely

.C]

[

event of oxygen-leakage into the vaste gas header. The TS requires that the concentration of oxygen be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume. An oxygen concentration above the specified limit vill actuate a local and control room alarm (TS 3.11.2.5).

3.2 Sampling and Analysis of Gaseous Effluents The program for sampling and analysis of gaseous vaste is prescribed in this Section and incorporates the basic requirements outlined in TS. Radioactive gaseous'vastes shall be sampled and analysed according to the sampling and analysis program of Table 3-3.

This table distinguishes two types-of gaseous releases (1) the vaste gas decay tank release and Containment PURGE are treated as BATCH RELEASESI and (2) routine releases from building ventilation system via the Station Vent are treated as continuous releases.

Containment pressure releases upon startup are considered batch relaamaa.

l 3.2.1 Station Vent Release All releases from the Station Vent are required to be continuously sampled for radioactivity. As specified in Table 3-3, the following samples and analysis are required:

once per week, analysis of an absorption media (e.g., charcoal cartridge) for I-1311 Davis-Besse ODCM 33 Revision 4, 1991 l

l

once per veek, analysis of a filter sample for all principal gamma emitters (particulate radioactive material):

once per month, analysis of a grab gas sample for all prinelpal gamma emitters (noble gas) and tritium once per month, analysis of a composite of the particulate samples of all releases for that month for gross alpha activitys once per quarter, analysis of a composite of the particulate samples for all releases for that month for St-89 and 901 continuous monitoring for noble gases (gross beta and gamma activity) (provided by RE-4598 AA and BA, Channel C as previously discussed in Section 3.1.1).

3.2.2 Vaste Gas Decay Tank Release and Containment PURGE (BATCH)

Table 3-3 requires that a grab gas sample be collected and analysed prior to each BATCH RELEASE from the Vaste Gas Decay Tanks (VGDT) or a Containment PURGE. The analysis shall include the identification of all principal gamma emitters (noble gas) and tritium.

For a p1anned Contalnment PURGE. the results oi the sample and analyais are used to establish the acceptable release rate and radiation monitor alarm setpoint in accordance with ODCH Section 3.3.

This evaluation is necessary to ensure compliance with the dose rate limits of Section 3.3.1.

3.3 Caseous Effluent Monitor Setpoint Delermination 3.3.1 Station Vent All releases of gaseous radioactive of,fluents are via the Sta'lon Vent. TS requires that alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of efiluents does not exceed the following limits.

The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY ahall be limited to the following for noble gas less than or equal to 500 mrem / year to the total body and less than or equal to 3000 ares / year to the skin, and iodine-131, tritium and all radionuclides in particulate form with half-lives greater than 8 days:

less than or equal to 1500 mres/ year to any organ. (The evaluation of the release rate of the radiolodines and particulates is based on the weekly continuous samplers. Refer to ODCH Section 3.6).

Vith the dose rate (s) exceeding the above limits, without delay restore the release rate to within the above limit (s).

This requirement is provided to ensure that the dose at the SITE BOUNDARY from gaseous effluents fron all units on the site vill be within the annual dose limits of 10 CTR Part 20 for UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B. Table II. These limits provide reasonable assurance that Davis-Besse ODCH 34 Revision 4, 1991

4 radioactive material discharged in gaseous effluents vill not result in the exposure of a HEMBER OF THE PUBLIC outside the SITE BOUNDARY to annual p

average concentrations exceeding the limits specified in Appendix B. Table II of 10 CFR Part 20 (10 CFR Part 20.106(a)). For MEMBERS OF THE PUBLIC vho may at times be within the $1TE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC vill be sufficiently lov to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified release limits restrict the corresponding gamma and beta doses above background to an individual at or beyond the UNRESTRICTED AREA boundary to <500 mres/ year to the total body or to (3000 mrem / year to the skin. These release limits also ren rict, at all times, the correspending thyroid doses above background to :

tild via the inhalation pathway to

$1500 area / year.

From a grab sample analysis of the applicable source (i.e., Station Vent, Vaste Ces Decay Tanks, or Containment atmosphere), the radiation monitoring alarm satpoint may be established by the following calculational method.

CR

500 (3-1)

SP

=

g g

=

3000 (3-2)

SP, CR, wheres SP" = limiting concentration in the effluent stream (i.e., setpoint of the monitor) corresponding to the release rate limit for the total body dose rate of 500 mrom per year (vC1/ml)

O' SP'

= limiting concentration in the effluent stream (i.e., setpoint of the monitor) corresponding to the release rate limit for the skin dose rate of 3000 mrom per year (uct/ml)

CR, = monitor response corresponding to a dose rate of one (1) mresvyear, total body (vC1/mi per ares /yr. total body)

CR,

= monitor response corresponding to a dose rate of one (1) areevyr, skin (vC1/mi per mresvyr, skin) 500 = total body dose rate limit (ares /yr) 3000 = skin dose rate limit (ares /yr)

The value for CR (monitor response corresponding to a dose rate of 1 Pr**

per year)'is dependent on the radionuclide distribution. Based on the measured distribution CR is calculated by the equations:

I Ci 1.67E+01

y/0

Vr

E(C,

K,)

Davis-Besse ODCM 35 Revision 4, 1991

and CR' (3-4)

=

1.67E+01

yA)

VT

E(C, * (L, + 1.1 M,)

wheres

= annual average meteorological dispersiop)to the controlling site yA) boundary location from Table 3-4 (sec/m Vr

= ventilation system flow rate for the applicable release point and monitor (liters / minute)

C'

= concentration of noble gas radionuclide (1) as determined by gama spectral analysis of grab sampic (vC1/ml)

K,

=totalbodydoseconvptsionfactorfornoblegasradionuclide(i) in mrenVyr per vCi/m (from Table 3-5)

L,

=betaskindoseconveystonfactorfornoblegasradionuclide(1) in mren/yr per vci/m (from Table 3-5)

M,

= gansna air dose conve[sion factor for noble gas radionuclide (1) in mrad /yr per uci/m (from Table 3-5) 1.1 = mrem skin dose per mrad game air dose (mrewinrad) 1.67E+01 = 1E+03 (ml/1) * (1/60) (min /sec) ne more limiting value (i.e., lower of the two values for SPcalculatedaboveisusedfo Vent monitors (RE-4598) sensitivities and read outs are in vCIAn1; however, the Containment Purge Exhaust Monitors (RE-5052) and the WGDT monitors (RE-1822) sensitivities and read outs are in counts per minute. @erefore, for RE-5052 and RE-1822, the setpoints in vCi/ml nust be corrected to an equivalent monitor count, p.ir minute. We monitor calibration curves are used for determining spectite radionuclide sensitivities (cp VvC1/ml).

otnervise, the monitor sensitivity for xe-133 may be used in lieu of the sensitivity values for the individual radionuclides. Because of its lower gamma energy and corresponding monitor response, the Xe-133 sensitivity provides a conservative value for alarm setpoint determination.

3.3.2 Conservative, Generic Alarm Setpoints Conservative alarm setpoints may be established, in lieu of the individual radionuclide evaluation as described above. This approach eliminates the need to adjust the setpoint periodically to reflect minor changes in radionuclide distribution or release flow rate. The alarm setpoint may be conservatively determined based on an assumed Kr-89 release. We Kr-89 total body dose conversion factor is the most limiting.

Davis-Desse occM 36 Revision 4, 1991

l Therefore, the more restrictive setpoint is based on the total body dose rate limit and may be calculated using equations (3-1) and (3-3). Again,

,h

(

the Xe-133 sensitivity is used for conservatism. 1he alarm setpoint is

(

established in accordance with the Radiation Monitor Setpoint Manual.

3.3.3 Gaseous Ef fluent Alarm Response - Evaluating Actual Release conditions The monitor alarm setpoint is used as the primary method for ensuring and demonstrating compliance with the release rate limits of Section 3.3.1.

Not exceeding alarm setpoints constitutes a demonstration that release rates have been maintained within the limits. When an effluent noble gas monitor exceeds the alarm setpoint, an evaluation of compliance with the release rate limits must be performed using actual release conditions. This evaluation requires collect'.ng a sample of the offluent to establish actual radionuclide concentrations and nonitor response. The following equations may be used for evaluating conpliance with the release rate limit of Section 3.3.1 for noble gases.

D

= 1.67E+01

y/9

VT

t( K,

C, )

(3-5) g b, = 1.67t+01

XA)

vr

t((L, + 1.1 M,)

C,) (3-6) where:

D

= total body dose rate (mrevyr) g D,

- skin dose rate (mresVyr) y/Q=atmosphericdispersiontothepontrollingSITEBOUNrpY

(]J location from Table 3-6 (sec/m )

L Vr - ventilation system release rate (liters / min)

C'

= concentration of radienuclide (i) as measured in the grab sample (vci/ml)

K'

= total body dose conversion factor for' noble gas radionuclide (1) in arevyr per pCiAn, f rom Table 3-5 L,

=betaskindoseconversiopfactorfornoblegasradionuclide (i) in arevyr per vCiAn, from Table 3-5 M,

=gammaairdoseconversionfactorforpoblegas radionuclide (1) in mrad /yr per vC1/m, from Table 3-5 1.1 = mrem skin dose per mrad gama air dose (mrevmrad) 1.67E+01 = 1E+03 (ml/1) * (1/60) (min /see) l l

(~'

l h)

Davis-Besse ODCM 37 Revision 4, 1991 l

3.4 Release Rate Evaluation - Waste Gas Decay Tank Releases and Containment V056t~ ~

ror a Waste Gas Decay Tank release or a containment PURGE, an evaluation of acceptable release rate shall be performed prior to the release. Based on the measured noble 9as concentration in the grab sample collected per the requirements of Table 3-3, the allowable release rate can be calculated by the following equation:

500 (3-7) pp" =

1.67E+01

yA)

I( K,

C,)

or 3000 (3-8)

RR

=

1.67E+01

X/Q

I((t, + 1.1 M,)

C, )

where RR" = allowable release rate so as not to exceed a dose rate of 500 mre v yr, total body (liters / minute)

RR

= allowable release rate so as not to exceed a dose rate of 3000 mrenvyr, skin (liters / minute) 500 = total body dose rate limit (mrevyr); Section 3.3.1 noble gas 3000 = skin dose rate limit (mrevyr) Section 3.3.1 noble gas or RR ) as calculated above should be used for The lesser value (RR@able r%1 ease rate for the PURGE or WGtyr release.

establishing the al1 3.5 Quantifying Releases - Noble Gases The determination of doses in the environment from releases is dependent on the mixture of the radioactive material. Also, NRC Regulatory Guide 1.21 requires reporting of individual radionuclides released in gaseous effluents. Werefore, the quantities of the individual radionuclides released in the gaseous effluents must be determined.

3.5.1 Quantifying Releases Usino Station Vent Noble Gas Monitor (RE-4598C) he quantification of the continuous gaseous effluents (noble gases) is based on the sampling and analysis of the Station Vent effluent. The monitor provides a measurement of gross radioactive material concentration in the effluent. As required by Table 3-3, a gas canple is collected at least monthly from the Station Vent. And, as discussed in ODCM Section 3.2.2, this gas sample is analyzed by gamma spectroscopy to identify principal gama emitting radionuclides (noble gases). The results of the l

sample analysis may be used to determine the radionuclide releases. mis simplified approach reasonably quantifies the continuous release provided that no atypical levels have been observed (e.g., alert setpoint being exceeded).

l Davis-Besse ODCM 38 Revision 4, 1991

Based on the average noble gas monitor reading over the release period, the individual noble gas radionuclide releases are quantified by the equation

\\

A' Q' = 1.0E+03 *

C

vr

T (3-9) t A, where Q,

= total activity released of radionuclide (i) (uct)

A,

- activity of radionuclide (1) from the gama spectral analysis of the grab sample from the release point (uci)

C

= average gross activity concentration over the release period as measured by the noble gas monitor, excluding any BA101 RELEASES (pCi/ml)

Vr

= ventilation system f1w rate (liters / min)

T

= total time of the release period (min) 1.0E+03

= milliliters per liter 3.5.2 Quantifying Releases with Inoperable Monitors with an inoperable radiation monitor on the Station vent (i.e., the RE-4598, Channel C), the once-per-8 hours grab samples provide the mechanism for the continued control and quantification of releases in accordance with TS requirements. Analysis of grab sanples provides the radioactive material k

concentrations in the effluent. We f1w measurement device, or flw estimate, and the release duration provide the total volume released. With these, the release rate and resultant total amount of radioactive material released can be determined, a.

Release Pate Evaluation. With an inoperable monitor, the demonstration of compliance with the release rate limit of Section 3.3.1 for noble gases suust be based on the periodic grab samples. Wese grab sanples provide a measurement of the noble gas concentration in the effluent stream. Equations (3-5) and (3-6) can be used for demonstrating that the measured release rate and corresponding calculated dose rate are within the limits, b.

Total Release Evaluation. he grab samples are also used to quantify the total releases. We measured noble gas radionuclide concentratiem in the grab samples are considered representative of the average effluent concentrations over the release period (i.e., elapsed time since last sample). The follwing equation may be used for determining the release quantities from any release point based on the grab sample analysis:

Q,

= 1.0E+03

vr

T

C, (3-10)

Davis-Besse ODCM 39 Revision 4, 1991

where:

0,

= total activity released of radionuclide (i) (uci)

Vr

= ventilation system release rate (1Anin)

T

= total time of release period (min) 1.0E+03

= milliliters per liter C'

= concentration of radionuclide (i) as measured in the grab sample (vCi/ml) 3.6 SITE BOUIRARY Dose Rate - Radiciodine and Particulates section 3.3.1 limits the dose rate to <1500 mrem /yr to any organ for gaseous releases of I-131, tritium and~all particulates with half-lives greater than 8 days. To denenstrate conpliance with this limit, an evaluation is perfo med at a frequency no greater than that corresponding to the sampling and analysis time period (nominally once per 7 days). The folicwing equation may be used for the dose rate evaluation:

D, y/Q

E( R,

Q, )

(3-11)

=

where D,

= dose rate to organ (o) over the sanpling time period (mrenvyr)

Y location

= atmospheric cispersion to the controlling SITE p)

M) for the inhalation pathway from Table 3-6 (sec/m R,

= dose parameter for radionuclide (i) for the controlling age vCiAn,via the inhalation pathway f rom Table 3-7 (mrenVyr per group

)

O,

- average release rate over the apptcpriate sampling period and analysis frequency for radionuclide (1), that is I-131, tritium or other radionuclide in particulate form with half-lives greater than 8 days (uci/sec) l 3.6.1 Sinclified Dose Rate Evaluation for Radiciodines and Particulates It is conservative to perform a sinplified evaluation of allowable releases l

by applying the I-131 dose factor to the collective releases for all measured radionuclides. By substituting 1500 mrenvyr for D and solving fat i

Q, an allowable release rate can be determined. Based on the annual average l

meteorological dispersion (see Table 3-6) and the dose factor for the most limitingpotentialpathway,agegry),andorgan(inhalation, child, thyroid

-R

= 1.62E+07 mrenvyr per vCi/m the allowable release rate (based on I-131)is44.1vCi/sec. An added conservatism multiplier of 0.8 has been included in this calculation to account for any potential dose contribution from other radioactive particulate material. For a 7-day period, which is the nominal sangling and analysis frequency, the cumulative release would be 26.7 C1. Therefore, as long as the releases in any 7-day period do not exceed 26.7 C1, no additional analyses are needed to verify coupliance with the Section 3.3.1 limits on allowable release rate.

Revision 4, 1991 Davis-Besse 00CM 40

3.7 Noble Gas Effluent Dose Calculations - 10 c n 50 3.7.1 UNRES'!RICTED AREA Dose - Noble Gases cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in this Section at least once per 31 days. 21s periodic assessment of releases of noble gases is to evaluate compliance with the quarterly dose limits and calendar year limits.

W e air dose due to noble gases released in gaseous effluents to areas at and beyond the $1TE BOUNDARY thall be limited to the following during any calendar quarter less than or equal to 5 mend for gamma radiation and less than or equal to 10 mrad for beta radiation, and i

during any calendar year: less than or equal to 10 mead for ganna radiation and less than or equal to 20 mrad for beta radiation.

With the calculated air dose from radioactive noble gases in gaseous effluents ex::eeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Ceaudasion within 30 days, pursuant to Section 7.3, a special Report that identifies the cause(s) for exceeding the

. limit (s) and defines the corrective actions that have been taken to reduce the. releases and the proposed corrective actions to be taken to assure that subsequent releases will be in eenplience with the above limits.

i W is specification la provided to implement the requirements of Section tO II.8, III.A and IV.A of Appendix I,-10 CFR Part 50. S e limits specified-above provide'the required operating flexibility and at the same time implement the guides set forth in-Section IV.A of Appendix I to assure that the releases.of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable.' ' his Section implements the requirements of Section III.A of ix ! that conformance with the guides of Appendix

! to be shown by cal ational procedures based on models and data such that the actual. exposure of an individual through the-appropriate pathways is

.unlikely to be substantially underestimated. We dose calculations established for calculating the doses due to the actual releass rates of radioactive noble gases in gaseous effluents are consistent with the-methodology pr w!,ded in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Rocine Releases of Reactor Effluents for the Purpose of '

Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from.

-Light-Water-Cooled Reactors," Revision 1,, July 1977.-

h e following equations may be used to calculate the gasma-air and-beta-air doses:

Dy = 3.17E-08

X/O

I (M,

O )

(3-12) g D6 = 3.17E-08

XA)

f (N,

Q, )

(3-13)

Davis-Besse 00CM 41 Revision'4, 1991 e

ve.

---..e---

~. -

m,-,-n-e

.N-

-v-w,,,,_

m

--a e,

~..--..-. w a

where Dy - air dose due to gama emissions for noble gas radionuclides (trad)

D6 = air dose due to beta emissions for noble gas radionuclides (mrad)

X/Q = atmospheric disprsion to the controlling $1TE BOUNDARY location (secAn, f rom Table 3-6)

Q,

- cumulative release of noble gas radionuclide (i) over the period of interest (vC1)

M'

= air dose factor due to gama emissions [ rom noble gas radionuclide (1) (mrad /yr per uciAn, f rom Table 3-5)

N'

=airdosefactorduetobetaemissionsf[omnoble gas radionuclide (i) (mrad /yr per vC1/m, from Table 3-5) 1/3.15t+07 (yr/sec) 3.17E-08 3.7.2 Sinclified Dose Calculation for Noble Gases In lieu of the individual noble gas radionuclide dose assessment presented above, the following simplified dose cale.11ational equations may be used for verifying compliance with the dose limits of Section 3.7.1.

(Refer to Appendix B for the derivation and justification of this simplified method.)

2. 0

3.17 t-00

y/Q

M,,,

t Q, (3-14)

Dy

=

and 2. 0

3.17 E-08

yA)

N,,,

I Q, (3-15)

DS

=

where M = 5.7E+02, effective' gamma-air dose factor from Appendix B (mrad /yr per vCi/m )

N,,=1.1+03,effectivep)ta-airdosefactorfromAppendixB (mrad /yr per vCi/m 2.0 = conservatinm factor to account for potential variability in the radionuclide distribution Davis-Besse ODCM 42 Revision 4, 1991

l 3.8 Radiciodine and Particulate Dose calculations - 10 CFR 50 3.8.1 UNRESTRICTED AREA Dose - Rhdioiodine and Particulates

/]

V A periodic assessment is required to evaluate compliance with the quarterly dose limit and the calendar year limit to any organ. Cumulative dose contributions for the current calendar quarter and current calendar year for I-131, tritium, and radioauclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in this Section at least once per 31 days.

We dose to a FEMBER OF 'n!E PUBLIC from I-131, tritium and all racionuclides in pcrticulate form with half-lives greater than 8 days in gaseous effluents released to areas at and beyond the SI'!T BOUNDARY shall be limited to the following:

- During any calendar quarter: less than or equal to 7.5 mrem to any organ, and During any calendar years less than or equal t 15 mrem to any organ.

t With the calculated dose from the release of iodine-131, v Atlum and radienuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the commission within 30 days, pursuant to Section 7.3, a special report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that p) subsequent releases will be in compliance with the above limits.

(

This requirement is provided to inclement the requirements of Section II.C, III.A, and IV.A of Appendix I, 10 CTR Part 50. We limits are the guides set forth in Section II.C of Appendix I.

@e actions specified provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." We 00CM calculational methods specified in this Section implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedure based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. We ODCM methods for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Coupliance with 10 CTR 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.

/~

t\\

Davis-Besse ODCM 43 Revision 4, 1991

We release rate specifications for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the (NRESTRICTED AREA. We pathways which are examined in the developnent of these calculations are:

- individual inhalation of airborne radionuclides,

- deposition of radionuclides into green leafy vegetation with subsequent consumption by man,

- deposition onto grassy areas where milk animals and meat producing animals grase with consumption of the milk and meat by man, and

- deposition on the ground with subsequent exposure of man.

W e following equation may be used to evaluate the maximum organ dose due to releases of iodine-131, tritium and particulates with half-lives greater than 8 days:

D,, p = 3.17E-08

W

Sr,

I (R,,

Q,)

(3-16)

Where:

D

= dose or dose comitment via controlling pathway (p) and age group (a) (as identified in table 3-6) to organ (o),

including the total body (mrem)

W

= atmospheric dispersion paramenter to the controlling location (s) as identified.in table 3-6 W

= X/Q atmospheric dispersion for inhalation thway and H-3 dose contribution via other pathways (sec/m )

W

=D/Q,atmosphericdepositionforvegetapion, milk and ground plane exposure pathways (m' )

=dosefactorfororgan(o)forradionuclide(1), (mrenVyr per R,,

vC1/m ) or (m' - mrenvyr per vCi/sec) from Table 3-7 for each age group (a) and the applicable pathway (p) as identified in Table 3-6.

Values for R were derived in accordance with the methodsdescribedinNURbb-0133.

- cumulative release over the period of interest for radionuclide Q,

(i) - I-131 or radioactive material in particulate form with half-lives greater than 8 days (vCi).

SF

= annual seasonal correction factor to account for the fraction of the year that the applicable exposure pathway does not exist.

1) For milk and vegetation exposure pathways: A six month fresh vegetation and grazing season (May through October) limits exposure through this pathway to half the year = 0.5 1.0

2) For inhalation and ground plane exposure pathways:

=

3.17E-08 = 1/3.15E+07 (yr/sec)

Davis-Besse ODCM 44 Revision 4, 1991

4-ne age group with the highest potential dose via the controlling pathway

(

should be used for evaluating the maximm exposed indiv! dual. his

('

determination is based on a comparison of the age group pathway dose conversion factors (Table 3-7).

Only the controlling age group and pathway identificd in Table 3-6 need be evaluated for compliance with the limits of Section 6.8.1.

3.8.2 Simplified Dose Calculation for Radiolodines and Particulates In lieu of the individual radionuclide (I-131 and particulates) dose assessment presented above, the following simplified dose calculation may be used for verifying compliance with the dose limits of Section 3.8.1.

IQi (3-17)

D,,, - 3.17E-08

W

Sr,

R where D,,, - maximum organ dose (mrem)

R

- I-131 dose parameter for the thyroid for the identified controlling pathway

-4.76E+10,childthyrpiddoseparameterforthe vegetable pathway (a - mres/yr per vCi/sec)

We ground plane exposure and inhalation pathways need not be considered when the tbove simplified calculational method is used because of the overall negligible contribution of these pathways to the total thyroid dose.

It is recognized that for some particulate radionuclides (e.g., Co-60 and Cs-137), the ground exposure pathway may represent a higher dose contribution than either the vegetation or milk pathway. Ilowever, use of the I-131 thyroid dose factor for all radionuclides will maximize the organ dose calculation, especially considering that no other radionuclide has a higher dose factor for any organ via any pathway than I-131 for the thyroid via the vegetable or milk pathway.

The location of exposure pathways (critical receptors) and the corresponding maxima organ dose calculation should be based on the pathways identified by the annual land use census (Section 5.0) and as identified in Table 3-6.

3.9 Gaseous Effluent Dose Projection As with liquid effluents, gaseous effluents require " processing" if the projected dose exceeds specified limits. This requirement implements the requirements of 10 CFR 50.36a on maintaining and using the appropriate radweste processing equipnent to keep releases ALARA.

1he CASEQUS RADNASTE TREA2DT1' SYSTDt (i.e., Waste Gas Decay Tank) shall be used to reduce noble gas levels prior to discharge when the projected air dose due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY would exceed 0.2 mrad for gama radiation and 0.4 mrad for beta radiation in a 31 day period (i.e., one quarter of the design objective rate).

O Davis-Besse 00CM 45 Revision 4, 1991 i

l

i The VDfrIIATICH EXHAUST TREATMD1T SYSTDI shall be used to reduce radiciodine and particulate effluents, prior to their discharge, when the projected dose due to gaseous effluents releases to areas at or beyond the SITE BOUNDARY would exceed 0.3 mrem to any organ in a 31 day period, rigure 3-1 presents the gaseous ef fluent release points and the CASEOUS PADASTE and VD(TILATICH EXHAUST TREATMDfT SYSTDtS applicable for reducing ef fluents prior to release.

With the gaseous waste being discharged without treatment and in excess of the limits, in lieu of a Licensee Event Report prepare and submit to the comission within 30 days, pursuant to Section 7.3 a Special Report that includes the following information

- Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reasons for the inoperability,

- Actions taken to restore the inoperable equipnent to OPERABLE status, and

- Sumary description of actions (s) taken to prevent a recurrence.

W e requirements that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This requirement implements the requirements of 10 CTR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CrR Part The specified limits governing the use of appropriate portions of the 50.

systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CrR Part 50, for gaseous effluents.

If the GASECUS RADWASTE and VDfrILATICH EXHAUST TREATMDfr SYSTDtS are not being used, dose projections shall be performed at least once per 31 days using the following equations:

Dy,

= Dy * (31/d)

(3-18)

D6

- D6 * (31/d)

(3-19)

P f3-20)

D,,,,

= D,,, * (31/d) wheret Dy'

= gamma-air dose p:ojection for current 31 day pro W

n (mrad) gama-air dose to date for current calendar quarter (mrad)

Dy

=

Dep = beta-air dose projection for current 31 day projection (mrad)

= beta-air dose to date for current calendar quarter (mrad)

D6 Davis-Besse 00CM 46 Revision 4, 1991

D

= maximum organ dose projection for current 31 day projection (mrem)

D

= maximum o.gan dose to date for current calendar quarter as determined by equation (3-16) or (3-17) (mrem) d

= number of days to date in current calendar quarter 31

= number of days in projection O

O Davis-Besse 00CM 47 Revision 4, 1991

D,,,, = maxinua organ dose projection for current 31 day projection (area)

D

- maxinua organ dose to date for current calendar quarter as determined by equation (3-16) or (3-17) (area) d

- number of days to date in current calendar quarter 31

= number of days in projection O

Davis-Besse COCM 47 Revision 4, 1990

O O

O~

~

Table 3-1 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY PARAMETER ACTION 1.

Vaste Gas Decay System (provides automatic isolation) a.

Noble Gas Activity Monitor 1

(1)

Radioactivity Measurement A

b.

Effluent System Flow Rate 1

(1)

System Flow Rate Measurement B Measuring Device 2.

Vaste Gas System (provides alarm function) a.

Oxygen Monitor 1

(2)

% Oxygen D

3.

Containment Purge Monitoring System (provides automatic isolation)

Noble Gas Activity Monitor 1

(1)

Radioactivity measurement C

a.

Revision 4, 1991 48 Davis-Besse ODCM

. TABLE 3-1 (Continued)

RADIk-ffIVE CASE 005 EFFLUENT MONITORING INSTRUMENTATION MINIMUM-CHANNELS OPERABIE APPLICABILITY PARAMETER ACTION INSTRUMEttr l

4.

Station Vent Stack l

(provides alarm function)

Noble Gas Activity Monitor 1

(1)

Radioactivity Heasurement C

a.

b.

Iodine Sampler Cartridge 1

(1)

Verify Presence of E

Cartridge Particulate Sampler Filter 1

(1)

Verify Presence of Filter E

c.

B I

d.

Effluent System Flov 1

(1)

System Flow Rate Measurement Rate Measuring Device l

Sampler Flov Rate Measuring Device 1

(1)

Sampler Flow Rate B

e.

Measurement Revision 4, 1991 49 Davis-Besse ODCM O

O O.

g I '

TABLE 3-1 (Continued)

O TABLE NOTATION

%]

(1) During radioactive vaste gas releases via this pathvay.

(2) During additions to the vaste gas surge tank ACTION A Vith the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, the contents of the tank may be released to the environment provided that prior to initiating the release 1.

At least two independent samples are analyzed in accordance with Table 3-3 for analyses performed with each batch 2.

At least two independent verifications of the release rate calculations are performed:

3.

At least two independent verifications of the discharge valving are performed.

ACTION B Vith the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continuo provided the flow rate is estimated at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION C Vith the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathvay may continue provided grab samples are taken at least once (O) per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION D Vith the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, additions to the vaste gas surge tank may continue provided another method for ascertaining oxygen concentrations, such as grab sample analysis, is implemented to provide measurements at least once per four(4) hours during degassing and daily during other operations.

ACTION E Vith the number of channels OPERABLE less than required by the minimum channels OPERABLE requirement, effluent releases via this pathway may continue provided samples are continuously collected with auxiliary sampling equipment, as required in Table 3-3.

Davis-Bessee ODCH 50 Revision 4,1991

(

l l

l TABLE 3-2 RADI0 ACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL CHECK CHECK CALIBRATION TEST INSTRUMENT 1.

Vaste Gas Decay System R' ' '

Q'3#

Noble Gas Activity Monitor p'**

P a.

b.

Effluent System Flow Rate p' *

fyA R

Q 2.

Containment Purge Vent System a.

Noble Gas Activity Monitor D' * '

p;M

R' ' '

Q'*'

3.

Station vent Stack R' ' '

Q'*'

a.

Noble Gas Activity Monitor D' * '

M W'*'

N/A N/A fVA b.

Iodine Sampler W'*'

N/A N/A tyA c.

Particulate Sampler d.

System Effluent Flow Rate Measurement Device D'*'

N/A R

rya e.

Sampler Flow Rate W'*'

N/A R

N/A Measurement Device 51 Revision 4, 1991 Davis-Besse ODCM 9

9 e..'

's.

1 J

'et TABLE 3-2 (Continued)

TABLE NCPTATION (1) During radioactive waste gas releases via this pathway.

(2) During additions to the waste gas surge tank.

(3) Se CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the alaravtrip setpoint..

.(4) We CHAlelEL FijNCTICNAL' TEST shall also demonstrate that control room alarm annunciation' occurs if the instrument indicates measured levels above the alarnVtrip setpoint.

(5) We initial CHANNEL CALIBRATION for-radioactivity measurement instrumentation shall be performed using one or more of they reference standards certified by the National' Institute of standards and Technology or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. Sese standards should permit calibrating the system over its intended range of energy and rate capabilities, for subsequent CHAlt4EL CALIBRATION, sources that have been-releated to 1

the initial. calibration should be used, at intervals of at least once-per eighteen months. For high range sonitoring instrumentation, where calibration with a radioactive source is impractical, an electronic calibration may be substituted for the radiation-source calibration.

(6) We CHAlelEL CALIBRATION shall include the use of standard gas' o

samples containing'a nominal:

1..One volume percent oxygen, balance nitrogen) and 2; Four volume percent oxygen, balance nitrogen.

(7) During containment purges.-

(8)' When used in a continuous mode.

.(P) Prior--to each release, e_

L (R) At least once per 18 months (550 days).

(Q) At least once per 92 days.

(D): At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(M) At least once per 31 days.

(W) At least once per 7 days.

Davis-Besse 52 Re'/ision 4,1991

TABLE 3-3 RADIOACTIVE GASEQUS WPSIE SAMPLDG APO ANP1YSIS Pf0 GRAM MiniEILEm Lower Limit Of Gaseous Release Type Sampling Analysis Type of Detection jLID)

Frequency Frequency Activity Analysis (pCi/al)

P P

Each Each Principal Q uma Emitters

1.0E-04 Waste Gas Decay Release Release Grab Sample H-3 1.0E-06 5

P Containment Purge Each Purge Each Purge Principal Cammma Emitters

1.0E-04 Grab Sample H-3 1.0E-06 M

M Station Vent Stack Grab Sample Principal Canana Dnitters" 1.0E-04 B-3 1.0E-06 W

Continuous

Charcoal I-131 1.0E-12 Sample W

Continuous" Particulate Principal camma 1.0E-ll Sample Daitters' M

Continuous' Composite Particulate Gross Alpha 1.0E-ll Sample Q

Continuous

Composite Particulate St-89, Sr-90 1.0E-11 Sample Continuous" Noble Gas Noble Gases Monitor Gross Beta or Gansna 1.0E-06 Davis-Besse ODCM 53 Revision 4, 1991 O

O O-

i TABLE 3-3 (Continued) e TABLE NOTATION A.

The LLD is the smallest concentration of radioactive material in a sample that vill be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radio-chemical separation):

4.66 s, E

V

2.22 a Y

exp(-Aot)

LLD is the lower limit of detection as defined above (as pCi per unit mass or volume);

s is the standard deviation of the background counting rate or of the ebunting rate of a blank sample as appropriate (as counts per minute):

E is the counting efficiency (as counts per transformation)

V is the sample size (in units of mass or volume):

2.22 is the number of transforinations per minute per picoeuriel Y is the fractional radioch'emical yield (when applicable);

A is the radioactive decay constant for the particular radionuclide; at for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

It should be recognized that the LID is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular sensurement, b.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with sections 3.3.1 and 3.0.

Davis-Besse ODCM 54 Revision 4, 1991 0

1 TABLE 3-3 Montinued)

TABLE I frATION We principal gama emitters for which the LLD specification will apply are c.

exclusively the following radionuclides:

Kr-87, Kr-08, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, re-59, Co-58, co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions, his list does not mean that only these nuclides are to be detected and reported. Other peaks which are measured and identified, together with the above nuclides, shall also be identifed and reported. Nuclides which are below the LLD for the analyses should be reported as "less than" the nuclide's LLD and should not be reported as being present at the LLD level for the nuclide. We "less than" values shall not be used in the required dose calculations. When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Semiarciual Ef fluent and Waste Di'iposal Report.

d.

Frequency notation P - Prior to each release.

M - At least once per 31 days.

W - At least once per 7 days.

Q - At least once per 92 days.

O Davis-Besse ODCM 55 Revision 4, 1991

Table 3-4

' Land-Use census sumary Pathway Locations and Atmospheric Dispersion Parameters Distance' x,a Dp Sector (meters)

Pathways Age Group (secMJ Jg J

_N.

880**

inhalation child 9.15E C.4E-09 telt 870 inhalation child 1.27E-06 1.47E-08 NE 900 inhalation child 1.26E-06 1.585-08 plE*

g.

ESE*

-~

~

,SE*

Sst

'2,900**

vegetation child 6.80E-08 7.90E-10 t

S 1,450**

vegetation child 1.21E-07 2.46E-09 ssW 2,180**

vegetation child-6.45E-08 1.19E-09 SW 1,340**

vegetation child

-2.10E-07 3.94E Wsw 4,270**

cowAnilk infant 5.71E 5,31E-10 W

1~,050**

vegetation child 5.72E-07 8.87E-09 NNN-

-'3,290** -

_ vegetation child 6.28E-08 5.182-10 NW 2,040**

vegetation' child 8.25E-08 -7.28E-10 testi 1,210**

vegetation

. child 2.70E-07.

1.92E *LSince these sectors are located over marsh areas and Lake Erie, no' ingestion-or inhalation pathways are present.

- %ese values-are a change to this table as a result of the 1990 Land Use census.

Davis-Besse 0001 56 Revision 4, 1991 4

I

(:

l.=

o Table 3-5 Dose factors for Noble Gases

O Total Body Skin Gamma Air Beta Air Gama Dose Beta Dose Dose Factor Dose Factor Nuclide Factor K Factor L M

N (mreavy[)pkr (mrenvy[)phr (mraddr'per (mrad /fy)per vC1/m vCiA vC1/m )

Ci/m Kr-83m 7.56E-02 1.93E+01 2.88E+02 Kr-85m 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 1.47E+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-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 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-135m 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 O

Dose factors taken from NRC Regulatory Guide 1.109 Davis-Besse ODCM 57 Revision 4, 1991 0

Table 3-6 Controlling Locations, Pathways and Atmospheric Dispersion for Dose calculations Atmospheric Dispersion ODCM Controlling X/Q DG)

Section Location Pathway (s)

Age Group (sec/m')

( m" 3.3.1 SITE BOUNDARY Noble Gases N/A 1.83E-06 N/A Noble Gas NNE direct exposure 3.3.1 SITE BOrrNDARY inhalation child 1.68E-06 N/A Radiciodine NNE

& Particulates 3.7.1 SITE BOUNDARY gamma-air N/A 1.83E-06 N/A NNE beta-air 3.0.1 cow cow infant 5.71E-08 5,31E-10 4270m, Wsw milk tCTES:

1.

All meteorological dispersion values have been taken from Stone and Webster report, Handbook for 00CM X/Q and DJ Calculations, October 1983.

2.

The noble gas, direct exposure x/Qs are based on the decayed, undepleted values.

3.

The inhalation pathway X/Qs are based on the decayed, depleted values.

O

- s-eesse ecs Se eevisien 4,1ee

11st31o 3-7 Gassous Efflu2nt Pathway DQ p Cormnitment Fcctors

+

i l

R'** Inh 21ctionPathwayDosarpetors-ADULT (mre v yr per vCi/m )

e muettee bone Liver Thgre &4 824 net..

banne Cl* Lt.2 f. So4Y..

..*..a 1.26t*3 1.264*3 L.264*3 1.264 3 1.268*3 1.268*3 m.3 c.14 1 42t*4

1. 4 La

3 3.41t*3

3. 41t

3 3.41t*3 3.41t*3 3.41t*3 me*24 L.028*4 1.023 4 1.028*4 1.028*4 1.022*4 1.02 t *4 1.021*4 Pall 1.32t*4 7.718*4 g.648*4 1.0Lt*4 l.95t*1 2.29t.1 1.448*4 3.J28*3 1.004*2

tall 3.96t**

9.84t*3 1.40t*6 1.748 4 6.30E*3 en.54

%. 2 4 t *0

1. 30t *C 9.44t*3 2.02t*4 1.038*1 anale 7.21t*4 6.034 3 3.943*3 re*S$

2.464*4 1.70t*4 Fe*ie 1.10E*4 2.788*4 1.028*4 1.Segel 1.063*4 3.70tel 3.144*4 6.71t*2 6.928*2 0a*17 9.28t*$ 1.06tel 2.07t*3 1.188*3

o.54 5.97 t *4 2.418+l L. 468 *4 1.152*4 co.60 Ngas) 4.32tel 3.14t*4 1.78tel 1.343 4 1.458*4

5. b34

3 1,238 4 9.128*2 N1*49

1. 54t *0 2.104*1 4.628*0 6.788 3 4.30t*4 6 tSt.L cua64 1.468 0 4.90t*4 4.64t+l 1.348*4 4.668*4 En=43 3.244*4 1.034*l 4 228*2 9.20t*2 1.638 1 4.528 3 ta*69 3.3st*2 6.81t*2 1 048*4 1 318 4 St*42 2.325*2 2.4LE*2 4t*43

1. 64t

3 3 138*2 St* 64 L.J88*1 st'el 1.644*4 S. tot *4 1.3St*S ka*64 3,348*9 1.938 2 3.w?t*2 LD*e4 1.704*2 2.564*2 La.99 1.40t*4

3. 504 *l 4.721*3 It-99

3. 04 8 *1 9.604*4 1.228 5 6.104*4 Stato 1.93t*1 3.61t*4 1.913*$ 2.504 0 Itatt 6.19t*1 1.654*4

4. 3ct *4 2.9 14 *1 St*92 6.748*0 1.70t*1 5.064*S 1.61t*1 T*90

2. 09t

3 1.928*) 1.333*0 1.023*2 Yat te 2.6tt*1 1.708*4 3.813*l 1.244*4 Yatt 4.621*l

=

1.87t*4 1.318 4 1.021 1 Y 92 1.038*1 4.45t*4 4.223*l 2.618*0 Y'93 9.448*1 5.421*4 1.7 7t *4 1.508*5 2.33t*4 Es*95 1.07tel 3.442*4 2.97t*L 7.67t*4 5,231*$ 9.044 *0 sta9?

9.644*1 l.163*L

=

7.748*3 5.058*1 1.043*$ 4.21E*3 Mb*41 1.41t*4 7.028*3 6.letal

2. 6ct

3 2.42t*2 2.05t*2 Mb 97 2.225*1 5.623*2 2.91E*2 9.128*4 2.468+l 2.304*1 1.21t*2 maatt

=

4.423*2 7.648*2 4.164*3 3.704*2 te 99m 1.033*3 2.918-3 5.90t*4 1.048 3 3.99t*2 ts* tot 4.188 5 6.021*1 S.438*3 1.058*1 t.108*$ 6.544*2 av.103 1.138*3 1 023 *C 1.10t *4 4,42t*4 3.113*1 Ru=101 7.904 1 L.34tel 9.364*4 9.121*5 8.725*3 2w.104 4.91E*4

=

a LA*10&a Rh* LOG 1.97t*4 4.634*4 3.028*l S. 948*3 Aq*LLCm 1.048*4 1.00t*4 2.444*4 4.064*l 1.248*4 En*124 3.123*4 1.89t*2 7.ll1*L 1.748*4 1.01t*$ 1.2 64 *4 33 123 1.144*4 1.95t*3 5.60t*1 toallSe 3.423*3 1.368 *3 1.01E*3 1.24t*4 1.148*5 7.068 4 4.67t 2 teat 27a 1.264*4 5.77t*3 3.29t*3

4. 548 *4 9.60t*l 1.508'l 1.57t*3 Te*127 1.60E*0 4.428 1 1.064 *C 5.104 *C 6.51**3 5.74t*4 3.10t*1 te*L2te 9.764*3 4.67t*3 3.44t*3 J.664*4 1.168 4 3.638*l L. 568

3 te* 121 4.944 2 2.39E*2 3.90E*2 1.s73 1 1.948*3 1.17t*2 1.248 2 te*131m 4.998 1 4.344*L $.50t*1 3.09t*2 1.464*6 5.543*l

2. 9CE

L te*131 1.1L2 *2 1.958 3 9.364*3 4.J7t*2 1.394*3 !.6446L 3.lDb=3 ft*t32 2.60E*2 2.158*2 1.90E*2 1.448*3 2.844*l 5.10t*5 1.625*2 7.69t*3 S.20E*3

!*130 4.548*3 L.343*4 1.148*4

2. 09 t

4 6.268*3 2.05t*4 2

LJ L 2.52864 3.544 *4 1.19t*7 6.138*4 4.044*2 1.164*3

!*132 1.164*3 3.264*3 1.14t+l 5.14t*3 8.548*3 4.522*3 1*133 8.648*3

1. 4e8 *4 2.138*4 2.54t*4 1.0 L2 *C 4.158*2 1 134 6.448*2 1.738*3 2.948 *4 2.71t*3

$.218*3 2.57t*3 2*L35 2.644 *3 6.948 *3 4.668+S 1.118*4 2.871*l 9.762 4 1.044*4 7.20 Eel Ca*L34 3.738*1 0.644*$

=

4.568*4 1.208*4 1.173*4 1.108+l Cs*L34

3. 90E *4 1.468*5 2.222*5 7.12t*4 0.60$*3

4. 20E *1 ca 137

4. ?lt et 4.2ttel

4. 608*2 4.444*1 1.648 3 3.244*2 Ca

L 34 3.313*3 4.211 2

=

6.221*4 2.76t*3 4.964*2 2.748*2 se.L39 9.362*1 6.664 4 1.67t*L 1.271*6 2.148*$ 2.l?t*3 6a.160 3.90t*4 4.904 *1 7.005 5 1.94t*3 1.164 7 3.364*3 6a*141 1.0nt*1 1.533*1 1.644*3 2.298 5 1.19t*3 64*142 2.638*2 2.708'S L.J44*9 4.548+S 4.148*1 La*160 3.448*2 1.748 *2 6.338*3 2.113*3 7.733 2 64*142 6.431*1 3.108 1 6.268*3 3.625+l 1.208'l 1.S38*3 Ce*141 1.99t*4 L.358*4 6.044*L 7.98t*4 2.285*1 1.$35*1 Ca=143 1.464*2 1.344*2 4.484*l 7.788*4 4.164*1 1.648*l Ca*l44

3. 4 3t *4 1.438*4 2.16t*3 2.lltel 2,00E*1 4.64t*2 pt*L43 9.3 64 *3 3.75t=3 1.058 3

1. 02.t

3 2.134 4 1.538 3 Pe

164 1.0 LA

2 1.21t*2 3.562*3 2..itel 1.733*$ 3.61E*J e4*147 S.278*3 6.10t*3

+

J.60t*4 1.lltel

2. 464 *C s*197 8.644 *0 1.088 *0 7.0ct*L 3.*64*4 1.198*5 1.248*1 wt*239

2. 30E

2 2.168*1 Davis-Besse 0DC1 59 Revision 4, 1991

Tablo M (cmtirn.**d)

Inhalation Pathway Dose FCc[ ors - TE2NMER R

(aree/yr per pCi/tt )

I w it.4 6=.

uvat n 7t.i4 It.*ev..

t, 03 12.2 7.soor

.*.a..

f~'

nei 1.27t*3 i.27t*3 1.27t*3 1.27t*3 1.27t*

i 27t*3 C.14 2.60t*4 4.87t*3 4.875*3 4.87t*3 4.87t*3 4.471*3 4.87t*3 se 24 1.34t*4 1.388*4 1.388 4 1.368 *4 1.364 4 1.344*4 1.348*4 Po ll 1.898*4 1.108*1 9.264*4 7.168*4 7.50t*4 3.07t*L 2.10t*4 3.008*3 1.354*2 etall 5.11t*4 1.27t*4 1.988,4 6.648*4 8.40t*3 g4.%4 1.70t*0

1. 79t *C L.128 4 5.748*4 3.528 4 in*i4 Fe*st

3. 34t.4 2.Jat*4 1 24t*5 6.J9t*3 S.548*1 Featt 1.19t*4 3.70E*4 1.138*6 1.78t*1 L.4Jt*4

al?

6.928*4 5.864*l 3.148*4 9.204*2 2.07t*3 Ja*g4 1.344 4 9.128 4 2.788*3 4 11t*4 Oaaec 8.72t*6 2.19t*4 L.948*4 4063 1.80t+l 4.34t*4 3.07t*$ 1.428*4 L. 948 *4 unael 2.188 *0 2.938'l 9.J68*3 3.67t*4 1.27t*L 2.038*0 Ou*64 6.41t*0 1.11t*4 6.143*4 8.484*L ta*65 3.84t *4 1.34t*S 648 4 L.24t*6 4.66t*4 6.348 4 In*49 4.838*2 9 20t*3

6. '8*2 1.54t*3 3.854*I 6.4 68 *3 Sr*82 1.828*4 3.448*3 Statl 4.338'2 8t*44 Statt 1.838*L 1.90tel Ra*se 1.77t*4 8.40s 4 l.46t*2 2.928*5 2.728 2 ka*84 3.531*2 3.364*? 2.33t*2 Ra*49 st*89 4.34t*1 2.42t=3 1.11t*t 1.218*4 St*90 L.04t*8 1.414*? 7.45t*l 4.648*6 statt 8.6Ct'l 6.078*4 2.59tel 3.11t *0 a

st*92 9.528'O 2.74t*4 1.49tel 4.044 1 Y*90 2.98t*3 2.934*l

5. 8 94 *$

8.008*1 Y*9 La 3.704 1 3.204 3 3.02 ten 1.428*2 Y=9L 6.41t*$

2. 94 t *4 4.098+l 1.771'4 Y *9 2 1.47t*1 2 648*4 1.658+l 4.J9t

L Y'93 1.31t*2 4.325*4 5.79tel 3.728*0 2t*91 1.468*5

4. 54t *4 6.744*4 3.69t *4 1.49tel 3.194*4 4t*97 1.388*2 2.728*1 4.128*1 1.30tel 6.306*5 1.264*n maa9t L.46t*4 1.038*4 1.00E*4 7.lltel 9.648*4 1.644*3 a

ND*97 3.148*1 I.78t*2 9.128 2 3.93t*3 2.17s*3 2.648 2 1.69t*2 4.118*2 1.548*l

2. 69e *t 3.228*L mo*99 5.164 2 1.15t*3 6.133*3 4.99t*3 tc'99e 1.34t*3 3.448 3

\\

ta=101 1.128*$ 8.40s.5 1.828 3 4.47t*2 8.728 7 8.248*4 7.438*3 7.838*5 1.098 1 8.964*2 Ruot03 2.10t*3

1. 412 *C 4.822*4 9.044*4 4.344*L Au*105 1.128 *0 1.904*5 1.618'? 9.60sel L.248*4 Ru 104 9.448*4 R4*LQ3e La*106 2.loE*4

6. 71t *4 2.738*5 1.99t*3 Ag*lLQs 1.J88 4 1.311 4 1a 124 4.30E*4 7.948*2 9.748*l 3.4 14 *6 3.9eEet 1.644*4 seat 25 7.348*4 0.08t*2 7.04t*L 2.744*6 9.9 18 *4 1.728*4 Teal 25e 4.644*3 2.248*3

1. 60E

3 5.)64*4 7.504*4 4.678*3 Teat 27e

1. 60E *4 8 668*) 4.388*3 6.548 4 1.664*4 1.19E*1 2.188*3 te=127 2.012*C 9.121 1 1.428*0 7.288*0 1.123*4

4. 04E.4 4.428 1 te*129e 1.39t*4 6.54t*3 4.544*3 1.19t *4 1.988*4 4.0 54 *l 2.25t*3 to 129 7.108 2 3.348*2 1.188*3 2.644*1 3.30t*3 1.628*3 1.768 2 f e* L 3 La 9.64tel 4.01E*1 7.25tel 4.Jt4*2-2.384 9 6 21t*$ 4.021*1 Te*L31 1.$48*2 4.318 3 1.244 2 6.188 2 2.344*3

1. 5 La *1 1.048*3 te*132 3.608*2

2. 90t

2 2.468*2 1.958*3

4. 4 94*$ 4.63t*9 3.19t*2 9.128 *3 7.17t*3 tallo 6.248*3 L.79t*4

1. 4 9t *4 2.714*4 6.49E*3 2.664*4 t*131 3.543*4 4.91t*4 1.468 *7 8.40S*4 t.27t*3 1.548*3

[*L32 1.19t*3 4.348*3 1 litet 6.923*3 1.024 4 6.228*3

!*133 1.225*4 2.01t*4 3.928*4 3.594*4 2.043*4 8.4cs*2

!*134 8.84t*2 2.322+1 3.958 4 3.648*3 6.958*3 3.49E*3 f*131 3.70E*3 9.444*3 6.215*1 1.4 98 *4 3.753*5 1.464*t 9.764*3 $.49t*$

Ca*134 S.025,5 1.138*4 1.10t*S

1. 78E *4 1.09t*4 1.37169 CaelJ6 5.11t*4 1.948*5 3.048*$ 1.218*$ 4.64t*3 3.118*5 Ca*137 4.708*$ 8.444*5

+

6.623+2 7.87t*1 2.70E*1 4.464*2 Cs.134 4.664*2 4.544 2 8.648*4 4.464 *3 6.458*3 3.90S*J 84 131 1.344 *0

9. 44t *4 2.268 1 2.038*4 2.19tel 3.l28*3 6e*160

$.478 4 4.70s 1 9.444 5 3.298+3 7.464 4 4.748*3 6e*141 1.42 Rat

1. 06E

4 2.271*3 3.148 1 1.913*3 64 142 3.70s 2 3.70tel 2.164*4 4.878 5 6.268 1 La*140 4.79t*2 2.344*2 1.022 4 1 20E*4 1.064*1 f.a 142 9.60s.1 4.258*1 8.648 *3 6.14t*1

1. 2 64 *$

2.171*3 ce*L41 4.844 4

1. 90t

4

8. 644 *1 1.10tel

3. 514 *t 2.164*1 ce*143 2.668*2

1. 94 A

2

1. 2 8 t *4 1.348*7

4. 644 *S 3.623*1 ca 444 4.89t*4 2.023*4

=

3.09t*3 4.838*$ 2.144*$ 4.621*2 Pr*l43 1.348*4 5.31t*3 1.0 L2

2 1.754,3 2.318*4 3.185*3 Pr*144 4.308*2 1.743*2 a

5.028*3 3.721*$ 1.628 *1 S.131*2 m4*L47 7.864*3 8.568*3 I

4.748*4

1. 7 71*$ 3.438 0 w*t47 1.204*1 9.764*C

1. 00a *2

6. 4 9e *4 4.322*1 L.778 1 wo*2.9 3.368 2 3.ltt*1 Davis-Besse ODCM 60 Revision 4, 1991

Tablo M (contitud)

Inhalction Pathway Dose Fpctors - CHILD R

i (arevyr per vCiAn )

l Nac14de Some

1. leet th 3.eea O!

t.1.!

?.Sudy

. T.F# 68 t18H8Y. *.. 9...

l 1.128*3 1.12t*3 1.12t*3 1.128*3 1 12t*3 4.128*3 wa)

C=44 3.59t*4 6.73t*3 6.738 3 6.738*3 6.73863 6.738*3 6.738*)

24*24 1.4Lt*4 1.61t*4 4 61t*4 1.6Lt*4 4.61t*4 L.41t*4 1.618*4 P*32 2.604*4 1.14tel 4.224*4 9.848 4 8.558*! 3.434*t L.70t*4 4 08t*3 1.548*2 7 tall 4.29t*4 en*le 1.00t*4 1.38t*4 2 29t*4 9.31t*3

=

t 66t *C L.47t.0 1.21t*4 1.J33*l 3.128 4 en = %

Fe.se 4.748 4 2.528*4 t.11tel 2.87t*3 7,77t*3 Fe6 2.075 4 1.248 4 1.27t*6 7.07t*4 1.67t*4 te

U 9.038 2 5.07tel 1.328 4 1.07t*3 L.77t*1

!a.58 1.11t*6 3.448*4 1.168*3 a

1.31t*4 Ce=60 7 071*4 9.421*4 2.26t*4 n=43 4.214*l 4.638*4 2.?ltel 6.231*3 2.80t *4 un=68 2.99t*0 2.96tal 8.18t*3 8.40t*4 1.648 4 1.998*0 6.01t*0 9.584*3 3.47t*4 1.078*0 Cu-64 ta 45 4.26t *4 1.13t+l 7.14t*4 9.95t'l L.438*4 7.038*4 ta*69 6.705 2 9.668 2

$.854*2 L.423*3 1 01t*4 8.9' b 3 St*82 2 09t*4 tr-43 4.748*2 St*84 5.48t*2 8t*81 2.53t*1 1.94t'l Amate 7.99t*3 1.148 9 l.62t*2 Am*68 8.721*l 3.668*2 3.45t*2 As*89 L.498 4 2.90E*2 treet 5.394*l 2.164 4 1.67tel 1.728*4 8tato 1.01t*8 1.48t*7 3.438*l 4.44t*6 tr*91 1.211 2 5.J34*4 1.744 5

4. 59t *0 tr-92 1.118el 2.40t*4 2.42tel S.254*L Y.90 4.11t*3 2.628*$ 2.68tel L 11t*2 Yatta 1.07t*l 2.411*) n.723 3 1.848*2 Y.91 9.14tel 2.632 4 t.848 5 2.448*4 Y*t2 2.048*n 2.314 4 2.398 5 $.81t*1 Y'93 1.868*2 7.448*4 1.498*1 1.11t*0 1s. )$

1.90tel 4.16t*4 1.948*4 2.238*6 4.11t*4 3.704*4 ar-97 1.888 2 2.728*L 3.09t*1 1.13t el 3.ltt el n. 60t *1 Ns.9%

2.358 4 9.18t*3 8.62t*3 6.148*l 3.708*4 6.554*3 ND*97 4.29t*L 7.10t*2 4.llt*2 3.428*3 2,788*4 3.60E*2 1.72t*2 3.921*2 1.35t*l L.27tel 4.264+1 mo.99 tc.99e

't.784 3 3.484*3 l.07t*2 9.51t*2 4.812+) 9.778 2 teat 01 8.10tal 8.lltel 1.458*3 $.848*2 1 638*1 1.Cet*3 tue103 2.19t*3 7.038*) 6.42t+l 4.448*4 L.07t*3 av.105 1.538*0 1.344 *3 3.19t*4 9.954*4 1.558*l

>. 106 1.364*l L.844*l 1.431 7 4.29tel

1. 69t *4 ga.103a

=

an.104 Aq 11Cn 1.698 4 L.148*4 2.!!!*4 S.44t*6 1.00tel 9.148*)

18 124 5.748*4 7.44t*2 1.24t

2 3.24t*6 1 64t*$ 2.00t*4 Statil 9.44t*4 7.59t*2 9.10t

L 2.325*4 4.038 4 2.07t*4 a

Te*lita 6.738 3 2.338*3 1.922*3 4.77tel

3. J44 *4 9.148*2 teal 27a 2.498 4 8.158*3 4.07E*3 6.362*4 1.48t*6 7.14a*4 3.028*3 Teall?

2. ??t *0 9.51t*1 1 94 tea 7.07t.0 1.00t*4 5.428*4 6.1 L2 1 fe.129m 1.928 4 6.454*) 6.338 3 1.034*4 1.764 4 1.822*l 3.044*3 Teal 29 9.77t*2 3.508 2 7.148 2 2.57t*1 2.938*3 2.558*4 2.388 2 te*131a 1.348 2 1.928*1 9.778*1 4.00E*2 2.04tel 3.04tel 5.07t*l te 131 2.175 2 4.444*3 1.70s 2 S.448*2 2.058*3 1.331*3 6.19tal Te=L32 4.8L8*2 2.728*2 3.47t+2 1.77t*3 3.77tel L.34tel 2.633*2 tan 30 8.188

3 1.64t *4 1.854*4 2.45t*4 1.112 *3 8.444*3

(*131 4.0 tA*4 4.818*4 1.628*7 7.888*4 2 848*3 2.73t*4 1 13J 2.122*3 4.07t*3 1.94t*9 6.258 3 3.208*3 1.848*3 3.L33 1.664*4 2.038 4 3.85t*4 1.34t*4 l.488 3 7.70t*3

!*tle L.171*3 2.164*) 1.078 4 3.30t*3 1.55t*2 9.954*2 1 131 4.922*3 8.738*) 7.928*1 1.34t*4 4.448*3 4.148*3 Ca*L34 6.51t*$ 1.01E *4

3. 30E el 1.28tel 3.858*3 2.214*t Ca*136 6.51t*4 1.718*1 9.llt*4 1.458 4 4.188*3 1.144*l Ca*137 9.07tel 8.218 5 2.823*l 1.048 5 3 628*3 1.244*5 es.134 6.338*2 8.40s.2 6.221 2 6.8ttel 2.70t*2 S.lltel Se*439 L. 44 t *0 9.844*4 8.624*4 S.77t*3 S.77t*4 S.37t*J se*L40 7.40t*4 6.484*l 2.11t*1 1.748*6 1.023*l 4.338+l me 14L L.964=L 1.098 4 9.47t 5 2.922*3 4.758 2 6.368 3 84.n42 1.00t*2 3.608*1 2.918 5 1.648*3 2.744*0 2.79t*3 La*140 6.448 3 2.218*2

1. 8 3 t *S 2.268*1 7.llt*1 1.4*l42 1.30t*0 4 11E=1 8.70t*3 7.59t*4 1.29tal co 141 3.128*4 1.95t*4 8.55t*3 S.44sel 5.664 4 2.90E*J co.14) 3.644 2 1.99t*2 8.36t*1 1.;ltel L.27t*1 2,873+1 ca.144 6.778 6 2.128 6 1.17 t *6.1. 2 04 7 3.495 5 3.615*5 Frat 43 L.89t*4 9.llt*3

3. 00t

1 4.33tel 9.738*4 9.14t+2 Prel44 S.944*2 1.85t*2 9.77t*3 1.17t.3 1 97t*2 3.003 3

i.147 1.048*4 8.738*3 4.415*3 3.298+l 8.21t*4 6.81t*2 u-L87 1.63tel 9.665*0 4..;t*4 1.10t *4 4.338*0

'op.239 4.644*2 1.348*L 9.738*1 1.818*4 4.4ct*4 2.318+t Davis-Besse ODCM 61 Revision 4, 1991

Tablo 3-7 (continued) 4 Inhalction pathway Dose FOptors - IN D NT R,

(

(ares /yr per pCIAn )

%elade tone Liver Thtnit tienet..

Lung 03.LL2 f.gesy p

n.3 4.478 4 6.478 0 6.47t*2

6. 41t 4 6.47t*2 6.47t*2 ble 2.658*4 S.3L8 0

5. !!8 0 5.318+e 5.21t*3 1.318*3 5.318 4

(

no.44 1,04t*4 1.ME

4 1.06t*4

1. ME *4

1. M4 *4 1.064*4 1.M t*4 poll 2.034*6 1.1J8*l 8.6L8*4 7.748*4 gr.91 5.71t*1 1.32tel

1. Jet *4 3.578*2 e,918e1

%.

4 2.538*4 4.94t*3 L.00t*6 7.068*3 4.968*3 v.n 1.54t*0 1.10t *0 1.Jlt*4 7.173 4 2.212 1 rool

!.97t*4 1.178 4 4.69t*4 1.098 0 3.33863 f e et 1.368 4 2.35t*4 1.02t*6 2.4e8 4 9.448 4

> l?

4.51E*2 3.79tel 4.064 0 4.418 4

o te 1.221*3 7.77t's 1.118 4 1.628*3

=

o.ec 4.02E4 4.513*4 3.198*4 1.188 *4 up6) 3.398 4 2.04 t

4 2.09tel 2.423*3 L.164*4 ut*65 2.39t*0 2.448 1 8.12 *3 S. Cited 1.236 1 cv.64 1.488 4 3.98t*0 9.308 4 1.10t*4 7.748*l 4

In*65 1 918*4 6.268*4 3.258*4 6,47tel 1.143*4 3.11t*4

&nett 1.394 4 9.678 2 4.028*2

1. 4 7 t 4 1.328*4 7.184 4 tr*02 1.338*4 tr*41

=

3.41t*2 Gr*44 4.00tel tr.el

2. 048*1 Ab 44 1.908*9 3.044*) 8.828*4 maatt 1.571 4 3.39t*

2.878 4 6.428.2 A3 49 3.21t*2 1 2.M24 Sr.tt 3.94tel 2.031*4 6.404*4 1.148*4 Stato 4.09t*7 i.121*7 1.3LI*l 2.898 4 5t*91 9.148 4 5.264 4 7.348*4 3.444 4 Sr.42 L.Citet 3.3at.4 1 40t*5 J.118 1 Y 90 3.29t d 2.698 0 1.048*5 0.82tet Yetta 4.07t*1 2.794 4 2.354 4 1.39t*2 V-91 S.44t*l 2.458 4 7.038*4 1.57t*4 Y.92 1.648*L 2.45t*4 L 27tel 4.61t*1 T-91 1.10t*2 7.648*4 1.67899 4.071 4 Es.95 1.15tel 2.798=4 3.118*4 L.758 4 2.17t*4 2.038*4 tr*97

1. lot *2 2.541*n 2.59E*1 1.104*l 1 404*l 1.178 4

=

no.91 1.l?t*4 6.438 4 4.728 0 4.79tel 1.27t*4 3.784 4 unet?

3.428 1 7.29t*2

$.708 2 3.328 4 2.69t*4 2.638 4 no.99 1.658 4 2.618*2 1.354*l 4.473 4 3.23t*L

=

ta=99m 1.404 4 2.843 0 3.113 2 4.11g*2 2.034 4 3.728 2 te 001 6.111 0 8.238 5 9.79t*4 1.444 4 8.448 4 0.128 4 Au*LO3 2.028 4 4.248*3 5.122 4 1.61t*4 6.798 4 Rua105 1.22t*0 8.99t*1 L.57t*4 4.644*4 4.10t*1 Ru 106

8. 64t *4 1.07tel 1.164*7 1.644*l 1.09t*4 Ra.103m

=

En.406

=

A4 110m 9.982 0 7.225 0 1.09t*4 3.67t *4

3. 30t *4 5.008 4 St*124 3.79t*4 1.544 4 1.01**2 2.684*4 5.91t*4 1 208*4 saal25 1.178 4 4.778 2 4.238*L 1.645 4 L.47E*4 1.09t*4 tod2 Se 4.764 0 1.598 4 1.423 4 4.4 7t *6 1.19t*4 6.548*2 te=L27m 1.67E*4 6.508 4 4.87t*3 3.75t*4 1.318 4 2.738*4 2.078 4 to.127 2.238*0 9.53tal 1.858 0 4.864 0 L.038 4 2.444*4 4.498 1 tenWe 1.41t*4 4.098 4 9.478 4 3.18t*4 1.644*4 6.90t*4 2.238 0 to.tn 7.884*2 3.478 4 6.75tal 1.75tal 3.008 0 2.638 4 1.888 2 to.131a 1.07t*3 5.50sen 0.934*l 2.658*2 L.998 5 1.19E*l 3.638*l 7e031 1.748 2 S.228 4 L.544*2 3.995*2 2.064*3 0.325 4 5.00t*3 te.132 3.71E*2 2.27t+2 2.798*2 1.038*3 3.404*5 4.418 4 1.764*2 1030 6.364*3 1.29t*4

1. 40t *4 1.134 *4 1.993 4 $.571*3 1 131 3.79t*4 4.448*4 1.448*? 5.144*4 1.044 4 1.964*4 t=132 1.698 4 3.548 0 1.49tel 3.954 0 1.908 4 1.264 4 1 133 1.32t*4 1.924 4 3.564*4 2.244*4 2.144*3 5.608 0

(*L34 9.21**2 1.04E 0 4.454*4 2.098 0 1.29t*3 6.458*2

!.135 3.864*3 7.60t*3 6.944*$ 8.478 0 1.638 0 2.771 4

=

Ca*134 3.944*l 7.032*1 1.904*S

7. 971 *4 1.338*3 7.45t*4 Coal 36 4.831'4 1.358*5 5.645 *4 1.138*4 L. 4 344 1.29t*4 cs*137 5.498 5 6.123*1 1.728*S 7.138*4 1.338 4 4.554*4 Ca 134 5.054*2 7.812+2 4.10t*2 6.548 4 4.764*2 3.90t*2 a

34*139 1.444*C 9.848 4 5.925 4 S.954 4 S.10t*4 4.J08 2 84 140 5.604 *4 5.604*1 1.J44*L 1.60E*4 3.444*4 2.90t*3 e4 141 1.57t 1 1.004 4 6.508 1 2.97t*3 4.75t*3 4.978 0 64442 3.948 4 3.30E*l 1.90E*5 1.558*3 4.93t*2 L. 9484 L4 140 1.058*2 2.004 4 1.644 4 8.444*4 5.11t*1 La 042 L.038*0 3.77tal 8.223 4 5.953*4 9.048 4 Co*141 2.77t*4 1.478 4 5.255 0 S.17t*5 2.144 *4 1.994 4 ca.143 2.93t*2 1.931 4 1.642*t 1.164*5 4.973 4 2.215*1 ce*144 3.19t *4 1 214 4 5.388*1 9.644*4 1.448 5 1.744*l pt-143 1.408*4 5.24t*3

1. 9 7E 4 4.J3tel 3.722*4 6.99t 4 pr t44 4.798 0 3.454 2 6.723 4 1.41t*3 4.288*3 2.41t*3 m.s 447 7.94t0 8.1Jt*3 3.111 0 3.222*5 1.12t *4 1.00t*2 w.147 L. 304

L 1.022 0 3.464*4 3.544*4 3.128*0 up q 39 3.71E*2 3.321*L 6.628*1 S.45t*4 J. 4 9t *4 1.445*l Davis-Besse 62 Revision 4, 1991

Tablo 3-7 (ccritinued)

R,,, Grass - Cow - Milk Ppthway Dos 3 rectors - ADULT (nyevyr per #CIAL ) for H-3 and C-14 (a

mra vyr per uCi/sec) for others i

m*ellee tone Laver thyrote' Elener Lun9 ct*LLI 7.6mer n.3 s.43t+2 v. 6 38 *2 7.638 2 s.438 2 s.43t+2 s.634 2 0 14 3.43tel 1.268*4 1.26t*4 f.268*4 1 244*4 1,2 63 *4 1.244*4 me 24 2.144*4 2.548 4 2.548*4 2 648*4 2.548*4 2.54t*4 2.544*6 P 3J 1.fttato 1.068*9 1.928*9

6. 60t *4

.r*ll 1.718*4 6.304*3 3.638 4 7.20t*4 2.468*4 4n 14 4.40t*6

2. lot *4 2.37t'? 1.604 :4 enate 4 23t*3 1.34t*3 1.35t=1 1.51t*4 a

a Fe ll 2.11t*7 8.73t*)

9 61t*6 9.95t*4 4.044*6 roast 2.948 1 1.00te?

4 95tet 2.33t*4 2.64t*f

e*g1 1.248*6 1.234*1 2.13t*4 ce*S4 4.728 4 4.l?tet L,06tet 0a 60 1.648*3 3.Det*4 3.62Eet N4*43 6.738*9 4.668*4 9.73tet 2.2b8 4

=

Nia65

3. 70t

1 4.41t*2 L.225 0 2.19t*2 2 41t*4 Cua64 4.048*4 1.054*6 1.138*4

=

in*69 1.27t*9 4.368*9 2.928*9 2.754*9 1.91t*9 ta*69 4t*42 3.128*1 3.25t+1 at*43 1.498 1 1 038 1 at*44 Staal 2.19t*9 Ra*S6 5.11t*4 1.218*9 Re se Rg*p3 stadt 1.454*9 2.338 4 4 164*7 st*90 4.648*10 1.35t*9 1.18t*10 statt 3.13t*4 8.49t*l 1.27t*3 Stat 2 4.49t*1

9. 644 *C 2.118*2 Y*to 1.07t*1 7.50tel 1.908 *0 T*tta Y*tt

4. 60t* 3 4.738*4

2. Jot *2 Y'92 1.421 5 9.49 ten L.let*6 Ya93 2.338*1

?.198 3 6.438*3 Statt 9.468*2 3.038*2 4.764 2 9.62tel 2.058*2 4:*91 4.268 1 4.39t*2 1.304*L 2.64t*4 3.938*2 Whatl 4.25t*4 4.59t*4 4.544*4 2.19t*4 2.478 4 Mb*97 l.418*9 2.12t+1 me* 99

$,72t*1 1.45t*1 4.aot*6 teat 9e 3.2 $1 *C 9.19E *0 1.40t*2

4. 50t *0 3.44t*3 1.17t*2 74*101 eu*103 1.025*3 3.494*3 1.198'l 4.39E*2 Au*10$

4.178 4 1.118 2 5.244*1 3.34t*4 tua104 2.048*4 3+948*4 4.328*4 2.54t*3

.RA*103e gnat 06 64*l10s 1.438*7 $.J9t*1 1.064*4 2.20t*l0 3.20t*7

+

sn*124 2.37te? 4.664*$ 4.248*4 2.QCE*? 7.31t*4 1 028*1 se t25 2.048*1 2.264*$ 2.048*4 1.54t*1 2.258 4 4.664 *6 te*L2Se 4.638*1 S.904*6 4.90E*4 4.638*1

6. 50t

7 2.188 *6 feel 21s

4. ltt e ?

1,644*f 1.17E*7 1.464*4 1.144*4 S.54t*6 te*t27 6.728*2 2.41t*2 4.948*2 2.748*3 1.304*4 1.45t*2 to 129e

4. 04t * ? 2.2SE*f 2.048*7 2.528*4 3.048*4 9.l?t*6 te*L29 Te* L 3 La 3.6 LS*5 1.77E*l

2. Dot *5 1.79t *4 1.718*1 1.471*S te*g3g fo t3J

2. 39t >4 1.158 6 1.7 tt *6 1.494*7 '

7.32t*? 1.458*4

!*130 4.264*5 1.268*4 1.07t*4 1.964 6 1.044*4 4,944*$

!*131 2.968*4 4.24t*4 1.39t*11 7.27t*4 1.123*4 2.432*4 a

!*132 L. 648 1 4.378 1 1.134*1 6.97t a l 4.241 2 1.138*1

!.L33 3.9 7t *4 6.90S*6 1.01t*9 1.20467 6.208*4 J.10S *6

!* L14 t 135 1.398 4 3.638 4 2.604 *4 S.438 4 4.10t*4 1.344 4 Co 134 S.454*9 1.348*10 4.358*9 1.644*9 2.35t*4 1.10E*10 Ca*136 2.618*4 1.038*9 1.748*6 7.47t*7 1.178*4 f. 428 *4 Ca*L37

7. 388 *9 1.0 La elo 3.438 *9 1.148*9 L.958*4 6.418 9 Ca*L34 B4 139 4.70s 4 4.34t*4 1.344 9 Seet60 2.698*7 3.348,4 1.154*4 1.938 4 $ 54t*7 1.764*4 se*Let teag42 Leal 40 4.49t*C 2.268*0 1.664 5 5.97ta!

La.142 3.038 4 Ce*141 4.648*3 3.2?t*3 1.124*3 1.25t*7 3.71t*2 a

Co*L43 4.19t*L 3.09t*4 1.344*1 1.168*4 3.423*0 Ce=144 3.14tel 1.50tel 4.47t*4 1.21t*4 1.928*4 Frat 43 1.193*2 4.37E 1 3.648'L

6. 964 *$

F.44t >0 Pr*L44 m4*t47 9.428*4

1. 09t

2 6.378*1 5.23t+S 6.*23*0 eatet 6.564*3 S.44t*3 g,gog*4 g,9,28 3 1.9 wp.239 3.664 *o

3. b08 1 n.424 0 7, j,3g,4 gg Davis-Besse ODCM 63 Revision 4, 1991

Table 3-7-(continued)

R(10)r Grtss - Cow - Milk Pathway Dos *) r:ctcrs - T12NAGE:R (mJouVyr per uCi/13) for N-3 and C-14 (m a arenvyr per uCi/sec) for othere

% glade tone I.1 vet Thytote Elhof Wag 03 4 &2 f.amog

9. us

2 9.94:*2 9.md

9. m *2 9.us

2

9. m.*2 c.it -

6. = *i i. as el

.auel-i.nad i. m el 1.24 *i

1. m oi wead4 4.448*6 4.444*6 4.444*6 4.444 4 4.444*4 4.448*6 4.440*4 P.12.

3.19t*10 1.958*9 2.664*9 1.228*9 itet) 2.784*4 1.104*4 7.138*4 g,40get 8,00t*4 sin al4 4.408'?

4+47t**

2.87t*7 2.788*4 gn*g6 7.518 4 9.508 4 4.94t=1 1.334*3 to.45 4.45t*7 3.168*7 2.00t*7 3.37te?

1. 3H + 4 fe-49 1.204 *7 1.21t*4 3.8Jt

7 2.87t*4

4. 648 + f

a*l1 2.458 6 4.19sef 17764*4 Ca.it 7.954 4 1.10t*0 1.8 34 *7 i

ce40 2.78t'?

3.628*0 4,268*?

i m143 L.18t*10 4.JS8 *4 -

4.J38*4 4.018 4 N1*45 6.78tal 4.H84' 4.708*0 3.944*4 l

Cu 64 4.39t*4 1.09t*l

=

3.334*6 2.028*4 2A 6,1 2.11t*9 7.31t*9 4.644*9 3,304*9 3 448*9 j

w.6 6t*02 l.6e4*7 8t*43 4.

1.91t*4 St*S4 4t*05

-4.138*9 as*e6 7.00t*4 2.22889 an 48 m3 39 st*e9 2.47t*9 3.148 4 7.664*7 stato 6.418*10 L.464*9 1.638*10 stat!

5.75t*4 2.41t*l 2.2M 4 st*92 8.958*4 ~

2. 20E 4 3.e 1a

2 f.90 4.308*3 4.078*4 3.lca 0 Y 91a y*tt -

1. lM *4 6.448*4 4.245*2 V'93 1.00E*4 -

3.758*0 2.90s.4 y*tl 4.308 1 4.3 13 *4 1.100*2 2t*95

-1.658*3 1.22t 4 7.678*2 1.20see 3.59t*2 st*97 7.758 1 4.83t*1 2.12 tat 4.15s+4 7,068*2

-~ -.

Nuott 1,414el

7. tot *4 7.578 4 3,343 4 4.30g*4 mm*97 4.344 4 4.548,7 ne*M 1.048*4 8.168,7 - 5.69t *4 Tc.99m

-5.64t*0 -1.57t 4 2.344 4 e. 734 *0 4.034*4 2.044*2 i

74*111 8v*LQ3

!.Siti) 4.408 4 '

1.528+l 7.75t*2 as*109 4,575 4 1.97t*2 4 264 0 6.00s.4

=

au*LQ6 3.758*4 7.238*4 1.00E*4 4.738*)

3A*103m Ap* lot -

A9antom 9.634*? 9.llta?

4.744*8 2.548*10 1.544*7

- I statie 4.lM*7 4.468'l 1.042*$

4.01g*7 9.254*4 1.79E*?

- st*t21 - 3.6M*?

3. 99E * $ 3.498*4 3.21t*7 2.e44 *4 - 3,548*4

+

7e* t25m - 3.00E*7 1.008*7' 4. 39E *4 8.048*7 4.028*4 fe.127a 4.448*7 2.ME* 7 3.018*7 3.424*4 2.10g*4. 8,00g*?

tean27. 1.248*3 4.415*2 8.598*2 S.044*3 9.618 *4

2. 68E

2 fe.129m 1.118*4 4.10E*7.3.87t*7 4.625's

' 4, gM.4 3.7p3 7 to.129.

1.67t*9 2,13g*9 Te 131s 6.571*$ 3.118 4 4.744 4 3.298*4

2. $ 35 *?

2.634*l 7e 134

=

Teal 32 4.288*4 - 2.715 4 2.448*4

2. 60E * ?

4.lete?

2. S$8 *4 3 130 --

7,498*$ 2.17t*4 1.77t*4 J.344 4 4.673 4 8.448*$

I*131

$. 3854 7.5384 2.20Est! 1.30E*9 4.498 4 4.048*e L

tal32.

.2.908 1

7. 5M

1 2.544*1 1.20E*0

'3.31s*1 2.723 1 3*t33 7.244*4 1.238*7 1.728*9 2.154*?

9. J08 +4 3.758*4 g

gl4

!*l35

2. 4 7E *4 ' 4.3 38 *4 4.0es*4 1.008 l 7.0 38 *4 2.358*4 Co.134

'9.41E*9 2.315*10 7.344.'9 2.tos *9 2.07t*4 1.07t*10-ca*434 4.458*4 1.758 9 9.138*4 1.50E*4 1.4 La *4 1.188*9 Ca*L37.1.344*10 1.70s*10 t

4.048*9 2.358 *9 2.33t*4 4.20E *9 c4 134 se*139 4.49t*8 7.758*1 2.538*9 ne*L40 4.858*7 $.958*4 2.028*4 4.00t*4 7.49e*7 3.133*4 Se.14%

g *342 1,4*440 4.048*0 1.964=0 2.27tel 1.05t*0 LA *142 2.238*7 ce*L41 4.87t*3 5.922 4 2.798 4 1.69t*7 4.818 4 co.443.

7.69tel 5.60t*4 2.ll4*L

+

4.648*4 6.258*0 t

ce*144 4.548*$ 2.72t*$

1.638*$

1. Ha *4 3.544*4 1

l Pr*l43 2.928 4 1.478 4 L 778+l 9.618*$ l.4 58

L pt*144 ree*n47 1.818*2 4.978 4 1.164 4 1.113*$

1. ist 4 w te)

1. 20E *4 9,7M4 2.654 4 3.434*3

- o*239 6.998 0 4.lH a

2.07bo 1.044*l 3.664=1

Table 3-7 (continued)

R(ic), Grass - Cow - Milk P;thway Dose Factors - 0111D

+

(m{esvyr per vC1/m.1) for H-3 and C-14 (m e area /yr per uC1/sec) for others nuo ude Sea.

uur env.r.ie llen.,

w 0:* u.2 t.8.er e.....*

.*3 1.iit*3

t. lite) 1.lts*3 1.57t*

i.s7t*3 i,ltt*3

- le 1.658*4 3.29t'l 3.298+l 3.19E*l 3.29tel 3.29t*4 3.294*l me*25 9,238 6 9.231 6 9.238*6 9.338*6 9.238*6 9.238*4 9.235*4 Pa12 7.??t*10 3.64t*9 2.15t*9

3. 00t *9 1.66t*4 1.558*4 1,038*5 p,41t*4 1.028'l croit 2.( Jef 5.47t*6 en 14 1.768*1 1.544*4

1. 3 - 1* J L. Set *J enal6

1. 90t *0 2.9't*'

.?

re*$1 L.128*8 5.9

3. 2 5t * ?

8. 108 *1 1.848*f fe*lt 4.201 0 1.9. '8 1,618+f 2.03t*G 1.71887 Co*lf 3.444 3.148*? 1.778 4 a

Ze.64 4.28t*3 1.04t*f 3.72tof 4.328*1 Zs'oc 2.39t*8 1.21t*4 N6*63 2.96t*10

1. 5 9t *9 1.07t *4 1.01t*9 N6 64 1.664*0 1,16 tan L,9tte 9.188 2 7.lll*4 1 821*5 Cu*44 3.34t*4 4.$64*4 taael 4.138*9 L.108*10 6.944*9 1.934*9 6.854*9 ta*49 2.148*9 Or*e2 1.158 4 Gree) 4.69tal trat4

=

Eg.85 8.77t*9 LD*64 l.64t*4 5.J98*#

La*64 i

Ra e9 se'et 6.628*9 2.56t*8 1.89t*4 trato 1.122*11 L.Intet 2.438*10 trati 1.415*1 3.128 5 5.338*3 sr*92 2.19t *C 4.14t+1 4.164*2 y*90 3.228*2 9.158 6 8.418 0 yatta a

j y*tt 3.91t*4 5.21t*4 1.J4t

3 y*92 2.464*4 7.104 *O

?.038*4 y*13 1.045 0 1.57t*4 2.90t*2 tr*99 3.848*) 8.488*2 1.21t*3 4.4Llet 1.824*2 tr*91 1.89t*0 2.728*1 3.915*L 4.138*4 1.61t*1 up*11 3.188 5 1.J48+l 1.164*l 2.29ted 8.644 +4 wn*97 1.458*6 Mo*99 8.J9t*1 1.??t*8 6.84t*7 2.0$4 *1 tc*99u 1.29tel 2.54t*1 3.64t*2 1.29t*1 1.44t*4

4. 20L.2 Tc=104 Ruan03 4.29t*3 1.088 *4 t.11t*l L.45t*3 Av 401 3.828*3 3.363*2 2.49f *C 1.39t*3 av.104 9.244*4 1 258'l 1.448 4 1.11t*4 na

103m kn*104 ngattom 2.09t*4 1.41t*4 2.438*8 1.644+10 1.138*8 54*t24 L.09t*4 1.41E*4 2.40tel 6.038*7 4.79E*4 3.818*7 Sb'129 8.708'?

1.412*6 8.068*4 4.854*7 J.048*4 1.828*1 fe* L2 5e 7.348*1 2.00t*1 2.07t*f 7.12t*1 9.644*4 te*127m 2.068*4 5.604*? 4.97t*7 5.932*8 1.648 4 2.478+?

te*L27 3.064*3 4.238 2 2.123*3 8.11E*)

1.208*$ 4.54t*2 te*129m 2.728 4 7.61S*7 8.788*7 0.00t*8 3.323 4 4.232*1 2.87t*9 To.129 6.122*8 Te* 13 La 1.60t*4 S.53E+S 1.144*4

5. 35t *4 2.244*7 5.998 5 te*L31 TeatI2 1.023*? 4.928*4 4.188*4 4.208'?

4.55t*7 S.464*4

(*130 1.755*4 3.54t *6

3. Fot *4 1.198 6 1.664 4 1.828 4

(*l31

1. 30t *9 1.31E*9 4.342+11 2.138*9 1.17t*4 7.464*4

!al32 6.848*1

1. 2 64 *C 5.858*1 1.938 0 1.448 *0 S.808 1 3+133 L.764*7 J.188*? 4.044*9 3.638*1 8.778 4 8.238*6

(*134

!*t39 5.644*4 1.058*l

9. 30E *4 1 611*$

e,004*4

4. 97t *4 Cael34 2.264*10 3.718*l0 1.114*10 4.132 9

2. 00t *4 f.838 9 Ca*L36

1. 00t *9 2.7 64 *9 1.47t*9 2.19teg 9.70t*7

1. 79t *9 C4 137 3.213*L0 J.098*10 1.0 LE

LO 3.628 9 1.932 4 4.55t*9 C4

134

&a.139 2.144*7 1.232*l 4.19E*9 Sa ito 1 17t*4 1.035*l 3.J44*4 4.123*4 1,94t*1 6 e48*4 64 141 Sa*142 La*140 1.934*4 6.748 0 1.64t*$ 2.278 0 La*142

2. 8 L8 *4 Ce*141 2.19t*4 1.09t*4 4.788*3 1.364*? 1.628*3 Ce*L43 1.898 2 1.021*l 4.29tel

1. 50t *4 1.484 *1 Coal 44 1.622 4 1.098 5 2.823*1 L.338 4 8.664*4 Prale) 7 234'2 2.17t*2 1.17t*J 7.80t*$

3.59t*L Pr L44 me*l47 4.45t*2 3.60t*2 1.718el 2.79t*l 1.Tes*2 a

w.187 2.914*4 1.728*4 2.421*4 7.73t*)

wp*239 4.728*4

1. 2 3t =0 3.578 0 9.144*4 8.644*n Davis-Besse ODCM 65 Revision 4, 1991

Tablo 3-7 (continued)

R(io), Grcas - Cow - Milk Pathway Does rectors - INTMT (myevyrperuCi/m3) for H-3 and C-14 (m

aren/yr per uCi/sec) for others mus14ee tone f.4 ret ThTT914 EAdEST..

t, mag 02 8,I !

T.8edt.

- .a.

2.384 3 4.38483 2.30E*3 2.388*3 2.388*3 2.388*3 w3 C 44 3.23t*4 6.89tel 6.898'l 6.89tel 4.89tel 4.894 5 6.89tet me*24 1.615*1 1.61t*1 1.61467 1.61t*1 1.64191 1.61t*7 1 61t*?

P 12 1.totet! 9.428*9 2,17t*9 6.21t*9 1.05t*5 2.30t*4 2.05tel

4. 7 t* *4 1.6tE*l

.'r

g t 3.89t*1 anale 8.63t*6 4.43tet 8.838*4 a

=

1.21t*2

'in*64 2.764*2 2,918 *0 S.538*3

+

Fe.gl 1.21t*8 8.72t*7 4.27t*1 1 1LE*? 2.33te?

r elt 2.25t*8 3.93t*8 1.164*8 1.844*8 1.51t*4 e

8.918*6

!a*ll 3.058*1

1. 46t

7 2.43t*7 Oeal8 6.058+f 6.Mt

7 8.41t*7 00 60 2.10t*8 2.08t*8 us.43 3.498 10 2.164*9 1.07B*8 6.21t*9 at*65 3.11t*0 3.97t*1 3.028*1 1.818'l 1.84tel 3.17tel cua64 3.81t*4 8.69t *4

=

Imael S.55t*9 1.90t+10 9.23t*9 1.6ttono 4.788*9 Za*49 7.36t.9 Gr*02 1,944*8 8g.8) 9.95t*L 4t*84 gr.8g 2.228610 Rbete

$.698 8 1.10t*LO R8 48 R3 89 Br*49 1.26t*10 J.99tes 3.61t*8 Steto 1.228*tt 1.528 9 3.102+10 tr 91 2.54E+l 3.448*l 1.064*4 Sr*92 4.614*C 5.015*1 1.?3t*1 Y.90 6.80t*2 9.398+l 1.82t+1

=

a Ya9 La a

=

y*g1 7.33t*4 5.264*4 1.958*3 Y'92 S.221 4

9. 9 7t *0 1.4?tal Ya93 2.25t*0 1.788*4 6.132 2 3r*95 4.83t*3 1.664 1 1.79t*3 8.284*l 1.184*3 Ir*97 3.99t*0 6.8SE*1 4.91**L 4.37t*4 3 138 1 me*95 S.931*l 2.444*l 1.75tel 2.c44*8 1.418*5 ht.91 3.708 4 2.128*8 no*99

,2.69t*1 1.llt'l 2.17t*4 6.984*? 4.13tet

+

Tc 99e 5.978 2 2.90t*1 1.61**4

?.154*2 Ts*Lol

=

3 8u 103 8.69t*3 1.81E*4 1.044*l 2.91S*3 Gu*tOS 8.04t*3 5.928 2 3.21t*0 2.718 3 Gual06 1.90tel 2.25tel L.448 4 2.388 4 8A*LQ3a

+-

an.goe 44*n10s 3.46t*4 2.82t*4 4.038*4 1.444*10 1.844*4 sa.424 2.098 4 3.068*6 1.564*5

1. 31E *4 6.444*4 4.49t*1 e

sa.L29 1.4*t*8 1.454*6 1.87tel 9.388*? 1.99t*4 3.07E*?

=

te*L23e 1.11t*4 S.04t=1 5.07E*?

7.18Ee? 2.044*7 to 127s 4.218*4

1. 40E *4 4.222*4 1.04 t +9

1. 70E *4 S.10t*7 fe.127 6.504*3 2.18t*3 5.29E*3 1.19t*4 1.364 5 3 40t*3

=

to 129e S.lttet 1.928 4 2.158*4 1.40t*9 3.344*4 8.6JE*7 Teal 29 2.088*9 1.754*9 5.168*9 1.644 7 To.131a 3.34t*4 1.364*4 2.765 4 9.318*4 2.298 7 1.125 *4 Te*131 To*132 2.10s*? 1.04t*? 1.544*7 6.51t*7 3.81s*7 9.723 6

!.130 3.60E*4 7.928*4 4.888*4 8.70E*4 1,70E *4 3.18E 4 1 131 2.723*9

3. 21t *9 1.0S4

12 3.75t*9 1.133*4 1.41**9

!.L32 1.423*0 2.89t*0 1.354 2 3.228*0 2.344*O 1.03t*0

!.433 3.728*7 5.41E*1 9.644*9 4.344*?

9.168 4 1.944*7

!.134 1.0La*9 t*435 1.212 5 2.415*5 2.164*? 2.69tel 4.144*4

8. 80E *4 Ca

13 3.65t*10 4.00t*10 L.758*t0 7.184*9 1.893 4

4. 8 7t *9 Ca*136 1.944*9 S.??t*9

2. 30E *1 4 ?Oted 8.76E*7 2.154*9 Ca.137 S.1$8*10 6.023*LO 1.622+10

6. 55 t *9

1. 88t *4

4. 2 7t *9 Ca*134 6a* LJ1 4.558.?

2.848*l L.328 4 sa.140 2.418*4

2. 4 tt el l.73t*4 1.44t*l S.923*7 1.24t*?

6a 141 ga.g42 L.a 140 4.038*L 1.ltE*L 1.87tel 4.094*0 La 142 5.21S*4 Co*L41 4.338 4 2.648*4 8.118*3 1.37t*1 ?.11t*3 Co*l43 4.00E*2 2.653*l 7.728*1 1.553*4 3.023*1 Co.644 2.J38 4 9.928*l 3.8 52 *$

1.J3t*4 1.30E*l Prat 43 1.49t*3 S.59t*J V

2.048*2 7.89t*$ 1.41t*1

,,.44.

m4*t47 8.828*2 9.068*2 3.49t*2 l.748*$ 1.llt*L u*te?

6 124*4 4.264*4 2,5cE *4 1.47t*4 a

mo*239 3.644*1 3.253 0

4. 4 9d *0 9.408 4 1.848*0 Davis-Besse ODCM 66 Revision 4, 1991

Tablo 3-7 (continued)

R(io), Grass - Ccw - Mest pat *Way Do32 rcctors - ADULT (mycm/yr per u01/m3) for H-3 and C-14 (a

mrem /yr per uci/sec) for others en7.ms ' Ele =v*.. 9...

c *uI t.***r.

.*eline sen.

uv.r

(

ual 3.J$4*2 3.25t*2 3.2$4*4 3,2$t*3 3,253*) 3.25t*2

le 3.338*$ 6.648*4 6.468*4 6.448*4 4.66t*4 6.664*4 4.668*4 me J4 1.048 3 1 848*3 1.848*3 1.444*3 1.64 Pan t.betal

1. 64 5 *)

Pa l2 4.45t*9 2 49t*4 5.238*8 L. 00t *4 Ot*$1 4 228*3 L.544*3 9.34t*3 g.7st*6 1.07B*3 mnale 9.1SE*6 2.721 4

2. 60t

7 1.75t*4 nn.54 Fe ll 2.934*4 2.02t*0 1.131*8 1.168*4 4.72t*7 fealt 2.671 9 4 27tet 4.758 3 2,098 9 2.6o4*t ce*lf 5.64t*4 1.438 8 9.37t*4 Oe'l4 1.038*7

=

3.70t*6 4.104*7 Oeaho 7.128 7 4.41t*9 1.668 4 m1*63 1.49t*10 1.38t*9 2.73 tat 6.338*4 ut*gg Cu.64 2.95t*7 7.45t*7 2.52t*1 1.395 7

!a*45 3.96t*4 1.138 9 7.l?t*e ga*eg 1.13:*4 $.11t*6 Sr*02 Statl 1.448*3 1.268*3 Gr*04 Sr*49 R4*64 4.87t*8 9.608*7 2.278*4 RD*64 La*e9 S:*49 3.0!!*8 4.44t*1 6.65t*4 Sr*to 1.24t*10 3.598 4 3.01t*9 Statt tr*92 1.J88*9 y*to 1.078*2 1.138*4 2 848*0 y*9 La y*91 1,138 6 6.244 6 3.034 4 y*92 y*t) 2.048 1 st*95 1.44t*6

6. 04 t * $

9.46t*1 1.91t*9 4.09tel St*97 1.838*$ 3.698 4 S.541 6 1 148 0 1.498 4 Mb*99 2,29t*4 %.Ill*6 1.268*4 7.75t*$ 6.e48*1 un*97 me*99 1.09t*S 2.448*1 Tc*tte 2.528*$ 2.07t*4

+

gg. tog Rw.103 1.06t*0 a

4.038*8 1.231*n0 4.llt*?

pu.LQ5 Ru 106 2.bCa*9 5.404*9 1.81t+11 3.548 8 Rh*103e Rh*104 6g* 110m

6. 69t *4 4.19t*6 1.228*7 2 122*$ 3.67t*4 ab.124 1.904*7 3.74t*1 4.60t*4 1.14te? S.621*6 7.838 4

$b*l25 1.918*7 2.131*$ L 94t*4 t.47to? 2.totet 4.544*4 te=125e 3.598 4 L.Jotes 1.04 tee 1.444*9 t.431 9 4.stt*7 teal 27a 1.128*9 3 99tet 2.818*8 4.53t*9 3.748*9 1.J64*4 fe*127 1.09t*9 2,105 4 Te*L29m 1.148*9 4 27E*8 3.938*8 4 77t*9 5.764*9 1.818*4 te*129 Te* t3 La 4.11262 2.212*2 3.50t*2 2.248*3 2.19t*4 1.848*2 to.g31 Teal 32 1.404*4 9.078 5

1. 00E *4 8.738*4 4.294 7 4.lltel t L30 2.318 4 6.944*4 1.648 4 1.044*S 3.94g.6 2.748*6

(*131 1.048*7

1. 54 t

7 S.054 *9

2. 64t

7 4.073 4 8.838 4 g.g32 tal33 4.30tal 7.47t*1 1.104*2

1. 30E *0 6.728 1 2.28tal

!*134 2*IJS Ca*134

4. 3 7E *4 1.$64 9 5.048*4 1.644*4 2.748 7 1.288*9 Cael36 1.144*7 4.671*?

2.604 *7 3.164 4 S. 30t *4 3.364*7 C4*137 0.728 0 1.198 9 4.053*8 L.J54*4 2.31t*7 7.81t*4 Cs*130 64*L39 Bea!40 2.648*7 3.418 4 t.231 4 2.071 4 5.92t*7 1.49E *4 Be 141 Se*142 Lael40 3.608*2 1.813*2 1.333*3 4.198 3 Le*142 ce*t41 1.40t*4 9.448 3 4.408 3 3.623*7 1.08t*3 ca.143 J.09t*2 1.55tet 6.604*3

$ 76t*3 1.71t*3 Co 144 1.448*4

6. 09t

1 3.61t*S 4.938 4 7.0j8 4 trale) 2.138*4 4.548*3 4.938*)

+

9.J3t*7 1.045*3 Pr*144 me*147 7.042 *3 4.Let*3 4.788*)

3.935 7 4.908 2 s.107 2.164 2 L.4tt'2 5.923*0 6.J28*3 wp 239 2.144 I 2.518*J 7.848*2 1.158 3 L.J98*2 I

Davis-Besse ODCM 67 Revision 4, 1991

~

Tablo 3-7 (continued)

R(io), Grass - Cow - Meat Pathway Dose Factors - TEWAGER

)

(m (a[eaWr per uCi/m3) fcr M-3 and C-14 l

mree/yr per vCi/sec) for others w 1ie.

e u,

...r....n,. tar 844,

w.,

..utt 9.n.47

.3 1.uan i.use2 1 us*2 t.uin 1.m o 1.ueo

\\

Colo 2.818'l 9.628*4 S.628*4 9.628*4 S.628*4 S.638*4 1.428*4 4

me*J4 1.478 3 1 41t*3 1.478 4 147kd 1.478 3 1.47t*3 1.478'3 p.32 3.93t*9 2.448*8

. 1.244*3 8.078 4 9.losel S.654 0 3.30E*4 1.528*0 Cr*11 3.148 0 an.84 6.94t*4 2.048 4 1 438*7 1 18t *4 g.g a

re.S$

2.333 0 1.69t.8 1.07t*4 7.30E*? 3.93487 re.59 4.t38 4 4.948 4 1.87t*4 1 188*9 4.928'8 3 57 4.53t*4 8.458 *7 7.598*4 2 54 ~

t 1.418*7 1.83t*7 4.94s's 3.254 *?

00 60 7.60tet n.348*4 n4 43 1.524*to 1.07t*9

.L.7ttet 5.11t*4 mg 65 a

g.64 2.4tt*7 6.108 4 4.878.$ t.138*7 33 45 -

2.50E*4 8.498*0 5.544*8

3. 6M *e 4.054*8

=

go.49 Br.82 9.f#8'I St.43 St.64 Sr.49 As.e4 4.04t*8

=

4.015 *1 1.912 4 an.as ng.g9 l

3r.09 2.544*8 1.034'? 7.298 4 Stato 8 058*9 2.164*4 t.99E 4 st*91 1.10s.9 St*92 Y.90 0.9 stet 7.40tel

2. 4 Jt *0 Y.9 ta Y.9n 9.544*l 3.933*4 2.564*4 y.92 Y 93 1.69t*?

3:*95 1.518*4 4.748'l

6. He e t l

gr.97 1.138 5 3.021 4 1.10g *9 3.273+1

4. H4*4 l

=

4.188 4 1.198 4 l

Nb.99 1.798 4 9.948'l 9.644*l 4.254*9 5.47t*l m.9 7 n

. n. 99

/j\\

8.90E*4 2.064*l 1.61t*$ 1.71t*4

\\

ts.99.

u.501 Au.103 0.60E*7 3.038*4

7. t38 *9 3.64861 gg.to9 ev.106 2.168*9 4.llE*9 3.133 11 2.97t*4 Aa.103m An*104

=

Agattom 3.068 *4 4.198 4 9 144*4

=

1.353*9 2.913 4 54424 1.62887 2.904*5 3.478*4 1.41t*7 3.348 *4. 6.315 4 ab 125 1.54C+7 1.74tel 1.49t*4 5.278*7 1.228*4 3.MS *4 te*L25m 3.038*4 1.09t*4 8.47t*7 0.944*4 4.054*?

To.527e 9.458*4 1.344*4 3.24a 4 3.428*9.

2.3 58 *9 4.528 4 Te=127 1.754 4 te.129m 9.548*4 3.544 *4 3.004 4 4.01E *9

3. 60t *9 4 523*4 fe*129 To.131a 3.744*2 -1.80S*2 2.714*2 1.888*3 1 438 4 1.lcse2 to.131 To.132 4.tl4*4 7.244 *l 7.448*1

4. 97t *4 2.30E*1 6.848*l

!.tJO 1.898*4 5.44E*4 4.47t*4 8.448*4 4.215 4 2.198 4

!.131 8.994*4 1.254*7 3.644*9 2.164*7 2.4es*4 6.738*4 2 132 1 433 31984 4.10E 4 8.Sitet 1.072'0 4.618*1 1.044*1 g.134 2 135 Ca 134 5.238*4-4.23t*9 3.918*4

1. 49t *4 1.53E*1 5.718*4 Ca

IJ4 -

9.228*4 3.634*7 1.97t*7 2.11t*4 2.923*4 2.444*7 Ca.137 7.244*4 - 9.634 4 3.245 4 1.273*4 1.3 73 *7

3. J48 *4

.ca.134 i

te.139 es.140 2.348*7 2.91t*4 9.888*3 1.944*4 3.67t*7 1.938*4

3. 141 go.g42 t.a.140 2.144*2 1.455 2 4,354e2 3.478 3 La.142 Cadet 1.188*4 7.844 0 3.70E*3 2.238*7 9.038*2

-Co.143 1.764 2 1.208*1 l.748*)

3.elg*3 1.434 3 Co 144 1.238 4 S.004*5 3.044*l 3.093 4 4.60E*4 k

Pr.143 1.79t*4 7.158*3 4.144*3 b-5.90E*7 8.925*2 pg.144 Nei.147 4.244*3

4. 79 t 4 1.988 4 2.4 4.068 3 3,9,5s*7

. 187 1.813 2 1.448 3 g *0 5.178 3 mp.239 2.234 4 2.111 2 e.418 2 1,39ael 1.t75 3 Davis-Besse CDCM 68 Revision 4, 1991

bg Table 3-7 (continued)

R(io), Grc:s - Cow - Meat Pathway Dose Factors - CHM (mrenVyr per uCi/M3) for H-3 and C-14 (m2 a mrevyr per aci/sec) for others

. 7....

I4 hey..

mellde tone Livet Th 7914 temq GI+1LI f.4ety n=3 2.348*2 2.345*2 2.34882 3.34t*2 2.34tel J.348*2 C+44 S.29tel

1. 06t *1 1.064*5 1.068+l 1.06tel 4 068*5 1.064*l me.24 2.348 3 2.348a4 2 348 3 2.348 3 2.348 3 2,345 3 3.348 8 Pan!

?.413*9 3.47t*4 2 058*4 2.464*4 Crell 4.89t*3 1.348*) 8.93t*3 4.478+l 4.818*3 a

an.64 7.99t*4 2.241*6 4.108*6 2.138*4 nnate re.lt 4.13 1 2.42t*4 8.3ttee 4.49t*f

?.518*1 reelt 3.748 4 6.12t *8 1.77t*8 4.37t*4 3.058*4 to.51 8.923*6 4.458*7 1.20t**

Co $4 L.65t*?

9. bot

f 5.044*1 6.93t*?

Ce*60

=

3.348+g 2.048*4 N4*43 2.91t*10 1.164*9 t.038*4 9.918*4 mg.48 Cu 64 3.248*1 7.428 7

=

L.328 5 1.964*?

ta'el 3.79t*4 1.00t*9 6.30t*e 1.768*4 6.228*4 In*69 8t*42 1.564*3 sg.43 Br*64 8t*0$

5.748*4 tp* e4 3.71tet 3.848*4 no.e4 Ra*49 St=49 4.828*8 1.464*f 1.348*1 Stato 1.048*10 1.40t*4

2. 64 8 *9 Sr*96 1.0Lt*9 Stata Yato L.708*2

=

4.448*$ 4.658 0 y.9 La Y 91 1.814*4 2.41**4 4.438*4 y.92 Y.9) 1.858 7 traft 2.648*4 5.49t*l 8.438+l 6.148*4 S.244*l tr*97 2.84Eal 4.104 4 S.49t*4 6.214*1 2.438*6 Mba95 3.098*4 1.204 *4 1.138*4 2.238*9 8.618*$

Mb.97 1.25tel me.99 2.6ftel

=

1.038*l 3.09t*4 gg.99m Ts*101 Au 103 1.868*4 3.128 4 4.02t>9 5.94t*1 ea*10$

Ru = 104.

4.448*9 5.994*9 6.90t*l0 1.548*4 gn.10 3m Rh*104 49 110m 0.40t*4 S.671 4 1.064*7 4.714*4 4.534*4

+

Sa*124 2.938*7 3.00S*l 4.465 *4 1.628*7 1.438*4 1.0JE*7 sh*125 2.81t*7 2.19t*1 2.644*4 1.594+7 6.008*7

5. t68 *4

+

fe*125e 1.49t*4

1. 54 8 *4 1.60t*4 5.464*4

7. 59t e t to 12?m 1.778*9 4.788 *4 4.2 68 *4 1.064*9 1.464 9 2.11t*4 te*121 1.213 9 1.664 0 Te*L29m 1.81t*4 5.044*4 5.82864 1.308*9 2.20t*9 2.408*4 to*121 to* L3 La 7.00t*2 2.41E *2 4.948*2 2.348*3 9.818*3 2.548*2 Te=131 Teal 3J 2.09t*4 9.27tel 1.358*4 8.40S*4 9.33R*6 1.128 4 t=130 3.39t*4 4.858 6 7.548 4 1.028*l 3.208 6 3.534*4 1 131 1.644*? 1.478*7 5.$23 *9 2.748*1

1. 4 9t *4 9.498 4

+

g.132 ta133 4.648*1 9.244 1 1.538*2 1.388 0 3.333 1 3.128 1 g.g34

!*135 Ce=134 9.222*4 1.818*4

4. 4 9E *4 1.648*4 0.154 4 3.19t*4 Coal 34 1.8 94 *1 4.37E*7 2.338*? 3.478*4 1.544 *4 2.838*7 Ca*137 1.33869 1.248*9 4.164*4 1.50t*4 f.99t*6 1 048*4 Ce*L34 n.*139 Se*160 4.39t*7 3.855*4 1.21E*4 2.29t*4 2.221*7 2.544*4 Se 141 ge.g42 Le*160 S.412*2 1.ttta!

5.27t+2 6.348*3 La.142 04 141 2.228*4 1.11E*4 4.445*3 1.348*1

1. 644*3 Co.143 3.308 2 L.79t*1 1.814*3 2.622*2 2.594*3

+

Ce*tte 2.322 *4 7.264*$

4.028*1 1.89t*4 1.248*1

=

Pr=443 3.398 4 1.022 4 S.$15*3 3.648+f 1.648*3 Fral64 m4*tet 1.17t *4 9.448*3 S.204*3 1.504*7 7.34t*2 w 687 3.3 64 *2 L.99t*2 2.19t *0 4.928 3 l

um.s**

4 70E*l 3.023*2 4.718 2 2.232+3 2.133*2 1

Davis-Besse ODCM 69 Revision 4, 1991

Tablo 3-7 (continued)

R Vegetation PathWjiy Dose Factors - ADULT l

,(a[es/yr per pC1/*a ) f r S3 and C-14 (a e aresVyr per vC1/sec) for others wuie naae to.t tavrou simet i

i si'u.: f. 0*r

(~

.*)

4. at

3

.2nd s.3ud s.2u'i

3. lues s.2ud e.it e.in st ta ian*e sand la9

tanei

5. in d

2. >u *i s.,n e t s

..

n uel sauel tsa d lau'l l saud s ou's b4 4 40t*9 4.?)tet

4. geg e W 4ttet etall 4.799*4 1.038*4 4 49t*4 g.gttet

4. H4 *4 a

ten 44 J.ittet 9.ift*1 9.64t*4 S.945

an*te 4.6 8t e l 2.045 4

1. t at o

3. 0 54 *0 to

H 2.Det*6 1.458 4 8.068*1 0.29tet 3.3?t4 fe*99 1.2?t*4 8.Ht *4 8.J14*)

9. Ma te 4.448*4 coalf 1.lftet 3.978 4 4.998+f re*te 3.09t*f

6. Mi'8 4.928st

e.60 1.67t*8 a

3.Let*9 3.694*4 m.43

1. 04 t e t 0 7.218*4

4. lot *4 3.494*8 N6 68 6 1884

1. 99t *0 3.038*3 3.664*0 cw*64 9.27t*3 3.348*4 7.90e*5 4.354*3 In49 3 19t*8 1.018 4 6.188*0

6. 3H +4 4.868 *4 1n49 0.758 4 46784 1.094*l 2.81t*6 L. 4 64 4 Brael 8.73t*4 1.518*4 at.4 3 4.635*0 3.818*0 gg.e4 St*45 2.19t*G u

e4

+

4.J2387 1 028*8 n.se kg.gt statt 9.968*9

t. tog *9 3.568 4 4t*90 6.058*l4 1.?ltete 1.4u44 St*96

3. Jot *l a

8. 828 *e L.294*4 st*9i 4.2?t*3 8.468*3 1 858*1 Y 90 1.338*4 1.414 *0

3. lu si Yatta 5.438 9 4.1L3 0 v*96 S.438 4 3.824*9 L.3ttel ve9t 9.018*!

4.las*4 8.633*4 Y*93 1.?ttel 4.638 4 4.404*0 Sto9t 1.194*6 3.818 4 9.97tel L.21t*9 3.886*l It*91 3.338 4 6.138 4 L.028*2 4.casef 3.000 4 ND*ft 1.428*$ I.91t*4 I. $lt *4 4.00$*4 4.31884

.b*tt 2.90E*4 f.348*?

9.564*?

3.713*3 3.644*1 Mo*99 6.288 4 1 418'?

1.44801 L.198 4

(

teat 9e 3.0u 'o 8.663*0 1.324 4 4.244*0 9.128*3 L.10S*4 teaton Re*LO) 4.006 4 1 838*f

1. 6 tt *$

l.0f t *4 pue10%

l.J984 6.968 4 3.3as*4 3.433*t Rue 104 1.91t*8 3.738 4 1.214*LO 3.445*?

u403e u.got A9 81Cas 1.064*1 9.764 4 1.928ef 3.90g *9 9 aus4 St*124 4.048 4

4. 96t 4 2.832*l

8. cat *?

3.963*9 4.418*1 Stellt 1.364 4 4.824 4 4 3954 n.068*4 8.508 4 3.254*f feattle 9.664*?

3. lot *f 3.90t*? 3 938*e 3.M4 *e 1.294*f te427e 3.496*8 1.218*4 6.938ef 3 424*9 4.3?t*9 4.2hof teen 31 1.764 4 3.01888 4.2?id 3 358*4 4.848*l 1.254 4 3.l'd*4 9.60tst 4.fttet 1.068*9 te* lite a

4.20s*9 4.038st fe429 6.4 14 *4

2. 50E *4 1.108*4 2 794*3 6.028 4 4 628*4 feel 3La 9.125*l 4.444*l 1.048 4 48244 4.434*? 3.738 4 te*134 te=132 4.298 4 3.ftid 3.068 4 3 671**

1.31t*3 3.eas *6 1430 3.944 4 4.!?t4 9.90E*f 4.625 4 1.014 4 4.618 4

!*L38 4.098 4 4.168*8 3.?tt 4 0 5 948*4 3.054*? 6,638*?

!*02 S.144*L 1.648 4 S.344*3 2.484 2.99t e n 3 348 8 8 1433 2.128 4 3.698 4 1.433 4 6.644 4 3.3 tg 4 16244

(* n4 1.064*4

3. HE *4 S.004 3 4.594*4 3.5Lt*f 1.038*4 2*t,1 4.00t *4 4.07tel 1.048*4 1.718 4 t.3ttel 3.944*4 Cs434 4.6444 4 11t*l0 3.594*9 1.19t*9 1.S44 4 9.078 4 ca

L M 4.20E*?

1. H4 *4 9.24tet 1.3?te? 4.364 4 1.19t*4 Caen37 6.364 4 4.708*9 2.958*9 9.018*4 1.648*4

5. 70E *9 ca.130 64

H9 2.958 3 3.102 4 1 944*l L.1984 9.238*3 8.644 4 Seal 40

1. 2H +4 L.624*8 9.49t*4 9.288*4 3.654 4 9.438 4 beel41 g6*l4J 1440 1.9?to 9.924*4 7.2sget 3.62868 1.4 141 1.408*4 6.318 4 4.643 1 4.888 9 e'len

1. 944 *l 1.33tel a

6 1?t*4 5.0e4 4 1.814*4 Ce

lO

1. cot o 7.42t+l 32644 2.77tet 4.215 4 ce.144 3.29t*f 4.34t*?

6.164 4 4.114*le 1.7784 Prel43 6.344*4 2.648*4

~

&.47t*4 2.784*4 3.148 4 pt*ges m4*14?

3.348*4 1.64t*4 8.28864 1.08t*4 2.215 4 molt?

3.428*4 1.19t*4 1.09tet 1.123*4 l

.p*239 1.428 4 6.408 4 4.37t*2 2.478 4 ?.?!8 4 l

Davis-Besse ODCM 70 Revision 4, 1991

Table 34 (continued)

R,,, V093tation Pathway Doss Factors - TEDEER (areW yr per uci/m3) for H-3 and C-14 (m2

mrenVyr per uCifsec) for Mhere ana416de tone 1.49er

?> trend tJewt Imag 08 01.3

f. Her w.)

3.898 4 3.198 4 3.898 4 3.694 4 3.998e3 3.6Mel C44 1.494 4 3.9108% 3.118+l 3.918+l 3.918el 3.918 4 3.918 4 me*34 4.463 4 3.468'l I.468 4 3.464*l 3.46tel 3.45tel 3.464*6 g,35t*8 6.33889 P*ll s.6Lt+9 9 : 968 H 3.444*4 1.364*4 4.06t*4 g 04tet 6.J0t*4 4 tall 4.834*0 1.35t*4 9.3ttog e.9?t*9 unde 1.638 4 eindt l.414 4 9.54t*3 3.564 4 Fedl 3.368 4 3.318 4 4.468*4

9. Hast 6.388*1 Peat 1.818*8 4.338*4 6.334*4 9.948 4 1.438 4 4.79t*f 3.348*e 3.00869 Co*it 4.368*t coelt 6.048*0 1.01t*4 3.49t*0 3.345*9 9.60tet Co*60 e443 8.61t*10 1.138*9

!.81888 li458 8 ut*65 S.738 4 1.33460 3.978 4 3.338*0 3.13t*4 e. 404 4 6.814*l 3.954*3 Cwe64 eneet 4.348*4 1.478 4 9.418*8 6.338 8 6.Ma *4 ta49 4 4964 1.l64*l L.03tel 3 tet*l 4.098 4 1.338 4 Breel 3.018 4 ag*g3 Sr*44 ggegl 3.738*0 4.088 1 1.388*8 as.e4 an.es kb*89 1.00E*9 4.338 4 3r*09 1.81t*10 3.11t* LO L.88344 tr**0 9.81848 6.368*6 1.19t*4 trett 3.998 9 L.0ttet 1.698 4 trott 3.97t*3 n 038+4 3.348*3 V*to 1.348*4 t* t ta 1.438*9 3.668 7 3.334 9 3.11tet t*tt 1.018 4 3,338 4 3.458*3 tela e,4t 0 t*93 16384 4.948*4 4.478 0 4.01t*$

1.37t*9 3.768*l tr*99 4.748 4 1.49tel Statt 3.C98*3 6.1 Lt

4 9.365 4 4.654*1 3.048 4 14 95 1.93886 1.068 4 1.038'l

4. llbe 8.648*4 1.99t*3 3.448*1 1.90t*?

pert?

3.698 4 6.47E*1 1.318'?

L.o3t+f 1 0944 8.148*6 ano*99 L.1384 4.19t*C 4.9HO 9.79t*4 ts*99s 3.10t*0 1.54t*4 Ts*101 3.438*1 5.tet*e 2.948 4 peano) 69784 4.31t*3 4.04t*4 1.948'l tue105 S.000+1 1.448*10 3.90t*1 9.978 0 twe06 3.09t*8 hh*103e Rh*104 3.14tet 4.045 4 0.148 4 69.LaQs 4.538*? 4.444*?

1.358 0 3.114*9 6.033*?

3e.634 1.558*4 3.85t4

3. llt *l 84435 3.148*e 3.348*4 2.044*6 1.648*e 1.664*9 1.00tet 4.37t*e n.948*1 te* Laps 1.444*4 S.34tet 4.348*1

1. 3 7t *9 6.664*1 toon3te 4.518 4 1.944*4 1.318 4 3.348*9 e

4,19tel L.stt4 tenit 1.438 4 1.938*3 3.148 4 3.30t*4 1.368*9 9.81t*1 to 639is 3.67t*4 L.368 4 1.188*4 1.545 4 7e.139 4.338*4 3.338 4 4.458*4 3.61t*l 3.408 4 1.81t*4 3.334et 3.308*$

te*L31a 4.4*8'l 4.084*l 6.00tel 4.388 *4 te*L31 teel 33 3,90s,4 3.478 4 3.608 4 3.37t*?

t.833*1 3.334 4 7.87t*$ 4.098 4 8 130 3.548 4 L.0344 8.35t+f 1.let*4 3,13tet 5.19t*?

1*L31 1.70s.1 1.004 8 3.148*10 1.85tes l,91t*t 4.87tet

!*133 S.188*1 1.368*3 4.875 4 3.148 4 3,838 4 1.033 4 1433 1,97t *4 3.344 4 4.648*4 8 M4 *6 3.368 4 9.138 8 1434 4.995*l 3.548 4 4.248*3 4.018*4 8435 3.68864 9.448*4

6. lot *4 1.308 4 4.058*l 3.538*4 8.30t*9 3.03t4 3.0e8 4 1.148 4 Ce*L34 f.09t*9 1.678 4 0 9.198'? 1.45te? L.364*? 4.138 4 Co-L36 4.39te? 1.69E*0 4.598 4

1. 7 H +9 L.938*4

4. 6M *9 Ce437 4.01840 1.35840 Ca*L30 3.648 5 1.344 4 3,47t*g 3.0e8 4 6e*L39 3.77t*3

1. 95t 4 1.718*4 1*148+l 3.138*4 6.913 4 te nto L.388*0 1.69t*$

beel41 ge*g43 4.0 stet 3.348 4 Lael40 1.008 4 8.644 4 4.738*0 1.438 5 La*143 1.201*4 1.698 4 9.300 's 3.168*4 8.864*4 Cealen 2.83t+l 1.64tel 3.048'? 1.638 4 3.068 4 ce*443 9.37t*3 6.838 4 4.33 *10 3.838 4 Co*L44 S.378 4 3.19t*f 1.308't 1.648 4 3.348*0 3.884 4 Prel43 f.13t*4 3.648*4 Prel44 3.33t*4 1.438 4 3.364 4 e4*l41 3.634*4 3.94t*4 a

1.644*6 1.038 4

{

u*ts?

3.568 4 3.9ut*4 3.10t*f 1.248 4 4.090 4 ap439 4.308 4 4.30t*3 Davis-Besse ODCM 71 Revision 4, 1991

Tablo 3 7 (continued)

R.. Gr p* are V yr per uCi/sec) Plan 3 Pathway Dose Factors s

(m

= Anea

6., Or,en Nel Cole he*34 L.318+1 Poll Cr*ll 4.644*4 me*le g,3eg.,

Ma*le 9.058 4 fe<ll fe ll 3,983 4 Co*la 3,433 3 Ce40 3.168 40 ut*63 stati 3,litel Cbe64 6.098 9 EA*4l f.4 Stet ta*69 Greal 4,393 3 St*44 3.038*l tr'el hate 3,933 4 heet 3,393 4 h*st 3,333 6 treet 3.168*4 Sr*90 8t*91 3.194*6 8t*93 t y get Y 90 4.464e3 V*tta 1.01886 f*tt g,oss g Y*ta 4,soset Y*93 8.85t+l Etatl 3,444 4 trat?

3,64g,4 h*H 4.368+4 M** H 4,0$8 *4 te* Hu 1.333 5 te*l08 3,043 4 tu*103 n pts.g Re*LOS 4.36tel Swalce 4,313.g th*103a th*106 A4*l10B 3.4?tet fe* late

1. 9 6t *4 te*L37e

9. g 73,4 teat 37 3,00s.3 fe*139e 3.00 gel 7e 139

3. 60s +4 te* 13 ta 4.03t+4 teal 31 3.933 4 fe*133 4.333,4

!*130 6,g3g.4 3alli 1.73sef

!*133 n,343 6

!*133 3.478+4 l*134 4,4 Men 1 135 3.568*4 Ca*L34 4.758+9 Ca*134 1,4 M.e ca.137 1.044 e te Ca*L34 3.598 *4 De*L39 1.06ael he*140 3.054,1 Se*144 4, g g,4 Sa*143 4,4M e4 La*140 1,313 1 La*143 7.363,g Ce*144 4.364 1 Ce*443 3,33t+4 Ce*l44 6,958*1 Prent)

Pral44 1,333 3 M*141 3,4ce.e s*181 3.J68*6 NP*339 1.71864 Davis-Besse ODCM 73 Revision 4, 1991

8 lL

?y LNW DAVIS-BESSE NUCLEAR POWER STATION

[

8 ATMOSPHERIC RADIDACTIVE RELEASE PATHWAYS

==::: r,.

i

~

g H

gi -- -

=

O~

{

~ '"

tt

@d 0-E v=,

__t

- -..w 1

. = = - -

g 7

i y

l

--. ~. g g

=

T T

g _.

g 1

T T

-lFICFIFl 3

1 m

d

,2,

'12.

[F] _.,

M*k i

[f]-.

I

~

=*

{*.}

,_ T @.-

R.'-

.. -u E

I_-

~ -7[F] f] _-i.H~xf__ " "

3_

- r.

n

.H.

-i p

=

M -'*t E

')}))) se -@

2 7- ;--

E E

]

L-

'c.,

]

0 4,.,

l_,

F i

!N-d' sa

p. 4 p~ga.-

g g

g

.: = _

.

-=-

5 w

o n

g os se si er o e at: ams e FOR INFORMAIION ONLY G

O O

O.

4 4.0 SPECIAL DOSE ANALYSES 4.1 Doses To Public Due To Activities Inside the SITE BOUNDARY In accordance with Technical Specification 6.9.1.11 and ODCH Section 7.2, the Semiannual Effluent and Vaste Disposal Report submitted within 60 days after January 1 and July 1 of each year shall include an assessment of radiation doses from radioactive liquid and gaseous effluents to HEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.

In special instances HEMBERS OF THE PUBLIC are permitted access.to the radiological controlled area within the Davis-Besse station. Tours for the public are conducted with the assurance that no individual vill receive an appreciable dose (i.e., small fraction of the 40 CFR 190 dose standards).

The Visitor Center located inside the Davis-Besse Administration Building (DBAB) is also accessible to HEMBERS OF THE PUBLIC. Considering the frequency and duration of the visits, the resultant dose vould be a small fraction of the calculated maximum SITE BOUNDARY doses. The dose from gaseous effluents as modeled for the DBAB Visitor Center is considered the controlling factor when evaluating doses to MEMBERS OF THE PUBLIC from activities inside the SITE BOUNDARY.

For purposes of assessing the dose to HEMBERS OF THE PUBLIC in accordance with Technical Specification 6.9.1.11 and ODCH Section 7.2. the following exposure assumptions may be used Exposure time for maximum exposed visitor of 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months (4 visits, 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per visit).*

Annual average meteorougical dispersion (conservative, default use of maximum SITE BOUNDARY dispersion) from Table 3-7.

The equations in ODCH Section 4.3 may be used for calculating the potential dose to a HEMBER OF THE PUBLIC for activities inside the SITE BOUNDARY.

Based on these assumptions, this dose vould be at least a factor of 400 less than the maximum SITE BOUNDARY air dose as calculated in ODCH Section 3.7.

There are no areas onsite accessible to the public vhere exposure to liquid effluents could occur. Therefore, the modeling of ODCH Section 2.4 conservatively estimates the maximum potential dose to HEMBERS OF THE PUBLIC.

Based on a maximum conservative estimate.

4.2 Doses to MEMBERS OF THE PUBLIC 40 CFR 190 As required by Technical Specification 6.9.1.11 and ODCH Section 7.2, the Semiannual Effluent and Vaste Disposal Report shall also include an assessment of the radiation dose to the likely most exposed HENBER OF THE PUBLIC for reactor releases and other nearby uranium fuel cycle sources I

j_

Davis-Besse ODCH 75 Revision 4, 1991

e (including dose contributions from effluents and direct radiation from onsite sources).

For the likely most exposed HEMBER OF THE PUBLIC in th vicinity of the Devis-Besse site, the sources of exposure need consider t the radioactive effluents and direct exposure contribution from Davis-Besse.

No other fuel cycle facilities contribute significantly to the cumulative dose to a HEMBER OF THE FUBLIC in the immediate vicinity of the site.

Fermi-2 is the closest fuel cycle facility located about 20 miles to the NNV. Due to environmental dispersion, any routint releases from Fermi-2 vould contribute insignificantly to the potential doses in the vicinity of Davis-Besse.

The correlation of measured plant effluents with pathway modeling of this ODCH provide the primary method for demonstrating / evaluating compliance with the limits specified belov (40 CFR 190).

However, as appropriate, the results of the environmental monitoring program may be used to provide additional data on actual measured levels of radioactive material in the actual pathways of exposure. ODCH Section 4.2.3 discusses the methodology for correlating measured levels of radioactive material in environmental pathvay samples with potential doses. Also, results of the land use census may be used to determine actual exposure pathvays and locations.

The annual (calendar year) dose or dose commitment to any HEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel i

cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than 1

or equal to 75 mrem.

Vith the calculated doses from the releases of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Sections 2.4.1, 3.7.1, and 3.0.1, evaluations should be made including direct radiation contributions from the rector units and from outside storage tanks to determine whether the above limits of this Section have been exceeded.

If such is the case, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Section 7.3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a HEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathvayo and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations.

If the estimated dose (s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has nnt already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190.

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

O Davis-Besse ODCH 76 Revision 4, 1991

4 This requirement.is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The requirement requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem.

n For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a HEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses f rom the reactor units at:d outside storage tanks are kept small. The Special Report vill describe a course of action that should result in the limitation of the annual dose to a HEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits.

For'the purposes of the Special Report, it may be assumed that the dose commitment to the HEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, vith the exception that the dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If a dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordanco with the provisions of 40 CFR Part 190.11 and 10 CFR.Part 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other dose requirements for dose limitation of 10 CFR Part 20, as addressed in Sections 2.2 and 3.3.1.

An individual is not considered a HEMBER OF THE PUBLIC during any period in.

O which.he/she is engaged in carrying out any operation that is a part of the nuclear fuel. cycle.

F 4.2.1 Effluent Dose Calculations For purposes of implementing the above requirements of determining the cumulative dose contribution from liquid and gaseous offluents in accordance with Sections 2 and 3 and the reporting requirements of Section 7, dose calculations for Davis-Besse may be performed using the calculational methods contained within this ODCHF the conservative controlling pathvays and locations of Table 3-7 or the actual.pathvays and locations as-identified by.the land use~ census may be used.- Liquid pathway doses.may be calculated using equations in ODCH Section 2.4.

Doses due to releases of i

radiolodines,itritium and particulates are calculated based on_ equations in-Section 3.8.

The following equations may be used for calculating the dose to HEMBERS OF THE PUBLIC from releases of. noble gases:

D

= 3.17E-08

U

X/O-.* I (K

- Q, )

( 4-_1 )

i tb g

and D,

= 3.17E-08

U

X/Q

I ((L, + 1.1 M,)

0, )

(4-2)

Davis-Besse;ODCM 77 Revision 4, 1991

-....,-,-m-

--. -.._, i.. ----.....

m

...m r

M-,M. ~

where total body dose due to gamme emissions for noble gas D

=

radionuclides (mrem) skin dose due to gamma and beta emissions for noble gas D

=

radionuclides (mrem) duration of exposure (hr/yr, default values in Table 4-1)

U

=

atrnospheric dispersion to the of fsite location (secM )

y/Q

=

Q' cumulative release of noble gas radionuclide (i) over the period

=

of interest (vC1) total body dose factor due to gamma emissions from noble gas K'

=

radionuclide (i) from Table 3-6 (mrenvyr per vCi/m')

L' skindosefactorduetobetaemissionsfromnoblepas

=

radionuclide (i) from Table 3-6 (mrenvyr per uC1/m )

M' gm air dose factoy)for noble gas radionuclide (i) f rom Table 3-6

=

(mrad /yv per vC1/m 8760 =

hours per year mrem skin dose per mrad gamma air dose (mrenVmrad) 1.1

=

1/3.15E+07 yr/see 3.17E-08

=

Average annual meteorological dispersion parameters or meteorological conditions concurrent with the release period under evaluation may be used (e.g., quarterly averages or year-specific annual averages).

4.2.2 Direct Exposure tbse Determination - Onsite Sources Any potentially significant direct exposure contribution from onsite sources to offsite individual doses may be evaluated based on the results of.the environmental measurements (e.g., TLD, ion chamber nasurements) or by the use of a radiation transport and shielding calculacional method.

Only during atypical conditions will there exist any potential for significant onsite sources at Davis-Besse that would yield potentially significant offsite doses to a MEMBER CF THE PUBLIC). However, should a situation exist whereby the direct exposure contribution is potentially significant, onsite measurements, offsite measurements and calculationni techniques will be used for determination of dose for assessing 40 CrR 190 compliance.

The following simplified method may be used for evaluating the direct dose based on onsite or site boundary measurements:

D,6

= D,,0 (X,,0):

(4-3)

(X 0)#

O Davis-Besse ODCM 78 Revision 4, 1991

where D' ' e =

direct radiation dose measured at location B (onsite or site boundary) in sector 0 0

0=

extrapolated dose at location L in same sector 0 X

0=

distance to the location L from the radiation source X,, e =

distance to location B from the radiation source 4.2.3 Dose Assessment Based on Radiological Environmental Monitoring Data Normally, the assessment of potential doses to MEMBERS Or 'nlE PUBLIC must be calculated based on the measured radioactive effluents at the plant.

The resultant levels of radioactive material in the offsite environment are so minute as to be undetectable. The calculational methods as presented in this ODCM are used for modeling the transport in the environment and the resultant exposure to offsite individuals.

The results of the radiological environmental monitoring program can provide input into the overall assessment of impact of plant operations and radioactive effluents. With measured levels of plant related radioactive material in principal pathways of exposure, a quantitative assessment of potential exposures can be performed. With the monitoring program not identifying any measurable levels, the data provides a qualitative assessment - a confirmatory demonstration of *he negligible impact.

i Dose modeling can be simplified into three basic parameters that can be applied in using environmental monitoring data for dose assessment.

D=C*U*DF (4-4) where dose or dose commitment D

=

concentration in the exposure media, such as air concentration C

=

for the inhalation pathway, or fish, vegetation or milk concentration for the ingestion pathway individual exposure to the pathway, such as br/yr for direct U

=

exposure, kg/yr for ingestion pathway dose conversion factor to convert from an exposure or uptd a Dr an individual dose or dose commitment

'Ihe appicability of each of these basic modeling parameters to the use of environmental monitoring data for dose assessment is addressed below:

O Davis-Besse ODCM 79 Revision 4, 1991

Concentration - C The main value of using environmental s epling data to assess potential doses to individuals is that the data represents actual measured levtis of radioactive material in the exposure pathways. This eliminates one main uncertainty in the modeling - the release from the plant and the transport to the environmental exposure medium.

Environmental samples are collected on a routine frequency (e.g., weekly airborne particulate samples, monthly vegetable samples, annual fish sam-ples). To determine the annual average concentration in the environmental medium for use in assessing cumulative dose for the year, an average con-centration should be determined based on the sampling frequency and measured levels.

C, I(C,

t)/365 (4-5) where:

C, average concentration in the sampling medium for the year concentration of each radionuclide (i) measured in the C

individual sampling medium period of time that the measured concentration is considered t

representative of the sampling medium (typically equal to the sampling frequency; e.g., 7 days for weekly samples, 30 days for monthly samples).

If the concentration in the sampling medium is below the detection capabilities (i.e., less than lower limits of detection -LLD), a value of zero should be used for C,

( C, - 0 ).

Exposure - U Default exposure values (U) as recommended in Regulatory Guide 1.109 are presented in Table 4-1.

These values should be used only when specific data applicable to the environmental pathway being evaluated is unavailable.

Also, the routine radiological environmental monitoring program is designed to sample / monitor the environmental media that would provide early indications of any measurable levels in the environment but not necessarily levels to which any individual is exposed. For examtsle, sediment samples are collected in the area of the liquid diccharge:

typically, no individuals are directly exposed. To apply the measured levels of radioactivity in samples that are not directly applicable to exposure to real individuals, the approach recommended is to correlate the location and measured levels to actual locations of exposure. Hydro-logical or atmospheric dilution factors can be used to provide reasonable correlations of concentrations (and doses) at other locations. The other alternative is to conservatively assume a hypothetical individual at the sampling location. Doses that are calculated in this manner should be presented as hypothetical and very conservatively determined - actual O

Davis-Besse ODCM 80 Revision 4, 1991

exposure would be much less. Samples collected from nearby wells or actual water supply intake (e.g., Port Clinton) should be used for estimating the potential drinking water doses, other water samples collected, such as near field dilution area, are not applicable to this pathway.

Dose rectors - Dr The dose factors are used to convert the intake of the radioactive material to an individual dose commitment. Values of the dose factors are presented in NRC Regulatory Guide 1.109. The use of the Regulatory Guide 1.109 values applicable to the exposure pathway and maximum exposed it.dividual is referenced in Table 4-1.

4.2.4 Use of Environmental TLD for Assessing Doses Due to Nob 1e Gas Releases Thermoluminescent dosimeters (TLD) are routinely used to assess the direct exposure component of radiation doses in the environment. However, because routine releases of radioactive material (noble gases) are so low, the resultant direct exposure doses are also very low. A study

performed for the NRC concluded that it is possible to determine a plant contribution to the natural background radiation levels (direct exposure) of around 10 mrem per year (by optimum methods and high precision data). Therefore, for routine releases from nuclear power plants the use of TLD is mainly confirmatory - ensuring actual exposures are within the expected natural background variation.

For releases of noble gases, environmental modeling using plant measured releases and atmospheric transport models as presented in this ODCM represents the best method of assessing potential environmental doses,

\\

(

However, any observed variations in TLD measuremento outside the norm s

should be evaluated.

NUREG/CR-0711, Evaluation of Methods for the Determination of X-and Gamma-Ray Exposure Attributable to a Nuclear racility Using Environmental TLD Measurements, Gail dePlangue, June 1979, USNRC.

O Davis-Besse ODCM 81 Revision 4, 1991

Table 4-1 Reconrnended Exposure Rates in Lieu of site Specific Data

Table Reference Exposure T1.aay Maximum Exposed Exposure Rates for Dose factors 1

l Age Group from RG 1.109 Liquid Releases rish Adult 21 kg/y E-11 Drinking Water Adult 730 1/y E-11 Bottom Sediment Teen 67 h/y L-6 AtJncspheric Releases Inhalation Teen 8,000 m'/y E-8 Direct Exposure All 6,100 h/y**

N/A (ODCM Table 3-7)

Leafy Vegetables Child 26 kg/y E-13 Truits, vegetables & Grain Teen 630 kg/y E-12 Milk Infant 330 1/y E-14 Adapted f rom Regulatory Guide 1.109, Table E-5 Net exposure of 6,100 h/y is based on the total 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months per year adjusted by a 0.7 shielding factor as recomended in Regulatory Guide 1.109.

O Davis-Besse ODCM 82 Revision 4, 1991

5.0 AsSEsSMorr or IMD USE CDJSUS DATA A land use census (LUC) is conducted annually in th* vicinity of the Davis-Besse site. This census fulfills two main purposes:

1) meet requirements of TS (as required by 10 CrR 50, Appendix 1,Section IV.B.3) for identifying controlling location / pathway for dose assessment of ODCM I

l Section 3.8.1; and (2) provide data on actual exposure pathways for assessing realistic doses to MEMBERS OF THE PUBt!C.

t 5.1 Land Use Census as Required by TS As required by TS, a land use census shall be conducted during the growing season at least once per twelve months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agricultural authorities, he land use census shall identify within a distance of 8 km (5 miles) the location, in each of the r

16 rateorological sectors, of the nearest milk animalj(heneafest t

residence and the nearest garden of greater than 50 m 500 ft ) producing l

broad leaf vegetation. This requirement is provided to ensure that changes in the use of UNRESTRICTED ARIAS are identified and that modifications to the monitoring program are made if required by the results of this census, 21scensussatisfiestherequirementsofSectionIV.B.3ofAppendix!po i

10CrRpart50. Restricting the census to gardens of greater than 50 m (500 ft ) provides assurance that significant exposure pathways via leafy vegetables, will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables, assumed in Regulatory Guide 1.109 for consumption by a child.

3

-To determine this minimum garden size, the following assumptions were made (1) 20% of the garden was used for growing broad leaf vegetation (i.e,

g similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m.

The data from the land use census is used for updating the locatiorv' pathway for dose assessment and for updating the Radiological Environmental Monitoring Program, The results of the land use census shall be included 1

in the Annual Radiological Environmental Operating Report pursuant to-Section 7.1.

l With a land use census identifying a location (s) that yields a calculated dose or dose comitment greater than the values currently being calculated l

in Sections 3.8.1, in lieu of a Licensee Event Report, identify the new locations (s)'in the next Semiannual Effluent and waste Disposal Report, pursuant to section 7.2. With a land use census identifying a locations (s) that yields a calculated dose or dose comitment (via the same exposure pathway) 20 percent greater than that at a location from which samples are currently being obtained in accordance with Section 6.1, add the new locations (s) if practical -(and readily obtainable) to the Radiological Environmental Monitoring Program within 30 days. The sampling locations (s), excluding the control station location, having a lower calculated dose or dose comitment(s), via the same exposure pathway, may be deleted from this monitoring program.

In lieu of a Licensee Event Report and pursuant to Section 7.2, identify the new location (s) in the next Semiannual Effluent and Waste Disposal Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new

?

location (s).

O Davis-Besse ODCM 83 Revision 4, 1991 f

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The following guidelines shall be used for assessing the results from the land use census to ensure compliance with this Section.

A.

Data Compilation A.1 Locations and pathways of exposure as identified by the land use census will be compiled for comparison with the current locations as presented in Table 3-4.

A.2 Changes from the previous year's census will be identified.

Also, any location / pathway not currently included in the Radiological Environmental Monitoring Program (Table 6-2) will be identified.

A.3 Historical, annual average meteorological dispersion parameters

( x/Q, D/Q) for any new location (i.e., location not previously identified and/or evaluated) will be determined. All locations should be evaluated against the same historical meteorological data set.

D.

Relative Dose Significance B.1 ror all new locations, the relative dose significance will be determined by applicable pathways of exposure.

B.l.1 Relative dose calculations should be based on a generic radionuclide distribution (e.g., Davis-Besse USAR gaseous effluent source term or past year actual effluents). An 1-131 source term dose may be used for assessment of the maximum organ ingestion pathway dose because of its overwhelming contribution ta the total dose relative to the other particulates.

B.1.2 The pathway dose equations of the ODCM should be used.

C.

Data Evaluation C.1 The controlling location used in the ODCM Table 3-4 will be verified.

If any location / pathway (s) is identified with a higher relative dose, this location / pathway (s) should replace the previously identified controlling location / pathway in Table 3-4.

If the previously identified controlling pathway is no longer present, the current controlling location / pathway should be determined.

C.2 Any changes in either the controlling location / pathway (s) ef tha OCCM dose calculations (Section 3.7 and Table 3-4) or the Radiological Environmental Monitoring Program (ODCM Section 6.0 and Table 6-2) shall be reported to NRC in accordance with ODCM Section 5.1 and 7.2.

O Davis-Besse ODCM 84 Revision 4, 1991

5.2 Land Use Census to support Realistic Dose Assessment n e Land Use census (LUC) provides data needed to support the special dose analyses of the ODCM Section 4.0.

Activities inside the SITE DOUtOARY should be periodically reviewed for dose ascessment as required by TS 6.9.1.11 (ODCM Section 4.1).

Assessment of realistic doses to MEMBERS OF

'nIE PUBLIC is required by Section 4.0 for demonstrating compliance with the EPA Environmental Dose Standard, 40 CrR 190 (ODCM Section 4.2).

To support these dose assessments, the LUC shall include (a) areas within the SITE BOU!OARY that are accessible to the public; and (b) use of Lake Erie water on and near the site. %e scope of the LUC shall include the following:

Assessment of areas onsite that are accescible to MEMBERS or THE PUBLIC. Particular attention should be give to assessing exposure times for visits to the Davis-Besse Administration Building. Data should be used for updating ODCM Table 4-1.

Data on Lake Erie use should be obtained from local and state officials. Reasonable efforts shall be made to identify individual irrigation and potable water users, and industrial and comercial water users whose source is Lake Erie. This data is used to verify the pathways of exposure used in ODCM Section 2.4.

l t-('

Davis-Besse ODCM 85 Revision 4, 1991 l

I 6.0 RADIOLOGICAL DNIRCNMDRAL MONI'IORING PRCGPAM The Radiological Environmental Monitoring Pro ram (REMP) required by TS provides measurements of radiation and of rad oactive materials in those exposure pathways and for those radionuclides which lead to the higher potential radiation exposures of individuals resulting irom the station operations. The sampling and analysis program described in this Section was developed to provide representative measurements of radiation and

]

radioactive materials resulting from station operation in the principal i

4 pathways of exposure of MEMBERS Or THE PUBLIC. This monitoring program implements Sections IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent controls by verifying that the measurable concentrations of radioactive e terials and levels of radiation are not higher than expected on the basis of the ef fluent measurements and the modeling of the environmental exposure pathways. Guidance for the development of this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring.

6.1 Program _ Description 6.1.1 General The REMP shall be conducted as specified in Table 6-1.

This table describes the minimum environmental media to be sampled, the sample collection frequencies, the number of representative samples required, the characteristics of the sampling locations, and the type and frequency of sample analysis. Table 6-2 provides a detailed listing of the sample locations for Davis-Besse which satisfy the requirements of Table 6-1.

Maps for each site listed in Table 6-2 are contained in Appendix C.

The specific locations used to satisfy the requirements of Table 6-1 may be changed as deemed appropriate by the Radiological Environmental Supervisor.

The changes shall be reported in the Annual Radiological Environmental Operating Report and the Semiannual Effluent and Waste Disposal Report as required by Sections 7.1 and 7.2, respectively.

If the changes are to be permanent, Table 6-2 and Appendix C shall be updated.

Note:

For the purpose of implementing Section 5.1, sampling locations will be modified, to reflect the findings of the land use census as described in ODCM Section 5.1.

6.1.2 Program Deviations With the REMP not being conducted as specified in Table 6-1, in lieu of a Licensee Event Report, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by TS 6.9.1.10 and Section 7.1, a description of the reasons for not conducting the program as required and plans for preventing a recurrence.

O Davis-Besse ODCM 86 Revision 4, 1991

6.1.3 Unavailability of Milk or Broad Leaf Vegetation Samples with milk or frer.h leafy vegetable samples unavailable from one or more of the sample locations required by Table 6-1, identify locations for obtaining replacen.ent samples and if practical add them to the REMP within 30 days. The locations from which samples were unavailable may then be deleted from the monitoring program. In lieu of a Licensee Event Report and pursuant to TS 6.9.1.11 and Section 7.2, identify the cause of the unavailability of samples and identify and the new locations (s) for obtaining replacement samples in the next Semiannual Effluent and Waste Disposal Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new locations (s).

6.1.4 Seasonal Unavailability, Equipment Malfunctions, Safety Concerns with specimens unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons, every effort will be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule will be documented in the Annual Radiological Environmental Operating report pursuant to TS 6.9.1.10 and section 7.1.

6.1.5 Sample Analysis REMP sampics shall be analyzed pursuant to the requirements of Table 6-1 and the detection capabilities required by Table 6-3.

Cumulative potential dose contributions for the current calendar year from radionuclides detected in environmental samples shall be determined in accordance with

(

the methodology and parameters in this ODCM.

6.2 Reporting Levels 6.2.1 General h e reporting levels are based on the design objective doses of 10 CrR 50, Appendix I (i.e., levels of radioactive material in the sampling media corresponding to potential annual doses of 3 mrem, total body or 10 mrem, taximum organ from liquid pathways; or 5 mrem, total body, or 15 mrem, maxirnum organ for gaceous effluent pathweys - tha annual limits of Sections 2.4.1, 3.7.1 and 3.8.1).

These potential doses are modeled on the maximum exposure or consumption rates of NRC Regulatory Guide 1.109.

We evaluation of potential doses should be based solely on radioactive material resulting from plant operation.

LO Davis-Besse ODCM 87 Revision 4, 1991 i

l 6.2.2 Exceedance of Reporting Levels With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 6-4 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Comission within 30 days, pursuant to section 7.3, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to MEMBER OF THE PUBLIC is less than the calendar year limits of Sections 2.4.1, 3.7.1 and 3.8.1, When more than one of the radionuclides in Table 6-3 are detected in the sampling medium, this report shall be submitted its concentration (1) +

Concentration (2) +...> 1.0 reporting level (1) reporting level (2)

When radionuclides other than those in Table 6-4 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MI2tBER OF THE PUBLIC is equal to or greater than the calendar year limits of Sections 2.4.1, 3.7.1 and 3.8.1.

The method described in section 4.2.3 may be used for assessing the potential dose and required reporting for radionuclides other than those listed in Table 6-4.

A special report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental operating Report.

6.3 Interlaboratory Comparison Program Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Comission. The requirement for participating in an approved Interlaboratory Comparison program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmntal sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid for tb purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

A sumary of the results obtained as part of the required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental operating Report pursuant to TS 6.9.1.10 and Section 7.1.

With analyses not being performed as required, report the correctiva actions taken to prevent a recurrence to the Comission in i.he Annual Radiological Environmental Operating Report pursuant to TS 6.9.1.10 and Section 7.1.

Davis-Besse ODCM 88 Revision 4, 1991

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TABLE 6-1 RADIOLOGICAL EN7IRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Representative Type and Frequency and/or Sample Samples and Sample Locations' Collection Frequency of Analysis 1.

DIRECT RADIATI M 27 routine monitoring stations

. Quarterly c.umma dose quarterly either with two or more dosi-meters or with one instrument for measuring and recording dose rate continuously, placed as follows:

an inner ring of stations, generally one in each meteorological sector in the general area of the SITE BOUNDARY; an outer ring of stations, one in each meteorological sector in the 6-to 8-km range from the site, excluding the sectors over Iake Erie; the balance of the stations to be placed in special interest areas such as populations centers, nearby residences, schools, and in 1 or 2 areas to serve as control stations.

DAVIS-BESSE ODCM 89 Revision 4, 1991

TABLE 6-1 (Continued)

RADIOLOGICAL DNIR0tPEMAL FKNI*KRITG PROGFRf Exposure Pathway

! Amber of Representative Type and Frequency and/or Sanple Samples and Sample Iocations*

Collection Fregaency of Analysis 2.

AIRDORIE Radiciodine and Samples from 5 locations, Continuous sampler Radiciodine Cannister:

Particulates placed as follows:

operation with sample I-131 analysis weekly.

collection weekly, or 3 samples from close to the more frequently if Particulate Sampler:

SITE BOUNDARY, in different required by dust Gross beta radioactivity sectors, generally from areas loading.

analysis following filter of higher calculated annual change;* Gama isotopic average groundlevel DA).

analysis of composite (by location) quarterly.

1 sample from the vicinity of a nearby community, generally in the area of higher calculated annual average groundlevel DA).

1 sample from a control location, 15-30 km from the site.

3.

WATERDCRE a.

Surface 2 samples Weekly camposite Tritium and g -

(untreated water) sasple (Indicator isotopic

analysis of location should be a caposite sar:ple monthly.

composite) b.

Ground Sample from one source Quarterly Gama Isotopic

arvi tritium analysis only if likely to be affected*

quarterly.

Davis-Besse ODCM 90 Revision 4, 1991 O

O O.

-1

,i l

2 TABLE 6-1 (Continued)

RADIOUXIICAL DFmOWENDL MONIl0 RING PROGRAM Exposure Pathway Number of Representative Type and Frequency j

and/or Sample

- Samples and Sagle Incations*

Collection Frequency of Analysis c.

Drinking 1 sample from the nearest Weekly mimite

. Gross beta on monthly (Treated water) source.

sample.

em posite. Trititsa and gasuna isotopic l

1 cample from a control analysis on quarterly location.

composite. I-131 analysis on each r=posite iden the dose calculated for the l

ccuposite 1Aen the dose i

calculated for the j

ctWien of the water l-is greater than 1 stem per year.

{

d.. Sediment from i sample from area with Semiannually ca=== isotopic analysis

j Shoreline existing 'or potential semiannually.

I recreational value.

l-4.

INGESTION a.

Milk If available, camples from Semiumthly when c - isotopic

and I-131 l

animals up to 2 locations animals are on analysis semiennthly when within 8 km distance having pasture, monthly animals are on pasture; j'

the highest dose potential.

at other times monthly at other times.

.1 sagle' front milking aniemis j

i at a control location 15-30 km i

distant and generally in a less prevalent wind direction.

1 i

3 i

Davis-Besse ODCM 91 Revision 4, 1991 i.

)

1

TABLE 6-1 (continued)

RADIOLOGICAL ENVIRONMENTAL NONITORING P30 GRAM Exposure Pathway Number of Representative Type and Frequency and/or Sample Samples and Sample Locations

Collecticrt Frc.qaency of Analysis isotcpic analysis

b.

Fish 1 sample each of 2 -rcially 1 sar:iple in season.

ranana and/or recreationally en edible portions.

important species in vicinity of site.

I sample of same species in areas not influenced by plant discharge.

c.

Food Products Onples of up to 3 different Monthly when available. Gama isotopic

an:1 I-131 (Broad leaf kinds of broad leaf vegatation analysis.

vegetation) growth in two different offsite locations of higher predicted annual average grourxi-level D t f

if milk sampling is not performed.

I sample of each of the similar Monthly when available. Gama isotopic

and I-131 broad leaf vegetations grown analysis.

15-30 km distant in a less prevalent wind direction if milk sampling is not performed.

Davis-Besse ODCM 92 Revision 4, 1991 O

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l TABLE 6-1 (Continued) o TABLE toTATICU

' Specific parameters of distance and direction sector from the centerline of one reactor, additional description where pertinent are provided for each and every sample location in Table 6-2.

Ref er to !UREG-0133, " Preparation of Radiological eBffluent Technical Specifications for 11uclear Power Plants", October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, llovember 1979.

It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be possible or practicable to continue to ootain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program. In lieu of a 1.icensee Event Report and pursuant to specification 6.9.1.11, and Section 7.2 identify the cause of the unavailability of samples for that pathway and identify the new locations (s) for obtaining replacement samples in the next Semlannual Effluent and waste Disposal Report and also include in the report a revised figure (s) and table for the ODCM reflecting the new locations (s).

one or more instruments, cuch as a pressurized ion changer, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters.

For the purposes of this table, a therm 31uminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a

(~}

packet are considered as two or more dosimeters, rilm badges shall not be used

()

as dosimeters for measuring direct radiation. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g.,

at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

' Airborne particultte sample filters shall be analyzed for grost beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay.

If gross beta activity in air particulata samples is greater than ten times the yearly mean of control samples, gama isotopic analysis shall be performed on the individual samples.

"Gama isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the ef fluents f rom the facility.

' Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gt Jient or recharge properties are suitable for contamination, Davis-Besse ODCM 93 Revision 4, 1991 pO l

Table 6-2 Sampling Locations Appendix C Type of Location Page Reference Locati_on*

Location Description T-1 C-3 I

site boundary, 0.6 mile i

DJE of Station.

T-2 C-4 I

Site boundary, 0.9 mile E of Station.

T-3 C-5

.t Site boundary, 1.4 miles ESE of Station near mouth of Toussaint River.

T-4 C-6 I

Site boundary, 0.0 mile S of Station.

T-5 C-7 Main entrance to site, 0.7 mile W of Station.

T-6 C-0 I

Site boundary, 0.5 mile

!NE of Station.

T-7A C-9 I

Sand Beach main entrance, 0.9 mile IM of Station.

T-7D C-9 I

Sand Beach residence, 0.8 mile 14M of Station.

T-0 C-10 I

rarm, 2.7 miles WSW of Station.

T-9 C-11 C

Oak Harbor substation, 6.8 miles SW of Station.

T-10 C-12 I

site boundary, 0.5 mile SSW of Station.

T-11 C-13 C

Port Clinton Water Treatment plant, 9.5 miles SE of Station.

T-12 C-14 C

Toledo Water Treatment Plant, 23.5 miles WrM of Station.

Water samples are collected 11.3 miles IM of site.

T-25 C-15 I

rarm, 3.7 miles S of Station.

I = Indicator locations; C = Control locations.

Davis-Besse ODCM 94 Revision 4, 1991

g on Appendix C Type of Location Page Reference Location

Location Description T-27 C-16 C

Crane Creek State Park, 5.3 miles WIM of Station.

T-28 C-17 I

Davis-Besse Water Treatment Plant, onsite.

T-33 C-18 I

Lake Erie within a 5 mile radius.

T-35 C-19 C

take Erie greater than a 10 mile radius.

T-37 C-20 C

rarm, 13 miles SW of Station.

T-40 C-21 I

Site boundary, 0.7 mile SE of Station.

T-41 C-22 I

Site Boundary, 0.6 mile SSE of Station.

T-42 C-23 I

Site boundary, 0.8 mile SW of Station.

T-44 C-24 I

Site boundary, 0.5 mile WSW of Station.

T-46 C-25 I

Site boundary, 0.5 mile IM of Station.

T-47 C-26 I

site boundary, 0.5 mile N of Station.

T-48

(, 27 I

Site boundary, 0.5 mile NE of Station.

T-50 C-28 I

Erie Industrial Park Water Treatment Plant, 4.5 mile SE of Station.

I = Indicator locations; C - Control locations.

Davis-Besse ODCM 95 Revision 4, 1991

e Table 6-2 (continued)

Sampling Locations Appendix C T)Te of Location Page Reference Location

Location Description T-52 C-29 I

rarm, 3.7 miles S of Station.

T-54 C-30 I

rarm, 4.0 miles SW of Station.

T-55 C-31 I

rarm, 5.0 miles W. of Station.

T-57 C-32 C

rarm, 22 miles SSE of Station.

T-67 C-33 I

site boundary, 0.3 mile t3M of Station.

T-60 C-34 I

site Boundary, 0.5 miles WtM of station T-91 C-35 I

Siren Post No. 1100, 2.5 miles SSE of Station.

T-112 C-36 I

state Route 2 and Thompson Road, 1.5 miles SSW of Station.

T-151 C-51 I

State Route 2 and Humphrey Road,1.8 miles W!M of Station.

I = Incidator locations; C = Control locations.

Davis-Desse ODCM 96 Revision 4, 1991 l

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Table 6-3 IINER LIMITS OF utim nCN (LID)*

Airborne Particulate Water or Gas Fish Milk Food Products Sediment Analysis (pCi/1)

(oci/m')

(pCi/kg. wet)

(oci/1)

( pCi Ag, wet)

(oCIAg, dry)

Gross Eeta 4*

1.0E-02

H 2000**

" Mn 15 130

" FE 30 260

'** **Co 15 130

::N 30 260

" ZR 15

*I i

7.0E-02 1

60 d

' ' ' ' * "Cs 15( 10* ),18 6.0E-02 130 15 60 150

- - Ba 15 15 This list does not mean that only these nuclides are !n be detected and reported.

PUTE:

Other peaks which are measurable and identifiable, together with the abtne nuclides, shall be l

i identified ard reported.

If no drinking water pathway exists, a value of 3000 pCi/L may be used.

(

97 Revision 4, 1991 Davis-Besse ODCM

a TABLE 6-3 (Continued)

TABLE ICTATICt1 A.

The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal, for a particular measurement system (which may include radiochemical separation):

4.66 s, Ea V

2.22

Y

exp(-Aot) where LLD is the lower limit of detection as defined above (as pC1 per unit mass or volume).

s is the standard deviation of the background counting rate or of the ebunting rate of a blank sample as appropriate (as counts per minute).

E is the counting efficiency (as counts per transformation).

V is the sample size (in units of mass or volume).

2.22 is the number of transformations per minute per picoeurie.

g Y is the fractional radiochemical yield (when applicable).

A is the radioactive decay constant for the particular radionuclide.

At is the elapsed time between end of the emple collection period and time of counting.

Typical values of E, V, Y and at should be used in the calculations.

O Davis-Besse ODCM 98 Revision 4, 1991

O O

O'

~

TABLE 6-4 REPORTING LEVELS EUR RADIOACTIVITY C0tO2HPATIOG IN EINIRCIM2TIAL SAMPLES Reporting Ievels Water Airborne Particulate Fish Milk Vegetables Analysis

( pCi/1) or Gases (pCi/m')

(pCi/kg, ret)

(pCi/1)

(pCi/kg, wet)

EI-3 2.0E+04 Mn-54 1.0E+03 3.0E+04 Fe-59 4.0E+02 1.0E+04 Co-58 1.0E+03 3.0E+04 Co-60 3.0Et02 1.0E+04 Zn-65 3.0E+02 2.0E+04 Zr-Nb-95 4.0E+02 I-131 2.0E+00 9.0E-01 3.0E+00 1.0E+02 CS-134 3.0E+01 1.0E+01 1.0E+03 6.0E+01 1.0E+03 CS-137 5.0E+01 2.0E+01 2.0E+413 7.0E+03 2.0E+03 3.0E+02 Ba-La-140 2.0E+02 For drinking water samples, this is the 40 CFR 141 vr21ue. If no :1rinking water pathway exists, a value of 30,000 pCi/ liter ray be used.

Revision 4, 1991 99 Davis-Besse ODCM

e 7.0 ADMINISTRATIVE CONTROLS 7.1 Annual Radiological Environmental Operating Report Routine Radiological Environmental Operating reporte covering the operation of the unit during the previous calendar year shall be submitted prior to

[

May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with the preoperational studies, with operational controls, as appropriate, and with previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the tesults of land use censuses as required in Section 5.1.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all

=--

radiation measurements taken during the period pursuant to the locations c

[

specified in Sections 6.1 and Appendix C of this ODCM, as well as L

summarized and tabulated results of these analyses and measurements.

In the event that sone individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the followingt a summary description of the radiological environmental monitoring programs at least two legible maps covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Section 6.31 and discussions of all analyses in which the LLD required by Table 6-3 was not achievable.

7.2 Semiannual Effluent and Vaste Disposal Report Routine Effluent and Vaste Disposal Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Semiannual Effluent and Vaste Disposal Reports (Semiannual Reports) shall include a summary of the quantities of radioactive liquid and gatenue effluents and solid vaste released from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Vastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Vater-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

O Davis-Besse ODCM 100 Revision 4, 1991

4.

The Semiannual Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data qt collected over the previous year. This annual summary may be either in the g

form of an hour-by-hour listing on magnetic tape of vind speed, vind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of vind speed, vind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY during the reporting period. All assumptior.. used in making these assessments, i.e.,

specific activity, exposure time, and location, shall be included in these reports. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in this ODCH.

The Semiannual report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed HEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation."

The Semiannual report shall include the following information for each class of solid vaste (as defined by 10 CFR Part 61) shipped offsite during the report period:

V(h a.

Container volume, b.

Total curie quantity (specify whether determined by measurement or estimate),

Principal radionuclides (specify whether determined by measurement c.

or estimate),

d.

Source of vaste and processing employed (e.g., devatered spent resin, compressed dry vaste, evaporator bottoms).

Type of container (e.g.. Type A, Type 3, Large Quantity), and e.

f.

Soliditication agent or absorbent (e.g., cement, urea formaldehyde).

The Semlannual Reports shall include a list and deteription of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Semiannual Reports shall include any changes made during the reporting period to the PROCESS CONTROL' PROGRAM (PCP) and to the ODCH, as well as a listing of new locations for dose calculations and pursuant to Section 5.1.

7.3 Special Reports o

I Special Reports shall be submitted to the U. S. Nuclear Regulatory Commission (NRC) in accordance with 10 CFR 50.4 vithin the time period Davis-Besse ODCH 101 Revision 4, 1991 j

specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable references a.

Dose or dose commitment exceedences to a HEMBER OF THE PUBLIC f rom radioactive materials in liquid effluents released to UNRESTRICTED AREAS ( Sec tion 2.4.1).

b.

The discharge of radioactive liquid vaste without treatment and in excess of the limits in Section 2.5.

c.

The calculated air dose from radioactive gases exceeding the limits in Section 3.7.1.

d.

The esiculated dose from the release of iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding the limits of Section 3.8.1.

e.

The discharge of radioactive gaseous vaste without treatment and in excess of the limits in Section 3.9.

f.

The calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding the limits of Section 4.2.

g.

The level of radioactivity as the result of plant effluents in an environmental sampling medium exceeding the reporting levels of Table 6-4 (Section 6.2.2).

7.4 Major Changes to Radioactive Liquid and Caseous Vaste Treatment Systems Licensee initiated major changes to the radioactive vaste systems (liquid and gaseous):

1.

Shall be reported to the Commission in the update to the Safety Analysis Report. The discussion of each change shall contains a.

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59; b.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; c.

A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; d,

An evaluation on the change, which shovs the predicted releases of radioactive materials in liquid or gaseous effluents and/or quantity of solid vaste that differ from those previously predicted in the license application and amendments thereto; e.

An evaluation of the change, which shows the expected maxirum exposures to individuals in the UNRESTRICTED AREA and the general population that differ from those previously estimated in the license application and amendments thereto; Davis-Eesse ODCH 102 Revision 4, 1991

f.

A comparison of the predicted releases of radioactive materials, in gm liquid and gaseous etfluents, to the actual releases for the period Q

prior to when the changes are to be mades g.

An estimate of the exposure to plant operating personnel as a result of the changel and h.

Documentation of the fact that the change was reviewed and found accaptable by the Station Review Board.

2.

Shall become etfective upon review and acceptance by the Station Reviev Board.

7.5 Definitions 7.5.1 BATCH RELEASE - The discharge of liquid wastes of a discrete volume.

7.5.2 CHANNEL CALIBRATION - A channel calibration shall be the adjustment, as necessary, of the channel output such that it responds with necessary range and accuracy to known values of the parameters which the channel monitors. The channel calibration shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Channel calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire chennel is calibrated.

7.5.3 CHANNEL CHECK - A channel check shall be the qualitative assessment of channel behavior during operation by observation. This determination O

shall include, where possible, comparison of the channel indication V

and/or status with other indications and/or status derived from independent instrument channels monitoring the same parameter.

7.5.4 CHANNEL FUNCTIONAL TEST - A channel functional test shall bet Analog Channels - The injection of a simulated signal into the a.

channel ao close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions, b.

Bistable Channels - The injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions.

7.5.5 COMPOSITE SAMPLE - A sample in which the method of sampling employed results in a specimen which is representative of the liquids released.

7.5.6 GASEOUS RADVASTE TREATHENT SYSTEH - A system that is designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off gases and providing for decay for the purpose of reducing the total radioactivity prior to release to the environment.

O Davis-Besse ODCH 103 Revision 4, 1991

7.5.7 LOVER LIMIT OF DETECTION (LLD) - The LLD is the smallest concentration of radioactive material in a sample that vill be detected with 95%

probability, with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

LLD.

4.66 Sb E

V

2.22

Y

exp(- Aot) where LLD is the lover limit of detection as defined above (as pCi per unit mass or volume):

Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute):

E is the counting efficiency (as counts per transformations);

V is the sample size ( in units of mass or volume):

2.21 is the number of transformations per minute per picoeurtei Y is the fractional radiochemical yield (when applicable);

A is the radioactive decay constant for the particular radionuclides at for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

-It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

7.5.8 MEMBER OF THE PUBLIC - Member (s) of the public shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreations, occupational, or other purposes not associated with the plant.

7.5.9 OPERABLE - OPERABILITY - A system, subsystem, train, component or device shall be operable or have operability when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal vatec, lubrication or other auxiliary device to perform its function (s), are also capable of performing their related support functions (s).

O Davis-Besse ODCM 104 Revision 4, 1991

7.5.10 PURGE-PURGING - the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration

['~5g

(_,/

or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

7.5.11 SITE BOUNDARY - The site boundary shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

7.5.12 SOURCE CHECK - A source check shall be the observation of channel upscale response when the channel sensor is exposed to a radioactive source.

7.5.13 UNRESTRICTED AREA - An unrestricted area shall be any area at or beyond the SITE boundary, access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation or radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes. The definition of unrestricted area used in implementing the Radiological Effluent Technical Specifications has been expended over that in 10 CFR 100.3(a), but the unrestricted area does not include areas over vater bodies. The concept of unrestricted areas, established at or beyond the SITE BOUNDARY, is utilized in the Technical Specifications and the ODCM to keep levels of radioactive materials in liquid and gaseous effluents as lov as is reasonably achievable, pursuant to 10 CFR 50.36a.

7.5.14 VENTILATION EXHAUST TREATMENT SYSTEM - a ventilation exhaust treatment

(~'N system is a system that is designed and installed to reduce radioactive

\\~ l material in particulate form in effluents by passing ventilation or vent exhaust gases through UEPA filters for the purpose of removing particulates from the gaseous exhaust stream prior to release to the environment.

Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be ventilation exhaust treatment system components.

7.5.15 VENT-VENTING - the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

Davis-Besse ODCM 105 Revision 4, 1991 1

APPENDIX A Technical Basis for Simplitled Dose Calculations Liquid Effluent Releases O

O 1

Davis-Besse ODCM Revision 4, 1991

t APPENDIX A 9

Technical Basis for Simplified Dose Calculations Liquid Effluent Releases

]

Overview To simplify the dose calculation process, it is conservative to identify a controlling, dose-significant radionuclide and to use its dose conversion factor in the dose calculations. Using the total release (i.e., the cumulative activity of all radionuclides) and this single dose conversion factor as inputs to a one-step dose assessment yields a dose calculation method which is both simple and conservative.

Cs-134 is the controlling nuclide for the total body dose.

It has the highest total body dose conversion factor for all the radionuclides listed in Table 2-5.

Therefore, the use of its dose conversion factor in the simplified dose assessment method for evaluating the total body dose is demonstrably conservative.

The selection of the maximum organ dose conversion factor for use in the simplified calculation requires censideration of the prevalence of the radionuclides in the effluents. An examination of the Table 2-5 factor vill show that the Nb-95 dose factor for the GI-LLI represents the highest value (1.51E+06 mrem /hr per pCi/ml); and the P-32 bone factor (1.39E+06) is similarly high.

Hovever, neither of these tvo radionuclides are of significance in the Davis-Besse effluents. Nb-95 it not typically measured in the liquid-effluente and P-32 analyses are not even performed.

(NRC has 9

categorically determined that P-32 is not a significant radionuclide in liquid effluents from nuclear power plants and does not require the special radiochemical analyses needed for identification and quantification.) The next highest dose conversion factor is for Cs-134, liver, with a value of 7.llE+05 mrem /hr per UCi/ml. And, Cs-134 is a prevalent radionuclide in the liquid effluents from Davis-Besse. Therefore, it is recommended that the Cs-134 liver dose conversion factor be used for the simplified maximum organ dose assessment.

O Davis-Besse ODCH A-1 Revision 4, 1991

8 Simplified Method For evaluatinE compliance with the dose limits of Section 2.41, the folloving simplified equations may be used:

Total Body 1.67E-02

VOL D,,

=

- ~***'

Dr

Z where:

dose to the total body (mrem)

D

=

volume of liquid effluents released (gal)

VOL

=

average Collection Box release flow (gal / min)

Dr

=

10, near field dilution Z

=

A'**'*34b' 5.81E+05 mrenVhr per vCi/ml, the total body ingestion dose

=

factor for Cs-134 IC, total concentration of all radionuclides (vci/ml)

=

1 hr/60 min 1.67E-02

=

substituting the values for Z and the Cs-134 total body dose conversion

~

factor, the equation simplifies to:

9.70 E+02

VOL

IC (A-2)

D

=

Maximum organ 1.67E-02

VOL max (cs-134,11ver)

A where:

maximum organ dose (mrem)

D,,,

7.llE+05 mrenyhr per Ci/ml, the liver ingestion dose A

=

i C'- 13 4 11" ' )

factor for Cs-134 O

Davis-Besse ODCM A-2 Revision 4, 1991

Substituting the values for Z and the Cs-134 liver dose conversion factor, the s

equation simplifies tot 1.19 E+03

VOL

]

D,,,

IC,

(A-4)

=

Dr Tritium should not be included in the simplified analysis dose assessment for liquid releases, The potential dose resulting from normal reactor releases of H-3 is relatively negligible. But, its relatively higher abundance would yield resulting simplified doses that would be overly conservative and unrealistic. Excluding tritium has essentially no impact on the conservative g

use of this recommended simplified method. Furthermore, the release of tritium is a function of operating history and is essentially unrelated to radwaste system operations.

O O

i Davis-Besse ODCM A-3 Revision 4, 1991

o O

APPENDIX B Technical Basis for Effective Dose Factors Gaseous Radwaste Effluents e

O Davis-Besse ODCM Revision 4, 1991

W t'

a

-e-,

2 APPENDIX B f

Technical Basis for Effective Dose Factors i

4 Gaseous Radweste Effluents Overview Dose evaluations for releases of gaseous radioactive effluents may be simplified by the use of an effective dose factor rather than radionuclide-specific dose factors. These effective dose factors are applied to the= total radioactive release to approximate the various doses in.the-l

environmenti 1.e.,c the total body, ganna-air, and beta-air doses.. The

-l effective dose factors are based on the typical radionuclide distribution'in the gaseous radioactive effluents. This approach reduces the analyses to a single multiplication (K M

, or N

) times the-quantity of radioactive'

- gases released,' rather tOk,inhihidual- &M1yses for each radionuclide and.

summing the results to determine the dose. Yet the approach provides a reasonable estimate of the-actual doses since under normal operating conditions there is expected to be minor variations in the radionuclide-distribution..

q Determination-of Effective Dose Factors 1

-Effective dose transfer factors are calculated by the following equations:-

j

-g K,,, = E ( K,

f, )

(B-1) wheres.

)

D 1

K" the effective total body dose factor due to ganu9, emissions-.

=

from all' noble gases released (mrenVyr per vCi/m ' effective)-

1

- K' '

the total--body dose factor _due to gamma emissions from each noble gas ~ radionuclide' (i)- released-(mrenvyr per C1/m', from' l

Table:3-6)'

~

the fractionaliabundance ofl noble gas radionuclide (i)

fy relative to-the total noble gas _ activity.

= I((L + 1.1Mg)

f )

.(B-2)

-(L +-1.1-M) gx 3

p

.where

~

the effective-skin dose factor due to beta and ganna

- ( L+ 1.1M ) ' ' ~

. emissions from all noble gases released:(mrenVyr per 3

lpC1/m,. effective).

' the> skin dose factor due to beta and gamma emissions f rm:

( L, +1 ~ 1M, )

=

--each poble gas radionuclide (1) released.-(mrenvyr per 3

pci/m,'from Table 3-6)

~

j i

I.

L 3

c:

l-i 1

Ib Davis-Besse ODCM B-1 Revision 4, 1991

)

M,,, = I( M,

f,)

(B-3) wheret the ef fective air dose f actor due to gamma emissions f rom all M

noble gases released (mrad /yr per vCi/m*, effective)

=

M' the air dose factor due to gamma emissions f[om each noble gas

=

radionuclide (i) released (mrad /yr per vCi/m, f rom Table 3-6)

N,,, = E( N,

f,)

(B-4) where the effective air dose factor due to beta emissions from all N

noble gases released (mrad /yr per vC1/m', effective)

=

the air dose factor due to beta emissions from each noble gas N'

radionuclide (i) released (mrad /yr per pC1/m', from Table 3-6)

=

Normally, past radioactive effluent data would be used for the determination of the effective dose factors. However, the releases of noble gases from Davis-Besse have been exceedingly insignificant. Therefore, in order to ensure overall conservatism in the rnodeling, the USAR estimate of radionuclide concentrations at the site boundary (summarized in Table B-1) has been used as the initial typical distribution. The effective dose factors derived from this distribution are presented in Table B-2.

Application To provide an additional degree of conservatism, a factor of 2.0 is introduced into the dose calculation when the effective dose factor is used. This conservatism provides additional assurance that the evaluation of dnses by the use of a single effective dose factor will not significantly underestimats any actual doses in the environment.

For evaluating compliance with the dose limits of Technical Specification 3.11.2.2 the followicq sitplified equations may be used:

Dy = 214 3.17E-08

yA)

M,,,

IQ, (B-5) and DS = 2.0

3.17E-08

X/Q

N,,,

IQ, (B-6)

O Davis-Besse ODCM B-2 Revision 4, 1991

where:

i air dose due to gamma emissions for the cumulative release of all Dy

=

noble gases (mrad) air dose due to beta emissions for the cumulative release of all D6

=

noble gases (mrad) atmospheric dispersion to the controlling site boundary (sec/m')

y/Q

=

M,,, =

5.7E+02, effective gamma-air dose factor (mrad /yr per C1/m')

N,,, =

1.lE+03, effective beta-air dose factor (mrad /yr per vCi/m')

0, cumulative release for all noble gas radionuclides (vci) conversion factor (yr/sec) 3.17E-08

=

conservatism factor to account for the variability in the effluent 2.0 data.

Combining the constants, the dose calculation equations simplify to Dy = 3.61E-05

WQ

IQ, (B-7) and DS = 7.20E-05

y/Q

E0, (B-8)

O The effective dose factors are used for the purpose of facilitating the timely assessment of radioactive effluent releases, particularly during periods when the computer or ODCM software may be unavailable to perform a detailed dose assessment.

O Davis-Besse ODCM B-3 Revision 4, 1991

TABLE B-1 Default Noble Gas Radionuclide Distribution

of Gaseous Effluents Traction of Total

'A / I A,)

3 Containment Station Waste Gas Nuclide Vessel Purge Vent Decay Tank Total Ar-41 0.0003 0.004 0.004 0.003 Kr-85 0.12 0.012 0.034 0.06 Xe-131m 0.02 0.009 0.008 0.017 Xe-133m 0.005 0.011 0.011 0.008 Xe-133 0.86 0.94 0.92 0.83 0.004 0.0034 0.06 Xe-135m Xe-135 0.002 0.02 0.02 0.021 Total 1.0 1.0 1.0 1.0 tore:

Data adapted from Davis-Besse USAR Section 11.3, Table 11.3-13 and Table 11.3-14.

Kr-83m, Kr-05m, Kr-87, Kr-88 and Xe-138 have been excluded because of their negligible fractional abundance (i.e., < 1%).

O Davis-Besse ODCM B-4 Revision 4, 1991

TABLE B-2

=

[V s Y

Effective Dose Factors - Noble Gas Effluents Total Body Skin Dose Gamma Air Beta Air Dose factor

;' actor -

Dose Factor Dose Factor

' Isotope Fractional K (L+1.lMP he)rM(kfkdpr per (hfkdp)r per N

(kfbmptper Abundance (mre vC1/g) vCi/m )

vCi/m vCi/m )

m Ar-41 0.003'

-2,65E+01.

3.87E+01 2.79E+01 9.84E+00

.Kr 0.06.

9.96E-01

.8.15E+01

-1.03E+00 1.17E+02 Xe-131m-0.017 1.55E+00-

.1.10E+01 2.65E+00 1.88E+01 Xe-133m 0.008'-

2.00E+00.

1.08E+01-

.2.6]E+00 1.18E+01-

.Xe-133 0.83 2.44E+02 5.76E+02 2.93E+02 8.72E+02-Xe-135m 0.06 1.87E+02 2.64E+02

'2.02E+02 4.43E+01 Xe-135 0.02 3.62E+01 7.94E+02 4~.03E+01 5.16E+01 WrAL 1.~ 0 4.98E+02 9.89E+02 5.69E+02 1.12E+03

~

n.

(

i l l.

N.

j;

- j i

2

O Davis-Besse'ODCM B-5 Revision'4, 1991 a

O APPENDIX C Radiological Environmental Monitoring Program Sample Location Maps O

r e

Davis-Besse ODCM Revision 4, 1991 l

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Davis-Besse CDCM C-1 Revision 4, 1991

l l

Sampling locations (excepting thosc on the site periphery)

Davis-Besse Nuclear Power Station, Unit No.1.

i NW NNW N

NNE G

y 1-> s

<s-Ne l

g WNW i

g ?>

y ENE Oreg Lake Erie 1-n g

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O APPENDIX J Justifications i O l l Davis-Besse ODCH Revision 4 0

Safety Evaluation for the Davis-Besse Ra,diological Effluent Technical Specifications Amendment (,-~)s q.J overview Revision to the Davis-Besse Appendix A and Appendix B Technical Specifications are proposed which vill implement the regulatory requirement of 10 CFR 50, Appendix I on ALARA for radioactive effluents and other NRC regulations and criterio on radioactive maswcial monitoring instrumentation, radioactive material control, and radiologien1 environmental monitoring. In keeping with NRC guidelines, all radiological requirements are being deleted from Appendix B and placed in Appendix A. This proposed amendment is a revision to a previously submitted amendment to the NRC dated March 16,1979 (Serial No. 488). The major areas that are addressed in the revtsed submittal are as follovs Liquid and gateous effluent monitoring instrumentation -- operation and f-~-). \\--( periodic operability checks; . Liquid and gaseous radiiactive material releases--maximum release rates, quarterly dose limits ar.1 yearly dose limits; - Sampling and enalysis requirements ini batch and continuous radioactive material releases; Operation requirements on the liquid radvaste treatment systems Curie inventory limit on outside temporary liquid storage tanks; Maximum allovable oxygen concentration in the vaste gas system; UAVIS-BESSE, UNIT 1 J-l (3 V

Requirements to assure all solid waste meets applicable burial site tequirements; g Radiological environmental monitoring program--minor revisions to reflect current prog :a and currant NBC guidelin s. Changes have also been made to Section 6 of Appendix A to reflect the applicable administrative controls needed for the Section 3/4 revisions. A notable addition to the amendment is the inclusion of a requirement for an v.f-site Dose Calculation Manual (ODCM),.nd a Process Control Program The ODCM and PCP are not licensed documents but are referenced in (PCP). S the Technical Specification as present.ing acceptable methods for evaluat-The ing compliance with applicable Technical Specification requirements. ODCM provides calculational methods for deterisining radioactive effluent instrumentation darm setpoints, and for evaluating releases of radioac-tive effluents and corresponding doses. The ODCM also includes the The PCP sampling locations for the environmental monitoring program. presents the methods used to verify that vaste (dewatered resins) as processed for disposal meets appropriate shipping and burial ground Changes may be made to these documents without NRC approval; regulations. review by the SRB is required. Safety Evaluation An evaluation of the revised amendment has been performed to assure that the revisions as proposed do not involve an unreviewed safety question as defined in 10 CFR 50.59. The three criteria of 10 CTR 50.59 for the unreviewed safety question determination are addressed below, Probability of occurrence or the consequences of an accident or i) malfunction of equipment important to safety previe"aly evaluated in the safety analysis report. O r J' C AM S 2 ESSE. '1 NIT '.

j radiation Except for the addition of the turbine building liquid effluent s moottor (for which an FCR bas already been initiated), no plant equipment modifications are required by the proposed amendment. Certain procedural requirements will need to be developed but these address routine radioac-tive material effluents and controls; no accident procedures are involved. ii) Probability for accident or malfunction of a different type than any evaluated previously in the SAR may be created. For reasons as stated in response to item (i) above, the proposed amend-ment does not directly or indirectly pose a probability for an accident or The amendment will implement the NRC regulations for routine malfunction. releases and. controls of radioactive material. The amendment does not address any engineered safety features of the plant des!gu. iii) Margin of safety as defined in the basis for any technical specifi-cation is reduced. The proposed The proposed amendment does not reduce the margin of safety. amendment addresses routine releases and control of radioactive material; are involved. Several except as noted in item (i), no plant modification: operating procedure changes may be needed, but these chaages will be only for routine operations and will have no impact on accident probability or consequences. For the reasons discussed above for each of the crite.ria of 10 CTR 50.59, it is concluded that the amendment as proposed doca-cot involve an unre-viewed safety question. l J-3 2 AVIS-BESSE. UNIT 1

~ Servtce Water System--Radiological Ef fluent Monitoring Requirements O The setytce water system is classified as a non-radioactive system, being removed from radioactive systems by two boundaries. Radioactive systems are serviced by the component cooling water system interface; and, the service water system provides cooling to the component cooling water system through closed loop heat ex-bangers. Therefore, any leaks from radioactive systems into the plant water systems would first be identified by the monitoring of the component cooling water system prict to any additional unexpected leakage into the service water system. As a prudent measure, the service water system is monitored in accordance with the NRC guidance of Standard Review Plan, Section 11.5. However, because this system is a non-radioactive system and is separated from radioactive systems through two closed-loop boundaries, no Technical Specification requirements are needed for routine monitoring and analysis for radioac-tive effluents. O

9 J-*

DAVIS-BESSE. UNIT 1

Radioactive Effluent Instrumentation--Actomatic l' solation Feature O. r The radioactive effluent monitoring instrumentation at Davis-Besse does include provisions as called for in the NRC Standard Radiological not Effluent Technical Specifications for automatic isolation should any of the following conditions exist: circuit failure, downscale failure, or instrument not set in operate mode. Even though the automatic isolation features do not exist, administrative controls have been established such that should any of these conditions exist the control of radioactive effluents would not be significantly impacted. Esseatially all releases of liquid radwaste are controlled as individual batch releases with predetermined allowable release conditions. Thereby the radiation monitor serves saf nly as a back-up; primary control is established by the prerelease radiological analyses and evaluations. To assure the availability of the back-up monitoring, the status of the instruments is checked once per shift by the control room operators. Indicator lights on the instrument panel are checked to verify operability. An indicator would illuminate should a fr.ilure occur such as the ones delineated above.- Therefore, in addition to the administrative controls on allowable releases, the verifi-cation of instrument operability prior to releases of radioactive effluents and the "once per shift" status check by the control room operators provides adequate assprance of the proper control of the radioactive effluents. e e 1 l J-5 DAVTS-BESSE, UNIT 1

~ Technical Bases for Eliminating Curie Inventory i 1.imit for Gaseous Waste Decay Tang The NRC Standard Technical Specifications include a limit for the amount of radioactivity that can be stored in a single waste gas decay tank. This curie inventory limit is established to assure that in the event of a tank failure releasing the radioactive content to the environment the resulting total body dose at the site boundary would not exceed 0.5 rem. For Davis-Besse the inventory limit in the waste gas storage tank has been m ermined to be approximately 45,000 curies (Xe-133, equivalent), An allowable. primary coolant radioactivity concentration is established by the Technical Specificati.ons which limit the primary coolant radioactivity concentrations to 100/E with E being the average energy of the radioactiv-ity in Mev. This equation yields an upper primary coolant gross activity limit of about 200 pCi/ml. By applying this activity concentration limit to the total liquid volume of the primary system, a total activity limit can be' determined. For Davis-Besse the primary system volume is about 56,000 gallons, which yields a limiting total inventory of approximately 41,000 C1. By assuming a typical radionuclide distribution an equivalent Xe-133 inventory can be determined. Table 1 provides the typical radionuclide The (noble gases) oistribution and the Xe-133 equivalent concentration. equivalent concentration is determined by multiplying the radionuclide concentration by the ratio of the nuclide total body,cose factor to the Xe-133' total body dose factor. Summing all the indivioual radionuclide equivalent concentrations provides the overall Xe-133 equivalent concen-For determining concentration in a vaste gas decay :.ank, a tration. conservative assumption of 48 hours decay in desassing the primary systesa The data has been used to correct the primary coolant concentrations. show that the equivalent concentration (decsy corrected) is less than the gross concentration (i.e.,16 pCi/gm total in primary coolant versus 12 pCi/gm equivalent). The resulting Xe-133 equivalent curie inventory for WGDT input is approximately 31,000 C1. J-6 DAVIS-BESSE, UNIT 1 1

~ Therefore, even if the total primary system at the maximum Tech Spec allowable concentration was degassed to a single, waste gas decay tank, the tank curse inventory would be well below the 45,000 Ci limit. Based on this evaluation, the curie inventory limit on a single waste gas storage tank has not been included as a Technical Specification requirement. t' ci i O <-y J-7 DAVIS-BESSE, UNIT 1

o Table 1 s Xe-13', Effective Concentration Primary

Half-life Concentration Reg Ouide 1.109 Ratio Xe-133 Coolant

@ 48 br decay TB Dose Factor TB DF Effective C (pCi/0M) (pCi/ml) arem/yr Xe-133 DF @ 48 hr dec pCi/m (pCi/ml) 3 7.6x10~8 Kr-83M 2.0-02 1.9 hr 1.2x10-3 4.1 Kr-85M 1.1-01 4.5 br Kr-85 7.4-02 10.7 yr 7.4x10-2 1.6x10'5 0.06 4.4x10-3 5.2x10-3 20. Kr-87 5.8-02 76.3 nin 1.5x10-2 52. Kr-88 1.9-01 2.84 hr 1.7x10-2 57. Kr-89 4.8-03 3.16 min y:;-131M 8.4-02 12 days 7.5x10-2 9.2x10-5 0.32 2.4x10-2 2133M 2.0-01 2.2 days 1.1210-1 2.5x10-4 0.86 9.5x10-2 X2-133 1.5+01 5.3 days 1.2x10'1 2.9x10-4 1.0 1.2x10+1 3.1x10-3 ti, Xe-135M 1.3-02 16 min 'X2-135 3.3 01 9.1 br 8.5x10-3 1.8x10'3 6.2 5.3x10-2 1.4x10-3 4.8 X,-137 8.7-03 4 min 8.8x10'8 30 X3-138 4.3-02 17 min 1.2x10'1 Total-1.6x10'1 1.2x10+1 CAdapted from Davis-Besse Evaluation of Compliance with Appendir 7 to 10 CFR 50, June 4, 1976. O J-B DAVIS-SESSE. LHIT 1

a m t Lovar Limit of Datection--D3ccy Correction Factor o The equation and definition of the lover limit of detection in the NRC Standard

vhich is Radiological Effluent Technical Specification include the term e s

used to decay correct the analysis. The LLD is further defined as an a priori l (bef ore the f act) limit representing the capabilities of a measurement system I and not an a poster'iori (after the fact) limit for a particular measurement. Providing a decay correction for an evaluation of the capabilities of a system does not appear appropriate. It may be appropriate to decay correct certain the analyses of specific samples to determine radionuclide concentrations at Even in this case, such a correction is not appropriate for time of release. Analyses are performed prior to any release; and, the sample batch releases. the same rate as the batch from which the sample vas taken. vill be decaying at For continuous releases, decay correcting analyses of samples obtained over a specified sampling interval must take into account the accumulation of radioactivity in the sampling medium, the decay during the sampling interval and, especially for short lived radionuclides, equilibrium or quasi-equilibrium conditions that may be achieved. O Short-lived radionuclides vill tend to reach an equilibrium value ir. the medium as a function of source input and half-life. A single decay sampling correction to adjust for sampling interval vill provide an unacceptable Equilibricia concentrations must be considered if analyses are to overestimate. be indicative of actual release quantities. Employing exp (-Mt) to adjust for radioactive decay between the end of sampling to use the sare and the time of analysis is straightforvard. However, to attempt As a term to adjust the decay during the sampling period is not proper. practical mat ter, when the half-life of a radionuclide is long relative to the sampling time and the time betveen sampling and analysis, i.e., minimal decay, the correction term vill be near unity. In that event, the correction term is relatively unimportant. Davis-Besse, Unit 1 J-9 l

B At the other extrese, vhan tha helt-life of c rcdionuclide is much shorter than the sampling time or the time between the end of sampling and the analysis, the term exp(- Aot) could be used to adjust for decay between the end of sampling and the analysis. However, it vould not be appropriate in that case to use the sam term to at tempt to adjust for decay during sampling. The relationship betveen the radioactivity in a sample at the end of sampling and activity concentration in the medium being sampled is somevhat more involved. To explain this in the simplest condition, assume the radionuclide concentration is constant in the medium being sampled and that the mediun is sampled at a constant rate. In the instance of water sampling, the relationship between the activity concentration in the water being sampled and the activity concentration in the vater samplo at the end of sampling ist At C3=C2 t 1-e (3) where C1 = radionuclide concentration in the vater being sampled 2 = radionuclide concentration in the vater sample at the end of sampling C t = duration of sampling A = radionuclide decay constant when A t >> 1, C = C XI g 2 In the separate case of sampling a radionuclide in air by filtering the air and analyzing radioactive material collected on the filter, the radioni-.ide of interest is concentrated. Absent diluent air in the semple bei analyzed, the relation between radioactivity on the sample media and radionuclide concentration in the air being sampled is q = C F (1-e ) (2) g A Davis-Besse, Unit 1 J-10

o vhoro radionuclide concentration in the air being sampled C, radioactivity on the sample media (assuming 100% collection efficiency) 1 = sampler flov rate (volume / time) F ra ionuclide decay constant A d dutation of sa'mpling t = when A t > > 1, C, = q A/ F, This merely recognires that the rate of loss from the filter by radioactive decay equals the rate of collection onto the filter at equilibrium. The NRC proposed equation appears to incorporate an adulterated way of encouraging analysis soon after the end of sampling and to encourage efficient sample concentration or radiochemical extraction. Although not rigorous, it combines both objectives in a simple and thus practical vay, provided the decay correction is not extrapolated to a time earlier than the end of sampling. A more nearly rigorous vay of determining the activity concentration (or minimum detectable activity) in the medium being sampled is to assess the LLD in the sample at the time of analysis. Then the activity concentration in the medium being sampled can be calculated with the product of exp(-Xot) for decay between the end of sampling and the analysis and one of the equations derived herein for ' the relation between the medium being sampled and the activity in the sample at the end of sampling. However, this method is not very practical or necessary considering the types of sampling and analysis at nuclear power plants, the significant radionuclides, and the offsite potential doses. The bulk of radioactivity is'rcleased as batch releases with all sampling and analysis performed prior to release. Therefore, no decay corrections are applicable. It is in the sampling and analysis of continuous releases that the accumulation and decay of the radioactive material may need to b3 considered. The use of NRC's guidance for decay correction to the mid-point of the sampling period can grossly overestimate actual release qualities of short-lived radionuclides, while providing little improvement for Davis-Besse, Unit 1 J-11

the quantification of the longer half-life radionuclides that are the major dose contributors. overall, it may be appropriate to decay correct a certain analysis to account for radionuclide decay during the sampling period. However, simple decay ] correction to the mid-point of sampling vill grossly overestimate any short-lived radionuclides that may be detected. More consideration needs to be given by the NRC to address this problem. In any case, the use of a decay correction factor in defining a lover limit of detection is inappropriate. The ulD is a measurement of the capability of the measurement system and should not be used to try to establish a regulatory position on sampling and decay correction for quantificatici of releases. 7 = 0 Davis-Besse, Unit 1 J-12

_ _.. -. - -. ~ -. - ~.. c,' ~ Vaste Gas Decay System and Ventilation Systes--Operability Requiremencs O r .Q At Davis-Besse, the operation of the waste gas decay system is essentially' continuous, similar.to the routine operation of-such a systes at other PWRs. The syst.es consists of a surge tank which receives the waste gases from the primary' system, dual compressors-(one in-service and the other in reserve), and three waste-gas hold-up tanks (one in-service, one isolated for gas decay, and the third in reserve). Once the system is on-line with 'i a waste gas decay tank receiving primary systes, gases for the surge tank, operation is' automatic; no operator actions are required. The operating philosophy at Davis-Besse is to essentially operate the waste gas syrtes continuously. Not only is this philosophy prudent from an ALARA standpoint, but it is also conservative and protective from an operational stradpoint. Having to periodically evaluate primary system off-gas activity levels and anticipate unexpected increases in radioactiv-ity would be an unnecessary burden in determining needed waste gas system .O' -operation. -g For the ventilation systems, the' operating philosophy.is similar to that for the waste gas system; operation is continuous. But for the ventila-tion systems, the reasons for continuous operation are even more straight-Areas within the plant must be provided with outside air in forverd. order to provide an inside environment _ suitable for continued oc.upancy. I Without continuous ventilation system operation,- heat, humidity.'and airborne radioactive material levels would increase,and worker occupancy-

would be = jeopardized.

As-described in the Davis-Besse Appendix I evaluation,.the ventilar. ion-systems contain HEPA filters for removal of airborne radioactive particu-(As evaluated late material prior to release to the outside environment. for Appendix I. compliance, only the waste gas VMnt includes charcoal L filters for removal of radiciodines) The operation of the systems can essentially be considered a passive-eperation. No active operational i-J J-13 MVIS-BESSE. UNIT 1 l -~ - =

o ~ procedures are required for nonnal system operation for removal of strborne radioactive material. Davis-Resse's operating philosophy (and operating procedures) for the waste gas s,ystem and ventilation systems is a commitment in itself to the routine continuous operation of the systems. Having to commit to such a requirement (in lieu of a technical specification requirement on opera-tion) without appropriate consideratien of system down-time and plant shut-down (where operation may not be needed or feasible) is unacceptable and not in keeping with the principles of ALARA. Including special technical specifications that would impose additional procedures and periodic surveillance requirements in excess of those already established (which at present assure appropriate operation) is unnecessary and excessive. O O J-14 DAVIS-BESSE, UNIT 1

^ 1-f- Radiological Effluent Dose Analysis'-- Meteorology for Short Ters Releases Except for the vast.e gas decay tank (WGDT) releases and the containment purges releases, gaseous effluents from the Davis-Besse Station are from ventilation systems and are considered continuous releases. Most of the radioactive material in gaseous effluents is released from the WGDT. However, because of the essentially random nature of WGDT releases (i.e., no prescribed diurnal time, frequency or duration), the dose analysis of these releases is better modeled by the use of annual average meteorologi-cal conditions rather than short term meteorology. Containment purges are so infrequent that special meteorological analyses are not wa rranted; reasonable evaluations of off-site doses can be provided by the use of annual average meteorological conditions. O O DAVIS-BESSE, UNIT 1 J-15

L Radiological Envirorimental Reporting Levels Only the radionuclides licted in Table 3.12-2 of the proposed Radiological Effluent Technical Specifications (see note) for Davis-Besse are con 91dered in the reporting requirements for elevated levelt, of radioactive material in environmental sampling media. The radionuclides listed are those '. hat are dominant in the plant effluents and contribute essentially all of the environmental dose. Other radionuclides vill be present in plant effluents, but their contribution to the calculated total environmental dose vill be minor compared to the contribution of the radionuclides listed in Table 3.12-2. (see note) Even the contents of the NRC's Standard RETS reflect this position; not all pathvays include reporting levels for all the radionuclides listed (e.g., no reporting levels are presented for Co-58, Co-60, or Fe-59 for the milk, airborne p eticulate. or vegetable pathway). This very selective identification of pathway and important radionuclides reflects the very well defined concept of significant radionuclides for each particular pathway. Based on past experience in monitoring plant effluents and environmental sampling media, it can be stated with confidence that for the routine operation of Davis-Besse the radionuclides listed in Table 3.12-2 (see note) with applicable. reporting levels by the identified pathways are the only radionuclides that need be considered when evaluating potential doses in the offsito environment. Also, even if reporting levels were included for other radionuclides, the values would be higher than those for the significant radionuclides and would have a very minor role in determining actual reporting requirements. The reporting levels for the significant radionu:lides vould be reached well before any identified levels of other radionuclides vould even be . controlling.- Note: ' Table 3.12-2 has been incorporated into Section 6 of the ODCM, Revision 4, Table 6 4. DAVIS-BESSE, LNIT 1 J-16 O l 1

f Technical Basis for Eliminatina Curie Inventory Limit of Outside Liquid Tanks l At Davis-Besse, outside liquid tanks that potentially contain radioactive material are limited to the borated water storage tanks (2 tanks # 550,000 gallons) and the primary water storage tank (1 tank, 6 140,000 gallous). The borated water storage tanks are part of ECCS and are of seismic These tanks are designed to withstand extremely adverse environ-design. mental conditions and for purposes of this evaluation can be considered rupture proof. Also, overflow from the tanks is piped back to radwaste. For these reasons, it is considered unnecessary to impose curie inventory limits on these tanks. The primary water storage tank is used for normal make-up and letdown to Water contained within the W storage tank in typi-the primary system. Prior cally processed primary coolant or clean (non-radioactive) water. Oo to adding primary system water to the W storage tank,-the water is processed by avaporation and demineralization. This processing limits the levels of radicactivity in the tank. Past naspling and analysis of the tank has indicated only detectable levels of tritium, no other radionu-clides have been identified. Also, the overflow on th3 W storage tank is Therefore, due to the processing of any radioactive piped to radwaste. waste prior to addition to the W storage tank and the piping of the overflow to radwaste, the probability of any abrormal discharges to the . environment that could exceed the concentrations of_10 CFR 20, Appendix B, Table II, Column 2 at the nearest drinking water supply is extremely remote. Bacause of the design of the BWST and the design and operating conditions of the W.torage tank, it is considered unnecessary to impose curie S inventory limits on these tanks. Having to routinely sample and analyze for radioactivity concentration imposes an undue burden on plant personnel without providing any additional assurance of the public health and

safety, i

I J-17 DAVIS-BESSE, UNIT 1

Sampling Frequency for I-131: Significance of Power Changes and Increases in Coolant Activity Levels ~ The NRC guidance on effluent monitoring for I-131 (RETS Table 4 11-1 unote c) calls for increased sampling frequency for I-131 during increases ;or decreases) in reactor power level and increases in primary coolant level or noble gas effluent activity level. By system design, releases of radioactive material from plant operation are minor. Trying to identify small increases in I-131 releases that may (or may not) be associated with power changes is unnecessary. To evaluate the potential significance of increases in I-131 releases associated with power changes and the effect that may (or may not) be associated with power changes and the effect that sampling time may have on actual quantification of releases, the following example situation is evaluated. Consider a power increase (or decrease) on the first day of a seven (7) day sampling period that leads to an increase in I-131 release rate by a factor of 10 for one (1) day. After this one day increase, the release rate returns to the steady-state condition for the remaining 6 days of the sampling period. To evaluate the amount of I-131 on the sampling cartridge as a function of sampling time and concentration, the following equation is used q, = C, F ( 1-e" i' ) mA where: O 0, = quantity of activity on collection snedium C, = air concentration of radionuclide i A, = decay constant for radionuclide i t = sample time m - correction factor for collection efficiency Assuming 100% collection efficiency, at the end of the one day increase the total amount of activity (I-131) on the collection cartridge is determined to be 9.54 C.F. (For this example, the steady-state I-131 concentration is designated as C and the one day increase is 10 C.) For the remainder of the sampling pehiod with a concentration equal to,C,, the I-131 activity on the collection cartridge is equal to 4.66 C,F. By decaying the activity on the collection cartridge for the one day increase to the end of the sampling period and adding this quantity to 4.66 C,F, the total I-131 activity is determined to be 10.3 C,F. f Incorporated into ODCM, Revision 4, Table 3-3. DAVIS-BESSE, UNIT 1 J-18

If this value is decay corrected to the mid-point of the sampling period is accordance with the guidance of Regulatory Guide 1.21, the I-131 activity which is used to determine the release quantity is equal to 14.0 ) C T. g If a similar analysis is performed for the case of analyzing the collection cartridge at the end of the one day increase end analyzing a new cartridge at the end of 6 days sampling (constituting a 7 day sampling period), the total activity (decay corrected to mid-point of sampling periods) is detemined to be 16.0 C F. g By not analyzing the collection cartridge at the end of the one day increase, the total quantity of I-131 is underestimated by 14%. This The later into the analysis represents a somewhat worse case situation. If the sampling period that the increase occurs, the less the error. increase in release rate occurs af ter the mid point of the 7 day saarpling period, the actual talessa will be overestimated. Over a period of time involving numerous increases and decreases in effluent level, the rules of probability dictate that the overestimations and underestimations will tend to cancel out, providing an overall closer approximation to actual releases. Both the NRC in-plant measurement program and a study by EPRI* have indicated th.it minor increases in 1-131 releases may be associated with However, these reactor power changes and the iodine spiking phenomenon. studies also indicate that averall such increases are minor, not being a As was concluded significant contributor to the total releases of I-131. by the EPRI study for other PWRs, the main source of I-131 releases at Dr.vis-Besse is associated with containment purges. Since Regardless of the source, the total I-131 releases are negligible. initial start-up of Davis-Besse, the annual releases of I-131 have been and calculated maximum individual doses less than 0.01 less than 0.06 C g Even considering a hypothetical 14% increase for saarpling periods area. the effect on that may include iodine spiking in the primary coolant, O

EPRI NP-939, " Sources of Radiciodine at Pressurized Wacer Reactors".

Science Applications, Inc., November 1978 l J-19 DAVIS-BESSE, UNIT 1

total releases and calculated doses is still negligible. The actual increase will be even more insignificant considering the fact that the major source of I-131 at Davis-Besse is from containment purges. Based on a review of, plant operating data and the above analysis of the I-131 release quantification as a function of concentration and sampling time, it is concluded that for Davis-Besse, a sampling frequency based on power changes and increases in primary coolant I-131 concentrations is not justified. Determining the releases (and the insignificant environmental doses of these releases) on a weekly basis is sufficient verification of the negligible impact of plant operation. Trying to " fine tune" these releases is not justified considering the manpower and saterial costs casociated with the additional sampling and analysis. O 4 O J-20 l DAVIS-BESSE, UNIT 1

O Coodensate Deeineralizer Backwash Receiving Tank - Ra'dioactivity Control i i The discharge from the condensate desineralizer backwash receiving tank is controlled on a batch-by-batch basis in lieu of continuous radioactive i effluent monitoring. This method of operation has been determined to provide better control over the disnharge of the backwash receiving tank, preventing any unanticipated, unevaluated releases of radioactively contaminated secondary-side clean-up resins to the on-site settling basin. Prior to discharge, the contents of the backwash receiving tank are sampled and analyzed for radioactivity. As required, radioactively contaminated resins are transferred to radwaste for processing and dispos-al as radioactive seterial. The condensate desineralizer backwash receiving tank discharge line as originally designed included a radiation monitor. However, because of the nature of the resin-slurry mixture and the accumulation of resin beads in the monitor line, the radiation monitor has failed to provide the reliable ( ) f. indication of-radioactivity and control as originally intended. For this reason, it bas been determined that the sampling and analysis of each batch prior to discharge is needed to identify and evaluate radioactive contamination resulting from minor steam generator tube leaks (or residual radioactive saterial from previous leaks) that sight otherwise go unde-tected and unevaluated by a gross radiation effluent monitor, l The conomeoste desineralizer backwash receiving tank discharges to an l l on-site settling basin. No resin discharges are ma,d,e directly to the i Therefore, even in the event of personnel error off-site environment. resulting in the inadvertent discharge of unacceptably radioactive, 1 l contaminated resins to the settling basin, no off-site releases would 1 All resins and radioactive material would be retained on-site occur. l within the settling basin. Appropriate follow-up measures could then be initiated to centrol the radioactive material and prevent any potential for releases to the off-site environment in excess of the regulatory limits. g J-21 DAVIS-BESSE, UNIT 1

--_w___ o Controlling the discharge of the condensate demineralizer backwash receiv-ing tank on a batch-by-batch basis provides adequate control over the releases of any radioactive material to the off-site environment from this pathway. Also, the discharge is to an on-site settling basin, represent-ing an additional passive barrier from release off-site. Even in the unlikely event of Personnel error, by discharging to an on-site settling basin and its isolation from the off-site environment, the probability of unwanted, unevaluated releases of radioactive material to the off-site environment is exceedingly remote. Any additional protective measures provided by a continuous radiation monitor (for which operational perfor-mance and reliability are unhkaly, based on past experience) are not considered needed. O O DAVIS-BESSE, UNIT 1 J-22

n Lower Limit Of Detection Def_inition And Application To Detection capabilities For Ce-144 The lower limit of detection (LLD), as defined in the Radiological Effluent Technical Specifications (RETS) is an "... a priori (before the f act) limit representing the capabilities of a measurement system and not as a posteriori (af ter the f act) limit for a particular measurement." As defined by this definition applicable to the detection capability for radioactive efflue'l..nalysis, the LLD is a statistical analysis of a background spectrum and represents the detection limits for a radionuclide if it is the only r-dionuclide present above background. LLDs should be determ ned bned %n an analysis of a blank (or background) sample. However, even with this definition and application of LLD, it can be increasingly difficult to achieve a predesignated LLD value for particular radionuclides as the photon abundance (i.e., decay yield) decreases. To address this problem, specific radionuclides have been identified in the RETS as being the principal radionuclides for which the required LLD aust be met. For the analysis of staples of liquid radioactive affluents, an LLD of 5 x 10~7 pCi/ml is required. For the principal samma emitters listed, all have characteristic gammas with energy levels and abundances that provide for sufficient analytical sensitivity yielding LLDs within the required value-of 5 x 10~7 pCi/a1 - except Ce-144. With a 10.8% abundance and an energy level of 133.5 kev, being able to meet the LLD of 5'u 10~7 pCi/al requiras optimum conditions--conditions which cannot be repeatedly achieved for an operational radiochemistry program at Davis-Besse. The low gamma yield is a major factor; however, with,an energy level which is located within the Compton continuum, the detection capability for Ce-144 even for a blank, background sagle is significantly higher com-pared with other so-called principal gsama emitters. The equation for LLD in the Davin-Besse RETS ist 4.66 Sb E V 2.22 Y O DAVIS-BESSE, UNIT 1 J-23 I

where: the standard deviation of the background counting rate S a b I A/T m background counting rate R = counting time T = counting efficiency E = sample size V = 2.22 = conversion factor (transformations per minute per picoeurie) fractional che.mical yield (when applicable) Y a s By substitution _of typical values in this equation, the LLDs for different principal samma emitters can be compared. For analysis of a typical background sample at Davis-Besse, the ratio of the LLDs for Ce-144 and Co-60 is about 5.35; for Ce-144 and Mn-54 the ratio is 8.34. These large ratios are demonstrative of some of the ' relative difficulties in achieving an LLD of 5 x 10-7 pCi/mi for Ce-144 compared with other principal samma emitters. Examining the equation of LLD, two emin factors can be altered in an attempt to improve the detection capability - count.ing time and detector (Altering sample size is not considered realistic since efficiency. larger samples would pose operational and standcrd calibration problems. It can also be shown that increasing sample volume does not strongly influence efficiency for counting on contact with the detector face due in part to sample self-shielding and decreased relative efficiency for the increased volume). J-24 DAVIS-BESSE, UNIT 1 - ^ ~ ' - - -' - - - - - - -

LLD improves at best as the square root of the counting time. Therefore, increasing the counting time from 2000 seconds to 5000 seconds would only D provide a 1.6 reduction in LLD. A 5000 second count is considered to be a Q reasonable maximum for radioactive effluent analysis. Having to extend to longer counting times would introduce a potential operational delay without commensurate improvement in detection capability. An improvement in the efficiency is negated in part by the corresponding increase in background count rate. A comparison of 5 GeLi detectors with relative efficiencies ranging from 7.2% to 22% was performed at the University of Michigan *. For a 500 m1 sample contact with the detectors, the 15% relative efficiency detector demonstrated the highest photopeak efficiency in this energy than did the 21% and 22% relative efficiency detectors. Some unexplainable dif ferences may be due to inherent manufacturers specifications; however, a valid conclusion is that increasing the detector efficiency provides little if any improvement in detection capability, especially in the low energy range (<200 kev). Therefore, the analysis of effluent samples at Davis-Besse with a 10% relative efficiency GeLi and a 5000 second counting time provides a detection system that is not only practical for an operational radio-chemistry program but can also be considered as representative of state-of-the-art for routipe, general purpose radionuclide detection. Since the required LLD of 5 x 10' 4/Ci/ml can not be met on a routine basis for Ce-144, therefore the LLD Ce-ll4 will be 2.0 x 10" sici/ml (Table 4.11-l** footnote b. ). D. M. Minnema, C. G. Hudson and J. D. Jones. "A Comparison of Ge(Li) Detectors with Different Efficiencies for Low-Level General Purpose Counting"; University of Michigan, 1978.

- Incorporated into ODCM, Revision 4, Table 3-3.

i O DAVIS-BESSE, UNIT 1 J-25 l u l l l [ -}}

ML20084U591

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1.0 INTRODUCTION

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ML20084U591

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