ML14063A471

Offsite Dose Calculation Manual, Revision 14

ML14063A471
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Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	08/06/2013
From:	Mcrill J Dairyland Power Cooperative
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DAIRYLAND POWER COOPERATIVE LA CROSSE BOILING WATER REACTOR OFFSITE DOSE CALCULATION MANUAL REVISION 14 PREPARED BY

• A;e DATE gjpport Engineer, Licensing REVIEWED BY Kcs2 K-*T-Health PhysicsB upMan DATE APPROVED BY _________ f________

Manager

TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

........................................................... 1 1.1 Purpose ......................................................................................................................... 1 1.2 Definitions 1 2.0 OFFSITE DOSE CALCULATIONS ...............................................

2.1 Compliance with the Limitations for Liquid Effluent Releases .................................................. 4 2.1.1 Calculation of Liquid Effluent Monitor Alarm Settings .................................................. 4 2.1.2 Calculation of Instantaneous Allowable Release Rates ........................ 4 2.1.3 Calculation of Liquid Effluent Dose Contribution ......................................................... 7 2.1.4 Calculation of Dose Commitments from Liquid Effluents ......................................... 8 2.2 Compliance with the Limitations for Gaseous Effluent Releases ......................................... 9 2.2.1 Calculation of Gaseous Effluent Monitor Alarm Settings ......................................... 9 2.2.2 H-3 and Particulates ............................................... 9 2.2.3 Calculation of Release Limits for H-3 and Particulates with Half-Lives Greater than 8 Days ........................................................................... 9 2.2.4 Alarm Setpoint Calculations for H-3 and Particulates with Half-Lives Greater than 8 Days ........................................................................... 10 2.2.5 Calculation of Gamma and Beta Air Dose Commitments ...................... 15 2.2.6 Calculation of Dose Commitments due to Gaseous Releases ................... 20 3.0 RADIOACTIVE EFFLUENT CONTROL PROGRAM 21 3.1 Program Requirements .................................................................................................... 21 3.2 Liquid Effluents ............................................................................................................... 21 3.2.1 Sampling and Analysis .................................................................................... 21 3.2.2 Liquid Effluent Release Limitation ....................................................................... 23 3.2.3 Liquid Effluent Instrumentation ........................................................................... 24 3.3 Gaseous Effluents 26 3.3.1 Reactor Building Ventilation ............................................................................... 26 3.3.2 Stack Effluent Sampling and Analyses ............................................................... 26 3.3.3 Stack Effluent Release Limitation ...................................... 28 3.3.4 Instrumentation .................................................. 29 3.4 Total Dose to a Member of the Public ......................................... 32 3.5 Radioactive Effluent Control Reporting Requirements ..................................................... 37 3.5.1 Radioactive Effluent Release Report ................................................................... 37 3.5.2 RECP Non-Conformance Reporting ................................................................... 38 4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ......................... 34 4.1 Program Requirements .................................................................................................... 34 4.2 REMP Description .......................................................................................................... 34 4.3 REMP Lower Limits of Detection ............................................ 36 4.3.1 Calculation of Lower Limits of Detection ................................. 36 4.4 Interlaboratory Comparison Program ............................................................................. 38 4.5 Radiological Environmental Monitoring Reporting Requirements ....................................... 38 4.5.1 Radiological Environmental Monitoring Report ...................................................... 38 4.5.2 REMP Non-Conformance Reporting ....................................................................... 38 ODCM Rev. 14 1

TABLE OF CONTENTS LIST OF TABLES 2.1 Infant Dose Factors Pi, (Inhalation) for H-3 and Particulate Gaseous Release Monitor Alarm Setpoint Determinations ................................................. 14 3.1 Radioactive Liquid Waste Sampling and Analysis Requirements for Batch Releases ..................................................... 21 3.2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirem ents ................................................................................................................. 25 3.3 Radioactive Gaseous Waste Sampling and Analysis ............................................................ 27 3.4 Radioactive Gaseous Effluent Monitoring Instrumentation ................................................. 30 3.5 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements ........................................................................................... 31 4.1 Radiological Environmental Monitoring Program ............................................................. 35 4.2 Environmental Sample Analyses Lower Limits Values (LLD) .............................................. 37 4.3 Reporting Levels for Radioactivity Concentrations in Environmental Sam ples ........................................................................................................................ 39 LIST OF FIGURES 1.1 Site Map, Including Effluent Release Boundary .............................................................. 3 2.1 Liquid Release Monitor Alarm Setpoint Determination ...................................................... 6 2.2 H-3 and Particulate Gaseous Release Monitor Alarm Setpoint Determination ......................................................... 12 2.3 H-3 and Particulate Gaseous Release Monitor Alarm Setpoint Summary ............................ 13 ODCM ii Rev. 14 I

1.0 INTRODUCTION

1.1 Purpose The Offsite Dose Calculation Manual (ODCM) contains the methodology and parameters used in (1) the calculation of offsite doses resulting from radioactive liquid and gaseous effluents from LACBWR, and (2) the calculation of liquid and gaseous effluent monitoring Alarm/Trip Setpoints.

The ODCM also contains the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs.

1.2 Defintions CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter 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. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.

b. Bistable channels - the injection of a real or simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.

EFFLUENT RELEASE BOUNDARY The Dairyland Power Cooperative property line within the 1109-feet (338-meter) radius EXCLUSION AREA is the EFFLUENT RELEASE BOUNDARY (See Figure 1.1.)

EXCLUSION AREA The EXCLUSION AREA is defined as the area within an 1109-feet (338-meter) radius from the centerline of the Reactor Building. This was the area established per 10 CFR 100 as the EXCLUSION AREA for plant siting and operation.

ODCM 1 of 39 Rev. 14 I

MAXIMUM PERMITTED CONCENTRATION (MPC)

The limiting liquid effluent concentration value 10 CFR 20, Appendix B, Table 2, Column 2.

MEMBER OF THE PUBLIC MEMBER OF THE PUBLIC shall mean an individual in a CONTROLLED or UNRESTRICTED AREA. However, an individual is not a MEMBER OF THE PUBLIC during any period in which the individual receives an occupational dose.

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) and when all necessary attendant instrumentation, controls, a normal or an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

ODCM 2 of 39 Rev. 14 I

FIGURE 1.1 SITE MAP INCLUDING EFFLUENT RELEASE BOUNDARY T-METEO.OLOGICAL TOWER..'/

O0 400 Soo0'

! p I Er )-f)

I'tv 0.1A/ "° /"?P-"°'°r (

r"*"

ODCM 3 of 39 Rev. 14 I

2.0 OFFSITE DOSE CALCULATIONS 2.1 Compliance with the Limitations for Liquid Effluent Releases 2.1.1 Calculation of Liquid Effluent Monitor Alarm Setpoints To assure compliance with the limitations of Section 3.2.2a, Radioactive Effluent Control Program (RECP), the liquid effluent monitor alarm setpoint is calculated as a function of the maximum effluent flow rate and the minimum dilution flow rate. The following equation is used to calculate the setpoint:

af <(2.1) k(F-+f) -

Where:

C= the effluent concentration limit implementing 10 CFR 20 for LACBWR, in pCi/ml.

a = the setpoint (in CPS above background) of the liquid effluent monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release is inversely proportional to the volumetric flow of the effluent line (f) and is proportional to the volumetric flow of the dilution stream plus the effluent stream (F + f). The setpoint represents a value which, ifexceeded, could result in concentrations exceeding the limits of 10 CFR 20.

k = the conversion factor, CPS per pCi/ml, for the liquid waste effluent monitor based upon most recent calibration of the monitor.

f= the effluent line volumetric flow setpoint as measured at the liquid effluent monitor location, in gallons per minute.

F= the dilution stream, LACBWR and Genoa Station No. 3 (G-3) Circulating Water volumetric flow, in gallons per minute.

Since f << F, Equation 2.1 is satisfied when the following discharge line liquid effluent monitor setpoint is met:

kCF a< kCF (2.2) f 2.1.2 Calculation of Instantaneous Allowable Release Rates LACBWR's liquid radwaste is released in batches. In order to assess the required liquid effluent monitor alarm setpoint, a, the following step-by-step method for obtaining data will be performed. The form presented in Figure 2.1 may be used as a worksheet for these calculations. The alarm setpoint calculation may be performed on an annual basis if the setpoint is determined to be sufficiently conservative so as to prevent exceeding 0.5 MPC at the discharge point where MPC is the isotope weighted effluent concentration release limit for a typical LACBWR waste batch based on 10 CFR 20, appendix B, Table 2, Column 2 values.

ODCM 4 of 39 Rev. 14 I

1) Go to Figure 2.1. Enter the date on the form.

2) Enter the concentration Ci (pCi/ml) for each isotope, i, in a typical LACBWR waste batch.

3) The values of f and F are determined and recorded at the top of Figure 2.1.

F is the minimum volumetric dilution flow rate during release at the LACBWRIG-3 outfall which is equal to the LACBWR Circulating Water flow rate plus the G-3 Circulating Water flow rate in gallons per minute. The value f is the maximum radioactive liquid release flow rate (GPM) for the batches discharged during the period. A value of 17 GPM is normally specified for f.

4) The quantities ECi and ZCi/MPCi are determined and recorded.

5) The monitor conversion factor, k, determined at last primary calibration is recorded on Figure 2.1, in CPS (net) per pCi/ml.

6) The alarm setpoint, a (in CPS with a factor of 0.5 for conservatism), for the liquid effluent monitor measuring the radioactivity concentration in the effluent line is then determined by:

0.5kF T. C1 a- MPCi (2.3) f Z C1/MPC1 ODCMV 5 of 39 Rev. 14 I

FIGURE 2.1 LIQUID RELEASE MONITOR ALARM SETPOINT DETERMINATION Date Maximum Liquid Release Rate for Period, f= GPM Minimum Dilution Flow Rate for Period, F= GPM CPS (net)

Monitor Conversion Factor, k pCi/ml Liquid Release Monitor Alarm Setpoint, 0.5kFXT Ci a= f Y,FE C/

Ci/MPCi CPS above background ODCM 6 of 39 Rev. 14 I

2.1.3 Calculation of Liauid Effluent Dose Contribution To demonstrate compliance with the limitations of Section 3.2.2b, dose contributions are calculated at a maximum interval of once every calendar quarter for all radionuclides identified in liquid effluents released to unrestricted areas using the methodology presented in NRC Regulatory Guide 1.109, Rev. 1, October 1977. This methodology takes the form of the following general equation:

DaT= Z* (AaiT Z1Th Ci/Fj) (2.4)

Where:

Dat = the cumulative dose commitment to the total body or any organ T, of an individual in age group a, from the liquid effluents released in batches m, expressed in mRem.

C.. = the total quantity of radionuclide i, released by batch j, in Ci.

AaiT = the site-related ingestion dose commitment factor to the total body or any organ T, of an individual in age group a, for each identified principal gamma and/or beta emitter, in mRem-gal-min'-Ci 1 .

Fj = the average dilution water flow rate during batch release j, in gallons per minute.

Equation 2.4 requires the use of a dose factor AajT for each nuclide, organ and individual in age group a, which includes the factors which determine the ultimate dose received such as pathway transfer factors (e.g., bioaccumulation factors),

pathway usage factors, ingestion dose factors and dilution factors. The following philosophy and site-specific conditions determine the site-specific factors incorporated into the liquid effluent dose calculation model:

1) Liquid Dose Pathways Due to LACBWR's status as a fresh water site, there is no invertebrate pathway. The drinking water pathway is not included, since the nearest community which obtains its drinking water supply from the Mississippi River is located at Davenport, Iowa, which is 195 miles downstream. The drinking water pathway represents < 0.01% of the dose to any organ. The irrigated foods pathway is not included since the river water is not used for irrigation in this area and the shoreline deposits pathway is insignificant for the Mississippi River. The only significant dose pathway is the dose commitment due to ingestion of fish from the Mississippi River waters.

2) Dilution The liquid effluent flow from the waste tanks is diluted by the combined Circulating Water flow for both LACBWR and G-3. For offsite dose calculations, no dilution by the Mississippi River flow is considered. Also, batch discharges of liquid effluent normally take place on average during less than 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> per month (< 5% of the time). Therefore, no fish in the river ODCM 7 of 39 Rev. 14 I

are continuously exposed to a radioactive environment produced by LACBWR liquid effluent as assumed in the calculation of the published bioaccumulation factors for fish.

Based on the above site-specific criteria, the dose factor AaI, is defined as follows:

AaiT=Ko(UFa)(BFj)(DFaiT) (2.5)

Where:

Ko = a units conversion constant, 5.03 E+5 =

(1 E+12 pCi/Ci x 0.2642 gal/ I) / (8760 hrs/yr x 60 min/hr).

UFa = fish consumption usage factor for an individual in age group a, in kg/yr.

BFj = the bioaccumulation factor in fish for nuclide i, in pCi/kg per pCi/ I.

DFaiT = the ingestion dose factor for age group a, for nuclide i, in organ T, in mRem/pCi.

2.1.4 Calculation of Dose Commitments from Liquid Effluents The equations for this calculation have been formatted on a computer-based spreadsheet. The values of UFa, BFj, and DFai, specified in NRC Regulatory Guide 1.109 Rev. 1, October 1977, and the constant K, have been entered on the spreadsheet.

To perform the calculation the following information is entered in the appropriate cells of the spreadsheet for each liquid batch released during the period of interest:

1) Date

2) Release interval, hrs

3) Waste volume, gal

4) Circulating Water flow rate, GPM

5) Activity concentration of each isotope i, in waste, pCi/ml.

The spreadsheet program will then calculate and display the total quarterly dose in mRem to the total body and each organ of an individual in each age group.

The cumulative calendar year doses and the percentage of the limits set forth in Section 3.2.2b, are also calculated. This spreadsheet will also print the data tables for the liquid effluent section of the annual Radioactive Effluent Release Report.

ODCMI 8 of 39 Rev. 14 I

2.2 Compliance with the Limitations for Gaseous Effluent Releases 2.2.1 Calculation of Gaseous Effluent Monitor Alarm Setpoints To assure compliance with the limitations of Section 3.3.3.a, alarm setpoints are established for the gaseous effluent monitor. These setpoints are calculated or checked annually, or as required by procedure, to confirm that the current setpoints are set correctly for one or two stack blower operation.

The Dry Cask Storage project was completed in September 2012. All spent fuel has been removed from the plant and transferred to the ISFSI. For this reason, noble gas (Kr-85) releases will no longer be analyzed for.

2.2.2 H-3 and Particulates The following mathematical relationship shall be used to implement the limitation for H-3 and Particulates with T 1/2 > 8 days alarm setpoints:

DpT= Zi PiTQPi( x/Q) (2.10)

Where:

DpT = the dose rate to organ T, of an individual at or beyond the EFFLUENT RELEASE BOUNDARY, due to H-3 and particulates with half-lives greater than 8 days. This value is to be less than 1500 mRem/yr.

PiT = the dose parameter for organ T, for radionuclide i, for the inhalation pathway, in mRem-m 3 per pCi-yr.

x/Q= the atmosphere dispersion coefficient in sec/m 3.

Qpi = release rate of nuclide i, in pCi/sec.

2.2.3 Calculation of Release Limits for H-3 and Particulates with Half-Lives Greater than 8 days Since it is impractical to measure instantaneous release rates for radionuclides, the alarm setpoints for radionuclides are expressed in terms of total accumulated activity on sample media for a specified sampling time, AT, which is monitored as pCi by the stack effluent monitor.

Equation 2.11 is used to calculate the release rate limit for all H-3 and particulates with half-lives greater than 8 days. This equation is based on the dose rate to an infant due to inhalation of these radionuclides. In accordance with NUREG-01 33, the infant will always receive the maximum dose rate. The atmospheric dispersion coefficients (x/Q) used are 6.05 E-5 sec/m 3 for the calculation of the FAST alarm setpoint and 3.9 E-6 sec/m 3 for the SLOW alarm setpoint.

ODCM 9 of 39 Rev. 14 I

2.2.4 Alarm Setpoint Calculations for H-3 and Particulates with Half-Lives Greater than 8 days 1500 mRem/yr QPT= TZi (x/Q)(Rpi)PiT(inhalation) (2.11)

Where:

QPT = the maximum allowed total release rate of a typical mixture of radionuclides in pCi/sec conservatively derived from the allowed annual average dose rate to organ T, and very conservative x/Q.

Rp= the ratio of the activity of nuclide i, to the total activity of all nuclides in a typical mixture being released.

x/Q = the atmospheric dispersion coefficient as given above for FAST or SLOW alarm respectively, in sec/m 3.

Resolution of the Pi, term in Equation 2.11 yields:

PiT(inhalation)=(106 pCi/pCi)(BR)(DFAiT) (2.12)

Where:

DFA*T =the inhalation dose factor for an infant, for the ith radionuclide, for organ T, in mRem/pCi.

BR = infant breathing rate, in m 3/yr.

To calculate the alarm setpoint in terms of total pCi deposited on filter sample media, the following equation is used:

Qsa- LowestFsQPT*xAF (2.13)

Where:

Qsa = the activity in pCi (deposited on sample media in sample time AT) which initiates an appropriate alarm in the stack effluent monitor.

QpT = pCi/sec F= stack flow rate, cc/sec AF = total flow through sample media (cc) in sample time AT, corrected to stack gas conditions. AT is normally 7 days.

ODCM 10 of 39 Rev. 14 I

The procedure outlined below is used to calculate the release limits for radionuclides. This will be done at least annually.

NOTE: This procedure is applicable for the determination of either FAST or SLOW alarms by utilizing the appropriate value for x/Q in the equation.

1) Start on Figure 2.2. Enter the date, the alarm setpoint being calculated, (FAST or SLOW) and the appropriate x/Q value to be used.

2) Enter the average release rate for the period, Qp,, in pCi/sec, of each identified radionuclide. At the bottom of the form, compute and enter the sum 7-QPi-

3) In the column labeled Rpi, enter the ratio of the average period release rate of nuclide i, to the average total period release rate, TQpi, for the period.

4) For each organ T, as noted at the top of the form, calculate and enter the value of (x/Q)(Rpj)PiT(inhalation) for each nuclide. PiT(inhalation) values are found on Table 2.1. At the bottom of the column, for each organ, enter the value of 1(x/Q)(Rpi)(Pi,) for that organ.

5) Go to Figure 2.3. Enter the date and the alarm setpoint being determined.

6) Using the equation at the top of Figure 2.3, calculate the release rate limits, QpT, for each organ T.

7) Select the lowest value of QPT, enter at the bottom of Figure 2.3 under the appropriate blower operation. Multiply the QPT number times the total sample flow through the sample media, cc, and divide this by the appropriate blower flow rate, cc/sec, to determine the Qsa, in pCi, and use these as alarm setpoints.

ODCMV 11 of 39 Rev. 14 I

FIGURE 2.2 H-3 AND PARTICULATE GASEOUS RELEASE MONITOR ALARM SETPOINT DETERMINATION Date 3

Alarm being calculated (FAST or SLOW) x/Q sec/m *

• For FAST alarm use x/Q of 6.05 E-5 sec/mr3; for SLOW alarm use x/Q of 3.90 E-6 sec/ M3 .

• • PT(inhalation) values found in Table 2.1.

ODCM 12 of 39 Rev. 14 I

FIGURE 2.3 H-3 AND PARTICULATE GASEOUS RELEASE MONITOR ALARM SETPOINT

SUMMARY

Alarm being calculated (FAST or SLOW) Date 1500 mRem/yr 1 x/Q )(Rpi)PiT PT= *i( (inhalation)

Q= the maximum allowed total release rate of a typical mixture of radionuclides in pCi/sec conservatively derived from the allowed annual average dose rate to organ T, and very conservative x/Q.

- p IfP, halatonl .... '*,X .

Whole Body Bone Liver Thyroid Kidney Lung GI-LLI

• From Table 2.1 I

One-Blower Operation Two-Blower Operation

-'sa - Lowest QptxAF

.sa = Lowest QpxAF 3.304E+7 1.650E+7 Qsa uCi/sec x cc Qsa = uCi/sec x cc 1.650 E+7 cc/sec 3.304 E+7 cc/sec Qsa =

pCi Qsa = pCi Where:

AF = corrected total flow through sample media, cc.

ODCM 13 of 39 Rev. 14 I

TABLE 2.1 INFANT DOSE FACTORS PIT(INHALATION) FOR H-3 AND PARTICULATE GASEOUS RELEASE MONITOR ALARM SETPOINT DETERMINATIONS In Units of mRem-m 3/pCi-yr

'" '*' *;° '"::°'*' ,.-...- .. .. . .. . .. .. . .. . . .......

d~Nilide- ý.W-hole Body ~it-.'Tyod~ "Bn' Kin'L L: ng, .. 7;'**;-;

GL.. *'

H-3 6.47 E+2

• 6.47 E+2 6.47 E+2 6.47 E+2 6.47 E+2 6.47 E+2 CO-60 1.18 E+4 8.02 E+3 4.51 E+6 3.19 E+4 SR-90 2.59 E+6 4.09 E+7 1.12 E+7 1.31 E+5 CS-134 7.45 E+4 3.96 E+5 7.03 E+5 1.90 E+5 7.97 E+4 1.33 E+3 CS-1 37 4.55 E+4 5.49 E+5 6.12 E+5

• 1.72 E+5 7.13 E+4 1.33 E+3 CE-144 1.76 E+5 3.19 E+6 1.21 E+6

• 5.38 E+5 9.84 E+6 1.48 E+5 Values in this table are derived from Tables E-5 and E-10 in App. E of NRC Regulatory Guide 1.109, Revision 1, October 1977.

• No data available.

ODCM 14 of 39 Rev. 14 I

2.2.5 Calculation of Gamma and Beta Air Dose Commitments In accordance with the RECP, the gamma and beta air dose commitments are to be calculated once per calendar quarter and yearly. Equations 2.14 and 2.15 are used to perform these calculations.

To demonstrate compliance with the limitations of Section 3.3.3.c, dose contributions are calculated for H-3, and particulates with half-lives greater than 8 days, identified in gaseous effluents released to unrestricted areas using the methodology presented in NRC Regulatory Guide 1.109, Rev. 1 October 1977.

This methodology takes the form of the following general equation:

DTa(r, ) Z MiTaW( r, e)Qi (2.16)

Where:

DTa(r, e) = the dose commitment to organ T,of an individual in age group a, at a distance r, in sector 6 from the release point, due to the release to the atmosphere of radionuclides in mRem.

W(r, e) = the average dispersion parameter for estimating the dose to an individual at the receptor location (r, 8), for the period of release, in sec/m 3 or m 2 as required by the characteristics of the exposure pathway.

Q0= the total activity of each radionuclide i, in gaseous effluents for the release period of interest, in pCi.

MPi = the dose conversion factor for exposure pathway P to organ T, of an individual in age group a, for each identified radionuclide i. The units of MPiTa are (mRem-m 2)/pCi or (mRem-m 3)/pCi-sec) as required so that the product MPTaW(r, E) is mRem/pCi.

Equation 2.16 may be expanded to the following form where each term is the incremental dose received via one of the three major dose pathways.

DTa(r, 0)= D9a(r, e)+ D'a(r, e)+Dia(r, e) (2.17)

Where the first term on the right is the external dose from direct exposure to activity deposited on the ground plane, the second term is the dose from inhalation of radionuclides in air, and the third term is the dose from ingestion of foods contaminated by atmospheric releases of radionuclides.

ODCM 15 of 39 Rev. 14 I

Applying the methodology of NRC Regulatory Guide 1.109 Rev. 1, equation 2.17 is expanded as follows:

DTa(r, 6)=j M9a Q1(D/Q)(r, 8)+E. M0a Qi( x/Q)(r, 8)

+ .iMivQa ( D/Q )(r, 8)+(M 4aQ 41+M TQT)( x/Q )(r, 8)

M~ a Q1( D/Q )(r, Dm Om/Q Dm Dm (rO8 i M)+(M 4TaQ1 4 +MTTaQT)( x/Q )(r, 8)

M ))+(MDM xDM Z-i Omi Qi( D/Q )(r, 8)(1TQ14+MTTaQT)( x/Q )(r, e)

+ Zi MToa Qi( DLD/Q )(r, 8)+(M1DLTaQl +MTTaQT)(

DL x/Q )(r, 6) (2.18) 4 4 Where:

(x/Q)(r, 8)= the annual average atmospheric dispersion factor for a receptor at the distance r, in sector E, from the release point, in sec/m3. For LACBWR, the value for this term is conservatively taken to be the largest historical (1983-1987) undecayed/undepleted x/Q for a real receptor and is 1.82 E-6 sec/m 3 .

(D/Q)(r, 8)= 1.82 E-9 m 2 . This is based on the relationship D/Q = Vdx/Q where Vd = the deposition velocity in m/sec. Vd is generally < 1 E-3 m/sec for dry deposition of submicron aerosols which may be released from the LACBWR facility during SAFSTOR (Ref.

Whicker, F. W., and Schultz, V., Radioecology: Nuclear Energy and the Environment, Vol. II, CRC Press, Inc., Boca Raton, Florida, 1982.

MGi~a =

1.06 SF DFGj, (1 - exp(-Atb))/Ai and according to R.G. 1.109 the dose to all internal organs (T)for all age groups (a) is taken to be the same as the total body dose.

MAi Ta = 3.17 E-2 BRa DFAia for the ingestion pathway (DV) for produce (non-leafy-vegetables, fruits, and grains):

MDViTa = 1.1 E2 DFIiTa UV"a fg exp(-Aith)(r(1 - exp(-AEite))/YvAEi

+ Bi, (1 - exp(-Aitb))/ PA,)

for all radionuclides except C-14 and H-3.

MDV 14Ta 22 DFI14Ta UVa fg p for C-14 MDVTTa= 12 DFITTa UVa fd/H for tritium ODCM 16 of 39 Rev. 14 1

for the ingestion pathway (Dm) for milk:

MDmiTa = 1.1 E2 DFIiTa Uma Fmi QF exp(-Ai tf)(fp fs(1 - exp(-Aith)) + exp(-Ai th))

x (r(1 - exp(-AEi te))/Yv AE + Bi, (1 - exp(-Ai tb))/PAi) for all radionuclides except C-14 and H-3.

MDm14ra = 22 DF11 4Ta Urma Fmi QF p(exp(-A1 4tf)) for C-14 M~mTTa = 12 DFITTa Uma Fmi QF exp(-ATtf)/H for tritium for the ingestion pathway (DM) for meat:

MDMiTa = 1.1 E2 DFliTa UMa F. QFexp(-AitS)(fpfS(1 - exp(-Aith)) + exp(-Aith))

x (r(1 - exp(-AEite))/YvAEi + Bi, (1 - exp(-Aitb))/PAi) for all radionuclides except C-14 and H-3.

MDM1 4 Ta = 22 DF114Ta UMa Ff14 QF p(exp(-A14t4)) for 0-14 MOMTTa = 12 DFIT*a UMa Fir QF exp(-ArTts)/H for tritium for the ingestion pathway (DL) for leafy vegetables:

MDLiTa = 1.1 E2 D~liTa ULa f, exp(-Aith)(r(1 - exp(-AEite))/YVAEi

+Biv(1 - exp(-A,\tb))/PAi) for all radionuclides except C-14 and H-3.

MDL1 4 Ta = 22 DFI1 4Ta uLa feP for 0-14 MDLTTa = 12 DFITTa ULa fjH for tritium ODCM 17 of 39 Rev. 14 I

The values used for the various parameters in the above equations are those recommended in NRC Regulatory Guide 1.109, Rev. 1, for the maximum exposed individual.

1.0 E6 pCi/PCi DFGiT mRem-m 2/pCi-hr from table E-6 in R.G.

DFAia mRem/pCi inhaled from table E-7 thru E-10 in R.G.

DFA14Ta mRem/pCi inhaled from table E-7 thru E-10 in R.G.

DFATa mRem/pCi inhaled from table E-7 thru E-10 in R.G.

DFIlja mRem/pCi ingested from tables E-11 thru E-14 in R.G.

DFI14Ta mRem/pCi ingested from tables E-11 thru E-14 in R.G.

DFITa mRem/pCi ingested from tables E-1 1 thru E-14 in R.G.

SF = 0.7 dimensionless attenuation factor accounting for shielding by residential structures A hrr1 radiological decay constant for nuclide i.

S=1.31x10 5 Hr period of long-term buildup for activity in soil (nominally 15 yrs) 3.17x1 02 pCi-yr/pCi-sec BRa m3/yr inhalation rate for age group a. Table E-5 in R.G.

1.1x10 2 pCi-yr/pCi-hr UVa kg/yr consumption rate of produce for individual in age group a. Table E-5 of R.G.

fg = 0.76 Dimensionless fraction of produce ingested that is grown in garden of interest.

time delay between harvest of vegetation or crops and ingestion.

=0 for pasture grass by animals

= 2160 for stored feed by animals

= 24 for leafy vegetables by man

= 1440 for produce by man r = 0.2 dimensionless fraction of deposited activity retained on crops, leafy vegetables, or pasture grass.

AEi = Ai + Aw hr-1 effective removal rate constant for radionuclide i from crops.

Aw = .0021 hr1 removal rate constant for activity on plant or leaf surfaces by weathering (- to 14 day half-life) te hr period of crop, leafy vegetable, or pasture grass exposure during growing season.

ODCM 18 of 39 Rev. 14 I

Paar~te mnsin .. Descriptiqn/Sourc.

= 720 for grass-cow-milk-man pathway

= 1440 for crop/vegetation-man pathway Y, kg/m 2 agricultural productivity (measured in wet weight)

= 0.7 for grass-cow-milk-man pathway

= 2.0 for produce or leafy vegetables ingested by man pCi/kg in vegetation per pCi/kg in soil for nuclide i.

,dimensionless Table E-1 in R.G.

P = 240 kg/m 2 effective surface density of soil (dry weight) 22 pCi-yr-m 3/pCi-kg-sec ratio of the total annual release time for C-14 to the p dimensionless total annual time during which photosynthesis occurs with the condition that p< 1.0

= 1.0 for continuous C-14 releases.

12 pCi-g-yr/pCi-kg-sec H = 8.0 g/m3 average absolute humidity of the atmosphere at location (r, 0)

U a m

liters/yr consumption rate of milk for individual in age group a.

urnlites/yrTable E-5 of R.G.

factor for estimation of activity of nuclide i in milk from Fmi day/I that in animal feed (pCi/I in milk per pCi/d ingested by the animal) Table E-1 in R.G.

=50 feed or forage consumption rate (wet weight) by milk Q = 50kg/day cow or beef cattle tf = 48 hr transport time from animal feed-milk-man.

fp = 0.5 dimensionless fraction of the year that animals graze on pasture.

fs= 1.0 dimensionless fraction isopatr of daily feed that is pasture when the animal is on pasture rate ofE-5 meat poultry for individual in u Ma kg/yr consumption age group a. Table of &

R.G.

factor for estimation of activity of nuclide i in meat Ffi day/kg from that in animal feed (pCi/kg in meat per pCi/day ingested by the animal) Table E-1 in R.G.

= 480 hr average time from slaughter of meat animal to consumption of meat ULa kg/yr consumption rate of leafy vegetables for individual in age group a. Table E-5 in R.G.

grown in garden of

=1.0 dimensionless interest. of leafy vegetables fraction ODCM 19 of 39 Rev. 14 I

2.2.6 Calculations of Dose Commitments due to Gaseous Release In accordance with the RECP, the maximum commitment to a MEMBER OF THE PUBLIC from H-3 and all radionuclides in particulate form with half-lives greater than 8 days shall be determined at least quarterly.

To perform this calculation Equation 2.18 has been formatted on a computer-based spreadsheet. The quantity in curies of each nuclide (i) released to the atmosphere from the LACBWR facility during the calendar quarter is entered in the appropriate cell of the spreadsheet. The spreadsheet program calculates and displays the total quarterly dose in mRem to the total body and each organ of an individual in each of four age groups and the cumulative calendar year dose to the total body and each organ. It also determines the maximum exposed organ (and its dose) for each age group each quarter and the dose to the maximum exposed organ in all age groups. The quarterly and cumulative calendar year doses to the maximum exposed organ are compared to the limits and the relation in terms of percent of the limit is displayed. The maximum incremental organ dose received through each of the three major pathways is also determined for each age group each quarter.

ODCM 20 of 39 Rev. 14 I

3.0 RADIOACTIVE EFFLUENT CONTROL PROGRAM 3.1 Program Requirements The Radioactive Effluent Control Program (RECP) shall conform to the guidance of 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. This program shall establish the requirements for monitoring, sampling, and analysis of radioactive gaseous and liquid effluents released from LACBWR to ensure the concentrations in effluents released to areas beyond the EFFLUENT RELEASE BOUNDARY conform to 10 CFR Part 20, Appendix B, Table 2, Columns 1 and 2. It shall provide limitations on the annual and quarterly dose commitment to a MEMBER OF THE PUBLIC from radioactive effluents in conformance with Appendix I of 10 CFR Part 50.

The limitations of operability of gaseous and liquid monitoring instrumentation, including surveillance test and setpoint determination in accordance with Section 2.0, Offsite Dose Calculations, will be included in this program.

Requirements for the Reactor Building Ventilation System, including filtration and elevated stack release of exhausted air is included in Section 3.3.1.

In accordance with provisions of 40 CFR 190, the restrictions and surveillance requirements for total dose to any MEMBER OF THE PUBLIC from all LACBWR related sources and dose pathways are presented in Section 3.4.

3.2 Liquid Effluents 3.2.1 Sampling and Analysis All liquid effluent releases at LACBWR will be in batch form. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analysis, each batch shall be isolated and then thoroughly mixed, to assure representative sampling. The radioactive content of each batch of radioactive liquid waste to be discharged shall be determined in accordance with Table 3.1.

The results of pre-release analyses shall be used in accordance with the Offsite Dose Calculations methodology to assure that the concentration at the point of release is maintained within the limits specified in this RECP.

ODCM 21 of 39 Rev. 14 1

TABLE 3.1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS REQUIREMENTS FOR BATCH RELEASES Ty"of Activity Pricipl Aa~yi~~& ".Satmpling '2Mninm n1i7:

~ Freue Each~. discharge - pirt Principal Gamma emitters(b) Prior to discharge Eacharge prior to Each discharge - prior to 2 Gross Alpha Prior to discharge discharge 3 Tritium Prior to discharge Each discharge 4 Sr-90 and Fe-55 Beta emitters Prior to discharge Quarterly Composite(a)

NOTES:

(a) A composite sample is one made up of individual samples which are proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquid release.

(b) The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Co-60, Cs-134, and Cs-137. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

(c) Methods of calculating the Lower Limits of Detection (LLD) shall be contained in approved I procedures and are calculated in accordance with criteria of NUREG-0473, Rev. 2.

ODCM 22 of 39 Rev. 14 I

3.2.2 Liquid Effluent Release Limitation a) Concentration - the concentration of radioactive material released in liquid effluents at any time to areas beyond the EFFLUENT RELEASE BOUNDARY shall be limited to concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2.

If the concentration of radioactive material released beyond the EFFLUENT RELEASE BOUNDARY exceeds the above limits, restore the concentration to within the above limits without delay.

This limit is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2.

b) Dose - the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to areas beyond the Effluent Release Boundary shall be limited to:

Calendar Quarter Calendar Year

< 1.5 mRem total body < 3 mRem total body

< 5 mRem to any organ < 10 mRem to any organ The cumulative dose contribution from liquid effluent shall be determined at least once per calendar quarter in accordance with Section 2.0, Offsite Dose Calculations. If this calculated dose exceeds the above limits, prepare and submit to the Commission, within 30 days, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions which have been or will be taken to assure that subsequent releases shall be in compliance with the above limits.

This limit is provided to implement the requirements of Sections II.A, IlI.A, IV.A and Annex of Appendix 1,10 CFR Part 50. The dose calculations in Section 2.0 implement the requirement in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

ODCM 23 of 39 Rev. 14 1

3.2.3 Liquid Effluent Instrumentation The following radioactive liquid effluent monitoring instrumentation channels shall be OPERABLE, with their alarm setpoints set to ensure that the limits of Section 3.2.2a, are not exceeded, at all times when releasing liquid radioactive effluents.

" Liquid Radwaste Effluent Line Monitor and;

• Liquid Radwaste Effluent Line Flow Meter The alarm setpoints for this monitor will be determined and adjusted using methodology in Section 2.0, Offsite Dose Calculations.

The radioactive liquid effluent instrumentation is provided to monitor the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents with the alarm setpoints set to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part 20.

a) Surveillance Requirements - each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 3.2.

b) Corrective Action

1) With the Liquid Radwaste Effluent Line Monitor channel alarm/trip-point setpoint less conservative than that required by Section 3.2.2a, immediately suspend the release or change the setpoint so that it is acceptably conservative.

2) With the Liquid Radwaste Effluent Line Monitor NOT OPERABLE, or if its alarm setpoint is found to be less conservative than required, suspend release of liquid radioactive effluent without delay. Effluent releases may be resumed without the Liquid Radwaste Effluent Line Monitor OPERABLE, provided that at least two independent samples are analyzed and that at least two technically qualified members of the staff independently verify the release rate calculations. If the monitor is not operable for more than 30 continuous days, explain in the next annual Radioactive Effluent Release Report.

3) With the flow meter NOT OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.

ODCM 24 of 39 Rev. 14 I

TABLE 3.2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Auce - b  ;

-'tidh Liquid Radwaste Prior to Prior to Quarterly At least once per Effluent Line Monitor discharge discharge (See Note 4) (See Note 1) 18 months (e oe3 (See Note 1) (See Note 3)

Liquid Radwaste At least once per Effluent Line Flow (See Note 2) N/A N/A 18 months Meter (See Note 5)

NOTES:

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that an alarm indication occurs if any of the following conditions exist:

• Instrument indicates measured levels at the alarm setpoint.

• Instrument indicates a downscale (circuit failure) failure.

(2) The CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days in which continuous, periodic, or batch releases are made.

(3) The CHANNEL CALIBRATION shall include the use of a known liquid radioactive source positioned in a reproducible geometry with respect to the sensor. The source will have the gamma emitting radionuclide mixture and activity concentration which would normally be measured by the channel during batch discharges.

(4) Background radiation may be used for the source check.

(5) The CHANNEL CALIBRATION will be in accordance with approved procedures.

ODCM 25 of 39 Rev. 14 I

3.3 Gaseous Effluents 3.3.1 Reactor Buildinq Ventilation Normal air discharge from LACBWR is made as an elevated stack release. Air is swept through the Turbine and Reactor Building and then discharged out the stack where the particulate activity is monitored prior to release.

If at any time a verified FAST alarm condition occurs on either of the stack particulate effluent monitors, filtration of the Reactor Building Ventilation System exhaust SHALL COMMENCE as soon as practicable through a set of HEPA particulate filters to reduce the amount of radioactive particulates being released to the environment.

With Reactor Building Ventilation System exhaust being discharged without filtration during periods of a verified FAST alarm condition on either of the stack particulate effluent monitors, prepare and submit to the Commission within 30 days a Special Report which discusses the circumstances and what action will be taken to prevent a recurrence.

3.3.2 Stack Effluent Sampling and Analyses The radioactive gaseous discharge from LACBWR will be sampled and analyzed in accordance with Table 3.3.

ODCM 26 of 39 Rev. 14 I

TABLE 3.3 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS Continuous(b) Weekly(a) Particulate Sample Principal Gamma Emitters(c)

Continuous(b) Quarterly Particulate Sample Sr-90 Stack Composite Effluent Continuous(b) Weekly(a) Particulate Sample Gross Alpha Monthly Monthly H-3 NOTES:

(a) The filter sample shall be changed at least weekly, and filter analyses shall be completed within seven (7) days.

(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.

(c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Co-60, Zn-65, Cs-134, Cs-137, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable and measurable, together with those of the above nuclides, shall also be analyzed and reported in the annual Radioactive Effluent Release Report.

(d) Lower Limits of Detection (LLD) are determined in accordance with approved procedures and are calculated in accordance with criteria of NUREG-0473, Revision 2.

ODCM 27 of 39 Rev. 14 I

3.3.3 Stack Effluent Release Limitation a) Instantaneous Dose Rate - the dose rate due to radioactive materials released in gaseous stack effluents to areas beyond the EFFLUENT RELEASE BOUNDARY shall be limited to:

• The dose rate limit for H-3 and for all radionuclides in particulate form with half-lives greater than 8 days shall be < 1500 mRem/year to any organ.

The dose rate due to H-3 and for all radioactive materials in particulate form with half-lives > 8 days in gaseous stack effluents shall be determined to be within the above limits in accordance with Section 2.0, Offsite Dose Calculations, by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 3.3.

If the dose rate(s) exceeds the above limits, without delay decrease the release rate to within the above limit(s).

This instantaneous dose rate limit is provided to ensure that the dose rate at any time at the EFFLUENT RELEASE BOUNDARY from gaseous effluents from LACBWR will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table 2, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, outside the EFFLUENT RELEASE BOUNDARY to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR Part 20. For individuals who may at times be within the EFFLUENT RELEASE BOUNDARY, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the EFFLUENT RELEASE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the EFFLUENT RELEASE BOUNDARY to < 500 mRem/year to the total body or to < 3000 mRem/year to the skin. These release rate limits also restrict, at all times, the corresponding organ dose rate above background to an individual via the inhalation pathway to < 1500 mRem/year.

b) Dose from Radionuclides - the dose to a MEMBER OF THE PUBLIC from H-3, and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents released to areas beyond the EFFLUENT RELEASE BOUNDARY shall be limited to:

Calendar Quarter Calendar Year

< 7.5 mRem to any organ < 15 mRem to any organ The cumulative dose contributions shall be determined at least once per calendar quarter in accordance with Section 2.0, Offsite Dose Calculations.

ODCM 28 of 39 Rev. 14 I

With the calculated dose from the release of H-3 and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions which have been taken or will be taken to reduce these releases in gaseous effluents during remaining quarters so that the cumulative dose during each subsequent quarter and during the calendar year will be within the above limits.

This limit is provided to implement the requirements of Sections II.C, Ill.A, IV.A and Annex of Appendix I, 10 CFR Part 50. The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section IIl.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

3.3.4 Instrumentation The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.4 shall be OPERABLE with their alarm setpoints set to ensure that the limits of Section 3.3.3a are not exceeded.

a) Gaseous Effluent Instrumentation Surveillance Requirements -

Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 3.5.

b) Corrective Action

1) With a radioactive gaseous effluent monitoring instrumentation channel alarm setpoint less conservative than that required, declare the channel inoperable or change the setpoint so that it is acceptably conservative.

2) With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION required by Table 3.4. Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

ODCM 29 of 39 Rev. 14 I

TABLE 3.4 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

• • At all times, unless alternate monitoring is available ACTIONS:

A. With 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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

B. With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided continuous collection of samples with auxiliary sampling equipment is initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

ODCM 30 of 39 Rev. 14 I

TABLE 3.5 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES:

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that an alarm indication occurs if any of the following conditions exist:

a. Instrument indicates measured level above the alarm setpoint on one channel.

b. Instrument indicates a failure by a Low Flow and Low Count Rate signal.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that the alarm occurs if the flow instrument indicates measured levels below the minimum and/or above the maximum alarm setpoint.

(3) The CHANNEL CALIBRATION shall be conducted in accordance with approved I procedures.

ODCM 31 of 39 Rev. 14 I

3.4 Total Dose to a Member of the Public The dose equivalent to any MEMBER OF THE PUBLIC due to release of radioactivity and radiation, shall be limited to < 25 mRem to the total body or any organ (except the thyroid, which is limited to < 75 mRem) over a period of one calendar year.

With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the calendar year dose limits specified in Sections 3.2.2b, 3.3.3b, or 3.3.3c, a determination should be made, including direct radiation from Reactor Building and radioactive waste storage tanks to determine if the above limits have been exceeded. If these limits have been exceeded, prepare and submit a Special Report (including an analysis which estimates the radiation exposure to a MEMBER OF THE PUBLIC for the calendar year) to the Director, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding these limits. Ifthe release condition resulting in the excess has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190. Submittal of the Special Report is considered a timely request, and a variance is granted until staff action on the request is complete.

Cumulative dose contributions from liquid and gaseous effluents shall be determined quarterly and annually in accordance with Section 2.0, Offsite Dose Calculations.

Cumulative dose contributions from direct radiation from the reactor containment or radioactive waste storage tanks shall be determined once per year in accordance with Section 4.0, Radiological Environmental Monitoring Program.

This requirement is provided to meet the dose limitations of 40 CFR 190. Whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix 1,10 CFR Part 50, a Special Report will be submitted which describes a course of action which should result in the limitation of dose to a real individual for 12 consecutive months to within the 40 CFR 190 limits.

For conservatism, for compliance with this limit, the maximum total dose to any MEMBER OF THE PUBLIC will be assumed to be the sum of the maximums from each dose pathway even though the actual maximally exposed individual for each of the pathways could not be the same person.

The maximum potential dose to a MEMBER OF THE PUBLIC from direct radiation from the Reactor Building and radioactive waste storage tanks is determined by TLD dosimeters located at various locations around the perimeter of the LACBWR access controlled area and the EFFLUENT RELEASE BOUNDARY for the environmental monitoring program. For compliance with this limit, the actual maximum possible exposure to an actual MEMBER OF THE PUBLIC from direct radiation may be determined from maximum possible exposure times relative to the continuous exposure dose measured by the TLDs. Conservative maximum possible exposure times will be determined by actual observation of the areas of interest by LACBWR management and/or security personnel.

ODCM 32 of 39 Rev. 14 1

3.5 Radioactive Effluent Control Reporting Requirements 3.5.1 Radioactive Effluent Release Report Paragraph (a)(2) of 10 CFR 50.36a, "Technical Specifications on Effluents from Nuclear Power Reactors," requires that a report be made to the Commission annually. The report shall specify the quantity of each of the principal radionuclides released to unrestricted areas by liquid or gaseous effluents during the previous year. With the exception of the collection of hourly meteorological data, the information submitted shall be in accordance with Appendix B of Regulatory Guide 1.21 (Revision 1) dated June 1974 with data summarized on at least a quarterly basis. The Radioactive Effluent Release Report shall be submitted by March 1 of each year.

This same report shall include an assessment, performed in accordance with the ODCM, of radiation doses to members of the public from radioactive liquid and gaseous effluents released beyond the effluent release boundary. This report shall contain any changes made to the ODCM during the previous twelve months.

3.5.2 RECP Non-Conformance Reportinq

1) With the Radiological Effluent Control Program not being conducted as specified in Sections 3.2.2b, 3.3.1, 3.3.3b, and 3.4; prepare and submit to the Commission, within 30 days, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions which have been or will be taken to assure that subsequent releases shall be in compliance with the stated limits.

Refer to Sections 3.2.2b, 3.3.1, 3.3.3b, and 3.4 for specific information.

ODCMV 33 of 39 Rev. 14 I

4.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.1 Program Requirements The Radiological Environmental Monitoring Program (REMP) shall conform to the guidance of Appendix I to 10 CFR Part 50. The REMP shall provide the requirements for monitoring, sampling, analyzing, and reporting radiation and radionuclides in the environment resulting from the LACBWR facility and/or its effluents. These requirements have been established to ensure the measurements of radiation and of radioactive material in potential exposure pathways to MEMBERS OF THE PUBLIC are performed. Various environmental samples will be taken within the area surrounding LACBWR and in selected controlled or background locations. An Interlaboratory Comparison Program shall be established to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in the environmental sample matrices are performed, as part of quality control for environmental monitoring.

The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from plant effluents. This monitoring program theory supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental samples are performed to demonstrate that the results are reasonably valid.

4.2 REMP Description Radiological environmental monitoring samples will be collected and analyzed in accordance with Table 4.1. The specific sample locations are listed in HSP-03.01. Section 3 of the Health and Safety Procedures (HSPs) shall contain procedures to provide specific guidance to the HP technicians in the collection and analysis of each environmental sample.

ODCM 34 of 39 Rev. 14 I

TABLE 4.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

,~Xp-sure Pathwaýy' Nri~ber2i"of", ,"'igj-

-S cl~cin~rqueiyo and omSi'mpde Samples Type'an.~ b, na y.

AIRBORNE Three (3) Continuous operation of 1) Analyze each filter for gross beta radioactivity > 24 PARTICULATES sampler with sample hours following filter change. Perform gamma isotopic collection as required by dust analysis on each sample when gross beta activity is >

loading, but at least weekly. 10 times the control sample (La Crosse).

2) A composite of particulate filters from each location will be gamma analyzed at least once per quarter.

2. DIRECT Eight (8) At least semiannually 1) Gamma dose - at least semi-annually.

RADIATION

3. WATERBORNE Two (2) Monthly 1) Gamma isotopic analysis monthly on each sample.

(River Water) 2) Tritium analysis on composite sample from each location quarterly.

4. RIVER SEDIMENT Two (2) Semi-annually 1) Gamma isotopic analysis on each sample.

5. INGESTION

a. Fish One (1) sample of At least semi-annually 1) Gamma isotopic analysis of the edible portions of two (2) different each sample.

species in area important as a recreational or commercial species.

b. Milk As Determined Obtain sample ifabnormal 1) Gamma isotopic analysis on each sample.

stack particulate release occurs.

c. Vegetation As Determined Obtain sample if abnormal 1) Gamma isotopic analysis of the edible portion of stack particulate release each composite sample.

occurs.

NOTE: (a) Exact sample locations are listed in HSP-03.01.

ODCM 35 of 39 Rev. 14 I

4.3 REMP Lower Limits of Detection (LLD)

The sampling techniques and counting equipment used for the analysis of samples collected as requirements of the REMP will meet LLDs calculated in accordance with criteria of NUREG-0473, Rev. 2. LACBWR's LLDs are calculated as follows and are essentially the same as those found in NUREG-0473, Rev. 2. Table 4.2 lists these values.

4.3.1 Calculation of Lower Limits of Detection:

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):

LLD= 4.66 Sb E x V x2.22 x y x exp (-AAt)

Where:

LLD = priori lower limit of detection as defined above (as picocurie per unit mass or volume).

Sb = standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute).

E= counting efficiency (as counts per gamma).

V= sample size (in units of mass or volume).

2.22 = number of transformations per minute per picocurie.

Y= gamma abundance for isotope of interest.

A= radioactive decay constant for the particular radionuclide.

At = elapsed time between sample collection (or end of the sample collection period) and time of counting.

ODCM 36 of 39 Rev. 14 I

TABLE 4.2 ENVIRONMENTAL SAMPLE ANALYSES LOWER LIMITS OF DETECTION (LLD) VALUES OZ.4 Gross Beta 6 1 E-2 H-3 3500(2000)(a)

Mn-54 15 130 Co-60 15 130 Zn-65 30 260 Cs-134 15 5 E-2 130 15 150 Cs-137 18 6 E-2 150 18 180 NOTE: (a) For drinking water.

ODCM 37 of 39 Rev. 14 I

4.4 Interlaboratorv Comparison Proqram An Interlaboratory Comparison Program will be established to ensure that the analyses being performed to comply with the REMP is accurate. A suitable offsite laboratory will be used to supply NIST traceable or equivalent standard spiked sample media for analysis. The offsite laboratory will supply a report to DPC of the comparison results. The Interlaboratory Comparison Program will be conducted annually. The results of this comparison will be included in the Radiological Environmental Monitoring Report.

4.5 Radiological Environmental Monitoring Reporting Requirements 4.5.1 Radiological Environmental Monitoring Report The Radiological Environmental Monitoring Report shall be submitted annually to the Administrator of the Regional Office of the NRC. This report shall include summarized and tabulated results, including interpretations and analysis of data trends, of environmental samples taken during the previous calendar year. In the event that some 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 report shall also include the following: a summary description of the Radiological Environmental Monitoring Program, a map of all sampling locations keyed to a table giving distances and directions from the plant, the results of the Interlaboratory Comparison Program, and a discussion of all analyses in which the LLD was not achievable. The Radiological Environmental Monitoring Report shall be submitted by March 1 of each year.

4.5.2 REMP Non-Conformance Reporting

1) With the Radiological Environmental Monitoring Program not being conducted as specified in Table 4.1, prepare and submit to the Commission, in the Radiological Environmental Monitoring Report, a description of the reasons for not conducting the program as required, analysis of the cause of unexpected results, and the plans for preventing a recurrence.

2) With the Interlaboratory Comparisons not being performed, report the corrective actions taken to prevent a recurrence to the Commission in the Radiological Environmental Monitoring Report.

3) With radiological environmental sample analysis in excess of the reporting levels listed in Table 4.3, when averaged over any calendar quarter, prepare and submit to the Commission a Special Report within 30 days, with a description of the reasons for exceeding these reporting levels.

ODCM 38 of 39 Rev. 14 I

TABLE 4.3 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES H-3 20,000 ......

Mn-54 1,000 -- 30,000 --

Co-60 300 -- 10,000 Zn-65 300 -- 20,000 --

Cs-134 30 10 1,000 60 Cs-1 37 50 20 2,000 70 ODCM 39 of 39 Rev. 14 I