Text

Arkansas Nuclear One, Units 1 and 2, Attachment 2 to 0CAN020901, Offsite Dose Calculation Manual, Revision 16
	ML090641023
Person / Time
Site:	Arkansas Nuclear
Issue date:	02/26/2009
From:	; Entergy Operations
To:	; Office of Nuclear Reactor Regulation
References
	0CAN020901
	Download: ML090641023 (200)
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Attachment 2 to OCAN020901 OFFSITE DOSE CALCULATION MANUAL REVISION 16 Offsite Dose Calculation Manual Arkansas Nuclear One Revision 016 Record of Revision Page 1 of 1 Rev. Pages Approved No. Affected (Implemented)

Date 014-00-0 1,-5,-10, 38, 46, 65, 70, 82, 103, 104, 106, 119, 124, 11/30/01 131, 142 (12/4/01)014-01-0 3, 108, 109, 110 6/24/02 (6/24/02)014-02-0 1, 3, 5, 38, 39, 42-48, 103, 104, 106, 107, 108-110 8/3/04 (8/18/04)014-03-0 5, 43, 104, 106, 169, 170, 172 12/1/04 (12/15/04) 016 2, 36, 39, 40, 46, 50 9/17/07 9/26/07 TABLE OF CONTENTS PAGE

1.0 INTRODUCTION

........................................................

5 2.0 LIQUID EFFLUENTS

....................................................

6 2.1 Radioactive Liquid Effluent Monitor Setpoint ....................

6 2.2 Liquid Dose Calculation

......................................

.8 2.2.1 Dose Calculations for Aquatic Foods .......................

8 2.2.2 Dose Calculations for Potable Water ......................

10 2.3 Liquid Projected Dose Calculation.............................

12 3.0 GASEOUS EFFLUENTS

..................................................

12 3.1 Gaseous Monitor Setpoints

....................................

12 3.1.1 Batch Release Setpoint Calculations

............

........ 13 3.1.2 Eberline SPING (Final Effluent)

Monitor Setpoint Calculations

..................................

14 3.2 Airborne Release Dose Rate Effects ...........................

16 3.2.1 Noble Gas Release Rate .................................

16 3.2.2 1-131, Tritium and Particulate Release Dose Rate Effects ......................................

19 3.3 Dose Due to Noble Gases ......................................

19 3.3.1 Beta and Gamma Air Doses from Noble Gas Releases ...............................................

20 3.4 Dose Due to 1-131, Tritium and Particulates in Gaseous 1.Effluents

............................................

20 3.4.1 Total Dose from Atmospherically Released Radionuclide

............................................

22 3.5 Gaseous Effluent Projected Dose Calculation

....................

33 3.6 Dose to the Public Inside the Site Boundary .................

.34 3.6.1 Liquid Releases ........................................

34 3.6.2 Airborne Release........................................

35 TABLE OF CONTENTS (Continued)

4.0 ENVIRONMENTAL

SAMPLING STATIONS -RADIOLOGICAL

.....................

FIGURES Figure 4-1 Radiological Sample Stations (Far Field) ...............

Figure 4-1A Radiological Sample Stations (Near Field) ...............

Figure 4-1B Radiological Sample Stations (Site Map) .................

Figure 4-2 Maximum Area Boundary for Radioactive Release Calculation (Exclusion Areas)...................

PAGE 36 38 39 40 41 TABLES Table 4-1 APPENDICES Appendix 1 1.0 2.0 Environmental Sampling Stations -Radiological

.........Radiological Effluent Controls -Unit 1 .................

DEFINITIONS

...............................................

RELEASE CONTROLS.........................................

2.1 Radioactive

Liquid Effluent Monitoring Instrumentation

..................................

2.2 Radioactive

Gaseous Effluent Monitoring Instrumentation...................................

2.3 Radioactive

Liquid Effluents...

................

2.3.1 Concentration

..............................

2 .3 .2 Dose .......................................

2.3.3' Liquid Radwaste Treatment

..................

2.4 Radioactive

Gaseous Effluents.....................

2.4.1 Dose Rate .....................................

2.4.2 Dose -Noble Gases .........................

2.4.3 Dose -Iodine-131, Tritium, And Radionuclides in Particulate Form ..........

2.4.4 Gaseous

Radwaste Treatment

.................

2.5 Radioactive

Effluents

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

2.5.1 Total

Dose...................................

43 52 53 56 58 64 75 75 81 82 83 83 87 88 89 90 90 TABLE OF CONTENTS (Continued)

PAGE 2.6 Radiological Environmental Monitoring

.............

92 2.6.1 Radiological Environmental Monitoring Program Description

........................

92 2.6.2 Land Use Census ............................

102 2.6.3 Interlaboratory Comparison Program .........

104 3.0 ADMINISTRATIVE CONTROLS .................................

105 3.1 Programs .........................................

105 3.2 Radioactive Effluent Release Report ...............

106 3.3 Annual Radiological Environmental Operating Report ...........................................

108 3.4 Special Reports ...................................

109 4.0 UNIT 1 MAIN STEAM LINE RADIATION MONITORING INSTRUMENTATION

...........................

... ..............

110 Appendix 2 Radioactive Effluent Controls -Unit 2 ..................

113 1.0 DEFINITIONS

...........................................

114 2.0 RELEASE CONTROLS .......................................

117 2.1 Radioactive Liquid Effluent Monitoring Instrumentation

..................................

119 2.2 Radioactive Gaseous Effluent Monitoring-)

Instrumentation

..................................

125 2.3 Radioactive Liquid Effluents

......................

139 2.3.1 Concentration..............................

139 2 .3 .2 Dose .......................................

144 2.3.3 Liquid Radwaste Treatment

...................

145 2.4 Radioactive Gaseous Effluents

.....................

146 2.4.1 Dose Rate ..................................

146 2.4.2 Dose -Noble Gases .........................

150 2.4.3 Dose -Iodine-131, Tritium, And Radionuclides in Particulate Form ..........

152 2.4.4 Gaseous Radwaste Treatment

..................

154 TABLE OF CONTENTS (Continued)

PAGE 2.5 Radioactive Effluents

............................

156 2.5.1 Total Dose ..................................

156 2._6 Radiological Environmental Monitoring

............

158 2.6.1 Monitoring Program .........................

158 2.6.2 Land Use Census ............................

168 2.6.3 Interlaboratory Comparison Program .........

170 3.0 ADMINISTRATIVE CONTROLS .................................

171 3.1 Programs .........................................

171 3.2 Radioactive Effluent Release Report ..............

172 3.3 Annual Radiological Environmental Operating Report ............................................

174 3.4 Special Reports ..................................

175 LICENSE NO. TITLE: PAGE: 4 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016

1.0 INTRODUCTION

The Offsite Dose Calculation Manual (ODCM) provides guidance for making release rate and dose calculations for radioactive liquid and gaseous effluents from Arkansas Nuclear One -Units 1 and 2. The methodology is drawn from NUREG-0133, Rev. 0.Parameters contained within this manual were taken from NUREG-0133 and Regulatory Guide 1.109 except as noted for site specific values. These numbers and the calculational method may be changed as provided for in the Technical Specifications.

The following references are utilized in conjunction this manual concerning the indicated subjects: with the limitations included in Subject ANO-I ANO-2 Process Control Program (PCP)Radioactive Effluent Controls Program Annual Radiological Environmental Monitoring Report Radioactive Effluent Release Report Offsite Dose Calculation Manual NMM RW-105 TS 5.5.4 TS 5.6.2 TS 5.6.3 TS 5.5.1 NMM RW-105 TS 6.5.4 TS 6.6.2 TS 6.6.3 TS 6.5.1

2.0 LIQUID

EFFLUENTS 2.1 Radioactive Liquid Effluent Monitor Setpoint The Radioactive Liquid Effluent Instrumentation Limitation requires that the radioactive liquid effluents be monitored with the alarm/trip setpoints adjusted to ensure that the limits of the radioactive liquid effluent concentration limitations are not exceeded.

These concentrations are for the site. The alarm/trip setpoint on the liquid effluent monitor is dependent upon the dilution water flowrate, radwaste tank flowrate, isotopic composition of the radioactive liquid to be discharged, a gross gamma count of the liquid to be discharged, background count rate of the monitor, and the efficiency of the monitor. Due to the fact that these are variables, an adjustable setpoint is used. The setpoint must be calculated and the monitor setpoint set prior to the release of each batch of radioactive liquid effluents.

The following methodology is used for the setpoint determination for the following monitors.ANO-I: RE-4642 Liquid Radwaste Monitor ANO-2: 2RE-2330 Liquid Radwaste Monitor 2RE-4423 Liquid Radwaste Monitor 1) A sample from each tank (batch) to be discharged is obtained and counted for gross gamma (Cs-137 equivalent) and a gamma isotopic analysis is performed.

2) A dilution factor (DF) for the tank is calculated based upon the results of the gamma isotopic analysis and the Maximum Permissible Concentration (MPC) of each detected radionuclide.

DF is calculated as follows: DF Z. (C,/MPC.)

+ C /MPCN i iTNG TNG where: DF = dilution factor;C. = concentration of isotope "i", (pCi/ml);1 MPC, = maximum permissible concentration of isotope "i'", (from 10 CFR 20, Appendix B, Table II, column 2 in pci/ml);C = total concentration of noble gases (pCi/ml);

and TNG MPC = 2 x 10-4 (pCi/ml) per Appendix 1, Limitation L2.3.I.A and TNG Appendix 2, Limitation L2.3.I.A.3) The dilution water flowrate is normally the number of ANO-I circulating water pumps in operation at the time of release. Each circulating water pump has an approximate flowrate of 191,500 gpm (this flowrate may be reduced due to throttling of circ water pump flow and/or circ water bay configuration).

However, under specific conditions and under strict controls, lower dilution water flowrates utilizing service water and cooling tower blowdown flowrates may be used.LICENSE NO. TITLE: PAGE: 6 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016

4) The theoretical release rate, Fm, of the tank (batch) to be released is expressed in terms of the dilution water flowrate, such that for each volume of dilution water released, a given volume of liquid radwaste may be combined.

This may be expressed as follows: F= DV/DF m where: F = theoretical release rate (gpm);m DV = Dilution volume (gpm). When ANO-I circulating water pumps are running, DV is the number of ANO-I circulating water pumps in operation multiplied by the approximate flowrate of an ANO-I circulating water pump (normally 191,500 gpm) or an indicated flow rate. The minimum total flow rate shall be greater than or equal to 100,000 gpm. Otherwise DV is dilution volume provided by service water and cooling tower blowdown flowrate;and DF = dilution factor as calculated in Step 2 above.NOTE In the above equation, the theoretical release rate (F )approaches zero as the dilution factor increases.

The actual flowrate (F A) will normally be equal to the theoretical release rate for high activity releases.For low activity releases, the theoretical release rate becomes large and may exceed the capacity of the pump discharging the tank. In these cases, the actual release rate may be set to the maximum flowrate of the discharge pump.5) The monitor setpoi'nt is calculated by incorporating the monitor reading prior to starting the release (i.e., background countrate), and a factor which is the amount of increase in the release concentration that would be needed to exceed the radioactive liquid concentration limitation.

The monitor setpoint is expressed as follows: M = A*(K*F /F ) + B L M A where: M = monitor setpoint (CPM);A = allocation fraction for the specific unit. (Typically, these values are set at 0.45, but may be adjusted up or down as needed. However, the total site allocation can not exceed 1.0.)K = monitor countrate (CPM) expected based on the gross activity of the release. (This value is obtained from a graph of activity (pCi/ml) versus output countrate for the monitor (CPM));

F M/FA= number of times the activity would have to increase to exceed the radioactive liquid effluent-concentration limitation; and B= background countrate (CPM) prior to the release.To permit the computer to calculate the setpoint, an equation for the expected countrate (K) is expressed as follows: S A K = Slope

10 + Offset where: Slope = Log of the detector response in CPM Log of activity concentration in OCi/ml S = Gross gamma (Cs-137 equivalent) activity for the tank A (pCi/ml);

and Offset = detector response (CPM) for the minimum detectable sample activity calculated from the calibration data.NOTE I I&C personnel use varying concentrations of Cs-137 i to determine the response curve; therefore, a Cs-137 i equivalent activity must be used to accurately predict I the countrate.

Combining terms, the equation for determining the monitor setpoint may be expressed as follows: SA ML = A * [(Slope

10 + Offset)* F M/F ] + B 2.2 Liquid Dose Calculation The "dose" or "dose commitment" to an individual in the unrestricted area shall be less than or equal to the limits specified in Radioactive Liquid Effluents-Dos6 Appendix 1 and Appendix 2 Limitations.

The dose limits are on a per reactoribasis.

This value is calculated using the Adult as the maximum exposed individual via the aquatic foods (Sport Freshwater Fish) and the potable water pathways.2.2.1 Dose Calculations for Aquatic Foods The concentrations of radionuclides in aquatic foods are assumed to be directly related to the concentrations in water.The equilibrium ratios between the two concentrations are called "bioaccumulation factors".

Two different pathways are calculated for aquatic foods: sport and commercial freshwater fish.The internal dose "d" from the consumption of aquatic foods in pathway "p" to organ "j" of individuals of age group "a" from all nuclides "i" is computed as follows: (See Chapter 4 of NUREG-0133 and Regulatory Guide 1.109-12, equation A-3).-X.t 1p d (r,O,a,j)

= Y. {ll00*e *B.}*M*U*F-'*Q*D p 1 I a i aij The total dose from both aquatic food pathways is then: D(r,O,a,j)

= E d (r,e,a,j)P where: r user-selected distance from the release point to the receptor location, in kilometers.

It may be different from the controlling distance specified for the potable water pathway (0.4 km);0 user-selected sector (one of sixteen 22.5' sectors surrounding the reactor site, designated N, NNE, NE,... etc). This sector may be different from the controlling sector specified for the potable water pathway (S);a user-selected age group: infant, child, teen, adult.It is the same controlling age group used in the potable water pathway (adult);user-selected organ: bone, liver, total body, thyroid, kidney, lung,. GI-LLI. It is the same controlling organ used in the potable water pathway (liver);}= represents the concentration factor stored in the database;NOTE Only one concentration factor is needed to represent the two pathways since sport and commercial use the same bioaccumulation factor for a given pathway.1100 = factor to convert from (Ci/yr)/(ft 3/sec) to pCi/liter; 1 decay constant of nuclide "i" in hr-1;tp environmental transit time, release to receptor;NOTE This value should be set to 0 hours0 days 0 hours 0 weeks 0 months (i.e., no decay correction) for the above equation in orderl to be consistent with the equation presented in Chapter 4 of NUREG-0133.

For maximum individual I dose calculations, this value is set to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , which is the minimum transit time recommended by Regulatory Guide 1.109, Appendix A, 2.b.B. bioaccumulation factor for nuclide "i", in pCi/kg per*1 pCi/liter.

Cesium has a site specific number based on carnivorous and bottom feeder sport fish of 400 pCi/kg per pCi/liter (0CAN048408, dated April 13,.1984); Niobium has a site specific number based upon freshwater fish of 300 pCi/kg per pCi/liter.

M dimensionless mixing ratio (reciprocal of the dilution factor) at the point of exposure;U a annual usage factor that specifies the intake rate a for an individual of age group "a", in kilograms/year.

The program selects this usage factor in accordance with the controlling age group"a" as specified previously by the user;F = average flow rate in ft 3/sec. This value is based on total dilution volume for the quarter divided by time into the quarter;Q, = number of curies of nuclide "i" released; and D .= ingestion dose factor for age group "a", nuclide "i", aij and organ "j", in mrem per pCi ingested.

The program selects the ingestion dose factor according to the user-specified controlling age group "a" and controlling organ "j".2.2.2 Dose Calculations for Potable Water The dose "D" from ingestion of water to organ "j" of individuals of age group "a" due to all nuclides "i" is calculated as follows (See Chapter 4 of NUREG-0133 and NRC Reg. Guide 1.109-12, equation A-2):

NOTE The potable water pathway is used only during the time that the Russellville Water System is using the Arkansas River as a water source. The Russellville Water Works will notify ANO when they are using the Arkansas River as a water source.D (r,0, a,j) = il{ll00*e I*M*u*F-*Q*D a 1 aij where: r user-selected distance (0.4 km) from the release point to the receptor location, in kilometers.

It may be different from the controlling distance selected for the aquatic food pathway;0 user-selected sector; (one of the sixteen 22.50 sectors surrounding the reactor site, designated N, NNE, NE, ... etc.). It may be different from the controlling sector for the aquatic food pathway;a = user-selected age group (infant, child, teen, adult).The same controlling age group is used for all liquid pathways (adult);= user-selected organ (bone, liver, total body, thyroid, kidney, lung, GI-LLI). The same controlling organ is used for all liquid pathways (liver).{} = the expression in brackets represents the concentration factor stored in the database;1100 = factor to convert from (Ci/yr)/(ft 3/sec) to pci/liter; M = dimensionless mixing ratio (reciprocal of the dilution factor) at the point of exposure;X. = decay constant of nuclide "i" in hr-1; and i t = environmental transit time, release -to receptor.NOTE This value is set to 0 hours0 days 0 hours 0 weeks 0 months (i.e., no decay correction) for the above equation to be consistent with the equation presented in Chapter 4 of NUREG-0133.

U = annual usage factor that specifies the intake rate a for an individual of age group "a", in liters/year.

The program selects this usage factor according to the user-specified controlling age group "a";F average flow rate in ft 3/sec; this value is based on total dilution volume for one quarter divided by time into the quarter;Q. = number of curies of nuclide "i" in the release; and I D = ingestion dose factor, for age group "a", nuclide aij"i", and organ "j", in mrem per pCi ingested.

The program selects the ingestion dose factor according to the user-specified controlling age group "a" and controlling organ "j 2.3 Liquid Projected Dose Calculation The quarterly projected dose is based upon the methodology of Section 2.2 and is expressed as follows: D = 92*(DQc + D RP)/T where: D Q = quarterly projected dose (mrem);92 = number of days per quarter;D = cumulative dose for the quarter (mrem);Qc D = dose for current release (mrem); and RP T = current days into quarter;3.0 GASEOUS EFFLUENTS 3.1 Gaseous Monitor Setpoints NOTE.Sections 3.1.1 and 3.1.2 below detail two methods of calculating setpoints at ANO. These methods cover two different sets of monitors of which only one will be in-service at any one time.

3.1.1 Batch

Release Setpoint Calculations 3.1.1.a This section applies to the following gaseous radiation monitors (These releases are also monitored by the SPING monitors in Section 3.1.2): ANO-I RE-4830* Waste gas holdup system monitor RX-9820 Unit 1 Containment Purge SPING ANO-2 2RE-8233 Containment purge 2RX-9820 Unit 2 Containment Purge SPING 2RE-2429*

Waste gas holdup system monitor These monitors provide automatic isolation for the waste gas holdup systems.The setpoints to be used during a batch type of release (i.e., reactor building [containment]

purge, release from the waste gas holdup system or any other non-routine release) will be calculated for each release before it occurs.3.1.l.b The basic methodology for determining a monitor setpoint is based upon the expected concentration at the monitor (CM ). This is in turn based upon the fraction of an MPC assigned to this release point.Batch releases are maintained below the assigned MPC fraction by controlling the release rate. The calculated value of S may not exceed the equivalentý of 1 MPC at site boundary.

If value of S for RX(2RX)-9820 is less than SPING channel 5 high alarm setpoint, then high alarm setpoint may be used as a default value. If value of S for RE-4830 and 2RE-2429 is less than 50,000 counts/min, then 50,000 counts/min may be used as a minimum setpoint.

If value of S for 2RE-8233 is less than 1,000 counts/min, then 1,000 counts/min may be used as a minimum setpoint.S = 1.2 * (CM*K) + (2.0

B)where: S = monitor setpoint (counts/min);

C M Xe-133 equivalent concentration at the monitor (pCi/ml);K conversion factor determined from response curve of monitor (counts/min per pCi/ml,).

This value is 1.0 when calculating S for RX(2RX)-9820.2.0 factor to accommodate random count rate fluctuations;

B background count rate at the monitor (counts/min).

.1.2 = Safety Factor to correct for instrument uncertainties.

LICENSE NO. TITLE: PAGE: 13 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016

3.1.2 Eberline

SPING (Final Effluent)

Monitor Setpoint Calculations 3.1.2.a This section applies to the following gaseous radiation monitors: ANO-I RX-9820 Containment Purge RX-9835 Hydrogen Purge/Emerg.

Pen. Room Vent.RX-9830 Fuel Handling Area RX-9825 Radwaste Area ANO-2 2RX-9820 Containment Purge 2RX-9840 PASS Bldg.2RX-9845 Aux. Bldg. Ventilation 2RX-9835 Hydrogen Purge/Emerg.

Pen. Room Vent.2RX-9830 Fuel Handling Area 2RX-9825 Radwaste Area 2RX-9850 Low-Level Radwaste Storage Building The determination of setpoints for the above monitors is based on an assigned fraction of the MPC of noble gas activity at the site boundary, (Xe-133 equivalent) released from the above release points.The total of these fractions is always less than 1.00. The assigned fractions are based on the vent flow rates, atmospheric dilution-rate, and the ventilation system(s) in operation.

NOTE The fact that an effluent monitor is in alarm does not necessarily mean that radioactive gases are being released at such a rate that the MPC limit is being exceeded.

The alarm would indicate that radioactive gases are being released at a rate that is exceeding the fractional allocation of an MPC allotted to that particular release point. Consideration must be given to the release rate of radioactive gases via all of the release pathways.The initial fractions of an MPC allocated to the release points are given below. The allocations may be changed as needed, to allow for operational transients, but may not exceed a site total of 1.00.

Monitor Monitor Fractional Number Name, Allocation RX-9820 RX-9835 RX-9830 RX-9825 Monitor Number Containment Purge Hydr. Purge/Emerg.

Pen.Rm. Vent.Fuel Handling Area Radwaste Area Monitor Name 0.1000 0.0001 0.1500 0.2000 Fractional Allocation 2RX-9820 Containment Purge 0.1000 2RX-9840 'PASS Bldg. 0.O100 2RX-9845 Aux. Bldg. Ventilation 0.0100 2RX-9835 Hydr. Purge/Emerg.

Pen. 0.0001 Rm. Ventý.2RX-9830 Fuel Handling Area 0.1500 2RX-9825 Radwaste Area 0.2000 2RX-9850 Radwaste Storage Bldg. 0.0100 NOTE I The setpoints to be used during a' batch release (i.e., reactor building [containment]

purge, I release from the waste gas holdup system) will be l I calculated for each release before it occurs.3.1.2.b SPING monitor setpoints may be calculated as follows: Setpoint (pCi/cc) = A

Xe-133 eq (pCi/cc)L ý*l.32l5E-9*TMPC where: A = allocation fraction (the fraction of an MPC at the site boundary (of noble gas Xe-133 eq activity) assigned to the particular release point).;Xe-133 eq = Xenon-133 equivalent concentration; F = discharge flow of the particular' release point in CFM;1.3215E-9

= 2.8317E-2(cm/f) 28E-6(sec/m

3) 7 cf)* 60 (sec/min)

J where: 2.8E-6 = the annual average gaseous dispersion factor (corrected for radioactive decay)as defined in Section 2.3 of the ANO-2 SAR; and TMPC total MPCs at site boundary.3.2 Airborne Release Dose Rate Effects 3.2.1 Noble Gas Release Rate 3.2.1.a To calculate the noble gas release dose rate, the average ground-level concentration of radion'uclide"i" at the receptor location must first be determined from the following equation. (See Regulatory Guide 1.109-2.0 equation B-4).x. (8) 3.17 x 104

Q

DlX/Q(0)1 1 where: x. (0) .= average ground level concentration in 1 pCi/m 3 of nuclide "i" at the user-specified controlling distance in sector 0 (1.05 km);(8) = one of the sixteen 22.50 sectors surrounding the reactor site, designated N, NNE, NE, ... etc. (WNW);3.17xi0 4= number of pCi per Ci divided by the number of seconds/year; Q. = release rate of nuclide "i" in curies/yr and 1 DlX/Q(8) = annual average gaseous dispersion factor (corrected for radioactive decay) in the sector at angle "8" at the receptor location in sec/m 3.This value is 2.8E-6 sec/m 3 for short term releases.The annual dose to the total body and skin due to noble gas can be calculated according to Sections 3.1.2.b and 3.2.1.c.3.2.1.b Annual Total Body Dose Rate The annual average total body dose rate to the maximally exposed individual is calculated as follows: DT(0) = RBPF

S,

Yi [xi(8)

DFBi]

where: T D (0) total body dose rate due to immersion in a semi-infinite cloud of gas at the controlling distance in sector "0", in mrem/yr. The program computes one total body dose rate value for each sector in which the user has specified a controlling distance and reports only the maximum value;0 one of sixteen 22.5' sectors surrounding the reactor site, designated N, NNE, NE, .... etc.;(WNW);RBPF = Reactor Building (Containment)

Purge Factor'-This factor is used to calculate the length of time (fractional duty cycle) that the purge fans will be'in operation.

It is calculated by comparing the highest dose rate (DOSER) to its applicable release limit, taking into account the allocation factor for the release point (RBPF = Allocation*

Limit/DOSER)

.This factor is calculated.only for Unit One and Two Reactor Building (Containment,)

Purges.For all other releases, this factor is set to 1.0;S = dimensionless attenuation factor F accounting for the dose reduction due to shielding by residential structures.

The NRC recommended value is 0.7 (for maximum individual);

x. (0) = average ground-level concentration of.1 nuclide "i" at the receptor location in the sector at angle "O" from the release point, as defined in Section 3.2.1.a; and DFB. = total body dose factor fora semi-1 infinite cloud of radionuclide "i", which includes the attenuation of 5 g/cm 2.of tissue, in mrem-m 3/pCi-yr 3.2.1.c Annual Skin Dose Rate The annual dose rate to the skin of the maximally exposed individual due to noble gases is calculated as follows. (See Regulatory Guide 1.109-20, equation B-9)

D s(0) = RBPF [I.II*S

I(xi (0)*DFY) + Z (xi (0) *DFSi) I F 1 1 1 _J where: D S(0) skin dose due to immersion in a semi-infinite cloud of gas at the user-specified controlling distance in sector "0", in mrem;NOTE The program computes a skin dose value for each I sector in which the user has specified a I controlling distance, but prints out only the maximum value.RBPF Reactor Building (Containment)

Purge Factor as defined in Section 3.2.1.b.1.11 average ratio of tissue to air energy absorption coefficient; S F dimensionless attenuation factor accounting for the dose reduction due to shielding by residential structures.

The value is 0.7 (for maximum individual);

x, () is the average ground-level 1 concentration of nuclide "i" at the receptor location in the sector at angle "0" from the release point, as defined in Section 3.2.1;0 = one of sixteen 22.50 sectors surrounding the reactor site,-designated N, NNE, NE, ... etc., (WNW);DFY gamma air dose factor for a semi-infinite cloud of radionuclide "i", in mrad-m 3/pCi-yr; and DFS. beta skin dose factor for a 1 semi-infinite cloud of radionuclide"i", which includes the attenuation by the outer "dead" layer of skin, in mrem-m 3/pCi-yr.

3.2.2 1-131, Tritium and Particulate Release Dose Rate Effects The annual dose rate to the maximally exposed individual for 1-131, tritium and radionuclides in particulate form with half-lives greater than eight days is calculated as follows: TOT- -DR7 DR TOT RBPF * [RI + DRG + DRM where: RBPF = Reactor Building (Containment)

Purge Factor as defined in Section 3.2.1.b;DRI dose rate to the controlling age group (infant)associated with the inhalation of radioiodines and particulates, as calculated in Section 3.4.1.b;G DR -dose rate from direct exposure to activity deposited on the ground plane, as calculated in Section 3.4.1.a; and M DR -dose rate to the controlling age group (infant) and the controlling organ for ingestion of food (milk), as calculated .in Section 3.4.1.d.Calculation of the annual dose rate considers the infant as the most restrictive age group. The organs that are considered as, contributing to the dose rate are: skin, bone, liver, total body, thyroid, kidney, lung, and GI-LLI. The food pathway for the infant is considered to be from milk only. All three pathways will contribute to the total body dose, while the skin will be affected by only the ground plane pathway. The other organs are affected only by the inhalation and food pathways.3.3 Dose Due to Noble Gases The air dose in unrestricted areas due to noble gases released in gaseous effluents shall be less than or equal to 5 mrad for gamma radiation and 10 mrad for beta radiation for any calendar quarter for each unit. The objective of less than or equal to 10 mrad of gamma radiation and 20 mrad of beta radiation for a calendar year per unit (2.5 mrad and 5 mrad respectively per quarter) should be used for planning releases.

NOTE The following equations have been simplified from equations in NUREG-0133, Rev. 0, in that there are no free-standing stacks at ANO.The equations were further simplified in that there are no long term (i.e., continuous) releases.

The individual stack vents are sampled weekly and are assigned a release period of 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months per sample (i.e., considered as short term (batch) releases).

Individual samples are to be taken for each waste gas tank release and reactor building (containment), purge.3.3.1 Beta and Gamma Air Doses from Noble Gas Releases Using the average ground level concentration of radionuclide"i" at the receptor location calculated in Section 3.2.1.a, the associated annual gamma or beta air dose may be calculated by the following equation. (See Regulatory Guide 1.109-20 equation B-5.)or = iI X(O)*(DFY or DFP)1 0 0 1 1 where: Dy(0) or DP(0) = the gamma or beta air dose for the controlling distance in sector "0". (Only the maximum value is reported), and Y DF. or DF. gamma or beta air dose factors for a uniform semi-infinite cloud of nuclide"i", in mrad-m 3/pCi-yr.3.4 Dose Due to 1-131, Tritium, and Particulates in Gaseous Effluents The calculational methodology for determining the dose to an individual from 1-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released to unrestricted areas as specified in the Appendix 1 and Appendix 2 Limitations is in this section.The child is the controlling age group unless stated otherwise.

The inhalation and ground plane pathways are considered to exist at all locations.

The grass-cow-milk, grass-cow-meat, and vegetation pathways are used where applicable.

It is assumed that iodines are in the elemental form.

A dispersion parameter of 2.8E-6 sec/M 3 (per ANO-2 SAR, Section 2.3.4.4) is used for "w" in the inhalation pathway since the majority of gaseous activity released from the site is within the 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time frame (i.e., reactor building [containment]

purges and waste gas decay tanks).The equation is: TOT G I V L M F D D + D + D + D + D + D where: DTOT= total dose;DG D Dv L D M D F D dose contribution from ground plane deposition as calculated in Section 3.4.1.a;dose contribution from inhalation of radioiodines, tritium, and particulates

(>8 days) as calculated in Section 3.4.1.b;dose contributions from consumption of vegetation (defined as produce) for humans and stored feed for cattle. See Section 3.4.1.c for calculations; dose contributions from consumption of fresh leafy vegetables (defined as garden products) for humans and pasture grass for cattle. See Section 3.4.1.c for calculations; dose contribution from consumption of cow's milk; and NOTE I Consumption by the cow of both stored feeds and pasture I grasses is taken into account when calculating this dose I contribution.

Concentration factors for both food sources I are calculated.

= dose contribution from consumption of meat.NOTE I Consumption by the cow of both stored feeds and pasture I grasses is taken into account when calculating this dose contribution.

Concentration factors for both types of animal 1 are calculated.

LICENSE NO. TITLE: PAGE: 21 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016

3.4.1 Total

Dose from Atmospherically Released Radionuclide After the calculation of the concentration factors from the applicable parts of Section 3.4.1, the maximum individual dose as calculated for controlling age group "a" and controlling organ "j", in sector 0 at the controlling distance "r" is given from: DG(r,O,j,a)

D (r,0,j,a)D V(r,O, j,a) =D L(r,0,j,a)

=M D (r,0,j,a)

=D (r,0,j,a)

=(Section 3.4.1.a)(Section 3.4.1.b)v v I DFI. U C. (r,0)j3a 1 i for ground plane deposition for inhalation for produce for leafy vegetables for cow's milk i i i L L DFI. U aC (r,0)3a a m m DFI. U Cm(r,8)ija a 1 F F DFI. .U C. (r,0)ija a 1 for meat where: a = controlling age group adult);(infant, child, teen, or j controlling organ (bone, liver, total body, kidney, lung, or GI-LLI);thyroid, r = user-selected distance from the release point to the receptor location in a particular sector, in kilometers. (The controlling distance is the same for all airborne pathways, 1.05 km.);0 = one of sixteen 22.50 sectors surrounding the reactor site, designated N, NNE, NE, ... etc., (WNW);DFI.. =13a dose conversion factor for ingestion of nuclide "i", organ "j", and age group "a", in mrem/pCi;NOTE I Values used in these tables are taken from I Tables E-11 through E-14 of Regulatory Guide 1.109. DFI.. is selected according to i3a I the controlling organ and age group as specified I I in the database.,

V L M F U, U, UO , U a a a a ingestion rates for produce, leafy vegetables, cow's milk, and meat, respectively, for individuals in age group ý'a". Values used are taken from Table E-5 of Regulatory Guide 1.109.);V L M F Ci, D., C., D. = concentration of nuclide "i" for Produce, leafy vegetables, cow's milk, and meat, respectively, in pCi/kg or pCi/liter.

The program calculates that maximum individual dose for each sector surrounding the plant in which the user has specified a controlling distance for each of the following pathways:

A) ground plane deposition; B) inhalation and the ingestion of; C) produce;D) leafy vegetables; E) cow's milk; and F) meat.Only the receptor point receiving the maximum dose value is printed.3.4.1.a Dose from Ground Plane Deposition The dose DG from direct exposure to activity deposited on the ground plane is calculated as follows (see Regulatory Guide 1.109-24, equations C-1 and C-2):[-- -Xitb -7 D G(r,O,j,a)

= 1S*1.0 x 1012* I Xi 1 * (1 -e )} )'* DOQ(r,O)

Q.* DFG..where: r user-selected distance from the release point to the receptor location in a particular sector, in kilometers.

The controlling distance is the same for all airborne pathways (1.05 km);o = one of sixteen 22.50 sectors surrounding the reactor site, designated N, NNE, NE ...etc., (WNW);a user-selected age group (infant, child, teen, adult) which is the same controlling age group used for all airborne pathways (child);user-selected organ (bone, liver, total body, thyroid, kidney, lung, GI-LLI) which.is the same controlling organ used for all airborne pathways;= represents the concentration factor stored in the database;S F dimensionless attenuation factor accounting for the dose reduction due to shielding by residential structures.

The value is 0.7 (for maximum individual);

1.0 X 1012 = number of pCi per Ci;K. = decay constant of nuclide "i" in hr-1;1 t b length of time over which the accumulation is evaluated (nominally 15 years which is the approximate midpoint of facility operating life or 1.31 x 105 hours0.00122 days 0.0292 hours 1.736111e-4 weeks 3.99525e-5 months );DOQ(r,O) average relative deposition of the effluent at the receptor location "r" in sector "0", considering depletion of the plume during transport, in m Q = release of nuclide "i" in curies, and 1 DFG.. = open field ground plane dose conversion factor for organ "j" (total body or skin)from radionuclide "i", in mrem-m 2/pCi-hr.The dose factor is selected according to the user-specified controlling age group"a" and controlling organ "j" 3.4.1.b Dose from Inhalation of Radionuclides in Air I The dose D to organ "j" of age group "a" associated via inhalation of radioiodines and particulates is (see Reg. Guide 1.109-25, Equations C-3 and C-4): DI(r,0,j,a)

= 3.17 x 10'

Ra*a i 'I

D2DPX/Q(r,0)

DFAia.where: r *user-selected distance from the release point to the receptor location in a particular sector, in kilometers.

The controlling distance is the same for all airborne pathways (1.05 km);0 one of sixteen 22.50 sectors surrounding the reactor site, designated N, NNE, NE ...etc., (WNW);user-selected organ (bone, liver, total body, thyroid, kidney, lung, GI-LLI) and is the same controlling organ as that used for all airborne pathways;LICENSE NO. TITLE: PAGE: 24 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 a = user-selected age group (infant, child, teen, adult) and is the same controlling age group as that used for all airborne pathways;3.17 x 10' = number of pCi/Ci divided by the number of seconds/year; R = annual air intake for individuals in age a group "a" (in m 3/year). The air intake factor is selected in accordance with the user-specified controlling age group;Q. = release of nuclide "i" in curies;D2DPX/Q(r,O)

= annual average atmospheric dispersion factor of the radionuclide at the receptor location "r" in sector "0" (in sec/M 3) as calculated; and NOTE This includes depletion (for radioiodines and particulates) and radioactive decay of the plume.DFA.. = inhalation dose factor for radionuclide oaa ,organ "j", and age group "a". The inhalation dose factor is selected in, accordance with the user-specified controlling age group "a" and controlling organ " j".3.4.1.c Dose from Nuclide Concentrations in Vegetation NOTE To reduce the computational overhead of the computer, the calculations for dose resulting from nuclide concentrations in forage, produce and leafy vegetables is performed in three steps.First, the concentration factors (CF) are computed and stored in the database.

The concentration factor includes all the parameters that are considered constant for each nuclide and agricultural activity, such as the radioactive decay constant, removal rate constant, exposure time, etc.

Second, the deposition rate from the plume is multiplied by the concentration factor and the nuclide activity to produce the nuclide concentration as follows: C (rO) = CF.

DOQ(rO)

Q.1 1 1 where: V C.(r,O) = concentration of nuclide "i" at the 1 receptor location (r,O);CF. = concentration factor of nuclide "i";DOQ(r,O) = relative deposition of nuclide "i". For the short term dispersion option, DOQ isýreplaced by (F x DOQ), where F is the short term dispersion correction factor;Q. = quantity of nuclide "i" released in curies.1 For carbon-14 and tritium, the nuclide concentration is calculated from the concentration factor times the decayed and depleted X/Q for radioiodines and particulates (D2DPX/Q), times the quantity of nuclide"i" released in curies. For the short term dispersion option., D2DPX/Q is replaced by F x D2DPX/Q, where F is the short term dispersion correction factor.v C (r,O) = CF

D2DPX/Q(r,O)

Q for tritium, and T T T CFV 1 4 (r,O) = CF 1 4

D2DPX/Q(r,O)

Q 1 4 for carbon-14 Third, the nuclide concentrations for a particular pathway (produce, leafy vegetables, cow's milk, and meat) are summed and multiplied by: 1) the ingestion rate for a particular age group and 2) the dose conversion factor: D(r,O,j,a)

= .E (DFI..

U

C.(r,O)l 1 13a a 1 where: r = user-selected distance from the release point to the receptor location in a particular sector, in kilometers (1.05 kin);8 = one of sixteen 22.5' sectors surrounding the reactor site, designated N, NNE, NE... etc., (WNW);

user-selected organ (bone, liver, total body, thyroid, kidney, lung, GI-LLI), and is the same controlling organ as that used for all airborne pathways;a user-selected age group (infant, child, teen, adult), and is the same controlling age group as that used for all airborne pathways;DFI. dose conversion factor for ingestion of 13a nuclide "i", organ "j", and age group "a", in mrem/pCi, according to the controlling organ and age group;U = annual ingestion rate of food in a a particular pathway (kilograms/year or liters/year) for individuals in age group"a", according to the controlling age group; and V C. (r,0) concentration of nuclide "i" at the 1 receptor location (r,O).3.4.1.c.1 Calculating Vegetation Concentration Factors NUREG-0133 calculations for radioiodines and particulate radionuclides (except tritium and carbon-14), the concentration factor of nuclide"i" in and on vegetation is estimated as follows:-x ith CF. = CONST*( r )*e *f yv, where: v CF. = concentration factor of radionuclide 1"i" in vegetation (forage, produce, or leafy vegetables), in m 2-hr/kg;CONST = 1.14 x 108 number of pCi per Ci (1012)divided by the number of hours per year (8760);r = is the fraction of deposited activity retained on crops, leafy vegetables, or pasture grass, from airborne radioiodine and particulate deposition:

r = 1.00 for radioiodines r = 0.20 for particulates Y = agricultural productivity (yield or v vegetation area density), in kg (wet weight)/m 2: Y = 2.0 kg/M 2 for stored animal s feed for grass-animal-man pathways Y = 0.7 kg/M 2 for pasture grass p for grass-animal-man pathways Y 2.0 kg/M 2 for leafy vegetation (fresh) for crop/vegetation-man pathways Y 9 2.0 kg/M 2 for garden produce g (stored vegetables) for crop/vegetation-man pathways X. is the decay constant of nuclide "i" 1 in hr-1;th is a holdup time that represents the time interval between harvest and consumption of the food, in hours: th 0 hours0 days 0 hours 0 weeks 0 months for pasture grass consumed by animals th 2160 hours0.025 days 0.6 hours 0.00357 weeks 8.2188e-4 months for stored feed consumed by animals th 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for leafy vegetables consumed by humans t h 1440 hours0.0167 days 0.4 hours 0.00238 weeks 5.4792e-4 months for produce consumed by humans f = is the fraction of leafy vegetables or produce grown in garden of interest: f 0.76 for the fraction of produce ingested, grown in garden of interest. (This is f in equation C-13 of g Regulatory Guide 1.109)f.

f 1.00 for the fraction of leafy vegetables grown in garden of interest. (This is f, in equation C-13 of Regulatory Guide 1.109)f 1.00 for all other pathways 3.4.1.c.2 Concentration Factor for Carbon-14 For carbon-14, the concentration factor in and on vegetation is estimated as follows (see Regulatory Guide 1.109-26, equation C-8): CFV 1 4 = 2.2 x 10'

p where: CFv 1 4 = concentration factor of carbon-14 in and on vegetation, in m 2-hr/kg; and p is defined as the ratio of total annual release time (for C-14 atmospheric releases) to the total annual time during which photosynthesis occurs (taken to be 4400 hours0.0509 days 1.222 hours 0.00728 weeks 0.00167 months ), under the condition that the value of "p" should never exceed unity. For continuous C-14 releases,"p" is taken to be unity.(Thus, the value of 2.2 x 107 is stored for CFV 1 4 in lieu of a site specific value for "p".')3.4.1.c.3 Concentration Factor for Tritium The concentration factor for tritium in vegetation is calculated from the tritium concentration in air surrounding the vegetation (see Regulatory Guide 1.109-27, equation C-9): v 1.2 X 107 CFT H where: v CFT concentration factor for tritium in vegetation (in m 2-hr/kg); and H absolute humidity at the location of the vegetation, in g/m 3.(The regulatory default value for "H" is 8.0 grams/m 3.)(Thus, the value 1.5 x 106 is stored for CFv in T lieu of a site specific value for "H".)

3.4.1.c.4 Nuclide Concentrations in Produce and Leafy Vegetables The concentrations in and on produce and leafy vegetables of all radioiodine and particulate nulcides "i" (except carbon-14 and tritium) are calculated as follows: C. (r,O) = CFv

DOQ(rýe)

Q. for produce; and 1 1 1 L C C. (r,e) =

DOQ(rO)

Q. for leafy 1 1 1 vegetables where: v CF. = concentration factor of nuclide "i" in 1 produce;L CF. = concentration factor of nuclide "i" in I leafy vegetables;(Note that the difference between CF. and CF.1 1 are the values for. th and fl.)DOQ(r,O) = relative deposition of the radionuclide "i" at the receptor (r,0); and, Q. release of nuclide "i" (in curies).1 The C-14 and H-3 nuclide concentrations are calculated from the concentration factors times the decayed and depleted radioiodine relative deposition D2DPX/Q times the fraction grown in the garden of interest (f = 0.76, f, = 1.0): g C (r,O). = CFv

D2DPX/Q(r,O)

Q

f T T T g C (r,G) = CF L

D2DPX/Q(r,O)

Q

f, for T T .T tritium CV1 4 (r,O) = CFVl 4

D2DPX/Q(r,O)

Q14

f g L L C 14 (r, 0) = CFL 1 4

D2DPX/Q (r, 8)

Q14

f, for carbon-14 3.4.1.d Nuclide Concentration in Cow's Milk The radionuclide concentration in cow's milk is dependent upon the quantity and contamination level of feed consumed by the animal. The concentration is estimated (see Regulatory Guide 1.109-27, equations C-10 and C-lI) as follows: 4 ~t if m *[ v'* Q Cm(r,O) {Fm*Q *e *[f *f *CF + (1 -f )*CF.' + f *(l -f )*CF. ]}*D(rO)

Q m F p s 1 p i. p s I where: C m (r,e) is the concentration of nuclide "i" in 1 cow's milk at the receptor location (r,O), in pCi/liter;

{ the expression in bracketsrepresents the concentration factor. (Note that the concentration factor for cow's milk involves two different vegetation concentration factors (see below).);F = average fraction of the cow's daily intake m of radionuclide "i" (which appears in each liter of milk), in days/liter; Q amount of feed consumed by the cow per day, F in kg/day (wet weight);X. decay constant of nuclide "i" in hr-1;1 t = average transport time of the activity from the feed into the milk and to the receptor (a value of 2 days is assumed);f = fraction of the year that cows graze on p pasture;f = fraction of daily feed that is pasture s grass when the cow grazes on pasture;v CF. vegetation concentration factor of nuclide 1-"i" on pasture grass with the holdup time th = 0 days, in pCi/kg. (Refer to the explanation of the vegetation concentration factor calculation);

VI CF. = vegetation concentration factor of nuclide 1"i" in stored feeds with the holdup time th= 90 days, in pCi/kg. (Refer to the explanation of the vegetation concentration factor calculations);

D(r,O) = relative deposition DOQ(r,O) of the radionuclides, except carbon-14 and tritium. For carbon-14 and tritium, the decayed and depleted dispersion factor D2DPX/Q(r,O) for radioiodines and particulates (in sec/m 3) is used; and Q. is the release of nuclide "i" in curies.1 3.4.1.e Nuclide Concentration in Meat The radionuclide concentration in meat is dependent upon the quantity and contamination level of feed consumed by the animal. The concentration is estimated (see Regulatory Guide 1.109-27, equations C-11 and C-12) as follows: 1s f [f f*Fv+(1 f+V C.(r,O) = {Ff*Q *e *[fp *fs*cF+ (1 -fp)*CF. + f *(l --f )*CF ]}*D(r,O)*Qi fF p s 1 p i p 5 where: NOTE I All parameters used in this pathway are for I beef cattle.Cf (r,O) concentration of nuclide "i" in animal 1 flesh at the receptor location (r,e) in pCi/liter; the expression in brackets represents the concentration factor (Note that the concentration factor for meat involves two different vegetation concentration factors);Ff average fraction of the animal's daily intake of radionuclide "i" which appears in each kilogram of flesh (in days/kg);Q = amount of feed consumed by the animal per, day in kg/day (wet weight);= decay constant of nuclide "i" in hr-1;1 C t average time from slaughter of the animal S to consumption by humans (20.days);

f p fraction of the year that animals graze on p pasture;f = fraction of daily feed that is pasture 5 grass when the animal grazes on pasture;CF. vegetation concentration factor of nuclide 1"i" on pasture grass with the holdup time th = 0 days in pCi/kg. (Refer to the explanation of the vegetation concentration factor calculation);

CF. = vegetation concentration factor of nuclide 1"i" in stored feeds with the holdup time th= 90 days, in pCi/kg. (Refer to the explanation of the vegetation concentration factor calculation);

D(r,O) = relative deposition DOQ(rO) of the radionuclides, except carbon-14 and tritium. For carbon-14 and tritium, the decayed and depleted dispersion factor D2DPX/Q(r,O) for radioiodines and particulates (in sec/M 3) is used;Q. = is the release of nuclide "i" (in curies).1 3.5 Gaseous Effluent Projected Dose Calculation 3.5.1 The quarterly projected dose is based upon the methodology of Sections 3.3 and 3.4, and is expressed as follows: 7D c+D p7 FDQC +DRP D = = I _I *92 LT where: D Qp Quarterly projected dose (mrem);DQC = cumulative dose for the quarter (mrem);D RP = dose for current release (mrem);T = current days into quarter; and 92 = number of days per quarter.

3.6 Dose to the Public Inside the Site Boundary 3.6.1 Liquid Releases*Dose to the public inside the site boundary due to liquid releases will be due to ingestion of fish caught from the discharge canal *and exposure to sediment along the discharge canal bank while fishing.3.6.l.a Dose Due to Ingestion of Fish Dose due to ingestion Of fish is calculated using the methodology given in Section 2.2, Liquid Dose Calculation.

3..6.1.b Dose Due to Exposure to Shoreline Sediments Dose from external exposure to shoreline sediments is calculated from equation A-7 of Regulatory Guide 1.109, Rev. 1, 10/77.U *M *W -ap p R a 110,000 1* IQi*Ti*D *[exp(-X t )]*[l-exp(-it b)]11 apj i aipj p FL where: Rap, = is the total annual dose to organ "j" of individuals of age group "a" from all of the nuclides "i" in pathway in mrem/yr;U is the usage factor that specifies exposure ap time for the maximum individual of age group "a" in hours from Table E-5 of Regulatory Guide 1.109. 67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months for shoreline recreation for a teen was chosen.Adult is the controlling age group for ingestion but the maximum usage factor (teen) was used rather than the adult factor to ensure a conservative dose estimate;M = is the mixing ratio (reciprocal of dilution p factor);W = is the shoreline width factor from Table'A-2 of Regulatory Guide 1.109. The discharge.canal value of 0.1 was chosen;F = is the flow rate of the liquid effluent in ft 3/sec. This was determined by: F(ft 3/sec) = waste volume (gal/yr)*

[134 ga-rftl*F1Yrh*F8 1 hr L1gal IL 860 hj ~60 e Q.1 is the release of nuclide "i" in Ci/yr;T. is the radioactive half-life of nuclide 1"i", in days, from Radioactive Decay Data Tables, Technical Information Center, U. S.Dept. of Energy, 1981;Daipj is the dose factor specific to age group"a", nuclide "i", and organ "j" from Table-E-6 of Regulatory Guide 1.109;X. = is the radioactive decay constant of 1 nuclide "i" in hr-1;t is the average transit time for nuclides to p reach the point of exposure.

A value of 0 hours0 days 0 hours 0 weeks 0 months was chosen due to the proximity of the discharge canal to the plant; and t = is the period of time for which sediment is exposed to the contaminated water in hours.The mid-point of plant operating life, 15 years was chosen per Regulatory Guide 1.109.3.6.2 Airborne Release 3.6.2.a Dose Due to Noble Gases Dose to fisherman at the discharge canal can be calculated by the ratio of dispersion factor for the discharge canal (1.6E-4 sec/M 3 from Table 2-45 SAR, Unit 1, 100 meters downwind in a southerly direction) and the usage factor of 67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months of shoreline recreation to the values used in Section 3.3 of this manual.---1 F-- --1 Dose at discharge canal DT(e)0 1.6E-4 67 hr L2ý.8E61 L 8 6 7 0 hr-T where D (0) is the noble gas dose calculated by Section 3.3.3.6.2.b Dose Due to Iodine, Tritium and Particulates from Gaseous Effluents Section 3.4 calculates total dose for iodine, tritium and particulates as the sum of: TOT G I V L M F D = D + D + D + D + D + D where: D = ground plane deposition; D = inhalation; D = consumption of vegetation; D = consumption of fresh leafy vegetables; m D = consumption of milk; and F D = consumption of meat and poultry;The only contributions relevant to fishing activities at the discharge canal are ground plane deposition and inhalation.

As DG and DI are not independently available, a conservative estimate can be obtained by using the same correction factor developed for noble gas dose to the total dose calculated in Section 3.4 for iodine, tritium and particulates.

Depletion of the plume as it travels downwind can be ignored since the fraction remaining in the plume at 100 meters (discharge canal) and 1046 meters (site boundary) are both greater than 90% according to Figure 3 of Regulatory Guide 1.111.The only activity inside the plant site by members of the public that might contribute a significant dose is fishing along the banks of the discharge canal.Travel alongpublic roads would involve short exposure time and tours of the facility are conducted according to radiological control procedures enforced at the plant to control exposure.

Fishing is the only uncontrolled activity.4.0 ENVIRONMENTAL SAMPLING STATIONS -RADIOLOGICAL Environmental samples will be collected as specified in the Appendix 1 and Appendix 2 Limitations.

The approximate locations of selected sample sites are shown on Figures 4-1, 4-1A and 4-1B for illustrative purposes.Table 4-1 lists the approximate distances and directions of the sample stations from the plant.

FIGURES FIGURE 4-1 SAMPLE STATIONS RADIOLOGICAL I ANrkansas Nuclear One.REMP Sample Locations.(Fai Field)

FIGURE 4-1A RADIOLOGICAL SAMPLE STATIONS FIGURE 4-1B RADIOLOGICAL SAMPLE STATIONS Arkansas Nuclear One REMP Sample Location,..Site. Map FIVI 4-:21 AMAIMUM. AREA 4B0UNDARY FOTZ RADbI6ALTIVE RELEASEý'X ALUATO__;LU'SIOl ARAS)CASES -- t046;METER RADIUS IIDS ENP OqF ,DI SARG E- ýC ANAL (POIT A)POINT A LICENSE NO. TITLE: PAGE: 41 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 TABLES TABLE 4-1 Environmental Sampling Stations -Radiological Sample Station Number: 1 Approximate ,Direction and Distance from Plant: 880 -.0.5 miles Sample Types: 1) Airborne radioiodines

2) Airborne particulates

3) Direct radiation Sample Station Location: The TLD is on a pole near the meteorology tower approximately

0.6 miles

east of ANO.Sample Station Number: 2 Approximate Direction and Distance from Plant: 2430 -0.5 miles Sample Types: 1) Airborne radioiodines

2) Airborne particulates

3) Direct radiation Sample Station Location: IF traveling from ANO, THEN go approximately

0.2 miles

west toward Gate 4. Turn left (at the east end of the sewage treatment plant) and go approximately 0.1 miles. Turn right and go approximately 0.1 miles. The sample station is on the right.Sample Station Number: 3 Approximate Direction and Distance from Plant: 50 -0.7 miles Sample Types: 1) Direct radiation Sample Station Location: IF traveling west on Highway 333, THEN go approximately 0.35 miles from Gate 2 at ANO. TLD is located on utility pole on south side of Highway 333 S.IF traveling east on Highway 333,.THEN go approximately

0.9 miles

from junction of Highway 333 and Flatwood Road. TLD is located on utility pole on south side of Highway 333 S.Sample Station Number: 4 Approximate Direction and Distance from Plant: 181' -0.5 miles Sample Types: 'l) Direct radiation Sample Station Location: Go approximately 0.25 miles south from bridge over intake canal. Turn right onto May Road. Proceed approximately

0.1 miles

west of May Cemetery entrance.

The TLD is located on a utility pole on the south 'side of May Road.

TABLE 4-1 Environmental Sampling Stations -'Radiological Sample Station Number: 6 Approximate Direction and Distance from Plant: 1110 -6.8 miles Sample Types: 1) Airborne radioiodines

2) Airborne particulates

3) Direct radiation Sample Station Location: Go to the Entergy local office which is located off Highway 7T in Russellville, AR (305 South Knoxville Avenue). The sample station is against the east wall of the back lot.Sample Station Number: 7 Approximate Direction and Distance from Plant: 2100 -19.0 miles Sample Types: 1) Airborne radioiodines

2) Airborne particulates

3) Direct radiation Sample Station Location: Turn west at junction of Highway 7 and Highway 27 in Dardanelle, AR. Proceed to junction of Highway 27 and Highway 10 in Danville, AR. Turn right onto Highway 10 and proceed a short distance to the Entergy supply yard, which is on the right adjacent to an Entergy substation.

The sample station is in the southwest corner of the supply yard.Sample Station Number: 8 Approximate Direction and Distance from Plant: Sample Types: 1) Surface water (composite) 1660 -0.2 miles 2) Shoreline sediment 243. -0.9 miles 3) Fish 2120 -0.5 miles Sample Station Location: Plant discharge canal Sample Station Number: 10 Approximate Direction and Distance from Plant: 950 -0.5 miles (intake canal)Sample Types: 1) Surface water (grab)Sample Station Location: Surface water (grab) is collected at plant intake canal.Sample Station Number: 13 Approximate Direction and Distance from Plant: 273' -0.5 miles Sample Types: 1) Broad leaf vegetation Sample Station Location: IF traveling west from ANO toward Gate 4, THEN go approximately

0.4 miles

from ANO. Turn right onto Flatwood Road. Go a short distance.

The sample may be collected from either side of Flatwood Road.

TABLE 4-i Environmental Sampling Stations -Radiological Sample Station Number: 14 Approximate Direction and Distance from Plant: 700 -5.1 Sample Types: 1) Drinking water Sample Station Location: From junction of Highway 7 and Water Works Road, go approximately

0.8 miles

west on Water Works Road. The sample station is on the left at the intake to the Russellville city water system from the Illinois Bayou.Sample Station Number: 16 Approximate Direction and Distance from Plant: 287' -5.5 miles Sample Types: 1) Shoreline sediment 2) Fish Sample Station Location: Panther Bay, located on the south~side of the AR River across from the mouth of Piney Creek.Sample Station Number: 36 Approximate Direction and Distance from Plant: 1530 -0.02 miles Sample Types: 1) Pond water 2) Pond sediment Sample Station Location: The sample station is at the Wastewater Holding Pond on the ANO site east of the discharge canal.Sample Station Number: 55 Approximate Direction and Distance from Plant: 208° -16.5 miles Sample Types: 1) Broadleaf Vegetation Sample Station Location: From Dardanelle, travel south on Highway 27. Go approximately 15..5 miles to the intersection of Highways 27 and 154. The sample station is located at this intersection.

Sample Station Number: 56 Approximate Direction and Distance from Plant: .2640 -0.4 miles Sample Types: 1) Airborne radioiodines

2) Airborne particulates 3J Direct Radiation Sample Station Location: If traveling west from ANO, the sample station is located at'the west end of the sewage treatment plant near the facility blower building.

TABLE 4-1 Environmental Sampling Stations -Radiological Sample Station Number: 57 Approximate Direction and Distance from Plant: 208' -19.5 miles Sample, Types: 1) Drinking water Sample Station Location: Go to Danville and turn left on Fifth Street. Go approximately three blocks. The Danville public water supply treatment facility is located on the left.Sample Station Number: 58 Approximate Direction and Distance from Plant: 220 -0.3 miles Sample Types: 1) Groundwater Sample Station Location: GWM,- 1; North of Protected Area on OCA. South of Outside Fab-Shop, on the west side of access road.Sample Station Number: 59 Approximate Direction and Distance from Plant: 185' -0.1 miles Sample Types: 1) Groundwater Sample Station Location: GWM -2; South of Protected Area on OCA. Near Security Barriers and discharge canal.Sample Station Number: 60 Approximate Direction and Distance from Plant: 2060 -0. 1 miles Sample Types: 1) Groundwater Sample Station Location: GWM -3; South of Protected Area on OCA.' West of Station Number 59 near wood line.Sample Station Number: 61 Approximate Direction and Distance from Plant: 245' -0.1 miles Sample Types: 1) Groundwater Sample Station Location: GWM -4; West of Protected Area on OCA. Edge of parking lot, east'of equipment laydown area.LICENSE NO. TITLE: PAGE: 46 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 TABLE 4-1 Environmental Sampling Stations -Radiological Sample Station Number: 108 Approximate Direction and Distance from Plant: 306' -0.9 miles Sample Types: 1) Direct radiation Sample Station Location: IF traveling from Highway 333, THEN turn south onto Flatwood Road and go approximately 0.4 miles. The TLD is on a utility pole on the right.IF traveling north on Flatwood Road, THEN go approximately

0.4 miles

from, sample station 109. The.TLD is on a utility pole on the left.Sample Station Number: 109 Approximate Direction and Distance from Plant: 2910 -0.6 miles ,Sample Types: 1) Direct radiation Sample Station Location: IF traveling west from ANO toward Gate 4, THEN go approximately

0.4 miles

and turn right onto Flatwood Road. Go approximately 0.2 miles. The TLD is on a utility pole on the right across from the junction of Flatwood Road and Round Mountain Road.Sample Station Number: 110 Approximate Direction and Distance from Plant: 1380 -0.8 miles Sample Types: 1) Direct radiation Sample Station Location: From bridge over intake canal, go south approximately 0.25 miles. Turn left and go approximately 0.25 miles. Turn right on Bunker -Hill Lane. The TLD is on the first utility pole on the left.Sample Station Number: i1 Approximate Direction and Distance from Plant: 1200 -2.0 miles Sample Types: 1) Direct radiation Sample Station Location: From junction of Highway 64 and Highway 326 (Marina Road), go approximately

2.1 miles

on Marina Road. The TLD is on.a utility pole on the left just prior to curve.

TABLE 4-1 Environmental Sampling Stations -Radiological Sample Station Number: 116 Approximate Direction and Distance from Plant: 3180 -1.8 miles Sample Types: 1) Direct radiation Sample Station Location: Go one block south of the west junction of Highway 333 and Highway 64 in London, AR., The TLD is on a utility pole north of the railroad tracks.Sample Station Number: 125 Approximate Direction and Distance from Plant: 460 -8.7 miles Sample Types: 1) Direct radiation Sample Station Location: While traveling north on Highway 7, turn left onto Water Street in Dover, AR. Go one block and turn left onto South Elizabeth Street. Go one block and turn right onto College Street. The TLD is on a utility pole at the southeast corner of the red brick school building, which is located on top of hill.Sample Station Number: 127 Approximate Direction and Distance from Plant: 1000 -5.2 miles Sample Types: 1) Direct radiation Sample Station Location: The TLD is located on Arkansas Tech Campus on N. Glenwood Streetf If traveling south on. State Highway 7 from Interstate 40, turn right on N. Glenwood.

Follow N. Glenwood for approximately 0.6 miles. The TLD is located on a utility pole (with a No Parking, sign on it) across from the northeast corner of Paine Hall.Sample Station Number: 137 Approximate Direction and Distance from Plant: 1510 -8.2 miles Sample Types: 1) Direct radiation Sample Station Location: At junction of Highway 7 and Highway 28 in Dardanelle, AR, go approximately

0.2 miles

on Highway 28. The TLD is on a speed limit sign on the right in front of the Morris R. Moore Arkansas National Guard Armory.Sample Station Number: 145 Approximate Direction and Distance from Plant: 280 -.0.6 miles Sample Types: 1) Direct radiation Sample Station Location: The TLD is located near the west entrance to the RERTC on a utility pole on the north side of State Highway-333.

LICENSE NO. TITLE: PAGE: 48 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 TABLE 1 Environmental Sampling Stations -Radiological Sample Station Number: 146 Approximate Direction and Distance from Plant: 45' -0.6 miles Sample Types: 1) Direct radiation Sample Station Location: The TLD is located on the south end of the east parking lot at the RERTC. The TLD is located on a utility pole.Sample Station Number: 147 Approximate Direction and Distance from Plant: 610 -0.6 miles Sample Types: 1) Direct radiation Sample Station Location: The TLD is located on the west side of Bunker Hill Road, approximately 100 yards from the intersection with State Highway 333.Sample Station Number: 148 Approximate Direction and Distance from Plant: 1220 -0.6 miles Sample Types: 1) Direct radiation Sample Station Location: If traveling east from ANO, turn right on Bunker Hill Road. Travel south for approximately 0.25 miles to the intersection with Scott Lane. The TLD is located on the county road sign post.Sample Station Number: 149 Approximate Direction and Distance from Plant: 156' -0.5 miles Sample Types: 1) Direct radiation Sample Station Location: If traveling south on Bunker Hill Road, turn right on May Road. Travel approximately 0.4 miles. The TLD is located on a "Notice" sign on the north side of May Road.Sample Station Number: 150 Approximate Direction and Distance from Plant: 2050 -0.6 miles Sample Types: 1) Direct radiation Sample Station Location: If traveling south on Bunker Hill Road, turn right on May Road. Travel approximately 0.8 miles. The TLD is located just past the .McCurley Place turn off on the north side of May Road on a utility pole.Sample Station Number: 151 Approximate Direction and Distance from Plant: 225' -0.4 miles Sample Types: 1) Direct radiation Sample Station Location: If traveling west from ANO, turn south on plant road along the est side of the sewage treatment plant. The TLD is located at the end of this road, near the lake on a metal post.

TABLE 4-1 Environmental Sampling Stations -Radiological Sample Station Number: 152 Approximate Direction and Distance from Plant: 338° -0.8 miles Sample Types: 1) Direct radiation Sample Station Location: If traveling west on State Highway 333 from the RERTC, travel approximately 0.7 miles. The TLD is located on the south side of State Highway 333 on a road sign post.Sample Station Number: 153 Approximate Direction and Distance from Plant: 304' -9.2 miles Sample Types: 1) Direct radiation Sample Station Location: Travel State Highway 64 West to Knoxville Elementary School. The TLD is located near the school entrance gate on a utility pole.Sample Station Number: STR -1 Approximate Direction and Distance from Plant: 1200 -0.33 miles Sample Types: 1) Stormwater Runoff Sample Station Location: East side of GSB drainage ditch near lift station.Sample Station Number: STR -2 Approximate Direction and Distance from Plant: 351° -< 0.10 miles Sample Types: 1) Stormwater Runoff Sample Station Location: Inside protected area near Sally Port from drainage ditch along fence.Sample Station Number: STR -3 Approximate Direction and Distance from Plant: 0.20 -0.13 miles Sample Types: 1) Stormwater Runoff Sample Station Location: Outside Protected Area near Sally Port from drainage ditch along fence.Sample Station Number: STR -4 Approximate Direction and Distance from Plant: 102' -0.10 miles Sample Types: 1) Stormwater Runoff Sample Station Location: East side of Oily Water Separator from storm drain.Sample Station Number: STR -5 Approximate Direction and Distance from Plant: 1700 -< 0.10 miles Sample Types: 1) Stormwater Runoff Sample Station Location: West side of discharge canal from storm drain.

APPENDICES Legend BL#.#.# = Limitation Bases Number BS#.#.# = Surveillance Limitation Bases Number L#.#.# = Limitation Number S#.#.# = Surveillance Limitation Number APPENDIX 1 Radioactive Effluent Controls UNIT 1

1.0 DEFINITIONS

OPERABLE -OPERABILITY 1.1 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 water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its functi6n(s) are also capable of performing their related support function(s).

CHANNEL TEST 1.2 A channel test is the injection of an internal or external test signal into the channel to verify its proper response, including alarm and/or trip initiating action, where applicable.

INSTRUMENTCHANNEL CHECK 1.3 An instrument channel check is a verification of acceptable instrument performance by observation of its behavior and/or state; this verification includes comparison of'output and/or state of independent channels measuring the same variable.INSTRUMENT CHANNEL CALIBRATION 1.4 An instrument channel calibration is a test, and adjustment (if necessary), to establish that the channel output responds with acceptable range and accuracy to known values of the parameter which the channel measures or an accurate simulation of these values. Calibration shall encompass the entire channel, including equipment actuation, alarm or trip and shall be deemed to include the channel test.SOURCE CHECK 1.5 A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.LIQUID RADWASTE TREATMENT SYSTEM 1.6 A liquid radwaste treatment system is a system designed and used for holdup, filtration, and/or demineralization of radioactive liquid effluents prior to their release to the environment.

GASEOUS RADWASTE TREATMENT SYSTEM 1.7 A gaseous radwaste treatment system is any system designed and installed

'to reduce radioactive gaseous effluents by collecting gases from radioactive systems and providing for decay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

DEFINITIONS VENTILATION EXHAUST TREATMENT SYSTEM 1.8 A ventilation exhaust treatment system is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be ventilation exhaust treatment systems.PURGE -PURGING 1.9 Purge or purging is the controlled process of discharging air or gas from a confinement to reduce the airborne radioactivity concentration in such a manner that replacement air or gas is required to purify the confinement.

MEMBER(S)

OF THE PUBLIC 1.10 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 the 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 recreational, occupational or other purposes not associated with the plant.EXCLUSION AREA 1.11 The exclusion area is that area surrounding ANO within a minimum radius of 0.65 miles of the reactor buildings and controlled to the extent necessary by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

UNRESTRICTED AREA 1.12 An unrestricted area shall be any area beyond the exclusion area boundary.FREQUENCY NOTATION 1.13 The frequency notation specified for the performance of Surveillance Limitations shall correspond to the following intervals:

P Completed prior to each release S Shift At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D Daily At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W Weekly At least once per 7 days M Monthly At least once per 31 days Q Quarterly At least once per 92 days SA Semiannual At least once per 184 days R Refueling At least once per 18 months N/A Not Applicable BATCH RELEASE 1.14 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.CONTINUOUS RELEASE 1.15 A "Continuous" release is the discharge of liquid waste of a.non-discrete volume, e.g. from a. volume of a system that has an input flow during the continuous release.LICENSE NO. TITLE: PAGE: 55 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016

2.0 RELEASE

CONTROLS LIMITATION L2.0.l The Limitation and Action requirements shall be applicable during the conditions specified for each limitation.

L2.0.2 Adherence to the requirements of the Limitation and/or associated Action.within the specified time interval shall constitute compliance with the Limitation.

In the event the Limitation is restored prior to the expiration of the specified time interval, completion of the action statement is not required.BASES BL2.C.l This limitation establishes the general requirements applicable to Limitations.

Limitations are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a Limitation is not met, the licensee shall follow any remedial Action permitted by the Limitation until the condition can be met. This limitation establishes the applicability statement within each individual limitation as the requirement for when (i.e., in which specified conditions) conformance to the Limitation is required for safe operation of the facility.

The Action requirements establish those remedial measures that must be taken within specified time limits when the requirements of a Limitation are not met. The Action requirements specify the remedial measures that permit continued operation of the facility which is not ,further restricted by the time limits of the Action requirements.

In this case, conformance to the Action requirements provides an acceptable level of safety for unlimited continued operations as long as the Action requirements continue to be met. The specified time limits of the Action requirements are applicable from the point in time it is identified that a Limitation is not met. The time limits of the Action requirements are also applicable when a system or component is removed from service for surveillance testing or investigation of operational problems.

Individual limitations may include a specified time limit for completion of a Surveillance Limitation when equipment is removed from service.In this case, the allowable outage time limits of the Action requirements' are applicable when this limit expires if the surveillance has not been completed.

BL2.O.2 This limitation establishes the general requirements applicable to Limitations.

Limitations are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limitation is not met, the licensee shall follow any remedial Action permitted by the Limitation until the condition can be met. This limitation establishes that noncompliance with a limitation exists when the requirements of the Limitation are not met and the associated Action requirements have not been implemented within the specified time interval.

Trhe purpose of this limitation is to clarify that (1) implementation of the Action requirements constitutes compliance with a limitation and (2) completion of the remedial measures of the Action requirements is not required when compliance with a Limitation is restored within the time interval specified in the associated Action requirements.

SURVEILLANCE LIMITATION S2.0.1 Surveillance Limitations shall be applicable during the conditions specified for individual Limitations unless otherwise stated in an individual Surveillance Limitation.

S2.0.2 Each Surveillance Limitation shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.BASES BS2.0.1 This surveillance limitation establishes the general requirements applicable to Surveillance Limitations.

Surveillance Limitations are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained and that the limitations will be met. This surveillance limitation establishes the requirement that surveillances must be performed during the conditions for which the requirements of the Limitation apply unless otherwise stated in an individual Surveillance Limitation.

The purpose of this surveillance limitation is to ensure that the surveillances are performed to verify the operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is an a specified condition for which the associated Limitations are applicable.

Surveillance Limitations do not have to be performed when the facility is in a condition for which the requirements of the associated Limitation do not apply unless otherwise specified.

BS2.0.2 This surveillance limitation establishes the general requirements applicable to Surveillance Limitations.

Surveillance Limitations are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained and that the limitations will be met. This surveillance limitation establishes the limit for which the specified time interval for Surveillance Limitations may be extended.It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance (e.g.., transient conditions or other ongoing surveillance or maintenance activities).

It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at a refueling frequency and are specified with an 18-month surveillance interval.

It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances.

This surveillance limitation is based upon engineering judgement and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Limitations.

This provision is sufficient to ensure that the reliability ensured through surveillance activities is-not significantly degraded beyond that obtained from the specified surveillance intervals.

2.1 RADIOACTIVE

LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITATION L2.1.1 The radioactive liquid effluent monitoring instrumentation shown in Table 2.1-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of Appendix 1, Limitation L2.3.I.A are not exceeded.APPLICABILITY:

During releases via this pathway.OBJECTIVE:

To provide instrumentation for radioactive liquid releases.ACTIONS: 1. With alarm/trip setpoints less conservative than required by the above limitation, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, until the setpoint is changed to an acceptably conservative value.2. With less than the minimum number of channels operable, take the action shown in Table 2.1-1. Return the instruments to operable status within 30 days or, in lieu of any other report, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected.

Table 2.1-1 Radioactive Liquid Effluent Monitoring Instrumentation Minimum Operable Channels Instrument Applicability Action 1. Liquid radwaste effluent monitor (automatic termination)

2. Liquid radwaste effluent flow monitor 1 During releases via this pathway (DRVTP).1 DRVTP 2 Table 2.1-1 (Continued)

Table Notation ACTION 1 With the number of channels operable less than required, effluent releases may be resumed provided that prior to initiating a release a. At least two independent samples of the tank's contents are analyzed in accordance with Appendix 1, Limitation S2.3.1.A.!

& S2.3.I.A.2;

b. At least two technically qualified members of the facility staff independently verify that the computer input data is correct and;c. At least 2 members of the facility staff independently verify the discharge valve lineup.Otherwise, suspend release of radioactive effluents via this pathway.ACTION 2 With the number of channels operable less than' required, 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.

RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATION S2.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated operable by performance of the channel check, source check, channel calibration, and channel test at the frequencies shown in Table 2.1-2.APPLICABILITY:

OBJECTIVE:

Applies to the instrumentation in the liquid radwaste system that is used to limit the amount of radioactivity released to the environs.To provide surveillance limitations for the instruments required in Appendix 1, Limitation L2.1.1.

Table 2.1-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Limitation Channel Source Channel Channel Check Check Calibration Test Instrument Liquid radwaste effluent line Radiation monitor (automatic termination)

Flow monitor R R Q N/A N/A Notation*During releases via this pathway**A check source is not required if the background activity is greater than the activity of the check source.P Prior to release D Daily Q Quarterly R Every 18 months N/A Not Applicable RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION BASES BL2.1.1 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 for these instruments shall be calculated in accordance with the methods in this manual to ensure that the alarm/trip will occur prior to-exceeding the limits of 10 CFR Part 20.BS2.1.1 To ensure that the instrumentation for the liquid radwaste system is operable.The channel test demonstrates that automatic isolation of this pathway and control room alarm annunciation occur if the instrument indicates measured levels above the trip setpoint.

The channel test also demonstrates that alarm annunciation occurs if any of the following conditions exist: A. Power to the detector is lost.B. The instrument indicates a downscale failure.C. Instrument controls are not set in the operate mode.The initial channel calibration is performed using one or more of the 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. These standards permit calibrating the system over its intended range of energy and measurement range. For subsequent channel calibration, sources thathave been related to the initial calibration are used.

RELEASE CONTROLS 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITATION L2.2.1 The radioactive gaseous effluent monitoring instrumentation shown'in Table 2.2-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of Appendix 1, Limitation L2.4.I.A are not exceeded.APPLICABILITY:

As shown in Table 2.2-1.OBJECTIVE:

To provide instrumentation for radioactive gaseous releases.ACTIONS: 1. With a channel alarm/trip setpoint less conservative than required, declare the channel inoperable.

2. With less than the minimum number of channels operable, take the action shown in Table 2.2-1. Return the instruments to operable status within 30 days or, in lieu of any other report, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected.

Table 2.2-1 Radioactive Gas Effluent Monitoring-Instrumentation Minimum Channels Operable Applicability Parameter Instrument Action 1. Waste Gas Holdup System Noble gas activity monitor (provides alarm and automatic termination of'release) 1 1 During rel~ases via this pathway (DRVTP)Radioactivity System flow 1 Effluent flow monitor 2. Auxiliary Building Ventilation System a) Noble gas activity monitor b) Iodine sampler c) Particulate sampler d) Effluent flow monitor e) Sampler flow monitor DRVTP 2 1 1 1 DRVTP DRVTP DRVTP DRVTP Radioactivity Verify presence of cartridge Verify presence-of filter.System flow 3 4 4 2 1 1>ý DRVTP Sample flow 2 Table 2.2-1 (Continued)

Radioactive Gaseous Effluent Monitoring Instrumentation Minimum Channels Operable Applicability Parameter Instrument

3. Spent Fuel Pool Area Ventilation System a) Noble gas activity monitor b) Iodine sampler c) Particulate sampler d) Effluent flow monitor e) Sampler flow monitor 4. Reactor Building Purge and Ventilation System a) Noble gas activity monitor b) Iodine sampler c) Particulate sampler d) Effluent flow monitor e) Sampler flow monitor 1 1 1 1 1 DRVTP DRVTP DRVTP DRVTP DRVTP Radioactivity Verify presence of cartridge Verify presence of filter System flow Sample flow Radioactivity Verify presence of cartridge Verify presence of filter System flow Sample flow Action 3 4 4 2 2 3, 5 4.1 1 DRVTP DRVTP DRVTP DRVTP DRVTP 1 1 2 2 Table 2.2-1 (Continued)

Radioactive Gaseous Effluent Monitorinq Instrumentation Minimum Channels Operable Instrument

'Applicability Parameter Action 5. Emergency Penetration Room Ventilation System a) Noble gas activity monitor b) Iodine sampler c) Particulate sampler d) Effluent flow monitor e) Sampler flow monitor 1 1 DRVTP DRVTP DRVTP DRVTP Radioactivity Verify presence of cartridge Verify presence of filter System flow 3 4 4 2 1 I DRVTP ,Sample flow 2 Table 2.2-1 (Continued)

Table Notation ACTION 1 With the number of channels operable less than required, the contents of the tank may be released to the environment provided that prior to initiating the release: a. At least two independent samples of the tank's contents are analyzed, and b. At least two technically qualified members of the facility staff independently verify the computer input data, and c. At least 2 members of the facility staff independently verify the correct discharge valve lineup.Otherwise, suspend release of radioactive effluents via this pathway.ACTION 2 With the number of channels operable less than required, 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 .ACTION 3 With the number of channels operable less than required, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 . During the performance of required source checks of radioactive gaseous effluent monitoring instrumentation for plant ventilation systems, these requirements need not be implemented if the instrument is restored to an operable condition within four hours.ACTION 4 With the number of. channels operable less than required, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 2.4-1. A time period of up to four hours is allowed to continue releases via the affected pathway without auxiliary sampling equipment in operation.

If at the conclusion of the four-hour period normal instrumentation is not restored or auxiliary sampling equipment is not in operation, then releases via the pathway shall be immediately suspended.

The allowable four-hour time period is applicable if the instrumentation is discovered to be inoperable for any reason and whenever it is necessary to render the instrumentation inoperable in order to perform the weekly surveillance (i.e., filter changeout) required by ODCM Appendix 1 Table 2.4-1.ACTION 5 When purging the reactor building, immediately suspend purging if less than the required number of monitoring channels are operable.Purging may be resumed provided that prior to initiating the purge: a. At least two independent samples of the reactor building atmosphere are analyzed, and b. At least two technically qualified members of the facility staff independently verify the computer input data.

RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATION S2.2.1 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated operable by performance of the channel check, source check, channel calibration, and channel test at the frequencies shown in Table 2.2-2.APPLICABILITY:

OBJECTIVE:

Applies to the instrumentation in the gaseous radwaste system that is used to limit the amount of activity released to the environs.To provide surveillance limitations for the instruments listed in. Appendix 1, Limitation L2.2.1.

Table 2.2-2 Effluent Monitorina Instrumentation Surveillance Radioactive Gaseous Limitations Channel Functional Test Instrument Channel Check Source**Check Channel Calibration

1. Waste Gas Holdup System a. Noble Gas Activity Monitor (provides automatic termination of release)b. Effluent Flow Monitor 2. Auxiliary Building Ventilation System a. Noble Gas Activity Monitor b. Effluent Flow Monitor c. Sampler Flow Monitor d. Iodine Sampler Cartridge e. Particulate Sampler Filter P N/A M N/A N/A N/A N/A R R Q N/A R Q D*D*W* (1-)W* (1)R R N/A N/A N/A N/A N/A N/A Table 2.2-2 (Continued)

Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Limitations Channel Channel Source** Channel Functional Check Check Calibration Test Instrument

3. Spent Fuel Pool Area Ventilation System a. Noble Gas Activity Monitor b. Effluent Flow Monitor c. Sampler Flow Monitor d. Iodine Sampler Filter e. Particulate Sampler Filter 4. Reactor Building Purge System a. Noble Gas Activity Monitor b. Effluent Flow Monitor c. Sampler Flow Monitor d. Iodine Sampler Filter e. Particulate Sampler Filter D*D*D*W* (1)W* (1)D*D*D*W* (i)M N/A N/A N/A N/A M N/A N/A N/A R R R N/A N/A R R R N/A Q N/A N/A N/A N/A P N/A N/A N/A W* (1) N/A N/A N/A Table 2.2-2 (Continued)

Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Limitations Channel Channel Source** Channel Functional Check Check Calibration Test Instrument

5. Emergency Penetration Room Ventilation System a. Noble Gas Activity Monitor b. Effluent Flow Monitor c. Sampler .Flow Monitor d. Iodine Sampler Filter e. Particulate Sampler Filter D*D*D*W*(l)W* (1)M N/A N/A N/A N/A R R R N/A N/A Q N/A N/A N/A -N/A Table 2.2-2 (Continued)

Table Notation*During releases via this pathway.**A check source is not required if the background activity is greater than the activity of the check source.P Prior to release D Daily W Weekly M Monthly Q Quarterly R Once per 18 Months N/A Not applicable (1) Verify presence of cartridge or filter only.

RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION BASES BL2.2.1 The radioactive gaseous effluent instrumentation is provided to monitor and 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 this manual to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.105.BS2.2.1 To ensure that the instrumentation for the gaseous radwaste system is operable.The channel test demonstrates that control room alarm annunciation occurs if any of the following conditions exist: 1. The instrument indicates measured levels above the alarm/trip setpoint.2. Power to the detector is lost.3. The instrument indicates a downscale.

failure.4. Instrument controls are not set in the operate mode.For the waste gas holdup system noble gas activity monitor, the channel test also demonstrates that automatic isolation of the release pathway occurs if the instrument indicates above the trip .setpoint.

The initial channel calibration is performed using one or more of the 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. These standards permit calibrating the system over its intended range of energy and measurement range. For subsequent channel calibration, sources that have been related to the initial calibration are used.

RELEASE CONTROLS 2.3 RADIOACTIVE LIQUID EFFLUENTS 2.3.1 Concentration LIMITATION L2.3.l.A The concentration of radioactive material released to the discharge canal shall be limited to the concentration specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the total concentration released shall be limited to 2 x 10-4 pCi/ml.APPLICABILITY:

At all times OBJECTIVE:

To ensure that the limits of 10 CFR 20 are met.ACTION: 1. With the concentration of radioactive material released exceeding the above limits, immediately initiate action to restore concentration to within limits and provide notification to the Commission within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In lieu of any other report, prepare and submit a Special Report within 30 days pursuant to Appendix 1, Section 3.4 Limitations.

RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENTS SURVEILLANCE LIMITATION 52.3.1.A.l Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analyses program of Table.2.3-1.$2.3.1.A.2 The results of the radioactivity analyses shall be used in accordance with this manual to assure that the concentrations at point of release are maintained within the limits of Appendix 1, Limitation L2.3.I.A.APPLICABILITY:

At all times OBJECTIVE:

To ensure that the limits of Appendix 1, Limitation L2.3.I.A are met.

Table 2.3-1 RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSES PROGRAM Sampling Minimum Type of Lower Limit Liquid Release Frequency Analyses Activity of Detection Type Frequency Analyses (LLD)(pCi/ml) (a)A. Batch Waste P P Release (d) y isotopic (g) 5 x 10-7 (b)Each Batch Each Batch 1-131 1 x 10-6 P M Dissolved and I x 10-5 One Batch/M Entrained Gases (Gamma Emitters)

_P M H-3 1 X 10-5 Each Batch Composite (c)Gross Alpha 1 x 10-7 P Q Sr-89, Sr-90 5 x 10-8 Each Batch Composite (c)Fe-55 1 x 10-6 B. Continuous Waste Release (e) Continuous D y isotopic (g) 5 x 10-7 (b)(f) (h) (f)1-131 1 x 10-6 M M Dissolved and 1 X 10-5 Grab Sample Entrained Gases (Gamma Emitters)Continuous M H-3 1 x 10-5/(if) (h) Composite (c)Gross Alpha 1 x 10-7 Continuous Q Sr-89, Sr-90 5 x 10-8 (f) (h) Composite (c)Fe-55 1 x 10-6 Table 2.3-1 (Continued)

TABLE NOTATION a. The Lower Limit of Detection (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 radio-chemical separation):

LLD -4.66 Sb E

V

2.22 *Y

exp(-XAt)where LLD is the lower 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 (in 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 picocurie Y is the fractional radiochemical yield (when applicable)

X is the radioactive decay constant for the particular radionuclide At is the elapsed time between sample collection (or end of the sample collection period) and time of counting Typical values of E, V, Y and At should be used in the calculation.

It should be recognized that the LLD is an a Priori (before the fact) limit representing the capability of measurement system and not an a Posteriori (after the fact) limit for a particular measurement.

TABLE 2.3-1 (Continued)

TABLE NOTATION b. For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations.

Under these circumstances, it will be more appropriate to calculate the concentration of such radionuclides using observed ratios with those radionuclides which are measurable.

c. A composite sample is one in which the quantity of liquid sampled is 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 liquids released.d. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling, each batch shall be isolated and mixed to ensure representative sampling.e. A continuous release is the discharge of liquid waste of a non-discrete volume, e.g. from a volume of a system that has an input flow during the continuous release.f. If continuous samples cannot be obtained, then grab samples shall be collected and analyzed at least 1/24 hrs when the specific activity of the secondary coolant is 0.01 pCi/ml IDE (1-131) or at least 1/12 hrs when the specific activity of the secondary coolant is >0.01 pCi/ml IDE (1-131). Grab sample frequency may be increased due to plant conditions.

g. The principal gamma emitters for which the LLD limitation will apply are exclusively the following radionuclides:

Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144.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.

Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.h. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously-in proportion to the rate of flow of the effluent stream. Prior to analysis, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be reperesentative of the effluent release.P Prior to Release D Daily M Monthly Q Quarterly LICENSE NO. TITLE: PAGE: 79 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENTS BASES BL2.3.I.A This limitation is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limit provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a member of the public. The concentration limit for noble gases is based upon the assumption that Xe-133 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.BS2.3.I.A This limitation is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limit provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures greater than the Section II.A design objectives of Appendix I, 10.CFR Part 50, to an individual.

The concentration limit for noble gases is based upon the assumption that Xe-133 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission of Radiological Protection (ICRP) Publication

2.

RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENTS 2.3.2 Dose LIMITATION L2.3.2.A The dose commitment to a member of the public from radioactive material in liquid effluents released from ANO-I to the discharge canal shall be: 1. During any calendar quarter less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ, and 2. During any calendar year less than or equal to 3 mrem to the total body and less than or equal to 10 mrem to any organ APPLICABILITY:

At all times OBJECTIVE:

To ensure that the dose limits of 10 CFR 50, Appendix I, Section IV.A, are met.ACTION: 1. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a Special Report to the Commission within 30 days, pursuant to Appendix 1, Section 3.4 Limitations.

BASES BL2.3.2.A Limitation L2.3.2 provides assurance that releases of liquid effluents will result in concentrations far below the limits of 10 CFR 20.The limitation provides the required operating flexibility and at the same time assures that the release of radioactive material in liquid effluents will be kept "as low as reasonably achievable".

RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENTS 2.3.3 Waste Treatment LIMITATION L2.3.3.A The appropriate parts of the liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid waste prior to their discharge when it is projected that the cumulative dose during a calendar quarter due to liquid effluent releases would exceed 0.18 mrem to the total body or 0.625 mrem to any organ; The provisions of this limitation do not apply to the laundrytanks due to their incompatibility with the radwaste system.APPLICABILITY:

At all times OBJECTIVE:

To assure that the amount of radioactive material in liquid effluents will be "as low as reasonably achievable." ACTION: 1. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report, prepare and submit a Special Report to the Commission within 30 days per Appendix 1, Section 3.4 Limitations.

BASES BL2.3.3;A The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the guide set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

The values of 0.18 mrem and 0.625 mrem are approximately 25%of the yearly design objectives on a quarterly basis. The yearly design objectives are given in 10 CFR 50, Appendix I, Section II.

RELEASE CONTROLS 2.4 RADIOACTIVE GASEOUS EFFLUENTS 2.4.1 Dose Rate LIMITATION L2.4.l.A The dose rate in unrestricted areas (see Figure 4-2) due to radioactive materials released in gaseous effluents from the site shall be: 1. For noble gases: Less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin.2. For iodine-131, for tritium and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.During periods of reactor building purging the dose rate may be averaged over a one hour interval.APPLICABILITY:

At all times OBJECTIVE:

To ensure that the dose rate in unrestricted areas from gaseous effluents will be within the limits of 10 CFR 20.ACTION: 1. With the dose rate(s) exceeding the above limits, without delay restore the release rate to within the above limit(s).SURVEILLANCE LIMITATIONS S2.4.I .A. 1 S2.4. 1.A.2 The dose rate, due to noble gases in gaseous effluents shall be determined in accordance with this manual to be within the limits of Appendix 1, Limitation L2.4.I.A.The dose rate in unrestricted areas, due to iodine-131, tritium, and all radionuclides in particulate form with half-lives greater than 8 days released in gaseous effluents, shall be determined in accordance with this manual to be within the required limits by using the results of the sampling and analyses program, specified Table 2.4-1.in APPLICABILITY:

At all times OBJECTIVE:

To ensure that the dose rate, at any time, in unrestricted areas from gaseous effluents will be within the dose limits Of 10 CFR 20.

TABLE 2.4-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSES PROGRAM-j Minimum Lower Limit of Gaseous Release Type Sampling Analyses Type of Detection (LLD)Frequency Frequency Activity Analyses (CiCi/ml) (a)P P A. Waste Gas Each Tank Each Tank Principal Gamma Emitters (b) 1 x 10-4 (g)Storage Tank Grab Sample B. Reactor Bldg. P P Principal Gamma Emitters (b)- 1 x 10-4 (g)Purge Each Purge Each Purge Grab Sample H-3 1 x 10-6 C. Unit Vents M (c) (d) M Principal Gamma Emitters (b) 1 x 10-4 (g)Grab Sample H-3 1 x 10-6 W (f) 1 x 10-12 (Auxiliary Bldg.) Continuous (e.) Charcoal 1-131 Sample (Spent Fuel Pool WA(f) Principal Gamma Emitters (b) 1 x 10-11 Area Ventilation)

Continuous (e) Particulate (1-131, Others)Sample (Rx Bldg. M 1 x 10-11 Ventilation)

Continuous (e) Particulate Gross Alpha Sample (Emergency Q Penetration Room Continuous (e) Composite l x 10-11 Ventilation)

Particulate Sr-89, Sr-90 Sample 1 x 10-6 Continuous (e) Noble Gas Noble Gases Gross (Xe-133 equiv.)Monitor Beta or Gamma TABLE 2.4-1 (Continued)

TABLE NOTATION a. See definition in Table 2.3-1, Table \otation.b. The principal gamma emitters for which the LLD limitation will apply are exclusively the following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.

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.

Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.c. Tritium grab samples shall be taken from the Reactor Building ventilation exhaust at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.d. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel area, whenever spent fuel is in the spent fuel pool.e. 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 Appendix 1, Limitation L2.4.l.A, L2.4.2.A, and L2.4.3.A.f. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from the sampler).g. For certain radionuclides with low gamma yield or low energies, or for certain radionuclides mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these circumstances, the LLD may be increased inversely proportional to the magnitude of the gamma yield (i.e., (1 x 10- /I), where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated .in this manner for a specific radionuclide, be greater than 10% of the MPC value specified in 10 CFR 20, Appendix B, Table II, Column 1.P Prior to Release W W~ekly M Monthly Q Quarterly RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS Dose Rate BASES BL2.4.l.A This limitation is provided to ensure that, at any time, the dose rate due to gaseous effluents from all units on the site will be within the limits of 10 CFR 20 for unrestricted areas.This limitation applies to the release of gaseous effluents from all reactors at the site.BS2.4.l.A.l/2 This limitation provides for sampling and analyses to ensure that Appendix 1, Limitation L2.4.I.A is met.LICENSE NO.075 TITLE: OFFSITE DOSE CALCULATION MANUAL PAGE: REV: 86 of 175 016 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS 2.4.2 Dose -Noble Gases LIMITATIONS L2.4.2.A The dose due to noble gases released in gaseous effluents from ANO-I to unrestricted areas (see Figure 4-2) shall be: 1. During any calendar quarter, less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and 2. During any calendar year, less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

APPLICABILITY:

At all times OBJECTIVE:

To ensure that the design objective doses of 10 CFR 50, Appendix I, Section IV.A, are not exceeded.ACTION: 1. With the calculated dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a Special Report to the Commission within 30 days, pursuant to Appendix 1, Section 3.4 Limitations.

BASES BL2.4.2.A Limitation L2.4.2.A implements the design guides specified in 10 CFR 50, Appendix I, Section II, and the limiting condition for operation as set forth in Section IV.A of Appendix I.The limitations provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A, Appendix I, to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." 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, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20[10 CFR Part 20.106(b)].

For individuals who may at times be within the exclusion area boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary.

RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS 2.4.3 Dose -Iodine-131, Tritium, and Radionuclides in Particulate Form LIMITATIONS L2.4.3.A The dose to a member of the public from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from ANO-I to unrestricted areas (see Figure 4-2) shall be: 1. During any calendar quarter, less than or equal to 7.5 mrems to any organ, and 2. During any calendar year, less than or equal to 15 mrems to any organ.APPLICABILITY:

At all times OBJECTIVE:

To ensure that the dose limits of 10 CFR 50, Appendix I, Section IV.A, are met.ACTION: 1. With the calculated dose from the release of iodine-131, tritium and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report, prepare and submit a Special Report to the Commission within 30 days, pursuant to Appendix 1, Section 3.4 Limitations.

BASES BL2.4.3.A Limitation L2.4.3.A implements the design guides set forth in 10 CFR 50, Appendix I, Section II.C, and the limiting conditions for operation as set forth in Appendix I,Section IV.A.This limitation provides 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 reasonably achievable".

RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS 2.4.4 Gaseous Radwaste Treatment LIMITATIONS L2.4.4.A Ventilation exhaust treatment systems shall be used to reduce radioactive materials in gaseous waste prior to discharge when the prodjected doses due to gaseous effluent releases from ANO-I to unrestricted areas (see Figure 4-2) would exceed 0.625 mrad for gamma radiation and 1.25.mrad for beta radiation over a calendar quarter; or when the projected doses due to iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days would exceed 1.0 mrem to any organ over a calendar quarter.L2.4.4.B When degasifying the reactor coolant system, the gaseous radwaste treatment system shall be utilized to process the degassing effluent to reduce the concentration of radioactive materials prior to discharge when the projected doses due to gaseous effluent releases from ANO-l. to unrestricted areas (see Figure 4-2) would exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation.

over a calendar quarter.APPLICABILITY:

At all times OBJECTIVE:

To assure that the amount of radioactive material in gaseous effluents is "as low as reasonably achievable." ACTION: 1. With gaseous waste being discharged without treatment and in excess of the above limits, in lieu of any other report, prepare and submit to the Commission within 30 days a Special Report, per Appendix 1, Section 3.4 Limitations.

BASES BL2.4.4.A/B The requirement 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 reasonably achievable." The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

The values 0.625 mrad, 1.25 mrad, and 1.0 mrem are approximately 25% of the yearly design objectives on a quarterly basis. The yearly design objectives are given in Appendix 1, Limitations L2.4.2.A and L2.4.3.A.

RELEASE CONTROLS 2.5 RADIOACTIVE EFFLUENTS 2.5.1 Total Dose LIMITATIONS L2.5.I.A The calculated doses from the release of radioactive materials in liquid or gaseous effluents shall not exceed twice the limits of Appendix 1, Limitation L2.3.2.A, L2.4.2.A, or L2.4.3.A.APPLICABILITY:

At all times OBJECTIVE:

To ensure that the limits of 40 CFR 190 are not exceeded.ACTION: 1. With the calculated doses exceeding the above limits, prepare and submit a Special Report pursuant to 10 CFR Part 20.405C.2. If the limits of 40 CFR 190 have been exceeded, obtain a variance from the Commission to permit further releases in excess of 40 CFR 190 limits. A variance is granted until staff action on the request is completed.

SURVEILLANCE LIMITATION S2.5.I.A Cumulative dose contributions and dose projections for liquid effluents and for gaseous effluents shall be determined in accordance with this manual at least once per 31 days.APPLICABILITY:

At 'all times OBJECTIVE:

To ensure that the requirements of 10 CFR 50, Appendix I, Section III.A are met.

RELEASE CONTROLS RADIOACTIVE EFFLUENTS Total Dose BASES BL2.5.I.A This limitation is provided to meet the dose limits of 40 CFR 190 that have now been incorporated into 10 CFR Part 20. The limitation requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites, containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level.The Special Report will describe a course of action that should result in limiting the annual dose to a member of the public to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities within a radius of 8 km must be considered.

If the dose to any member of the public is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed.

The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limits in 10 CFR 20, as addressed in Appendix 1, Section 2.3 & 2.4 Limitations.

An individual is not considered to be a member of the public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.BS2.5.1.A limits of L2 .4.4 .A, These calculations provide the dose values to be compared to the Appendix 1, Limitations L2.3.2.A, L2.3.3.A, L2.4.2.A, L2.4.3.A, and L2.5.I.A.LICENSE NO. TITLE: PAGE: 91 of 175 075 OFFSITE DOSECALCULATION MANUAL REV: 016 RELEASE CONTROLS 2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING

2.6.1 Radiological

Environmental Monitoring Program Description LIMITATION L2.6.l.A The radiological environmental monitoring samples shall be collected pursuant to. Table 2.6-1 and shall be analyzed pursuant to the requirements of Tables 2.6-1 and 2.6-2. The sample locations shall be listed in Table 4-1.APPLICABILITY:

Applies at all times..OBJECTIVE:

To provide information( on the radiological effects of station operation on the environment.

ACTION: 1. With the radiological environmental monitoring program not being conducted as specified in Table 2.6-1, prepare and submit to the Commission in the Annual Radiological Environmental Operating Report a~description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

2. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at one or more of the locations specified in Table 2.6-1 exceeding the limits of Table 2.6-3 when averaged over any calendar quarter, prepare and submit to. the Commission, within 30 days from the end of the affected quarter, a Special Report which includes an evaluation of any release conditions, environmental factors or. other aspects which caused the limits of Table 2.6-3 to be, exceeded, and defines the actions taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than.the calendar year limits of Appendix 1, Limitations L2.3.2.A and L2.4.2.A.'When more than one of the radionuclides in Table 2.6-3 are detected in the sampling medium, this Special Report shall be submitted if: Concentration (1) + Concentration (2) + ... > 1.0 reporting level (1) reporting level (2)

RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING Radiological Environmental Monitoring Program Description SURVEILLANCE LIMITATIONS (Continued)

When radionuclides other than those in Table 2.6-3 are detected and are the result of plant effluents, this Special Report shall be submitted if the potential annual dose to a member of the public is equal to or greater than the calendar year limits of Appendix 1, Limitations L2.3.2.A and L2.4.2.A.

This 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.3. With milk or fresh leafy vegetable samples unavailable from any of the sample locations required by Table 2.6-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the causes of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Radioactive Effluent Release Report and also include in the report a revised Table 4-1 reflecting the new location(s).(I TABLE 2.6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample 1. AIRBORNE a. Radioiodine and Particulates Number of Sample Location(a) 4 Locations 2 Samples close to Site Boundary, in (or near) different sectors with the highest calculated annual average groundlevel D/Q.1 Sample from the vicinity of a community having the highest calculated annual average groundlevel D/Q.1 Sample from a control location 15-30 km.(10-20 miles)distance (c).Sample and Collection Frequency(a)

Continuous operation of sampler with sample collection as required by dust loading but at least once per 14 days.Type and Frequency of Analyses Radioiodine canister.Analyze at least once per 14 days for 1-131.Particulate sampler: Analyze for gross beta radioactivity following filter change(b)

TABLE 2.6-1 (Continued)

Exposure Pathway and/or Sample Number of Sample Location.a)

Sample and Collection Frequency(a)

Type And Frequency of Analyses Gamma dose. Once per 92 days.2. DIRECT RADIATION(d) 24 Locations-16 inner ring stations with two or more dosimeters in each meteororological sector in the general area of the Site Boundary-8 stations with two or mote dosimeters in special interest areas such as population centers, nearby residences, schools, and in 1 -2 areas to serve as control locations.

Once per 92 days Exposure Pathway and/or Sample 3. WATERBORNE Number of Sample Location(a)

TABLE 2.6-1 (Continued)

Sample and Collection Frequency(a)

Type and Frequency of Analyses a. Surface Water b. Drinking Water c. Sediment from Shoreline 1 Indicator location (influenced by plant discharge) 1 Control location (uninfluenced by plant discharge) 1 Indicator location (influenced by plant discharge) 1 Control location (uninfluenced by plant discharge) 1 Indicator location (influenced by plant discharge) 1 Control location (uninfluenced by plant discharge)

Once per 92 days.Once per 92 days.Once per 365 days Gamma isotopic(e) and Tritium analyses once 92 days.1-131, gross beta, gamma isotopic(e) and tritium analyses once per 92 days.Gamma isotopic(e) analysis once per 365 days.

TABLE 2.6-1 (Continued)

Exposure Pathway and/or Sample 4. INGESTION a. Milk b. Fish c. Food Products Number of Sample Location(a) 1 Indicator sample location within 8 km.distant if commercially available.

1 Control sample locati6n at a distant of >8 km. when an indicator exists.1 Sample of commercially and/or recreationally important species in vicinity of plant discharge.

1 Sample of same species in area not influenced by plant discharge.

1 Sample of broadleaf (edible or non-edible) near the Site Boundary from one of the highest anticipated annual average groundlevel D/Q sectors, if milk sampling is not performed.

1 Sample location of broadleaf vegetation (edible or non-edible) from a control location 15 -30 km. distant, if milk sampling is not performed.

Sample and Collection Frequency(a)

Once per 92 days.Once per 365 days.Three per 365 days.Type and Frequency of Analyses Gamma isotopic(e) and 1-131 analyses once per 92 days.Gamma isotopic(e) on edible portions once per 365 days.Gamma isotopic(e) and 1-131 analyses three times per 365 days.

TABLE 2.6-1 (Continued)

TABLE NOTATION a The ODCM shall include, in a table and figures, specific parameters of distance and direction from the centerline of one reactor, and additional description where pertinent, for each sample location in Table 2.6-1. Refer to NUREG-0133, "Preparation of Radiological Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment-Branch Technical Position, Revision 1, November 1979.Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunctions, every effort shall be made to complete corrective action before the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report.It is recognized that, at times, it may not be possible or practical to continue to obtain samples of the media of choice at the most desirable 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 the next Annual Radioactive Effluent Release Report, identify the cause of the unavailability of samples for that pathway and identify the new location(s) for obtaining replacement samples, and also include in the report a revised figure(s) and table(s) for the ODCM reflecting the new location(s).

b Particulate sample filters should be analyzed for gross beta 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 or water is greater than ten times the yearly mean of control samples for any medium, gamma isotopic analysis-should be performed on the individual samples.c The purpose of this sample is to obtain background information.

d One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate-continuously may be used in place of, or in addition to, integrating dosimeters.

For the purpose of this table, a thermoluminescent dosimeter may be considered to be one phosphor and two or more phosphors in a packet considered as two or more dosimeters.

Film badges should not be used for measuring direct radiation.

e Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

Table 2.6-2-MAXIMUM VALUES OF THE LOWER LIMITS OF DETECTION (LLD (a))Analyses Gross Beta H-3 Mn-54 Fe-59 Co-58, 60 Zn-65 Zr-95 Nb-95 1-131 Cs-134 Cs-137 Ba-140 La-140 Water (pCi/1)4 (b)2000 (d)15 30 15 30 30 15 1 (e)15 Airborne Particulate or Gas (pCi/m 3)Fish (pCi/kg,wet)

Milk (pCi/l)Food Products (pCi/kg, wet)Sediment (pCi/kg, dry)1 x 1 0-2(c)130 260 130 260 7 x 1 0-2(f)5 x 1 0-2(g)6 x 10- 2(g)130 150 1 18 60 15 60 60 80 150 180 18 60 15 (a)(b)(c)(d)(e)(f)(g)See definition of LLD in table notation of Table 2.3-1.LLD for drinking water.Applicable to airborne particulate only.LLD for drinking water. If no drinking water pathway exists, a value of 3000 may be used.LLD for drinking water. If no drinking water pathway exists, a gamma isotopic LLD value of 15 pCi/l may be used.Applicable to airborne gas only.Applicable to airborne particulate only during gamma isotopic analysis.LICENSE NO. TITLE: PAGE: 99 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 TABLE 2.6-3 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Analyses Water (pCi/1)Airborne Particulate or Gases (pCi/m 3)Fish (pCi/kg,wet)

Milk (pCi/l)Food Products (pCi/kg,wet)

H-3 Mn-54 Fe-59 Co-58 Co-60 Zn-65 2 x 104 (a)1 x 103 4 x 102 1 x 103 3 x 102 3 x 102 4 x 102 (b)2 30 50 2 x 102 (b)3 x 10~4 1 x 104 3 x 104 1 x 104 2 x "104 Zr-Nb-95 1-131 0.9 3 Cs-134 Cs-137 10 20 1 x 103 2 x 103 60 70 1 x 102 1 x 103 2 x 103 Ba-La-140 3 x 102 (b)(a) For drinking water samples.(b) Total for parent and daughter.

RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING Radiological Environmental Monitoring Program Description SURVEILLANCE LIMITATIONS S2.6.1.A The results of analyses performed on the radiological environmental monitoring samples shall be summarized in the Annual Radiological Environmental Operating Report.APPLICABILITY:

Applies at all times.OBJECTIVE:

To provide information on the radiological effects of station operation on the environment.

BASES BL2.6.I.A

& BS2.6.I.A The radiological monitoring program required by this limitation 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 the station operation.

This monitoring program thereby supplements the radiological effluents 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 initially specified monitoring program will be effective for at least the first three years of commercial operation.

Following this period, program changes may be initiated based on operational experience.

The detection capabilities required by Table 2.6-2' are state of the ar~t for routine environmental measurements in industrial laboratories.

RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING 2.6.2 Land Use Census LIMITATION L2.6.2.A A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence, and the nearest garden* of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles from the ANO-I reactor building.APPLICABILITY:

OBJECTIVES:

Applies at all times This limitation will identify changes in use of the unrestricted areas.ACTION: 1. With a land use census identifying a location(s) which yields a calculated dose commitment due to 1-131, tritium, and radionuclides in particulate form greater than the values currently being calculated in Appendix 1, Limitation S2.5.1.A and Appendix 2, Limitation S2.4.3.A submit location description in the Radioactive Effluent Release Report per Appendix 1, Section 3.2 Limitations.

2. With a land use census identifying a location(s) which yields a calculated dose commitment (via the same exposure pathway) greater than at a location from which samples are currently being obtained in accordance with Appendix 1, Limitation L2.6.I.A, identify the new location in the Radioactive Effluent Release Report per Appendix 1, Section 3.2 Limitations.

The new location shall be added to the radiological environmental monitoring program within 30 days, if possible.

The sampling location having the lowest calculated dose commitment (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

Broad leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.LICENSE NO. TITLE: PAGE: 102 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING Land Use Census SURVEILLANCE LIMITATION s2.6.2.A The land use census shall be conducted at least once per 24 months between the dates of June 1 and October 1, by door-to-door survey, aerial survey, or by consulting local agricultural authorities.

The results of the land use census shall be included in the Annual Radiological Environmental 0perating'Report.

APPLICABILITY:

Applies at all times OBJECTIVES:

This limitation will identify changes in use of the unrestricted areas.BASES BL2.6.2.A

& BS2.6.2.A This limitation is provided to ensure that changes in the'use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathway 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. To determine this minimum garden size, the following assumptions were used, 1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/square meter.y RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING

2.6.3 Interlaboratory

Comparison Program LIMITATION L2.6.3.A Analyses shall be performed on radioactive materials supplied as part of Interlaboratory Comparison Program which has been approved by NRC.APPLICABILITY:

OBJECTIVE:

Applies to the off-site radiochemistry laboratory.

To provide independent checks on the accuracy of the measurements of radioactive material in environmental samples.ACTION: 1. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.SURVEILLANCE LIMITATION S2.6.3.A The results of analyses performed as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Appendix 1, Section 3.3 Limitations.

APPLICABILITY:

Applies to the off-site radiochemistry laboratory.

OBJECTIVE:

To provide independent checks on the accuracy of the measurements, of radioactive material in environmental samples., BASES BL2.6.3.A

& BS2.6.3.A 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 sample matrices are performed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

3.0 ADMINISTRATIVE

CONTROLS 3.1 PROGRAMS LIMITATION L3.1.1 The following program shall be established, implemented, and maintained:

A. Radioactive Effluent Controls Program (Refer to Unit 1 Technical Specification 5.5.4)B. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative , measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.

The program shall (1) be contained in this manual, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1. Monitoring, sampling, analysis', and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in this manual, 2. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census, 3. Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

ADMINISTRATIVE CONTROLS 3.2 RADIOACTIVE EFFLUENT RELEASE REPORT**LIMITATIONS L3.2.1 The Radioactive Effluent Release Report shall be in accordance with Unit 1 Technical Specification 5.6.3 requirements.

A. The Radioactive Effluent Release Report shall include a summary 'of the quantities of radioactive liquid and gaseous effluents and solid waste release from the unit. The data will be summarized following the format of Regulatory Guide 1.21, Rev. 1.B. The Radioactive Effluent Release Report shall include the following information for all unplanned releases to unrestricted areas of radioactive material in gaseous and liquid effluents:

1. A description of the event and equipment involved.2. Cause(s) for the unplanned release.3. Actions taken to prevent recurrence.

4. Consequences of the unplanned release.C. The Radioactive Effluent Release Report shall contain a description of any changes to the ODCM and PCP made during the period of the report.(Refer to Unit 1 Technical Specification 5.5.1.c for potential reporting requirements.)

D. The Radioactive Effluent Release Report shall contain: I 1. A summary of the hourly meteorological data collected over the previous calendar year. In lieu of including this summary in the report, the data may be retained by the Licensee for NRC review and noted as such in the report.2. A summary of radiation doses due to radiological effluents during the previous calendar year calculated in accordance with the methodology specified in this manual.3. The radiation dose to members of the public due to their activities inside the site boundary.

This calculated dose shall include only those dose contributions directly attributed to operation of the unit and shall be compared to the limits specified in 40 CFR 190.**A single submittal may be made for ANO. The submittal shall combine sections common to both units. The submittal shall specify the releases of radioactive material from each unit.

ADMINISTRATIVE CONTROLS 3.2 RADIOACTIVE EFFLUENT RELEASE REPORT**LIMITATIONS (Continued)

E. The Radioactive Effluent Release Report shall include a description of licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid) during the previous calendar year.******This information may be included in the periodic SAR update in lieu of inclusion in this report.LICENSE NO. TITLE: PAGE: 107 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 ADMINISTRATIVE CONTROLS 3.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*LIMITATIONS L3.3.1 The Annual Radiological Environmental Operating Report shall be in accordance with Unit 1 Technical Specification 5.6.2 requirements.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.A. The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.B. The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples and of all environmental radiation measurements required by this manual taken during the reporting period. 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. If the missing data becomes available, it, shall be submitted as soon as possible in a supplementary report.C. The Annual Radiological Environmental Operating 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 reactor buildings; the results of the Land Use Census required by Appendix 1, Section 2.6.2 limitations, and the results of the Interlaboratory Comparison Program participation required by Appendix 1, Section 2.6.3 limitations.

A single submittal may be made for ANO.common to both units.The submittal should combine sections ADMINISTRATIVE CONTROLS 3.4 SPECIAL REPORTS LIMITATIONS L3.4.1 Special reports shall be submitted to the Administrator of the appropriate Regional Office within the time period specified for each report.These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference limitation.

A. Radioactive Effluents; Appendix 1, Section 2.3, 2.4, 2.5 Limitations and NMM RW-105.This report shall include the following:

1. Description of occurrence.

2. Identify the cause(s) for exceeding the limit(s)3. Explain corrective action(s) taken to mitigate occurrence.

4. Define action(s) taken to prevent recurrence.

5. Summary of consequence(s) of occurrence.

6. Describe levels exceeding 40 CFR 190 in accordance with 10 CFR 20.405(c), as applicable.

B. Radiological Environmental Monitoring Sample Analysis;Appendix 1, Section 2.6.1 Limitations.

C. An unplanned offsite release during any one hour period of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents.

The report of an unplanned offsite release of radioactive material shall be submitted within 30 days of the occurrence and shall include the following information:

1. A description of the event and equipment involved.2. Cause(s) for the unplanned release.3. Actions taken to prevent recurrence

4. Consequences of the unplanned release.

LIMTAIO UNIT 1 MAIN STEAM LINE RADIATION MONITORING INSTRUMENTATION LIMITATION L4.1 The Main operable whenever APPLICABILITY:

OBJECTIVE:

Steam Line Radiation Monitoring Instrumentation shall be with a minimum measurement range from 10-1 to 104 mR/hr, the reactor is above the cold shutdown condition.

Unit 1 Main Steam Line Radiatidn Monitors To delineate the condition of the Unit.1 main steam line radiation monitors to assure reactor safety.I ACTION: 1. With the less than minimum number of channels operable, take action shown in Table 4.1-1, Note 1.SURVEILLANCE LIMITATION S4.1 The Main Steam Line Radiation Monitors shall be demonstrated operable by performance of the- channel check, channel test, and channel calibration at the frequency shown in Table 4.1-2.APPLICABILITY:

OBJECTIVE:

Applies to the Main Steam Line Radiation Monitors.To provide surveillance limitations for the instruments listed Appendix 1, Limitation L4.1.LICENSE NO. TITLE: PAGE: 110 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 Table 4.1-1 Instrumentation Limitinq Conditions for Operation No. of Channels for system trip No. of Channels Min.operable channels Min.degree of redundancy Operator action if conditions of column 3 or 4 cannot be met Functional Unit 1. Main Steam Line 1/steam line N/A 1/steam line 0 Note 1 Notes: 1. With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements,Initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and restore the inoperable Channel(s) to OPERABLE status within 7 days of the event.

Table 4.1-2 Instrument Surveillance Requirements Channel Description

1. Main Steam Line Radiation Monitors S -Each Shift M -Monthly R -Once every 18 months Check S Test M Calibrate R APPENDIX 2 RADIOACTIVE EFFLUENT CONTROLS UNIT 2

1.0 DEFINITIONS

OPERABLE -OPERABILITY 1.1 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 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).

CHANNEL CALIBRATION 1.2 *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 1.3 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 1.4 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.

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

C. Digital computer channels -The exercising of the digital computer hardware using diagnostic programs and the injection of simulated process data into the channel to verify operability.

SOURCE CHECK 1.5 A source check shall be the qualitative assessment of channel response when the channel sensor is exposed to the radioactive source.LICENSE NO. TITLE: PAGE: 114 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 DEFINITIONS LIQUID RADWASTE TREATMENT SYSTEM 1.6 A liquid radwaste treatment system is a system designed and installed to reduce radioactive liquid effluents from the unit. This is accomplished by providing for holdup, filtration, and/or demineralization of radioactive liquid effluents prior to their release to the environment.

GASEOUS RADWASTE TREATMENT SYSTEM 1.7 A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents from the plant by collecting offgases from radioactive systems and providing for decay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

VENTILATION EXHAUST TREATMENT SYSTEM 1.8 A ventilation exhaust treatment system is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment.

Such a system is not considered to have any effect on noble gas effluents.

Atmospheric cleanup systems that are Engineered Safety Feature (ESF) actuated are not considered to be ventilation exhaust treatment systems.MEMBER(S)

OF THE PUBLIC 1.9 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 recreational, occupational or other purposes not associated with the plant.PURGE -PURGING lAI0 Purge or purging is the controlled process of dis'charging air or gas from a confinement to reduce airborne radioactive concentrations in such a manner that replacement air or gas is required to purify the confinement.

EXCLUSION AREA 1.11 The exclusion area is that area surrounding ANO within a minimum radius of 0.65 miles of the reactor buildings and controlled to the extent necessary by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

DEFINITIONS UNRESTRICTED AREA 1.12 An unrestricted area shall be any'area at or beyond the exclusion area boundary.FREQUENCY NOTATION 1.13 The frequency notation specified for the performance of Surveillance Limitations shall correspond to the following intervals:

P S D W M Q SA R N/A Shift Daily Weekly Monthly Quarterly Semiannual Refueling Completed prior to each release At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 18 months Not Applicable BATCH RELEASE 1.14 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.CONTINUOUS RELEASE 1.15 A "Continuous" release is the discharge of liquid waste of a.non-discrete volume, e.g. from a volume of a system that has an input flow during the continuous release.

2.0 RELEASE

CONTROLS LIMITATION L2.0.1 The Limitation and Action requirements shall be applicable during the conditions specified for each limitation.

L2.0.2 Adherence to the requirements of the Limitation and/or associated Action within the specified time interval shall constitute compliance with the Limitation.

In the event the Limitation is restored prior to the expiration of the specified time interval, completion of the action statement is not required.BASES BL2.0.1 This limitation establishes the general requirements applicable to Limitations.

Limitations are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a Limitation is not met, the licensee shall follow any remedial Action permitted by the Limitation until the condition can be met. This limitation establishes the applicability statement within each individual limitation as the requirement for when (i.e., in which specified conditions) conformance to the Limitation is required for safe operation of the facility.

The Action requirements establish those remedial measures that must be taken within specified time limits when the requirements of a Limitation are not met. The Action requirements specify the remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the Action requirements.

In this case, conformance to the Action requirements provides an acceptable level of safety for unlimited continued operations as long as the Action requirements continue to be met. The specified time limits of the Action requirements are applicable from the point in time it is identified that a Limitation is not met. The time limits of the Action requirements are also applicable when a system or component is removed from service for surveillance testing or investigation of operational problems.

Individual limitations may include a specified time limit for completion of a Surveillance Limitation when equipment is removed from service.In this case, the allowable outage time limits of the Action requirements are applicable when this limit expires if the surveillance has not been completed.

BL2.0.2 This limitation establishes the general requirements applicable to Limitations.

Limitations are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limitation is not met, the licensee shall follow any remedial Action permitted by the Limitation until the condition can be met. This limitation establishes that noncompliance with a limitation exists when the requirements of the Limitation are not met and the associated Action requirements have not been implemented within the specified time interval.

The purpose of this limitation is to clarify that (1) implementation~of the Action requirements constitutes compliance with a limitation and (2) completion of the remedial measures of the Action requirements is not required when compliance with a Limitation is restored within the time interval specified in the associated Action requirements.

SURVEILLANCE LIMITATION S2.0.1 Surveillance Limitations shall be applicable during the conditions specified for individual Limitations unless otherwise stated in an individual Surveillance Limitation.

S2.0.2 Each Surveillance Limitation shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25% of the specified surveillance interval.BASES BS2.0.1 This surveillance limitation establishes the general requirements applicable to Surveillance Limitations.

Surveillance Limitations are requirements relating to test,) calibration, or inspection to ensure that the necessary quality of systems and components is maintained and that-the limitations will be met. This surveillance limitation establishes'the requirement that surveillances must be performed during the conditions for which the requirements of the Limitation apply unless otherwise stated in an individual Surveillance Limitation.

The purpose of this surveillance limitation is to ensure that the surveillances are performed to verify the operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is an a specified condition for which the associated Limitations are applicable.

Surveillance Limitations do not have to be performed when the facility is in a condition for which the requirements of the associated Limitation do not apply unless otherwise specified.

BS2.0.2 This surveillance limitation establishes the general requirements applicable to Surveillance Limitations.

Surveillance Limitations are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained and that the limitations will be met. This surveillance limitation establishes the limit for which the specified time interval for Surveillance-Limitations may be extended.It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance (e.g., transient conditions or other ongoing surveillance or maintenance activities).

It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at a refueling frequency and are specified with an 18-month surveillance interval.

It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances.

This surveillance limitation is based upon engineering judgement and the recognition that the most probable result of any.particular surveillance being performed is the verification of conformance with the Surveillance'Limitations.

This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance intervals.

RELEASE CONTROLS 2.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITATION L2.1.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 2.1-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of Appendix 2, Limitation L2.3.I.A are not exceeded.APPLICABILITY:

During releases via this pathway.ACTION: 1. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above limitation, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, until the set point is changed to an acceptable conservative value.2. With less than the minimum number of monitoring instrumentation channels operable, take the action shown in Table 2.1-1.3. Return the instruments to operable status within 30 days or, in.lieu of any other report, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected.

9 TABLE 2.1-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE INSTRUMENT

1. Gross Radioactivity Monitor(s)(provides alarm and automatic termination of release)a. Liquid Radwaste Effluent Line APPLICABILITY ACTION 1 During Releases Via This Pathway (DRVTP)1 2. Flow Monitor(s)

a. Liquid Radwaste Effluent Line DRVTP 2 TABLE 2.1-1 (Continued)

TABLE NOTATION ACTION 1 With 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: a. At least two independent samples are analyzed;and b. At least two technically qualified members of the Facility Staff independently verify the release rate computer input data; and c. At least two technically qualified members of the Facility Stiff independently verify the discharge valve lineup.Otherwise, suspend release of radioactive effluents via this pathway.ACTION 2 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 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.

RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATIONS S2.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated operable by performance of the channel check, source check,channel calibration, and channel functional test at the frequencies shown in Table 2.1-2.

TABLE 2.1-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATIONS CHANNEL INSTRUMENT CHECK 1. Gross Radioactivity Monitor(s)(provides alarm and automatic isolation)

a. Liquid Radwaste Effluents Line D*SOURCE CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST 2. Flow Monitor(s)

R R Q a. Liquid Radwaste Effluent Line N/A N/A* During releases via this pathway** A source check is not required if the background activity is greater than the activity of the check source.P Prior to release D Daily Q Quarterly R Every 18 months N/A Not Applicable Unit 2 RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION BASES BL2.1.1 & BS2.1.1 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 of liquid effluents.

The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in this manual to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.For the radioactive liquid effluent instrumentation surveillance requirements, the channel test demonstrates that automatic isolation of this pathway and control room alarm annunciation occur if the instrument indicates measured levels above the trip setpoint.

The channel test demonstrates that alarm annunciation occurs if any of the following conditions exist: A. Power to the detector is lost.B. The instrument indicates a downscale failure (local alarm only on 2RE-4423) .The initial channel calibration is performed using one or more of the 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. These standards permit calibrating the system over its intended range of energy and measurement range. For subsequent channel calibration, sources -that have been related to the initial calibration are used.

I Unit 2 RELEASE CONTROLS 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITATION L2.2.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 2.2-1 shall be operable with their alarm/trip setpoints set to ensure that the limits of Appendix 2, Limitation L2.4.I.A are not exceeded.APPLICABILITY:

During releases via this pathway.ACTION:.1. With the following gaseous effluent monitoring instrumentation channels alarm/trip setpoint less'conservative than required by the above limitation, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel.a. Waste Gas Holdup System Noble Gas Activity Monitor (during periods of gaseous releases.)

b. Containment Purge and Ventilation System Noble Gas Activity Monitor (during periods of containment building purge.)2. With less than the minimum number of monitoring instrumentation channels operable, take the action shown in Table 2.2-1.3. Return the instruments to operable status within 30 days or, in lieu of any other report, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected.

Unit 2 1 TABLE 2.2-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT

1. Waste Gas Holdup System a. Noble Gas Activity Monitor (provides alarm and automatic termination of release)b. Effluent System Flow Monitor-2. Containment Purge and Ventilation System a. Noble Gas Activity Monitor b. IodineSampler Cartridge c. Particulate Sampler Filter d. Effluent System Flow Monitor e. Sampler Flow Monitor OPERABLE

APPLICABILITY PARAM.ETER 1 1 During Releases Via This Pathway (DRVTP)DRVTP 1 1 1 DRVTP DRVTP DRVTP DRVTP DRVTP Radioactivity System Flow Radioactivity Verify Presence of Cartridge Verify Presence of Filter System Flow Sampler Flow ACTION 2 3, 5 4 4 2 2 1 Unit 2 RADIOAC INSTRUMENT

3. Spent Fuel Area Ventilation System a. Noble Gas Activity Monitor b. Iodine Sampler Cartridge c. Particulate Sampler Filter d. Effluent System Flow Monitor e. Sampler Flow Monitor 4. Auxiliary Building Area Ventilation System a. Noble Gas Activity Monitor b. Iodine Sampler Cartridge c. Particulate Sampler Filter TABLE 2.2-1 (Continued)

TIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE APPLICABILIITY PARA METER 1 1 DRVTP DRVTP DRVTP DRVTP DRVTP Radioactivity Verify Presence of Cartridge Verify Presence of Filter System Flow Sampler Flow Radioactivity Verify Presence of Cartridge Verify Presence of Filter System Flow Sampler Flow ACTION 3 4 4 2 2 3 4 4 2 2 1 1 1 1 DRVTP DRVTP DRVTP DRVTP DRVTP d.e.Effluent System Flow Monitor Sampler Flow Monitor Unit 2 1 RADI INSTRUMENT

5. Auxiliary Building Extension Ventilation System a. Noble Gas Activity Monitor b. Iodine Sample Cartridge c. Particulate Sampler Filter d. Effluent System Flow Monitor e. Sampler Flow Monitor 6.. Radwaste Storage Building HVAC Exhaust System a. Noble Gas Activity Monitor b. Iodine Sample Cartridge c. Particulate Sampler Filter d., Effluent System Flow Monitor e. Sampler Flow Monitor TABLE 2.2-i (Continued)

OACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE APPLICABILITY PARAME DRVTP DRVTP DRVTP DRVTP DRVTP DRVTP DRVTP DRVTP DRVTP DRVTP Radioactivity Verify Presence of Cartridge Verify Presence of Filter System Flow Sampler Flow Radioactivity Verify Presence of Cartridge Verify Presence of Filter System Flow Sampler Flow ACTION 3 4 4 2-2 6 7 7 8 8 I Unit 2 1 RADI INSTRUMENT

7. Emergency Penetration Room Ventilation System a. Noble Gas Activity Monitor b. Iodine Sample Cartridge c. Particulate Sampler Filter d. Effluent System Flow Monitor e. Sampler Flow Monitor TABLE 2.2-1 (Continued)

OACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS OPERABLE APPLICABILITY ,TER 1 1 1 1 DRVTP DRVTP DRVTP DRVTP DRVTP Radioactivity Verify Presence of Cartridge Verify Presence of Filter System Flow Sampler Flow ACTION 3 4 4 2 2 Unit 2 TABLE 2.2-1 (Continued)

TABLE NOTATION ACTION 1 With 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: a. At least two independent samples of the tank's contents are analyzed; and b. At least two technically qualified members of the Facility Staff independently verify the computer input data; and c. At least two technically qualified members of the Facility Staff independently verify the discharge valve lineup.Otherwise, suspend release of radioactive effluents via this pathway.ACTION 2 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 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .ACTION 3 With the number of channels operable less than required by the Minimum Channels operable requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 . During the performance of required source checks of radioactive gaseous effluent monitoring instrumentation for plant ventilation systems, these requirements need not be implemented if the instrument is restored to an operable condition within four hours.ACTION 4 With the number of channels operable less than required by the Minimum Channels operable requirement, effluent releases via this pathway may continue provided samples are collected with auxiliary sampling equipment.

Iodine sample cartridges and particulate sample filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing in accordance with Table 2.4-1. A time period of up to four hours is allowed to continue releases via the affected pathway without auxiliary sampling equipment in operation.

If at the conclusion of the four-hour period normal instrumentation is not restored or auxiliary sampling equipment is not in operation, then releases via the pathway shall be immediately suspended.

The allowable four-hour time period is applicable if the instrumentation is.discovered to be inoperable for'any reason and whenever it is necessary to render the instrumentation inoperable in order to perform the weekly surveillance (i.e., filter change6ut) required by ODCM Appendix 2 Table 2.4-1.

I Unit 2 1 TABLE 2.2-1 (Continued)

TABLE NOTATION ACTION 5 With the number of channels operable less than required by the Minimum Channels operable requirement, suspend all operations involving movement of fuel assemblies or CEAs within the pressure vessel.ACTION 6 With the number of channels operable less than required by the Minimum Channels operable requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-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 .Otherwise., suspend all compaction activities within the Radwaste Storage Building.ACTION 7 With the number of channels operable less than required by the Minimum Channels operable requirement, effluent releases via this pathway may continue provided samples are collected with auxiliary sampling equipment.

Iodine sample cartridges and particulate sample filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing in accordance with Table 2.4-1. Otherwise, suspend all compaction activities within the Radwaste Storage Building.ACTION 8 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 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Otherwise, suspend all compaction activities within the Radwaste Storage Building.

Unit 2 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATION S2.2.1 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated operable by performance

of the channel check, source check, channel calibration, and channel functional test at the frequencies shown in Table 2.2-2.

.Unit 2 1 TABLE 2.2-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATIONS CHANNEL CHECK SOURCE CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST INSTRUMENT

1. Waste Gas Holdup System a. Gas Activity Monitor (provides alarm and automatic termination of release)b. System Effluent Flow Monitor 2. Containment Purge and Ventilation System a. Gas Activity Monitor b. Iodine Sampler Cartridge c. Particulate Sampler Filter d. System Effluent Flow Monitor e. Sampler Flow Monitor R R Q N/A N/A D*W* (2)W* (2)D*D*N**N/A N/A N/A N/A R N/A N/A R R M (1), P N/A N/A N/A N/A I Unit 2 TABLE 2.2-2 (Continued)

I RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATIONS INSTRUMENT

3. Spent Fuel Area Ventilation System a. Gas Activity Monitor b. Iodine Sampler Cartridge c. Particulate Sampler Filter d. System Effluent Flow Monitor e. Sampler Flow Monitor 4. Auxiliary Building Area Ventilation System a. Gas Activity Monitor b. Iodine Sampler Cartridge c. Particulate Sampler-Filter

d. System Effluent Flow Monitor e. Sampler Flow Monitor CHANNEL CHECK SOURCE CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST W*W* (2)W* (2)D*D*M**N/A N/A N/A N/A R N/A N/A R R Q N/A N/A N/A N/A m*W* (2)W* (2)D*D*N/A N/A N/A N/A R N/A N/A R R Q N/A N/A N/A N/A I Unit 2 TABLE 2.2-2 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATIONS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST 5. Auxiliary Building Extension Ventilation System a. Gas Activity Monitor D* M** R Q b. Iodine Sampler Cartridge W*(2) N/A N/A N/A c. Particulate Sampler Filter W*(2) N/A N/A N/A d. System Effluent Flow Monitor D* N/A R N/A e. Sampler Flow Monitor D* N/A R N/A 6. Radwaste Storage Building HVAC Exhaust System a. Gas Activity Monitor D* M** R Q b. Iodine Sampler Cartridge W*(2) N/A N/A N/A c. Particulate Sampler Filter W*(2) N/A N/A N/A d. System Effluent Flow Monitor D* N/A R N/A e. Sampler Flow Monitor D* N/A R N/A Unit 2 1 TABLE 2.2-2 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE LIMITATIONS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST 7. Emergency Penetration Room Ventilation System a. Gas Activity Monitor D* M** R Q b. Iodine Sampler Cartridge W*(2) N/A N/A N/A c. Particulate Sampler Filter W*(2) N/A N/A N/A d. System Effluent Flow Monitor D* N/A R N/A e. Sampler Flow Monitor D* N/A R N/A Unit 2*TABLE 2.2-2 (Continued)

TABLE NOTATION*During releases via this pathway.**A source check is not required if the background activity is greater than the activity of the check source.(1)- During Containment Building ventilation operations.

(2) Verify presence of cartridge or filter only.P Prior to release D Daily W Weekly ,M Monthly Q Quarterly R Every 18 months N/A Not Applicable LICENSE NO. TITLE: PAGE: 137 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 IUnit 2 1 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION BASES BL2.2.1 & BS2.2.1 The radioactive gaseous effluent instrumentation is provided, to monitor and control, as applicable, ,the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm/trip setpoints for these instruments shall be calculated in accordance with the procedures in this manual to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.For the radioactive gaseous effluent instrumentation surveillance requirements, the channel functional test demonstrates that control room alarm annunciation occurs if any of the following conditions exist: A. The instrument indicates measured levels above the alarm/trip setpoint.B. Power to the detector is lost.C. The instrument indicates a downscale failure.For the containment purge and the waste gas holdup system noble gas activity monitors, the channel functional test also demonstrates the automatic isolation of the release pathway occurs if the instrument indicates above the trip setpoint.The initial channel calibration is performed using one or more of the 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. These 'standards permit calibrating the system over its intended range of energy and measurement range. For subsequent channel calibration, sources that have been related to the initial calibration are used.

lUnit 2 RELEASE CONTROLS 2.3 RADIOACTIVE LIQUID EFFLUENTS 2.3.1 Concentration LIMITATION L2.3.l.A The concentration of radioactive material 'released from the site in liquid effluents to the discharge canal shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration released shall be limited to 2 x. 10-4 pCi/ml.APPLICABILITY:

At all times.ACTION: 1. With the concentration of radioactive material released exceeding the above limits, immediately initiate actions to restore concentrations to within the above limits. Provide notification to the Commission within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and in lieu of any other report, submit a Special Report pursuant to Appendix 2, Limitation L3.4.I.A within 30 days.SURVEILLANCE'LIMITATIONS

$2.3.1.A.I Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analyses program of Table 2.3-1.$2.3.1.A.2 The results of the radioactivity analyses shall be. used in accordance with the methods in this manual to assure that the concentrations at the point of release are maintained within the limits of Appendix 2, Limitation L2.3.I.A.

Unit 2 TABLE 2.3-1'RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSES PROGRAM Liquid Release Sampling Minimum Type of Lower-Limit Type Frequency Analyses Activity of Detection Frequency Analyses (LLD)(uCi/ml) (a)A. Batch Waste P P 7 isotopic (g) 5 x 10-7 (b)Release (d)Each Batch Each Batch __1-131 lxlO-6 Dissolved and P Entrained 1 x 10-5 M Gases One Batch/M (Gamma Emitters)P M H-3 ix 5 Each Batch Composite (c)Gross Alpha -7 1 x 10 P Q Sr-89, Sr-90 5 x 107-8 Each Batch Composite (c)Fe-55 1 x 10-6 B. Continuous y isotopic (g) 5 x 10- (b)Waste Releise (e)(g 5(b Continuous D (f) (f) '1-131 1i x 10- 6 Dissolved and M Entrained 1 x 10-5 M Gases Grab Sample (Gamma, Emitters)Continuous M H-3 1 x 105 (f) (h) Composite (c)Gross Alpha 1 x 70-Q Sr-89, Sr-90 -8 1 Continuous 5 x 10 (f) (h) Composite (c)Fe-55 1 x 10-6 Unit 2 Table 2.3-1 (Continued)

TABLE NOTATION 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 Sb LLD =E

V* 2.22

Y

exp (-%At))Where: LLD is the lower limit of detection as defined above (as picocurie 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 transformation)

V is the sample size (in units of mass or volume).2.22 is the number of transformations per minute per picocurie.

Y is the fractional radiochemical yield (when applicable).

X is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).Typical values of E, V, Y, and At shall be used in the calculation.

It should be recognized that the LLD is an a priori (before the fact) limit representing the capability of a measurement system-and not an a posteriori (after the fact) limit for a particular measurement.

Unit 2 Table 2.3-1 (Continued)

/TABLE NOTATION (Continued)

b. For certain mixtures of gamma emitters, it may not be possible to measure radionuclides in concentrations near their sensitivity limits when other nuclides are present in the sample in much greater concentrations.

Under these circumstances, it will be more appropriate to calculate the concentration of such radionuclides using observed ratios with those radionuclides which are measurable.

c. A composite sample is one in which the quantity of liquid sampled is 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 liquids released.d. A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling, each batch shall be isolated and mixed to assure representative sampling.e. A continuous release is the discharge of liquid waste of a non-discrete volume, e.g. from a volume of a system that has an input flow during the continuous release.f. If continuous samples cannot be obtained, then grab samples shall be'collected and analyzed at least 1/24 hrs when the specific activity of'the secondary coolant is 0.01 pCi/ml IDE (1-131) or at least 1/12 hrs when the specific activity of the secondary coolant is >0.01 PCi/ml IDE (1-131). Grab sample frequency may be increased due to plant conditions.

g. The principal gamma emitters for which the LLD limitation will apply are exclusively the following radionuclides:

Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. 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.

Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.h. To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream.'Prior to analysis, all samples taken for the composite~shall be thoroughly mixed in order for the composite sample to be reperesentative of the effluent release.P Prior to release D Daily M Monthly Q Quarterly Unit 2 RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENTS Concentration BASES BL2.3.l.A

& BS2.3.1.A This limitation is provided to ensure that the concentration of radioactive materials released in liquid waste effluents in unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limit provides additional assurance that the levels of radioactive materials in bodies of water in unrestricted areas will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a member of the public, and (2) the limits of 10 CFR Part 20.106(e) to the population'.

The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-133 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication

2.

Unit 2 RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENTS 2.3.2 Dose LIMITATION L2.3.2.A The dose commitment to a member of the public from radioactive materials in liquid effluents released from ANO-2 to the discharge canal shall be limited: 1. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and 2. During any calendar year to less than or equal to 3 mremto the total body and to less than or equal to 10 mrem to any organ.APPLICABILITY:

At all times.ACTION: 1. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of any other report submit a Special Report pursuant to Appendix 2, Limitation L3.4.I.A within 30 days.SURVEILLANCE LIMITATIONS

$2.3.2.A Cumulative dose contributions from liquid effluents shall be determined in accordance with this manual at least once per 31 days.BASES BL2.3.2.A

& BS2.3.2.A This limitation provides assurance that releases of liquid effluents will result in concentrations below the limits of 10 CFR 20.The limitation provides the required operating flexibility and at the same time assures that the release of radioactive material in liquid effluents will be kept"as low as is reasonably achievable." The equations specified in this manual 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 CFR Part 50, Appendix I", Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", April 1977.

Unit 2 RELEASE CONTROLS RADIOACTIVE LIQUID EFFLUENTS 2.3.3 Liquid Radwaste Treatment LIMITATION L2.3.3.A The liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from ANO-2 to the discharge canal, would exceed 0.18 mrem to the total body or 0.625 mrem to any organ in any calendar quarter.APPLICABILITY:

At all times.ACTION: 1. With radioactive liquid waste being discharged without treatment and in excess of the above limits, in lieu of any other report, submit a Special Report pursuant to Appendix 2, Limitation L3.4.I.A within 30 days.SURVEILLANCE LIMITATIONS

$2.3.3.A Doses due to liquid releases shall be projected at least once per 31 days in accordance with this manual.BASES BL2.3.3.A

& BS2.3.3.A The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

The values of 0.18 mrem and 0.625 mrem are approximately 25%.of the yearly design objectives on a quarterly basis. The yearly design objectives are given in 10 CFR 50, Appendix I, Section II.

Unit 2 RELEASE CONTROLS 2.4 RADIOACTIVE GASEOUS EFFLUENTS 2.4.1 Dose Rate LIMITATION L2.4.1A The dose rate due to radioactive materials released in gaseous effluents from the site to unrestricted areas (see Figure 4-2) shall be limited to the following:

1. For noble gases: Less than or equal to the 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin.2. For iodine-131, for tritium and for all radionuclides in particular form with half lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.During periods of containment purging the dose rate may be averaged over a one hour interval.APPLICABILITY:

At all times.ACTION: 1. With the dose rate(s) exceeding the above limits, without delay restore the release rate to comply with the above limit(s).SURVEILLANCE LIMITATIONS S2.4.I.A.I The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of this manual.S2.4.I.A.2 The dose rate due to iodine-131, tritium, and all radionuclides in particulate form with half-lifes greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methods and procedures of this manual by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 2.4-1.

F Unit 2 TABLE 2.4-1 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSES PROGRAM Gaseous Release Type Sampling Frequency A. Waste Gas Storage Tank B. Containment Purge C. Unit Vents (Auxiliary Bldg. Ext.)(Spent Fuel Pool Area Ventilation)(Containment Bldg.Ventilation)(Radwaste Area Ventilation)(Low-Level Radwaste Storage Building HVAC Exhaust Ventilation)(Emergency Penetration Room Ventilation)

P Each Tank Grab Sample P Each Purge Grab Sample M (c) (d)Grab Sample Minimum Analyses Frequency P Each Tank P Each Purge M W(f)Charcoal Sample W(f)I ParticulateI I Sample Type of Activity Analysis Principal Gamma Emitters(b) 1 x 10-4 (g)Principal Gamma Emitters(b).

.1 x 10-4 (g)H-3 1 x 10-6 Principal Gamma Emitters(b)

I 1 x 10-4 (g)H-3 1 x 10-6 Lower Limit of Detection (LLD)(uCi/ml) (a)Continuous (e)1-131 1 x 10-12 Continuous (e)Principal Gamma Emitters(b)

(1-131, Others)1 x 10-11 1 x 10-"1 1 x 10-n Continuous(e)

Continuous (e)I M I Particulatel I Sample'I Q I Composite I Particulatel Sample Noble Gas Monitor Gross alpha Sr-89, Sr-90 Noble Gases Gross Beta or Gamma 1 x 10-6 (Xe-133 equiv.)Continuous (e)

I Unit 2 1 TABLE 2.4-1 (Continued)

TABLE NOTATION a. The Lower Limit of Detection (LLD) is defined in Table Notation (a.) of Table 2.3-1 of Appendix 2, Limitation L2.3.l.A.b. The principal gamma emitters for which the LLD limitation will apply are exclusively the following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, .and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.

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..

Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLD's higher than' required, the reasons shall be documented in the Radioactive Effluent Release Report.c. Tritium grab samples shall be taken from the Containment Building ventilation exhaust at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.d. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel area, whenever spent fuel is in the spent fuel pool.e. 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 Appendix 2, Limitations L2.4.I.A, L2.4.2.A, and L2.4.3.A.f. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from the sampler).g. For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these circumstances, the LLD-may be increased inversely proportional to the magnitude of the gamma yield (i.e., 1 x 10-4/I, where I is the photon abundance expressed asa decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the MPC value specified in 10 CFR 20, Appendix B, Table II, Column I.P Prior to release W Weekly M Monthly Q Quarterly Unit 2 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS Dose Rate BASES BL2.4.l.A

& BS2.4.l.A This limitation is provided to ensure that the dose at any time in unrestricted areas from gaseous effluents from all units on the site will be within the limits of 10 CFR Part 20.105(b).

This limitation applies to the release of gaseous effluents from all reactors at the site.

Unit 2 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS 2.4.2 Dose -Noble Gases LIMITATION L2.4.2.A The dose due to noble gases released in gaseous effluents from ANO-2 to unrestricted areas (See Figure 4-2) shall be: 1. During any calendar quarter, less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and 2. During any calendar year, less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.

APPLICABILITY:

At all times.ACTION: 1. With the calculated dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of any other report, submit a Special Report pursuant to Appendix 2, Limitation L3.4.I.A within 30 days.SURVEILLANCE LIMITATIONS

$2.4.2.A Cumulative dose contributions for noble gases for the current calendar quarter and current calendar year shall be determined in accordance with this manual at least once per 31 days.

IUnit 2 1 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS Dose -Noble Gases BASES BL2.4.2.A

& BS2.4.2.A This limitation is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10 CFR Part 50.The limitation implements the guides'set forth in Section II.B of Appendix I.The action statements 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." The Surveillance Limitations implement the requirements 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 a member of the public through appropriate pathways is unlikely to be substantially underestimated.

The dose calculations established in this manual for calculating the doses due to the actual release rates of radioactive noble gases in gaseous 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 CFR Part 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.111,"Methods for Estimating AtmosphericTransport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The equations in this manual provided for determining the air doses at and beyond the site boundary are based upon the historical average atmospheric conditions.

Unit 2 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS 2.4.3 Dose -Iodine-131, Tritium, And Radionuclides In Particulate Form LIMITATION L2.4.3.A The dose to a member of the public from iodine-131, from tritium, and from all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from ANO-2 to unrestricted areas (see Figure 4-2) shall be: 1. During any calendar quarter, less than or equal to 7.5 mrems to any organ, and 2. During any calendar year, less than or equal to 15 mrems to any organ.APPLICABILITY:

At all times.ACTION: 1. With the calculated dose from the release of iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report, submit a Special Report pursuant to Appendix 2, Limitation L3.4.I.A within 30 days.SURVEILLANCE LIMITATIONS

$2.4.3.A Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with this manual at least once per 31 days.

Unit 2 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS Dose -Iodine-131, Tritium, And Radionuclides In Particulate Form BASES BL2.4.3.A

& BS2.4.3.A This limitation is provided to implement the requirements of Sections II.C, III.A, and IV.A of Appendix I, 10 CFR Part 50.The limiting conditions for operation are the guides set forth in Section II.C of Appendix I. The action statements 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".

The calculational methods in this manual specified in the Surveillance Limitations implement the requirements 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 a member of the public through appropriate pathways is unlikely to be substantially underestimated.

The calculational methods in this manual 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,"Calculation of Annual Doses to Man from Routine 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 These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.

The release rate limitations for iodine-ll, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent on the existing radionuclide pathways to man in the areas at or beyond the site boundary.The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

Unit 2 RELEASE CoNTROLS RADIOACTIVE GASEOUS EFFLUENTS 2.4.4 Gaseous Radwaste Treatment LIMITATION L2.4.4.A The ventilation exhaust treatment systems shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent doses from ANO-2 to unrestricted areas (see Figure 4-2) would exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter; or when the projected doses due to iodine-131, tritium, and radionuclides in particulate form with half-lives greater than 8 days would exceed 1.0 mrem to any organ over a calendar quarter.L2.4.4.B When degasifying the reactor coolant system, the gaseous radwaste treatment system shall be used to reduce radioactive material in gaseous waste prior to their discharge when the projected gaseous effluent doses for ANO-2 to unrestricted areas (see Figure 4-2) would exceed 0.625 mrad for gamma radiation and 1.25 mrad for beta radiation in any calendar quarter.APPLICABILITY:

At all times.ACTION: 1. With gaseous waste being discharged without treatment and in excess of the L2.4.4.A limits, in lieu of any other report, submit a Special Report pursuant to Appendix 2, Limitation L3.4.I.A within 30 days.2.' With gaseous waste being discharged without treatment and in excess of the L2.4.4.B limits, in lieu of any other report, submit a Special Report pursuant to Appendix 2, Limitation L3.4.I.A within 30 days.SURVEILLANCE LIMITATIONS S2.4.4.A/B Doses due to gaseous releases from the site shall be projected at least once per 31 days in accordance with this manual.

I Unit 2 1 RELEASE CONTROLS RADIOACTIVE GASEOUS EFFLUENTS Gaseous Radwaste Treatment BASES BL2.4.4.A/B

& BS2.4.4.A/B The requirement 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 limitation implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the 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 CFR Part 50, for gaseous effluents.

This limitation applies to gaseous radwaste from ANO-2.LICENSE NO. TITLE: PAGE: 155 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 Unit 2 RELEASE CONTROLS 2.5 RADIOACTIVE EFFLUENTS 2.5.1 Total Dose LIMITATION L2.5.l.A The calculated doses from the release of radioactive materials in liquid or gaseous effluents shall not exceed twice the limits of Appendix 2, Limitations L2.3.2.A.I, L2.3.2.A.2, L2.4.2.A.I, L2.4.A.2, L2.4.3.A.I, or L2.4.3.A.2.

APPLICABILITY

At all times.ACTION: 1. With the calculated doses exceeding the above limits, prepare and, submit a Special Report pursuant to 10 CFR Part 20.405c.2. If the limits of 40 CFR 190 have been exceeded, obtain a variance from the Commission to permit further releases in excess of 40 CFR 190 limits. A variance is granted until staff action on the request is complete.SURVEILLANCE LIMITATIONS

$2.5.1.A Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Appendix 2, Limitations

$2.3.2.A, S2.4.2.A, and$2.4.3.A, and in accordance with this manual.

Unit 2 RELEASE CONTROLS RADIOACTIVE EFFLUENTS Total Dose BASES BL2.5.I.A

& BS2.5.I.A This limitation is provided to meet the dose limits of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525. The limitation requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action that should result in limiting the annual dose to a member 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 member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered.

If the dose to any member of the public is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provision 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 requirements for dose limits in 10 CFR Part 20, as addressed in Appendix 2, Section 2.3 and 2.4 Limitations.

An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

FUnit 2 RELEASE CONTROLS/2.6 RADIOLOGICAL ENVIRONMENTAL MONITORING

2.6.1 Monitoring

Program LIMITATION L2.6.I.A The radiological environmental monitoring samples shall be collected pursuant to Table 2.6-1 and shall be analyzed pursuant to the requirements of Table 2.6-1 and 2.6-2. The sample locations shall be shown in Table 4-1 in this manual.APPLICABILITY:

At all times.ACTION: 1. With the radiological environmental monitoring program not being conducted as specified in Table 2.6-1, prepare and submit to the Commission in the Annual Radiological Environmental Operating Report a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

2. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at one or more of the locations specified in Table 2.6-1 exceeding the limits of Table 2.6-3 when averaged over any calendar quarter, prepare and submit to the Commission, within 30 days from the end of the affected quarter, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table 2.6-3 to be exceeded, and defines the actions taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of Appendix 2, Limitations L2.3.2.A, L2.4.2.A, and L2.4.3.A.

When more than one of the radionuclides in Table 2.6-3 are detected in the sampling medium, this Special Report shall be submitted if: Concentration (1) + Concentration (2) + ...... > 1.0 reporting level (1) reporting level (2)When radionuclides other than those in Table 2.6-3 are detected and are the result of plant effluents, this Special Report shall be submitted if the potential annual dose to a member of the public is equal to or greater than the calendar year limits of Appendix 2, Limitations L2.3.2.A, L2.4.2.A, and L2.4.3.A.

This 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.

Unit 2 RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING Monitoring Program LIMITATION (Continued)

3. With milk or fresh leafy vegetable samples unavailable from any of the sample locations required by Table 2.6-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days.The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the causes of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Radioactive Effluent Release Report and also include in the report a revised Table 4-l reflecting the new location(s).

I Unit 2 1 TABLE 2.6-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway and/or Sample 1. AIRBORNE a. Radioiodine and Particulates Number of Sample Location(a) 4 Locations 2 Samples close to Site Boundary, in (or near) different sectors with the highest calculated annual average groundlevel D/Q.1 Sample from the vicinity of a community having the highest calculated annual average groundlevel D/Q.1 Sample from a control location 15-30 km.(10-20 miles)-distance(c)*

Sample and Collection Frequency(a)

Continuous operation of sampler with sample collection as required by dust loading but at least once per 14 days.Type and Frequency of Analyses Radioiodine canister.Analyze at least once per 14 days for 1-131.Particulate sampler: Analyze for gross beta radioactivity following filter change(b)

I Unit 2 1 TABLE 2.6-1 (Continued)

Exposure Pathway and/or Sample 2. DIRECT RADIATION(d)

Number of Sample Location(a) 24 Locations-16 inner ring stations with two or more dosimeters in each meteororological sector in the general area of the Site Boundary-8 stations with two or more dosimeters in special interest areas such as population centers, nearby residences, schools, and in 1 -2 areas to serve as control locations.

Sample and Collection Frequency(a)

Once per 92 days Type and Frequency of Analyses Gamma dose. Once per 92 days.

I Unit 2 1 TABLE 2.6-1 (Continued)

Exposure Pathway and/or Sample Number of Sample Location(a)

Sample and Collection Frequency.(a)

Type and Frequency of Analyses 3. WATERBORNE

a. Surface Water b. Drinking Water c. Sediment from Shoreline 1 Indicator location (influenced by plant discharge) 1 Control location (uninfluenced by plant discharge) 1 Indicator location (influenced by plant discharge) 1 Control location (uninfluenced by plant discharge) 1 Indicator location (influenced by plant discharge) 1 Control location (uninfluenced by plant discharge)

Once per 92 days.Once per 92 days.Once per 365 days.Gamma isotopic(e) and Tritium analyses once 92 days.1-131, gross beta, gamma isotopic(e) and tritium analyses once per 92 days.Gamma isotopic(e) analysis once per 365 days.

Unit 2 1 TABLE 2.6-1 (Continued)

Exposure Pathway and/or Sample 4. INGESTION a. Milk b. Fish c. Food Products Number of Sample Location(a) 1 Indicator sample location within 8 km.distant if commercially available.

1 Control sample location at a distant of >8 km. when an indicator exists.1 Sample of commercially and/or recreationally important species in vicinity of plant discharge.

1 Sample of same species in area not influenced by plant discharge,.

1 Sample of broadleaf (edible or non-edible) near the Site Boundary from one of the highest anticipated annual average groundlevel D/Q sectors, if milk sampling is not performed.

1 Sample location of broadleaf vegetation (edible or non-edible) from a control location 15 -30 km. distant, if milk sampling is not performed.

Sample and Collection Frequency(a)

Once per 92 days.Once per 365 days.Three per 365 days.Type and Frequency of Analyses Gamma isotopic(e) and 1-131 analyses once per 92 days.Gamma isotopic(e) on edible portions once per 365 days.Gamma isotopic(e) and 1-131 analyses three times per 365 days.

I Unit 2 1* TABLE 2.6-1 (Continued)

TABLE NOTATION a The ODCM shall include, in a table and figures, specific parameters of distance and direction from the centerline of one reactor, and additional description where pertinent, for each sample location in Table 2.6-1. Refer to NUREG-0133,' "Preparation of Radiological Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunctions, every effort shall be made to complete corrective action before the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report.It is recognized that, at times, it may not be possible or practical to continue to obtain samples of the media of choice at the most desirable 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 the next Annual Radioactive Effluent Release Report, identify the cause of the unavailability of samples for that pathway and identify the new location(s) for obtaining replacement samples, and also include in the report a revised figure(s) and table(s) for the ODCM reflecting the new location(s).

b Particulate sample filters should be analyzed for gross beta 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 or water is greater than ten times the yearly mean of control samples for any medium, gamma isotopic analysis should be performed on the individual samples.c 'The purpose of this sample is to obtain background information.

d One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters.

For the purpose of this table, a thermoluminescent dosimeter may be considered to be one phosphor and two or more phosphors in a packet considered as two or more dosimeters.

Film badges should not be used for measuring direct radiation.

e Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

I Unit 2 TABLE 2.6-2 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD(a))Analyses Water (pCi/i)Airborne Particulate or Gas (pCi/mr 3)Fish (pCi/kg, wet)Milk (pCi/l)Food Products (pCi/kg, wet)Sediment (pCi/kg, dry)Gross Beta H-3 Mn-54 Fe-59 Co-58, 60 Zn-65 Zr-95 Nb-95 1-131 Cs-134 Cs-137 Ba-140 La-140 4 (b)2000 (d)15 30 15 30 30 15 1 (e)15 18 60 15 1 x 1 0-2(C)130 260 130 260 7 x 1 0-2(f)5 x 1 0-2(g)6 x 10- 2(g)130 150 1 15 18 60 15 60 60 80 150 180 (a)(b)(c)(d)(e)(f)(g)See definition of LLD in table notation of Table 2.3-1.LLD for drinking water.Applicable to airborne particulate only.LLD for drinking water. If no drinking water pathway exists, a value of 3000 may be used.LLD for drinking water. If no drinking water pathway exists, gamma isotopic LLD value of 15 pCi/l may be used.Applicable to airborne gas only.Applicable to airborne particulate only during gamma isotopic analysis.

Unit 2 TABLE 2.6-3 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Airborne Particulate ter or gases- Fish Milk Food Produ./1) (pCi/m 3) (pCi/kg, wet) (pCi/l) (pCi/kg, w Wa (pCi cts et)Analyses H-3 Mn-54 Fe-59 Co-58 Co-60 Zn-65 Zr-Nb-95 1-131 Cs-134 Cs-137 Ba-La-140 2 x 104 (a)1 x 103 4 x 102 1 x 103 3 x 103 3 x 102 4 x 102 (b)2 30 5o 2 x 102 (b)3 1 3 1 2 x x x x x 10 4 10 4 10 4 10 4 10 4 0.9 10 20 1 x 103 2 x 103 3 60 70 3 x 102 (b)1 x 102 1 x 103 2 x 103 (a) For drinking water samples.(b) Total for parent and daughter.

1 Unit 21 RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING Monitoring Program SURVEILLANCE LIMITATIONS

$2.6.1.A The results of analyses performed on the radiological environmental monitoring samples shall be summarized in the Annual Radiological Environmental Operating Report.BASES BL2.6.I.A

& BS2.6.I.A The radiological monitoring program required by this limitation 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 the station operation.

This monitoring program thereby supplements the radiological effluents 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 initially specified monitoring program will be effective for at least the first three years of commercial operation.

Following this period, program changes may be initiated based on operational experience.

The detection capabilities required by Table 2.6-2 are state of the art for routine environmental measurements in industrial laboratories.

Unit 2 RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING 2.6.2 Land Use Census LIMITATION L2.6.2.A A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence, and the nearest garden* of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of five miles.APPLICABILITY:

At all times.ACTION: 1. With a land use census identifying a location(s) which yields a calculated dose commitment due to 1-131, tritium, and radionuclides in particulate form greater than the values currently being calculated in Appendix 2, Limitation S2.4.3.A, submit location description in the Radioactive Effluent Release Report per Appendix 2, Section 3.2 Limitations.

2. With a land use census identifying a location(s) which yields a calculated dose commitment (via the sample exposure pathway)greater than at a location from which samples are currently being obtained in accordance with the Appendix 2, Limitation L2.6.I.A, identify the new location in the Radioactive Effluent Release Report per Appendix 2, Section 3.2 Limitations.

The new location shall be added to the radiological environmental monitoring program within 30 days, if possible.

The sampling location having the lowest calculated dose commitment (via the same exposure pathway) may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

SURVEILLANCE LIMITATIONS S2.6.2.A The land use census shall be conducted at least once per 24 months between the dates of June 1 and October 1 by door-to-door survey, aerial survey, or by consulting local agricultural authorities.

The results of the land use census shall be included in the Annual Radiological Environmental Operating Report.*Broad Leaf vegetation sampling may be performed at the site boundary in the direction sector with the highest D/Q in lieu of the garden census.

Unit 2 RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING Land Use Census BASES BL2.6.2.A

& BS2.6.2.A This limitation is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathway 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. To determine this minimum garden size, the following assumptions were used: 1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/square meter.

I Unit 2'. I RELEASE CONTROLS RADIOLOGICAL ENVIRONMENTAL MONITORING

2.6.3 Interlaboratory

Comparison Program LIMITATION L2.6.3.A Analyses shall be performed on radioac'tive materials supplied as part-of the Interlaboratory Comparison Program which has been approved by NRC.'APPLICABILITY:

At all times.ACTION: 1. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence.

to the Commission in the Annual Radiological Environmental Operating Report.SURVEILLANCE LIMITATIONS

52.6.3.A The results of analyses performed as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.BASES BL2.6.3.A

& BS2.6.3.A 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 sample matrices are performed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

Unit 2 3.0 ADMINISTRATIVE CONTROLS 3.1 PROGRAMS LIMITATION L3.1.1 The followingprogram shall be established, implemented, and maintained:

A. Radioactive Effluent Controls Program (Refer to Unit 2 Technical Specification 6.5.4)B. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1)representative , measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.

The program shall (1) be contained in this manual, (2)conform to the guidance of Appendix I to 10 CFR Part 50, and (3)include the following:

1. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in this manual, 2. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census, 3. Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

Unit 2 ADMINISTRATIVE CONTROLS 3.2 RADIOACTIVE EFFLUENT RELEASE REPORT*LIMITATION L3.2.1 The Radioactive Effluent Release Report shall be in accordance with Unit 2 Technical Specification 6.6.3.A. The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The data will be summarized following the format of Regulatory Guide 1.21, Revision 1.B. Any changes in the ODCM and PCP shall be included in the Radioactive Effluent Release Report for the period in which the change(s) was made effective. (Refer to Unit 2 Technical Specification 6.5.1.c for potential reporting requirements.)

C. The Radioactive Effluent Release Report shall include the following information for all unplanned releases to unrestricted areas of radioactive materials in gaseous and liquid effluents:

1. Description of the occurrence.

2. Identify the cause(s) for exceeding the limit(s).3. Explain corrective actions taken to mitigate occurrence.

4. Define action(s) taken to prevent recurrence.

5. Summary of consequence(s) of occurrence.

A single submittal may be made for ANO. The submittal shall combine sections common to both units. The submittal shall specify the releases of radioactive material from each unit.

I Unit 2 1 ADMINISTRATIVE CONTROLS RADIOACTIVE EFFLUENT RELEASE REPORT*LIMITATION (Continued)

D. The Radioactive Effluent Release Report shall contain: 1. A summary of the hourly meteorological data collected over the previous-calendar year. In lieu of including this summary in the report, the data may be retained by the licensee for NRC review and noted as such in the report.2. A summary of radiation doses during the previous calendar the methodology specified in due to radiological effluent year calculated in accordance with this manual..3. The radiation dose tomembers of the public due to their activities inside the site boundary.

This calculated dose shall include only those dose contributions directly attributed to operation of the unit and shall be compared to the limits specified in 40 CFR 190.E. The Radioactive Effluent Release Report shall contain a description of licensee initiated major changes to the radioactive waste systems (liquid, gaseous and -solid) during the previous calendar year.**This information may be included in the periodic SAR update in lieu of inclusion in this report.LICENSE NO. TITLE: PAGE: 173 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 Unit 2 ADMINISTRATIVE CONTROLS 3.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*LIMITATION L3.3.1 The Annual Radiological Environmental Operating Report shall be in accordance with Unit 2 Technical Specification 6.6.2. The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of year.A. The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental monitoring program for the reporting period. If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.B. The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of all radiological environmental samples and of all environmental radiation measurements required by this manual taken during the reporting period. 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. If the missing data becomes available, it shall be submitted as soon as possible in a supplementary report.C. The Annual Radiological Environmental Operating 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 reactor buildings; the results of the Land Use Cen'sus required by Appendix 2, Section 2.6.2 Limitations; and the results of the Interlaboratory Comparison Program participation required by Appendix 2, Section 2.6.3 Limitations.

A single submittal may be made for ANO.common to both units.The submittal should combine sections LICENSE NO. TITLE: PAGE: 174 of 175 075 OFFSITE DOSE CALCULATION MANUAL REV: 016 Unit 2 ADMINISTRATIVE CONTROLS 3.4 SPECIAL REPORTS LIMITATION L3.4.1 Special reports shall be submitted to the Administrator of the Regional Office within the time period specified.

for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference limitation.

A. Radioactive Effluents; Appendix 2, Limitations L2.3.I.A, L2.3.2.A, L2.3.3.A, L2.4.2.A, L2.4.3.A, L2.4.4.A, L2.4.4.B, and L2.5.I.A.This report shall include the following:

1. Description of occurrence.

2. Identify the cause(s) for exceeding the limit(s)3. Explain corrective action(s) taken to mitigate occurrence.

4. Define action(s) taken to prevent recurrence.

5. Summary of consequence(s) of occurrence.

6. Describe levels exceeding 40 CFR 190 in accordance with 10 CFR 20.405(c).

B. Radiological Environmental Monitoring Sample Analysis; Appendix 2, Limitation L2.6.I.A.C. Unplanned Offsite Release during one hour period of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents.

This report shall be submitted within 30 days of the occurrence of the event and shall include the following information:

1. Description of the 6ccurrence.

2. Identify the cause(s) of exceeding the limit(s).3. Explain'corrective action(s) taken to mitigate occurrence.

4. Define action(s) taken toprevent recurrence.

5. Summary of the consequence(s) of occurrence.

Attachment 3 to, OCAN020901 EN-RW-105, "PROCESS CONTROL PROGRAM" REVISION 1 Procedure Contains NMM REFLIB Forms: YES E] NO Z Effective Procedure Owner: Mark L. Carver Executive Sponsor: Kevin Walsh 'Date Title: Manager, Fleet Radwaste Title: VP -Operations 7/30/2008 Site: Echelon -HQN Site: Waterford 3 Exception Site Site Procedure Champion Title Date*ANO David Moore RPM N/A BRP N/A N/A GGNS Roy Wilson RPM IPEC Dennis Loope RPM JAF John Solowski RPM PLP Chuck Sherman RPM PNPS Jack Priest RPM RBS Brad Houston RPM VY Sam Wender IV RPM W3 Blake Pilutti RPM N/A NP N/A N/A N/A HQN Mark L. Carver Manager, Fleet Radwaste Site and NMM Procedures Canceled or Superseded By This Revision Palisades Nuclear Plant, Process Control Program (PCP) -Rev. 11 Process Applicability Exclusion All Sites: E]Specific Sites: ANO El BRP El GGNS El IPECLI JAF El PLP El PNPS El RBS E] VY E] W3 E] NP E]Change Statement* The revision to EN-RW-105 included a major re-write to provide updates to allow VTY, PLP and IPEC to be included as sites implementing this Process Control Program.* The following are updates: o Waste management practices section replaced dry and liquid waste management o Waste stream sampling methods and frequency section added o Waste classification section added o Quality Control section added o Dewatering section added, o Waste packaging section added o Miscellaneous section added for special tools/equipment&

training requirements

Exception Dates to permit appropriate change management training and to facilitate outage schedules.

EN-RW-105, Rev. 1 TABLE OF CONTENTS Section 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 Title Page P U R PO S E ..............................................................................

3 REFERENCES

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3 D E FIN IT IO N S ..........................................................................

5 RESPONSIBILITIES

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9 D E T A ILS ..............................................................................

..10 INTERFACES

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20 R EC O R D S .............................................................................

20 OBLIGATION AND REGULATORY COMMITMENT CROSS-REFERENCES

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21 ATTACHMENTS

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21 EN-RW-105, Rev. 1

1.0 PURPOSE

The Process Control Program (PCP) requires formulas, sampling, analyses, test and determinations to be made to ensure that the processing and packing of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61 and 71, State Regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste. The scope of a PCP is to assure that radioactive waste will be handled, shipped, and disposed of in a safe manner in accordance with approved site or vendor procedures, whichever is applicable.

[GGNS UFSAR, Chapter 16B.1 / TRM -7.6.3.8 paragraph 1]1.1 The purpose of this document is to provide a description of the solid radioactive waste Process Control Program (PCP) at all the Entergy fleet sites. The PCP describes the methods used for processing, classification and packaging low-level wet radioactive waste into a form acceptable for interim on-site storage, shipping and disposal, in accordance with 10 CFR Part 61 and current disposal site criteria.1.2 To ensure the safe operation of the solid radwaste system, the solid radwaste system will be used in accordance with this Process Control Program to process radioactive wastes to meet interim on-site storage, shipping and burial ground requirements.

1.3 This document addresses the process control program in the context of disposal criteria, on-site processing and vendor processing requirements.

1.4 The Process Control Program implements the requirements of 1 0CFR50.36a and General Design Criteria 60 of Appendix A to 1 OCFR Part 50. The process parameters included in the Process Control Program may include but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.1.5 This document does NOT address the requirements for 1 OCFR Part 61.56 (waste characteristics) for material sent to intermediate processors, because the final treatment and packaging is performed at the vendor facilities.

2.0 REFERENCES

[1] EN-QV-1 04, "Entergy Quality Assurance Program Manual Control"[2] Title 49, Code of Federal Regulations

[3] Title 10, Code of Federal Regulations, Part 20[4] Title 10, Code of Federal Regulations, Part 61[5] Title 10, Code of Federal Regulations, Part 71, Appendix H [QAPM, Section A.1I.c]EN-RW-105, Rev. 1 2.0 continued[6] Low-Level Waste Licensing Branch Technical Position on Radioactive Waste Classification, 11 May 1983[7] Disposal Site Criteria and License[8] Waste Processor Acceptance Criteria[9] EN-LI-100, "Process Applicability Determination"[10] NRC Information and Enforcement Bulletins* NRC Information Notice 79-19: Packaging of Low-Level Radioactive Waste for Transport and Burial.NRC Information Notice 80-24: Low-Level Radioactive Waste Burial Criteria.NRC Information Notice 80-32: Clarification of Certain Requirements for Exclusive-Use Shipments of Radioactive Materials.

NRC Information Notice 80-32, Rev. 1: Clarification of Certain Requirements for Exclusive-Use Shipments of Radioactive Materials.

NRC Information Notice 83-05: Obtaining Approval for Disposing of Very-Low-Level Radioactive Waste -1 OCFR Section 20.302.1P NRC Information Notice 83-10: Clarification of Several Aspects Relating to Use of NRC-Certified Transport Packages.* NRC Information Notice 83-33: Non-Representative Sampling of Contaminated Oil.* NRC Information Notice 84-50: Clarification of Scope of Quality Assurance Programs for Transport Packages Pursuant to 1 OCFR 50 Appendix B.* NRC Information Notice 84-72: Clarification of Conditions for Waste Shipments Subject to Hydrogen Gas Generation.

NRC Information Notice 85-92: Surveys of Wastes Before Disposal from Nuclear Reactor Facilities.

NRC Information Notice 86-20: Low-Level Radioactive Waste Scaling Factors, 1OCFR 61.0. NRC Information Notice 86-90: Requests to Dispose of Very Low-Level Radioactive Waste Pursuant 1 OCFR 20.302 o NRC Information Notice 87-03: Segregation of Hazardous and Low-Level Radioactive Wastes 0 NRC Information Notice 87-07: Quality Control of On-Site Dewatering/

Solidification Operations by Outside Contractors EN-RW-105, Rev. 1 2.0 continued[11] NRC Information and Enforcement Bulletins (continued)

NRC Information Notice 89-27: Limitations on the Use of Waste Forms and High Integrity Containers for the Disposal of Low-Level Radioactive Waste* NRC Information Notice 92-62: Emergency Response Information Requirements for Radioactive Material Shipments* NRC Information Notice 92-72: Employee Training and Shipper Registration Requirements for Transporting Radioactive Materials* NRC Generic Letter 89-01, "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program".[12] Nureg-0800 Standard Review Plan Section 11.4 Revision 2, Solid Waste Management Systems.[13] NRC Waste Form Technical Position, Revision 1 Jan 24 1991.[14] NRC SECY 94-198 Review of Existing Guidance Concerning the Extended Storage of Low-Level Radioactive Waste.[15] EPRI TR-1 06925 Rev-1, Interim On-Site Storage of Low Level Waste: Guidelines for Extended Storage -October1996

[16] NRC Branch Technical Position On Concentration Averaging And Encapsulation Jan 17 1995[17] Commitment Documents (U-2 and U-3)* IPN-99-079, "Supplement to Proposed Changes to Technical Specifications Incorporating Recommendations of Generic Letter 89-01 and the Revised 10 CFR Part 20 and 10 CFR Part 50.36a.* Appendix B Technical Specifications, Section 4.5 [IP, RECS ODCM Part 1]3.0 DEFINITIONS

[1] Batch -A quantity of waste to be processed having essentially consistent physical and chemical characteristics as determined through past experience or system operation knowledge by the Radwaste Shipping Specialist.

A batch could be a waste tank, several waste tanks grouped together or a designated time period such as between outages as with the DAW waste stream. An isolated quantity of feed waste to be processed having essentially constant physical and chemical characteristics.(The addition or removal of water will not be considered to create a new batch).EN-RW-105, Rev. 1 3.0 continued[2] Certificate of Compliance

-Document issued by the USNRC regulating use of a NRC licensed cask or issued by (SCDHEC) South Carolina Department of Health and Environmental Conservation regulating a High Integrity Container.

[3] Chelatinq Agents -EDTA, DTPA, hydroxy-carboxylic acids, citric acid, carbolic acid and glucinic acid.[4] Compaction

-The process of volume reducing solid waste by applying external pressure.[5] -Confirmatory Analysis -The practice of verifying that gross radioactivity measurements using MCA are reasonably consistent with independent laboratory sample data.[6] Dewatered Waste -Wet waste that has been processed by means other than solidification, encapsulation, or absorption to meet the free standing liquid requirements of 1 OCFR Part 61.56 (a)(3) and (b)(2).[7] De-waterinI

-The removal of water or liquid from a waste form, usually by gravity or pumping.[8] Dilution Factor -The RADMAN computer code factor to account for the non-radioactive binder added to the waste stream in the final product when waste is solidified.

[9] Dry Waste -Radioactive waste which exist primarily in a non-liquid form and includes such items as dry materials, metals, resins, filter media and sludges.[10] Encapsulation

-Encapsulation is a means of providing stability for certain types of waste by surrounding the waste by an appropriate encapsulation media.[11] Gamma-Spectral-Analysis

-Also known as IG, MCA, Ge/Li and gamma spectroscopy.'

[12] Gross Radioactivity Measurements

-More commonly known as dose to curie conversion for packaged waste characterization and classification.

[13] Homogeneous

-Of the same kind or nature; essentially alike. Most Volumetric waste streams are considered homogeneous for purposes of waste classification.

[14] Incineration

-The process of burning a combustible material to reduce its volume and yield an ash residue.[15] Liquid Waste -Radioactive waste that exist primarily in a liquid form and is contained in other than installed plant systems, to include such items as oil, EHC fluid, and other liquids. This waste is normally processed off-site.EN-RW-105, Rev. 1 3.0 continued[16] Low-Level Radioactive Waste (LLW) -Those wastes containing source, special nuclear, or by-product material that are acceptable for disposal in a land disposal facility.

For the purposes of this definition, low-level radioactive waste has the same meaning as in the Low-Level Waste Policy Act, that is, radioactive waste not classified as high-level radioactive waste, transuranic waste, spent nuclear fuel, or by-product material as defined in section 11 e.(2) of the Atomic Energy Act (uranium or thorium tailings and waste).[17] Measurement of Specific Radionuclides

-More commonly known as direct sample or container sample using MCA data for packaged waste characterization and classification.

[18] Operable -A system, subsystem, train, component or device SHALL be OPERABLE or have OPERABILITY when it is capable of performing its specified functions(s), and when all necessary attendant instrumentation, controls, electrical power, 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).

[19] Prequalification Program -The testing program implemented to demonstrate that the proposed method of wet waste processing will result in a waste form acceptable to the land disposal facility and the NRC.[20] Processing

-Changing, modifying, and/or packaging radioactive waste into a form that is acceptable to a disposal facility.[21] Quality Assurance/Quality Control -As used in this document, "quality assurance" comprises all those planned and systematic actions necessary-to provide adequate confidence that a structure, system, or component will perform satisfactorily in service.Quality assurance includes quality control, which comprises those quality assurance actions related to the physical characteristics of a material structure, component, or system to predetermined requirements.

[22] Reportable Quantity Radionuclides (RQ) -Any radionuclide listed in column (1) of Table 2 of 49CFR Part 172.101 which is present in quantities as listed in column (3) of Table 2 of 49CFR Part 172.101.[23] Sampling Plan -A program to ensure that representative samples from the feed waste and the final waste form are obtained and tested for conformance with parameters stated in the PCP and waste form acceptance criteria.[24] Scaling Factor -A dimensionless number which relates the concentration of an easy to measure radionuclide (gamma emitter) to one which is difficult to measure (beta and/or alpha emitters).

EN-RW-105, Rev. 1 3.0 continued[25] Significant Quantity -For purposes of waste classification all the following radionuclide values SHALL be considered significant and must be reported on the disposal manifest.* Any value (real or LLD) for radionuclides listed in Appendix G to 1 OCFR20 (H-3, C-14, 1-129, Tc-99).* Greater than or equal to 1 percent of the concentration limits as listed in 1 OCFR Part 61.55 Table 1.* Greater than or equal to 1 percent of the Class A concentration limits listed in 1 OCFR Part 61.55 Table 2..Greater than or equal to 1 percent of the total activity.Greater than or equal to 1 percent of the Reportable Quantity limits listed on 49CFR Part 172.101 Table 2.[26] Solidification

-The conversion of wet waste into a free-standing monolith by the addition of an agent so that the waste meets the stability and free-standing liquid requirements of the-disposal site.[27] Special Radionuclides

-The RADMAN computer code term for radionuclides listed in Appendix G to 10CFR20 (i.e., H-3, C-14,1-129

& Tc-99)[28] Stability

-Structural stability per 1 OCFR61.2, Waste Form Technical Position, and Waste Form Technical Position Revision 1. This can be provided by the waste form, or by placing the waste in a disposal container or structure that provides stability after disposal.

Stability requires that the waste form maintain its structural integrity under the expected disposal conditions.

[29] Training -A systematic program that ensures a person has knowledge of hazardous materials and hazardous materials regulations.

[30] Type A Package -Is the packaging together with its radioactive contents limited to Al or A2 as appropriate that meets the requirements of 49CFR Part 173.410 and Part 173.412, and is designed to retain the integrity of containment and shielding under normal conditions of transport as demonstrated by the tests set forth in 49CFR Part 173.465 or Part 173.466 as appropriate.

[31] Type B Package -Is the packaging together with its radioactive contents that is designed to retain the integrity of containment and shielding when subjected to the normal conditions of transport and hypothetical acdident test conditions set forth in 10CFR Part 71.EN-RW-105, Rev. 1 3.0 continued[32] Volume Reduction

-any process that reduces the volume of waste. This includes but is not limited to, compaction and incineration.

[33] Waste Container

-A vessel of any shape, size, and composition used to contain the'waste media.[34] Waste Form -Waste in a waste container acceptable for disposal at a licensed disposal facility.[35] Waste Stream -A Plant specific and constant source of waste with a distinct radionuclide content and distribution.

[36] Waste Type -A single packaging configuration and waste form tied to a specific waste stream.4.0 RESPONSIBILITIES

[1] The Vice President Operations Support (VPOS) is responsible for the implementation of this procedure.

[2] Each site Senior Nuclear Executive (SNE) is responsible for ensuring that necessary site staff implements this procedure.

[3] The Low Level RadWaste (LLRW) Focus Group is responsible for evaluating and recommending changes and revisions to this procedure.

[4] Each site RP Department

-Radwaste Supervisor

/ Specialist (title may vary at the site's respectively) has the overall responsibility for implementing the PCP and is responsible for processing and transportation is tasked with the day-to-day responsibilities for the following:

Implementing the requirements of this document.0 Ensuring that radioactive waste is characterized and classified in accordance with 1 OCFR Part 61.55 and Part 61.56.Ensuring that radioactive waste is characterized and classified in accordance with volume reduction facility and disposal site licenses and other requirements.

Designating other approved procedures (if required) to be implemented in the packaging of any specific batch of waste.Providing a designated regulatory point of contact between the Plant and the NRC, volume reduction facility or disposal site.EN-RW-105, Rev. 1

4.0 continued

Maintaining records of on-site and off-site waste stream sample analysis and Plant evaluations.

Suspending shipments of defectively processed or defectively packaged radioactive wastes from the site when the provisions of this process control program are not satisfied.

5.0 DETAILS

An isotopic analysis SHALL be performed on every batch for each waste stream so that the waste can be classified in accordance with 10CFR61. The isotopic and curie content of each shipping container SHALL be determined in accordance with 49CFR packaging requirements.

The total activity in the container may be determined by either isotopic analysis or by dose-rate-to-curie conversion.

5.1. Precautions

and Limitations

[1] Precautions (a) Radioactive materials SHALL be handled in accordance with applicable radiation protection procedures.(b) All radioactive waste must be processed or packaged to meet the minimum requirements listed in 1 OCFR Part 61.56 (a) (1) through (8).(c) If the provisions of the Process Control Program are not satisfied, suspend shipment of the defectively processed or defectively packaged waste from the site. Shipment may be accomplished when the waste is processed

/ packaged in accordance with the Process Control Program.(d) The generation of combustible gases is dependent on the waste form, radioactive concentration and accumulated dose in the waste. Changes to organic inputs (e.g. oil) to waste stream, may change biogas generation rates.[2] Limitations (a) Only qualified personnel will characterize OR package radioactive waste OR radioactive materials for transportation or disposal.(b) All site personnel that have any involvement with radioactive waste management computer software SHALL be familiar with its functions, operation and maintenance.

EN-RW-105, Rev. 1

5.2. Waste

Management Practices[1] Waste processing methods include the following: (a) Present and planned practice is NOT to solidify or encapsulate any waste streams.(b) Waste being shipped directly for burial in a HIC (High Integrity Container) is dewatered to less than 1 percent by volume prior to shipment.(c) Waste being shipped directly for burial in a container other than a HIC is dewatered to less than 0.5 percent by volume prior to shipment.(d) IF solidification is required in the future, THEN at least one representative test specimen from at least every 10th batch of each type of radioactive waste will be checked to verify solidification.

(1) IF any specimen fails to verify solidification, THEN the solidification of the batch under test SHALL be suspended until such time as additional test specimens can be obtained, alternative solidification parameters can be determined, and a subsequent test verifies solidification.

If alternative parameters are determined, the subsequent tests shall be verified using the alternative parameters determined.

(2) IF the initial test specimen from a batch of waste fails to verify solidification, THEN provide for the collection and testing of representative test specimens from each consecutive batch of the same type of waste until at least 3 consecutive initial test specimens demonstrates solidification.

The process SHALL be modified as required to assure solidification of subsequent batches of waste.[2] Operation and maintenance of dewatering systems and equipment include the following: (a) Present and planned practice is to utilize plant personnel supplemented by vendor personnel or contracted vendor personnel, to operate AND maintain dewatering systems and equipment (as needed to meet disposal site requirements).(b) All disposal liners are manufactured by.and purchased from QA-approved vendors.[3] ALARA considerations are addressed in all phases of the processes involving handling, packaging AND transfer of any type OR form of radioactive waste (dewatered or dry).Resin, charcoal media, spent filter cartridges AND sludges are typically processed within shields. Sluiceable demineralizers are shielded when in service. Radiation exposure and other health physics requirements are controlled by the issuance of a Radiation Work Permit (RWP) for each task.EN-RW-105, Rev. 1

5.3. Waste

Stream Samplinq Methods and Frequency[1] The following general requirements apply to Plant waste stream sampling: (a) Treat each waste stream separately for classification purposes.(b) Ensure samples are representative of or can be correlated to the final waste form.(c) Determine the density for each new waste stream initially or as needed (not applicable for DAW and filters).(d) Perform an in-house analysis for gamma-emitting radionuclides for each sample sent to an independent laboratory.(e) Periodically perform in-house analysis for gamma emitting radionuclides for comparison to the current data base values for gamma emitters. (The current database is usually based on the most recent independent laboratory results.)(f) Resolve any discrepancies between in-house results AND the independent laboratory results for the same or replicate sample as soon as possible.(g) Maintain records of on-site and off-site waste stream sample analysis and evaluations.

[2] When required, waste stream samples should be analyzed, re-evaluated and if necessary, shipped to a vendor laboratory for additional analysis.

The same is true when there is a reason to believe that an equipment or process change has significantly altered the previously determined scaling factors by a factor-of 10.Specific examples include but are not limited to: Changes in oxidation reduction methods such as zinc, injection, hydrogen water chemistry, Changes in purification methods including media specialization, media distribution, ion/cation ratios,-* Changes in fuel performance criteria including fuel leaks* Other changes in reactor coolant chemistry.

Sustained, unexplained, changes in the routinely monitored Beta/Alpha ratios, as determined by Radiation Protection,* When there is an extended reactor shutdown (> 90 days).* When there are changes to liquid waste processing, such as bypassing filters, utilizing filters or a change in ion exchange media. -When there are changes to the waste stream that could change the biogas generation rate.EN-RW-105, Rev. 1 5.3 continued[3] The following requirements apply to infrequent or abnormal waste types: (a) Infrequent OR abnormal waste types that may be generated must be evaluated on a case-by-case basis.(b) The RP Department Supervisor

/ Specialist responsible for processing AND shipping will determine if the waste can be correlated to an existing waste stream.(c) IF the radioactive waste cannot be correlated to an existing waste stream, THEN the RP Department Supervisor

/ Specialist responsible for processing and shipping SHALL determine specific off-site sampling and analysis requirements necessary to properly classify the material.[4] Specific sampling methods and data evaluation criteria are detailed in EN-RW-1 04 for specific waste streams.5.4. Waste Classification

[1] General requirements for scaling factors include the following: (a) The Plant has established an inferential measurement program whereby concentrations of radionuclides which cannot be readily measured are estimated through ratio-ing with radionuclides which can be readily measured.(b) Scaling factor relationships are developed on a waste stream-specific basis.These relationships are periodically revised to reflect current independent lab data from direct measurement of samples. The scaling factor relationships currently used by the sites are as follows:* Hard to detect ACTIVATION product radionuclides and C-14 are estimated by using scaling factors with measured Co-60 activities.

Hard to detect FISSION product radionuclides and H-3, Tc-99 and 1-129 are estimated by using scaling factors with measured Cs-1 37 activities.

Hard to detect TRANSURANIC radionuclides are estimated by using scaling factors with measured Ce-1 44 activities.

Where Ce-1 44 cannot be readily measured, transuranics are estimated by using scaling factors with measured Cs-137 activities.

Second order scaling of transuranics is acceptable when Cs-137 and Ce-144 are not readily measurable.

EN-RW-105, Rev. 1 5.4 continued[2] General requirements for the determination of total activity and radionuclide concentrations include the following: (a) The activity for the waste streams is estimated by using either Gross Radioactivity Measurement OR Direct Measurement of Radionuclides.

Current specific practices are as follows: DAW -Gross radioactivity measurement in conjunction with the RADMAN computer codes, other approved computer codes or hand calculation.

Filters -Gross radioactivity measurement in conjunction with the FILTRK computer code, other approved computer codes or hand calculation.

All Other Waste Streams -Direct measurement of radionuclides in conjunction with the RADMAN computer codes, other approved computer codes or hand calculation.(b) Determination of the NRC waste classification is performed by comparing the measured or calculated concentrations of significant radionuclides in the final waste form to those listed in 1 OCFR Part 61.55.5.5. Quality Control[1] The RADMAN computer code provides a mechanism to assist the Plant in conducting a quality control program in accordance with the waste classification requirements listed in 1 OCFR Part 61.55. All waste stream sample data changes are written to a computer data file for future review and reference.

[2] Audits and Management Review includes the following: (a) Appendix G to 1OCFR20 requires conduct of a QC program which mustinclude management review of audits.(b) Management audits of the Plant Sampling and Classification Program SHALL be periodically performed to verify the adequacy of maintenance sampling and analysis.(c) Audits and assessments are performed and documented by any of the following:

Radiation Protection Department

Quality Assurance Department

Qualified Vendors (d) Certain elements of the Entergy Quality Assurance Program Manual are applied to the Process Control Program. [OAPM, Section A.l.c]EN-RW-105, Rev. 1

5.6. Dewatering

Operations

[1] Processing requirements during dewatering operations include the following: (a) All dewatering operations are performed per approved Plant or vendor operating procedures and instructions.(b) Dewatering limitations and capabilities are verified by vendor Topical Reports or Operating and Testing Procedures.

[2] Dewatered resin activity limitations include the following: (a) Dewatered resins will not be shipped off-site that have activities which will produce greater than 1.OE+8 rads total accumulated dose over 300 years. This is usually verified by comparing the'container specific activity at the time of shipment to the following concentration limits for radionuclides with a half-life greater than five years:* 10 Ci (0.37 TBq) per cubic foot.* 350 uCi (12.95 MBq) per cubic centimeter

5.7. Waste

Packaging Waste in final form will be packaged in accordance with Title 10 and Title 49 of the Code of federal regulations and in accordance with current burial site criteria as is detailed in EN-RW-102.

EN-RW-105, Rev. 1

5.8. Administrative

Controls[1] Information on solid radioactive waste shipped off-site is reported annually to the Nuclear Regulatory Commission in the Annual Radioactive Effluent Release Report as specified by the Offsite Dose Calculation Manual (ODCM) or Technical Specification.

[ANO1 Technical Specifications -5.6.3] [ANO2 Technical Specifications -6.9.3].[WF3 Technical Specifications -6.9.18] [GGNS ODCM -5.6.3.c] [JAF Technical Specifications -5.6.3] [PLP ODCM, Appendix A -IV. A].[2] All changes to the PCP SHALL be documented.

All records of reviews performed SHALL be retained as required by the Quality Assurance Program. The documentation of the changes SHALL [GGNS UFSAR, Chapter 16B.1 / TRM -7.6.3.8 paragraph 2]: (a) Contain sufficient information to support the change with appropriate analyses or evaluations justifying the change..(b) Include a determination that the change will maintain the overall conformance of the solidified waste product (if applicable) to existing requirements of Federal, State or other applicable regulations.

[3] All changes in the Process Control Program and supporting documentation are included in each site's next Annual Radiological Effluent Release Report to the Nuclear Regulatory Commission.

[ANO ODCM -L3.2.1.B]

[VTY TRM 6.12][4] The changes to EN-RW-1 05 SHALL become effective upon r'eview and, acceptance by the site's General Plant Manager except as listed below: (a) For Grand Gulf Nuclear Station, the changes to RW-1 05 SHALL be accomplished as specified in Grand Gulf Nuclear Station Technical Requirements Manual (TRM) Section 7.6.3.8. The changes SHALL become effective upon review and acceptance by the On-site Safety Review Committee (OSRC) and the approval of the GGNS Plant General Manager. [GGNS UFSAR, Chapter 16B.1 / TRM -7.6.3.8 paragraph 2](b) For River Bend Nuclear Station, the procedure approval along with changes to RW-1 05 SHALL be accomplished per the River Bend Nuclear Station Technical Requirements, Section 5.5.14.1.

The changes SHALL become effective upon review and acceptance by approval from the River Bend Nuclear Station Plant Manager or Radiation Protection Manager. [RBS Technical Requirements

-5.5.14.1, 5.5.14.2 & 5.8.2]EN-RW-105, Rev. 1

5.8 Continued

(c) For Waterford 3, the procedure approval along with changes to RW-1 05 SHALL be accomplished per Waterford 3 Technical Specifications 6.13.2. The changes SHALL become effective upon review and acceptance by the Waterford 3 General Plant Manager. [WF3 Technical Specifications-6.13.2.b](d) For James A. FitzPatrick Nuclear Station, the procedure approval along with changes to EN-RW-1 05 SHALL be accomplished per the James A. FitzPatrick Station Technical Specifications, Section 5.6.3. The changes SHALL become effective upon review and acceptance through approval from the James A.FitzPatrick Nuclear Station On-Site Safety Review Committee.

[JAF UFSAR, Chapter 11.3.5](e) For Vermont Yankee, Changes to the Process Control Program SHALL become effective after review and acceptance by the (OSRC) On-Site Safety Review Committee and the Site VP.(f) For IPEC, Changes to the Process Control Program SHALL become effective after final review and acceptance by the On-Site Safety Review Committee (OSRC).5.9. Vendor Requirements

[1] Vendors performing radwaste services under 1OCFR61 and 10CFR71 requirements will be on the Entergy Qualified Supplier's List (QSL). [QAPM, Section A.1.c][2] Vendors performing radwaste services on-site are to comply with the following: (a) Dewatering and solidification services SHALL have a NRC-approved Topical Report or other form of certification documenting NRC approval of the processes and associated equipment/containers.(b) All vendor procedures utilized for performing on-site radwaste processing services (to assure compliance with 10 CFR Parts 20, 61 and 71, State Regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste) will be reviewed per the requirements of EN-LI-1 00, technically by the applicable site's Radiation Protection organization and only be accepted per the approvals specified in Section 5.8 [4].(c) All changes to vendor procedures for ongoing on-site radwaste services will be reviewed technically by the site's Radiation Protection organization and screened per the requirements of EN-LI-1 00. Significant procedural changes will require the approvals specified in Section 5.8 [4]. During screening, the level of significance for procedural changes on equipment and process parameters may warrant the full 1 OCFR50.59 documentation and approval process.EN-RW-105, Rev. 1

5.9 continued

(d) Plant management SHALL review vendor(s) topical reports and test procedures per applicable requirements in Section 5.8.NOTE The PCP does not have to include the vendor's Topical Report if it has NRC approval, or has been previously submitted to the NRC.(e) Plant management review will assure that the vendor's operations and requirements.

are compatible with the responsibilities and operation of the Plant.(f) Training requirements and records listed in Section 5.10 also apply to contracted vendors.5.10. Miscellaneous

[1] Special tools and equipment (a) Frequency of Use and Descriptions Required tools and equipment will vary depending on the specific process and waste container that is used. The various tools and equipment which may be required are detailed in specific procedures developed to govern activities described in this document.[2] Pre-requisites (a) Maintenance of Regulatory Material Ensure that a current set of DOT, NRC, EPA and applicable State regulations, vendor processing facility and disposal site regulations and requirements are maintained at the site and are readily available for reference.

The use of web based regulations is acceptable.(b) Representative Radionuclide Sample Data Ensure that representative radionuclide sample data is on file for each active waste stream. Unless operation conditions or changes in processing methods require increased sample frequency, data is considered to be current if it meets the requirements of EN-RW-1 04.EN-RW-105, Rev. 1 5.10 continued (c) Initial and Cyclic Training* A training program SHALL be developed, implemented and maintained for all personnel involved in processing,, packaging, handling and transportation of radioactive waste to ensure radwaste operations are performed within the requirements of NRC Information Bulletin 79-19 and 49CFR Part 172.700 through Part 172.704." Training requirements and documentation also apply to contracted on-site vendors.NOTE Cyclic training is defined as within three years for DOT, and two years for IATA (d) Specific employee training is required for each person who performs the following job functions

[1 72.702(b)].

, Classifies hazardous materials.

Packages hazardous materials.

Fills, loads and/or closes packages.* Marks and labels packages containing hazardous materials.

Prepares shipping papers for hazardous materials.

Offers or accepts hazardous materials for transportation.

Handles hazardous materials.

Marks or placards transport vehicles.* Operates transport vehicles.* Works in a transportation facility and performs functions in proximity to hazardous materials which are to be transported." Inspects or tests packages.EN-RW-105, Rev. 1 5.10 continued (e) Cyclic training is defined as within three years for DOT & within two years for IATA.Copies of training records are required for as long as a person is employed and 90 days thereafter.

The records should include, as a minimum, the following:

Trainee's name and signature* Training dates* Training material or source reference* Trainer's information

6.0 INTERFACES

[1] EN-LI-1 00, "Process Applicability Determination"[2] EN-RW-104, "Scaling Factors"[3] EN-QV-1 04, "Entergy Quality Assurance Program Manual Control" 7.0 RECORDS[1] Documentation of pertinent data required to classify waste and verify solidification will be maintained on each batch of processed waste as required by approved procedures.

[2] Documentation will also be maintained to ensure that containers, shipping casks, and methods of packaging wastes meet applicable Federal regulations and disposal site criteria.

The records of reviews performed and documents associated with these reviews will be maintained as QA records.EN-RW-105, Rev. 1

8.0 OBLIGATION

AND REGULATORY COMMITMENT CROSS-REFERENCES Document Document NMM Procedure Site Applicability Section Section ANO ODCM L3.2.1.B 5.8 [3] ANO ANO1 Technical Specifications 5.6.3 5.8 [1] ANO ANO2 Technical Specifications 6.9.3 5.8 [1] ANO RBS Technical Requirements 5.5.14

RBS RBS Technical Requirements 5.5.14.1 5.8 [3] RBS 5.8 [4] (b)RBS Technical Requirements 5.5.14.2 5.8 [4] (b) RBS RBS Technical Requirements 5.8.2 5.8 [4] (b) RBS WF3 Technical Specifications 1.22

WF3 WF3 Technical Specifications 6.9.18 5.8 [1] WF3 WF3 Technical Specifications 6.13.2.b 5.8 [4] (c) WF3 JAF ODCM 6.2.1 5.8 [1] JAF JAF Technical Specifications 5.6.3 5.8 [1], 5.8 [4] JAF JAF FSAR Chapters 7 and 5.8 [4] JAF 11 11759 -NRC IN 79-19 All

WF3 GGNS UFSAR, Chapter 16B.1 / 7.6.3.8 1.0 GGNS TRM paragraph 1 GGNS ODCM 5.6.3.c 5.8 [1] GGNS GGNS FSAR 11.4.5.S2 5.9 [2](a) GGNS GGNS FSAR 11.4.2.3AS7 5.9 [2](a) GGNS IPN-99-079 All

IPEC-Appendix B Technical Section 4.5, *Specifications RECS ODCM IPEC Part 1 PLP Technical Specifications 5.5.15 5.8 [4] PLP PLP ODCM Appendix A -5.8 [1] PLP IV. A NRC Letter 1.98.091 All

VY NRC Letter 1.88.078 All

VY VY TRM 6.12 5.8(3) VY QAPM Section A.1.c

All'* Covered by directive as a whole or by various paragraphs of the directive.

9.0 ATTACHMENTS

None EN-RW-1 05, Rev. 1 '

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REVIEWS I APPROVALS Action'Taken Actipn.Date it:LatNe First Namee Faclity `Approal Note: all fields are taken from the TIMX600 panel for documents with an associated routing/approval (TIMX600) panel.

ML090641023

Text

1.0 INTRODUCTION

4.0 ENVIRONMENTAL

2.1 Radioactive

2.2 Radioactive

2.3 Radioactive

2.3.1 Concentration

2.4 Radioactive

2.4.4 Gaseous

2.5 Radioactive

2.5.1 Total

1.0 INTRODUCTION

2.0 LIQUID

3.1.1 Batch

3.1.2 Eberline

3.4.1 Total

0.6 miles

0.2 miles

0.9 miles

0.1 miles

0.4 miles

0.8 miles

0.4 miles

0.4 miles

2.1 miles

0.2 miles

1.0 DEFINITIONS

2.0 RELEASE

2.1 RADIOACTIVE

2.6.1 Radiological

2.6.3 Interlaboratory

3.0 ADMINISTRATIVE

1.0 DEFINITIONS

2.0 RELEASE

2.6.1 Monitoring

2.6.3 Interlaboratory

1.0 PURPOSE

2.0 REFERENCES

4.0 continued

5.0 DETAILS

5.1. Precautions

5.2. Waste

5.3. Waste

5.6. Dewatering

5.7. Waste

5.8. Administrative

5.8 Continued

5.9 continued

6.0 INTERFACES

8.0 OBLIGATION

9.0 ATTACHMENTS

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