Revision as of 12:38, 29 October 2019

Text

Amend 16 to Odcm.
	ML17290A995
Person / Time
Site:	Columbia
Issue date:	12/31/1993
From:	; WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	;
Shared Package
ML17290A993	List: GO2-94-051, Forwards WNP-2 Radioactive Effluent Release Rept Jan-Dec 1993 & Odcm; ML17290A994; ML17290A995;
References
	PROC-931231, NUDOCS 9403080127
	Download: ML17290A995 (212)
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WASHINGTON PUBLIC POWER SUPPLY'YSTEM WNP-2

.OFFSITE DOSE CALCULATION MANUAL, q+0228 05080~~7 o500059'7 }' .~

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'HENDHENT NO. 16

.DECEHBER 1993 .

OFFSITE DOSE CALCULATION HANUAL LIST OF EFFECTIVE PAGES Pa e Amendment Pa e Amendment 9 20 9 16 21 9 9 22 9 lv 9 23 9 9 24 14 Vl 9 25 9 Vl 1 16 26 9 1 15 27 9 2 15 28 15 3 15 29 9 4 15 30 9 5 15 31 ll 6 15 32 9 7 15 33 9 7a 15 34 ll 8 ll 34a ll 9 15 35 16 10 15 35a 16 ll 15 36 13 12 15 37 9 13 9 38 9 14 9 39 9 15 15 40 16 9 41 17 15 42 18 15 '43 19 LEP-1

AHENDHENT NO. 16 DECEHBER 1993 OFFSITE DOSE CALCULATION HANUAL LIST OF EFFECTIVE PAGES Pa e Amendment Pa e Amendment 45 9 73 ll 46 9 74 9 47 15 75 15 48 9 76 9 49 9 77 9 50 9 78 9 51 ll 79 9 52 ll 80 9 53 9 81 9 54 ll 82 9 55 ll 83 9 56 15 84 9 57 ll 85 15 58 ll 86 9 59 11 87 9 60 ll 88 9 61 ll 89 62 9 90 .9 63 15 91 9 64 15 9la 16 65 9 92 9 66 9 93 67 9 94 9 68 9 95 16 69 9 96 16 70 9 97 10 71 9 98 9 72 9 99 9 LEP-2

AMENDMENT NO. 16 DECEMBER 1993 OFFSITE DOSE CALCULATION MANUAL LIST OF EFFECTIVE PAGES Pa e Amendment Pa e Amendment 100 12 126 100a 12 127 10 101 15 128 9 102 ll 129 9 103 9 130 10 104 10 131 9 105,. 10 132 9 106 15 133 107 12 134 108 12 135 109 15 136 110 12 137 110a 12 138 10 111 15 139 112 9 140 113 9 141 114 9 142 115 9 143 11 116 10 144 9 117 10 145 118 9 146 119 10 147 120 10 148 121 16 149 122 13 150 16 123 9 151 124 9 152 10 125 13 153 LEP-3

AMENDMENT NO. 16 DECEMBER 1993 OFFSITE DOSE CALCULATION MANUAL LIST OF EFFECTIVE PAGES Pa e Amendment Pa e Amendment 154 ll 170 9 155 16 171 9 156 ll 172 10 157 9 9 173'74 158 15 9 159 9 175 9 160 9 176 9 161 16 177 9 162 9 163 9 164 9 165 9 166 16 167 9 168 9 169 9 LEP-4

AHENDHENT NO. 9

.JANUARY 1992 OFFSITE DOSE CALCULATION MANUAL TABLE 'OF CONTENTS Section Title ~Pe e

1.0 INTRODUCTION

2.0 LIQUID EFFLUENT DOSE CALCULATION ................. 1 2.1 Introduction 1 2.2 Radwaste Liquid Effluent Radiation Honitoring System 2 2.3 10 CFR 20 Release Rate Limits . 2 2.3.1 Pre-Release Calculation . 3 2.3.2 Post-Rel ease Cal cul ation 4 2.3.3 Continuous Release 5 2.4 10 CFR 50, Appendix I, Release Rate Limits 5 2.4.1 Projection of Doses .

2.5 Radwaste Liquid Effluent Dilution Ratio Alarm Setpoints Calculations 9 2.5.1 Introduction 9 2.5.2 Methodology for Determining the Maximum Permissible.Concentrati on (HPC) Fraction 9 2.5.3 Methodology for the Determination of Liquid Effluent Monitor S etpoint 10 2e6 Verification of Compliance with 10 CFR 50, Appendix I, and 10 CFR 20, Appendix B . 12 2.7 Methods for Calculating Dose to Man from Liquid Effluent Pathways 13 2.7.1 Radiation. Doses . 13 2.7.2 Plant Parameters ~ ~ ~ ~ ~ ~ ~ ~ 17

AMENDMENT NO. -16 C

DECEMBER 1993, OFFSITE DOSE CAL'CULATION MANUAL TABLE OF CONTENTS Section Ti tl e ~Pe e 2.8 Compliance With Technical Specification 3. 11. 1.4 ' . . . . . . . . 18 2.8.1 Maximum Allowable Liquid Radwaste Activity in Temporary Radwaste Hold-Up Tanks . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 2.8.2 Maximum Allowable Liquid Radwaste in Tanks That Are Not Surrounded by Liners, Dikes or Walls . . . . . . . . . . . . . . . . . . . . . 21 2.9 Liquid Process Monitors and Alarm Setpoints Calculations . . . . . 22 2.9.1 Standby Service Water (SW) Monitor . . . . . . . . . . . . . . . . 23 2.9.2 , Turbine Building Service Water (TSW) Monitor . . . . .'. . . . . . 23 2.9.3 Turbine Building Sumps Water (FD) Monitor . . . . . . . . . . .' . 24 3.0 GASEOUS EFFLUENTS DOSE CALCULATIONS . . . . . . . . . . . . . . . . 34 3.1 I ntroduction 34 3.2 Gaseous Effluent Radiation Monitoring System . . . . . . . . . . 35a 3.2. 1 Hain Plant Release Point . . . ; . . . . . . . . . . . . . . . . 35a 3.2.2 Radwaste Building Ventilation Exhaust Monitor . . . . . . . . . . . 36 3.2.3 Turbine Building Ventilation Exhaust Monitor . . . . . . . . . . . 37 3.3 10 CFR 20 Release Rate Limits . . 38 3.3.1 N oble Gases . 38 3.3.2 Radioiodines and Particulates . . . . . . . . . . . . ... . . . . . 39 3.3.2.1 Dose Parameter for Radionuclide i (P;) . . . . . . . . . . . . . . 41 3.4 10 CFR 50 Release Rate Limits . . . . . . . . . . . . . . . . . . . 42 3.4.1 Noble Gases (Requirement for Operability 6.2.2.2 (3.11.2.2) . . ; . 43 3.4.2 Radioiodines, Tritium and Particulates Requirement for Operability 6.2.2.3 (3.11.2.3) . . . . . . . . . . . . . . . . . . . . . . . . 45 3.4.2.1 Dose Parameter for Radionuclide i (R) 48

AMENDMENT NO. 9

,JANUARY 1992

'FFSITE DOSE CALCULATION MANUAL I

TABLE OF CONTENTS Section Title ~Pa e 3.4.3 Annual Dose at Special Locations 55 3.5 Compliance with Requirement for Operability 6.2.2.4 (3. 11.2.4) 55 3.5.1 Projection of Doses . 56

'3.6 Calculation of Gaseous Effluent Monitor Alarm Setpoints . . . . . . 56 3.6.1 Introduction 56 3.6.2 Setpoint Determination for All Gaseous Release Paths . . . . . . . 56 3.6.2. 1 Setpoints Calculations Based on Whole Body Dose Limits . . . . . . 57 3.6.2.2 Setpoints Calculations Based on Skin Dose Limits . . . . . . . . . 60 4.0 COMPLIANCE WITH 40 CFR 190 92 4.1 Requirement for Operability . . . . 92 4.2 ODCH Methodology for Determining Dose and Dose Commitment from Uranium Fuel Cycle Sources . . . . . . . . . . . . 92

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

. . . . . . . .

4.2.1 Total Dose from Liquid Effluents 93 4.2.2 Total Dose from Gaseous Effluents . . . . . . ....... . . . . . . . 93 4.2.3 Direct Radiation Contribution . . . . . . . . ~ . . . . . . . . . . 93 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING............... 93 5.1 Radiological Environmental Monitoring Program (REHP) . . . . . . . 94 5.2 Land Use Census . 95 5.3 Laboratory Intercomparison Program 96 5.4 Reporting Requirements 97 J

6.0 CONDUCT OF TESTS AND INSPECTIONS IN SUPPORT OF WNP-2 RADIOACTIVE EFFLUENT AND RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAMS . 113 6.1 Instrumentation, in Support of WNP-2 Radioactive Effluent Monitoring Requirement for Operability . . . . . . . . . . . . . 115 111

AMENDMENT NO. 9 JANUARY 1992 OFFSITE DOSE, CALCULATION MANUAL TABLE OF CONTENTS Section Title ~Pe e 6.2 ,

Requirement for Operability in Support of the Radioactive Effluent Monitoring Programs . . . . . . . . . . . . . . . . . . 128 6.3 Requirement for Operability in Support of the Radiological Environmental Monitoring Programs . . . . . . . . . . . . . . . . 148 6.4 Radiological Environmental Operating/Radioactive Effluent Release Report Requirements and Control of Changes . . . . . . . 163 6.5 B ases . 168

AMENDMENT NO. 9 JANUARY 1992 OFFSITE DOSE CALCULATION MANUAL LIST OF TABLES Section Ti tl e Pacae Section 2.0 2-1 Fish Bioaccumulation Factors (BF;) and Adult Ingestion Dose Conversion Factors (DF,) . . . . . . . . . . . . . . . . . . . . . 25 2-2 Ingestion Dose Factors (A;;) for Total Body and Critical Organ . . . 28 2-3 Input Parameters Used to Calculate Maximum Individual Dose From Liquid Effluents . . . . . . . . . . . . . . . . . . . . . . . 31 Section 3.0 Dose Factors for Noble Gases and Daughters . . . , . . . . . . . . 62 Distances (Miles) to Typical Controlling Locations as Measured from Center of WNP-2 Containment Building . 63 3-3 WNP-2 Long-Term Average Dispersion (X/g) and Deposi tion (D/g)

Values for Typical Locations 64 3-4 Dose Rate Parameters. Implementation of 10 CFR 20, Airborne Releases 65 3-5a Dose Rate Parameters. Implementation of 10 CFR 50, Airborne Releases - Age Group: Adult 67 3-5b Dose Rate Parameters. Implementation of 10 CFR 50, Airborne Releases - Age Group: Teen . 68 3-Sc Dose Rate Parameters. Implementation of 10 CFR 50, Airborne Releases - Age Group: Child 69 3-5d Dose Rate Parameters. Implementation of 10 CFR 50, Airborne Releases - Age Group: Infant . . . . 70 3-6 Input Parameters for Calculating R,. 71 3-7 Input Parameters for Calculating RM, 72 3-8 Input Parameters for Calculating R", 73

AMENDMENT NO. 9 E JANUARY 1992 OFFSITE DOSE CALCULATION MANUAL LIST OF TABLES Section Title ~Pe e 3-9 Input Parameters Needed for Calculating Dose to the Maximum Individual from WNP-2 Gaseous Effluent . . . . . . . . . . . . . . 74 3-10 Reactor Building Stack X/Q and D/Q Values . 76 3-11 Turbine Building or Radwaste, Building X/Q and D/Q Values 80 3-13 Characteristics of WNP-2 Gaseous Effluent Release Points 84 3-14 References for Values Listed in Table 3-9 . 85 3-15 Design Base Percent Noble Gas (30-Minute Decay) 86 3-16 Annual Doses at Typical Locations Source: WNP-2 Gaseous Effluent . 87 3-17 Annual Occupied Air Dose at Typical Locations . 88 Section 5.0 5-1 Radiological Environmental Monitoring Program Plan 100 5-2 WNP-2 REHP Locations 104 5-3 Environmental Radiological Monitoring Program Annual Summary . . . . . 111 5-4 Reporting Levels for Nonroutine Operating Reports . . . . . . . . . . . 112 Section 6.0 6.1.1.1-1 (3.3.7.11-1) Radioactive Liquid Effluent- Monitoring Instrumentation . . . . . . . . . . . . . . . . 117 6.1.1.1.1-1 (4.3.7.11-1) Radioactive Liquid Effluent Monitoring Instrumentation Periodic Tests and Inspections 119 6.1.2.1-1 (3.3.7.12-1) Radioactive Gaseous Effluent Monitoring Instrumentation . . . . . . . . . . . . . . . . 122 vi

AMENDMENT NO. 16 DECEMBER 1993 OFFSITE DOSE CALCULATION MANUAL TABLE OF CONTENTS Section Ti tl e ~Pa e 6.1.2.1.1-1 (4.3.7.12-1) Radioactive Gaseous Effluent Monitoring Instrumentation Periodic Tests and Inspections Requirements . . . . . . . . . . . 125 6.2.1.1.1-1 (4.11-1) Radioactive Liquid Waste Sampling and Analysis Program . . . . . . . . . . . . . . . 130 6.2.2.1.2-1 (4.11-2) Radioactive Gaseous Waste Sampling and Analysis Program . . . . . . . . . . . . . . . 136 6.3.1.1-1 (3.12-1) Radiological Environmental Honitoring Program . 151 6.2.1.1-2,(3.12-2) Reporting Levels for Radioactivity Concentrations in Environmental Samples . 157 6.3.1.1.1-1 (4.12-1) Detection Capabilities for Environmental Sample Analysis . 158 LIST OF FIGURES FIGURE Ti tl e ~Pa e 2-1 Simplified Block Diagram of Liquid Waste System . ~ ~ ~ ~ 32 2-2 Simplified Block Diagram of Solid Radwaste System . ~ ~ ~ ~ 33 3-1 Site Boundary for Radioactive Gaseous and Liquid Effluents . . . . 89 3-2 Simplified Block Diagram of Gaseous Waste System ~ ~ ~ ~ 90 3-3 Simplified Block Diagram of Off-Gas Treatment System ~ ~ ~ ~ 91 3-4 Auxiliary Boiler 91a 5-1 Radiological Environmental Honitoring Sample Locations Inside of 10-Mile Radius 109 5-2 Radiological Environmental Monitoring Sample Locations Outside of 10-Mile Radius . 110 Radiological Environmental Monitoring Sample Locations Near Plant 2 ~ ~ ~ ~ ~ ~ ~ ~ ~

vi 1

AMENDMENT NO. 15 OCTOBER 1993

'.

0 INTRODUCTION The purpose of this manual is to provide the information and methodologies to be used by the -Washington Public Power Supply System to satisfy the.

requirements of 10 CFR 20. 106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50.

2.0 LI UID EFFLUENT DOSE CALCULATION The U.S. Nuclear Regulatory Commission's computer program LADTAP II can be used for dose analysis for liquid radioactive effluents from WNP-2 into surface waters. The analyses estimate radiation dose to individuals, population groups, and biota from ingestion (aquatic foods, water, and terrestrial irrigated foods) and external exposure (shoreline, swimming, and boating) pathways. The calculated doses provide for determining compliance with Appendix I to 10 CFR Part 50.

2. 1 Introduction Liquid radwaste released from WNP-2 will meet 10 CFR 20 limits at the point of discharge to the Columbia River. Actual discharges of liquid radwaste effluents will only occur on a Batch Basis, and the average concentration at the point of discharge will be only a small percentage of the allowed limits.

A simplified block diagram of the liquid waste management system and effluent pathways is contained in Figure 2-1. Solid radioactive wastes are disposed of by way of an approved disposal site., A simplified block diagram of the solid radwaste system is described in Figure 2-2.

The cumulative quarterly dose contributions due to radioactive liquid effluents released to the unrestricted areas will be determined once every 31 days using the LADTAP II computer code.

AMENDMENT NO. 15 OCTOBER 1993

'he. dose contributions will be calculated for all.,radionuclides identified in the released effluent based on guidelines provided by NUREG-0133.

The methods for calculating the doses are discussed in Section 2.4 of this manual.

I 2.2 Radwaste Li uid Effluent Radiation Monitorin S ste This monitoring subsystem measures the radioactivity in the liquid effluent prior to its entering the cooling tower blowdown line.

All radwaste effluent passes through a four-inch line which has an off-line sodium iodide radiation monitor. The radwaste effluent flow, variable from 0 to 190 gpm, combines with the 36-inch cooling water blowdown line, variable from 0 to 7500 gpm and is discharged to the Columbia River with a total flow based on MPC; total, and cooling water flushing needs.

The radiation monitor is located on the 437'evel of the Radwaste Building and has a minimum sensitivity of 10'Ci/cc for Cs-137. The radiation indicator has seven decades of range.

2.3 10 CFR 20 Release Rate Limits The requirements pertaining to discharge of radwaste liquid effluents to the unrestricted area are specified in Requirement for Operability 6.2. 1. 1 (3.1l.l.l):

"The concentration of radioactive material released from the site to unrestricted areas shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II,.Column, 2 for .

radionuclides other than noble gases, and 2 x 10'Ci/ml total activity concentration for all dissolved or entrained noble gases."

AMENDMENT NO. 15

'OCTOBER 1993

-In order to comply with the requirements stated above, limits will be set,to assure that blowdown line concentrations do not exceed 10 CFR 20, Appendix 8, Table II, Column 2 at any time. "

2.3. 1 Pre-Release Calculation The activity of the radionuclide mixture and the liquid effluent discharge rate will be determined in accordance with Supply System procedures. The effluent concentration is determined by the following equation:

Ci x fw ft where:

Conc, Concentration of radionuclide i in the effluent at point of discharge - pCi/ml.

C, Concentration of radionuclide i in the batch to be released

- pCi/ml.

fw Discharge flow rate from sample tank to the blowdown line-variable from 0 to 190 gpm.

fb Blowdown flow rate - variable from 0 to 7500 gpm.

Total discharge (ft = fb + fw) flow rate - variable from 0 to 7690 gpm.

The calculated concentration in the blowdown line must be less than the concentrations listed in 10 CFR 20, Appendix B. Before releasing the batch to the environment, the following equation must hold:

AMENDMENT NO. I5 OCTOBER 1993 (Conc,/HPC,.) g 1 (2) ie1 where:

Conc, The concentration of radionuclide i in the effluent at the point of discharge into the river.

maximum permissible concentration of nuclide i as listed in 10 CFR 20, Appendix B, Table II.

Total number of radionuclides in the batch.

2.3.2 Post-Release Calculation The concentration of each radionuclide in the unrestricted area, following the batch release, will be calculated as follows:

The average activity of radionuclide i during the time period of the release is divided by the Plant Discharge Flow/Tank Discharge Flow ratio yielding the concentration at the point of discharge:

Ckx fw Conc .

ik (3) where:

Colic;k The concentration of radionuclide i in the effluent at the point of discharge during the-release period k'-'pCi/ml).

Cik The concentration of radionuclide i in the batch during the release period k - (pCi/ml).

fw Discharge flow rate from sample tank to the blowdown line

- variable from 0 to 190 gpm.

AMENDMENT NO. 15 OCTOBER 1993 fb Blowdown flow rate - variable from 0 to 7500 gpm.

Total discharge (ft - fb + fw) flow rate -'variable from 0 to 7690 gpm.

To assure compliance with 10 CFR 20, the following relationships must hold:

(Conc,/MPC.,) g 1 (4) i~1 where the terms are as defined in Equation (2).

2.3.3 Continuous Release Continuous release of liquid radwaste effluent is not planned for WNP-2.

However, should it occur, the concentrations of various radionuclides in the unrestricted area would be calculated according to Equation (3) and Equa-tion (4). To show compliance with 10 CFR 20, the two equations must again hold.

2.4 10 CFR 50 A endix I Release Rate Limits Periodic Test and Inspection 6.2. 1.2. 1 (4.11. 1.2) requires that the cumulative dose contributions be determined in accordance with the ODCM at least once per 31 days. Requirement for Operability 6.2. 1.2 (3. 11. 1.2) specifies that the dose to a member of the public from radioactive material in liquid effluents released to the unrestricted area shall be limited to:

<1.5 mrem/Calendar quarter - Total Body and

~5.0 mrem/Calendar quarter - Any Organ.

AMENDMENT NO. 15 OCTOBER 1993 The cumulative dose for the calendar year shall be .limited to:

~3 mrem - Total Body and

~10 mrem - Any Organ.

I The maximum exposed individual is assumed to be an adult whose exposure pathways include potable water and fish consumption. The choice of an adult as the maximum exposed individual is based on the highest fish and water consumption rates shown by that age group and the fact that most of the dose from the liquid effluent comes from these two pathways.

The dose contribution will be calculated for all radionuclides identified in the liquid effluent released to the unrestricted area, using the following equation:

Dr = /{A,, g htsCF~) (5) i l~1 where:

Dr The cumulative dose commitment to the total body or morgan, 7', from liquid effluents for the total time period g ht,,

la1 ~

in mrem.

The length of the 2th time period over which C and F, are averaged for all liquid releases, in hours.

The number of rele'ases for the time period under consideration.

C The average concentration of radionuclide i in undiluted liquid effluent during time period ~t, from any liquid release, in pCi/ml.

AMENDMENT NO. 15 OCTOBER" 1993 The site-related ingestion dose-commitment factor to the total body or any organ r for each identified principle gamma and beta emitter listed in Table 2-2, in mrem/hr- per

,pCi/ml.

F, The near field average dilution factor for C, during any liquid waste release. This is defined as the ratio of the maximum undiluted I liquid waste flow during release to the product of the average flow from the site discharge structure to unrestricted receiving waters times 500.

While the actual discharge structure exit flow is variable from 0 to 17. 1 cfs (0 to 7690 gpm), a maximum flow value of 2.0 cfs will be used for dose calculation purposes in accordance with the NUREG-0133 requirement that the product of the average blowdown flow to the receiving water body, in cfs and the applicable factor (500), is 1000 cfs or less.

Liquid Radioactive Waste Flow fw Discharge Structure Exit Flow x 500 ft x 500 Th<<<<m A; the ingestion dose factors for the total body and critical organs, are tabulated in Table 2-2. It embodies the dose factor, fish bioaccumulation factor, pathway usage factor, and the dilution factor for the plant diffuser pipe to the Richland potable water intake. The following equation was used to calculate the ingestion dose factors:

where: (7)

A;, The composite dose parameter for total body or critical organ of an adult for nuclide i (in mrem/hr per pCi/ml).

AMENDMENT NO. 15 .

OCTOBER. 1993 I

A conversion factor: .h 1.14E+05 = (10 pCi/pCi) x (10 ml/liter)/8760 hr/yr..

730 liter/yr - which is the annual water consumption by the maximum adult (Table E-4 of Regulatory Guide 1. 109,

~

Revision 1).

Bioaccumulation factor for radionuclide i in fish - (pCi/Kg per pci/liter) (Table A-1 of Regulatory Guide 1. 109, Revision 1 and NUREG/CR-4013).

7a

AMENDMENT NO. 11 AUGUST 1992 DFi 'Adult ingestion dose conversion factor for. nuclide i - Total body or critical organ, 7, in'(mrem/pCi) (Table E-11 of Regulatory Guide 1. 109, Revision 1 and NUREG/CR-4013).

D Dilution factor from near field area (within one-quarter mile of the release point) to the Richland potable water intake - 100.

Uq = Adult fish consumption, 21 kg/yr (Table E-5 of Regulatory Guide 1. 109, Revision 1).

The values of BF, and DF, are listed in Table 2-1. Dilution assumptions, calculations, and LADTAP II input parameters are provided in Radiological Health Calculation Log 92-2.

The quarterly limits mentioned before represent one-half of the annual design objective of Section II.A of 10 CFR 50, Appendix I. If any of the limits (either that of the calendar quarter or calendar year) are exceeded, a special report pursuant to Section IV.A of 10 CFR 50, Appendix I, shall be filed with the NRC.

2.4. 1 Pro 'ection of Doses The projected doses due to releases of WNP-2 radwaste liquid effluents will be calculated for each batch, us'ing Equation (5) or LADTAP II. If the sum of the accumulated dose to date for the month and the projected dose for the remainder of the month exceeds the Requirement for Operability 6.2. 1.3 (3. 11. 1.3) limits, then the liquid. radwaste treatment .system shall. be used..

This is to ensure compliance with Requirement for Operability 6.2. 1.3 (3. 11. 1.3). This Requirement for Operability states that the liquid radwaste treatment system shall be maintained and the appropriate subsystem shall be used if the radioactive materials in liquid waste, prior to their discharge, when the dose, due to liquid effluent release to unrestricted areas when averaged over the month would exceed 0.06 mrem to total body or 0.2 mrem to any organ.

AMENDMENT NO. 15 OCTOBER 1993 2.5 Radwaste Li uid Effluent Dilution Ratio and Alarm Set pints Calculations 2.5. 1 Introduction The dilution alarm ratio and setpoints of the sample liquid effluent monitor are established to ensure that the limits of 10 CFR 20, Appendix B, Table II, Column 2, are not exceeded in the effluent at the discharge point (i.e.,

compliance with Requirement for Operability 6.2.1.1 (3. 11.1. 1), as discussed in section 2.3. 1 of this manual).

The alarm (HI) and the alarm/trip (HI-HI) setpoints for the liquid radwaste effluent monitor are calculated from the results of the radiochemical analysis of the effluent sample. The setpoints will be set into the radwaste monitor just prior to the release of each batch of radioactive liquid.

2.5.2 Methodolo for Determinin the Maximum Permissible Concentration HPC Fraction Radwaste liquid effluents can only be discharged to the environment through the four-inch radwaste line. The maximum radwaste discharge flow rate is 190 gpm. Prior to discharge, the tank is isolated and recirculated for at least thirty minutes, and a representative sample is taken from the tank. An isotopic analysis of the batch will be made to determine the sum of the HPC fraction (MPC,),based on 10 CFR 20 limits. From the sample analysis and the MPC values in 10 CFR 20, the HPC, is determined using the following equation.

MPC, =g HPC.,

where:

HPCq Total fraction of the Haximum Permissible Concentrations (HPCs) in the liquid effluent waste sample.

C( The concentration of each measured radionuclide i observed by the radiochemical analysis of the liquid waste sample (pCi/ml).

AMENDMENT NO. 15 OCTOBER 1993 HPCi The limiting concentrations of the appropriate radionuclide i from. 10 CFR 20, Appendix B, Table II,'olumn 2. For dissolved or entrained noble gases, the concentration shall be limited to 2.0E-04 pCi/ml total activity.

The total number of measured radionuclides in the liquid batch to be released.

If the HPC, is less than or equal to 0.8, the liquid batch may be released at any radwaste discharge or blowdown rate. If the HPC, exceeds 0.8, then a dilution factor (Fd) must, be determined. The liquid effluent radiation monitor responds proportionally to radioactivity concentrations in the undiluted waste stream. Its setpoint must be determined for diluted releases.

2.5.3 Hethodolo for the Determination of Li uid Effluent Monitor Set pints The measured radionuclide concentrations are used to calculate the dilution factor (Fd), which is the ratio of the total discharge flow rates (fw + fb) to the radwaste tank effluent flow rate (fw) that is required to assure that the limiting concentrations of Requirement for Operability 6.2.1. 1 (3. 11. 1. 1) are met at the point of discharge.

The dilution factor (Fd) is determined according to:

m I

x Fs (9)

HPC,.

where:

Fd The dilution factor required for compliance with 10 CFR 20, Appendix B, Table II, Column 2.

C, The concentration of each radionuclide i observed by radiochemical analysis of the liquid waste sample (pCi/ml).

10

AMENDMENT NO. 15 OCTOBER 1993 NPC, The limiting concentration of the appropriate radionuclide i from 10 CFR 20, Appendix 8, Table II, Column 2. For dissolved or entrained noble gases, the concentration shall be limited to 2.0E-04 pCi/ml total activity.

Fs The safety factor; a conservative factor used to compensate for statistical fluctuations and errors in measurements.

- For example, a safety factor (Fs) of 1.5 corresponds to a fifty (50) percent P) variation. The safety factor is 1.5.

The total number of measured radionuclides i in the liquid batch to be released.

The dilution which is required to ensure compliance with Requirement for Operability 6.2. l. 1 (3. 11. 1. 1) concentration limits will be set such that discharge rates are:

fw+fb fd 5 (10) fw and follows that:

fwg- fb Fd-1 (10a) or fb z fw(Fd-1) (10b) where:

Fd The dilution factor from Equation (9).

fw The discharge flow rate from the liquid radwaste tank to the blowdown line - variable from 0 to 190 gpm.

The cooling tower blowdown flow rate - variable from 0 to 7500 gpm.

The liquid effluent radiation monitor response is based on the results of the radiochemical analysis of the waste solution. Therefore the calculation for 11

AMENDMENT NO. 15 OCTOBER 1993 the radiation monitor's alarm (HI) and alarm/trip (HI-HI) setpoints are:

~ ~ ~

alarm (HI) = 0.80 CF~ + BKG + K[0.80 CF+ BKG ]'"

alarm /trip (HI-HI) = CF~ + BKG + K[CF~ + BKG ]'" (12) j where:

The dilution factor from Equation (10)

(C,.x E,) represents the count rate from the ie1 radionuclides in the liquid radwaste.

C; The concentration of each measured radionuclide i observed by radiochemical analysis of the liquid waste sample (pCi/ml ) .

Same as for Equation (9).

E, The radwaste effluent monitor's response to radionuclide i (count rate per pCi/ml).

BKg Background count rate of the radwaste effluent monitor.

K A constant to compensate for normal expected statistical Appendix variations in the liquid effluent radiation monitor count rate to reduce the chance of false alarms/trips K=3.

2.6 Verification of Com liance with 10 CFR 50 A endix I and 10 CFR 20 B

Verification of compliance with 10 CFR 50, Appendix I, and 10 CFR 20, Appendix B, limits will be achieved by following WNP-2 Plant Procedures for liquid discharge and the periodic application of the LADTAP II computer code.

12

AMENDMENT NO. 9

,JANUARY 1992 2.7 Methods for Calculatin Doses to Man From Li uid Effluent Pathwa s Dose models presented in NRC Regulatory Guide, 1. 109, Revision 1, as incorporated in the LADTAP II computer code, will be used for offsite dose calculation. The details of the computer code', and user instruction, are included in NUREG/CR-4013, "LADTAP II - Technical Reference and User Guide."

2.7. 1 Radiation Doses Radiation doses from potable water, aquatic food, shoreline deposit, and irrigated food pathways will be calculated by using the following equations:

a. Potable Water UM, R

pj 1 1 00 . t g,. D ...j e x p ( -A, ~ ) (13)

b. Aquatic Foods UMp pj 11 0 P B pD pjexp ( -A,.t,) (14)

c. Shoreline Deposits R., = 110,000 "',

UM,W g DTD.. [exp (-At,) (1

"-

exp(-1t,)] (15) 13

AMENDMENT NO; 9 JANUARY 1992

d. Irrigated foods For all radionuclides except tritium:

r [1 exp(- A~t.] f,B,.[l exp(-A,.t,) ]

R pj U gI i d e xp ( A h ) 0 YXG PA.i r [1 exp(-),E,.t.)

U animal ap gI F iAD sipj gFd,.exp(-A,.t)

'A~

, ]

f, B,.[l- exp(-A,.t) ] (16) iAw~AW pg l For tritium: .

Uveg animal <C n + C spj C vD spi

+ U D spj F A( v<F Aw~Aw) (17) ap ap where:

B;p The equilibrium bioaccumulation factor for nuclide i in pathway p, expressed as the ratio of the concentration in biota (in pCi/kg) to the radionuclide concentration in water (in pCi/liter), in liters/kg.

B; The concentration factor for uptake of radionuclide i from soil by edible parts of crops, in pCi/kg (wet weight) per pCi/kg dry soil.

C;A The concentration of radionuclide i in water consumed by animals, in pCi/liter.

Civ The concentration of radionuclide i in vegetation, in pCi/kg.

AMENDMENT NO. 15 OCTOBER 1993 The dose factor specific to a given age group a, radionuclide i, pathway p, and organ j, which can be used to calculate the radiation dose from an intake of a radionuclide, in mrem/pCi, or from exposure to a given concentration of a radionuclide in sediment,'xpressed as a ratio of the dose rate (in mrem/hr) and the areal radionuclide concentration (in pCi/m').

The deposition rate of nuclide i in pCi/m per hour.

The flow rate of the liquid effluent, variable from 0 to 2.0 cfs, for dose calculation purposes.

The fraction of the year crops are irrigated, dimensionless.

The stable element transfer coefficient that relates the daily intake rate by an animal to the concentration in an edible portion of animal product, in pCi/liter (milk) per pCi/day or pCi/kg (animal product) per pCi/day.

The mixing ratio (reciprocal of the dilution factor) at the point of exposure (or the point of withdrawal of drinking water or point of harvest of aquatic food), dimensionless.

The effective "surface density" for soil, in kg (dry soil)/m'Table E-15, Regulatory Guide 1. 109, Revision 1).

The consumption rate of contaminated water by an animal, in liters/day.

The consumption rate of contaminated feed or forage by an animal, in kg/day (wet weight).

The release rate of nuclide i in Ci/yr.

15

AMENDMENT NO. 9 JANUARY 1992 The'raction of deposited activity retained on crops, dimensionless (Table E-15, Regulatory Guide 1. 109, Revision 1).

The total annual dose to organ j of individuals of age group a from all of the nuclides i in pathway p, in mrem/yr.

The period of time for which sediment or soil is exposed to the contaminated water, in hours (Table E-15, Regulatory Guide 1. 109, Revision 1).

The time period that crops are exposed to contamination during the growing season, in hours (Table E-15, Regulatory Guide 1.109, Revision 1).

A holdup time that represents the time interval between harvest and consumption of the food, in hours (Table E-15, Regulatory Guide 1. 109, Revision 1).

The radioactive half life of nuclide i in days.

The average transit time required for nuclides to reach the point of exposure. For internal dose, t, is the total time elapsed between release of the nuclides and ingestion of food or water, in hours (Table E-15, Regulatory Guide 1.109, Revision 1).

A usage factor that specifies the exposure time or intake rate for an individual of age group a associated with pathway p, in hr/yr, l/yr, or kg/yr (Table E-5, Regulatory Guide 1.109, Revision 1).

The shoreline width factor, dimensionless (Table A-2, Regulatory Guide 1. 109, Revision 1).

16

AMENDMENT NO. 15 OCTOBER 1993 Y The agricultural productivity (yield), in kg (wet weight)/m'Table E-15, Regulatory Guide 1. 109, Revision 1).

The effective removal rate constant for radionuclide i from crops, in hr',. where A~ = A,. + A=, A,'is the radioactive decay constant, and A is the removal rate constant for N

physical loss by weathering (Regulatory Guide 1.109, Revision 1, Table B-15).

The radioactive decay constant of nuclide i in hr' 1100 The factor to convert from (Ci/yr)/(ft'/sec) to pCi/liter.

110)000 The factor to convert from (Ci/yr)/(ft'/sec) to pCi/liter and to account for the proportionality constant used in the sediment radioactivity model.

These equations yield the dose rates to various organs of individuals from the exposure pathways mentioned above.

2.7.2 Plant Parameters WNP-2 is a river shoreline site with a variable effluent discharge flow rate 0 to 7690 gpm. ,The population center nearest WNP-2 is the city of Richland, where drinking water withdrawal takes place. The applicable dilution factor is 50,000, using average river flow. The time required for released liquids to reach Richland, approximately 12 miles downstream, is estimated at 4.0 hours0 days 0 hours 0 weeks 0 months . Richland is the "realistic case" location, and doses calculated for the Richland location are typically applicable to the population as a whole.

Individual and population doses based on Richland parameters are calculated for all exposure pathways.

Only the population downstream of the WNP-2 site is affected by the liquid effluents released. There is no significant commercial fish harvest in the 50-mile radius region around WNP-2. Sportfish harvest is estimated at 14,000 kg/year.

17

AMENDMENT NO. 15

'

OCTOBER 1993 For irrigated foods exposure pathways, it can be .assumed that production

~

within the 50-mile radius region around WNP-2 is sufficient to satisfy

~ ~ ~

consumption requirements; Other relevant parameters relating to the irrigated foods pathways are defined as follows:

A 1 Add d Ad

~Food T e (liter/m'/mo) (kg/m') (Days)

Vegetation 150 5.0 70 Leafy Vegetation 200 1.5 70 Feed for Milk Cows 200 1.3 30 Feed for Beef Cattle 160 2.0 '30 Source terms are measured based on sampled effluent.

Table 2-3 summarizes the LADTAP II input parameters. Documentation and/or calculations of these parameters are discussed in detail in HPI 2.3, and Radiological Health Calculation Log 92-2.

2.8 Com liance with Technical S ecification 3.11.1.4 2.8. 1 Maximum Allowable Li uid Radwaste Activit in Tem orar Radwaste Hold-U Tanks The use of temporary liquid radwaste hold-up tanks is planned for WNP-2.

Technical Specification 3. 11. 1.4 states the quantity of radioactive material contained in any outside temporary tanks shall be limited.to the limits calculated in the ODCM such that a complete release of the tank contents would not result in a concentration at the nearest offsite potable water supply that would exceed the limits specified in 10 CFR Part 20 Appendix B, Table II.

18

AHENDHENT NO. 9

.JANUARY 1992 I

Equation '(18) will be used to calculate the curie limit for a temporary radwaste hold-up tank.'. The total tank concentration will be limited to less than or equal to ten (gl0) curies, excluding .tritium and dissolved or

,entrained gases.

Surveillance requirement 4. 11. 1.4, states that the quantity of radioactive material in"the hold-up tanks shall b'e determined to be within the limit by analyzing a representative sample of the tank's contents at least -once per 7 days when radioactive materials are being added to the tank.

where:

A

.e"'18) fi HPC, Aq Total allowed activity in tank (curies).

A; Activity of radioisotope i (curies).

HPC; Haximum permissible concentration of radionuclide i (10 CFR 20, Appendix B, Table II, Column 2).

Decay constant (years') radioisotope i.

Transit time of ground water from WNP-2 to WNP-1 well (WNP-2 FSAR Section 2.4) = 67 years.

Fraction of radioisotope fi = A,.

gA,.

Index for all radioisotopes in tank except tritium and noble gases.

Dispersion constant based on hydrological parameters, (2.4E+05 Ci per pCi/cc.)

19

,AMENDMENT NO. 9 JANUARY 1992 The total allowed activity (A,) is based on limiting WNP-1 well water to less than 1 MPC, of the entire liquid content of the tank spilled to ground and then. migrated via ground water to the WNP-1 well., The WNP-1 well is the location of maximu'm concentration since it is the nearest source of ground water and conditions are such that no spill of liquid should reach surface water. The 70-85 foot depth of the water table and the low ambient moisture of the soil requires a rather large volume of spillage for the liquid to even reach the water table in less than several hundred years. However, allowed tank activity (AT) is conservatively based on,all liquid radwaste in the tank instantaneously reaching the water table.

The hydrological analysis performed for the WNP-2 FSAR (Section 2.4) deter-mined that the transit time through the, ground water from WNP-2 to the WNP-1 well is 67 years for Strontium and 660 years for Cesium. These two radio-nuclides are representative of the 'radionuclides found in liquid radwaste.

Strontium is a moderate sorber and Cesium strongly sorbs to soil particles.

This calculation conservatively treats all radionuclides as moderate sorbers with a transit time of 67 years.

The concentration of each radionuclide in the well (CW;) is simply the con-centration in the tank (CT;) adjusted for radioactive decay during transit (e"') and divided by the minimum concentration reduction factor (CRF;).

Limiting well concentration to 1 MPC yields:

CW,. CT,.e (19)

.CRF "'PC,

.MPCi (Front Section 2.4 of WNP-2 FSAfti CRF . (4')

'a X a a)"

(20) 2V where:

Migration distance = 1 mile.

Volume of tank.

a, u, a, Dispersion constants.

20

AMENDMENT NO. 9

'JANUARY 1992 Combining Equations (19) and (20) yields:

CT,2V e x L) 'e a .a

"'4

)'I MPC (21)

Substituting A; for CT; V and reorganizing terms yields:

(4 m L)'" (e a e,)'" A.

MPC,.e

'~'22)

Making the following substitutions A,. =f,A~

L)'" (e )'"

K~-

(4 ~

2 e n x 10 'i/pCi = 2.4 x 10'i per Ci p

CC yields:

f.

K=

MPC.,e '"'r 2,8.2 Maximum K

f,.

MPC e '"'24)

Allowable Li uid Radwaste in Tanks That Are Not Surrounded b Liners Dikes or Walls Although permanent outside liquid radwaste tanks which are not surrounded by liners, dikes, or walls are not planned for WNP-2, Equation (18) will be used should such tanks become necessary in the future.

21

AMENDMENT NO. 9 JANUARY 1992 2;9 Li uid Process Monitors and Alarm Set pints Calculations I

As mentioned in Section 2.2 of this manual, all liquid radwaste effluent is discharged through a four-inch line 'that is monitored by an off-line sodium iodide radiation monitor. This monitor is located on the 437'evel of the Radwaste Building. All WNP-2 radwaste liquid effluent is discharged to the Columbia River th'rough the 36-inch Cooling Water Blowdown line. In addition to the liquid effluent discharge monitor there are three liquid streams that are normally nonradioactive but have a finite possibility of having radioactive material injected into them. These liquid streams are:

~ Standby Service Water (SW)

~ Turbine Building Service Water (TSW)

~ Turbine Building Sump Water (FD)

To prevent any discharges of radioactive liquid from these streams, radiation monitoring systems have been installed to detect any increase above the normal background concentration of radioactive material.

Alarm/setpoints are established to prevent any release of radioactive material in concentrations greater than 10 CFR 20 limits. The maximum radiation detector setpoint calculation for the three systems is based on the MPC; concentration of Cs-137 which is 2.0E-05 pCi/ml. The following equation is used to calculate the maximum setpoint:

Setpoint max. = [(2.0E-05 pCi/ml) (CF)] (25)

(in cpm or cps) where:

2.0E-05 pCi/ml = MPC limit for Cs-137 CF Monitor calibration factor - in cpm/ pCi/ml or cps/ pCi/ml 22

AMENDMENT NO. 9 JANUARY 1992 2.9. 1 Standb Service Water SW Monitor

,

The Standby Service Water Monitors (SW). are located on the 522'evel of the Reactor Building.

The meter is located in the main control room on panel P-604.

The flow rate through the monitor is variable, from zero (0) to .two (2) gpm with a normal flow of 1.0-1.5 gpm.

To ensure 10 CFR 20 limits are never exceeded, the alarm setpoint shall be established at 80% or less of the maximum setpoint plus background.

If the setpoint is exceeded, an alarm will activate in the main control room.

The control room operato} can then terminate the discharge and mitigate any uncontrolled release of radioactive material.

2.9.2 Turbine Buildin Service Water TSW Monitor This monitor is located on the 441'evel of the Turbine Building. The readout meter and recorder is located in the main control panel BD-RAD-24.

The flow rate through that monitor is variable, from zero (0) to six (6) gpm with a normal flow of 3-4 gpm.

To ensure 10 CFR 20 limits are never exceeded, the alarm setpoint shall be established at '80% or less of the maximum setpoint plus background.

If the setpoint is exceeded, an alarm will activate in the main control room.

The control room operator can then terminate the discharge and mitigate any uncontrolled release of radioactive material.

23

AMENDMENT NO. 14 AUGUST'1993 2.9.3 Turbine Buildin Sum s Water FO 'onitor There are, three detectors to measure the activity of each of the three nonradioactive sumps. .The monitors are located on the 441'evel of the Turbine Building. The readout meters and re'corder are located in the Radwaste Control Room Panel BD-RAD-41. The alarm/setpoint for these detectors is established by design at 80% of .the 10CFR Part 20, Appendix. B, Table II value for Cs-137. In the event the setpoint is exceeded, the sump discharge will be automatically diverted to the Radwaste system for processing.

Turbine building sumps, Tl, T2, and T3 are normally routed to the liquid radwaste system. Effluent from these turbine building sumps may be routed to the storm water system if analyses indicate no detectable radioactivity is present. Other inputs to the storm waste system, in addition to rain water, include water treatment filter backwashes, Service Building and Emergency Diesel Generator Building floor drains, HVAC air wash units, and condensed steam from plant steam leaks that collect on rooftops during cool weather.

The storm water system terminates in an unlined depression or pond located 1500 feet northeast of the plant. Releases to the storm drain pond are sampled as part of the Radiological Environmental Monitoring Program. Based on past experience, it is expected that there will be some accumulation of low levels of radioactive materials, particularly tritium, in the pond.

24

AMENDMENT. NO. 9

.JANUARY 1992 Table 2-1 (contd;)

Table 2-1 FISM 'BIOACCUMULATION FACTORS (BF;)"'ND ADULT INGESTION DOSE CONVERSION FACTORS DF Dose Conversion Factor (DF;)

Fish Bioaccumulation Nuclide ~F"" I Total Bone ~Th roid Liver GI Tract (pCi/kg per (mRem per pCi Ingested) pCi/liter)

M-3 9.0E-01 6.0E-08 (3) 6.0E-OS 6.0E-OS 6.0E-OS Na-24 1.0E+02 1.7E-06 1.7E-06 1.7E-06 1.7E-06 1.7E-06 P-32 1.0E+05 7.5E-06 1.9E-04 '(3) 1.2E-05 2.2E-05 Cr-51 2.0E+02 2.7E-09 (3) 1.6E-09 (3) 6.7E-07 Mn-54 4.0E+02 8.7E-07 (3) (3) 4.6E-06 1.4E-05 Mn-56 4.0E+02 2.0E-OS (3) (3) 1.2E-07 3.7E-06 Fe-55 1.0E+02 4.4E-07 2.8E-06 (3) 1.9E-06 1.1E-06 Fe-59 1.0E+02 3.9E-06 4.3E-06 (3) 1.0E-05 3.4E-05 Co-58 5.0E+Ol 1.7E-06 (3) (3) 7.5E-07 1.5E-05 Co-60 5. OE+01 4.7E-06 (3) (3) 2.1E-06 4.0E-05 Ni-65 1. OE+02 3.1E-OS 5.3E-07 (3) 6.9E-OS 1.7E-06 CU-64 5. OE+01 3.9E-OS (3) (3) 8.3E-OS 7.1E-06 Zn-65 2.0E+03 7.0E-06 4.8E-06 (3) 1.5E-05 9.7E-06 Zn-69m 2.0E+03 3.7E-08 1.7E-07 (3) 4.1E-07 2.5E-05 As-76 1.0E+02 4.8E-06 (3) (3) (3) 4.4E-05 Br-82 4.2E+02 2.3E-06 (3) (3) (3) 2.6E-06 Br-83 . 4.2E+02 4.0E-OS (3) (3) (3) 5.8E-OS Br-84 4.2E+02 5.2E-OS (3) (3) (3) 4. 1E-13 Rb-89 2.0E+03 2.8E-OS (3) (3) 4.0E-OS 2.3E-21 Sr-89 3.0E+01 S.SE-06 3.1E-04 (3) 4.9E-05 (3)

Sr-90 3.0E+Ol 1.8E-04 8.7E-03 (3) (3) 2.2E-04 Sr-91 3.0E+01 2.3E-07 5.7E-06 (3) 2.7E-05 (3)

Sr-92 3.0E+01 9.3E-OS 2.2E-06 (3) (3) 4.3E-05 Y-90 2.5E+01 2.6E-10 9.7E-09 (3) 1.0E-04 (3)

Y-91m 2. 5E+01 3.5E-12 9.1E-ll (3) (3) 2.7E-10 Y-91 2. 5E+01 3.8E-09 1.4E-07 7.8E-05 (3) (3)

Y-92 2. 5E+01 2.5E-11 8.5E-10 (3) 1.5E-05 (3) 25

'HENDHENT NO. 9 JANUARY 1992 Table 2-1 (contd-.)

Dose Conversion Factor (DF;)

Fish, Nuclide ~FF j Bioaccumulation'otal

~Bod Bone T~h roid Liver GI Tract (pCi/kg per (mRem per pCi Ingested) pCi/1 i ter)

Y-93 2. 5E+01 7.4E-11 2.7E-09 (3) (3) 8.5E-05 Zr-95 3. 3E+00 6.6E-09 3.1E-OS (3) 9.8E-09 3.1E-05 Nb-95 3.0E+04 1.9E-09 6.2E-09 (3) 3.5E-09 2.1E-05 Zr-97 3.3E+00 1.6E-10 1.7E-09 (3) 3.4E-10 1. IE-04 Nb-97 3.0E+04 4.8E-12 5. 2E-11 (3) 1. 3E- 1 1 4,9E-OS Ho-99 1.0E+01 8.2E-07 (3) (3) 4.3E-06 1.0E-05 Tc-99m 1.5E+01 8.9E-09 2.5E-10 (3) 7.0E-10 4.1E-07 Tc-101 1.5E+Ol 3.6E-09 2.5E-10 (3) 3. 7E-10 1.1E-21 Ru-103 1.0E+Ol S.OE-08 1.9E-07 (3) (3) 2.2E-05 RU-105 1.0E+Ol 6.1E-09 1.5E-08 (3) (3) 9.4E-06 Rh-105 1.0E+Ol 5.8E-OS 1.2E-07 (3) 8.9E-OS 1.4E-05 RU-106 1.0E+01 3.5E-07 2.8E-06 (3) (3) 1.8E-04 Ag-110m 2.3E+00 S.SE-OS 1.6E-07 (3) 1.5E-07 6.0E-05 Sb-124 1.0E+00 1.1E-06 2.8E-06 6.8E-09 5.3E-OS S.OE-05 Sb-125 1.0E+00 4.3E-07 1.8E-06 1.8E-09 2.0E-OS 2.0E-05 Sb-126 1.0E+00 4.2E-07 1.2E-06 7.0E-09 2.3E-OS 9.4E-05 Sb-127 1.0E+00 9.9E-OS 2.6E-07 3.1E-09 5.7E-09 5.9E-05 Te-127 4.0E+02 2.4E-OS 1.1E-07 8.2E-OS 4.0E-OS, 8 'E-06 Te-129m 4.0E+02 1.8E-06 1.2E-05 4.0E-06 4.3E-06 5.8E-05 Te-129 4.0E+02 7.7E-09 3.1E-OS 2.4E-OS 1.2E-OS 2.4E-OS Te-131m 4.0E+02 7.1E-07 1.7E-06 1.3E-06 8.5E-07 8.4E-05 Te-131 4.0E+02 6.2E-09 2.0E-08 1.6E-OS 8.2E-09 2.8E-09 Te-132 4.0E+02 1.5E-06 2.5E-06 1.8E-06 1.6E-06 7.7E-05 I-131 1.5E+01 3.4E-06 4.2E-06 2.0E-03 6.0E-06 1.6E-06 I-132 1.5E+01 1.9E-07 2.0E-07 1.9E-05 . 5.4E-07 1.0E-07 I-133 1.5E+Ol 7.5E-07 1.4E-06 3.6E-04 2.5E-06 2 'E-06 I-134 1.5E+01 1.0E-07 1.1E-07 5.0E-06 2.9E-07 2.5E-10 I-135 1.5E+01 4.3E-07 4.4E-07 7.7E-05 1.2E-06 1.3E-06 Cs-134 '.0E+03 1.2E-04 6.2E-05 (3) 1.5E-04 2.6E-06 Cs-136 2 'E+03 1.9E-05 6.5E-06 (3) 2.6E-05 2.9E-06 Cs-137 2.0E+03 7.1E-05 S.OE-05 (3) 1.1E-04 2.1E-06 26

AMENDMENT NO. 9 JANUARY 1992 Table 2-1 (contd.)

j Dose Conversion Factor (DF;)

Nuclide

'ish Bioaccumulation. Total (pCi/kg per pCi/liter)

'hid (mRem

\

Ii per pCi Ingested)

GI Tract Cs-138 2.0E+03 5.4E-OS 5.5E-OS (3) 1.1E-07 4. 7E-13 Ba-139 4.0E+00 2.8E-09 9.7E-08 (3) 6.9E-11 1.7E-07 Ba-140 4.0E+00 1.3E-06 2.0E-05 (3) 2.6E-OS 4.2E-05 La-140 2.5E+01 3.3E-10 2.5E-09 (3) 1.3E-09 9.3E-05 La-141 2.5E+01 1.6E-11 3.2E-10 (3) 9. 9E-11 1.2E-05 La-142 2.5E+01 1.5E-11 1.3E-10 (3) 5.8E-11 4.3E-07 Ce-141 1.0E+00 7. 2E-10 9.4E-09 (3) 6.3E-09 2.4E-05 Ce-143 1.0E+00 1.4E-10 1.7E-09 (3) 1.2E-06 4.6E-05 Ce-144 1.0E+00 2.6E-08 4.9E-07 (3) 2.0E-07 1.7E-04 Pr-143 2.5E+01 4.6E-10 9.2E-09 (3) 3.7E-09 4.0E-05 Nd-147 2.5E+01 4.4E-10 6.2E-09 (3) 7.3E-09 3.5E-05 Hf-179m 3.3E+00 4.8E-06 (3) (3) (3) 4.1E-OS Hf-181 3.3E+00 4.3E-06 (3) (3) (3) 4.1E-05 M-185 1.2E+03 1.4E-OS 4.1E-07 (3) 1.4E-07 1.6E-05 W-187 1.2E+03 3.0E-OS 1.0E-07 (3) 8.6E-OS 2,8E-05 Np-239 1.0E+01 6.5E-11 1.2E-09 (3) 1. 2E-10 2.4E-05

NRC NUREG/CR-4013.

'NRC NUREG/CR-4013.

"'No data listed in NUREG/CR-4013.,

(Use total body dose conversion factor as an approximation.)

27

AMENDMENT NO. 15 OCTOBER 1993 Table, 2-2 INGESTION DOSE FACTORS A, FOR TOTAL BODY AND CRITICAL'ORGAN (in mrem/hr per pCi/ml)

Liquid Effluent Total GI Nuclide ~Bod Bone ~Th roid Liver Tract H-3 1.8E-01 1.8E-01 1.8E-01 1.8E-01 Na-24 4. 1E+02 4. 1E+02 4. 1E+02 4. 1E+02 4. 1E+02 P-32 1.8E+06 4.6E+07 2.9E+06 5. 3E+06 Cr-51 1.3E+00 7.7E-01 3.2E+02 Mn-54 8.3E+02 4.4E+03 1.3E+04 Mn-56 1. 9E+Ol 1.6E+02 3.6E+03 Fe-55 1. 1E+02 6.7E+02 4.6E+02 2.6E+02 Fe-59 9.4E+02 1.0E+03 2.4E+03 8.2E+03 Co-58 2. 1E+02 9.0E+01 1.BE+03 Co-60 5.7E+02 2.5E+02 4.BE+03 Ni-65 7.5E+00 1.3E+02 1.7E+01 4. 1E+02 Cu-64 4.7E+00 1.0E+Ol 8.6E+02 Zn-65 3.4E+04 2.3E+04 7.2E+04 4.7E+04 Zn-69m 1.8E+02 8. 1E+02 2.0E+03 1.2E+05 As-76 1.2E+03 lelE+04 Br-82 2.3E+03 2.6E+03 Br-83 4.0E+Ol 5. BE+01 Br-84 5. 2E+Ol 4.1E-04 Rb-89 1:3E+02 1.9E+02 1.1E-l 1 Sr-89 6.4E+02 2.3E+04 3.6E+03 Sr-90 1.3E+04 6.3E+05 1.6E+04 Sr-91 1.7E+Ol 4.1E+02 2.0E+03 Sr-92 6.BE+00 1.6E+02 3. 1E+03 Y-90 1.6E-02 5.9E-01 6. 1E+03 Y-91m 2.1E-04 5.5E-03 -1.6E-02 Y-91 2.3E-01 8.5E+00 4.7E+03 Y-92 1.5E-03 5.2E-02 9. 1E+02 Y-93 4.5E-03 1.6E-01 5.2E+03 28

AMENDMENT NO. 9

.JANUARY 1992' Table 2-2 (contd.)

'otal GI Nuclide ~Bod Bone ~Th roid Liver 'Tract Zr-95 5.3E-02 ' 2. 5E-01 7.9E-02 2.5E+02 Nb-95 1.4E+02 4.5E+02 2.5E+02 1.5E+06 Zr-97 1.3E-03 1.4E-02 2.7E-03 8.8E+02 Nb-97 3.5E-01 3.7E+00 9.3E-Ol 3.5E+03 Mo-99 2. OE+Ol 1.1E+02 2.5E+02 Tc-99m 3. 3E-01. 9.2E-03 2.6E-02 1.5E+Ol Tc-101 1.3E-01 9.2E-03 1.4E-02 4.0E-14 RU-103 2.0E+00 4,7E+00 5.5E+02 RU-105 1.5E-01 3.7E-Ol 2.3E+02 Rh-105 1.4E+00 3.0E+00 2.2E+00 3.5E+02 Ru-106 8.7E+00 6. 9E+Ol 4.5E+03 Ag-110m 5.6E-01 1.0E-OO 9. 5E-'Ol 3.8E+02 Sb-124 3.6E+00 9.0E+00 2.2E-02 1.7E-Ol 2.6E+02 Sb-125 1.4E+00 5.8E+00 6.5E-02 6.5E+Ol 5.8E-03'.3E-02 Sb-126 1.4E+00 3.9E+00 7.4E-02 3.0E+02 Sb-127 3.2E-01 8. 4E-01 1.0E-02 1.8E-02 1.9E+02 Te-127 2. 3E+Ol 1.1E+02 7.9E+Ol 3.8E+01 8.3E+03 Te-129m 1.7E+03 1.2E+04 3.8E+03 4.1E+03 5.6E+04 Te-129 7.4E+00 3.0E+Ol 2.3E+01 ~

1.2E+01 2.3E+01 Te-131m 6.8E+02 1.6E+03 1.3E+03 8.2E+02 8. 1E+04 Te-131 5.9E+00 1.9E+Ol 1.5E+Ol 7.9E+00 2.7E+00 Te-132 1.4E+03 2.4E+03 1.7E+03 1.5E+03 7.4E-04 I-131 1.3E+02 1.5E+02 7.4E+04 2.2E+02 5.9E+Ol I-132 7.0E+00 7.4E+00 7.0E+02 2. OE+01 3.?E+00 I-133 2.8E+Ol 5.1E+01 1.3E+04 9.2E+Ol 8.1E+01 I-134 3.7E+00 4.0E+00 1.8E+02 1.1E+01 9.2E-03 I-135 1. 6E+O1 1.6E+Ol 2.8E+03 4.4E+01 4.8E+Ol Cs-134 5.8E+05 3.0E+05 7.2E+05 1.3E+04 Cs-136 9. 1E+04 3.1E+04 1.3E+05 1.4E+04, Cs-137 3.4E+05 3.8E+05 5.3E+05 1.0E+04 Cs-138 2.6E+02 2.6E+02 5.3E+02 2.3E-03 29

AMENDMENT NO. 9 JANUARY 1992 Table 2-2 (contd.)

Nucl i'e Total

~Bod Bone ~Th roid . Liver 'ract GI Ba-139 2. 9E'-'02 1. OE-00 7.2E-04 1.BE+00"'.4E+02 Ba-140 1. 4E+01 2. 1E+02 2.7E-01 La-140 2.0E-02 1.5E-01 7.9E-02 5.6E+03 La--141 9.7E-04 1.9E-02 6.0E-03 7.3E+02 La-142 9. 1E-04 7.9E-03 3.5E-03 2.6E+01 Ce-141 2.3E-03 3.0E-02 2.0E-02 7.7E+01 Ce-143 4.5E-04 5.5E-03 3.9E+00 1.5E+02 Ce-144 8.4E-02 1.6E+00 6. SE-01 5.5E+02 Pr-143 2.8E-02 5.6E-01 2.3E-01 2.4E+03 Nd-147 2.7E-02 3.8E-01 4.4E-01 2.1E+03 Hf-.179m . 4.2E+01 3.6E+02 Hf-181 3.BE+01 3.6E+02 M-185 4.0E+01 1.2E+03 4.0E+02 4.6E+04 M-187 8. 6E+Ol 2.9E+02 2.5E+02 8. 1E+04 Np-239 1.6E-03 3.0E-02 3.0E-03 6. OE+02

- No Ingestion Dose Factor (DF;) is listed in NUREG/CR-4013. (Total body dose factor value will be used as an approximation.)

30

AMENDMENT NO. 11 AUGUST 1992 TABLE 2-3 INPUT PARAMETERS USED TO CALCULATE MAXIMUM INDIVIDUAL DOSE FROM LI UID EFFLUENTS Drinkin Water River Dilution: 50,000 River Transit Time: 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Usage Factors: Adult -'730 1/yr Teenager 510 1/yr Child 510 1/yr Infant - 330 1/yr Boatin and A uatic Food River Dilution: 500 Transit Time: 2 hours Usage Factors: (Aquatic Food) Adult = 21 kg/yr Teenager = 16 kg/yr Child = 6.9 kg/yr Infant = 0 (Boating) Adult = 100 hr/yr Teenager 100 hr/yr Child = 85 hr/yr Infant - 0 Recreation River Dilution: 20,000 Shoreline Width Factor: 0.2 Usage Factors: Shoreline Activities: Adult 90 hr/yr Teenager - 500 hr/yr Child 105 hr/yr Infant = 0 Swimming: Adult = 18 hr/yr Teenager = 100 hr/yr Child = 21 hr/yr Irri ated Foodstuffs River Dilution: 50,000 River Transit Time: 4 hours Leafy

~Ve etabl es Milk Meat ~Ve etab'les Food Delivery Time: 14 days 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

20 days' 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Usage Factors:

Adult 520 kg/yr 310 1/yr 110 kg/yr 64 kg/yr Teenager 630 kg/yr 400 1/yr 65 kg/yr 42 kg/yr Child 520 kg/gr 330 1/yr 41 kg/yr 26 kg/yr Monthly Irrigation Rate: 180 1/m 200 1/m 160 1/m 200 Annual Yield: 5.0 kg/m'0 1.3 1/m'0 2.0 kg/m~

1/m'.5 kg/m Annual Growing Period: days days 130 days 70 days Annual 50-Mile Production: 3.5E+09 kg 2.8E+08 L 2.3E+07 kg 1.9E+06 kg 31

AMENDMENT NO. 9 JANUARY 1992

-SUHPS COOLING TOW=R Radwaste Bldg Waste Surge Waste" Sample BLOWDOWN LINE Turbine Bldg and Collector, Tanks (TWO)

Orgwell Tanks Floor Drain Floor Drain Collector Sample MISC WASTE Tank Tank Reactor Bldg Distillate Deter gent Tanks Orain Tanks (TWO)

Filters and Oemineralizers (Solid Waste)

Chemical Waste Tanks Shop Oecon Condensate Plant Chem Pumps Stor age Use Oecon Drain Tanks (TWO)

Reactor Bldg ~

COLUNBIA Turbine Bldg RIVER SINPLIFIEO BLOCK DIAGRAM OF LIQUID WASTE SYSTEN Figure 2-1 32

AMENDMENT NO.. 9 JANUARY 1992 MN R Condensate Phase Dewatering Disposal RWCU Separator Liners Site EDR/FDR Radwaste Bead Qrg Active Compactor Disposal Waste Site SIMPLIFIED BLOCK DIAGRAM OF SOLID RADWASTE SYSTEM Figure 2-2 33

AHENDHENT NO. 11 AUGUST 1992 3.0 GASEOUS EFFLUENTS DOSE CALCULATIONS

,The U.S. Nuclear Regulatory Commission's computer program GASPAR II can be used to perform environmental dose analyses for releases of radioactive efflu-ents from WNP-2 into the atmosphere. The analyses estimate radiation dose to individuals and population groups from inhalation, ingestion (terrestrial foods), and external exposure (ground and plume) pathways. The calculated doses provide information for determining compliance with Appendix I of 10 CFR Part 50. This computer code has the subroutine "PARTS" which can be used for calculating dose factors.

The NRC computer program GASPAR II supplements the ODCH in monthly, quarterly and annual dose equivalent determinations from gaseous effluents. The method which is normally employed to calculate the annual dose to the maximally exposed organ sums the dose to the maximally exposed organ for each quarter.

As a result, the maximum annual organ dose may not represent the maximum dose to any one particular organ for that particular year. Actual specific organ doses will be less than or equal to this calculated value.

Both the ODCH equations and the NRC GASPAR II computer program for estimating the highest dose to any organ for a particular age group provides conservatism in calculating maximum organ doses. This conservatism is recognized and is intentional.

3. 1 Introduction WNP-2 gaseous effluents are released on a continuous basis; in addition, batch releases also occur when containment and mechanical vacuum pump purges are performed and when the off-gas treatment system operates in the charcoal bypass mode. The gaseous effluents released from WNP-'2 'will meet Requirement for Operability at the site boundary.

34

AMENDMENT NO.'ll AUGUST 1992 Figure 3-' delineates the WNP-2 Site boundary, which for dose calculation .pur-poses, is considered circular with, a radius of 1.2 miles. There are several low occupancy unrestricted locations within the site boundary. These loca-tions, with the exception of the WNP-2 visitor center, are not continuously controlled by the Supply System. The locations are:

1. Wye burial site - normally controlled by DOE.

2. DOE train - two railroad lines pass through the site (approximately 3 miles of line). According to DOE, the train makes one round trip a day, through the site at an average speed of 20 mph, 5 days a week, 52 weeks/year.

3. BPA Ashe Substation - occupied 2080 hour0.0241 days 0.578 hours 0.00344 weeks 7.9144e-4 months s/year. These people are

. not normally controlled by the Supply System but are involved in activities directly in support of WNP-2.

34a

AHENDHENT NO. '6 DECEHBER 1993

4. WNP-'2 - Supply System Visitor. Center - .assumed occupied 8 hrs/yr

, by non-Supply System individuals.'.

WNP, occupied 2080 hrs/yr. This location is controlled by the Supply System. However, activities are not in direct support of WNP-2.

6. WNP-4 - occupied 2080 hrs/yr. This location is controlled by the Supply System. However, activities are not in direct support of WNP-2.

All other locations listed in Figure 3-1 support WNP-2 activities and are controlled by the Supply System. Figure 3-2 provides a simplified block diagram of the gaseous radwaste system for the reactor, turbine a'nd radwaste buildings. Figure 3-3 provides a simplified block diagram for the off-gas treatment system.

The Auxiliary Boiler supplies heating steam to the Reactor, Radwaste, Turbine and Service buildings when Seal Steam Evaporator B is not in operation. The Auxiliary Boiler and associated heating steam system vents to the atmosphere and provides a possible unmonitored source of radioactive effluent when in operation. Samples have shown 2.0 E+06 picocuries per liter of tritium activity to be present within the Auxiliary Boiler system. Using NRC Regulatory Guide 1.109 methodology with FSAR Low Population Zone (LPZ) X/Q values and assuming one gallon per minute (1 gpm) makeup flowrate for 180 days plus a one time complete boil-off of the total water inventory, the dose contribution from tritium would be less than one tenth of a millirem per year

(<0. 1 mrem/yr). Figure 3-4 provides a simplified diagram for the Auxiliary Boiler.

Air doses and doses to individuals at these locations were calculated based on the NRC GALE code design base mixture, location specific estimated occupancy, and X/Qs from XOQDOQ. (Note: Desert Sigmas were used in calculating X/Q and D/Q values, and are listed in Table 3-10 and 3-11). These doses are listed in Tables 3-16 and 3-17 along with the doses to the maximum exposed individual.

35

'MENDMENT NO. 16 DECEMBER 1993 The most likel ex osed member of the ubl'ic is considered to be residing in Taylor Flats (4.2 miles ESE of WNP-2). This is the closest residential area with the highest X/g and D/g values.

3.2 Gaseous Effluent Radiation Monitorin S stem 3.2.1 Hain Plant Release oint The Hain Plant Release is instrument monitored for gaseous radioactivity prior to discharge to the environment via the main plant vent release point.

Particulates and iodine activity are accumulated in filters which will be changed and analyzed as per Periodic Test and Inspection 6.2.2.1.2 (4. 11.2. 1.2) and Table 6.2.2. 1.2-1 (4. 11-2). The effluent is supplied from:

the gland seal 35a

AMENDMENT NO. 13 AUGUST 1993 exhauster, mechanical vacuum pumps, treated off 'gas; standby gas treatment,

'and exhaust air from the entire reactor'uilding's ventilation.

Two 100-percent capacity vanaxial fans supply 80,000 CFH ventilation air. One is normally operating, the -other is in standby. The radiation monitors are located on the ventilation exhaust plenum.

Effluent monitoring consists of a gamma spectroscopy system utilizing three in-line detectors to provide an isotopic analysis of the Elevated Release effluents. The low range (PRH-RE-1A) is a high efficiency, cryogenically cooled, high purity germanium detector located inside the duct at elevation 611'o monitor low level normal operation radioactivity. Low range response is 8.24 x 10'ps/yCi/cc. Two additional'detectors (PRH-RE-1B and PRH-RE-1C) are mounted in lead enclosures at elevation 618'7" for post-accident monitoring. They provide a range of 10 to 10 pCi/cc with one decade of overlap. All three have gross gamma Log Count Rate Meter ranges of 10 to 10'ps. PRH-LCRM-IA, -1B, and 1C are located on Radwaste Bldg.

elevation 525'n PRH-CP-1 and are recorded at PRH-RR-3 on BD-RAD-24 in the Hain Control Room. Isotopic information from all three detectors is available at E-CP-H13/P814 on PRH-COMP-3. Power is from battery-backed, reliable 120 VAC buses. This monitor has no control function but annunciates in the Hain Control Room. The alarm will initiate proper action as defined in the WNP-2 plant procedures.

3.2.2 Radwaste Buildin Ventilation Exhaust Monitor The radwaste building ventilation exhaust monitoring system monitors the radioactivity in the exhaust air prior to discharge. Radioactivity can originate from: radwaste tank vents, laboratory hoods, and various cubicles housing liquid process treatment equipment and systems.

The radwaste building exhaust system has three 50-'percent capacity exhaust filter units of 42,000 cfm capacity. Each exhaust unit has a medium-efficiency prefilter, a high efficiency particulate air filter (HEPA) and two centrifugal fans. Total exhaust flow will vary as the combined exhaust unit maintains a r adwaste building differential pressure of -0.25 inches H,O to the environment.

Particulate and iodine air sample filters are changed weekly for laboratory 36

AMENDMENT NO. 9 JANUARY 1992 V

analysis. After the particulate and iodine filters, the air sample streams

~ ~

are combined. in a manifold prior to being. monitored by a beta scintillator.

~

The beta scinti-llators, on the 487'evel are mounted in lead shielded chambers. 'he low'ange beta scintillator has an approximate response of 80 cpm/pCi/cc to Kr-85, and 50 cpm/pCi/cc to Xe-133 and a meter range of 10 - 10 cpm., The intermediate range has a response'from 10

- - 10'Ci/cc Xe-133 equivalent, and reads in panel meter units (PMU) with a meter range of 10 - 10'MU. The readouts and recorder are located in the main control room panel BD-RAD-24. Power is provided from 125 VDC divisional buses. This .

monitor has no control functions but annunciates in the main control room.

The alarm will initiate proper action as defined in the WNP-2 plant procedures.

3.2.3 Turbine Buildin Ventilation Exhaust Monitor This monitoring system detects fission and the activation products from the turbine building air which may be present due to leaks from the turbine and

~

other primary components in the building.

~

The turbine building main exhaust system consists of four roof-mounted cen-trifugal fans which draw air from a central exhaust plenum. Three fans operate continuously, with one in standby to provide a flow of 260,000 cfm.

A representative sample is extracted from the exhaust vent and passed through a particulate and charcoal filter. The air sample then passes to a beta scintillator.

The beta scintillators are mounted in lead shielded chambers. The low range beta scintillator has an approximate response of 80 cpm/pCi/cc to Kr-85, and 50 cpm/pCi/cc to Xe-133 and a meter range of 10 - 10 cpm. The intermediate range has a response from 10' 10'Ci/cc Xe-133 equivalent, and reads in panel meter units (PMU) with a meter range of 10 - 10'MU. The monitors are on the 525'evel of the radwaste building and the readouts and the recorder are .located in the main control room panel BD-RAD-24. Power is provided from 37

AHfNDHfNT NO. 9 JANUARY 1992 the 125 VDC divisional buses. This monitor has no control functions but annunciates, in the main control room. The alarm .will initiate proper .action r

as defined in the WNP-2 plant procedures.

3.3 10 CFR 20 Release Rate Limits Limits for release of gaseous effluents from the site to areas at and beyond the site boundary are stated in Requirement for Operability 6.2.2. 1 (3.11.2. I). The dose rate at these areas due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

(a) "The dose rate limit for noble gases shall be g500 mrem/yr to the total body and <3000 mrem/yr to the skin."

(b) "The dose rate limit for all radioiodines and for all radio-active materials in particulate form and radionuclides other than noble gases with half-lives greater than eight days shall be Z1500 mrem/yr to any organ."

3.3.1 Noble Gases In order to comply with Requirement for Operability 6.2.2. 1, (3. 11.2. I) the following equations must hold:

Whole body:

Ki [(~XQ) Q. + (~XQ) Q. )] (500 mrem/yr Skin:

,+

[(L. l.llL) ((X/Q) Q., + (7X Q), Q,,)] 5 3000 mrem/yr (2) 38

AMENDMENT NO. 9 JANUARV 1992 3.3.2 Radioiodines and Particulates Part "b" of Requirement for Operability 6.2.2. 1 (3. 11.2. 1) requires that the release rate limit for all radioiodines and radi'oactive materials in particulate form and radionuclides other than noble gases must meet the following relationship:

Any organ:

Z P [W 0 + W 0,.] <1500 mrem/yr (3)

I The term's used in Equations (1) through (3) are defined as follows:

K; The total body dose factor due to gamma emissions for each identified noble gas radionuclide i (mrem/yr per pCi/m').

The skin dose factor due to beta emissions for each identified noble gas radionuclide i (mrem/yr per pCi/m').

M; The air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad/yr per pCi/m'unit conversion constant of 1. 1 mrem/mrad converts air dose to skin dose).

p; The dose parameter for all radionuclides other than noble gases for the inhalation pathway, (mrem/yr per pCi/m') and for food and ground plane pathways, m'(mrem/yr per pCi/sec).

The dose factors are based on the critical individual organ and the most restrictive age group.

Q; The release rate of radionuclide i in gaseous effluent from mixed mode release. The main plant release point is a partially elevated mixed mode release (pCi/sec).

39

AMENDMENT NO. 9 JANUARY 1992

~ I Q;, The release rate of radionuclide i in gaseous effluent from all ground level'eleases (pCi/sec).

C (sec/m')'. 'For partially elevated mixed mode releases, from the ma'in plant vent release point. The highest calculated partially elevated annual average relative concentration for any area at and beyond the site boundary.

(sec/m ). For all Turbine Building and Radwaste releases.

The highest calculated ground level annual average relative concentration for any area at and beyond the site boundary.

Wg The highest calculated annual average dispersion parameter for estimating the dose to an individual at the controlling location due to all ground level releases.

W 9 (sec/m ). For the inhalation pathway. The location is at and beyond the site boundary in the sector of maximum concentration.

W 0

m'. For ground plane pathways. The location is at and beyond the site boundary in the sector of maximum concentration.

WM The highest calculated annual average dispersion parameter for estimating the dose to an individual at the controlling location due to partially elevated releases:

WM sec/m'. For inhalation pathway. The location is at and beyond the site boundary in the sector of maximum concentration.

WM m'. For ground plane pathways. The location is at and beyond the site boundary in the sector of maximum concentration.

40

AMENDMENT NO. 9 JANUARY 1992

"'he factors, L; and H;, relate"the radionucli'de airborne concentrations to

~

various ose ra es assuming a semi-infinite cloud. These factors are listed

~

in Table B-I of Regulatory Guide 1. 109, Revision 1, and in Table 3-1 of this'anual.

The values used in the equations for the implementation of Requirement for Operability 6.2.2. 1 (3. 11.2. 1) are based upon the maximum long-term annual average X/g at and beyond the site boundary. Atmospheric dispersion factors will be evaluated annually from the WNP-2 meteorological data base and if significantly different than preoperational data as is displayed in Tables 3-10 and 3-11, then the tables will be updated. This comparison began with 1989 data. Table 3-2 provides typical locations based on the current Land Use Census (LUC) with pathways for use in dose determinations. Table 3-3 provides these typical locations with long term X/g and D/g values which may be used if current annual averages are not available.

The X/g and D/g values listed in Tables 3-10 and 3-11 reflecting correctly acquired meteorological data, January 1, 1984 - January 1, 1990 may be utilized in GASPAR II Computer runs.

3.3 '. 1 Dose Parameter for Radionuclide i (P,)

The dose parameters used in Equation (3) are based on:

l. Inhalation and ground plane. (Note: Food pathway is not applicable to WNP-2 since no food is grown at or near the restricted area boundary.)

2. The annual average continuous release meteorology at, the .site boundary.

3. The critical organ for each radionuclide (thyroid for radioiodine).

4. The most restrictive age group.

Calculation of I P. ( Inhalation):

1 The following equation will be used to cal-

AMENDMENT NO. 9 JANUARY 1992 culate I (Inhalation).

P.

1 I

P. ( Inhalation) = K"(BR) DFA,. (mrem/yr per Ci/m') (5) where:

K" A constant of conversion, 10'pCi/Ci.

BR The breathing rate of the child age group, 3700 m'/yr.

DFA; The critical organ inhalation dose factor for the child age group for the ith radionuclide in mrem/pCi. The total body is considered as an organ in the selection of DFA;.

The inhalation dose factor for DFA; for the child age group is listed in Table E-9 of Regulatory Guide 1. 109, Revision 1, and Table 3-4 of this manual.

Resolving the units yields:

I P. = (Inhalation) = (3.7 x 10 ) (DFA,.) (mrem/yr per pCi/m') (6)

The P.

I (Inhalation) values for the child age group are tabulated in Table 3-4 1

of this manual.

3.4 10 CFR 50 Release Rate Limits The requirements pertaining to 10 CFR 50 release rate limits are specified in Requirement for Operability 6.2.2.2 (3. 11.2.2) and 6.2.2.3 (3. 11.2.3).

Requirement for Operability 6.2.2.2 (3.11.2.2) deals with the air dose from gases and requires that the air dose at and beyond the site boundary due 'oble to noble gases released in gaseous effluents shall be limited to the

AHENDHENT NO. 9 JANUARY 1992

'following:

(a) "During any calendar quarter, to <5 mrad for gamma radiation and to gl0 mrad for beta radiation."

C (b) "During any calendar year, to <10 mrad for gamma radiation and <20 mrad for beta radiation."

Requirement for Operability 6.2.2.3 (3. 11.2.3) deals with radioiodines, tritium, and radioactive materials in particulate form, and requires that the dose to an individual from radioiodines, tritium and radioactive materials in particulate form with half-lives greater than eight days in gaseous effluents released to unrestricted areas shall be limited to the following:

(a) "During any calendar quarter, to <7.5 mrem."

(b) "During any calendar year, to <15 mrem."

3.4.1

~ ~ Noble Gases Re uirement for 0 erabilit 6.2.2.2 3.11.2.2 The air dose at and beyond the site boundary due to noble gases released in the gaseous effluent will be determined by using the following equations.

a. During any calendar quarter, for gamma radiation:

3. 17 x 10

' [II, (X/0),0,, + (X/q),q,, + (X/0) 0, + (X/q) q,. <5 mrad (8)

I ]

During any calendar quarter, for beta radiation:

a 3.17 x 10 g .[(X/0),0,,,(+ (X/q),q.+ + (X/q) (9)

I N, (7X00) 0., q,.

] <10 mrad 43

AMENDMENT NO. 9 JANUARY 1992

b. ,During any calen'dar year, for gamma'adiation:

a 3.17 x 10 a g N,.[(~X0),0,. + (X/q) q,

+ (X/0) 0,. + (X/q) q,] <10 mrad (10)

I During any calendar year, for beta radiation:

3.17 x 10' N,. [(~X0),0.+ (X/q},q. + (~Xq) 0(

+'(X/q} q, ] <20 mrad (11) i where:

M; The air dose factor due to gamma emissions for each identified noble gas radionuclide, in mrad/yr per pCi/m'M,.

values are listed in Table 3-1).

The air dose factor due to beta emissions for each identified noble gas radionuclide, in mrad/yr per pCi/m'N,.

values are listed in Table 3-1).

For ground level release points. The highest calculated annual average relative concentration for area at and beyond the site area boundary for long-term releases (greater than 500 hr/yr). (Sec/m')

(x/q), For ground level release points. The relative concentration for areas at and beyond the site area boundary for short-term releases (equal to or less than 500 hr/yr).

(Sec/m')

(WQ) For partially elevated release points. The highest 44

AMENDMENT NO. 9 JANUARY 1992 calculated annual average relative concentration for areas at and beyond the site boundary for, long-term releases (greater than 500. hr/yr). (Sec/m')

(x/q) For partially elevated release points. The relative concentration for areas at and beyond the site boundary for short-term releases (equal to or less than 500 hr/yr).

(Sec/m )

%m The average release of noble gas radionuclides in gaseous effluents, i, for short-term releases (equal to or less than 500 hr/yr) from the main plant release point, in pCi.

Releases shall be cumulative over the calendar quarter or'ear, as appropriate.

The average release of noble gas radionuclides in gaseous effluents, i, for short-term releases (equal to or less than 500 hr/yr) from Radwaste and Turbine Building, in pCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

The average release of noble gas radionuclides in gaseous releases, i, for long-term releases (greater than 500 hr/yr) from the main plant release point,- in pCi. Release shall be cumulative over the calendar quarter or year, as appropriate.

The average release of noble gas'adionuclides 'in gaseous effluents, i, for long-term releases (greater than 500 hr/yr) from Radwaste and Turbine Building, in pCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

3. 17 x 10 = The inverse of the number of seconds in a year.

3.4.2 Radioiodines Tritium and Particulates Re uirement for Q erabilit 45

AMENDMENT NO. 9 JANUARY 1992, 6.2.2.3 '.11.2.3

,The following 'equation calculates the dose to an individual from radioiodines, tritium, and radioactive 'material in particulate form with half-lives. greater than eight days'n'aseous effluents released to the unrestricted ar'eas:

a. During any calendar quarter:

3 . 17 x 10

' R,.

[ M 0,. + w q,. + M,q,, + w,qwr

] 57 .5 mrem (12)

I

b. During any calendar year:

3.17x10'g R,.[W0,.+wq,. +W,q,,+w,q. ]515mrem (13)

I where:

Q;IQ;, The releases of radionuclides, radioactive materials in particulate form, and radionuclides other than noble gases in gaseous effluents, i, for long-term releases greater than 500 hr/yr, in pCi. Releases shall be cumulative over the calendar quarter or year, as appropriate (m's for mixed mode releases, g is for ground level releases).

qImI qtp The releases of radionuclides, radioactive materials in particulate form, and radionucl.ides other, than noble gases in gaseous effluents, i, for short-term releases equal to or less than 500 hr/yr, in pCi. Releases shall be cumulative over the calendar quarter or year as appropriate (m is for mixed mode releases, g is for ground level releases).

W,W The dispersion parameter for estimating the dose to an

0 AMENDMENT NO. 15 OCTOBER 1993 individual at the controlling location for long-term (greater than 500 hr.) releases (m is for mixed mode

-releases; g is for ground level releases).

(~X g) for the inhalation pathway, in sec/m'.

pwca W, - (~D g) for the food and ground plane pathways in meters .

Wm The dispersion parameter for estimating the dose to an individual at the controlling location for short-term (less than 500 hr.) releases (m is for mixed mode releases, g is for ground level releases).

Wm (~X q) for the inhalation pathway, in sec/m'.

(~D q) for the food and ground plane pathways in meters ~.

3.17 x 10 The inverse of the number of seconds in a year.

R; The dose factor for each identified radionuclide, i, in m (mrem/yr per pCi/sec) or mrem/yr per pCi/m ."

AMENDMENT-NO. 9 JANUARY 1992 5

3.4'.2. 1'ose'Parameter for Radionuclide i (R;)

h The R,. values used in Equations (12) and (13) of. this section are calculated separately for each of the following potential exposure pathways:

Inhalation Ground plane contamination Grass-cow/goat-milk pathway Grass-cow-meat pathway Vegetation pathway Monthly dose assessments for WNP-2 gaseous effluent will be'done for all age groups.

Calculation of I R. (Inhalation Pathway Factor)

R. (Inhalation) = K'BR) (DFA.), (mrem/yr per pCi/m') (14) where:

R 1

The inhalation pathway factor (mrem/yr per pCi/m ).

K' constant of unit conversion, 10'Ci/pCi.

(BR), The breathing rate of the receptor of age group (a) in meter'/yr. (Infant - 1400, child - 3,700, teen = 8,000, adult = 8,000. From P.32 NUREG-0133).

48

AMENDMENT NO, 9 JANUARY 1992

'(DFA;), The maximum organ inhalation dose factor for receptor of age "

group'a for the ith radionuclide (mrem/pCi). The -total body is considered as an organ in the selection of (DFA,.),.

(DFA;). values are listed in Tables E-7 through E-10 of Regulatory Guide 1. 109 manual, Revision 1 and NUREG/CR-4013.

I Values of R. are listed in Table 3-5.

1 G

Calculation of R. (Ground Plane Pathway Factor) 1 G

R. (Ground Plane) = K"K (SF)(DFG,.) (1-e "')/A,. (m x mrem/yr per pCi/sec) (15) where:

G R. = Ground plane pathway factor 2 (m x mrem/yr per pCi/sec).

K* A conversion constant of (10'Ci/pCi).

K A conversion constant - (8760 hr/yr).

The decay constant for the ith radionuclide (sec').

Exposure time, 6.31 x 10 sec (20 years).

DFG; The ground plane dose conversion factor for the ith radionuclide, as listed in Table E-6 of Regulatory Guide 1. 109, Revision 1 and NUREG/CR-4013 (mrem/hr per pCi/m ).

SF Shielding Factor (dimensionless)--0.7 if building is present, as suggested in Table E-15 of Regulatory Guide 1. 109, Revision 1.

G The values of R. are listed in Table 3-5 of this manual.

49

AMENDMENT NO. 9 JANUARY 1992 C

Calculation of R.- (Grass-Cow/Goat-Hilk Pathway Factor)

C R. (Grass-Cow/Goat-Milk'Factor) =

1

~~

K' "

Q(Usp)

F(r) (DFL,.).

f,f, (1-f,f,)e e

-A,t, X,. +A Y Y, (m' mrem/yr per pCi/sec) where:

K' constant of unit conversion, 10'Ci/pCi.

Q The -cow/goat consumption rate, in kg/day (wet weight).

U,p The receptor's milk consumption rate for age a, in liters/yr.

Yp The agricultural productivity by unit area of pasture feed grass, in kg/m'.

Y, The agricultural productivity by unit area of stored feed, in kg/m .

The stable element transfer coefficients, in days/liter.

Fraction of deposited activity retained on feed grass.

(DFL;), The maximum organ ingestion dose factor for the ith radionuclide for the receptor in age group a, in mrem/pCi (Tables E-11 to E-14 of Regulatory Guide 1. 109, Revision 1 and NUREG/CR-4013).

The decay constant for the ith radionuclide, in sec' 50

AMENDMENT NO.

1992 11'UGUST The decay constant'or removal of activity on leaf andplant surfaces by weathering, 5.73.x 10 sec'corresponding to a day half-life).

The transport time from pasture to animal, to milk, to receptor, in sec.

The transport time from pasture, to harvest, to animal, to milk, to receptor, in sec.

Fraction of the year that the cow/goat is on pasture (dimensionless).

Fraction of the cow/goat feed that is pasture grass while the cow is on pasture (dimensionless).

NOTE: For radioiodines, multiply R. value by 0.5 to account 1

for the fraction of elemental iodine available for deposition.

The input parameters used for calculating R.C are listed in Table 3-6. The i

individual pathway dose parameters for R. are tabulated in Tables 3-5a through 1

3-5d.

For Tritium:

In calculating RTC pertaining to tritium in milk, the airborne concentration rather than the deposition will be used:

C RT (Grass-Cow/Goat-Milk Factor) =

K"K'F gU (DFL,.), [0.75(0.5/H)] (mrem/yr per pCi/m') (17) where:

KA A constant unit conversion, 10 pCi/pCi.

51

AMENDMENT NO. 11

'UGUST 1992

'A constant of unit conversion, 10 gm/kg.,'bsolute humidity of the atmosphere, in gm/m'.

0.75 The fraction of total feed that is water.

0.5 The ratio of the specific activity of the feed grass water to the atmospheric water.

H Calculation of R. (Grass-Cow-Heat Pathway Factor) 1 H

R. (Grass-Cow-Meat 1

Factor)

K'p

, ~i(U.)

I W Ff(r) (DFL,.),

f,f, Y

(1-ff,)e Y,

"~

(18)

(m' mrem/yr per pCi/sec) where:

K', A constant unit conversion, 10 pCi/pCi.

The stable element transfer coefficients, in days/kg.

U, The receptor's meat consumption rate for age a, in kg/yr.

t, The transport time from pasture to receptor, in sec.

The transport time from crop field to receptor, in sec.

H NOTE: For radioiodines, multiply R. value by 0.5 to account 1

for the fraction of elemental iodine available for deposition.

The input parameters used for calculation M R. (18) are listed in Table 3-7.

H The individual pathway dose parameters for R. are tabulated in Tables 3-5a 1

through 3-5d.

52

AMENDMENT NO. 9 JANUARY 1992 For Tritium:

M In calculating the RT for tritium in meat, the airborne concentration is used rather than the deposition rate, The following equati,on is used to calculate I

the RTM values for tritium:

M R

T (Grass-Cow-Meat Pathway)-

K"K 0.75(0.5/H) ] (mrem/yr per pCi/m') (19)

[ F,DU (DFL.,(). ] [

Where the terms are as defined in Equations (16) through (18), R.M values for 1

tritium pertaining to the infant age group is zero since there is no meat consumption by this age group.

V Calculation of i (Vegetation Pathway Factor)

R.

V R.

1 (Vegetation Pathway Factor) =

(DFL.,)(. IU"fle "e + U. fe "~ (20) y (y y )

I 8 p ee L 0 p ]

(m' mrem/yr per pCi/sec) where:

K A constant of unit conversion, 10'pCi/pCi.

U a

The consumption rate of fresh leafy vegetation by the receptor in age group a, in kg/yr.

s The consumption rate of stored vegetation a by the receptor in age group a, in kg/yr.

53

AMENDMENT NO. 11 AUGUST 1992

'he fraction of the annual intake of fresh leafy vegetation grown locally.

The fraction of the annual intake of stored vegetation grown locally.

The average time between harvest of leafy vegetation and its consumption, in seconds.

The average time between harvest of stored vegetation and its consumption, in seconds.

Y The vegetation area density, in kg/m .

NOTE: For radioiodines, multiply R value by 0.5 to account for the fraction of elemental iodine available for deposition.

V The input parameters for calculation R.1 are listed in Table 3-8. The individual pathway dose parameters for R.V1 are tabulated in Tables 3-Sa through 3-5d.

All other items are as defined in Equations (16) through (18).

For Tritium:

In calculating the RT for tritium, the concentration of tritium in vegetation is based on airborne concentration rather than the deposition rate. The following equation is used to calculate R for tritium:

V R (Vegetation Pathway Factor)

K"K [(U,"f+ U:f ) (DFL,),] [0.75(0.5/H)] (mrem/yr per pCi/m ) (21) 54

AMENDMENT NO. 11 AUGUST 1992 Where'all terms have been defined above and in Equations (16) through (18),

-V the RT value for tritium is zero for the infant,.age group due to zero vegetation consumption rate by that age group'. The input'arameters needed for solving Equations (20) and (21) are listed in Table 3-8.

3.4.3 Annual Doses At S ecial Locations The Radioactive Effluent Release Report submitted within 60 days after January 1 of each year shall include an assessment of the radiation doses from radioactive gaseous effluents to "Members of the Public," due to their activities inside the site boundary during the report period.

Annual doses within the site boundary have been determined for several locations using the NRC GASPAR computer code and source term data from Table 11.3-7 of .the FSAR. These values are listed in Tables 3-16 and 3-17. Of the locations listed within the site boundary, only two, the DOE Train and WNP-2 Visitor Center are considered as being occupied by a "Member of the Public."

Annual doses to the maximum exposed "Member of the Public" shall be determined for an individual at the WNP-2 Visitor Center based on occupancy of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per year due to it being the higher of the two locations.

3.5 Com liance with Re uirement for 0 erabilit 6.2.2.4 3. 11.2.4 Requirement for Operability 6.2.2.4 (3. 11.2.4) states:

"The GASEOUS RADWASTE TREATMENT SYSTEM shall be in operation in either the normal or charcoal bypass mode. The charcoal bypass mode shall not be used unless the offgas post-treatment radiation monitor is OPERABLE as specified in Table 6. 1.2. 1-1 (3.3.7. 11-1)."

"RELEVANT CONDITIONS: Whenever the main condenser steam jet air ejector (evacuation) system is in operation."

Prior to placing the gaseous radwaste treatment system in the charcoal bypass mode, the alarm setpoints on the main plant vent release monitor shall be set to account for the increased percentages of short-lived noble gases. Noble gas percentages shall be based either on actual measured values or on primary 55

AMENDMENT NO. 15 OCTOBER 1993 coolant design'ase noble gas concentration percentages adjusted for 30-minute decay. Table 3-15

-

lists the percentage values for 30-minute decay.

3.5. 1 Pro 'ection of Doses The projected doses due to WNP-2 gaseous effluent releases will be determined at least once per 31 days as stated in Requirement for Operability 6.2.2.5 (3. 11.2.5). The projected dose when averaged over 31 days is not to exceed 0.3 mrem to any organ in a 31 day period to areas at and beyond the site boundary. Dose projection values will be determined by using a previous 31 day "Gaspar Output" (NRC Computer Code) for the site boundary and/or an area beyond the site boundary. Based on operating data, the projected dose should be adjusted accordingly to compensate for those anticipated changes in operations and/or source term values.

3.6 Calculation of Gaseous Effluent Monitor Alarm Set pints 3.6. 1 Introduction The following procedure is used to ensure that the dose rate in the unrestricted areas due to noble gases in the WNP-2 gaseous effluent do not exceed 500 mrem/yr to the whole body or 3000 mrem/yr to the skin. The initial setpoints determination was calculated using a conservative radionuclide mix obtained from the WNP-2 GALE code. While the plant is operating and sufficient measurable process fission gases are in the effluent, then the actual radionuclide mix will be used to calculate the alarm setpoint.

3.6.2 Set oint Determination for all Gaseous Release- Paths The setpoints for gaseous effluent are based on instantaneous noble gas dose rates. Sampling and analysis of radioiodines and radionuclides in particulate form will be performed in accordance with Requirement for Operability to ensure compliance with 10 CFR 20'nd 10 CFR 50 App'endix I limits. The three release points will be partitioned such that their sum does not exceed 100 percent of the limit. Originally, the setpoints will be set at 40 percent for 56

AHENDHENT NO.

1992 11'UGUST the Reactor Building, 40 percent for the Turbine Building and 20 percent for

~

,the Radwaste Building. ,These percentages, could vary at the plant discretion,

~

should the operational conditions, warrant such change. However, the combined

. releases due to variations in the setpoints will not result in doses which exceed the limit stated in Requirement for Operability. Both skin dose and whole body setpoints will be calculated and the lower limit will be used.

3.6.2. 1 Set pints Calculations Based on Whole Bod Dose Limits The fraction (m,) of the total gaseous radioactivity in each gaseous effluent release path j for each noble gas radionuclide i will be determined by using the following equation:

x,,

= (dimensionless)

H,,"

(22)

Hr, where:

The measured individual concentration of radionuclide i in the gaseous effluent release path (pCi/cc). j Hr, The measured total concentration of all noble gases identified in the gaseous effluent release path j (pCi/cc).

Based on Requirement for Operability 6.2.2. 1 (3. 11.2. 1), the maximum acceptable release rate of all noble gases in the gaseous effluent release path j is calculated by using the following equation:

F,. 500 (pCi%sec)

(23) x/g,. g (K,.) (m,,)

57

,AMENDMENT NO. 11

'AUGUST 1992 Qu'he where:

maximum acceptable release rate (pCi/sec) of all noble gases in the gaseous effluent release pa'th j (pCi/cc).

F; Fraction of total dose allocated to release path j.

500 Whole body dose rate limit of 500 mrem/yr as specified in Requirement for Operability 6.2.2. l.a (3. 11.2. l.a).

X/OJ Maximum normalized diffusion coefficient of effluent release path j at and beyond the site boundary (sec/m'). Turbine Building and Radwaste Building values are based on average annual ground level values. Main plant vent release values are for mixed mode and may be either short term or average annual value dependent upon type of release.

K; The total whole body dose factor due to gamma emission from noble gas nuclide i (mrem/yr per pCi/m') (as listed in

Table B-1 o'f Regulatory Guide 1. 109, Revision 1).

As defined in Equation (22).

Total number of radionuclides in the gaseous effluent.

Different release pathways.

The total maximum acceptable concentration (C~;) of noble gas radionuclides in the gaseous effluent r'elease path j (pCi/cc) will be, calculated, by using the following equation:

(24) 58

AMENDMENT NO. 11 AUGUST 1992 where:

CgJ The total allowed concentration of all noble gas radionuclides in .the gaseous effluent release path j (pCi/cc).

The maximum acceptable release rate (pCi/sec) of all noble gases in the gaseous effluent release path j.

RJ The effluent release rate (cc/sec) at the point of release.

To determine the maximum acceptable concentration (C;,) of noble gas radio-nuclide i in the gaseous effluent for each individual noble gas in the gaseous effluent (pCi/cc), the following equation will be used:

CgÃ~C<j(pC i /cc) (25) where:

x;; and C>> are as defined in Equations (22) and (24) respectively, the gaseous effluent monitor alarm setpoint will then be calculated as follows:

C.R.j. = g C,,E,,(cpm) (26) i ~ 1 where:

C.R.j Count rate above background (cpm) for gaseous release path j.

CU The maximum acceptable concentration of noble gas nuclide i in the gaseous effluent release path j pCi/cc.

59

AHENDHENT NO. 11 AUGUST 1992 E;; Detection efficiency of the gaseous effluent monitor j for noble gas i (cpm/pCi/cc).

3.6.2.2 Set pints Calculations Based on Skin Dose Limits The method for calculating the setpoints,to ensure compliance with the skin dose limits specified in"Requirement for Operability 6.2.2. l.a (3.11.2. l.a) is similar to the one described for whole body dose limits (Section 3.6.2. 1 of this manual), except Equation (27) will be used instead of Equation (23) for determining maximum acceptable release rate (gr).

F) 3000 (pCi/sec)

(27) where:

The maximum acceptable release rate of all noble gases in the gaseous effluent release path j in pCi/sec.

X/Q; The maximum annual normalized diffusion coefficient for release path j at and beyond the site boundary (sec/m').

F; Fraction of total allowed dose.

The skin dose factor due to beta emission for each identified noble gas radionuclide i in mrem/yr per pCi/m'L; values are listed in Table 3-1)'.

The air dose factor due to gamma emissions for each identified noble gas radionuclide, in mrad/yr per pCi/m'H,.

values are listed in Table 3-1).

60

AMENDMENT NO. 11 AUGUST 1992 1.'1 A conversion

factor to 'convert dose in mrad to dose equiv'alent in mrem.

3000 Skin dose rate limit of 3000 mrem/yr as specified in.

Requirement for Operability 6.2.2.1 (3.11.2.1).

61

Table 3-1 DOSE FACTORS FOR NOBLE GASES AND DAUGHTERS*

Total Body Gamma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor Radionuclide K. L. M. N.

(mrem/yr per ICi/m3) (mrem/yr per yCi/m3) (mrad/yr per ICi/m3) (mrad/yr per yCi/m3)

Kr-85m 1.17E+03** 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 ,

Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3. 12E+03 7.11E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4. 13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2. 69E+03 9.30E+03 3.28E+03 Dl

.% D

The listed dose factors are for radionuclides that may be detected in gaseous effluents. KZ C Pl K7 M

- 7.56E-02 = 7.56 x 10'.

~

lO O

~

The values listed above were taken from Table 8-1 of NRC Regulatory Guide 1. 109, Revision 1. The values were multiplied by 10'o convert picocuries to microcuries

AMENDMENT NO. 15 OCTOBER 1993 Table 3-2 DISTANCES 'MILES) TO TYPICAL CONTROLLING LOCATIONS S'MEASURED FROM CENTER OF WNP-2 CONTAINMENT BUILDING*

Location Distance'ector Dose Pathwa s (miles)

"

Site Boundary 1.2 Air dose measurement One 4.2 ESE Ground, vegetables, and inhalation Two 6.4 SE Ground, meat, and inhalation Three 4.5 ESE Ground, vegetables, and inhalation Four 4.1 ENE Ground, vegetables, and inhalation Five 4.3 NE Ground and inhalation Six ,7.2 ESE Ground, Cow milk, and inhalation

Typical locations and pathways are based on the current Land Use Census (LUC).

63

Table 3-3 WNP-2 LONG-TERN AVERAGE DISPERSION (X/Q)

AND DEPOSITION D VALUES FOR TYPICAL LOCATIONS X/Q X/Q X/Q 2.3 Days 8.0 Days No Decay Decay Decay Location Sector Distance Point of Release No De letion No De letion ~De 1eted (miles) (sec/m ) (sec/m ) (sec/m ),

Site Boundary SE 1.2 Reactor Bldg. 2.7E-07 2.7E-07 2.6E-07 2.0E-09 Turbine Bldg. 1.4E-05 1.3E-05 1.2E-05 1.2E-08 Radwaste Bldg. 1.4E-05 1.3E-05 1.2E-05 . 1.2E-08 One ESE 4.2 Reactor Bldg. 1.5E-06 1.5E-06 1.2E-06 6.0E-10 Turbine Bldg. 1.1E-06 1.0E-06 8.1E-07 6.0E-10 Radwaste Bld . 1.1E-06 1.0E-06 8.1E-07 6.0E-10 Two 6.4 Reactor Bldg. 3.7E-07 3.5E-07 3.4E-07 3.2E-10 Turbine Bldg. 7.2E-07 6.8E-07 5.1E-07 2.6E-10 Radwaste Bld . 7.2E-07 .6.8E-07 5.1E-07 2.6E-10 Three ESE 4.5 Reactor Bldg. 1.6E-06 1.5E-06 1.3E-06 5.1E-10 Turbine Bldg. 1.0E-06 9.8E-07 7.7E-07 5.1E-10 Radwaste Bld . 1.0E-06 9.8E-07 7.7E-07 5.1E-10 Four ENE 4.1 Reactor Bldg. 9.8E-07 9.3E-07 7.7E-07 3.8E-10 Turbine Bldg. 6.9E-07 6.5E-07 5.2E-07 3.7E-10 Radwaste Bld . 6.9E-07 6.5E-07 5.2E-07 3.7E-10 Five NE 4.3 Reactor Bldg. 6.8E-08 6.6E-08 6.6E-08 1.3E-10 Turbine Bldg. 6.7E-07 6.3E-07 5.0E-07 3.7E-10 m Radwaste Bldg. 6.7E-07 6.3E-07 5.0E-07 3.7E-10 OU

.A Wm X Six ESE 7.2 Reactor Bldg. 7.9E-07 7.1E-07 5.9E-07 1.9E-10 OZ U7 W Turbine Bldg.

Radwaste Bldg.

5.2E-07 5.2E-07 4.7E-07 4.7E-07 3.6E-07 3.6E-07 1.9E-10 1.9E-10 nx CrJ Vl

AHENDHENT NO. 9 JANUARY 1992 Table 3-4 DOSE RATE PARAHETERS IHPLEHENTATION OF 10 CFR 20 AIRBORNE RELEASES Child Dose Factor* PI 1

DFA, DFG Inhalation Nuclide ~mrem Ci ~~Ci m

~%sec'.8E-09 H-3 1.7E-07 0.0 6.3E+02 Na-24 1.3E-05 4.4E-06 2.9E-08 1.6E+04 Cr-51 2.9E-07 4.6E-06 2.6E-10 1.7E+04 Hn-54 2.6E-OS 4.3E-04 6.8E-09 1.6E+06 Hn-56 7.5E-05 3.3E-05 1.3E-08 1.2E+05 Fe-55 8.5E-09 3.0E-05 0.0 1.1E+05 Fe-59 1.8E-07 3.4E-04 9.4E-09 1.3E+06 Co-58 1.1E-07 3.0E-04 8.2E-09 1.1E+06 Co-60 4.2E-09 1.9E-03 2.0E-OS 7.0E+06 Cu-64 1.5E-05 9.9E-06 1.7E-09 3.7E+04 Zn-65 3.3E-OS 2.7E-04 4.6E-09 1.0E+06 Zn-69m 1.4E-05 2.7E-05 3.4E-09 1.0E+05 As-76 7.3E-06 1.9E-05 1.7E-07 7.0E+04 Br-82 5.5E-06 5.7E-06 2.2E-OS 2. 1E+04 Sr-89 1.5E-07 5.8E-04 6.5E-13 2.2E+06 Sr-90 7.9E-10 1.0E-02 2. 6E-12** 3.7E+07 Zr-95 1.2E-07 6.0E-04 5.8E-09 2.2E+06 Nb-95 2.3E-07 1.7E-04 6.0E-09 6.3E+05 Zr-97 1.1E-05 9.5E-05 6.4E-09 3.5E+05 Nb-97 1.6E-04 7.5E-06 5.4E-09 2.BE+04 Ho-99 2.9E-06 3.7E-05 2.2E-09 1.4E+05 Tc-99m 3.2E-05 '.3E-06 1.1E-09 4.BE+03 RU-106 2.2E-OS 3.9E-03 1.8E-09 1.4E+07 Ag-110m 3.2E-OS 1.5E-03 2.1E-OS 5.6E+06 Sb-124 1.3E-07 S.SE-04 1.5E-OS 3.3E+06 Sb-125 7.9E-09 6.3E-04 3.5E-09, 2.3E+06 Sb-126 6.5E-07 2.9E-04 1.0E-OS 1. 1E+06 Sb-127 2.1E-06 6.2E-05 6.6E-09 2. 3E+05 Te-127 2.1E-05 1.5E-05 1.1E-11 5.6E+04 Te-131m 6.4E-06 8.3E-05 9.9E-09 3. 1E+05 I-131 1.0E-06 4.4E-03 3.4E-09 1. 6E+07 I-132 8.4E-05 5.2E-05 2.0E-08 1. 9E+05 I-133 9.2E-06 1.0E-03 4.5E-09 3.7E+06 I-135 2.9E-05 2.1E-04 1.4E-OS 7.8E+05 Cs-134 1.1E-OS 2.7E-04 1.4E-OS 1.0E+06 65

AMENDMENT NO. 9

. JANUARY 1992 Table 3-4 DOSE RATE PARAMETERS IMPLEMENTATION'OF 10 CFR 20 AIRBORNE RELEASES Child Dose Factor* pI 1

DFG; Inhalation Nucl i de

~%sec'FA; mrem Ci ~modem Ci m hr

~~Ci m Cs-137 7.3E-10 2.5E-04 4.9E-09 9. 3E+05 Cs-138 3.6E-04 2.3E-07 2.4E-OS 8.5E+02 Ba-140 6.3E-07 4.7E-04 2.4E-09 1.7E+06 La-140 4.8E-06 6.1E-05 1.7E-OS 2.3E+05 Ce-141 2.4E-07 1.5E-04 6.2E-10 5.6E+05 Ce-144 2.8E-OS 3.2E-03 3.7E-10 1.2E+07 Nd-147 7.2E-07 8.9E-05 1.2E-09 3.3E+05 Hf-179m 3.7E-02 2.0E-05 NO DATA 7.4E+04 Hf-181 1.8E-07 6.0E-05 1.2E-08 2.2E+05 W-185 1.1E-07 1.9E-04 0.0 7.0E+05 Np-239 3.4E-06 1.7E-05 9.5E-10 6.4E+04

Maximum Organ

- No data is listed for Sr-90 in Table E-6 of Regulatory Guide 1. 109, Revision l. Y-90 values were used for dose conversion factor Sr-90.

66

AHENDHENT NO. 9 JANUARY 1992 Table '3-Sa DOSE PARAHETERS'FOR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES AGE GROUP: ADULT ORGAN OF

REFERENCE:

HAXIHUH ORGAN R(I), INDIVIDUAL-PATHWAY DOSE PARAHETERS FOR RADIONUCLIDES OTHER TKAN NOBLE GASES RADIO- INHALATION GROUND PLANE COW%ILK'H2.HREH/YR GOAT-HILK ANIMAL-HEAT VEGETABLES NUCLIDE (HREH/YR (H2.HREH/YR (M2.HREH/YR (H2.HREH/YR (H2.HREM/YR PER pCI/H3) PER pCI/SEC) PER pCI/SEC) PER pCI/SEC) PER pCI/SEC) PER UCI/SEC M 3 7.2E+02 0. OE-01 5.BE+02 1. 2E+03 2.4E+02 1.6E+03 NA 24 1.DE+04 1. 2E+07 1.2E+06 2.2E+05 7.2E-04 1.1E+05 CR 51 1.4E+04 4.7E+06 3.3E+06 5.BE+05 8.2E+05 2.3E+07 HN 54 1.4E+06 1.4E+09 1.4E+07 2.1E+06 1.5E+07 9.4E+08 MN 56 2.DE+04 9.DE+05 6.2E-02 1.1E-02 O.OE-01 2.DE+02 FE 55 7.2E+04 O.OE-OI 1.4E+07 2.2E+06 1.6E+08 1.9E+08 FE 59 I.DE+06 2.7E+08 1. IE+08 2.DE+07 9.BE+08 1.5E+09 CO 58 9.3E+05 3.8E+08 4.7E+07 7.6E+06 1.BE+08 B.OE+08 CO 60 6.DE+06 2.3E+10 1.7E+08 2.5E+07 B.DE+08 2.9E+09 CU 64 4.9E+04 6. 1E+05 I.DE+06 1.7E+05 1. IE-05 3.3E+05 ZN 65 8.6E+05 7.5E+08 2.7E+09 4.DE+08 7.DE+08 1.3E+09 ZN 69H 1.4E+05 . 1.3E+06 1.3E+07 2.4E+06 1.2E"03 1.4E+06 AS 76 1.5E+05 3.BE+06 2. IE+07 3.BE+06 2.9E+01 B.DE+06 BR 82 1.4E+04 2. 1E+07 1.9E+07 3.4E+06 7.DE+02 7.7E+05 SR 89 1.4E+06 2.2E+04 6.9E+08 2.DE+09 1.4E+08 1.5E+10 SR 90 2.9Et07 6.7E+06 3.4E+10 8.3E+10 8.9E+09 7.4E+11 ZR 95 1.BE+06 2.5E+08 4.6E+05 7.6E+04 9.2E+08 1.6E+09 NB 95 5. 1E+05 1.4E+08 1.3E+08 2.2E+07 3.6E+09 8.4E+08 ZR 97 5.2E+05 3.DE+06 1.4E+04 2.4E+03 6.4E-01 B.BE+06 NB 97 2.4E+03 I.BE+05 1.6E-09 2.9E"10 O.OE-OI B. IE-04 MO 99 2.5E+05 4.DE+06 2.9E+07 5.2E+06 1.2E+05 9.3E+06 TC 99M 4.2E+03 1.BE+05 2.BE+03 S.DE+02 3.6E-18 Z.ZE+03 RU106 9.4E+06 4.2E+08 7.3E+05 1. IE+05 I.DE+11 1.2E+10 AG110H 4.6E+06 3.5E+09 1.2E+10 1.BE+09 1.4E+09 4.4E+09 58124 2.5E+06 6.DE+08 3.5E+08 5.BE+07 2.7E+08 4.DE+09 58125 1.7E+06 2.4E+09 1.3E+08 I.BE+07 I.ZE+08 1.4E+09 58126 7.7E+05 8.4E+07 2.2E+08 4.DE+07 7.6E+07 1.6E+09 SB127 3.DE+05 1.7E+07 5.2E+07 9.3E+06 1.9E+06 1.2E+08 TE127 5.7E+04 3.DE+03 2.6E+04 4.7E+03 8.4E-09 2.DE+05 TE131H 5.6E+05 B.DE+06 8.9E+06 1.6E+06 1. IE+04 2.DE+07 I 131 1.2E+07 8.6E+06 3.4E+10 6. IE+10 1.2E+09 4.4E+10 I 132 1. IE+05 6.2E+05 3.9E+00 6.9E+00 O.OE-01 I.IE+03 I 133 2.2E+06 1.2E+06 2.5E+08 4.5E+08 2.4E+01 I.IE+08 I 135 4.5E+05 1.3E+06 5.5E+05 9.BE+05 1. 7E-15 1.4E+06 CS134 8.5E+05 6.9E+09 7.4E+09 2.7E+10 8.6E+08 1. DE+10 CS136 1.5E+05 1.5E+08 S.DE+08 2.2E+09 2.3E+07 4.6E+08 CS137 6.2E+05 1.3E+10 6.DE+09 2.1E+10 7. 1E+08 8.6E+09 CS138 6.2E+02 3.6E+05 I.OE-23 4.6E-23 O.OE-OO 3.0E-11 BA140 1.3E+06 2.1E+07 2.7E+07 4.BE+06 2.BE+07 7.3E+08 LA140 4.6E+05 1.9E+07 8.4E+04 1.5E+04 7.DE+02 3.3E+07 CE141 3.6E+05 1.4E+07 5.BE+06 I.DE+06 1.7E+07 9.3E+08 CE144 7.BE+06 7.DE+07 6.4E+07 9.6E+06 2.6E+08 1. IE+10 ND147 2.2E+05 8.5E+06 2.5E+05 4.6E+04 1.9E+07 5. IE+08 HF179H 1.6E+05 O.OE-OI O.OE-OI O.OE-OI O.OE-OI O.OE-OI MF181 4.BE+05 2. 1E+08 5.5E+05 9.3E+04 1. 2E+10 I.BE+09 N 185 4.5E+05 1.BE+04 2.4E+07 3.9E+06 1.9E+07 8.4E+08 NP239 1.2E+05 1. 7E+06 3.7E+04 6.7E+03 2.6E+03 1.6E+07 NOTE: The Y-90 ground plane dose factor was used for Sr-90.

The PARTS subroutine of GASPAR II was used to produce this table.

67

AMENDMENT, NO. 9 JANUARY 1992 J

Table 3-5b I

DOSE PARAMETERS'OR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES

. AGE GROUP: TEEN ORGAN OF

REFERENCE:

MAXIMUM ORGAN R(I),. INDIVIDUAL PATHWAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES RADIO- INHALATION GROUND PLANE COW-MILK GOAT&ILK ANIMAL-MEAT VEGETABLES

"

NUCLIDE (HREH/YR (M2.MREM/YR (H2.HREH/YR (H2.MREH/YR (H2.MREH/YR (H2.MREH/YR PER pCI/H3) PER pCI/SEC) PER pCI/SEC) PER pCI/SEC) PER pCI/SEC) PER pCI/SEC M 3 7.3E+02 O.OE-OI 7.5E+02'. 1.5E+03 1.5E+02 1.9E+03 NA 24 1.4E+04 1.2E+07 IE+06 3.9E+05 5.8E"04 1.DE+05 CR 51 2.1E+04 4.7E+06 3.9E+06 6.BE+05 4.4E+05 2.5E+07 MN 54 2.DE+06 1.4E+09 1.6E+07 2.3E+06 7.BE+06 9.BE+08 HN 56 5.7E+04 9.DE+05 2.3E-OI 4.1E-02 O.OE-OD 3.7E+02 FE 55 1.2E+05 O.OE-01 2.4E+07 3.BE+06 1.3E+08 3.DE+08 FE 59 1.5E+06 2.7E+08 1.3E+08 2.5E+07 5.5E+08 1.7E+09 CO 58 1.3E+06 3.BE+08 5.3E+07 8.7E+06 9.4E+07 8.3E+08 CO 60 8.7E+06 2.3E+10 2.1E+08 3.DE+07 4.3E+08 3.1E+09 CU 64 6. IE+04 6. IE+05 1.6E+06 2.7E+05 B.OE-06 2.7E+05 ZN 65 1.2E+06 7.5E+08 4.5E+09 6.7E+08 5.4E+08 2.DE+09 ZN 69M 1.7E+05 1.3E+06 2.1E+07 3.BE+06 9. IE-04 1. IE+06 AS 76 1.5E+05 3.BE+06 2.7E+07 4.9E+06 1.7E+01 5.3E+06 BR 82 I.BE+04 2. IE+07 2.BE+07 5.1E+06 4.9E+02 6. IE+05 SR 89 2.4E+06 2.2E+04 1.3E+09 3.7E+09 1.2E+08 2.4E+10 SR 90 3.3E+07 6.7E+06 5.1E+10 1.3E+11 6.2E+09 I.DE+12 ZR 95 2.7E+06 2.5E+08 5.BE+05 9.5E+04 5.3E+08 1.BE+09 NB 95 7.5E+05 1.4E+08 1.6E+08 2.7E+07 2.DE+09 9. 1E+08 ZR 97 6.3E+05 3.DE+06 2.1E+04 3.BE+03 4.6E-OI 7.DE+06 NB 97 3.9E+03 I.BE+05 1.9E-OB 3.3E-09 O.OE-OI 4.8E"03 HO 99 2.7E+05 4.DE+06 5.1E+07 9.2E+06 9.4E+04 1. IE+07 TC 99M 6. IE+03 I.BE+05 5.3E+03 9.5E+02 3.2E-IB 2. IE+03 RU106 1. 6E+07 4.2E+08 9.9E+05 1.5E+05 6.2E+10 1.5E+10 AGIIOM 6.BE+06 3.5E+09 1.4E+10 2.1E+09 7.6E+08 4.6E+09 SB124 3.BE+06 6.DE+08 4.5E+08 7.3E+07 1.6E+08 4.6E+09 58125 2.7E+06 2.4E+09 1.6E+08 2.3E+07 6.BE+07 1.6E+09 58126 1.2E+06 8.4E+07 2.BE+08 5. 1E+07 4.5E+07 1.BE+09 58127 3.2E+05 1.7E+07 6.9E+07 1. 2E+07 1.2E+06 1. 2E+08 TE127 8. IE+04 3.DE+03 4.BE+04 8. 6E+03 7.0E-09 I.BE+05 TE131H 6.2E+05 B.DE+06 1.3E+07 2. 3E+06 7.4E+03 1.5E+07 I 131 1.5E+07 8.6E+06 5.4E+10 9.7E+10 9.DE+08 6. IE+10 I 132 1.5E+05 6.2E+05 6.4E+00 1. 2E+01 O.OE-OO 9.3E+02 I 133 2.9E+06 1.2E+06 4.2E+08 7.5E+08 I.BE+01 9.6E+07 I 135 6.2E+05 1.3E+06 9.3E+05 1.7E+06 1.3E-15 1.2E+06 CS134 1. IE+06 6.9E+09 1.3E+10 4.6E+10 6.BE+08 1.6E+10 CS136 1.9E+05 1.5E+08 8.4E+08 3.BE+09 I.BE+07 7.DE+08 CS137 8.5E+05 1.3E+10 I. IE+10 3.BE+10 5.7E+08 1. 4E+10 CS138 8.6E+02 3.6E+05 I.BE-23 8. IE-23 O.OE-OO 2.7E-11 BA140 2.DE+06 2.1E+07 3.6E+07 6.4E+06 I.BE+07 B.BE+08 LA140 4.9E+05 1.9E+07 1. IE+05 2. IE+04 4.4E+02 2.4E+07 CE141 6. IE+05 1.4E+07 7.9E+06 1.4E+06 I.DE+07 I.IE+09 CE144 1.3E+07 7.DE+07 B.BE+07 1.3E+07 1.6E+08 1.3E+10 ND147 3.7E+05 8.5E+06 3.5E+05 6.2E+04 1,2E+07 6.1E+08 HF 179M 7. IE+04 O.OE"01 O.OE-OI O.OE-OI O.OE-OI O.OE-01 HF181 4.BE+05 2. IE+08 7.1E+05 1.2E+05 7.DE+09 2.1E+09 W 185 7.7E+05 I.BE+04 3.3E+07 5.4E+06 1.2E+07 I.DE+09 NP239 1.3E+05 1.7E+06 5.3E+04 9.6E+03 1.7E+03 1.4E+07 NOTE: The Y-90 ground plane dose factor was used for Sr-90.

The PARTS subroutine of GASPAR II was used to produce this table.

68

AMENDMENT NO. 9 JANUARY 1992 Table 3-5c DOSE 'PARAMETERS, FOR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES

'

AGE GROUP:" CHILD ORGAN OF .

REFERENCE:

MAXIMUM ORGAN R(I), INDIVIDUAL PATHWAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES RADIO- INHALATION GROUND PLANE COW-MILK GOAT-HILK ANIMAL-HEAT VEGETABLES HUCLIDE (HREH/YR (M2.HREM/YR '(H2.HREH/YR (H2.MREH/YR (M2.MREH/YR (M2.HREH/YR pCI/H3) pCI/SEC) pCI/SEC)

'ER PER PER PER pCI/SEC) PER pCI/SEC) PER pCI/SEC H 3 6.4E+02 0. OE-01 1. 2E+03 2. 4E+03 I.BE+02 2.9E+03 HA 24 1.6E+04 1. 2E+07 4.5E+06 B.DE+05 9.2E-04 1.6E+05 CR 51 1.7E+04 4.7E+06 2.5E+06 4.4E+05 2.2E+05 1.6E+07 MN 54 1.6E+06 1.4E+09 1. 1E+07 1.7E+06 4.3E+06 6.9E+08 HN 56 1.2E+05 9.DE+05 B.BE-01 1.6E-01 O.OE-OO 1. IE+03 FE 55 1. 1E+05 O.OE-01 6.1E+07 9.6E+06 2.5E+08 7.6E+08 FE 59 1.3E+06 2.7E+08 9.5E+07 1.7E+07 3.DE+08 1.2E+09 CO 58

'O

1. IE+06 3.BE+08 3.4E+07 5.6E+06 4.7E+07 5.3E+08 60 7. 1E+06 2.3E+10 1.4E+08 2.DE+07 2.2E+08 2. IE+09 CU 64 3.7E+04 6.1E+05 1.7E+06 2.9E+05 6.5E-06 2.2E+05 ZN 65 1.DE+06 7.5E+08 6.BE+09 1.DE+09 6.2E+08 3.DE+09 ZH 69H 1.DE+05 1.3E+06 2.2E+07 4.DE+06 7.2E-04 9.DE+05 AS 76 7.DE+04 3.BE+06 2.2E+07 4.DE+06 I.IE+01 3.3E+06 BR 82 2.1E+04 2.1E+07 5.BE+07 1.DE+07 7.6E+02 9.5E+05 SR 89 2.2E+06 2.2E+04 3.1E+09 9.2E+09 2.3E+08 6.DE+10 SR 90 3.BE+07 6.7E+06 1.DE+11 2.6E+11 9.BE+09 2. IE+12 ZR 95 2.2E+06 2.5E+08 4.2E+05 7.DE+04 3.DE+08 1.3E+09 NB 95 6. IE+05 1.4E+08 1.1E+08 1.BE+07 I.DE+09 6.2E+08 ZR 97 3.5E+05 3.DE+06 2. IE+04 3.BE+03 3.5E-01 5.2E+06 NB 97 2.BE+04 1.BE+05 4.2E-07 7.6E-OB O.OE-DI 8.2E"02 HO 99 1.3E+05 4.DE+06 8.7E+07 1.6E+07 1.2E+05 1.6E+07 TC 99H 4.BE+03 1.BE+05 7.4E+03 1.3E+03 3.4E-IB 2.2E+03 RU106 1.4E+07 4.2E+08 7.9E+05 1.2E+05 3.BE+10 1. 2E+10 AG110H 5.5E+06 3.5E+09 9.4E+09 1.4E+09 3.BE+08 3.DE+09 58124 3.2E+06 6.DE+08 3.3E+08 5.4E+07 B.BE+07 3.3E+09 58125 2.3E+06 2.4E+09 1.2E+08 1.7E+07 3.BE+07 1.2E+09 SB126 1.1E+06 8.4E+07 2.2E+08 4.DE+07 2.7E+07 1.4E+09 58127 2.3E+05 1.7E+07 5.5E+07 I.DE+07 7.2E+05 9.2E+07 TE127 5.6E+04 3.DE+03 5.9E+04 1.1E+04 6.7E-09 1.7E+05 TE131H 3.1E+05 B.DE+06 1.1E+07 2.1E+06 S.DE+03 9.BE+06 I 131 1.6E+07 8.6E+06 1.1E+11 1.9E+11 1.4E+09 1.2E+11 I 132 1.9E+05 6.2E+05 1.5E+01 2.7E+01 O.OE-OO 1.6E+03 I 133 3.BE+06 1.2E+06 9.9E+08 1.BE+09 3.3E+01 1.7E+08 I 135 7.9E+05 1.3E+06 2. IE+06 3.BE+06 2.3E-15 2. IE+06 CS134 1.DE+06 6.9E+09 2. DE+10 7.5E+10 8.3E+08 2. 6E+10 CS136 1.7E+05 1.5E+08 1.3E+09 6.DE+09 2. IE+07 1. IE+09 CS137 9.1E+05 1.3E+10 1.9E+10 6.BE+10 7.9E+08 2. SE+10 CS138 8.4E+02 3.6E+05 3.2E-23 1.4E-22 O.OE-OO 3.6E-II BA140 1.7E+06 2. 1E+07 5.6E+07 1.DE+07 2. 1E+07 1.4E+09 LA140 2.3E+05 1.9E+07 9.5E+04 1.7E+04 2.BE+02 1.6E+07 CE141 5.4E+05 1.4E+07 6.3E+06 1. IE+06 6.4E+06 9.DE+08 CE144 1.2E+07 7.DE+07 7.DE+07 1.1E+07 1.DE+08 1. 1E+10 ND147 3.3E+05 8.5E+06 2.BE+05 5.DE+04 7.4E+06 4.BE+08 HF 179M 7.4E+04 O.OE-01 O.OE-OI O.OE-01 O.OE-01 O.OE-OI HF181 2.2E+05 2. 1E+08 5.9E+05 9.9E+04 4.4E+09 I.BE+09 M 185 6.9E+05 1.BE+04 2.7E+07 4.3E+06 7.3E+06 8.3E+08 NP239 6.4E+04 1.7E+06 4.6E+04 8.3E+03 1. 1E+03 I.DE+07 NOTE: The Y-90 ground plane dose factor was used for Sr-90.

The PARTS subroutine of GASPAR II was used to produce this table.

69

AMENDMENT NO. 9 JANUARY 1992 i

Table 3-5d DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES

'AGE GROUP: INFANT ORGAN OF

REFERENCE:

MAXIMUM ORGAN R(I), INDIVIDUAL PATHWAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES RADIO- 'NHALATION GROUND PLANE COW%ILK GOAT-HILK ANIHALWEAT VEGETABLES NUCLIDE (HREH/YR (H2.HREH/YR (H2.HREH/YR 'H2.HREH/YR (H2.HREH/YR (H2.HREH/YR PER pCI/H3) PER pCI/SEC) PER pCI/SEC) PER pCI/SEC) PER pCI/SEC) PER pCI/SEC H 3 3.7E+02 O.OE-01 1. BE+03 3.7E+03 O.OE-OI O.OE-01 NA 24 ~ I.IE+04 1.2E+07 " 7.BE+06 1.4E+06 O.OE-OI O.OE-01 CR 51 1.3E+04 4.7E+06 2.2E+06 3.BE+05 O.OE-OI O.OE-OI HN 54 I.DE+06 1.4E+09 2.1E+07 3.1E+06 0. OE-01 O.OE-OI HN 56 7.2E+04 9.DE+05 1.3E+00 2.4E"01 O.OE"01 O.OE-OI FE 55 8.7E+04 O.OE-01 7.4E+07 1.2E+07 O.OE-OI O.OE-OI FE 59 I.DE+06 2.7E+08 1.BE+08 3.4E+07 O.OE-OI O.OE-OI CO 58 7.BE+05 3.BE+08 2.9E+07 4.BE+06 O.OE-OI O.OE-OI CO 60 4.5E+06 2.3E+10 1.2E+08 1.7E+07 O.OE-OI O.OE-01 CU 64 1.5E+04 6.1E+05 1.9E+06 3.2E+05 O.OE-OI O.OE-01 ZN 65 6.5E+05 7.5E+08 1. 2E+10 1.7E+09 O.OE-OI O.OE-OI ZN 69H 4.1E+04 1.3E+06 2.4E+07 4.3E+06 O.OE"01 O.OE-OI AS 76 2.7E+04 3.BE+06 2.2E+07 4.DE+06 O.OE-01 O.OE"01 BR 82 1.3E+04 2. IE+07 9.BE+07 I.BE+07 O.OE-OI O.OE-OI SR 89 2.DE+06 2.2E+04 6.DE+09 I.BE+10 O.OE-OI O.OE-O)

SR 90 1.6E+07 6.7E+06 1.2E+11 2.9E+11 O.OE-OI O.OE-OI ZR 95 I.BE+06 2.5E+08 4.DE+05 6.5E+04 O.OE-OI O.OE"01 NB 95 4.BE+05 1.4E+08 9.6E+07 1.7E+07 O.OE-OI O.OE-OI ZR 97 1.4E+05 3.DE+06 2.2E+04 4.DE+03 O.OE-OI O.OE-OI NB 97 2.7E+04 I.BE+05 I. IE-06 1.9E-07 O.OE-OI O.OE-01 HO 99 1.3E+05 4.DE+06 1.6E+08 2.BE+07 O.OE-01 O.OE-OI TC 99H 2.DE+03 I.BE+05 8.2E+03 . 1.5E+03 O.OE-01 O.OE-OI RU106 1.2E+07 4.2E+08 B.DE+05 1.2E+05 D.OE-OI O.OE-OI AG110H 3.7E+06 3.5E+09 8.2E+09 1.2E+09 O.OE-OI O.OE-01 58124 2.6E+06 6.DE+08 3.1E+08 5.1E+07 O.OE-OI O.OE-OI 58125 1.6E+06 2.4E+09 I.IE+08 1.6E+07 O.OE-OI O.OE-OI 58126 9.6E+05 8.4E+07 2.1E+08 3.7E+07 O.OE-01 O.OE-01 58127 2.2E+05 1.7E+07 5.5E+07 9.9E+06 O.OE-OI O.OE-OI TE127 2.4E+04 3.DE+03 6.BE+04 1.2E+04 O.OE"01 O.OE-OI TE131H 2.DE+05 B.DE+06 1.2E+07 2. IE+06 O.OE-OI O.OE-OI I 131 1.5E+07 8.6E+06 2.6E+11 4.7E+11 O.OE-OI O.OE-OI I 132 1.7E+05 6.2E+05 3.4E+01 6.1E+01 D.OE-OI O.OE-OI I 133 3.6E+06 1.2E+06 2.4E+09 4.3E+09 O.OE-OI O.OE-OI I 135 7.DE+05 1.3E+06 4.9E+06 8.9E+06 O.OE-OI O.OE-OI CS134 '.DE+05 6.9E+09 3.7E+10 1.4E+11 O.OE-DI O.OE-OI CS136 1.3E+05 1.5E+08 2.BE+09 1. 2E+10 O.OE"01 O.OE-01 CS137 6. 1E+05 1.3E+10 3.6E+10 1.3E+11 D.OE"01 O.OE-01 CS138 B.BE+02 3.6E+05, 1.2E-22 5.6E-22 O.OE-OI O.OE-OI BA140 1.6E+06 2. IE+07 1.2E+08 2. IE+07 O.OE-OI O.OE-OI LA140 1.7E+05 1.9E+07 9.4E+04 1. 7E+04 O.OE-01 O.OE-01 CE141 5.2E+05 1.4E+07 6.4E+06 1. IE+06 O.OE-OI O.OE-OI CE144 9.BE+06 7.DE+07 7. IE+07 1. IE+07 O.OE-OI O.OE-OI ND147 3.2E+05 8.5E+06 2.BE+05 S.DE+04 O.OE-OI O.OE-OI HF179H 2.BE+04 O.OE"01 O.OE-OI O.OE-OI O.OE-01 O.OE-01 HF181 8.4E+04 2. IE+08 5.9E+05 9.9E+04 O.OE-OI O.OE-OI M 185 6.3E+05 I.BE+04 2.7E+07 4.4E+06 O.OE"01 O.OE-OI NP239 6.DE+04 1.7E+06 4.7E+04 8.5E+03 O.OE-OI O.OE-OI NOTE: The Y-90 ground plane dose factor was used for Sr-90.

The PARTS subroutine of GASPAR II was used to produce this table.

70

AHENDHENT NO. 9 JANUARY 1992 Table 3-6 INPUT PARAHETERS FOR CALCULATING R; Parameter =-Val ue Table*

r (dimensionless) 1.0 for radioiodine E-15 0.2 for particulates E-15 F (days/1 i ter) Each stable element E-1 U., (liters/yr) --Infant 330 E-5

--Child 330 E-5

--Teen 400 E-5

--Adult 310 E-5 (DFL;), (mrem/pCi) Each radionuclide E-11 to E-14 Y~ (kg/m') 0.7 E-15 Y, (kg/m') 2.0 E-15 t, (seconds) 1.73 x 10'2 days) E-15 t (seconds) 7.78 x 10'90 days) E-15 gF (kg/day) 50 for cow E-3 6 for goat E-3 fs (dimensionless) 1.0 NUREG-0133 fp (dimensionless) 0.5 for cow Site specific 0.75 for goat Site specific

Of Regulatory Guide 1. 109, Revision 1 unless stated otherwise.

71

AHENDHENT NO. 9

, JANUARY 1992 Table 3-7 INPUT PARAHETERS FOR CALCULATING R; Parameter Value Table*

II r (dimensionless) 1.0 for radioiodine E-15 0.2 for particulates E-15 F, (days/kg) Each stable element E-1 U (kg/yr) --Infant 0 E-5

--Child 41 E-5

--Teen 65 .E-5

--Adult 110 (DFL;), (mrem/pCi) Each radionuclide E-11 to E-14 Y, (kg/m') 0.7 E-15 Y, (kg/m') 2.0 E-15 t, (seconds) 1.73 x 10'20 days) E-15 t (seconds) 7.78 x 10 (90 days) E-15 gF (kg/day) 50 E-3

Of Regulatory Guide 1. 109, Revision 1.

72

AMENDMENT NO; 11 AUGUST 1992 Table 3-8 INPUT PARAMETERS FOR CALCULATING R.

a

)

Parameter Value Table*

r,(dimensionless) 1.0 for radioiodine E-1 0.2 for particulates E-1 (DFL;) (mrem/pCi) Each radionuclide E-11 to E-14 UL a (kg/yr) -- Infant E-5

--Child 26 E-5

--Teen 42 E-5

--Adult 64 E-5 U'

(kg/yr) --Infant E-5

--Child 520 E-5

--Teen 630 E-5

--Adult 520 E-5 f (dimensionless) 0.42 Ref 2**

f, (dimensionl ess) 0.76 E-15 t (seconds) 8.6 x 10'1 day) E-15 t> (seconds) 5.18 x 10 (60 days) E-15 Y (kg/m') 2.0 E-15

Of Regulatory Guide 1.109, Revision 1.

- Refer to Table 3-14.

73

AMENDMENT NO. 9 JANUARY 1992 Table 3-9 E

INPUT PARAMETERS NEEDED FOR CALCUL'ATING DOSE SUMMARIES TO THE MAXIMUM

'INDIVIDUALAND THE POPULATION WITHIN 50 MILES FROM WNP-2 'GASEOUS EFFLUENT In ut Parameter Value Reference*

Distance to Maine (miles) 3000 Ref 1 Fraction of year leafy vegetables are grown 0.42 Ref 2 Fraction of year cows are on pasture 0.5 Ref 2 Fraction of crop from garden 0.76 Ref 3 Fraction of daily intake of cows derived from pasture while on pasture 1.0 Ref 2 Annual average relative humidity (%) 53.8 Ref 4 Annual average temperature (F') 53.0 Ref 5 Fraction of year goats are on pasture 0.75 Ref 2 Fraction of daily intake of goats derived from pasture while on pasture 1.0 Ref 2 Fraction of year beef cattle are on pasture 0.5 Ref 2 Fraction of daily intake of beef cattle derived from pasture while on pasture '1. 0 Ref 2 Population within 50 miles of plant by direction and radii interval in miles. 252,356 Ref 6 Annual 50-mile milk production (liters/yr) 2.8E+08 Refs 7 &9 Annual 50-mile meat production (kg/yr) 2.3E+07 Refs 7 & 9 Annual 50-mile vegetable production (kg/yr) 3.5E+09 Refs 7 & 9 Source terms Ref 8 74

AMENDMENT NO. 15 OCTOBER 1993 h

Table 3-9 (contd.')

In ut Parameter Value Reference*

X/g values by sector for each dis-tance ]recirculation, no decay) See Tables 3-10 (sec/m ) and 3-11 Ref 10 X/g values by sector for each dis-tance (recirculation, 2.26 days See Tables'3-10 decay, undepleted) (sec/m') and 3-11 Ref 10 X/g values by sector for each dis-tance (recirculation, 8.0 days See Tables 3-10 decay, depleted) (sec/m') and 3-11 Ref 10 D/g values by sector for each dis- See Tables 3-10 tance (1/m') ..and 3-11 Ref 10

References are listed in Table 3-14.

75

, AMENDMENT NO.

9

, JANUARY 1992 TABLE 3-10 REACTOR BUILDING STACK X/Q AND D/Q VALUES, A) NO DECAY, UNDEPLETED CHI/Q (SEC/HETER CUBED) FOR EACH SEGHENT SEGHENT BOUNDARIES IN HILES FROM THE SITE DIRECTION .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 FROM SITE 5 3.899E-07 1.486E-07 6.171E-OB 3.982E-OB 3.093E-OB 2.000E-OB 2.118E-07 1.769E-07 1. 196E-07 8.944E-OB SSM 2.557E-07 9.471E-OB 3.914E-OB 2.553E-OB 2.000E-OB 1.411E-OB 1.702E-07 1.431E-07 9.698E-OB 7.264E-OB 5'M 1.635E-07 6.378E-OB 3.299E-OB 2.517E-OB 1.999E"08 3.647E-OB 1.045E-07 7.704E-OB 5.209E-OB 3.894E-OB MSM 6.676E-OB 2.927E-OB 1.506E-OB 1. 122E-08 8.872E-09 1.668E-OB 5.532E-OB 4. 156E-08 2.808E-OB 2.09BE-OB M 6.588E-OB 2.996E-OB 1.509E-OB 1.090E-OB 8.368E-09 4.928E-09 2.837E-OB 2.330E-OB 1.569E-OB l. 170E-08 MNM 1.279E-07 5.746E-OB 3.018E-OB 2.258E-OB 1.781E-OB 1.324E-OB 5. 160E-08 4. 103E-08 2.750E-OB 2.044E-OB NM 2.294E-07 8.625E-OB. 3.624E-OB 2.423E-OB 1.934E-OB 1.543E-OB 9.519E-OB 7.785E-OB 5.228E-OB 3.891E-OB NNM 5.137E-07 1.770E-07 6.982E-OB 4.507E-OB 4.224E-OB 2.976E-OB 1.801E-07 1.479E-07 9.945E-OB 7.407E-OB N 6.024E-07 2.016E-07 8.063E-OB 5.264E-OB 4.120E-OB 2.146E-07 2.652E-07 1.430E-07 9.579E-OB 7. 115E-08 4.988E-07 1.690E"07 6.861E-OB 4.526E-OB 4.339E-OB 2.904E-07 1.966E-07 1.057E-07 7.066E-OB 5.243E-OB 3.347E-07 1. 195E-07 4.965E-OB 4.175E-OB 1.400E-07 3.198E-07 1.723E-07 9.247E-OB 6. 174E-08 4.576E-OB ENE 4.184E-07 3.067E-07 4.347E-07 9.267E-07 8.436E-07 4.052E-07 1.641E-07 8.817E-OB 5.893E-OB 4.371E-OB E 4.207E-07 3.460E-07 4.968E-07 1.027E-06 8.714E-07 4.159E-07 1.669E-07 8.906E-OB 5.928E-OB 4.385E-OB ESE 6.224E-07 5.205E-07 7.813E-07 1.572E-06 1.364E-06 5.365E-07 2.045E-07 1.403E-07 9.350E"08 6.922E-OB SE 5.045E-07 2.156E-07 1.174E-07 3.944E-07 6.347E-07 3.083E-07 2.738E-07 1.923E-07 1.289E-07 9.576E-OB SSE 4.591E-07 1.855E-07 7.985E-OB 5.319E-OB 4.237E-OB 3.085E-OB 2.635E-07 2. IBBE-07 1.475E-07 1. 100E-07 76

AMENDMENT NO. 9 JANUARY 1992 t

TABLE 3-10 (CONTD)

B) 2.260 DAY DECAY, UNOEPLETED CHI/O (SEC/HETER CUBED) FOR EACH SEGHENT SEGHENT BOUNDARIES IN HILES FROH THE SITE DIRECT IOK .5-1 1-2 2-3 3-4 4-5 5-10 10-20 '20-30 30-40 40-50 FROH SITE S 3.887E-07 1.477E-07 6.094E-OB 3.904E-OB 3.008E-OB 1.898E-OB 1.813E-07 1.424E-07 8.918E-OB 6.201E-OB SS'M 2.550E-07 9.411E-OB 3.866E-OB 2.504E-OB 1.947E-OB 1.341E-OB 1.493E-07 1. 190E-07 7.530E-OB 5.275E-OB SM 1.630E-07 6.338E-OB 3.255E-OB 2.463E-OB 1.939E-OB 3.438E-OB 9. 132E-08 6.300E-OB 3.969E-OB 2.776E-OB MSM 6.657E-OB 2.909E-OB 1.484E-OB 1.093E-OB 8.533E-09 1.521E-OB 4.618E-OB 3.210E-OB 1.995E-OB 1.381E-OB M 6.563E-OB 2.972E-OB 1.488E-OB 1.069E-OB 8.157E-09 4.721E-09 2.319E"08 1.757E-OB 1.077E-OB 7.365E-09 MN'M 1.275E-07 5.702E-OB 2.970E-OB 2.201E-OB 1.717E-OB 1.226E-OB 4.063E-OB 2.933E-OB 1.765E-OB 1. 194E-08 NM 2.287E-07 8.575E-OB 3.584E-OB 2.381E-OB 1.888E-OB 1.470E-OB 8.026E-OB 6.139E-OB 3.811E-OB 2.642E-OB NNM 5. 125E-07 1.760E-07 6.913E-OB 4.439E-OB 4.130E-OB 2.853E-OB 1.614E-07 1.269E-07 8.077E-OB 5.711E-OB K 6.011E-07 2.006E-07 7.988E"08 5.189E-OB 4.040E-OB 2.000E-07 2.381E-07 1.202E-07 7.574E-OB 5.313E-OB NNE 4.978E-07 1.682E-07 6.795E-OB 4.456E-OB 4.236E-OB 2.707E-07 1.714E-07 8.475E-OB 5.256E-OB 3.639E-OB 3.339E-07 1.188E-07 4.909E-OB 4.089E-OB 1.348E-07 2.908E-07 1.447E-07 7.008E-OB 4.277E-OB 2.924E-OB EHE 4.172E-07 3.040E-07 4.272E-07 8.948E-07 7.996E-07 3.720E-07 1.390E"07 6.706E-OB 4. 134E-08 2.827E-OB E 4.194E-07 3.430E-07 4.885E-07 9.909E-07 8.315E-07 3.861E-07 1.445E-07 7.067E-OB 4.346E-OB 2.986E-OB ESE 6.207E-07 5.158E-07 7.670E-07 1.523E-06 1.306E-06 5.046E-07 1.776E-07 1. 132E-07 7.012E-OB 4.846E-OB SE 5.030E-07 2. 142E-07 1. 159E-07 3.850E-07 6. 171E-07 2.946E-07 2.383E-07 1.554E-07 9.668E-OB 6.701E-OB SSE 4.577E-07 1.843E-07 7.887E-OB 5.214E-OB 4. 117E-08 2.911E-OB 2.219E-07 1.724E-07 1.072E-07 7.416E-OB 77

AHENDHENT NO. 9 JANUARY 1992 TABLE 3-10 (CONTD)

C) 8.000 DAY DECAY, UNDEPLETED CMI/O (SEC/HETER CUBED) FOR EACH SEGMENT SEGHENT BOUNDARIES IN HILES FROH THE SITE DIRECTION .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 FROM SITE 5 3.793E-07 1.434E-07 5.872E-OB 3.765E-OB 2.919E-OB 1.879E-OB 1.925E-07 1.497E-07 9.326E-OB 6.496E-OB SSM 2.479E-07 9.089E-OB 3.705E-OB 2.403E-OB 1.880E-OB 1.325E-OB 1.561E-07 1.226E-07 7.691E-OB 5.388E-OB SM 1.572E-07 6.070E-OB 3 ~ 115E-08 2.387E-OB 1.896E-OB 3.473E-OB 9.189E-OB 6.223E-OB 3 '71E-08 2.694E-OB MSM 6.375E-OB 2.776E-OB 1.416E-OB 1.057E-OB 8.356E-09 1.572E-OB 4.792E-OB 3.295E-OB 2.035E-OB 1.407E-OB M 6.471E-OB 2.914E-OB 1.449E-OB 1.047E-OB 8.037E-09 4.713E-09 2.534E-OB 1.922E-OB 1. 182E-08 8. 138E-09 MNM 1.255E-07 5.587E-OB 2.901E-OB 2. 171E-08 1.709E-OB 1.261E-OB 4.452E-OB 3.233E-OB 1.960E-OB 1.335E-OB KM 2.228E-07 8.309E-OB 3.451E-OB 2.300E-OB 1.837E-OB 1.471E-OB 8.579E-OB 6.505E-OB 4.009E-OB 2.769E-OB NNM 4.947E-07 1.686E-07 6.558E-OB 4.219E-OB 3.996E-OB 2.820E-OB 1.654E-07 1.272E-07 7.938E-OB 5.547E-OB N 5.785E-07 1.917E-07 7.566E-OB 4.927E-OB 3.863E-OB 2.032E-07 2.372E-07 1. 154E-07 7. 127E-08 4.939E-OB NNE 4.769E-07 1.602E-07 6.423E-OB 4.232E-OB 4.105E-OB 2.728E-07 1.696E-07 8. 150E-08 4.985E-OB 3.425E-OB 3.220E-07 1.141E-07 4.688E-OB 3.977E-OB 1.366E-07 2.947E"07 1.433E-07 6.805E-OB 4.121E-OB 2.806E-OB EIIE 4.056E-07 2.988E-07 3.849E-07 7.340E-07 6.588E-07 2.951E-07 1.033E-07 4.759E-OB 2.806E-OB 1.864E-OB E 4.072E-07 3.375E-07 4.406E-07 8. 152E-07 6.738E-07 3.000E-07 1.042E-07 4.785E-OB 2.822E-OB 1.877E-OB ESE 5.997E-07 5.068E-07 6.926E-07 1.240E-06 1.053E-06 3.916E-07 1.247E-07 7.545E-OB 4.463E-OB 2.978E-OB SE 4.883E-07 2.075E-07 1.122E-07 3.874E-07 6.185E-07 2.852E-07 2.217E-07 1.413E-07 8.648E-OB 5.940E-OB SSE 4.476E-07 1.796E-07 7.640E-OB 5.064E-OB 4.029E-OB 2.929E-OB 2.179E-07 1.669E-07 1.027E-07 7.085E-OB 78

AMENDMENT NO. 9 JANUARY 1992 ,

TABLE 3-10 (COHTD)

D) REACTOR BUILDING 0/Q RELATIVE DEPOSITION PER UHIT AREA (H**-2) BY DOWNWIND SECTORS SEGHEHT BOUNDARIES IH HILES DIRECTION .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 FROM SITE 5 4.044E-09 1.146E-09 3.717E-IO 1.874E-IO 1. 127E-10 4 '35E-11 3.868E-11 2.283E-11 1.219E-11 7.548E-12 SSW 2.643E-09 7.296E-IO 2.324E-IO 1. 165E-10 7 ~ OOOE-11 2 '16E-11 2 '63E-11 1 '96E-11 8 '26E-12 5.278E-12 SW 1.429E-09 4.068E-IO 1.407E-IO 6.799E-11 4.016E-11 3. 192E-11 2.386E-II 9.555E-12 5. 104E-12 3 . 160E-12 WSW 4.407E-IO 1.347E-IO 4.908E-II 2.400E-II 1.423E-II 1.224E-II 9.617E-12 3.865E-12 2.064E-12 1.278E-12 W 5.587E-10 1.780E-10 6.665E-II 3.253E-11 1.929E-11 7.707E-'2 6. 116E-12 3.617E-12 1.932E-12 1. 196E-12 WHW 1. 110E-09 3.459E-IO 1.262E-IO 6.186E-11 3.674E-11 2.357E-11 1.640E-II 6.963E-12 3.719E-12 2.302E-12 HW 2.199E-09 6.289E-10 2.051E-IO 1.036E-IO 6.242E-II 2.625E-11 2.528E-11 1.520E-II 8. 117E-12 5.025E-12 HHW 5. 161E-09 1.411E-09 4.463E-IO 2.231E-IO 1.382E-10 5.828E-11 5.329E-II 3. 186E-11 1.702E-II 1.053E-11 N 7.312E-09 1.932E-09 6.001E-IO 2.970E"10 1.774E-IO 1.307E-10 8.654E-11 3.430E-11 1.832E-11 1. 134E-11 NNE 6.688E-09 1.751E-09 5.437E-IO 2.675E-IO 1.637E-10 1.566E-IO 6.754E-II 2.677E-11 4.430E-11 8.851E-12 4.654E-09 1.223E-09 3.808E-10 1.931E-10 2.225E-IO 1.592E-IO 4.683E-11 1.873E-11 1.000E-11 6. 191E-12 EHE 4 .842E-09 1.277E-09 6. 137E-10 5.265E-IO 3 .056E-10 1.189E-IO 3 .440E-11 1.364E-II 7 .286E-12 4 .51 IE-12 E 4. 004E-09 1. 121E-09 6. 044E-10 5. 617E-10 3. 268E-10 1. 248E-10 3. 590E-11 1. 441E-11 7. 695E-12 4. 763E-12 ESE 6.270E-09 1.764E-09 9.704E-10 9.016E-IO 5.207E-IO 2.008E-10 5.788E-11 2.316E-I I 1.237E-11 7.659E-12 SE 5.027E-09 1.477E-09 5.218E-10 5.481E-10 6.894E-IO 2.662E" 10 7.839E-11 3.142E-11 1.678E-11 1.039E-I I SSE 4.426E-09 1.321E-09 4.452E-IO 2.267E-IO 1.366E-IO 5.692E-11 4.873E-II 2.896E-I I 1.547E-I I 9.573E-12 79

AMENDMENT NO. 9 JANUARY 1992 TABLE 3-11 TURBINE OR RAOWASTE BUILOIHG X/Q ANO 0/Q VALUES A) HO OECAY, UHOEPLETEO CMI/Q (SEC/METER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROM THE SITE DIRECT IOH .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 FROM SITE S 2.782E"05 7.806E-06 2.832E-06 1.567E-06 1.037E-06 5.081E-07 2. 113E-07 1.153E-07 7.771E-OB 5.794E-OB SSN 2.117E-OS 6.000E-06 2.195E-06 1.220E-06 8.099E-07 3.989E-07 1.671E-07 9.172E-OB 6.199E-OB 4. 631E-08 SW 1.211E-OS 3.404E-06 1.236E-06 6.834E-07 4.521E-07 2.214E-07 9.199E-OB 5.019E-OB 3.381E-OB 2.520E-OB NSM 6.468E-06 1.831E-06 6.680E-07 3.702E-07 2.451E-07 1.202E-07 5.001E-OB 2.729E-OB 1.837E-OB 1.369E-OB W 4.034E-06 1.113E-06 3.982E-07 2.186E-07 1.439E-07 6.994E-OB 2.873E-OB 1.555E-OB 1.043E-OB 7.751E-09 MHM 7.812E-06 2. 127E-06 7.518E"07 4.096E-07 2.682E-07 1.292E-07 5.239E-OB 2.809E-OB 1.873E-OB 1.387E-OB HM 1.386E-05 3.830E-06 1.370E-06 7.517E-07 4.944E-07 2.397E-07 9.809E-OB 5.290E-OB 3.538E-OB 2.624E-OB NNV 2.549E-OS 7.081E-06 2.548E-06 1.402E-06 9.242E-07 4.498E-07 1.849E-07 1.001E-07 6.703E-OB 4.976E-OB N 2.640E-OS 7.275E-06 2.599E-06 1.424E-06 9.356E-07 4.528E-07 1.845E-07 9.915E-OB 6.615E-OB 4.897E-OB 2.061E-OS 5.617E-06 1.986E-06 1.082E-06 7.085E-07 3.410E-07 1.379E-07 7.372E-OB 4.906E"08 3.626E-OB 1.800E-OS 4.929E-06 1.749E-06 9.543E"07 6.251E-07 3.009E-07 1.217E-07 6.502E-OB 4.323E-OB 3.193E-OB EHE 1. 715E-05 4. 677E-06 1; 656E-06 9. 030E-07 5.914E-07 2.848E-07 1.152E-07 6. 164E-08 4. 103E-08 3. 032E-08 E. 1.821E-05 4. 961E-06 1. 751E-06 9. 521E-07 6. 221E-07 2.982E-07 1. 198E-07 6.368E-OB 4. 221E-08 3. 111E-08 ESE 2.834E-OS 7.730E-06 2.730E-06 1.484E-06 9.699E-07 4.651E-07 1.870E-07 9.951E-OB 6.602E-OB 4.868E-OB SE 3.509E-OS 9.697E-06 3.466E-06 1.899E-06 1.247E-06 6.035E-07 2.459E-07 1.322E-07 8.823E-OB 6.534E-OB SSE 3.628E-OS 1.013E-OS 3.656E-06 2.015E-06 1.330E-06 6.485E-07 2.677E-07 1.453E-07 9.755E-OB 7.255E-OB 80

AHENDMENT NO. 9 JANUARY 1992 TABLE 3-11 (CONTD)

8) 2.260 DAY DECAY; UNDEPLETED CHI/O (SEC/HETER CUBED) FOR EACH SEGHENT SEGHENT BOUNDARIES IN HILES FROH THE SITE DIRECTION .5-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 FROH SITE 5 2.763E-OS 7.701E-06 2.766E-06 1.515E-06 9.933E-07 4.745E-07 1.848E-07 9.291E-OB 5.799E-OB 4.022E-OB SSM 2. 104E-05 5.933E-06 2.152E-06 1.186E-06 7.812E-07 3.766E-07 1.492E-07 7.615E-OB 4.802E-OB 3.355E-OB SM 1.203E-OS 3.361E-06 1.208E-06 6.623E-07 4.343E-07 2.077E-07 8.127E-OB 4. 111E-08 2.581E-OB 1.801E-OB MSM 6.405E-06 1.797E-06 6.466E-07 3.537E-07 2.313E-07 1.098E-07 4.210E-OB 2.083E-OB 1.286E-OB 8.86SE-09 M 4.001E-06 1.095E-OB 3.867E-07 2.097E-07 1.363E-07 6.412E-OB 2.424E-OB 1. 183E-08 7.228E-09 4.937E.09 MNW 7.732E-06 2.084E-06 7.250E-07 3.891E-07 2.510E-07 1.163E-07 4.274E-OB 2.033E-OB 1.222E-OB 8.256E-09 NM 1.376E-05 3.776E-06 1.337E-06 7.256E-07 4.724E-07 2.229E-07 8.513E-OB 4.216E-OB 2.614E-OB 1.810E-OB NNM 2.537E-05 7.013E-OB '2.506E-OB 1.369E-OB 8.966E-07 4.286E-07 1.683E-07 8.590E-OB 5.449E-OB 3.842E-OB N 2.626E-OS 7. 199E-06 2.551E-06 1.387E-06 9.044E-07 4.289E-07 1.659E-07 8.349E-OB 5.244E-OB 3.66BE-OB NNE 2.047E-OS 5.544E-06 1.941E-06 1.047E-OB 6.792E-07 3.187E-07 1.208E-07 5.960E-OB 3.687E-OB 2.548E-OB

1. 784E-05 4.844E-06 1. 696E-06 9. 137E-07 5.910E-07 2. 753E-07 1.025E-07 4.952E-OB 3. 013E-08 2. OSSE.OB ENE 1. 701E-05 4. 603E-06 1. 610E-06 8. 673E" 07 5. 613E-07 2. 621E-07 9.803E-OB 4. 756E-08 2.901E-OB 1.980E" 08 E 1.808E-OS 4.891E-06 1.707E-06 9.184E-07 5.938E-07 2.769E-07 1.037E-07 5.047E-OB 3.090E-OB 2.115E-OB ESE 2.813E"OS 7.623E-06 2.663E-06 1.434E-06 9.278E-07 4.336E-07 1.634E-07 8.016E-OB 4.941E-OB 3.402E-OB SE 3.486E-05 9.574E-06 3.389E-06 1.840E-06 1.197E-06 5.654E-07 2.165E-07 1.075E-07 6.672E-OB 4.615E-OB SSE 3.600E-OS 9.979E-06 3. 562E-06 1.942E-06 1.268E-06 6.016E-07 2.313E-07 1. 150E-07 7. 125E-08 4. 919E-OB 81

AMENDMENT NO. 9 JANUARY 1992 TABLE 3-11 (CONTD)

C) 8.000 DAY DECAY, DEPLETED CHI/O (SEC/HETER CUBED) FOR EACH SEGMENT SEGMENT BOUNDARIES IN MILES FROH THE SITE DIRECTION .5-1 '1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 '0-50 FROM SITE 5 2.487E-OS 6.658E-06 2.286E-06 1.213E-06 7.751E-07 3.540E-07 1.275E-07 5.998E-OB 3.588E-OB 2.411E-OB SSM 1. 892E-05 5. 121E-06 1. 774E-06 9.460E-07 6 067E-07

~ 2. 788E-07 1. OISE-07 4.823E-OB 2 ~ 906E-08 1. 964E-08 SM 1.082E-OS 2.904E-06 9.977E-07 5.294E-07 3.382E-07 1.545E-07 5.566E-OB 2.623E-OB 1.571E-OB 1.058E-OB MSM 5.777E-06 1.559E"06 5.378E-07 2.856E-07 1.824E-07 8.320E-OB 2.980E-OB 1.391E-OB 8.267E-09 5.523E-09 M 3.605E-06 9.486E-07 3.209E-07 1.688E-07 1.072E-07 4.847E-OB 1.714E-OB 7.924E-09 4.679E-09 3. 111E-09 MNM 6.978E-06 1.811E-06 6.046E-07 3.155E-07 1.992E-07 8.908E-OB 3.092E-OB 1.406E"08 8.205E-09 5.403E-09 NW 1.239E-05 3.267E-06 1.106E-06 5.816E-07 3.693E-07 1.668E-07 5.900E-OB 2.734E-OB 1.619E-OB 1.080E-OB NNM 2.280E-OS 6.047E-06 2.061E-06 1.089E-06 6.935E-07 3. 153E-07 1.129E-07 5.307E-OB 3 . 181E-08 2 . 145E-08 N 2.362E-05 6.211E-06 2. 101E-06 1. 105E-06 7.013E-07 3. 169E-07 1. 123E-07 5.225E-OB 3. 110E-08 2.086E-OB NNE 1.843E"05 4.793E-06 1.604E-06 8.381E-07 5.298E-07 2.377E-07 8.323E-OB 3.832E-OB 2.263E-OB 1.507E-OB 1.608E-OS 4.201E-06 1.409E-06 7.367E-07 4.655E-07 2.085E-07 7.255E-OB 3.311E-OB 1.939E-OB 1.282E-OB EIIE 1.532E-OS 3.988E-06 1.335E-06 6.977E-07 4.409E-07 1.976E-07 6.893E-OB 3.155E-OB 1.853E-OB 1.227E-OB E 1.628E-OS 4.232E-06 1.413E-06 7.366E-07 4.646E-07 2.075E-07 7.206E-OB 3.291E-OB 1.932E-OB 1.281E-OB ESE 2.534E-05 6.595E-06 2.203E-06 1. 149E-06 7.248E-07 3.241E-07 1.128E-07 5.170E-OB 3.045E-OB 2.024E-OB SE 3. 137E-05 8.274E-06 2.799E-06 1.471E-06 9.331E-07 4.210E-07 1.487E-07 6.892E-OB 4.086E-OB 2.729E-OB SSE 3.242E-OS 8.636E-06 2.949E-06 1.558E-06 9.928E-07 4.510E-07 1.609E-07 7.503E-OB 4.461E-OB 2.984E-OB 82

AMENDMENT NO. 9 JANUARY 1992 TABLE 3-11 (CONTO)

0) TURBINE OR RAOMASTE DEPOSITION, 0/Q.

RELATIVE DEPOSITION PER UNIT AREA (M"*-2) BY OOMNMINO SECTORS SEGMENT BOUHOARIES IH MILES 0 I RECT ION .5-1 1-2 2-3 3-4 4-5 5-10 10-20 '20-30 - 30-40 40-50 FROM SITE 5 2.664E-OB 5.457E-09 1.425E-09 6.398E-10 3 '20E-10 1.392E-10 4.027E-11 1.596E-11 8.523E-12 5.275E-12 5SM 1.853E-OB 3. 796E-09 9.909E-10 4.450E-IO 2. 518E-10 9. 682E-11 2.801E-11 1. 110E-11 5.928E-12 3. 669E-12 SW 1.160E-OB 2.375E-09 6. 201E-10 2. 785E-10 1. 575E-10 6. 058E-11 1. 753E-11 6. 947E-12 3. 710E-12 2. 296E-12 MSM 4. 652E-09 9. 529E-10 2.488E-10 1. 117E-10 6.321E-II 2.431E-11 7. 032E-12 2. 787E-12 1. 488E-12 9. 212E-13 W 4. 254E-09 8. 714E-10 2. 275E-10 1. 022E-10 5. 780E-11 2. 223 E-11 6. 430E-12 2. 549E-12 1. 361E-12 8. 424 E-13 WH1I 8.379E-09 1. 716E-09 4. 481E-10 2. 012E-10 1. 138E-10 4.378E-11 1. 266E-11 5. 020E-12 2. 681E-12 1. 659E-12 NM 1.761E-OB 3.608E-09 9.419E-10 4.230E-10 2.393E-10 9.203E-11 2.662E-11 1.055E-11 5.635E-12 3.488E-12

. NHU 3.707E-OB '.593E-09 1.982E-09 8.903E-IO 5.036E-IO 1.937E-10 5.603E-11 2.221E-11 1.186E-11 7.340E-12 N 4.270E-OB 8.746E-09 2.283E-09 1.025E-09 5.801E-IO 2.231E-10 6.454E-11 2.558E-11 1.366E-11 8.455E-12 HHE 3.448E-OB 7.062E-09 1.844E"09 8.280E-10 4.684E-10 1.801E-10 5.211E-11 2.065E-11 1.103E-11 6.827E-12 2.465E-OB 5.050E-09 1.318E-09 5.921E-10 3.349E-IO 1.288E-10 3.726E-11 1.477E-11 7.887E-12 4.881E-12 EHE 2. 235E-08 4. 579E-09 1. 195E-09 5.368E" 10 3. 037E-10 1. 168E-10 3.379E-I I 1.339E-11 7. 151E-12 4.426E-12 E 2.363E-OB 4.841E-09 1.264E-09 5.676E-10 3.211E-10 1.235E-10 3.572E-11 1.416E-11 7.560E-12 4.679E-12 ESE 3.810E-OB 7.804E-09 2.037E-09 9.150E-IO 5.176E-IO 1.991E-IO 5.759E-11 2.282E-11 1.219E-11 7.544E-12 SE 4.168E-OB 8.537E-09 2.229E-09 1.001E-09 5.663E-IO 2.178E-10 6.300E-I I 2.497E-11 1.333E-11 8.253E-12 SSE 3.672E-OB 7.521E-09 1.963E-09 8.818E-10 4.988E-10 1.918E-10 5.550E-ll 2.200E-ll 1.175E-11 7.270E-12 83

AHENDHENT NO. 9

, JANUARY 1992 Table 3-13 CHARACTERISTICS OF WNP-2 GASEOUS EFFLUENT RELEASE POINTS Reactor Radwaste Turbine

~Bui1din . ~Bui1din ~Bui1din Height of- release point

,above ground level (m) 70.6m 31.1 27.7 Annual average rate of air flow from release point (m /sec) 44.8 38.7 125.6 Annual average heat flow from release'oint 1.06 x 10 2.9 x 10' 9.1 x 10 (cal/sec)

Type and size of Duct Louver houses 4 Exhaust fans release point (m) 1.14 x 3.05 1.4 x 2.4 x 0.8 1.45 x 2.01 Each Each Effective vent area (m') 3.48 2 x 2.7 3 x 2.91 Vent velocity (m/sec)* 12.9 2 x 525 cfm** 14.4 Effective diameter (m) 1.0 (mr' area)

Building height (m) 70.1 70.1 70.1

Reactor Building exhaust in vertical direction. Radwaste and Turbine Building exhaust in horizontal plane.

- FSAR Drawing 6-41, 525 cfm x 2 out of 3, will run at any one time.

AHENDHENT NO. 15 OCTOBER 1993 Table 3-14 REFERENCES FOR VALUES LISTED IN TABLES 3-'8 and 3-9 Reference 1 U.S. Hap Reference 2 Site Specific Reference 3 'egulatory Guide 1. 109, Revision 1, Table E-15 Reference 4 Section 2.3, WNP-2 FSAR, Table 2.3-1 Reference 5 Section 2.3, WNP-2 FSAR, page 2.3-3 Reference 6 WNP-1 & WNP-2 Emergency Preparedness Plan Table 12. 1, Permanent Population Distribution, Rev 5, Feb. 88 Reference 7 1986 50-Hile Land Use Census, WPPSS REHP Reference 8 WNP-2 Effluent Analysis for Applicable Time Period Reference 9 Health Physics Calculation Log No. 93-2 Reference 10 NUREG/CR-2919 XOgDOg: Computer Program For The Heteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, September 1982.

85

AMENDMENT NO. 9 JANUARY )992 4

Table 3-15 DESIGN BASE PERCENT NOBLE GAS 30-MINUTE DECAY *

~lsoto e Percent of Total Activit Kr-83H 2.9 Kr-85M 5.6 Kr-85 0 Kr-87 15 Kr-88 18 Kr-89 0.2 Xe-131M 0.02 Xe-133M 0.3 Xe-133 &.2 Xe-135M 6.9 Xe-135 22 Xe-137 0.7 Xe-138 21

From Table 11.3-1 WNP-2 FSAR 86

AMENDMENT NO. 9 JANUARY 1992 TABLE 3-16

. ANNUAL DOSES AT TYPICAL LOCATIONS Source: WNP-2 Gaseous Effluent Whol e Thyroid Distance Occupancy Body Dose Dose Location Miles hrs r mrem r mrem r BPA Ashe Substation 0.5 N 2080 1. 1E+00 1.7E+00 DOE Train 0.5 SE* 78 6.7E-02 1.0E-01 Wye Burial Site 0.5 WNW 4.1E-03 6.5E-03 WNP-1 1.2 ESE 2080 3.8E-02 1.3E-01 WNP-4 1.0 ENE 2080 7.0E-02 1.1E-01 WNP-2 Visitor Center 0.08 ESE 8.6E-02 1.3E-01 Taylor Flats** 4.2 ESE 8760 3.1E-02 5.2E+00 Site Boundary*** 1.2 SE 8760 1. 1E+00 1.7E+00

The sector with the highest X/g values (within 0-0.5 mile radius) was used.

- Closest residential area representative of maximum individual dose from plume, ground, ingestion, and inhalation exposure pathways. Included for comparison.

- - Assumed continuously occupied. Actual occupancy is very low.

Doses from Inhalation and Ground Exposure pathways. No food crops.

87

-

AHENDHENT NO. 9 JANUARY 1992 TABLE 3-17 ANNUAL OCCUPIED AIR DOSE AT TYPICAL'LOCATIONS Location ~d Annual Beta Air dose Annual Gamma Air mrad Dose BPA Ashe Substation 8.9E-01 1.5E+00 DOE Train 5.3E-02 9.2E-02 Wye Burial Site 3.2E-03 5.7E-03 WNP-1 3.3E-02 2.8E-02 WNP-4 5.3E-02 8.5E-02 WNP-2 Visitor Center 7.0E-02 1.2E-01 Taylor Flats* 2.3E-02 1.4E-02 Site Boundary 8.7E-01 1.5E+00

Closest residential area.

88

AMENDMENT NO. 9 JANUARY 1992 Gorurns WNP.2 eer Pump House WN P.114 Pump House BentOn Swttsntnrt Station

~ wm M~ nm ew fl North r~ Oo Site Boundary b~ p, Qst WHY bg 0 WNA1 o Route 4 IL*Asne susstad on I a

>>

,o 0" WNP-2 Plant Support od'econdary Facility Access Wye Burtat ~

Ground stet Emergency tower ~R~ Opersttons Factuty P+

+@

Exctuston Botmdary SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS Figure 3-1 89

AMENDllENT NO. 9 JANUARY 1992 Main Condenser Air Ejector Off-Gas Condensers Preheaier s Water Off-Gas Off-Gas Separator Condenser Recombiners Gycol Off-Gas Off-Gas Cooler Mofstur e Pre-Filters Condenser Separator s Charcoal Coolers Dryer s Adsorber s Post Treatment Monitor After Elevated Filters Release SIMPLIFIED BLOCK DIAGRAM OF OFF-GAS TREATMENT SYSTEM Figure 3-3 91

AMENDMENT NO. 9 JANUARY 1992 Plenum-Monitors pH~tr Refueling Pump Rooms Main Steam Tunnels Plenum'o Exhaust Elevated Release Plenum Monitors Cond. Pump Rm Exhaust To Turb. Opr. Oeck Plenum Atmosphere Condenser Area RM Hot Mach. Shop Chem. Labs Filter To Oemin. Room Units Atmosphere Radwaste Proc. Area Control Room SIMPLIFIEO BLOCK DIAGRAM OF GASEOUS WASTE SYSTEM Figure 3-2 90

~

I s l ~ ~ ~ ~

AMENDMENT NO. 9 1992 'ANUARY 4.0 COMPLIANCE WITH 40 CFR 190 h

'4.l Re uirement For 0 erabilit Requirement for Operability 6.2.4. 1 (3. 11.4) states, "The annual (calendar year) dose or dose commitment to any Hember of the Public, due to release of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems."

4.2 ODCH Hethodolo for Determinin Dose and Dose Commitment from Uranium Fuel C cle Sources The annual dose or dose commitment to a Member of the Public for the uranium fuel cycle sources is determined as:

a) Dose to the total body due to the release of radioactive materials in .

liquid effluents..

b) Dose to any organ due to the release of radioactive materials in liquid effluents. I c) Air doses due to noble gases released in gaseous effluents.

d) Dose to any organ due to the release of radioiodines, tritium and radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents.

e) Dose due to direct radiation from the plant.

The annual dose or dose commitment to a Member of the Public from the uranium fuel cycle sources is determined whenever the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceed twice the limits in Requirement for Operability 6.2. 1.2.a, 6.2. 1.2.b, 6.2.2.2.a, 6.2.2.2.b, 6.2.2.3.a, or 6.2.2.3.b (3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a,

3. 11.2.2.b, 3. 11.2.3.a, or 3. 11.2.3.b). Direct radiation measurements will also be made to determine if the limits of Requirement for Operability 6.2.4.1 (3. 11.4) have been exceeded.

92

, AHENDHENT NO, 9 JANUARY 1992 uid Effluents

~

4.2.1

~ Total Dose from Li The annual dose to a Hember of the Public from liquid effluents will be determined using NRC LADTAP computer code, and methodology presented by Equation (5) in Section 2.4. It is assumed that'ose contribution pathways to

'

Member of the Public do not exist for areas within the site boundary.

4.2.2 Total Dose from Gaseous Effluents The annual dose to a Hember of the Public from gaseous effluents will be determined using NRC GASPAR computer code, and methodology presented by Equations (10), (11) and (13) in Section 3.4. Appropriate atmospheric dispersion parameters will be used.

4.2.3 Direct Radiation Contribution The dose to I

a Member of the Public due to direct radiation from the reactor plant will be determined using thermoluminescent dosimeters (TLDs) or may be calculated. TLDs are placed at sampl.e locations and analyzed as per Table 5-1. The direct radiation contribution will be documented in the Radioactive Effluent Release Report submitted 60 days after January 1 of each year.

TLD stations 1S-16S are special interest stations and will not be used for direct radiation dose determinations to a Member of the Public.

5. 0 RADIOLOGICAL ENVIRONMENTAL MONITORING Radiological environmental monitoring is intended to supplement radiological effluent monitoring by verifying that measurable concentrations of radioactive materials and levels of radiation in the environment are not greater than expected based on effluent measurement and dose modeling of environmental exposure pathways. The Radiological Environmental Monitoring Program (REHP) for WNP-2 provides for measurements of radiation and radioactive materials in those exposure pathways and for those radionuclides for which the highest potential dose commitment to a Member of the Public would result due to plant operations.

93

AMENDMENT NO, 9 JANUARY 1992 The WNP-2 REHP is designed to conform to regulatory guidance provided

~ ~

by Regulatory Guide 4;1, 4.8 and the Radiological Assessment

~

Branch Technical Position (BTP), taking into consideration certain site specific character-

~

istics. The unique nature, of the WNP-2 site on Federally owned and admin-istered land (Hanford Reservation) dedicated to energy facilities, research, waste management and as 'a natural reserve, forms the basis for many of the site specific parameters. Amongst the many site specific parameters con-sidered is demographic data such as:

1) No significant clusters of population including schools, hospitals, business facilities or primary public transportation routes are located within 8 km (5 mile) radius of the plant.

2) No private residences are located on the Hanford Reservation.

3) The closest resident is east of the Columbia River at a distance of approximately 4 miles.

Additional site information is available in the

~

WNP-2 Environmental Report, Operating License Stage. ~

Radiological environmental monitoring activities implemented by PPH 1. 11. 1 "Radiological Environmental Monitoring Program (REHP) Implementation Proce-dure", as detailed in the following sections, meet or exceed the criteria of the REHP plan as specified by Requirement for Operability, 6 .3 . 1. I (3 . 12. 1) .

5. 1 Radiolo ical Environmental Monitorin Pro ram REMP Environmental samples for the REHP are collected in accordance with Table 5-1.

This table provides a detailed outline of the environmental sampling plan including both Requirement for Operability and non Requirement for Operability items by sample type, sample location code, sampling and collection frequency, and type and frequency of analysis of samples collected within exposure pathway. Deviations from the sampling frequency detailed in Table 5-1 may occur due to circumstances such as hazardous conditions, malfunction of automatic sampling equipment, seasonal unavailability, or other legitimate reasons. When sample media is unobtainable due to equipment malfunction,

AHENDMENT NO. 16 DECEHBER 1993 special actions per program instruction shall be taken to ensure that corrective action is implemented, prior to the end of -the next sampling period.

In some cases, alternate sample collection may be substituted for the missing specimen. All deviations from the sampling plan Requirement for Operability items detailed in the sampling plan, Table 5-1, shall be documented and reported in the Annual Radiological Environmental Operating Report in accordance with PPH 1. 10.2, "Routine or Periodic Reports Required by Regulatory- Agencies", Regulatory Guide 4.8 and BTP.

In the event that it becomes impossible or impractical to continue sampling a media of choice at currently established location(s) or time, an evaluation shall be made to determine a suitable alternative media and/or. location to provide appropriate exposure pathway evaluations. The evaluation and any substitution made shall be implemented in the sampling program within 30 days of identification of the problem. All changes implemented in the sampling program due to unavailability of samples shall be fully documented in the next Radioactive Effluent Release Report and ODCH per PPH 1. 10. 1, "Reportable !

Events and Occurrences Required by Regulatory Agencies". Revised sampling plan table(s) and figure(s) reflecting the new locations and/or media shall be included with the documentation.

WNP-2 sampling stations are described in Table 5-2. Each station is identified by an assigned number or alphanumeric designation, meteorological sector (16 different, 22-1/2'ompass sections) in which the station is located, and radial distance from WNP-2 containment as estimated from map positions. Also included in Table 5-2 is information identifying the type(s) of samples collected at each station and whether or not the specific sample type satisfies a Requirement for Operability criteria. Figures 5-1 and 5-2 depict the geographical locations of each of the sample stations listed in Table 5-2.

5.2 Land Use Census A Land Use Census shall be conducted in accordance with the requirements of Requirement for Operability 6.3.2. 1 (3. 12.2). It shall identify within a distance of 8 km (5 miles) in each of the 16 meteorological sectors, the 95

AMENDMENT NO. 16 DECEMBER 1993 location of the nearest milk .animal, the nearest residence and the nearest

,garden of greater than 150m'500 ft ) producing broad leaf vegetation. Field activities pertaining to the Land Use Census (LUC) will be initiated during the growing season and completed no later than September 30 each year. 'he information obtained during the field survey is used along with other demographic data to assess population changes in the unrestricted area that might require modifications in the sampling plan to ensure adequate evaluation of dose or'dose commitment.

The results of the Land Use Census will be submitted no later than October 31 of each year for evaluation of maximum individual doses or dose commitment.

All changes, such as a location yielding a greater estimated dose or dose commitment or different location with a 20'ercent greater estimated dose or dose commitment than a currently sampled location, will be reported in the next Radioactive Effluent Release Report in accordance with PPH 1. 10.2 and Requirement for Operability 6.3.2. 1 (3. 12.2). The REHP plan, ODCH, will be changed to reflect new sampling location(s). The new sampling location(s) will be added to the REHP within 30 days.

The best available census information, whether obtained by aerial survey, door-to-door survey, or consultation with local authorities, shall be used to complete the Land Use Census and the census results shall be reported in the Annual Radiological Environmental Operating Report, in accordance with PPH

1. 10.2'and Technical Specification requirements.

5.3 Laborator Intercom arison Pro ram Analysis of REHP samples is contracted to a provider of radiological analytical services. By contract, this analytical service vendor is required to conduct all activities in accordance with Regulatory Guides 4. 1, 4.8, and

4. 15 and to include in each quarterly report, actions pertinent to their participation in the Environmental Protection Agency's (EPA) Environmental Radioactivity Laboratory Intercomparison Studies (Crosscheck) Program. A precontract award survey and annual audit at the contractor's facility ensure that the contractor is participating in the Crosscheck Program, as reported.

96

AMENDMENT NO. 10 JANUARY 1992 The results of the contractor's analysis of Crosscheck samples shall be included in the Annual Radiological Environmental Operating Report in accordance with PPH 1. 10.2 and Requirement for Operability 6.3.3. I (3. 12.3).

Besides the vendor's required participation in the EPA's Crosscheck Program, the Department of Health (DOH) of the State of Washington oversees an analytical program for the Energy Facility Site Evaluation Council (EFSEC) to provide an independent test of WNP-2 REMP sample analyses. The WNP-2/DOH split samples are analyzed by Washington State'.s Office of Public Health Laboratories and Epidemiology, Environmental Radiation Laboratory (ERL). The State's ERL participates in the EPA Crosscheck Program, as well as other federal participatory analytical quality assurance programs. The results of the ERL analysis and EPA Crosscheck data are included in an annual report, "Environmental Radiation Program, Environmental Health Surveillance, State of Washington" and is available for comparison with the WNP-2 data.

The Supply System participates in the International Intercomparison of Environmental Dosimeter Program. Results of this intercomparison program are reported in the REHP Annual Report, when available.

5.4 Re ortin Re uirements WNP-2 radiological environmental monitoring program activities are presented I

annually per PPH 1. 10.2 in the Annual Radiological Environmental Operating Report (AREOR). The approved report is submitted to the Administrator, Region V Office of Inspection and Enforcement, with copies to the Director, Office of Nuclear Reactor Regulation, and the State of Washington Energy Facility Site Evaluation Council (EFSEC) and Radiation Control Section, DOH, by Hay 1 of each year for program activities conducted the previous calendar year. The period of the first operational report begins with the date of initial criticality.

The annual report is to include the following types of information: a tabulated summary; interpretations and analyses'of trends for results of radiological environmental surveillance activities for the report period, including comparisons with operational controls, preoperational studies, and previous environmental surveillance reports as appropriate; an assessment of 97

AMENDMENT NO. 9 1992 'ANUARY e observed impacts of plant operation on the environment; a brief scription of the, radiological environmental monitoring program; maps representing sampling station locations, keyed to tables of distance and direction from reactor containment; results of the Land Use Census; and the results of analytical laboratory participation in the EPA's Crosscheck Program. The tabulated summary shall be presented in a format represented in Table 5-3. A supplementary report is required if all analytical results are not available for inclusion in the annual report within the specified time frame. The missing data shall be submitted as soon as possible upon receipt of- the results. Along with the missing data, the supplementary report shall include an explanation as to the cause for the delay in completion of the analysis within the report period.

A nonroutine radiological environmental operating report is required to be submitted within 30 days from the end of any quarter in which a confirmed measured radionuclide concentration in an environmental sample averaged over the quarter sampling period exceeds a reporting level. Table 5-4 specifies the reporting level (RL) for most radionuclides of environmental importance due to potential impact from plant operations. When more than one of the nuclides listed in Table 5-4 is detected in a sample, the reporting level is considered to be exceeded and a nonroutine report required if the following conditions are satisfied:

Concentration I Concentration 2 Reporting Level (I) Reporting Level (2)

For radionuclides other than those listed in Table 5-4, the reporting level is considered to have been exceeded if the potential annual dose to an individual is greater than or equal to the design objective doses of Appendix I, 10 CFR 50. When a nonroutine report on an unlisted (Table 5-4) radionuclide must be issued, it shall include an evaluation of any release conditions, environmental factors, or other aspects necessary to explain the anomalous sample results.

98

AMENDMENT NO. 9 JANUARY 1992 When it can be 'demonstrated, that the anomalous sample result(s) exceeding reporting levels is not the result of plant e'ffluents, 'a nonroutine report does not have to be submitted. ,A full discussion of the sample result'nd subsequent-evaluation or investigation of the anomolous result will be included in'the Annual Radiological 'Environmental Operational Report.

99

TABLE 5-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PLAN Sam le T e~~ Sam le Location Code*

Sampling and Collection Fre uenc

'ype of and Frequency Anal sis'

1. AIRBORNE

a. Particulates and 1, 4-9A, 21, 23, 40, Continuous sampling Particulate: Gross beta',

radioiodine 48 and 57 Weekly collection weekly; gamma isotopic, (5/12) quarterly composite (by location)

Radioiodine: I-131 analysis, weekly

b. Soil'0/7) 9A, 1, 7, 21, and 23 Annually Gamma isotopic', annually strontium-90 when requested" CD CD 101, 118 quarterly, or more Gamma isotopic often, as needed

2. DIRECT RADIATION TLD 1-9A, 10-25, 40-47, quarterly, annually TLD converted to exposure (34/56) 49-51, 53-56, quarterly, annual processing 1S-16S

3. WATERBORNE

a. Surface/ 26, 27, 28 and 29 Composite aliquots, Gamma isotopic, 'gross beta,.

Drinking'3/5 monthly monthly; tritium, quarterly composite strontium-90, iodine-131, when requested'01 Grab -samples weekly or Gamma isotopic, tritium more often, as needed

TABLE 5'-I (contd.)

RADIOLOGICAL ENVIRONHENTAL HONITORING PROGRAH PLAN Sam le T e" Sam le Location Code*

Sampling and Collection Fre uenc

'ype of and Frequency Anal sis'.

Ground water 31, 32, and 52 quarterly Gamma isotopic'nd tritium, (2/3) 'uarterly

c. Sanitation 102 Honthly Gamma isotopic Facility (O/I) Annually Alpha, beta, gamma isotophic Prior to discharge Alpha, beta, gamma isotopic

TABLE 5-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PLAN Sam le T e" Sam le Location Code*

Sampling and Collection Fre uenc

'ype and Frequency of Anal sis'.

Sediment from 33 and 34 Semiannually Gamma isotopic',

shoreline semiannually (I/2)

e. Sediment from 102 Semiannually Gamma isotopic~

sanitation facility (0/1)

4. INGESTION

a. Milk 9B, 36, 62, 64 and 96 Semimonthly during Gamma isotopic and (4/5) grazing season, iodine-131, monthly/

monthly at other times semimonthly strontium-90, when requested

b. Fish 30, 38, or 39 Annually, unless an Gamma isotopic', when (2/2) impact is indicated, sampled-then produce'C, semiannually'onthly

c. Garden 37 and 91 during growing Gamma isotopic, when (2/4) season in the sampled Riverview area of Pasco and a control near Grandview.

Annually for the apple sample collection at Station 91.

Samp e ocations are graphically depicted in Figures 5-1 and 5-2. m OU

'Deviations are permitted if samples are unobtainable due to hazardous conditions, seasonal availability, malfunction of automatic sampling equipment, or other legitimate reasons. All deviatiohs

-I OZ m be documented in the Annual Radiological Environmental Monitoring Report.

will g~M CU C)

Particulate sample filters will be analyzed for gross beta after at least 24-hour decay. If gross beta 4O ~

LD EO activity is greater than 10 times the yearly mean of the control sample, gamma isotopic analysis shall Ol on the individual sample. be'erformed

TABLE 5-1 (contd.)

Gamma isotopic means identification and quantification of gamma-emitting radionuclides that may be to the effluents of the facility. 'ttributable

'TLD refers to thermoluminescent dosimeter. For purposes of WNP-2 REHP, a TLD is a phosphor card (32mm x 45mm x 0.5mm) with eight individual read-out areas (four main dosimeter areas and four back-up dosimeter areas) in each badge case. TLDs used in RBIP meet the requirements of Regulatory Guide 4. 13 (ANSI N545-1975), except for specified energy-dependence response. Correction factors are available for energy ranges with response outside of the specified tolerances., TLD stations 1S-16S and 61 are special interest stations and are not included amongst the 34 routine TLD stations required by Requirement for Operability, Table 7.3. 1. 1-1 (3.12-1).

'Composite samples will be collected with equipment which is capable of collecting an aliquot at time intervals which are short relative to the compositing period. A composite sample is also one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow.

'Station 26, WNP-2 makeup water intake from the Columbia River, satisfies the Requirement for Operability criteria for upstream surface water and drinking water control samples. The discharge water (Station 27) samples are used to fulfill the Requirement for Operability criteria for a downstream sample.

However, they provide very conservative estimates of downstream concentrations. Drinking water samples are not routinely analyzed for I-131 from two week composite, but I-131 analysis will be performed when the calculated dose for the consumption of water is greater th'an 1 mrem per year to the maximum organ. When the gross beta result in drinking water is greater than ten times the mean of the previous month's data for the=:

control location or greater than 8 pCi/liter, Sr-90 analysis shall be performed.

'Hilk samples will be obtained from farms or individual milk animals which are located in sectors with high calculated annual average ground-level D/gs and high dose potential. There are no milk animals located within 5 km of WNP-2. If cesium-134 or cesium-137 is measured in an individual milk sample in excess of 30 pCi/1, then strontium-90 analysis shall be performed.

8 There are no commercially important species in the Hanford reach of the Columbia River. Host recreationally important species in the area are anadromous, primarily salminoids. Three species will normally be collected by electroshock technique in the vicinity of the plant discharge (Station 30). If m electroshocking produces insufficient fish samples, anadromous species may be obtained from Ringold Fish C3 Hatchery (Station 39). Control samples are normally collected from the Snake River, in the vicinity of Ice C: Pl Harbor Dam (salminoids may be obtained through the National Marine Fisheries Service at Lower Granite Dam). C t/)

Three species (same ones obtained from the Columbia River) will be collected from the control location.

any of the analytical results of the Columbia River fish samples are significantly higher than the results If LO C3 of the Snake River samples or the results of previous fish samples, sampling will be conducted semiannually.

TABLE 5-1 (contd.)

'Garden produce will routinely be obtained from farms or gardens using Columbia River water One sample of a root crop, leafy vegetable, and a fruit should be collected each =sample for'rrigation.

period if available. The variety of the produce sample will be dependent on seasonal availability.

10 Soil samples are collected to satisfy the requirements of the Site Certification Agreement (SCA),

WNP-2. If gamma isotopic results for an indicator sample are greater. than ten times the mean of the control station (station 9) data, the sample shall be analyzed for Sr-90.

The fraction in parenthesis under each sample type gives the ratio of the number of Requirement for 11 Operability sample locations to the total number of sample locations for the sample type that is currently included in the overall WNP-2 radiological environmental monitoring program.

TABLE 5-2 WNP-2 REMP LOCATIONS Station Sector Radial Miles' TLD ~AP AI SW DW GW SE HI FI GP ~SO 1 S 1.3 0 X X 2 NNE 0 1.8'.0 3 SE 4 SSE 9.3 0 5 ESE 7.7 0 X S 7.7 0 X WNW 2.7 0 X ESE 4.5 0 0 9A* WSW 30.0 0 0 98* WSW 35.0 X 9C WSW 33.0 0 10 E 3.1 0 11 ENE 3.1 X 12 NNW 6.1 X 13 SW 1.4 0 14 WSW 1.4 0 15 W 1.4 0 16 WNW 1.4 0 17 NNW 1.2 0 GO M Pl C:R KR LD fQ D

5-2 Miles'ABLE (Continued)

Station Sector Radial TLD ~AP AI SW DW GW SE NI F I GP ~SO 18 N 0 19 NE 1.8 20 ENE 1.9 21 ENE 1.5 X 22 E 2.1 23 ESE 3.0 X SE 1.9 0 25 SSE 1.6 26* E 3.2 0 0 27 E 3.2 X 28 SSE 7.4 0 0 29 SSE 11.0 30 E 3.3 31 ESE 32 ESE 1.2 33* ENE 3.6 34 ESE 3.5 36 ESE 7.2 37A SSE 17.0.

37B SSE 16.0 38* E 26.5 38A E 30.0

5-2 Station Miles'ABLE (Continued)

Sector Radial TLD AP/AI SW= DW GW SE MI FI GP 39 NE 4.4 X 4O SE 6.4 0 0 41 SE 5.8 0 42 ESE 5.6 0 43 E 5.8 0

-

44 ENE 5.8 0 45 ENE 4.3 0 46 NE 5.0 0 47 N 0.5 X 48 NE 4.5 0 49 NW 1.2 0 50 SSW 1.2 0 51 ESE 2.1 0 52 N 0.1 0 53 N 7.5 0 54 NNE 6.5 0 oo SSE 5.5 0 56 SSW 7.0 0 57 N 0.8 0

TABLE 5-2 (Continued)

Station Sector Radial Miles'LD ~AP AI SFW SW DM GW SE HI FI V GP SO 62 SE 10.9 X 64 ESE 9.9 0 91 ESE 4.4 96* WSW 36.0 0 101 ENE 0.3 X 102 SE 0.3 X 118 S 0.3 1S(71) N 0.3 X 2S(72) NNE 0.4 X 3S(73) NE 0.5 X 4S(74) ENE 0.4 X 5S(75) E 0.4 X 6S(76) ESE 0.4 X 7S(77) SE 0.5 X 8S(78) SSE 0.7 X 9S(79) S 0.7 X m.

Cs TD mm CZ lD lO Crd M

5-2 Miles'ABLE (Continued)

Station Sector Radial TLD ~AP AI SM DM GW SE HI F I GP ~SO 10S(80) SSM 0.8 X 11S(81) SM 0.7 12S(82) WSW 0.5 13S(83) W 0.5 14S(84) WNW 0.5 15S(85) NW 0.5 16S 86) NNM 0.4

Control location.

X - Sample collect'ed at station that is not included in the Requirement for Operability 0 - Radiological Environmental Requirement for Operability sample collected at station. (non-RETS) a - Estimated from center of MNP-2 Containment from map positions.

b - Included in sampling program to satisfy requirements for Site Certification Agreement with the State of Washington.

AP/AI = Air Particulate and Iodine SW - Surface Water (River Water)

DW = Drinking Mater GM = Ground Water SFW = Sanitation Facility Water SE = Shoreline Sediment HI = Nilk FI = Fish GP = Garden Produce SO = Soil m

Ca D mm CZ C3

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TABLE 5-3 ENVIRONMENTAL RADIOLOGICAL HONITORING PROGRAH ANNUAL

SUMMARY

'ame of Facility Docket No.

Location of Facility Reporting Period (County, State)

Location with Highest Hedium oI Type and All Indicator Annual Hean Number of .

Pathway Sampled Total Number Lower Limit Locations Mean f Control Locations Konroutine (Unit of of Analyses of Detection Mean (f) Distance and Hean (f) RepoIted Measurement Performed Rlrectlcn ~Ren e Air particulates Gross g 416 0.01 0.08 (200/312) Hiddletown 0.10 (5/52) 0.08 (8/104)

(pCI/m ) (0.05-2.0) 5 IAi. 340 (0.08-2.0) (1.05-1.40) y-Spec 32 137cR 0.01 0.05 (4/24) Smithvi 1 le 0.08 (2/4) LLD (0.03-0.13) 2.5 mi. 160'0.03-2.0) 131, 0.07 0.12 (2/24) Podunk 0.20 (2/4) 0.02 (2/4)

(0.09-0.18) 4.0 mi. 270 (0.10-0.31)

Fish (pCI/kg) g-Spec. 8 (wet weight) 137cR 130 LLD LLD 90 (1/4) 134ce 130 LLD LLD LLD 6', 130 180 (3/4) River Hile 35 See Column 4 LLD (150-225)

'SImmary Table is taken from the NRC's Branch Technical Position, Rev. 1, Kov. 1979. and provided for illustrative purposes only.

'Hean and range based upon detectable measurements only. Fraction of detectable measurements at specified locations is indicated in parentheses (f). m OU AX

+m 02 mw cps O

LD IQ w 44 Ul

TABLE 5-4 REPORTING LEVELS fOR NONROUTINE OPERATING REPORTS Reporting Level (RL)

Particulate Broad Leaf A~nal sis, Water or Gases Fish 111 (pci/1) (pCi/N ) (pCi/kg, wet) (pCi/1) (pCi/Kg, wet)

H-3 2x10*

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For drinking water samples. This is 40 CFR Part 141 value.

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AMENDMENT NO. 9 JANUARY 1992 6.0 CONDUCT OF TESTS AND INSPECTIONS IN SUPPORT OF WNP-2 RADIOACTIVE EFFLUENT AND RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAMS 113

AMENDMENT NO 9 JANUARY 1992

6.0 INTRODUCTION

NOTE: In accordance with Generic Letter 89-01, the following Limiting Conditions for Operations (LCO) have .,been relocated from the WNP-2 Technical Specifications to the ODCM. To differentiate between Technical Specifications and ODCM programs, the following title changes have been made:

Limiting Condition for Operation Requirement for Operability Appl i cabi 1 i ty Relevant Conditions Action Compensatory Measures Surveillance, Surveillance Requirements Periodic Tests and Inspections The following, Requirement for Operability are numbered sequentially as part of Section 6.0. The Technical Specification numbering has been retained in parenthesis to promote traceability. The above changes will conform to plant practices being developed with the WNP-2 Improved Technical Specifications Program. Further sections 1.0 and 4.0 of the WNP-2 Technical Specifications are to be followed in conforming to this section .and applicability statements 3.0. 1, 3.0.2, 3.0.3 and 3.0.4 of the WNP-2 Technical Specifications are to be followed as applied in the text of the Requirement for Operability.

114

0 AMENDMENT NO. 9 JANUARY 1992 6.1 INSTRUMENTATION IN SUPPORT OF WNP-2 RADIOACTIVE EFFLUENT MONITORING REQUIREMENT FOR OPERABILITY 115

AMENDMENT NO. 10 JANUARY 1992 I

6. 1 3 4. 3

~ INSTRUMENTATION 6.1.1~ 3 4.3.7.11

~ ~ RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION CONTROLS 6.1.1. 1 (3.3.7.11) The radioactive liquid effluent monitoring instrumentation channels shown in Table 6. 1. 1. 1-1 (3.3.7. 11-1) shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Requirement for Operability 6.2. 1. 1 (3. 11. 1. 1) are not exceeded. The alarm/trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters described in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

a 0 With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, immediately suspend the release of radio-active liquid effluents monitored by the affected channel or declare the channel inoperable.

b. With less than. the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the COMPENSATORY MEASURES shown in Table 6. 1. 1. 1-1 (3.3.7. 11-1).

Restore the inoperable instrumentation to OPERABLE status within the time specified in the COMPENSATORY MEASURES or, in lieu of a Licensee Event Report, explain why this inoperability was not corrected within the time specified in the next Radioactive Effluent Release Report.

c~ The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS and INSPECTIONS 6.1.1.1.1 (4.3.7.11) 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 operations at the frequencies shown in Table 6. 1. l. l. 1-1 (4.3.7. 11-1).

116

TABLE e.l.l. 1-1 3.3.7.11-1 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION MINIHUH CHANNELS COMPENSATORY, INSTRUMENT OPERABLE MEASURES

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line 100

b. Turbine Building Sump 1/Sump 101

2. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMOTIVE TERHINATION OF RELEASE

a. Service Water System Effluent Line 101

b. RHR Service Water System Effluent Line 1/Loop 101

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 102

b. Plant Discharge-Blowdown Line 102 m

M C7 am A7 K C)

R lO fQ D

'AMENDMENT NO. 9 JANUARY 1992

.TABLE 6.1.1.1-1 3.3.7.11-1 (Continued)

COMPENSATORY MEASURES COMPENSATORY With the number of channels OPERABLE less than required by MEASURE 100 the Minimum Channels. OPERABLE requirements, effluent releases via this pathway may 'continue for up to 30 days provided that prior to initiating a release:

a ~ At least two independent samples of the batch are analyzed in accordance with Periodic Tests and Inspections 6.2. 1. 1. 1 (4. 11. l. 1. 1) and 6.2. 1. 1.2 (4.11.1.1.2) and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and the discharge valve lineup; Otherwise, suspend release of radioactive effluents via this pathway.

COMPENSATORY- With the number of channels OPERABLE less than required by MEASURE 101 the Minimum Channel OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and are analyzed for gross radioactivity (beta or gamma) at a limit of detection of at least 10 microcurie/mL.

COMPENSATORY- With the number of channels OPERABLE less than required by HEASURE 102 the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that 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 performance curves generated in place may be used to estimate flow.

118

TABLE 6.1. .1.1-1 4.3.7.11-1 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION PERIODIC TESTS AND INSPECTIONS

.CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line 0- R(3) Q(1,2)

P

b. Turbine Building Sump R(3) Q(1 5)

2. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC TERMINATION OF RELEASE

a. Service Water System Effluent Line D R(3) Q(5)

b. RHR Service Water System Effluent Line D R(3)

Q(2)

3. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line N.A. '

D(4)

b. Plant Discharge-Blowdown Line D(4) N.A.

m GO

'K Pl C R C)

CQ D

AMENDMENT NO. 10 JANUARY 1992 TABLE 6. 1. 1. 1. 1-1 4.3.7. 11-1

"

(Continued)

TABLE NOTATIONS (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway occurs if the:

Instrument indicates measured levels above the alarm setpoint.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. High voltage abnormally low.

3. Instrument indicates a downscale failure.

4. Instrument controls not set in operate mode.

The initial CHANNEL CALIBRATION shall be performed using one or more reference standards certified by the National Institute of Science and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with HIST. These standards shall permit .calibrating the system. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when continuous, periodic, or bate) releases are made.

(5) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

l. Instrument indicates measured levels above the alarm setpoint.

2. High voltage abnormally low.

3. Instrument indicates a downscale failure.

120

(

AMENDMENT NO. 16 DECEMBER 1993

6. 1 3 4;3 INSTRUMENTATION 6.1.2 '

4;3.7.12 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION CONTROLS

6. 1.2.1 (3.3.7. 12) The radioactive gaseous effluent monitoring instrumentation channels shown in Table 6. 1.2. 1-1 (3.3.7. 12-1) shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Requirements for Operability 6.2.2.1 (3. 11.2. 1) are not exceeded. The alarm/trip setpoint of these channels shall be determined in accordance with the methodology and parameters described in the ODCM.

RELEVANT CONDITION: As shown in Table 6.1.2.1-1 (3.3.7.12-1).

COMPENSATORY MEASURES:

With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, immediately initiate action to suspend the release of radioactive gaseous effluents monitored by the .

affected channel or change the setpoint so it is acceptably conservative or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the COMPENSATORY MEASURES shown in Table 6. 1.2. 1-1 (3.3.7. 12-1).

Restore the inoperable instrumentation to OPERABLE status within the time specified in the COMPENSATORY MEASURES or, in lieu of a Licensee Event Report, explain why this inoperability was not corrected within the time specified in the next Radioactive Effluent Release Report.

C. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS

6. 1.2. 1.1 (4.3.7. 12) 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 operations at the frequencies shown in Table 6. 1.2. 1. 1-1 (4.3.7. 12-1).

121

AMENDMENT NO. 9 JANUARY 1992

6. 1 3 4.3~ INSTRUMENTATION 6.1.2~ 3 4.3.7.12

~ ~ RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION CONTROLS 6; 1.2. 1 (3.3.7. 12) The radioactive gaseous effluent monitoring instrumentation channels:shown in Table 6.1.2.1-1'3.3.7.12-1) shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Requirements for Operability 6.2.2. 1 (3. 11.2. 1) are not exceeded. The alarm/trip setpoint of-these channels shall be determined in accordance with the methodology and parameters described in the ODCM.

RELEVANT CONDITION: As shown in Table 6.1.2.1-1 (3.3.7.12-1).

COMPENSATORY MEASURES:

a ~ With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, immediately initiate action to suspend the release of radioactive gaseous effluents monitored by the affected channel or change the setpoint so it is acceptably conservative or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the COMPENSATORY MEASURES shown in Table 6. 1.2. 1-1 (3.3.7. 12-1).

Restore the inoperable instrumentation to OPERABLE status within the time specified in the COMPENSATORY MEASURES or, in lieu of a Licensee Event Report, explain why this inoperability was not corrected within the time specified in the next Semiannual Radioactive Effluent Release Report.

c ~ The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.1.2.1.1 (4.3.7.12) 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 operations at the frequencies shown in Table 6. 1.2. 1. 1-1 (4.3.7:12-'1,).

121

TABLE 6. 1.2. 1-1 3.3.7. 12-1 RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUHENTATION HINIHUH CHANNELS RELEVANT COHPENSATORY INSTRUHENT OPERABLE CONDITIONS HEASURES Hain Condenser Offgas Post-Treatment Radiation Honitor

a. Gross Gamma Detection Alarm and 110 Automatic Isolation of the Offgas System Outlet and Drain Valves

2. Hain Condenser Offgas Pre-Treatment Radiation Honitor .

a. Gamma Sensitive Ion-Chamber Located Upstream of Holdup Line

3. Hain Plant Vent Release Honitor

a. Low Range Activity Honitor 1 110

b. Iodine Sampler 1 112

c. Particulate Sampler 1 112

d. Effluent System Flow Rate Honitor 1 113

e. Sampler 'Flow Rate Honitor 1 113 Dl C7 am AK ill C)

LD LQ w CrJ CsJ

TABLE 6.1.2. 1-1 3.3.7.12-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION HINIMUM CHANNELS RELEVANT COMPENSATORY

,

INSTRUMENT OPERABLE CONDITIONS HEASURES Turbine Building Ventilation Exhaust Monitor

a. Noble Gas Activity Monitor

1) Low Range 1 110

2) Intermediate Range 1 110

b. Iodine Sampler 1 112 c ~ Particulate Sampler '1 112

d. Effluent System Flow Rate Monitor 1 . 113

e. Sampler Flow Rate Honitor 1 113

5. Radwaste Building Ventilation Exhaust

a. Noble Gas Activity Monitor

1) Low Range 1 110

2) Intermediate Range -1 110

b. Iodine Sampler 1 112 =

c. Particulate Sampler 1 112

d. Effluent System Flow Rate Measurement 1 115 Device 8

e. Sampler Flow Rate Monitor 113 CaK-

%D C Pl'l M

~O

AMENDMENT NO. 9

,. JANUARY 1992 TABLE 6. 1.2. 1-1 3.3.7. 12-1 (Continued)

TABLE NOTATIONS

At all times.

- During main condenser offgas treatment system operation.

8 Radwaste Building ventilation exhaust fan. There are 3 fans; WEA-FN-IA, WEA-FN-1B and WEA-FN-1C.

COMPENSATORY MEASURES COHPENSATORY- With the number of channels OPERABLE less than required by the MEASURE 110 Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed for noble gas gamma emitters within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

COMPENSATORY- With the number of channels OPERABLE less than required by the HEASURE 112 Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the channel has been declared inoperable samples are continuously collected with auxiliary sampling equipment as required in Table 6.2.2. 1.2-1 (4.11-2).

COMPENSATORY- With the number of channels OPERABLE less than required by the HEASURE 113 Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that the flow rate is estimated at least once per 4 hours.

COMPENSATORY- With the number of channels operable less than required by the HEASURE 114 Minimum Channels OPERABLE requirement, gases from the main condenser offgas treatment system may be released to the environment for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided:

a. The offgas treatment system is not bypassed, and

b. The offgas post-treatment monitor used in a pretreatment function shall be OPERABLE.*

COMPENSATORY- With the number of channels OPERABLE less than required by the HEASURE 115 Minimum Channels OPERABLE requirement, effluent releases via this pathway shall be terminated.

With the offgas post-treatment monitor in a pretreatment function unavailable or inoperable, install a temporary replacement ionization chamber for the pre-treatment monitor or be in HOT STANDBY within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

124

TABLE 6.1.2.1.1-1 4.3.7.12-1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION PERIODIC TESTS AND INSPECTIONS RE UIREHENTS MODES IN MHICH CHANNEL PERIODIC =-TESTS CHANNEL SOURCE CHANNEL FUNCTIONAL RAND INSPECTIONS INSTRUMENT CHECK CHECK CALIBRATION TEST A~IRt II tt

1. Hain Condenser Offgas Post-Treatment Radiation Monitor

a. Gross gamma detector alarm and 0 . R(2) Q(1) automatic isolation of the offgas system outlet and drain valves

2. Hain Condenser Offgas Pre-Treatment Radiation Monitor

a. Gamma sensitive ion chamber located D R(2) Q(1) upstream of holdup line

3. Hain Plant Release Monitor

a. Low Range Activity Monitor D R(2) Q(1)

b. Iodine Sampler M N.A. N.A. N.A.

c. Particulate Sampler M N.A. N.A. N.A.

d. Effluent System Flow Rate Monitor 0 N.A. R

e. Sampler Flow Rate Monitor D N.A. R m

'D a rn C

I Z C7 lO W EA 4l

TABLE 6. 1.2. 1. 1-1 4.3.7. 12-1 (Continued)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION PERIODIC TESTS AND INSPECTIONS RE UIREMENTS MODES -IN WHICH CHANNEL PERIODIC TESTS CHANNEL SOURCE CHANNEL FUNCTIONAL AND INSPECTIONS INSTRUMENT CHECK CHECK CALIBRATION TEST EEKE tREO

4. Turbine Building Ventilation Exhaust Monitor

a. Noble Gas Activity Monitor -

1) Low Range D H R(2) Q(1)

2) Intermediate Range 0 H R(2) Q(6)

b. Iodine Sampler W N.A. N.A. N.A.

c. Particulate Sampler W N.A. N.A. N.A.

d.

e.

Effluent System Flow Rate Monitor Sampler Flow Rate Monitor D

D '.A.N.A. R R

Q Q

5. Radwaste Building Ventilation Exhaust

a. Noble Gas Activity Honitor

1) Low Range D H R(2) Q(1)

2) Intermediate Range D H R(2) Q(6)

b. Iodine Sampler N.A. N.A. N.A.

c. Particulate Sampler W N.A. N.A. N.A.

d. Effluent System Flow Rate Monitor D(3) N.A. R(5) Q(4)

e. Sampler Flow Rate Monitor 0 N.A. R Q m

%D KX c= m 2o M Xl

~O lO

~ lO

AMENDMENT NO. 10 JANUARY 1992 TABLE 6. 1.2. 1. 1-1 4.3.7. 12-1 (Continued)

TABLE NOTATIONS ;

'* At all times.

- During main condenser offgas treatment system operation (1) The CHANNEL, FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

a. Instrument indicates measured levels above the alarm setpoint.

b. Circuit failure.

(2) The initial CHANNEL CALIBRATION shall be performed using one or more reference radioactive standards traceable to the NATIONAL INSTITUTE OF SCIENCE AND TECHNOLOGY (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range.

Subsequent CHANNEL CALIBRATION shall be performed using the initial radioactive standards or other standards of equivalent quality or radioactive sources that have been related to the initial calibration.

(~) The CHANNEL CHECK shall be performed by comparing a computer reading or power signal comparing each fan's local amperage reading with preestablished baseline values.

(4) The CHANNEL FUNCTIONAL TEST shall be performed by measurement of the phase currents for each fan.

The CHANNEL CALIBRATION shall be performed by using .a flow measurement device to determine the fan current to flow relationship.

(6) For the CHANNEL FUNCTIONAL TEST on the intermediate range noble gas activity monitors, demonstrate that circuit failures or instrument controls when set in the OFF position produce control room alarm annunciation.

127

e AMENDMENT NO. 9 JANUARY 1992.

6.2 REQUIREMENT FOR OPERABILITY SUPPORT OF RADIOACTIVE EFFLUENT MONITORING PROGRAMS 128

AMENDMENT NO. 9 JANUARY 1992 1

6.2 3 4.11 RADIOACTIVE EFFLUENTS 6.2.1 3 4.11.1 LI UID EFFLUENTS 6.2. 1. 1 CONCENTRATION RE UIREMENTS FOR OPERABILITY 6.2. 1. 1 (3. 11. l. 1) The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see ODCM Figure 3-1) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix 8, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.

For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10'icrocurie/ml total activity.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concen-tration to the above limits.

PERIODIC TESTS AND INSPECTIONS 6.2. 1. 1. 1 (4. 11. l. 1. 1) Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 6.2.1.1.1-1 (4.11-1).

6.2.1.1.2 (4.11.1. 1.2) The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Re'quirement for Operability 6.2. 1. 1 (3. 11. 1. 1).

129

AMENDMENT NO. 10 JANUARY 1992 TABLE -6.2.1.1.1-1 4.11-1 RADIOACTIVE LI UID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY TYPE (LLD)'vCi/ml)

FREQUENCY FREQUENCY ANALYSIS A. Batch Waste P Release Tanks' Each Batch Each Batch Emitters'xlO' Principal Gamma I-131 1 x10'xlO' P Dissolved and One Batch/M Entrained Gases (Gamma Emitters)

M H-3 lx10 Each Batch Composite'ross Alpha P Q Sr-89, Sr-90 Each Batch Composite" 5x10'x10 Fe-55 130

AMENDMENT NO. 10 JANUARY 1992 TABLE 6.2. 1. 1. 1-1 4. 11-1 (Continued)

TABLE NOTATIONS

'he LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95K probability with only 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

' '.22 x 10 ' 'xp (-%~7')

Where:

LLD is the "a priori" lower limit,of detection as defined, above, as microcuries per unit mass or volume, s, 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 disintegration, V is the sample size in units of mass or volume, 2.22 x 10's the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and

~7 for plant effluents is the elapsed time between the midpoint of sample collection and time of counting.

Typical values of E, V, Y, and ~r should be'used in the calculation.

It shouid be recognized that the LLD is defined as an a griori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

131

AMENDMENT NO. 9 JANUARY 1992 TABLE 6.2. 1. 1. 1-1 4. 11-1 (Continued)

TABLE NOTATIONS

'I

' batch'elease is the discharge of liquid wastes of a discrete volume.

Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed by a method described in the ODCH to assure, representative sampling.

'he principal gamma emitters for which the LLD specification applies include the following radionuclides: Hn-54, Fe-59, Co-58, Co-60, Zn-65, Ho-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to Technical Specification 6.9.1.11.

~

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 that is representative of the liquids released. This may be accomplished through composites of grab samples obtained prior, to discharge after the tanks have been recirculated.

132

AMENDMENT NO. 9 JANUARY 1992 6.2 (3/4.11) RADIOACTIVE EFFLUENTS

(/. I.l)'~lt E I E 6.2.1.2 DOSE RE UIREMENT FOR OPERABILITY 6.2. 1.2 (3. 11. 1.2), The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS (see ODCM Figure 3-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems to the total 'body and to less than or equal to 5 mrems to any organ, and

b. During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ.

RE EVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

a ~ With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective action to be taken to assure that subsequent releases will be in compliance with the above limits. This Special Report shall also include (1) the results of radiological analyses of the drinking water source and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR Part 141.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.2.1.2.1 (4.11.1.2) Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

133

AMENDMENT NO. 9 JANUARY 1992.

6.2 (3/4.11)~ RADIOACTIVE EFFLUENTS

.2.t l3/4.ll.l') ~UUI

~ ~ ~ EFFLUENT 6.2. 1.3 (3. 11.'1.3) LI UID RADWASTE TREATMENT SYSTEM A

RE UIREMENT FOR OPERABILITY 6.2. 1.3 (3. 11. 1.3) The liquid radwaste treatment system shall be OPERABLE.

The appropriate portions of the system shall be used to reduce the releases of radioactivity when the projected doses due to the liquid effluent, from each reactor unit, to UNRESTRICTED AREAS (see ODCH Figure 3-1) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31-day period.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

a ~ With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the liquid radwaste treatment system not in operation, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days pursuant to Technical Specification 6.9.2 a Special Report that includes the following information:

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

2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of actions(s) taken to prevent a recurrence.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.2. 1.3. 1 (4. 11. 1.3. 1) Doses due to liquid releases from each reactor unit to UNRESTRICTED AREAS shall be projected at least once per 31.days -in accordance with the methodology and parameters in the ODCH.

6.2. 1.3.2 (4. 11. 1.3.2) The installed liquid radwaste treatment system shall be demonstrated OPERABLE by meeting Requirement for Operability 6.2. 1. 1 (3.11.1.1) and 6.2.1.2 (3.11.1.2).

134

AMENDMENT NO. 9 JANUARY 1992 6.2 (3/4.11) RADIOACTIVE EFFLUENTS 6.'2.2 (3/4. 11.2) GASEOUS EFFLUENTS 6.2.2.1 (3.11.2.1) DOSE RATE RE UIREMENT FOR OPERABIL'ITY 6.2.2. 1 (3.11.2.1) The dose rate due to radioactive materials released in gaseous effluents from the site to are'as at and beyond the SITE BOUNDARY (see ODCH Figure 3-1) shall. be limited to the following:

a. For noble gases: Less than or equal to 500 mrems/yr to the total body and less than or equal to 3000 mrems/yr to the skin, and

b. For iodine-131, for iodine-133, for tritium, and for all radio-nuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

With the dose rate exceeding the above limits, immediately restore the release rate to within the above limit(s).

PERIODIC TESTS AND INSPECTIONS 6.2.2. 1. 1 (4. 11.2. 1.1) The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCH.

6.2.2. 1.2 (4. 11.2. 1.2) The dose rate due to iodine-131, iodine-133,. tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCH by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 6.2.2. 1.2-1 (4. 11-2).

135

TABLE 6.2.2.1.2-1 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM HINIHUH -LOWER LIMIT OF SAMPLING ANALYSIS TYPE OF DETECTION GASEOUS RELEASE TYPE FRE UENCY FRE UENCY ACTIVITY ANALYSIS (LLD)'imL

.

P P A. Primary Containment Each PURGE Each PURGE Principal Gamma lxlO and VENT and VENT Emitters'-3 PURGE and VENT Grab Sample lxlO'.

Hain Plant Vent Hb,d Principal Gamma lxlO Grab Sample Emitters'-3 lx10 C. Turbine Building H Principal Gamma lx10 Vents and Radwaste Grab Sample Emitters'-3 Building Vents lxlO ch D All Release Types Continuous's I-131 lxl0" listed in A, Sample 1-133 lxlO" B, and C above Principal Gamma Emitters lx10 W'articulate Continuous'ontinuous Sample H Gross Alpha lxlO Composite Par-ticulate Sample Continuous'ontinuousCharcoal Sr-89, Sr-90 lxlO Composite Par-ticulate Sample m Cl Noble Gas Noble Gases lx10 am Monitor Gross Beta or Gamma (Xe-133 equi val ent) C CA ED LD

~

AHENDHENT NO. 9.

JANUARY 1992 TABLE 6.2.2. 1.2-1 4. 11-2 (Continued)

TABLE NOTATIONS

'he LLD is defined, for purposes 'of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

4.66sb LLD E

' '.22 x 10 ' 'xp (-A~t)

Where:

LLD is the "a priori" lower limit of detection as defined above, as microcuries 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 disintegration, V is the sample size in units of mass or volume, s

2.22 x 10 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and

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

Typical values of E, V, Y, and ~t should be- used in the calculation.

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

137

AMENDMENT NO. 10 JANUARY 1992 TABLE 6.2.2. 1.2-1 4. 11-2 (Continued)

TABLE NOTATIONS .,

Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change .exceeding 15% of RATED THERMAL POWER within a 1-hour period.

'amples 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 sampler.

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and analyses shall be completed within 48 hou'rs of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE E(UIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.

'ritium grab samples shall be taken at least once per 7 days from the main plant vent stack to determine tritium releases in the ventilation exhaust from the spent fuel pool area whenever spent fuel is in the spent fuel pool.

'he 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 Requirement for Operability 6.2.2. 1, 6.2.2.2 and 6.2.2.3 (3.11.F 1, 3.11.2.2, and 3.11.2.3).

'he principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141 and Ce-144 in iodine and particulate releases.

This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report.

138

AMENDMENT NO. 9

. JANUARY 1992

6. 2 (3/4. 11) RAD IOACTIV EFFLUENTS 6.2.2 (3/4. 11.2) 'ASEOUS EFFLUENTS 6.2.2.2 (3.11.2.2) 'OSE - NOBLE GASES RE UIREMENT FOR OPERABILITY'.2.2.2 (3. 11.2;2) The air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see ODCM Figure 3-.1) shall be limited to the following:

a. 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,

b. 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.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

a ~ With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subse-.

quent releases will be in compliance with the above limits.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.2.2.2. 1 (4. 11.2.2) Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

139

AMENDMENT N0.-9 4 JANUARY 1992

6.2 (3/4.11)

~ RADIOACTIVE EFFLUENTS 6.2.2 (3/4. 11.2)

~ ~ ~ GASEOUS EFFLUENTS 6.2.2.3 (3.11.2.3) "DOSE - IODINE-131 IODINE-133 TRITIUH AND RADIONUCLIDES IN PARTICULATE FORH RE UIREHENT FOR OPERABILITY 6.2.2.3 (3.11.2.3) The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see ODCH Figure 3-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrems to any organ and,

b. During any calendar year: Less than or equal to 15 mrems to any organ.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

a ~ With the calculated dose from the release of iodine-131, iodine-133, 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 a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s),for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.2.2.3. 1 (4. 11.2.3) Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131; iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters-in the ODCH at least once per 31 days.

140

AHENDHENT NO. 9 JANUARY 1992'.2 (3/4.11) RADIOACTIVE EFFLUENTS'.2.2 (3/4. 11.2) GASEOUS EFFLUENT 6.2.2.4 (3.11.2.4) GASEOUS OFFGAS RADWASTE TREATMENT SYSTEM RE UIREMENT FOR OPERABILITY 6.2.2.4 (3.11.2.4) The GASEOUS OFFGAS 'RADWASTE TREATMENT SYSTEM* shall be in operation in either the normal or charcoal bypass mode. The charcoal bypass mode shall not be used unless the offgas post-treatment radiation monitor is OPERABLE as specified in Table 6. 1.2. 1-1 (3.3.7. 12-1).

RELEVANT CONDITIONS: Whenever the main condenser steam jet air ejector (evacuation) system is in operation.

COMPENSATORY MEASURES:

a 0 With the GASEOUS OFFGAS RADWASTE TREATMENT SYSTEM not used in the normal mode for more than 7 days, in lieu of a Licensee Event Report, prepare and submit to the Commission within, 30 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action(s) taken to prevent a recurrence.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.2.2.4. 1 (4. 11.2.4) The GASEOUS OFFGAS RADWASTE TREATMENT SYSTEM shall be verified to be in operation in either the normal or charcoal bypass mode at least once per 7 days whenever the main condenser steam "jet air ejector system is in operation. 'evacuation)

A GASEOUS OFFGAS RADWASTE TREATMENT SYSTEH shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

141

AMENDMENT NO. 9' j I JANUARY 1992

'.2 (3/4.11)

~ RADIOACTIVE EFFLUENTS 6.'2.2 (3/4.11.2), 'ASEOUS

~ ~ EFFLUENTS 6.2.2.5 (3. 11.2.5) VENTILATION EXHAUST TREATMENT SYSTEM RE UIREMENT FOR OPERABILITY 6.2.2.5 (3.11.2.5) The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM'shall be OPERABLE and shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from each reactor unit to areas at and beyond the SITE BOUNDARY (see ODCH Figure 3-1) when averaged over 31 days would exceed 0.3 mrem to any organ in a 31-day period.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

'a ~ With the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days, or with gaseous waste being discharged without treatment and in excess of the above limits, in lieu of a Licensee Event Report, prepare and submit to the Commission within 10 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems, and the reason for the inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action(s) taken to prevent a recurrence.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.2.2.5. 1 (4. 11.2.5. 1) Doses due to gaseous release to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCH.

6.2.2.5.2 (4.11.2.5.2) The VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE by operating the VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 10 minutes, .at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

142

AMENDMENT NO. 11 AUGUST 1992 6.2.2 (3/4.11) 'ADIOACTIVE EFFLUENTS 6.2.2 (3/4. 11.2) GASEOUS EFFLUENTS 6.2.2.6 (3.11.2.8) VENTING OR PURGING RE UIREMENT FOR OPERABILITY 6.2.2.6 (3.11.2.8) VENTING or PURGING of the Hark II containment drywell shall be through the standby gas treatment system or the primary containment vent and purge system. The first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of any vent or purge operation shall be through one standby gas treatment system.

RELEVANT CONDITIONS: All drywell vents and purges in Mode 1, 2, or 3, and when de-inerting.

COMPENSATORY MEASURES:

'a ~ With the requirements of the above specification not satisfied, suspend all VENTING and PURGING of the drywell.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.2.2.6. 1 (4. 11.2.8. 1) The containment drywell shall be determined to be aligned for VENTING or PURGING through the standby gas treatment system or the primary containment vent and purge system within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING or PURGING of the drywell.

6.2.2.6.2 (4. 11.2.8.2) Prior to use of the purge system through the standby gas treatment sys'em assure that:

a. Both standby gas treatment system trains are OPERABLE whenever the purge system is in use, and

b. Whenever the purge system is in use during OPERATIONAL CONDITION 1 or 2 or 3, only one of the standby gas treatment system trains may be used.

6.2.2.6.3 (4.11.2.8.3) The containment drywell shall be sampled and analyzed per Table 6.2.2. 1.2-1 (4.11-2) of Requirements for Operability 6.2.2.1 (3.11.2.1) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during VENTING and PURGING of the drywell through other than the standby gas treatment system.

143

AMENDMENT NO. 9 JANUARY 1992 6.2 (3/4.11) ~ RADIOACTIVE EFFLUENTS 6.2.3 (3/4.11.3)

~ ~ ~ SOLID RADIOACTIVE WASTE 6.2.3.1 (3.11.3) SOLID RADIOACTIVE WASTE RE UIREMENT FOR OPERABILITY 6.2.3.1 (3.11.3) Radioactive wastes shall be SOLIDIFIED or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transpor-tation requirements during transit, and disposal site requirements when received at the disposal site.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

a ~ With SOLIDIFICATION* or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures and/or the solid waste system as necessary to prevent recurrence.

b. With SOLIDIFICATION or dewatering not performed in accordance with the PROCESS CONTROL PROGRAM, (1) test the improperly processed waste in each container to ensure that it meets burial ground and shipping require-ments and (2) take appropriate administrative action to prevent recurrence.

c~ The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.2.3.1.1 (4.11.3) SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive wastes (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM.

a ~ If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICA-TION of the batch under test shall be suspended until such time as additional t6st specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.

SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

144

AMENDMENT NO. 9

.JANUARY 1992 PERIODIC TESTS AND INSPECTIONS Continued

b. If the initial test specimen-from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test spe'cimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Technical Specification 6.13, to assure SOLIDIFICATION of subsequent batches of waste.

e C. With the installed equipment incapable of meeting Requirement for Operability 6.2.3. 1 (3. 11.3) or declared inoperable, restore the equipment to OPERABLE status or provide for contract capability to process wastes as necessary to satisfy all applicable transportation and disposal requirements.

145

AMENDMENT NO. 9 f JANUARY )992 6.2 (3/4.11)~ RADIOACTIVE EFFLUENTS

.6.2.4 (3/4.11.4)'OTAL

~ ~ ~ DOSE RE UIREMENT FOR OPERABILITY 6.2.4. 1 (3. 11.4) The annual (calendar year) dose or dose commitment'to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body, or any organ, except the thyroid, which shall be limited to

,less than or equal to 75 mrems.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

'a ~ With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Requirement for Operability 6.2.1.2.a, 6.2.1.2.b, 6.2.2.2.a, 6.2.2.2.b, 6.2.2.3.a, or 6.2.2.3.b (3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b,

3. 11.2.3.a, or 3. 11.2.3.b), calculations shall be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of Requirement for Operability 6.2.4. 1 '(3. 11.4) have been exceeded. If such is the case, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered,a timely request, and a variance is granted until staff action on the request is complete.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

146

'MENDMENT JANUARY, NO. 9 6.2 (3/4.11)

~ RADIOACTIVE EFFLUENTS 6.2.4 (3/4. 11.4)

~ ~ TOTAL DOSE Continued 1992'.2.4.

PERIODIC TESTS AND INSPECTIONS

1. 1 (4. 11.4. 1) Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with PERIODIC TESTS AND INSPECTIONS 6.2.1.2.1, 6.2.2.2.1, and 6.2.2.3.1 (4.11.1.'2, 4.11.2.2, and 4.11.2.3), and in accordance with the methodology and parameters in the ODCM.

6.2.4.1.2 (4. 11.4.2) Cumulative dose contributions from direct radiation from unit operation shall be determined in accordance with the methodology and parameters in the ODCM.

147

AMENDMENT NO. 9 JANUARY 1992 6.3 RE(UIREHENT FOR OPERABILITY IN SUPPORT OF THE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 148

AMENDMENT NO. 9 JANUARY 1992 6.3 (3/4.12)

~ RADIOLOGICAL ENVIRONMENTAL MONITORING 6.3.1 (3/4.12.1)

~ ~ ~ MONITORING PROGRAM RE UIREMENT FOR OPERABILITY 6.3. 1.1 (3. 12. 1) The radiological environmental monitoring program shall be conducted as specified in Table 6.3. 1. 1-1 (3. 12-1).

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

'a ~ With the radiological environmental monitoring program not as specified in Table 6.3.1.1-1 (3.12-1), in lieu of a being'onducted Licensee Event Report, 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.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 6.3. 1. 1-2 (3. 12-2) when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose* to A MEMBER OF THE PUBLIC is less than the calendar year limits of Requirement for Operability 6.2. 1.2, 6.2.2.2 and 6.2.2.3 (3. 11. 1.2, 3. 11.2.2, and

3. 11.2.3). When more than one of the radionuclides in Table 6.3. 1. 1-2 (3. 12-2) are detected in the sampling medium, this report shall be submitted if:

reporting level (1) reporting level (2) >>o radionuclides other than those in Table 6.3. 1. 1-2 (3. 12-2) are

'hen detected and are the result of plant effluents, this 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 Requirement for Operability 6.2. 1.2, 6.2.2.2 and 6.2.2.3 (3. 11. 1.2, 3. 11.2.2 and

3. 11.2.3). This 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.

c~ With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 6.3. 1. 1-1 (3. 12-1),

identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days.

The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

149

AMENDMENT NO. '16 DECEMBER 1993 RADIOLOGICAL ENVIRONMENTAL MONITORING RE UIREMENT FOR OPERABILITY Continued COMPENSATORY MEASURES: (Continued)

The specific locations from which samples were unavailable may then be deleted from the monitoring program. In lieu of a Licensee Event Report and pursuant to Technical Specification 6.9. l. 11, identify the cause 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 figure(s) and table for the ODCM reflecting the new location(s).

d. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.3. 1. 1. 1 (4. 12. 1) The radiological environmental monitoring samples shall be collected pursuant to Table 6.3. l. 1-1 (3. 12-1) from the specific..locations given in the table and figure(s) in the ODCM,, and shall be analyzed pursuant to the requirements of Table 6.3. 1. 1-1 (3. 12-1) and the detection capabilities required by Table 6.3. 1. 1. 1-1 (4. 12-1).

150

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM*

EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES AND TYPE AND FREQUENCY AND/OR SAMPLE AND SAMPLE LOCATIONS'AMPLING COLLECTION FREQUENCY OF ANALYSIS

1. DIRECT RADIATION 34 routine monitoring stations Quarterly. Gamma dose quarterly.

either with two or more dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:

An inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDARY.

An outer ring of stations, one in each of the meteorological sectors of NE, ENE, E, ESE, SE in the 6- to 9-km range from the site, and one in each of the meteorological sectors of N, NNE, SSE, S, SSW in the 9- to 12-km range from the site.

The balance of the stations to be placed in special interest areas such as population centers, nearby residences, schools, and 1 or 2 areas to serve as control stations.

m PD

The number, media, frequency, and location of samples may vary from site to site. This table presents an C f11 acceptable minimum program for a site at which each entry is applicable. Local site characteristics must be examined to determine if pathways not covered by this table may significantly contribute to an individual's ~O dose and should be included in the sampling program. Q) ~

~ lO

TABLE 6.3. 1. 1.- 3. 12-1 (Continued)

RADIOLOGICAL ENVIRONHENTAL HONITORING PROGRAH*

EXPOSURE PATHWAY NUHBER OF REPRESENTATIVE SAHPLES SAMPLING AND TYPE AND FREQUENCY AND/OR SAHPLE AND SAHPLE LOCATIONS'. COLLECTION FREQUENCY OF ANALYSIS AIRBORNE Radioiodine and Samples from 5 locations: Continuous sampler Radioiodine Canister:

Particulates operation with sample 1-131 analysis weekly.

1 sample from close to the 1 collection weekly, or SITE BOUNDARY location, having high calculated annual average a more frequently if Particulate Sam ler:

required by dust Gross beta radioactivity-ground-level 0/Q. loading. analysis following filter change Three samples from close to the 3 Columbia River locations having the highest calculated D/Q One sample from the vicinity of Gamma isotopic a community having the analysis'f highest composite (by loca-.

calculated annual average tion) quarterly.

ground-level D/Q.

One sample from a control loca-tion, as for example 30-50 km distant and in the least prevalent wind direction.

3. WATERBORNE

a. 1 sample upstream Composite sample over isotopic Surface'.

1 sample downstream 1-month period.', analysis'onthly.

Composite for tritium analysis Ground Samples from 1 or 2 affected.'amma sources Quarterly.

quarterly.

isotopic I

only if likely to be Gamma tritium analysis'uarterly.

and C)

LD PO D

TABLE 6.3.1.1.- 3.12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM*

EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES AND/OR SAMPLE AND SAMPLE LOCATIONS'AMPLING AND COLLECTION FREQUENCY TYPE AND FREQUENCY OF ANALYSIS

3. WATERBORNE (Continued)

c. Drinking One sample of each of 1 to 3 of Composite sam~le over I-131 analysis on each the nearest water supplies that 2-week period when composite when the dose could be affected by its I-131 analysis is calculated for the discharge. performed, monthly -

consumption of the water composite otherwise. is greater than I mrem per year."

One sample from a control Composite for gross beta location. and gamma isotopic analysis'onthly.

Composite for tritium analysis quarterly.

d. Sediment from One sample from downstream area Semiannually. Gamma isotopic shoreline with existing or potential analysis'emiannually.

recreational value.

4. INGESTION

a. Hilk Samples from milking animals in Semimonthly when Gamma isotopic and I-131 3 locations within 5 km distance animals are on analysis semi-'monthly-having the highest dose poten- pasture, monthly at when animals are on tial. If there are none, then 1 other times. pasture; monthly at other sample from milking animals in times.

each of 3 areas between 5-16 km distant where doses are calculated to be greater than 1 mrem per year."

1 sample from milking animals at a control location, 30-50 km distant and in the least prevalent wind direction.

TABLE 6.3. I. I.-l 3. 12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM*

EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLES AND/OR SAMPLE AND SAMPLE LOCATIONS'AMPLING COLLECTION FREQUENCY AND TYPE AND FREQUENCY OF ANALYSIS

4. INGESTION (Continued)

b. Fish and 1 sample of each of three Sample annually, isotopic Invertebrates recreationally important species unless an impact is analysis'n edible (one anadromous and two resident) in vicinity of plant semiannually.'amma indicated, then portions.

discharge area.

1 sample of same species in areas not influenced by plant discharge.

c. Food Products 1 sample of each principal class At time of Gamma isotopic of food products from any area analyses'n edible that is irrigated by water in portion.

which liquid plant wastes have harvest.'onthly been discharged.

Samples of 3 different kinds of during growing Gamma isotopic" and broad leaf vegetation grown season. I-131 analysis.

nearest each of two different offsite locations of highest predicted annual average ground-level D/Q if milk sampling is not performed.

1 sample of each of the similar Monthly during growing Gamma isotopic and CD broad leaf vegetation grown 30- season. I-131 analysis. C: Pl AK 50 km distant in the least C:H prevalent wind direction if milk CA R

C) sampling is not performed. tD Nw

, AMENDMENT NO. 16 DECEMBER 1993 TABLE 6.3.'1;1-1 3.12-1 (Continued)

TABLE NOTATIONS

'pecific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be provided for each and every'sample location in Table 6.3.1. 1-1 (3. 12-1) in a table and figure(s) in the ODCM. Refer to NUREG-0133, "Preparation of Radiological Effluent 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 circumstances such as hazardous conditions, seasonal unavailability, and malfunction of automatic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to 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 practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. In lieu of a Licensee Event Report, identify the cause of the unavailability of samples for that pathway 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 figure(s) and table for the ODCM reflecting the new location(s).

'ne 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 purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor card with multiple readout areas; a phosphor card in a packet is considered to be equivalent to two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. (The number of direct radiation monitoring stations may be reduced according to geographical limitations. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to'btain optimum dose information with minimal fading.)

'irborne particulate sample filters shall be analyzed for gross beta radio-activity 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 particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

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

'he "upstream sample" shall be taken at a distance beyond significant influence of the discharge. The "downstream" sample shall be taken in an area beyond but near the mixing zone.

155

AMENDMENT NO. 9 JANUARY 1992 TABLE 6.3. 1. 1-1 3. 12-1 (Continued)

TABLE NOTATIONS

'pecific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be provided for l.

each and every sample location in Table 6.3. 1-1 (3.12-1) in a table and figure(s) in the ODCM. Refer to NUREG-0133, "Preparation of Radiological Effluent 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 circumstances such as hazardous conditions, seasonal unavailability, and malfunction of automatic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All devi ations 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 practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the radiological environmental monitoring program. In lieu of a Licensee Event Report, identify the cause of the unavailability of samples for that pathway and identify the new location(s) for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report and also include in the report' revised figure(s) and table for the ODCM reflecting the new location(s).

'ne 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 purposes of this table, a thermolumines-cent dosimeter (TLD) is considered to be one phosphor card with multiple readout areas; a phosphor card in a packet is considered to be equivalent to two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. (The number of direct radiation monitoring stations may be reduced according to geographical limitations. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.)

'irborne particulate sample filters shall be analyzed for gross beta radio-activity 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 particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

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

'he "upstream sample" shall be taken at a distance beyond significant influence of the discharge. The "downstream" sample shall be taken in an area beyond but near the mixing zone.

155

AMENDMENT NO. 11 AUGUST 1992

'ABLE 6.3. 1. 1-1 3. 12-1 (Continued)

TABLE NOTATIONS

' composite sample is one in which the quantity, (aliquot) of liquid is

'roportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In, this program composite sample aliquots shall .be collected at time interval's that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.

'roundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.

" The dose shall be calculated for the maximum organ and age group, using the

'f methodology and parameters in the ODCM.

any of the analytical results for Columbia River fish samples are significantly higher than the results of the Snake River samples or the results of previous fish samples, sampling will be conducted semiannually.

'f each harvest occurs more than once a year, sampling shall be performed during discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.

156

TABLE 6.3. 1. 1-2 3. 12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS ANALYSIS (pCi/L) OR GASES (pCi/m') (pCi/kg, wet) (pCi/L) (pCi/kg, wet)-

H-3"'n-54 2x10 1 x 10 3x10 Fe-59 4 x10' 1 x 10 Co-58 x 10 3 x 10 Co-60 3 x 1 x 10 10'x10 Zn-65 2 x 10 Zr-Nb-95 4 x 10'0 I-131 0.9 1 x IO~

Cs-134 10 1 x 10 60 1 x 10 Cs-137 50 20 2x10 70 2 Ba-La-140 2 x 10 3 x 10 x10'1)

For drinking water samples. The value given is the 40 pathway exists, a value of 30,000 pCi/L may be used.

CFR Part 141 value. If no drinking water

TABLE 6.3.1.1.1-1 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS'OWER LIMIT OF DETECTION (LLD)

WATER AIRBORNE PARTICULATE FISH MILK FOOD PRODUCTS SEDIMENT ANALYSIS (pCi/L) OR GASES (pCi/m') (pCi/kg, wet) (pCi/L) (pCi/kg, wet) (pCi/kg, dry)

Gross beta 1x10 H-3 2000*

Mn-54 15 130 Fe-59 30 260 Co-58,60 15 130 Zn-65 30 260 Zr-95 30 Nb-95 15 I-131 7 x 10 60 Cs-134 15 5 x 10 130 15 60 150 Cs-137 18 6x10 150 18 80 180 Ba-140 60 60 La-140 15 15

(*) If no drinking water pathway exists, a va1ue of 3,000 pCi/L may be used.

AHENDHENT NO. 9 h

JANUARY 1992 TABLE 6.3. 1. 1. 1-1 (4. 12-1) (Continued)

TABLE NOTATIONS

'his list does not mean that only these nuclides are to be considered.

Other peaks that are identifiable, together with those of the above also be analyzed and reported in the Annual Radiological Environmental nuclides,'hall Operating Report.

'equired detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the rec'ommendations of Regulatory Guide 4. 13, except for specification regarding energy dependence.

Correction factors shall be provided for energy ranges not meeting the energy dependence specification.

'he LLD is defined for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

4.66sb LLD-E V '.22 '

'xp(-A~t)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, s, 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 disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable, A is the radioactive decay constant for the particular radionuclide, and

~t for environmental samples is the elapsed time between sample collec-tion, or end of the sample collection period, and time of counting.

Typical values of E, V, Y, and ~t should be used in the calculation.

159

AMENDMENT NO. 9 JANUARY 1992 TABLE 6.3. l. l. 1-1 4. 12-1 Continued TABLE NOTATIONS It should be recognized that the LLD is defined as an a griori (before the fact) l-imit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

'LD for drinking water samples. If no drinking water pathway exists,'he LLD of gamma isotopic analysis may be used.

160

AMENDMENT NO. 16 DECEMBER 1993 E

'6.3 (3/4.12) RADIOLOGICAL ENVIRONMENTAL MONITORING 6.3.2 (3/4.12.2)

~ ~ ~ LAND USE CENSUS RE UIREMENT FOR OPERABILITY 6.3.2. 1 (3.12.2) A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the'location in each of the 16 meteorological sectors of the nearest milk animal< the nearest residence and the nearest garden* of greater than 50 m'500 ft ) producing broad leaf vegetation.

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

a 0 With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Requirement for Operability 6.2.2.3. 1 (4. 11.2.3), in lieu of a Licensee Event Report, identify the new location(s) in the next Radioactive Effluent Release I Report.

b. With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Requirement for Operability 6.3. 1. 1 (3. 12.1), add the new location(s) to the radiological environmental monitoring program within 30 days. The sampling location(s), excluding the control station location having the lowest calculated dose or dose commitment(s), 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. In lieu of a Licensee Event Report, identify the new location(s) in the next Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

c. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.3.2. 1. 1 (4. 12.2) The Land Use Census shall be conducted during the growing season at least once per calendar year using that information that will pro-vide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report.

Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/gs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 6.3.1. 1-1 (3. 12-1) shall be followed, including analysis of control samples.

161

AMENDMENT NO."9 JANUARY 1992 6.3 (3/4.12) RADIOLOGICAL ENVIRONMENTAL MONITORING 6.3.3 (3/4. 12.3) 'NTERLABORATORY COMPARISON PROGRAM,.

RE UIREHENT FOR OPERABILITY 6.3.3. 1 (3. 12.3) Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by Table 6.3.1.1-1 (3.12-1).

RELEVANT CONDITIONS: At all times.

COMPENSATORY MEASURES:

a ~ 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.

b. The provisions of Technical Specifications 3.0.3 and 3.0.4 are not applicable.

PERIODIC TESTS AND INSPECTIONS 6.3.3.1.1 (4. 12.3) The Interlaboratory Comparison Program shall be described in the ODCH. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.

162

AMENDMENT NO: 9 JANUARY 1992,

6. 4 RADIOLOGICAL ENV IRONHENTAL OPERATING/RADIOACTIVE EFFLUENT RELEASE REPORT RE(UIRENENTS CONTROL OF CHANGES TO THE:

RADIOACTIVE L I (U ID ~ GASEOUS ~ AND SOLID WASTE TREATMENT SYSTEMS 163

AMENDMENT,NO. 9 E

JANUARY 1992 6.4.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT Routine Radiological Environmental Operating Reports covering the operation of the uni.t during the previous calendar year shall be submitted prior to May 1 of each year.

The Annual Radiological 'Environmental Operating Reports shall interpretations, and an analysis of trends of the results of the include'ummaries, radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational controls as appropriate, and with previous'nvironmental surveillance reports, and an assessment of the observed impacts of the plant operation of the environment.

The reports shall also include the results of Land Use Censuses required by Requirement for Operability 6.3.2. 1 (3.12.2).

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM; as well as summa-rized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event, that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following: a summary description of the radiological environmental monitoring program, at least two legible maps*

covering all sampling locations keyed to a table giving distances and direc-tions from the centerline of the reactor; the results of license participation in the Interlaboratory Comparison Program, required by Requirement for Opera-bility 6.3.3.1 (3. 12.3); discussion of all deviations from the sampling schedule of Table 6.3. 1. 1-1 (3. 12-1); and discussion of all analyses in which the LLD required by Table 6.3.1. 1. 1-1 (4. 12-1) was not achievable.

One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

164

AMENDMENT NO. 9 JANUARY 1992 6.4.2. RADIOACTIVE EFFLUENT RELEASE REPORT The Routine Radioactive Effluent Release Report covering the operation of the unit shall'e submitted in accordance with 10 CFR 50.36a(a)(2).

The Radioactive Effluent Release Report shall include a summary of the quan-tities of radioactive liquid and gaseous effluents and solid waste released from 'the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid 'Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on' quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorolog-ical data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (ODCM Figure 3-1) during the report period. All assumptions used in making these assessments, i.e., specific activity, expo-sure time and location, shall be included in these reports. The meteorolog-ical conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCH).

The Radioactive Effluent Release Report shall also include once a year an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1. 109, Rev. 1, October 1977.

In lieu of submission with the first half year Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

165

AMENDMENT NO. 16 DECEMBER 1993 6.4.2 RADIOAC IVE EFFLUENT RELEASE REPORT (Continued)

The Radioactive Effluent Release Report shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:

a. Container volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g., dewatered spent resin,'ompacted dry waste, evaporator bottoms'),

e. Type of container (e.g., LSA, Type A, Type B, Large guantity), and f., Solidification agent or absorbent (e.g., cement, urea formaldehyde).

The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCH), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Requirement for Operability 6.3.2. 1 (3. 12.2).

6.4.3 MAJOR CHANGES TO RADIOACTIVE LI UID GASEOUS AND SOLID WASTE TREATMENT SYSTEMS*

Licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid):

a. Shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the POC. The discussion of each change shall contain:

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;

2. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

Licensees may choose to submit the information called for in this specification as part of the annual FSAR update.

166

AMENDMENT NO. 9 JANUARY 1992

3. A detailed description of the .equipment, components, and processes involved and. the interface with other plant systems;

4. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or 'quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

5. An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto;

6. A comparison of the predicted releases of radioactive mate-rials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

7. An estimate of the exposure to plant operating personnel as a result of the change; and-

8. Documentation of the fact that the change was reviewed and found acceptable by the POC.

b. Shall become effective upon review and acceptance by the POC.

Licensees may choose to submit the information called for in this specification as part of the annual FSAR update.

167

AMENDMENT NO. 9 JANUARY 1992'.5 BASES FOR RADIOACTIVE EFFLUENTS MONITORING REQUIREMENT FOR OPERABILITY 168

AMENDMENT NO. 9 JANUARY 1992

'B6. 1 INSTRUMENTATION BASES HON ITOR I NG INSTRUMENTATION B6.1.1 (3/4.3.7.11) RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION 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 and adjusted in accordance with the methodology and parameters in the ODCH to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the require-ments of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The purpose of tank level indicating devices is to assure the detec-tion and control of leaks that if not controlled, could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS.

86.1.2 (3/4.3.7.12) RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 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 and adjusted in accordance with the methodology and parameters -in the ODCH to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the require-ments of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

169

AHENDHENT NO. 9 JANUARY 1992 B6.2 (3/4. 11) RADIOACTIVE EFFLUENTS BASES 66.3.6 63/4.663 ~LI TB EFFL E I B6.2.1.1 (3/4.1.1.1) CONCENTRATION This Requirement for Operability is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation 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 HEHBER OF THE PUBLIC and (2) the limits of 10 CFR 20. 106(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the con-trolling radioisotope and its HPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

This Requirement for Operability applies to the release of radioactive materials in liquid effluents from all reactor units at the site.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).

Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Hanual, HASL-300 (revised annually), Currie, L. A., "Limits for gualitative Detection and guantitative Determination - Application to Radiochemistry," Anal. Chem. 40 586-93 (1968), and Hartwell, J. K.,

"Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARM-SA-215 (June 1975).

B6.2.1.2 3 4.11.1.2 DOSE This Requirement for Operability is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Require-ment for Operability implements the guides set forth in Section II.A of Appendix I. The COHPENSATORY HEASURES 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 liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reason-ably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCH 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 HEHBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCH for calculating the doses due to the actual release rates of radioactive materials 170

AMENDMENT NO. 9 JANUARY 1992 (3/4.tl.l.2) 'OE t'.2.1.2 i" d in liquid effluerits're 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.

This Requirement for Operability applies to the release of radioactive materials in liquid effluents from each reactor unit at the site.

B6.2. 1.3 (3/4. 11. 1.3) LI UID RADWASTE TREATMENT SYSTEM The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluent will be kept "as low as is reasonably achievable." This Requirement for Operability implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. 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.

This Requirement for Operability applies to the release of radioactive mate-rials in liquid effluents from each reactor unit at the site.

B6.2.2 (3/4. 11.2) GASEOUS EFFLUENTS B6.2.2. 1 (3/4. 11.2. 1) DOSE RATE This Requirement for Operability is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column l. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR 20. 106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, is provided in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/year to the total body or to less than or equal to 3000 mrems/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year.

171

AMENDMENT NO. 10 JANUARY 1992 B6.2.2. ~ 1 (3/4. 11.2.

~ ~ ~ 1) DOSE 'RATE Continued This Requirement for Operability applies to the release of radioactive .

materials in gaseous effluents from all reactor units at the site. ~

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs).

Detailed discussio'n of the LLD and other detection limits can be found in HASL Procedures Manu'al, HASL-300 (revised annually), Currie, L. A., "Limits for Qualitative Detection and Quantitative Determination - Application to Radio-chemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K. , "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

B6.2.2.2 (3/4. 11.2.2) DOSE - NOBLE GASES This Requirement for Operability is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Require-ment for Operability implements the guides set forth in Section II.B of Appendix I. The COMPENSATORY MEASURES 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 to UNRESTRICTED AREAS will be kept "as low as is reason-ably achievable." The PERIODIC TESTS AND INSPECTIONS requirements 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 appro-priate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCH 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 Han 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.

The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

This Requirement for Operability applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site.

B6.2.2.3 (3/4.11.2.3) DOSE - IODINE- 131 IODINE- 133 TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM This Requirement for Operability is provided to implement the requirements of Sections II.C, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Require-ment for Operability are the guides set forth in Section II.C of Appendi.x I.

The COMPENSATORY MEASURES statements provide the required operating flexi-bility 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 to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCH calculational methods specified in the Requirement for Operability implement the requirements in Section III.A of Appendix I that 172

AMENDMENT NO. 9 JANUARY 1992 B6.2.2.3 (3/4.11:2.3) DOSE - IODINE-131 'IODINE-133 TRITIUM AND RADIONUCLIDES'N PARTICULATE FORM Continued 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 under-estimated. The ODCH calculational methodology and parameters. 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 Han 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 prov'ide for determining the actual doses bas'ed upon the historical average atmospheric conditions or concurrent meteorology. The release rate specifications for iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at and 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.

This Requirement for Operability applies to the release of radioactive materials in gaseous effluents from each reactor unit at the site.

173

AMENDMENT NO. 9 JANUARY 1992 86.2 RADIOACTIVE EFFLUENTS BASES B6:2.2.4 and 6.2.2.5 (3/4. 11.2.4 and 3/4. 11.2.5) GASEOUS OFFGAS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM The OPERABILITY of the GASEOUS OFFGAS RADWASTE TREATMENT SYSTEM and the VENTI-LATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. 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 reason-ably achievable." This Requirement for Operability implements the require-ments of 10 CFR 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.

B6.2.2.6 (3/4.11.2.8) VENTING OR PURGING This Requirement for Operability provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of

~

10 CFR Part 20 for unrestricted areas. ~

~

86.2.3.1 (3/4.11.3)

~ ~ ~ ~ ~ SOLID RADIOACTIVE WASTE This Requirement for Operability implements the requirements of 10 CFR 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing 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 constit-uents, mixing and curing times.

174

AMENDMENT NO. 9 JANUARY 1992 B6.2 RADIOACTIVE EFFLUENTS BASES B6.2.4.1 (3/4.11.4) ~TOTAL DOS This Requirement for Operability is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR 'Part 20 by 46 FR 18525. The Requirement for Operability requires the preparation and submittal of a Special Report whenever the calculated doses from plant

,generated radioactive effluents and direct radiation exceed 25 mrems to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. 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 twice the dose design objecti,ves of Appendix I, and if direct radiation doses, from the reactor units and outside storage tanks are kept small. The Spec'ial Report will describe a course of action that should result in the limitation of 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 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 Part 190 until NRC staff action is com-pleted. 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 limitation of 10 CFR Part 20, as addressed in Requirement for Operability 6.2. l. 1 (3. 11. 1.) 1 and

. 6 .2.2. 1 (3 . 11.2. 1) . 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.

175

AMENDMENT NO. 9

.'ANUARY 1992

~ .

B6.3 (3/4.12) RADIOLOGICAL 'ENVIRONMENTA MONITORING BASES B6.3.1.1 (3/4.12.1)" MONITORING PROGRAM The radiological environmental monitoring program required by this Require-ment for Operability provides representative measurements of radiation and of radioactive materials in those exposure pathways and for'those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environ-mental Monitoring. The initially specified monitoring program will be effec-tive for at least the first 3 years of commercial operation. 'Following this period, program changes may be initiated on operational experience.

The required, detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 6.3. 1.1. 1-1 (4. 12-1) are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an h priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

Detailed discussion on the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., "Limits for gualitative Detection and guantitative Determination - Application to Radiochemistry," Anal. Chem. 40 586-93 (1968), and Hartwell, J. K.,

"Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June'975).

B6.3.2. 1 (3/4. 12.2) LAND USE CENSUS This Requirement for Operability is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the radiological environmental monitoring program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey or from consulting with local agricultural authorities shall be used. 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 50 m'rovides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1. 109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m'.

176

AMENDMENT NO. 9 JANUARY 1992

,C 86.3 (3/4.12)

~ ~ RADIOLOGICAL ENVIRONMENTAL MONITORING BASES B6.3.3.1 (3/4.12.3) INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environ-mental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.

177

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