ML17285B062

Amend 7 to Odcm.

ML17285B062
Person / Time
Site:	Columbia
Issue date:	02/26/1990
From:	WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:
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ML17285B059	List: ML17285B058 ML17285B060 ML17285B062
References
PROC-900226, NUDOCS 9003090248
Download: ML17285B062 (234)

Text

AMENDMENT NO. 7 December 1989 WNP-2 OFFSITE DOSE CALCULATION MANUAL (ODCN)

List of Effective Pages Pacae Amendment lv CHAPTER 1 CHAPTER 2 7

2a 7 7

4 3 5 3 6 6 7 6 7

7 10 3 ll 3 12 3 12a 3 12b 5 13 0 14 0 15 6 16 3 17 6 18 6 LEP-1 900SO9O248 900226 PDR ADOCK 05000397 R PNU

AMENDMENT NO. 7 December 1989

~Pa e Amendment 19 3 19a 3 20 3 21 3 22 3 23 3 24 2 25 2 26 6 27 6 28 5 28a 5 29 7 30 6 31 6 32 6 32a 6 32b 6 CHAPTER 3 33 7 33a 7 34 7 35 7 36 7 37 6 38 7 39 7 40 7 41 7 42 7 43 7 7

45 3 LEP-2

ANENONENT NO. 7 Oecember 1989 Pa<ac Amendment 46 3 47 6 48 6 49 0 50 7 51 0 52 7 53 7 54 3 54a 5 55 7 56 7 57 6 58 0 59 7 60 3 61 7 62 7 63 6 63a 7 64 7 65 7 66 7 67 7 68 3 69 3 70 7 71 6 72 3 73 0 74 0 75 0 76 0 77 0 LEP-3

AMENDMENT NO. 7 Oecember 1989

~Pa e Amendment 78 0 79 0 80 0 81 0 82 0 83 0.

84 0 85 0 86 7 87 0 88 7 89 7 90 1 90a 6 90b 6 CHAPTER 4 91 91 a CHAPTER 5 92 1 93 3 94 7 95 7 96 4 97 4 98 7 99 7 100 7 101 3 102 4 103 3 LEP-4

AMENDMENT NO. 7 December 1989

~Pa e Amendment 104 5 105 3 106 6 107 4 108 0 109 0 110 3 ill 7 LEP-5

Controlled Copy No.

WASHINGTON PUBLIC POWER SUPPLY SYSTEM WNP-2 OFFSITE 00SE CALCULATION MANUAL

AMENDMENT NO. 7 December 1989 OFFSITE DOSE CALCULATION MANUAL TABLE OF CONTENTS Section Title ~Pa e

1.0 INTRODUCTION

2.0 LIQUID EFFLUENT DOSE CALCULATION................ 2 2.1 Introduction . 2 2.2 Radwaste Liquid Effluent Radiation Monitoring System . ~ 2a 2.3 10 CFR 20 Release Rate Limits 3 2.3.1 Pre-Release Calculation 3 2.3.2 Post-Release Calculation .

2.3.3 Continuous Release . 5 2.4 10 CFR 50, Appendix I, Release Rate Limits . . . . . . . . . . . 6 2.4.1 Projection of Doses . . . . . . . . . . . . . . . . . . . . . . 9 2.5 Radwaste Liquid Effluent Dilution Ratio and Alarm Setpoints Calculations . .-9 2.5.1 Introduction . 9 2 '.2 Methodology for Determining the Maximum Permissible Concentra-.

tion (NPC) Fraction. 10 2 '.3 Methodology. for the Determination of Liquid Effluent Honitor Setpoint .

2.6 Verification of Compliance with 10 CFR 50, Appendix I, and 10 CFR 20, Appendix B. 12b 2.7 Methods for Calculating Dose to Man from Liquid Effluent p athways . 12b 2.7.1 Radiation Doses 13 2.7.2 Plant Parameters . 17 2.8 Compliance with Technical Specification 3.11.1.4 . 19 2.8.1 Maximum Allowable Liquid Radwaste Activity in Temporary Radwaste Hold-Up Tanks . 19 2.8.2 Maximum Allowable Liquid Radwaste in Tanks That Are Not Surrounded by Liners, Dikes, or Walls 22 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) Nonitor . 24 2.9.3 Turbine Building Sumps Water (FD) Monitor 24

AMENDMENT NO. 3 February 1986 Section Title ~Pa e 3.0 GASEOUS EFFLUENTS DOSE CALCULATION . . 33 3.1 I ntroduci ion e ~ ~ ~ ~ e ~ ~ ~ e ~ ~ ~ ~ ~ ~ e ~ ~ ~ e ~ ~ . 33 3.2 Gaseous Effluent Radiation Monitoring System . ~ ~ . 34 3.2.1 Main Plant Release Point . . . . . . . . . . . ~ ~ . 34 3.2.2 Radwaste Building Ventilation Exhaust Monitor ~ ~ . 35 3.2.3 Turbine Building Ventilation Exhaust Monitor . ~ ~ . 36 3.3 10 CFR 20 Release Rate Limits ~ ~ . 36 3.3.1 Noble Gases ~ ~ . 37 3.3.2 Radioiodines and Particulates ~ ~ . 37 3.3.2.1 Dose Parameter for Radionuclide i (Pl) ~ ~ 40 3.4 10 CFR 50 Release Rate Limits ~ ~ . 41 3.4.1 Noble Gases (Technical Specification 3.11.2.2) . . . . . . ~ ~ . 42 3.4.2 Radioiodines and Particulates (Technical Specification 3.11.2.3) . . . . . . . . . . . . ~ ~

3.4.2.1 Dose Parameter for Radionuclide i (R; ) ~ ~ . 47 3.4.3 Annual Dose at Special Locations . . . . . . . ~ ~ . 54 3.5 Compliance with Standard Technical Specifications 3.11.2.4 . 54 3.6 Calculation of Gaseous Effluent Monitor Alarm Setpoints ~ ~ . 54a 3.6.1 Introduction . . . . . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . 54a 3.6.2 Setpoint Determination for All Gaseous Release Paths... ~ ~ . 55 3.6.2.1 Setpoints Calculations Based on Whole Body Dose Limits . ~ ~ . 55 3.6.2.2 Setpoints Calculations Based on Skin Dose Limits . . . . ~ ~ . 58 4.0 COMPLIANCE WITH 40 CFR 190 . . 91 4.1 Technical Speci fication Requirement............ . 91 4.2 ODCM Methodology for Determining Dose and Dose Commitment from Uranium Fuel Cycle Sources. . . . . . . ~ ~ . 91 4.2.1 Total Dose from Liquid Effluents . ~ ~ . 91a 4.2.2 Total Dose from Gaseous Effluents. ~ ~ . 9la 4.2.3 Direct Radiation Contribution. . . . . 91a 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING . 92 5.1 Radiological Environmental Monitoring Program (REMP) . . 93 5.2 Land Use Census . 94 5.3 Laboratory Intercomparison Program . . . . . . . . . 95 5.4 Reporting Requirements... . 96 6.0 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT. .110

AMENDMENT NO. 7 December 1989 OFFSITE DOSE CALCULATION MANUAL LIST OF TABLES Tabl e Ti tl e ~Pe e Section 2.0 2-1 Fish Bioaccumulation Factors (BF;) and Adult Ingestion Dose Conversion Factors (DF;). . . . . . . . . . . . ~ . . . . 26 2-2 Ingestion Dose Factors (A;>) for Total Body and Critical Organ . 29 2-3 Input Parameters Used to Calculate'Max imum Individual Dose From Liquid Effluents ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 32 Section 3.0 3-1 Dose Factors for Noble Gases and Daughters . 60 3-2 Distances (Miles) to Typical Controlling Locations as Measured from Center of MNP-2 Containment Building ..... 61 3-3 HNP-2 Long-Term Average Dispersion (X/g) and Deposition (0/0)

Values for Typical Locations . 62 3-4 Dose Rate Parameters. Implementation of 10 CFR 20, Airborne Releases 63 3-5a Dose Rate Parameters. Implementation of 10 CFR 50, Airborne Releases Age Group: Adult 64 3-5b Dose Rate Parameters. Implementation of 10 CFR 50, Airborne Releases Age Group: Teen 65 3-5c Dose Rate Parameters. Implementation of 10 CFR 50, Airborne Releases Age Group: Child 66 3-5d Dose Rate Parameters. Implementation of 10 CFR 50, Airborne Releases Age Group: Infant . 67 3-6 Input Parameters for Calculating R. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 68 1

3-7 Input parameters for Calculating R. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 69 1

3-8 Input Parameters for Calculating R ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 70 1

AMENDMENT NO. 7 December 1989 Title ~Pe e Input Parameters Needed .for Calculating Dose to the Haximum Individual from WNP-2 Gaseous Effluent . 71 3-10 Reactor Building Stack X/Q and D/Q Values 73 3-11 Turbine Building X/Q and D/Q Values 77 r

3-12 Radwaste Building X/Q and D/Q Values . 81 3-13 Characteri st i cs of WNP-2 Gaseous Ef fluent Release Points... 85 3-14 References for Values Listed in Table 3-9 86 3-15 Design Base Percent Noble Gas (30-Minute Decay) 87 3-16 Annual Doses at Typical Locations Source: WNP-2 Gaseous Effluent 88 3-17 Annual Air Dose at Typical Locations . 89 Section 5.0 Radiological Environmental Honitoring Program Plan . ~ ~ ~ ~ ~ 98 WNP-2 REHP Locations . 102 Environmental Radiological Honitoring Program Annual Summary . 108 5-4 Reporting Levels for Nonroutine Operating Reports . . . . . . 109 LIST OF FIGURES

~Pi ere Title ~Pa e 2-1 Simplified Block Diagram of Liquid Waste System 32a 2-2 Simplified Block Diagram of Solid Radwaste System 32b 3-1 Site Boundary for Radioactive Gaseous and Liquid Effluents 90 3-2 Simplified Block Diagram of Gaseous Waste System . 90a 3-3 Simplified Block Diagram of Off-Gas Treatment System . 90b 5-1 Radiological Environmental Honitoring Sample Locations Inside of 10-Mile Radius . 106 Radiological Environmental Honitoring Sample Locations Outside of 10-Hi le Radius 107 iv

1. 0 INTRODUCTION

'he purpose of this manual is to provide the information and methodologies to be used by the Washington Public Power Supply System to ensure compliance with the dose requirements stated in the WNP-2 Effluents Technical Specifications.

ANENONENT NO. 7 Oecember 1989 2.0 LI UIO EFFLUENT DOSE CALCULATION The U.S. Nuclear Regulatory Commission's computer program LAOTAP 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. This design objective will be kept at all times. 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 efflu-ents released to the unrestricted areas will be determined once every 31 days using the LAOTAP II computer code. The maximum exposed individual is assumed to be an adult whose exposure pathways include potable water and fish consump-tion. The choice of an adult as the maximum exposed individual is based on the highest fish and water consumption rates shown by'hat age group and the fact that most of the dose from the liquid effluent comes from these two pathways.

AMENDMENT NO. 7 December 1989 The dose contributions will be calculated for all radionuclides identified in the released effluent. The calculations are based on guidelines provided by Nureg-0133 and the LAOTAP II computer code.

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

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

2a

AMENDMENT NO. 7 December 1989 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, (average, based on operating data is 808 gpm) and is discharged to the Columbia River with a total flow based on HPC; total, and cooling water flushing needs.

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

2.3 10 CFR 20 Release Rate Limits The requirements pertaining to discharge of radwaste liquid effluents to the unrestricted area are specified in Technical Specification 3.11.1.1:

"The concentration of radioactive material released from the site to unrestricted areas shall be limited to the concentrations specified in 10 CFR 20, Appendix 8, Table II, Column 2 for radionuclides other than noble gases, and 2 x 10 4q Ci/ml total activity concentra-tion for all dissolved or entrained noble gases."

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 Ca 1 culati on The activity of the radionuclide mixture will be determined in accordance with Supply System procedure PPH 12.5.3, Liquid Effluent Discharge Determination.

Liquid effluent discharge is determined and calculated according to PPH 12.11.1, Radiological Effluent Honitoring Gaseous and Liquid. The effluent concentration is determined by the following equation:

Ci xFw Ci ft

AMENOMENT NO. 3 February 1986 where:

ConC,. Concentration of radionuclide i in the effluent at point of discharge - qCi/ml.

P Ci Concentration of radionuclide i in the batch to be released - pCi/ml.

fw = Discharge flow rate from sample tank to the b'lowdown line - variable from 0 to 190 gpm.

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

ft = Total discharge flow rate - (ft = fb + fw)

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

g (ConC,.

1=1 WPC,. 1 (2) where:

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

MPCi Maximum permissible concentration of nuclide i as listed in 10 CFR 20, Appendix 8, Table II.

m = Total number of radionuclides in the batch.

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

AMENDMENT NO. 3 February 1986 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:

Cik x fw (33 elk where:

ConCik = 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 - (qCi/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.

ft = Total discharge lft = fb + fwl flow rate - variable from 0 to 7690 gpm.

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

g (ConC,. k/tlPC,.

i=1 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

~ ~ ~ ~

AHENDHENT NO. 6 November 1988 unrestricted area would be calculated according to Equation (3) and Equa-l 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 Technical Specification 4.11.1.2 requires that the cumulative dose contribu-tions be determined in accordance with the ODCH at least once per 31 days.

Technical Specification 3.11.1.2 specific's that the dose to a member of the public from radioactive material in liquid effluents'eleased to the unre-stricted area shall be limited to:

< 1.5 mrem/Calendar quarter Total Body and

< 5.0 mrem/Calendar quarter Any Organ.

The cumulative dose For the calendar year shall be limited to:

< 3 mrem Total Body and

< 10 mrem Any Organ. ~

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

D< =g(A,. ~ pat Ci F ) (5) i a=1 where:

Dz The cumulative dose commitment to the total body or organ,w, from liquid effluents for the total time period in mrem.

g 4tR E=]

AMENDMENT NO. 6 November 1988 The length of the ath time period over which C.15 and Fg, are averaged for all liquid releases, in hours.

The number of releases for the time period under consideration.

The average concentration of radionuclide, i, in undiluted liquid effluent during time periodht4 from any liquid release, in pCi/ml.

The site-related ingestion dose commitment factor to the total body or any organ ~ for each identified principle gamma and beta emitter listed in Table 2-2, in mrem/hr per uCi/ml.

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

Li uid Radioactive Waste Flow fw Discharge Structure Exit Flow x 500 ft x 500

~ tel <<Pl. ~ i *

't( t l hdtv @be 4 i 4-' . V ' * ~ 5 I AMENDMENT NO. 7 December 1989 The term

)T ,

A. the ingestion dose factors for the total body and critical organs, are tabulated in Table 2-2. It embodies the dose factor, fish bioac-cumulation 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:

Uw (7) where:

A)lT The composite dose parameter for total body or criti-cal organ of an adult for nuclide i (in mrem/hr per qCi/ml).

K A conversion factor:

1.14E+05 = (10 6 ~pCi/ Ci) x (10 3 ml/liter) hr/yr.

~

8760 Uw 730 li.ter/yr which is the annual water consumption by the maximum adult (Table E-4 of Regulatory Guide 1.109, Revision 1).

Fi 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).

Fi Adult ingestion dose conversion factor for nuclide i Total body or critical organ (mrem/pCi) (Table E-ll of Regulatory Guide 1.109, Revision 1 and NUREG/CR-4013).

'w Dilution factor from near fie1d area (within one-quarter mile of the release point) to the Richland potable water intake - 100.

UF Adult fish consumption, 21 kg/yr (Table E-5 of Regulatory Guide 1.109, Revision 1)..

8

AMENDMENT NO. 7 December 1989 The values of BFi and OF; are listed in Table 2-1. Oilution assumptions, calculations, and LAOTAP II input parameters are provided in Radiological Programs Calculation Log 88-3.

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 Ooses The projected doses due to releases of WNP-2 radwaste liquid,effluents will be calculated for each batch, using equation 5. If the sum of the accumulated dose to date for the month and the projected dose for the remainder of the month exceeds the technical specification 3.11.1.3 limits, then the liquid radwaste treatment system shall be used. This is to ensure compliance with Standard Technical Specification 3.11.1.3. This technical specification 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.

2.5 Radwaste Li uid Effluent Oi lution 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 8, Table II, Column 2, are not exceeded in the effluent at the discharge point (i.e.,

compliance with Standard Technical Specification 3.11.1.1, as discussed in section 2 '.1 of this manual).

AMENDMENT NO. 3 February 1986 The trip/alarm setpoint for the liquid radwaste effluent monitor is calculated from the results of the radiochemical analysis of the waste solution. The setpoint 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 (MPC)

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 MPC fraction (MPCf) based on 10 CFR 20 limits. From the sample analysis and the MPC values in 10 CFR 20, the MPCf is determined using the following equation.

m

'PCf 1 (8)

.i =1 i where:

MPCf Total fraction of the Maximum Permissible Concentra-tions (hlPCs) in the liquid effluent waste sample.

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

10

AMENDMENT NO. 3 February 1986 MPC. The limiting concentrations 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.

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

If the MCPf is less than or equal to 0.8, the liquid batch may be released at any radwaste discharge or blowdown rate. If the MPCf 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 Methodolo for the Determination of Li uid Effluent Monitor Set oint 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 Technical Specification 3.11.1.1 are met at the point of. discharge.

The dilution factor (Fd) is determined according to:

C ~

x Fs (9)

Where:

The dilution factor required for compliance with 10 CFR 20, Appendix 8, Table II, Column 2:

AMENDMENT NO. 3 February 1986 C. The concentration of each radionuclide (i) observed by radiochemical analysis of the liquid waste sample (qCi/ml).

MPC- The limiting concentration of the appropriate radionuclide (i) from 10 CFR 20, Appendix B, 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 (%) variation.

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

The dilution

~ ~ ~

which is required to ensure compliance with Technical

~ ~ ~

Specification 3.11.1.1 concentration limits will be set such that discharge

~

~ ~ ~ ~

rates are:

Fd ~

fw+ fb fw (10) and follows that:

fw- fb (loa) or

. fb ~ fw(Fd-1) (lob)

Where:

The dilution factor from equation 9.

12

NENDMENT NO. 3 February 1986 fw The discharge flow rate from the liquid radwaste tank to the blowdown line - variable from 0 to 190 gpm.

fb 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 the radiation monitor's alarm/trip setpoint is;;

SP = C + BKg + K [C + Bkg] /

Where:

SP Radiation monitor setpoint (count rate) i=1 (C; x Efi) represents the count rate from the radionuclides in the liquid radwaste.

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

Same as for equation 9.

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

12a

AMENDMENT NO. 5 April 1988 BKg Background count rate of the radwaste effluent monitor.

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

2e6 Verification of Com liance with 10 CFR 50 A endix I and 10 CFR 20 Appendix8 Verification of compliance with 10 CFR 50, Appendix I, and 10 CFR 20, Appen-dix B, limits wi 11 be achieved by following NNP-2 Plant Procedures for liquid discharge and the periodic application of the LADTAP II computer code.

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."

12b

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 R . = 1100 PZ U Mp Q,.D ,. exP(-X,.tp) (13) 1

b. Aquatic Foods U

R apj

. = 1100 ~ZQ.B.

F Mp

,. i ip 0 ..exp(-K.t aipj i p) (14)

c. Shoreline Deposits R

apj

. =

110,000 U M W QQ.T.D 1

i i aipj (exp(-X.t L i p) (1 - exp(-X.t )1 i bg (15)

d. Irrigated foods For all radionuclides except tritium:

r[l - exp(-XE.t )] fIB. [1 - exp(-~.t+)]

rLl - exp( aEite ]

ap

~ iA aipj }Fd,exP( With) y y 1 v Ei f IB,.r.l - exp(-h,.tb)]

(16)

P X.

1 i Aw~Aw 13

l1'

~

2.7.1 Radiation Ooses Radiation doses from potable water, aquatic food, shoreline deposit, and irrigated food pathways will be calculated by using the following equations:

a. Potable Water R

apj

. = 1100 ~Z(}.0 U

F Mp

..exp(-) i.t i aipj p

) (13)

b. Aquatic Foods R . = 1100 PZ0,.B,.

U Mp D, exp(-T,.t ) (i4) 1

c. Shoreline Oeposits R . = 1'10,000 ~g U M W Q,.T,.D, [exp(-X,.t ) (1 - exp(-X,.t>)] (15)

d. Irrigated foods For all radionuclides except tritium:

r[l - exp( ~Eite ~ I iv[1

- exp( ~i 4 j Ra

= U Z diexp(-h.th 1

r[l - exp(-XE.t )j 1 v Ei fIB,. [l - exp(-,.tb)]

P X.

1 i Aw~Aw 13

For tritium:

where:

1P The equilibrium bioaccumulation factor for. nuclide"i in pathway p, expressed as the ratio of the concen-tration in biota (in pCi/kg) to the radionuclide .

concentration in water (in pCi/liter), in liters/kg.

8. The concentration factor for uptake of radionuclide i 1V from soil by edible parts of crops, in pCi/kg (wet weight) per pCi/kg dry so'il.

C.

iAw The concentration of radionuclide i in water consumed by animals, in pCi/liter.

1V The concentration of radionuclide i in vegetation, in pCi/kg.

D alps 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, expressed as a rat'io of the dose rate (in mrem/hr) and the areal radionuclide concentration (in pCi/m ).

ANENOHENT NO. 6 November 1988

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

The flow rate of the liquid effluent, in ft /sec.

FT The fraction of the year crops are irrigated, dimensionless.

iA The stable element transfer coefficient that'elates the daily intake rate by an animal to the concen-tration in an edible portion of animal product, in pCi/liter (milk) per pCi/day or pCi/kg (animal pro-duct) 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 2 (Table E-15, Regulatory Guide 1.109, Revision 1).

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

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

Q; The release rate of nuclide i, in Ci/yr.

The fraction of deposited activity retained on crops, dimensionless (TabTe E-15, Regulatory Guide 1.109, Revision 1).

NENDMENT NO. 3 February 1986 The total annual dose to organ j of individuals of apJ age group a from all of the nuclides i in pathway p, in mrem/yr.

tb 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 contamina-tion during the growing season, in hours (Table E-15, Regulatory Guide 1.109, Revision 1).

h 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 r each the point of exposure. For internal dose, tp 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 ap intake rate for an individual of age group a associ-ated with pathway p, in hr/yr, g/yr, or kg/yr (Table E-5, Regulatory Guide 1.109, Revision 1).

16

ANENONENT NO. 6 November 1988 leased liquids to reach Richland, approximately 12 miles downstream, is esti-mated at 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Richland is the "realistic case" location, and doses cal-culated 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 MNP-2 site is affected by the liquid effluents released. There is no significant commercial fish harvest in the 50-mile radius region around MNP-2. Sportfish harvest is estimated 'at 14,000 kg/year.

For irrigated foods exposure pathways, it can be assumed that production with-in 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:

~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 2.0 130 Source terms are measured based on sampled effluent.

Table 2-3 summarizes the LAOTAP II input parameters. Oocumentation and/or calculations of these parameters are discussed in detail in R.P.I. 2.3, and Rad. Prog. calculation Log 88-3.

18

AMENDMENT NO. 3 February 1986 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-Up Tanks The use of temporary liquid radwaste hold-up tanks is planned for MNP-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 ODCN 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 Il.

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 ( 10) 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 be determined to be within the limit by E

analyzing a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank.

Kd T (18) where:

Total allowed activity in tank (curies).

A. Activity of radioisotope i (curies).

19

AMENDMENT NO. 3 February 1986 MPC. Haximum permissible concentration of radionuclide i (10 CFR 20, Appendix 8, 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 ~.

A fi =gni

~

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

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

19a

s Q1ENDHENT NO. 3 February 1986 The total allowed activity (AT) is based on limiting MNP-1 well water to less than 1 NPC. 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 maximum 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 (A ) is conservatively based on all liquid radwaste in the tank instantaneously reaching the water table.

The hydrological analysis performed for the MNP-2 FSAR (Section 2.4) deter-mined that the transit time through the ground water from MNP-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 (CWi) is simply the con-centration in the tank (CT.) adjusted for radioactive decay during transit

-Xt (e ) and divided by the minimum concentration reduction factor (CRF min).

Limiting well concentration to 1 HPC yields:

i l

=Z~ f"i min i i "2.4 fN- FAR 1/2 CRF .

min

= (4 < L) x Y g (20) 20

AMENDMENT NO. 3 February 1986 where:

L = Migration distance = 1 mile.

V = Volume of tank.

a ~ a r a = Dispersion constants.

x v z Combining Equations 19 and 20 yields:

CT 1

2V e i 1 (21)

(4~ L) (a~ a a ) MPC.

Substituting Ai for CTi Y and reorganizing terms yields:

(4mL) 3/2 (ax ay o )

1/2 A,

(22)

MPC. e ~i .

Making the following substitutions

=fi AT (4 n L) 3/2 (aaa )

1/2

= x 10 Ci/pCi = 2.4 x 10 Ci per pCi (23)

Kd CC 21

NENDMENT NO. 3 February 1986 yields:

d TZ()p~ +Et HPC,-e or Kd MPCie 2.8.2 Maximum Allowable Liquid Radwaste in Tanks That Are Hot Surrounded by Liners, Dikes, or Mails Although permanent outside liquid radwaste tanks which are not surrounded by liners, dikes, or walls are not planned for MNP-2, Equation 18 will be used should such tanks become necessary in the future.

~ ~

2.99

~ Li uid Process Monitors and Alarm Setpoints Calculations 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 level of the Radwaste Building. All WHP-2 radwaste liquid effluent is discharged to the Columbia River through the 36-inch Cooling Mater Blowdown line. In addition to the liquid effluent discharge monitor there are three liquid streams that are normally non-radioactive but have a finite possibility of having radioactive material injected into them.

These liquid streams are:

Standby Service Mater (SW)

Turbine Building Service Mater (TSW)

Turbine Building Sump Mater (FD) 22

Y AMENDMENT NO. 3 February 1986 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 10CFR20 limits. The maximum radiation detector setpoint calculation for the three systems is based on i concentration of Cs-'l37 which is 2.05-05 pCi/sIl. The fo11ow-ing 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 2.9.1 Standby 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-].5 gpm.

To ensure 10CFR20 limits are never exceeded, the alarm setpoint shall be established at 80% or less of the maximum setpoint plus background.

'f 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. 2 February 1985 Turbine Buildin Service Mater (TSM) 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 10CFR20 limits are never'xceeded, 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.

Turbine Buildin Sum s Mater (FD) Monitor - There are three detec-tors to measure the activity of each of the three non-radioactive sumps. The monitors are located on the 441'evel of the Turbine Building. The readout meters and recorder are located in the Rad-waste Control Room Panel BD-RAD-41.

The Turbine Building Sump Water Effluents are not released to the Columbia River. This effluent is discharged to the Storm Drain System which is an open pond by the WNP-2 Warehouse.

The hydrological analysis performed for the WNP-2 FSAR (Section 2.4) determined that the transmit time through the ground water from WNP-2 to the WNP-1 well is 67 years for strontium and 660 years for cesium.,

In the event the setpoint is exceeded, the sump water will be automatically routed to the radioactive waste system.

24

AMENDMENT NO. 2 February 1985 To prevent the sum of the sump water discharged from the three pumps from exceeding 10CFR20 limits, the alarm/setpoint will be estab-lished at 80% or less of the maximum setpoint plus background.

25

ANENOMENT NO. 6 November 1988 The shoreline width factor, dimensionless (Table A-2, Regulatory Guide 1.109, Revision 1).

Y The agricultural productivity (yield), in kg (wet weight) /m (Table E-15, Regulatory Guide 1.109, Revision 1).

XEi The effective removal rate constant for radionuclide.

i from crops, in hr,-1 where x E. = X . + x w' El 1

. is the radioactive decay constant, and x is 1 w the removal rate constant for physical loss by weathering (Regulatory Guide 1.109, Revision 1, Tab 1 e B-l 5) .

The radioactive decay constant of nuclide i, in hr 1100 The factor to convert from (Ci/yr)/(ft /sec) to pCi/1 i ter.

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

~ANP-2 is a river shoreline site with a variable effluent discharge flow rate 0 to 7500 gpm. The population center nearest NNP-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 17

AMENDMENT NO. 6 November 1988 Table 2-1 FISH BIOACCUMULATION FACTORS (BF.)

1 AND ADULT INGESTION DOSE CONVERSION FACTORS (DF.)

1 Dose Conversion Factor (DF 1 )

Fish Bioaccumulation Total GI Nuc 1 ide Factor (BF,-) Body Bone Thyroid Liver Tract (pCi/kg per (mRem per pCi Ingested) pCi/liter)

H-3 9.0E-01 6.0E-OB (3) 6.0E-OB 6.0E-OB 6.0E-OB 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 Hn-54 4.0E+02 8.7E-07 (3) (3) 4.6E-06 1.4E-05 Hn-56 4.0E+02 2.0E-OB (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 5.0E+01 4.7E-06 (3) (3) 2.1E-06 4.0E-05 Ni-65 1.0E+02 3.1E-OB 5.3E-07 (3) 6.9E-OB 1.7E-06 Cu-64 5.0E+01 3.'9E-08 (3) (3) 8.3E-OB 7.1E-06 Zn-65 2.0E+03 7.0E-06 '.8E-06 (3) 1.5E-05 9.7E-06 Zn-69m 2.0E+03 3.7E-OB 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-OB (3) (3) (3) 5.8E-OB Br-84 4.2E+02 5.2E-OB (3) (3) (3) 4.1E-13 Rb-89 2.0E+03 2.8E-OB (3) (3) 4.0E-OB 2.3E-21 Sr-89 3.0E+01 B.BE-06 3.1E-04 (3) (3) 4.9E-05 Sr-90 3.0E+01 1.8E-,04 8.7E-03 (3) (3) 2.2E-04 26

AMENOMENT NO. 6 November 1988 Table 2-1 (contd.)

Oose Conversion Factor (OF; )

Fish Bioaccumulation Total GI Nuclide Factor (BF,.) Body Bone Thyroid Liver Tract (pCi/kg per (mRem per pCi Ingested) pCi/liter)

Sr-91 3.0E+01 2.3E-07 5.7E-06 (3) (3) 2.7E-05 Sr-92 3.0E+Ol 9.3E-OB 2.2E-06 (3) (3) 4.3E-05, Y-90 2.5E+01 2.6E-10 9.7E-09 (3) (3) 1.0E-04 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 (3) (3) 7.8E-05 Y-92 2.5E+01 2.5E-11 8.5E-10 (3) (3) 1.5E-05 Y-93 2.5E+Ol 7.4E-11 2.7E-09 (3) (3) 8.5E-05 Zr-95 3.3E+00 6.6E-09 3.1E-OB (3) 9.8E-09 3.1E-05 Nb-95 3.Of~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 1.1E-04 Nb-97 3.0E+04 4.8E-12 5.2f-ll (3) 1.3E-11 4.9E-OB Mo-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+01 3.6E-09 2.5E-10 (3) 3.7E-10 1.1E-21 Ru-103 1.0E+Ol 8.0E-OB 1.9E-07 (3) (3) 2.2E-05 Ru-105 1.0E+01 6.1E-09 1.5E-OB (3) (3) 9.4E-06 Rh-105 1.0E+Ol 5.8E-OB 1.2E-07 (3) 8.9E-OB 1.4E-05 Ru-106 1.0E+Ol 3.5E-07 2.8E-06 (3) (3) 1.8E-04 Ag-110m 2.3E+00 8.8E-OB 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-OB B.OE-05 Sb-125 1.0E+00 4.3E-07 1.8E-06 1.8E-09 2.0E-08 2.0E-05 Sb-126 1.0E+00 4.2E-07 1.2f-06 7 'E.09 2.3E-OB 9.4E-05 Sb-127 1.0E+00 9.9E-OB 2.6E-07 3.1E-09 5.7E-09 5.9E-05 Te-127 4.0E+02 2.4E-OB 1.1E-07 8.2E-OB 4.0E-OB 8.7E-06 Te-129m 4.0E+02 1.8E-06 1.2E-05 4.0E-06 4.3E-06 .

5.8E-05'.7E-09 Te-129 4.0E+02 3.1E-OB 2.4E-OB 1.2E-OB 2.4E-OB 27.

AHENOMENT NO. 5 April 1988 Table 2-1 (contd.)

Oose Conversion Factor (OF; )

Fish Bi oaccumul ation Total GI Nuclide Factor (BF;) Body Bone Thyroid Liver Tract (pCi/kg per (mRem per pCi Ingested) pCi/1 i ter)

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-OB 1.6E-OB 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+01 7.5E-07 1.4E-06 3.6E-04 2.5E-06 2.2E-06 I-1 34 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 2.0E+03 1.2E-04 6.2E-05 (3) 1.5E-04 2.6E-06 Cs-136 2.0E+03 1.9E-05 .6.5E-06 (3) 2.6E-05 2.9E-06 Cs-137 2.0E+03 7.1E-05 8.0E-05 (3) 1.1E-04 2.1E-06 Cs-138 2.0E+03 5.4E-OB 5.5E-OB (3) 1.1E-07 4.7E-13 Ba-139 4.0E+00 2.8E-09 9.7E-OB (3) 6 'E-11 1.7E-07 Ba-140 4.0E+00 1.3E-06 2.0E-05 (3) 2.6E-OB 4.2E-05 La-140 2.5E+01 3.3E-10 2.5E-09 (3) 1.3E-09 9 'E-05 La-141 2.5E+01 1.6E-11 3.2E-10 (3) 9.9E-11 1.2E-05 La-142 2.5E+Ol 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-OB 4.9E-07 (3) 2.0E-07 1.7E-04 Pr-143 2.5E+Ol 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-05 Hf-181 3.3E+00 4.3E-06 (3) (3) (3) 4.1E-05 W-185 1.2E+03 1.4E-OB 4.1E-07 (3) 1.4E-07 ~ 1.6E-05 28

AMENDMENT NO. 5 April 1988 Table 2-1 (contd.)

Dose Conversion Factor (DF,.)

Fish Bioaccumulation Total GI Nuclide Factor (BF;) Body Bone Thyroid Liver Tract (pCi/kg per (mRem per pCi Ingested) pCi/liter)

H-187 1.2E+03 3.0E-OB 1.0E-07 (3) 8.6E-OB 2.8E-05 Np-239 1.0E+Ol 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.

(3)

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

28a

AMENDMENT NO. 7 December 1989 Table 2-2 INGESTION DOSE FACTORS (A ) FOR TOTAL BODY ANO CRITICAL ORGAN (in mrem/hr per Ci/ml)

Liquid Effluent Total Gi Nuc 1 i de ~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.BE+06 4.6E+07 2.9E+06 5.3E+06 Cr-51 1.3E+00 7.7E-01 3.2E+02 Hn-54 8.3E+02 4.4E+03 1.3E+04 Hn-56 1.9E+01 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.DE+03 2.4E+03 8.2E+03 Co-58 2.1E+02 9.0E+Ol 1.BE+03 Co-60 5.7E+02 2.5E+02 4.8E+03 Ni-65 7.5E+00 1.3E+02 1.7E+Ol 4.1E+02 Cu-64 4.7E+00 1.0E+01 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 1.1E+04 Br-82 2.3E+03 2.6E+03 Br-83 4.0E+01 5.BE+01 Br-84 5.2E+01 4.1E-04 Rb-89 1.3E+02 1.9E+02 1.1E-11 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+01 4. 1E+02 2.0E+03 Sr-92 6.BE+00 1. 6E+02 3.1E+03 29

I AMENDMENT NO. 6 November 1988 Table 2-2 (contd.)

Total Gi Nuclide ~Bod Bone ~Th roid Liver Tract 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-Ol 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 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-01 3.5E+03 Mo-99 2.0E+01 1.1E+02 2.5E+02 Tc-99m 3.3E-01 9.2E-03 2.6E-02 1.5E+01 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-Ol 3.7E-01 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-01 2.6Ew02 Sb-125 1.4E+00 5.8E+00 5.8E-03 6.5E-02 6.5E+01 Sb-126 1.4E+00 3.9E+00 2.3E-02 7.4E-02 3.0E+02 Sb-127 3.2E-01 8.4E-Ol 1.0E-02 1.8E-02 1.9E+02 Te-127 2.3E+Ol 1.1E+02 7.9E+Ol 3.8E+Ol 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+01 2.3E+01 1.2E+Ol 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+01 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 2.0E+01 3.7E+00.

I-133 2.8E+01 5.1E+Ol 1 ~ 3E+04 9.2E+Ol 8.1E+Ol I-134 3.7E+00 4.0E+00 1.8E+02 1.1E+01 9.2E-03 I-135 1.6E+01 1.6E+Ol 2.8E+03 4.4E+01 4.8E+01 30

AMENDMENT NO. 6 November 1988 Table 2-2 (contd.)

Total Gi Nociide ~Bod Bone ~Th roid Liver Tract Cs-134 5.BE+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.BE+05 5.3E+05 1.0E+04 Cs-138 2.6E+02 2.6E+02 5.3E+02 2.3E-03 Ba-139 2.9E-02 1.0E-OO 7.2E-04 1.BE+00 Ba-140 1.4E+Ol 2.1E+02 2.7E-Ol 4.4E+02 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-1 41 2.3E-03 3.0E-02 2.0E-02 7. 7EI01 Ce-143 4.5E-04 5.5E-03 3.9E+00 1.5E+02 Ce-144 8.4E-02 1.6E+00 ** 6.5E-Ol 5.5E+02 Pr-143 2.8E-02 5.6E-01 2.3E-01 2.4E+03 Nd-147 2 'E-02 3.8E-01 4.4E-01 2elEt03 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 W-187 8.6E+01 2.9E+02 2.5E+02 8.1E+04 Np-239 1.6E-03 3.0E-02 3.0E-03 6.0E+02

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

31

AMENDMENT NO. 6 November 1988 TABLE 2-3 INPUT PARAHETERS USED TO CALCULATE MAXIMUM INDIVIDUAL DOSE FROM LI UID EFFLUENTS Drinkin Water River Dilution: 50,000 River Transit Time: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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: 2,000 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: Adul t 90 hr/yr Teenager 500 hr/yr Child 105 hr/yr

. Infant 0 Swimming: Adul t 18 hr/yr Teenager 100 hr/yr Child 21 hr/yr Irri ated Foodstuffs River Dilution: 50,000 River Transit Time: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Leafy

~Ve etables Hilt Heat ~Ve etables Food Delivery Time: 14 days 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 20 days 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Usage Factors:

Adul t 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/yr 330 1/yr 41 kg/yr 26 kg/yr Monthly Irrigation Rate: 180 1/m~ 200 1/m2 160 1/m2 200 1/m2 Annual Yield: . 5.0 kg/m2 1 3 1/m2 2.0 kg/m2 1.5 kg/m2 Annual Growing Period: 70 days 30 days 130 days 70 days Annual 50-Mile Production: 3.5E+09 kg 2.BE+08 L 2.3E+07 kg 1.9E+06 kg 32

AMENDNFNT NO. 6 November 1988 SUMPS COOLING TOVER VASTE VASTE BLOWDQWN LINE RADVASTE BLDG SURGE TURBINE BLDG SAMPLE AND TANKS DRYWELL COLLECTOR (TVO)

TANKS FLOOR DRAIN FLOOR COLLECTOR DRAIN MISC VASTY TANK SAMPLE TANK REACTOR BLDG DISTILLATE TANKS (TWO)

DETERGENT.

DRAIN TANKS FILTERS 8 DEMINERAL-IZERS (SOLID VASTE)

CHEMICAL VASTE TANKS SHOP DECON CONDENSATE CHEM PUMPS STORAGE DECQN DRAIN PLANT TANKS REACTOR BLDG USE (TVO)

TURBINE BLDG COLUMBIA RIVER SIMPLIFIED BLCICK DIAGRAM GF LIQVID WASTE SYSTEM FIGURE 2-1 32a

AMENDMENT NO. 6 November 1988 DEMINERALIZERS Condensate RWCU Phase Dewatering Disposal EDR/FDR Separator Liners Site Radwaste Bead Ory Active Compactor. Disposal Waste Site SIMPLIFIED BLOCK DIAGRAM OF SOLID RADWASTE SYSTEM Figure 2-2 32b

~ 4 ~r AMENDMENT NO. 7 December 1989 3.0 GASEOUS EFFLUENTS OOSE 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.

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 instantan-eous technical specification requirement at the site boundary.

Figure 3-1 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 00E.

2. OOE train two railroad lines pass through the site (approximately 3 miles of line). According to 00E, 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.

33

AMENDMENT NO. 7 December 1989

3. BPA Ashe Substation occupied 2080 hour0.0241 days <br />0.578 hours <br />0.00344 weeks <br />7.9144e-4 months <br />s/year. These people are not normally controlled by the Supply System but are involved in activi-

'ties directly in support of MNP-2.

4. WNP-2 Supply System Visitor Center assumed occupied 8 hrs/yr by non-Supply System individuals.

5. WNP-1 occupied 2080 hrs/yr. This location is controlled by the Supply System. However, activities are not in direct support of MNP-2.

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

33a

AMENDMENT NO. 7 December 1989 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 and radwaste buildings. Figure 3-3 provides a simplified block diagram for the Off-Gas Treatment System.

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 to 3-12). These doses are listed in Tables 3-16 and 3-17 along with the doses to the maximum exposed individual.

The most likel ex osed member of the ublic 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/Q and D/Q values.

3.2 Gaseous Effluent Radiation Monitorin S stem 3.2.1 Main Plant Release Point The Main Plant Release is instrument monitored for gaseous radioactivity prior to discharge to the environment via the main plant vent release point. Par-ticulates and iodine activity are accumulated in filters which will be changed and analyzed as per Technical Specification 4.11.2.1.2 and Table 4.11.2. The effluent is supplied from: the gland seal exhauster, mechanical vacuum pumps, treated off gas, standby gas treatment, and exhaust air from the entire reactor building's ventilation.

Two 100-percent capacity vanaxial fans supply 80,000 CFM 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 low range beta scintillator, an intermediate range beta scinti llator and two ion chamber LOCA monitors. The beta'cinti 1-lators are mounted in thick 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 7 cpm. The intermediate 34

\

AMENDMENT NO. 7 Oecember 1989

-2 range has a response from 10 -10 3 pCi/cc Xe -133 equivalent, and reads in panel meter units (PMU) with a meter range of 10 -10 PMU.

The readouts and recorder are located in the main control room panel BD-RAO-24. Power is provided from 125 VOC divisional buses. This monitor has no control function but annunciates in the main control room. The alarm will initiate proper action as defined in the WNP-2 Plant Procedures.

3.2.2 Radwaste Bui ldin Ventilation Exhaust Monitor The radwaste building ventilation exhaust monitoring system monitors the radio-activity 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 prefi lter, a high efficiency particulate air filter (HEPA) and two centrifugal fans. Total exhaust flow will vary as the combined exhaust unit maintains a radwaste building differential pressure of -0.25 inches H20 to the environment.

Particulate and iodine air sample filters are changed weekly for laboratory 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. 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 7 cpm. The intermediate range has a response From 10 10 3 pCi/cc Xe -133 equivalent, and reads in panel meter units (PMU) with a meter range of 0 5 10 -10 PHU. The readouts and recorder are located in the main control room panel BD-RAO-24. Power is provided from 125 VOC 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.

35

AMENDMENT NO. 7 December 1989 3.2.3 Turbine Bui ldin 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 centri-fugal 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 passedthrough 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

-2 3 range has a response from 10 10 pCi/cc Xe -133 equivalent, and reads in panel meter units (PMU) with a meter range of 10 0 10 5 PHU. The mon-itors are on the 525'evel of the radwaste building and the readouts and the recorder are located in the main control room panel 80-RAD-24. Power is provided from the 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.

3.3 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 Technical Specification 3.11.2.1. 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 < 500 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 <1500 mrem/yr to any organ."

36

AMENOHENT NO. 6 November 1988 3.3.1 Noble Gases In order to comply with Technical Specification 3.11.2.1, the following equa-tions must hold:

Whole body:

Z Ki [(X/Q) 0. + (X/Q) 0. )] < 500 mrem/yr (1)

Skin g [(Li + 1.)H.)((X/0) Q. + (X/0) 0. )j < 3000mrem/yr (2) l 3.3.2 Radioiodines and Particulates Part "b" of Technical Specification 3.11.2.1 requires that the release rate limit for all radioiodines and radioactive materials in particulate form and radionuclides other than noble gases must meet the following relationship:

Any organ:

g P. Wt g. + W g. < 1500 mrem/yr (3)

The terms 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 uCi/m ).

L,. = The skin dose factor due to beta emissions for each iden-tified noble gas radionuclide i (mrem/yr per vCi/m ).

37

AMENDMENT NO. 7 December 1989 M) The air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad/yr per wCi/m (unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose).

Pi The dose parameter for all radionuclides other than noble gases for the inhalation pathway, (mrem/yr per p Ci/m )

and for food and ground plane pathways, m (mrem/yr per p Ci/sec). The dose factors are based on the critical individual organ and the most restrictive age group.

Qim 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).

Q,.

= The release rate oF radionuclide.i in gaseous effluent from all ground level releases (uCi/sec).

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

(X/Q) (sec/m 3 ). 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.

38

AMENDMENT NO. 7 December 1989 W The highest calculated annual average dispersion parameter g

for estimating the dose to an individual at the control-ling location due to all ground level releases.

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

W = m . For ground plane pathways. The location g

is at and beyond the site boundary in the sector of maximum concentration.

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

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

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

The factors, L; and H., relate the radionuclide airborne concentrations to dose rates assuming a semi-infinite cloud. These factors are listed 1'arious in Table B-l of Regulatory Guide 1.109, Revision 1, and in Table 3-1 of this manual.

The values used in the equations for the implementation of'echnical Specifi-cation 3.11.2.1 are based upon the maximum long-term annual average X/Q at and beyond the site boundary. Atmospheric dispersion factors will be evaluated 39

AMENDMENT NO. 7 December 1989 annually from the l<NP-2 meteorological data base and if significant different than preoperational data as is displayed in Tables 3-10 through 3-12, then the tables will be updated. This comparison will begin with 1989 data'able 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/Q and D/Q values which may be used if current annual averages are not available.

The X/Q and D/Q values listed in Tables 3-10 through 3-12 reflect correct ac-quired meteorological data up to 1983 and were utilized in the initial GASPAR Computer runs. Subsequent meteorological reports will use updated X/Q and D/Q averages Characteristic of NNP-2 gaseous effluent release points as listed in Table 3-13.

3.3.2.1 Dose Parameter for Radionuclide i (P.) l The dose parameters used in Equation 3 are based on:

l. Inhalation and ground plane. (Note: Food pathway is not applicable to NNP-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 P,.I ( Inhalation): The following equation will be used to calcu-late P. ( Inha 1 a t i on) .

1 P.

I (Inhalation) = KA (BR) DFA. (mrem/yr pervCi/m )

40

AMENDMENT NO. 7 December 1989 where:

A 6 K = A constant of conversion, 10 ppCi/ Ci.

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

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

The inhalation dose factor for OFA; 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:

P. = (Inhalation) = (3.7 x 10 )(OFA.) (mrem/yr per p Ci/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 Technical Specifications 3.11.2.2 and 3.11.2.3.

Technical Specification 3.11.2.2 deals with the air dose from noble gases and requires that the air dose at and beyond the site boundary due to noble gases released in gaseous effluents shall be limited to the following:

(a) "Ouring any calendar quarter, to < 5 mrad for gamma radiation and to < 10 mrad for beta radiation."

(b) "During any calendar year, to <10 mrad for gamma radiation and <20 mrad for beta radiation."'1

/t' AMENDMENT NO. 7 December 1989 Technical Specification 3.11.2.3 deals with radioiodines, tritium, and radio-active materials in particulate form, and requires that the dose to an indi-vidual 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) "Ouring any calendar quarter, to < 7.5 mrem."

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

3.4.1 Noble Gases Technical S ecification 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

~ Z H. (X/Q) Q.i + (X/q) q.i + (X/Q) Q. + (X/q) q. < 5 mrad (8) l During any calendar quarter, for beta radiation:

3.17 x 10 Z N. (X/Q) Q; + (X/q) q. + (X/Q) Q. + (X/q) q. <10 mrad (9)

b. Ouring any ca 1 enda r year, for gamma radi at i on:

3.17 x 10 Z H. (X/Q) Q. + (X/q) q. + (X/Q) Q. + (X/q) q. <10 mrad (10) 1 42

AMENDMENT NO. 7 December 1989 During any calendar year, for beta radiation:

3.17 x 10 Z N. (X/Q) Q. + (X/q) q. + (X/Q) Q. + (X/q) q. < 20 mrad (11) i where:

H.

1

= The air dose factor due to gamma emmissions for each identi f i ed noble gas radi onuc1 ide, in mrad/yr per p Ci/m (H. values are i listed in Table 3-1).

N.

1

= The air dose factor due to beta emissions for each iden-tified noble gas radionuclide, in mrad/yr per pCi/m (N-1 values are listed in Table 3-1).

(X/Q) 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)g 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 )

(X/Q) For partially elevated release points'he highest 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 )

43

AMENDMENT NO. 7 December 1989 qim 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 year, as appropriate.

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

Qim 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 >Ci. Release shall be cumulative over the calendar quarter or year, as appropriate.

Q. The average release of noble gas radionuclides 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.

-8 3.17 x 10 = The inverse of the number of seconds in a year.

3.4.2 Radioiodines Tritium and Particulates Technical S ecification 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 in gaseous effluents released to the unrestricted areas:

• 6 - 'b I *>> I ~ \ < w ll C A <<\'L4'I NENDNENT NO. 3 February 1986

a. During any calendar quarter:

3.17 x 10 Z R. I M iL mim 0. + w q.

mim

+ i}. gig + w q. gig < 7.5 mrem

b. During any calendar year:

3.'}7 x 10 Z R. i }} i} m im

                                        + .w m

q.im + }} g i}i ig

                                                               + w g

q. ig

                                                                           < 15 mrem          (13) where:

The releases of radionuc}ides, 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 cumu-lative over the calendar quarter or year, as appropriate (m is for mixed mode releases, g is for ground level releases). qim <<g The releases of r adionuclides, radioactive materials in particulate form, and radionuclides 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).

hhhf ~ rt=ddhhhl ~ hh.h-hAhh'h ~ -" h ~ +h NENDHENT NO. 3 February 1986 W III ~ W The dispersion parameter for estimating the dose to an g individual at the controlling location for long-term ( ) 500 hr.) releases (m is for mixed mode releases, g is for ground level releases). W = (77/) for the inhalation pathway, in sec/m . W = (57/) for the food and ground plane pathways in meters The dispersion parameter for estimating the dose to an individual at the controlling location for short-term (( 500 hr.) releases (m is for mixed mode releases, g is for ground level releases). w = (Y/q) for the inhalation pathway, in sec/m 3 . w = (57q) for the food and ground plane pathways in meters

              -8 3.17 x 10         = The    inverse of the number of seconds in       a year.

R,.

                 = The dose 2

factor for each identified radionuclide, i, in 3 m (mrem/yr per pCi/sec) or mrem/yr per pCi/m . 46

AMENDMENT NO. 6 November 1988 3.4.2.1 Dose Parameter for Radionuclide i (R.) 1 The R.. i values used in equations 12 and 13 of this section are calculated separately for each of the following potential exposure pathways: o Inhalation Ground plane contamination Grass-cow/goat-mi lk 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 (Inhalation R. Pathway Factor) I (Inhalation) I R. = (mrem/yr per pCi/m 3 )

                                                                   ~

K (BR) (OFA.) (14) where: RI The inhalation pathway factor (mrem/yr per pCi/m ). 1 A constant of unit conversion, 10 pCi/pCi. (BR) The breathing rate of the receptor of age group (a) in meter 3 /yr. (Infant = 1400, child = 3,700, teen = 8,000, adult = 8,000. From P.32 NUREG-0133).

w' a' -"8 la ~z AN'l>> ~ 'aw ace ( y ~ ~ ~s w ~ t ' +, 5 ~ << >>~ AMENOMENT NO. 6 November 1988 3.4.2.1 Oose Parameter for Radionuclide i (R.) 1 The i values R,. used in equations 12 and 13 of this section are calculated separately for each of the following potential exposure pathways: o Inhalation Ground plane contamination Grass-cow/goat-milk pathway Grass-cow-meat. pathway Vegetation pathway Monthly dose assessments for NNP-2 gaseous effluent wi 11 be done for all age groups. Calculation of R.l (Inhalation Pathway Factor) R. I (Inhalation) = K (BR) (OFA.) (mrem/yr per qCi/m ) (14) where: RI The inhalation pathway factor (mrem/yr per pCi/m ). 1 6 A constant of unit conversion, 10 pCi/pCi. (BR) The breathing rate of the receptor of age group (a) in meter 3 /yr. (,Infant = 1400, child = 3,700, teen = 8,000, adult = 8,000. From P.32 NUREG-0133). 47

t AMENDMENT NO. 6 November 1988 (OFA. ) 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;)a. (OFAi)a values are listed in Tables E-7 through E-10 of Regulatory Guide 1.109 manual, Revision 1 and NUREG/CR-4013. Values of R.I are listed in 1 Table 3-5. G Calculation of R. (Ground Plane Pathway Factor) G

                    =  A 8 R,.(Ground Plane)      K K   (SF)(DFG,.) (1-e      )/X,. (m    x mrem/yr per  uCi/sec) (15) where:

G R. = Ground plane pathway factor (m x mrem/yr per uCi/sec). A = K A conversion constant of (10 pCi/>Ci). 8 = K A conversion constant (8760 hr/yr). The decay constant for the ith radionuclide (sec ). t = Exposure time, 6.31 x 10 8 sec (20 years). OFGi The ground plane dose conversion factor for the ith radio-nuclide, 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 l. 48

G The values of R. 1 are listed in Table 3-5 of this manual. C Calculation of R. 1 (Grass-Cow/Goat-Milk Pathway factor) C R. (Grass-Cow/Goat-Milk Factor) = 1 qF(u ) L.) ff (1-ff)e if (16) m i a Y Y (m x mrem/yr per i@i/sec) where: 6 A constant of unit conversion, 10 pCi/>Ci. The cow/goat consumption rate, in kg/day, (wet weight). U ap The receptor's milk consumption rate for age (a), in liters/yr. Y The agricultural productivity by unit area of pasture feed 2 grass, in kg/m . Y s The agricultural productivity by unit area of stored feed, in kg/m2 The stable 'element transfer coefficients, in days/liter. r = Fraction of deposited activity retained on feed grass.

AMENDMENT NO. 7 December 1989 (OFLi) The maximum organ ingestion dose factor for the ith radio-nuclide for the receptor in age group (a), in mrem/pCi (Tables E-ll to E-14 of Regulatory Guide 1.109, Revision 1 and NUREG/CR-4013). The decay constant for the ith radionuclide, in sec The decay constant for removal of activity on leaf and plant surfaces by weathering, 5.73 x 10 -7 sec -1 (cor-responding to a 14-day half-life). tf = The transport time from pasture to animal, to milk, to receptor, in sec. th = The transport time from pasture, to harvest, to animal, to milk, to receptor, in sec. fP = Fraction of the year that the cow/goat is on pasture (dimensionless). f = 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 for the fraction of elemental iodine available for deposition. The input parameters used for calculating C R. 1 are listed in Table 3-6 and the M R. l values are tabulated in Table 3-7. For Tritium: C In calculating RT, pertaining to tritium in milk, the airborne concentration rather than the deposition will be used: C R T (Grass-Cow/Goat-Milk Factor) = K K F 0 U (OFL.) 0.75(0.5/H) (mrem/yr per qCi/m ) (17) 50

where: K = A constant unit conversion, 10 pCi/pCi. K = A constant of unit conversion, 10 gm/kg. H = Absolute 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. M Calculation of R. 1 (Grass-Cow-Meat Pathway Factor) M R. (Grass-Cow-Meat Factor) =

            ~

1

          , qF(u   )
                                                                   -X.tf 1

K F<(r)(DfL,.) (18) 1 w (m~ x mrem/yr per uCi/sec) where: I K = A constant unit conversion, 10 pCi/pCi. The stable element transfer coefficients, in days/kg. The receptor's meat consumption rate for age (a), in kg/yr. 51

AMENDMENT NO. 7 December 1989 tf = The transport time from pasture to receptor, in sec. The transport time from crop field to receptor, in sec. NOTE: For radioiodines, multiply R.M value by 0.5 to account for the fraction of elemental iodinh available for deposition. The input parameters needed for solving equation 18 are listed in Table 3-7. For Tritium: In calculating the RT for tritium in meat, the airborne concentration is used rather than the deposition rate. The following equation is used to calculate the R T values for tritium: M R T (Grass-Cow-Meat Pathway) = K K FfQFU (DFL.) 0.75(0.5/H) (mrem/yr per uCi/m ) Where the terms are as defined in equations 16-18, R,. values for tritium pertaining to the infant age group is zero since there is no meat consumption by this age group,. Calculation of i (Vegetation R. Pathway factor) V l (Vegetation Pathway Factor) R. = ULf 1L USf > h aL ag (2o) 2 (m x mrem/yr per >Ci/sec) 52

vqAlew,,w,be ilk+.ysvtse'<<A<< ~ AMENDMENT NO. 7 December 1989 where: I 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.

U a

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

fL = The fraction of the annual intake of fresh leafy vegetation grown 1 oca1 ly. fg = The fraction of the annual intake of stored vegetation grown locally. tL = The average time between harvest of leafy vegetation and its consumption, in seconds. th = The average time between harvest of stored vegetation and its consumption, in seconds. Y v

          =  The   vegetation area density, in kg/m 2 .

NOTE: For radioiodines, multiply R. value by 0.5 to account for the fraction of elemental iodinh available for deposition. All other items are as defined in equations 16-18. 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 fol-lowing equation is used to calculate R for tritium:

0 i

AMENDMENT NO. 3 February 1986 V RT (Vegetation Pathway Factor) = K K U f + U f (DFL.) 0.75(0.5/H) (mrem/yr per pCi/m ) (21) Where all terms have been defined above and in equations 16-18, the R.i value for tritium is zero for the infant age group due to zero vegetation consump-tion rate by that age group..The input parameters 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 radio-active 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 loca-tions 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 hours per year due to it being the higher of the two locations. 3.5 Com liance with Standard Technical S ecification 3.11.2.4 Standard Technical Specification 3.11.2.4 states:

            "The GASEOUS RADWASTE TREATMENT SYSTEM     shall be in opera-tion 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 3.3.7.11-1."
            "APPLICABILITY: Whenever the main condenser             jet air P'""." steam

AHENDHENT NO. 5 April 1988 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 coolant design base 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 Technical Specification 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 Honitor Alarm Set pints 3.6.1 Introduction The following procedure 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 54a

AMENDMENT NO- 7 December 1989 mrem/yr to the whole body or 3000 mrem/yr to the skin. The initial setpoints determination is calculated using a conservative radionuclide mix obtained from the WNP-2 GALE code. Once the plant is operating and sufficient measur-able 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 Oetermination 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 technical specifications to ensure compliance with 10 CFR 20 and 10 CFR 50 Appendix 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 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 technical specification. Both skin dose and whole body setpoints wi 11 be calculated and the lower limit will be used. 3.6.2.1 Set pints Calculations Based on Whole Bod Dose Limits The fraction ( << i) 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: ra, ij (dimensionless) ,(22) HT where: N 1J

         ~   =  The measured  individual concentration of radionuclide i in the gaseous  effluent release path j (>Ci/cc).

55

AMENDMfNT NO. 7 Oecember 1989 HT. Tj

            =  The measured    total concentration of all noble gases identified in the gaseous effluent release path j (uCi/cc).

Based on Technical Specification 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 m (vCi/sec) (23) X/Q. Z= (K,.)( i 1 where: QT>

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

nicall Fj = Fraction of total dose allocated to release path j. 500 = Whole body dose rate limit of 500 mrem/yr as specified in Tech-Speci f ication 3.11.2. la. X/Q. Haximum 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. Hain plant vent release values are for mixed'ode and may be either short term or average annual value dependent upon type of release. K. i

           =   The  total  whole body dose       factor  due  to gamma  emission from noble gas nuclide i (mrem/yr per uCi/m ) (as listed in Table B-l of Regulatory Guide 1.109, Revision 1).

56

AMENDMENT NO. 6 November 1988 ij As defined in equation 22. m = Total number of radionuclides in the gaseous effluent. j = Different release pathways. The total maximum acceptable concentration (CT.) of noble gas radionuclides Tj in the gaseous effluent release path j (MCi/cc) will be calculated by using the following equation: C . Tj

                                      = ~ (uCi/cc)

QT. Rj (24) where: CT. TJ

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

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

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

To determine the maximum acceptable concentration (Ci.) of noble gas radio-ij nuclide i in the gaseous effluent for each individual noble gas in the gaseous effluent (uCi/cc), the following equation will be used: ij = ~,ijCT.Tj Ci (qCi/cc) (25) 57

where: and CT< are as defined in equations 22 and 24 respectively, the gaseous effluent monitor alarm setpoint will then be calculated as follows: m C.R.j = p C..E..(cpm) (26) where: C.R.j = Count rate above background (cpm) for gaseous release path j. The maximum acceptable concentration of noble gas nuclide i in the gaseous effluent release path j. >Ci/cc. lj Detection efficiency of the gaseous effluent monitor j for noble gas i (cpm/I.Ci/cc). 3.6.2.2 Setpoints Calculations Based on Skin Dose Limits The method for calculating the setpoints to ensure compliance with the skin dose limits specified in Technical Specification 3.11.2.1a is similar to the one described for whole body dose limits (Section 3.6.2.1 of this manual), except Eq. 27 will be used instead of Eq. 23 for determining maximum accept-able release rate (QT-). Tj Fj 3000 (X/Q.)~ " ' (27) QTJ j i~ =1 (L. 1

                                          + 1.1N 1
                                                   )(   lj )

58

AMENDMENT NO. 7 December 1989 where: gT>

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

X/gj = The maximum annual normalized diffusion coefficient for release path j at and beyond the site boundary (sec/m ). F. = Fraction of total allowed dose. j The skin dose factor due to beta emission for each identified noble gas radionuclide i in mrem/yr per > Ci/m (L. values are listed in Table 3-1). The air dose factor due to gamma emmissions for each identified noble gas radionuclide, in mrad/yr per qCi/m (N. values 1 are listed in Table 3-1). 1.1 = A conversion factor to convert dose in mrad to dose equivalent in mrem. 3000 = Skin dose rate limit of 3000 mrem/yr as specified in Technical Specification 3.11,2.1. 59

able 3-1 DOSE FACTORS FOR HOBLE GASES AHD DAUGHTERS* Total Body Gamma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor Radionuclide K. L~ Hi H ~ (mrem/yr per pCi/m ) (mrem/yr per pCi/m ) (mrad/yr per pCi/m ) (mrad/yr per pCi/m ) Kr-85m 1.17E+03** 1 . 46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+Ol 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 3.27Et02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11 E+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

    *The  listed   dose  factors are for radionuclides that        may be   detected in gaseous effluents.
  ~7.56E.-.02    = 7.56 x 10 The values    listed   above were taken from Table B-1 6                        -1 of  HRC  Regulatory Guide 1.109, Revision
                                                                         -1

l. The values were multiplied by 10 to convert picocuries to microcuries

AMENDMENT HO. 7 December 1989 Table 3-2 DISTANCES (MILES) TO TYPICAL CONTROLLING LOCATIONS AS MEASURED FROM CENTER OF WNP-2 CONTAINMENT BUILDING" Location Distance Sector Dose Pathwa s (miles) Site Boundary 1.2. SE Air dose measurement. One 4.2 ESE Ground, vegetables, and inhalation. Two 6.4 SE Ground, vegetables, meat, cow milk, and inhalation. Three 4.8 SE Ground, vegetables, and inhalation. Four 4.1 ENE Ground, vegetables, and inhalation. Five 4.3 NE Ground and inhalation.

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

61

Tab -3 NNP-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 leted (miles) (sec/m ) (sec/m ) (sec/m ) (m+) Site Boundary SE 1.2 Reactor Bldg. 4.8E-07 4.8E-07 4.6E-07 7.6E-09 Turbine Bldg. 1.8E-06 1.8E-06 1.6E-06 1.5E-OB Radwaste Bldg. 1.8E-06 1.8E-06 1.6E-06 1.5E-OB One ESE 4.2 Reactor Bldg. ,7.9E-OB 7.7E-OB 6.4E-OB 5.3E-10 Turbine Bldg. 7.8E-OB 7.7E-OB 5.9E-OB 5.3E-10 Radwaste Bldg. 7.8E-OB 7.7E-OB 5.9E-OB 5.3E-10 Two SE 6.4 Reactor Bldg. 5.7E-OB 5.6E-OB 5. 1E-08 3. 2E-10 Turbine Bldg. 5.8E-OB 5.7E-OB 4.1E-OB 3.1E-10 Radwaste Bldg. 5.8E-OB 5.7E-OB 4.1E-OB 3.1E-10 Three SE 4.8 Reactor Bldg. 9.5E-OB 9.3E-OB 8.7E-OB 5.8E-10 Turbine Bldg. 9.6E-OB 9.5E-OB 7.2E-OB 5.8E-10 Radwaste Bldg. 9.6E-OB 9.5E-OB 7.2E-OB 5.8E-10 O% Four ENE 4.1 Reactor Bldg. 3.0E-OB 2.9E-OB 2.4E-OB 3.1E-10 (D O Vl (D M Turbine Bldg. 2.9E-OB 2.9E-OB 2.2E-OB 3.2E-10 3 O e M CT m Radwaste Bldg. 2.9E-OB 2.9E-OB 2.2E-OB 3.2E-10 LD M 00 lg) O

                                                                                                            ~

Five NE 4.3 Reactor Bldg. 2.7E-OB 2.7E-OB 2.5E-OB 2.9E-10 Turbine Bldg. 3.4E-OB 3.3E-OB 2.6E-OB 3.7E-10 Radwaste Bldg. 3.4E-OB 3.3E-OB 2.6E-OB 3.7E-10

AMENDMENT NO. 6 November 1988 Table 3-4 DOSE RATE PARAMETERS IMPLEMENTATION OF 10 CFR 20 AIRBORNE RELEASES Child Dose Factor* PI 1 DFA. DFGi Inhalation mrem/hr m~rem/ r Nuc1 ide sec" ~mrem/ Ci ~Ci /m ~Ci /m 3 H-3 1.8E-09 1.7E-07 0.0 6.3E+02 Na-24 1.3E-05 4.4E-06 2.9E-OB 1.6E+04 Cr-51 2.9E-07 4.6E-06 2.6E-10 1.7E+04 Hn-54 2.6E-OB 4.3E-04 6.8E-09 1.6E+06 Hn-56 7.5E-05 3.3E-05 1.3E-OB 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-OB 7.0E+06 Cu-64 1.5E-05 9.9E-06 1.7E-09 3.7E+04 Zn-65 3.3E-08 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-OB 2.1E+04 Sr-89 1.5E-07 5.8E-04 6.5E-13 2.2E+06 Sr-90 7.9E-IO 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 1.3E-06 1.1E-09 4.BE+03 Ru-106 2.2E-OB 3.9E-03 1.8E-09 1.4E+07 Ag-110m 3.2E-08 1.5E-03 2.1E-OB 5.6E+06 Sb-124 1.3E-07 8.8E-04 1.5E-OB 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-08 1.1E+06 Sb-127 2 'E-06 6.2E-05 6.6E-09 2.3E+05 63

AMENDMENT NO. 7 December 1989 Table 3-4 DOSE RATE PARAMETERS IMPLEMENTATION OF 10 CFR 20 AIRBORNE RELEASES Child Dose Factor" pI DFA. DFG. l 1 1 Inhalation mrem/hr m~rem/ r X -1 3 Nuclide sec m~rem/ Ci ~Ci /m ~C1 /lll 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-OB 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-OB 7.BE+05 Cs-134 1.1E-08 2.7E-04 1.4E-OB 1.0E+06 Cs-137 7.3E-10 2.5E-04 4.9E-09 9.3E+05 Cs-138 3.6E-04 2.3E-07 2.4E-OB 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-OB 2.3E+05 Ce-141 2 'E-07 1.5E-04 6.2E-IO 5.6E+05 Ce-144 2.8E-OB 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-OB 2.2E+05 tir-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, Revi-sion l. Y-90 values Mere used for dose conversion factor Sr-90. 63a

AMENDMENT NO. 7 December 1989 TABLE 3-5 a DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS. AIRBORNE RELEASES AGE GROUP: ADULT ORGAN OF

REFERENCE:

MAXIMUM ORGAN R( INDIVIDUAL PATHWAY DOS E PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES RADIO- INHALATION GROUND PLANE COW-MILK GOAT-HILK ANIHAL-HEAT VEGETABLES NUCLIDE (HREH/YR (H2.HREH/YR (M2.HREH/YR (M2.HREH/YR (M2.MREM/YR (H2.HREH/YR PER UCI/H3 PER UCI/SEC ) PER UCI/SEC) PER UCI/SEC) PER UCI/SEC) PER UCI/SEC H 3 7.2E&2 O.OE-01 5.BE&2 1.2E+03 2.4E&2 1.6E+03 NA 24 1.0E+04 1.2E+07 1.2E&6 2.2E+05 7.2E-04 1.1E&5 CR 51 1.4E+04 4.7E+06 3.3E+06 5.9E&5 8.2E+05 2.3E&7 HN 54 1.4E+06 1.4E+09 1.4&07 2.1E+06 1.Sf&7 9.4E+08 HN 56 2.0E+04 9.DE+05 6.2E-02 1.1E-02 0.0E-01 2.DE&2 FE 55 7.2E+04 O.OE-01 1.4E+07 2.2E+06 1.6&08 1.9&08 FE 59 1.Of+06 2.7E+08 I. If+08 2.DE+07 9.BE+08 1.5E+09 CO 58 9.3E+05 3.BE+08 4.7E+07 7.6E+06 1.8E+08 8.Of+08 CO 60 6.DE+06 2.3E+10 1.7E+08 2.5E+07 8.DE+08 2.9E+09 CU 64 4.9E+04 6.1f+05 1.0E+06 1.7E+05 1.1E-05 3.3E+05 ZN 65 8.6E+05 7.5E+08 2.7E+09 4.0E&8 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 3.8E+06 2.1E+07 3.BE+06 2.9E&1 8.Of+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 2.9E+07 6.7E+06 3.4E+10 8.3E+10 8.9E+09 7.4E+ll ZR 95 1.8E+06 2.5E+08 4.6E+05 7.6E+04 9.2E&8 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 8.BE+06 NB 97 2.4E+03 1.8E+05 1.6E-09 2.9E-10 O.OE-01 8.1E-04 HO 99 2.5E+05 4.DE&6 2.9E+07 5.2E+06 1.2E+05 9.3E+06 TC 99M 4.2E+03 1.8E+05 2.8E+03 5.0E+02 3.6E-18 2.2E+03 RU106 9.4E+06 4.2f+08 7.3E+05 1.1E+05 1.DE+11 1.2E+10 AGl lOH 4.6E+06 3.5f+09 1.2E+10 1.BE+09 1.4E&9 4.4E+09 SB124 2.5E+06 6.Of+08 3.5E&8 5.BE+07 2.7E&8 4.Of+09 SB125 1.7E+06 2.4E+09 1.3E+08 1.BE+07 1.2E+08 1.4E+09 SB126 7.7E+05 8.4E+07 2.2E+08 4.0E+07 7.6E+07 1.6E+09

$ 8127       3.0E+05       1.7E+07          5.2f+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       2.0E+05 TE131H       5.6E+05       8.DE&6           8.9E+06        1.6E+06          1.1E+04       2.0E+07 I 131        1.2E+07       8.6E+06          3.4E+10        6.1E+10          1.2E+09       4.4E+10 I 132        1.1E+05       6.2E+05          3.9E+00        6.9E&0           O.OE-OI       1.1E+03 I 133        2.2E+06       1.2E+06          2.5E+08        4.5E+08          2.4E+01       1. 1E+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       1.0E+10 CS136        1.5E+05       1.5E+08          5.0E+08        2.2E+09          2.3E+07       4.6E+08 CS137        6.2E+05       1.3E+10          6.0E+09        2.1E+10          7.1E+08       8.6E+09 CS138        6.2E+02       3.6E+05          1.0E-23        4.6E-23          O.OE-OO       3.0E-11 BA140        1.3E+06       2. 1E+07         2.7E+07        4.8E+06          2.BE+07       7.3E+08 LA140        4.6E+05       1.9E+07          8.4E+04        1.5E+04          7.0E+02       3.3E+07 CE141        3.6E+05       1.4E+07          5.BE+06        1.Of+06          1.7E+07       9.3E+08 CE144        7.BE+06       7.0E+07          6.4E+07        9.6E+06          2.6E+08       1.1E+10 ND147        2.2E+05       8.5E+06          2.5E+05        4.6f+04          1.9E+07       5.1E+08 HF 179M      1.6E+05       0.0E-01          O.OE-01        O.OE-Ol          O.OE-01       O.OE-01 HF181        4.BE+05       2. 1E+08         5.5E+05        9.3E+04          1.2E+10       1.BE+09 W 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. 64

                                                                                >>  i I+   Va  cw,   ~       'hf  k - All AMENDMENT NO. 7 December 1989 TABLE        3-5 b DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS, AIRBORNf RELEASES AGE GROUP: TEEN                   ORGAN OF 

REFERENCE:

MAXIMUM ORGAN R(I), INDIVIDUAL PATHMAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES RADIO- INHALATION GROUND PLANE COM-HILK GOAT-HILK ANIHAL-HEAT VEGETABLES NUCLIDE (HREH/YR (H2.MREH/YR (H2.HREH/YR (H2.HREH/YR (H2-HREH/YR (H2.HREH/YR PER UCI/H3) PER UCI/SEC) PER UCI/SEC) PER UCI/SEC) PER UCI/SEC) PER UCI/SEC H 3 7.3E+02 O.OE-01 7.5E+02 1.5E+03 1.5E+02 1.9E+03 NA 24 1.4E+04 1.2E+07 2.If+06 3.9E+05 5.8E-04 1.DE+05 CR 51 2.1E+04 4.7E+06 3.9E+06 6.BE&5 4.4E+05 2.5E&7 MN 54 2.DE+06 1.4E+09 1.6E+07 2.3E+06 7.BE+06 9.6E+08 MN 56 5.7E+04 9.DE+05 2.3E-01 4.1E-02 O.OE-OO 3.7E&2 FE 55 1.2E+05 0.0f-01 2.4E+07 3.BE&6 1.3E+08 3.Of+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&7 8.3E<8 CO 60 8.7E+06 2.3E+10 2.1E+08 3.DE+07 4.3E+08 3.If+09 CU 64 6.1E+04 6.1E+05 1.6E+06 2.7E+05 8.0E-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.If&7 3.BE+OS 9.1E-04 1.1E+06 AS 76 1.5E+05 3.BE+OS 2.7f+07 4.9E+06 1.7E&I 5.3E+06 BR 82 1 ~ BE+04 2.1E+07 2.BE+07 5.1E+06 4.9E+02 6.1E+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+ll 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.)E+08 ZR 97 6 'E+05 3.DE+06 2.1E+04 3.8E+03 4.6E-Ol 7.DE+06 NB 97 3.9E+03 1.BE+05 1.9E-OB 3.3E-09 O.OE-01 4.8E-03 HO 99 2.7E+05 4.DE+06 5.1E+07 9.2E+06 9.4E+04 1.1E+07 TC 99M 6.1E+03 1.BE+05 5.3E+03 9.5E+02 3.2E-18 2.1E+03 RU106 1.Sf&7 4.2E&8 9.9E+05 1.5E+05 6.2E+10 1.5E+10 AG110M 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&8 7.3E+07 1.Sf+08 4.Sf+09 SB125 2.7E+06 2.4E+09 1.6E+08 2.3E+07 6.BE+07 1.6E+09 SB126 1.2E+06 8.4E+07 2.8E+08 5.1E+07 4.5E+07 1.BE+09 SB127 3.2E+05 1.7E+07 6.9E+07 1.2E+07 1.2E+06 1.2E+08 TE127 8.1E+04 3.DE+03 4.BE+04 8.6E+03 7.0E-09 1.BE+05 TE131H 6.2E+05 8.0E+06 1.3E+07 2.3E+06 7.4E+03 1.5E+07 I 131 1.5E+07 8.6E+06 5.4E+10'.4E+00 9.7f+10 9.DE+08 6.1E+10 I 132 1.5E+05 6.2E+05 1.2E+01 O.OE-OO 9.3E+02 I 133 2.9E+06 1.2E+06 4.2E+08 7.5E+08 1.BE+01 9.Sf+07 I 135 6.2E+05 1.3E+06 9.3E&5 1.7E+06 1.3E-15 1.2E+06 CS134 1.1E+06 6.9E+09 1.3E+10 4.6E+10 6.Sf+08 1.6E+10 CS136 1.9E+05 1.5E+08 8.4E+08 3.BE+09 1.BE+07 7.DE+08 CS137 8.5E+05 1.3E+10 1.1E+10 3.BE+10 5.7E+08 1.4E+10 CS138 8.6E+02 3.6E+05 1.8E-23 8. 1E-23 O.OE-OO 2.7E-11 BA140 2.DE+06 2.1E+07 3.Sf+07 6.4E+06 1.8E+07 8.BE+08 LA140 4.9E+05 1.9E+07 1.1E+05 2.1E+04 4.4E+02 2.4E+07 CE141 6.1E+05 1.4E+07 7.9E+06 1.4E+06 1.Of+07 1. 1E+09 CE144 1.3E+07 7.0E+07 8.BE+07 1.3E+07 1.Sf+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 7.1E+04 O.OE-Ol O.OE-OI O.OE-Ol O.OE-01 O.OE-01 HF181 4.BE+05 2.1E+08 7.1E+05 1.2E+05 7.Of&9 2.1E+09 M 185 7.7E+05 1.8E+04 3.3E+07 5.4E+06 1.2E+07 1.0E+09 NP239 1.3E+05 1.7E+06 5.3E+04 9.SE+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. 65

AMENDMENT. NO. 7 Oecember 1989 TABLE 3-5 c DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS. AIRBORNE RELEASES AGE GROUP: CHILD ORGAN OF

REFERENCE:

MAXIMUM ORGAN R(I), INDIVIDUAL PATHMAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES RADIO- INHALATION GROUND PLANE CON-MILK GOAT-HILK ANIMAL-HEAT VEGETABLES NUCLIDE (MREH/YR (H2.HREH/YR (M2.HREM/YR (M2.HREH/YR (H2.HREM/YR (H2.MREH/YR PER UCI/M3) PER UCI/SEC) PER UCI/SEC) PER UCI/SEC) PER UCI/SEC) PER UCI/SEC H 3 6.4E+02 O.OE-01 1.2E&3 2.4E&3 1.BE+02 2.9E&3 NA 24 1. 6E+04 1.2E+07 4.5E+06 8.DE+05 9.2E-04 1.6E+05 CR 51 1.7E+04 4.7E+06 2.5E+06 4.4E+05 2.2E&5 1.6E+07 HN 54 1.6E+06 . 1.4E&9 1.1E+07 1.7E+06 4.3E+06 6.9E+08 MN 56 1.2E+05 9.DE+05 &.8E-01 1.6E-01 O.OE-OO 1.1E&3 FE 55 1.1E+05 O.OE-01 6.1E&7 9.6E+06 2.5E+08 7.6E+08 FE 59 1.3E&6 2.7E+08 9.5E+07 1.7E&7 3.DE&8 1.2E&9 CO 58 l. IE&6 3.8E+08 3.4E+07 5.6E&6 4.7E+07 5.3E+08 CO 60 7.)E+06 2.3E+10 1.4E&8 2.0E&7 2.2E&8 2.1E+09 Cu 64 3.7E+04 6.1E+05 1.7E+06 2.9E&5 6.5E-06 2.2E+05 ZN 65 1.0E+06 7.5E+08 6.BE+09 1.0E&9 6.2E+08 3.DE&9 ZN 69M 1.0E+05 1.3E&6 2.2E+07 4.DE+06 7.2E-04 9.0E+05 AS 76 7.DE+04 3.BE+06 2.2E+07 4.0E+06 1.1E+01 3.3E+06 BR 82 2.1E+04 2.1E+07 5.8E+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 1.0E+11 2.6E+ll 9.BE+09 2.1E+12 ZR 95 2.2E+06 2.5E+08 4.2E+05 7.DE+04 3.0E+08 1.3E+09 NB 95 6.1E+05 1.4E+08 1.1E+08 1.BE+07 I.DE+09 6.2E+08 ZR 97 3. 5E+05 3.DE+06 2.1E&4 3.8E+03 3.5E-01 5.2E+06 NB 97 2.8E+04 1.BE+05 4.2E-07 7.6E-OB O.OE-01 8.2E-02 MO 99 1.3E+05 4.DE&6 8.7E+07 1.6E+07 1.2E+05 1.6E+07 TC 99H 4.BE+03 1.BE+05 7.4E&3 1.3E+03 3.4E-18 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&9 3.8E+08 3.0E&9 58124 3.2E+06 6.DE+08 3.3E+08 5.4E+07 8.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 4.0E+07 2.7E+07 1.4E&9 SB127 2.3E+05 1.7E+07 5.5E+07 1.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 8.DE+06 1.1E+07 2.1E+06 5.0E+03 9.9E+06 I 131 1.6E+07 8.6E+06 1.1E+1 1 1.9E+11 1.4E+09 1.2E+l1 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 2.1E+06 3.8E+06 2.3E-15 2.1E&6 CS134 1.DE+06 6.9E+09 2.DE+10 7.5E+10 8.3&48 2.6E+10 CS136 1.7E+05 1.5E+08 1.3E+09 6.DE+09 2.1E&7 1.1E+09 CS137 9.1E+05 1.3E+10 I. 9E+10 6.BE+10 7.9E+08 2.5E+10 CS138 8.4Et02 3.6E+05 3.2E-23 1.4E-22 O.OE-OO 3.6E-11 BA140 1.7E+06 2. 1E+07 5.6E+07 1.0E+07 2.1E+07 1.4E+09 LA140 2.3E&5 1.9E+07 9.5E+04 1.7E+04 2.8E+02 1.6E+07 CE141 5.4E005 1.4E+07 6.3E+06 1.1E+06 6.4E+06 9.0EKB CE144 1.2E+07 7.0E+07 7.0E+07 1.1E+07 1.DE&8 1.1E+10 N0147 3.3E+05 8.5E+06 2.BE+05 5.DE&4 7.4E+06 4.BE&8 HF179H 7.4E+04 O.OE-Ol O.OE-01 O.OE-01 O.OE-Ol O.OE-01 HF181 2.2E+05 2. 1E+08 5.9E&5 9.9E+04 4.4E+09 1.BE+09 N 185 6.9E+05 1.BE+04 2.7E+07 4.3E+06 7.3E&6 8.3E+08 NP239 6.4E+04 1.7E+06 4.6E+04 8.3E+03 1.1E&3 1.0E+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. 66

AMENDMENT NO. 7 December 1989 TABLE 3-5 d DOSE PARAMETERS FOR 10 CFR 50 EVALUATIONS, AIRBORNE RELEASES AGE GROUP: INFANT ORGAN OF

REFERENCE:

MAXIMUM ORGAN

      ,  R(I),    INDIVIDUAL PATHMAY DOSE PARAMETERS FOR RADIONUCLIDES OTHER THAN NOBLE GASES RADIO-      INHALATION       GROUND PLANE       CON-MILK         GOAT-MILK     ANIMAL-HEAT   VEGETABLES NUCLIDE       (HREH/YR       (H2.HREH/YR       (H2.MREH/YR      (M2.HREH/YR    (H2.HREM/YR   (H2.HREH/YR PER    UCI/H3)   PER UCI/SEC)     PER    UCI/SEC)  PER    UCI/SEC) PER  UCI/SEC) PER.UCI/SEC H     3       3.7E+02         O.OE-01            1.Sf&3          3.7f&3         O.OE-Ol       O.OE-Ol NA 24          l.If+04         1.2E+07          7.Sf+06           1.4E+06       O.OE-01       O.OE-01 CR 51         1.3E+04         4.7E+06           2.2E+06          3.BE+05        O.OE-OI       O.OE-01 MN  54        1.DE+06          1.4E&9           2.1E+07          3.1E+06        O.OE-01       O.OE-01 MN  56        7.2f&4          9.0E+05            1.3E+00         2.4E-01        O.OE-01       O.OE-01 FE  55        8.7E+04         O.OE-OI           7.4E&7           1.2E+07        O.OE-01       O.OE-Ol FE  59        1.DE+06         2.7E+08            1.BE+08         3.4E+07        O.OE-Ol       O.OE-01 CO  58        7.BE+05         3.BE&8            2.9E+07          4.BE+06        O.OE-01       O.OE-OI CO  60        4.5E+06         2.3E+10           1.2E+08          1.7E+07        O.OE-Ol       O.OE-01 CU  64        1.5E+04         6.1E+05           1.9E+06          3.2E+05        O.OE-01       O.OE-01 ZN  65        6.5E+05         7.5E+08           1.2E+10          1.7E&9         O.OE-OI       O.OE-01 ZN 69H        4.1E+04         1.3E+06           2.4E+07          4.3E+06        O.OE-01       O.OE-01 AS 76         2.7E+04         3.BE+06           2.2E+07          4.DE&6         O.OE-OI       O.OE-OI BR 82         1.3E+04         2.If+07           9.BE&7           1.BE+07        O.OE-01       O.OE-01 SR 89         2.0E+06         2.2E+04           6.DE&9           1.BE+10        O.OE-OI       O.OE-OI SR 90         1.6E+07         6.7f+06           1.2E+11          2.9E+11        O.OE-OI       O.OE-01 ZR 95         1.BE+06         2.5E+08           4.DE+05          6.5E+04        O.OE-01       O.OE-01 NB 95         4.BE+05         1.4f&8            9.6E+07          1.7E+07        O.OE-01       O.OE-01 ZR 97         1.4E+05         3.0E+06           2.2E+04          4.DE+03        O.OE-01       O.OE-Ol NB  97        2.7E&4          1.Sf+05           1.1E-06          1.9E-07        O.OE-01       O.OE-01 MO  99        1.3E+05         4.DE+06           1.6E&8           2.8E+07        O.OE-01       O.OE-01 TC 99M        2.0E+03         1.BE+05           8.2E+03          1.5E+03        O.OE-OI       0.0f-01 RU106         1.2E007         4.2E+08           S.DE&5           1.2E+05        O.OE-01       O.OE-01 AGllOH        3.7E+06         3.5E+09           8.2E+09          1.2E+09        O.OE-01       O.OE-OI SB124         2.6E+06         6.DE+08           3.1E+08          5.1E+07        O.OE-OI       O.OE-Ol SB125         1.6E+06         2.4E+09           1.1E+08          1.6E+07        O.OE-01       O.OE-01 SB126         9.6E+05         8.4E+07           2.1E+08          3.7E+07        O.OE-OI       O.OE-01 SB127         2.2E+05         1.7E+07           5.5E&7           9.9E+06        O.OE-01       O.OE-01 TE127         2.4E+04         3.0E+03           6.BE+04          1.2E+04        O.OE-OI       O.OE-OI TE131H        2.DE+05         S.DE+06           1.2E+07          2.1E+06        O.OE-01       O.OE-01 I 131         1.5E+07         8.6E+06           2.6E+11          4.7E+l1        O.OE-01       O.OE-01 I 132         1.7E+05         6.2E+05           3.4E+01          6.1E+Ol        O.OE-01       O.OE-01 I 133         3.6E+06         1.2E+06           2.4E+09          4.3E+09        O.OE-01       O.OE-01 I 135         7.DE+05         1.3E+06           4.9E+06          8.9E+06        O.OE-Ol       O.OE-01 CS134         7.DE+05         6.9E+09           3.7E+10          1.4E+l1        O.OE-01       O.OE-Ol CS136         1.3E+05         1.5E+08           2.8E+09          1.2E+10        O.OE-01       O.OE-01 CS137         6.1E+05         1.3E+10           3.6E+10          1.3E+11        O.OE-Ol       O.OE-OI CS138         8.BE+02         3.6E+05           1.2E-22          5.6E-22        O.OE-Ol       O.OE-01 BA140         1.6E+06         2.1E+07           1.2E+08          2.1E+07        O.OE-01       O.OE-01 LA140         1.7E+05         1. 9E+07          9.4E&4           1.7E+04        O.OE-OI       O.OE-OI CE.141        5.2E+05         1.4E+07           6.4E+06          1. 1E+06       O.OE-OI       O.OE-01 CE144         9.BE+06         7.0E+07           7.1E+07          1.1E+07        O.OE-Ol       O.OE-01 ND147         3.2E+05         8.5E+06           2.BE+05          5.DE+04        O.OE-01       O.OE-Ol HF179H        2.BE+04         O.OE-01           O.OE-01          O.OE-01        O.OE-01       O.OE-01 HF181         8.4E+04         2. If+08          5.9E+05          9.9E+04        O.OE-01       O.OE-Ol N 185         6.3E+05         I.BE+04           2.7E+07          4.4E+06        O.OE-01       O.OE-OI NP239         6.0E+04         1.7E+06           4.7E+04          8.Sf+03        O.OE-01       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. 67

NENDNENT NO. 3 February 1986 Table 3-6 INPUT PARNETERS FOR CALCULATING R. Parameter Value Table* r (dimensionless ) 1.0 for radioiodine 0.2 for particulates E-15 F (days/liter) Each stable element E-1 U (liters/yr) Infant'30 E-5 Child 330 E-5 Teen 400 Adult 310 E-5 (DFL,. ) (mrem/pCi ) Each radionuclide E-ll to E-14 Yp (kg/m ) 0.7 E-15 Y (kg/m ) 2.0 E-15 tf (seconds) 1.73 x 10 5 (2 days) E-15 th (seconds) 7.78 x 10 (90 days) E-15 g (k g/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.

68

AMENDMENT NO. 3 February 1986 Table 3-7 INPUT PARAMETERS FOR CALCULATING R. Parameter Yalue Table* r (dimensionless ) 1.0 for radioiodine E-15

                                  . 0.2 for particulates        E-15 Ff (days/kg)                    Each   stable element U      (kg/yr ) Infant         0 Child Teen           65 Adul t         110 (DFLi )    (mrem/pCi )          Each   radionuclide    E-11  to E-14 Y     (kg/m  )                   0.7                         E-1 5 P

Y (kg/m ) 2.0 E-1 5 tf (seconds) 1.73 x 10 6 (20 days) E-15 th (seconds) 7.78 x 10 6 (90 days) E-15 (} (kg/day) 50 E-3

• Of Regulatory Guide 1.109, Revision 1.

69

AMENDMENT NO. 7 December 1989 Table 3-8 INPUT PARAMETERS FOR CALCULATING R. Parameter Value Table* r (dimensionless) 1.0 for radioiodine E-1 0.2 for particulates E-1 i a (mrem/pCi) Each radionuclide E-11 to E-14 U L a (kg/yr) Infant 0 Child 26 E-5 Teen 42 Adul t 64 U S a (kg/yr) Infant 0 E-5 Child 520- E-5 Teen 630 E-5 Adult 520 E-5 fL (dimensionless) 1.0 E-15 f (dimensionless) 0.76 E-15 tL (seconds) 8.6 x 10 (1 day) E-'15 th (seconds) 5.18 x 10 (60 days) E-15 Y (kg/m ) 2.0 E-'1 5 "Of Regulatory Guide 1.109, Revision 1 . 70

AMENDMENT NO. 6 November 1988 Table 3-9 INPUT PARAMETERS NEEDED FOR CALCULATING DOSE SUMMARIES TO THE MAXIMUM INDIVIDUAL AND THE POPULATION MITHIN 50 MILES FROM MNP-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 ('A) 53.8 Ref 4 Annual average temperature (Fo) 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) 2.8E408 Refs 7 5 9 Annual 50-mile meat production (kg/yr) 2.3E+07 Refs 7 8 9 Annual 50-mile vegetable production (kg/yr) 3.5E+09 Refs 759 Source terms Ref 8 71

AMENDh1ENT NO. 3 February 1986 Table 3-9 (contd.) In ut Parameter Yalue Reference X/g values by sector for each dis-tance Lrecirculation, no decay) See Tables 3-11 (sec/m~) through 3-12 Ref 10 X/(} values by sector for each dis-tance (recirculation, 2.2g days See Tables 3-11 decay, undepleted) (sec/m~) through 3-12 Ref 10 X/g values by sector for each dis-tance (recirculation, 8 0 days See Table 3-11 decay, depleted) (sec/m ) through 3-12 Ref 10 D/() values by sector for each dis- See Table 3-11 tance (1/m2) through 3-12 Ref 10

     *References are  listed in  Table 3-14.

72

Table 3-10 REIICTGR GUILE IIIG STACK X/II NIU 0/I/ VIILUES* a) tlo Deca Unde leted CIII/Q (sec/meter cubed) for each segment Direct ion From Site Se nt Boundaries in Hiles from the Site

                                     -2           2-S        1.103E-06       3.229E-07     1.153E-07     6.291E-OB     4.15IE-OB    2.056E-OO    6.109E-OB 4.956E-OB  3.370E-OB 2.530E-OB SSW          8.824E-07       2.569E-07    9.106E-OB      4.941E-08     3.243E-OS    1.607E-OS    5 '67E-08 4.304E-OB  2.930E-OO 2.201E-OO SW         7.484E-07       2.220E-07    8.257E-OB      4.646E-OB     3.098E-OB    4.101E-OG    6.486E-OS 4.274E-OB  2.917E-OB 2.195E-OO WSW          5.687E-07       1.717E-07    6.362E-OB      3.543E-OB     2.341E-OG    2.882E-OB    4.367E-OB 2.851E-OB  1.940E-OS 1.457E-OB M        2.201E-07       7.362E-OB    2.829E-OO      1.604E-OB     1.065E-OB    5.201E-09    2.986E-OS 2.489E-OB  1.695E-OS 1.274E-OB Wthl         3.037E-07       1.024E-07     3.926E-OB     2.208E-OO     1.459E-OB    7.801E-09    2.680E-OB 2.16BE-OB  1.471E-OB 1.104E-OO
   -N           9.434E-07       2.769E-07    9.967E-OB      5.427E-OB     3,563E-OB    1.789E-OB    3.036E-OB 2.344E-OO  1.582E-OS 1.183E-OO tltiW        3.010E-06       8.542E-07    3.077E-07      1.684E-07     1.121E<<07    5.498E-OB    5.529E-OB 4.004E-OS  2.706E-OB 2.023E-OO tl       3.675E-06       1.034E-06    3.712E-07      2.037E-07     1.343E-07    1.060E-07    8.208E-OB 4.484E-OB  3.033E-OB 2.269E-OB ttttE        2.430E-06       6.639E-07    2.313E-07      1.237E-07     8.113E-OB    9.852E-OB    5.49IE-OB 2.952E-OS  1.980E-OS 1.473E-OB tlE        1.308E-06       3.571E-07     1.242E-07     6.798E-OB     7.999E-OB    9.512E-OB    4.486E-OB 2.428E-OB  1.634E-OB 1.219E-OB OIE          1.086E-06       3.381E-07    2.229E-07      2.754E-07     2.056E-07    9.895E-OB    4.020E-OB 2. 168E-08 1.455E-OB 1.082E-OB E        1 ~ 218E-06     3.665E-07    2. 195E-07     2.582E-07     1.926E-07    9.269E-OB    3.768E-OS 2.036E-OG  1.369E-OB 1.020E-O&

ESE 2.409E-06 7.211E-07 4.124E-07 4.440E-07 3.242E-07 1.423E-07 5.594E-OB 3.335E-OB 2.231E-OB 1.656E-OO SE 3.043E-06 8.555E-07 3.108E-07 2.844E-07 3.417E-07 1.677E-07 8.760E-OS 5.311E-OB 3.5&6E-OB 2.680E-OB SSE 1.842E-06 5.373E-07 1.943E-07 1.064E-07 7.011E-OB 3.471E-OB 7.245E-OG 5.737E-OO 3.891E-OO 2.917E-OB 'Desert Sigmas, Building wake effect. All stability classes A through G.

Table 3-10 (contd.) b) 2.26-0a Deca Unde leted OII/f) (sec/meter cubed) for each segment Oirection From Site Se nt Boundaries in fftles from the Site

                                                                                     - 0
                                                                               '23E3E-08 S       1.101E-06      3.218E-07     1.146E-07 6.238E-OB     4.106E-OB  2.202E-OB    5.876E-OB 4.683E-OG 3.113E-OS  2.285E-OB SSW          O.OIOE-07      2.561E-07     9.054E-OS 4.900E-OB     3.208E-OB   1.580E-OS-  5.064E-OS 4.064E-OS 2.704E-OG  1.985E-OO SW         7.471E-07      2.212E-07     8.208E-OB 4.608E-OB     3.065E-OB               6.259E-OB 4.036E-OG 2.692E-OO  1.980E-OS WSW          5.678E-07      1 ~ 711E-07   6.326E-OB 3.515E-OS     2.3I7E-OB  2.827E-OO    4.213E-OS 2.691E-OO 1.789E-OB  1.313E-OB W       2.197E-07      7.334E-OB     2.810E-OB 1.589E-OB     1.053E-OB  5.107E-09    2.859E-OB 2.338E-OB 1.553E-OB  1.138E-OB WIIW         3. 031E-07     1.020E-07     3.899E-OB 2.187E-OB     1;442E-08  7.652E-09    2.570E-OB 2.039E-OB 1.350E-OB  9.877E-09 tlW       9.419E-07      2.760E-07     9.911E-OO 5.383E-OB     3.527E-OB   1.760E-OB   2.929E-OG 2.223E-OB 1.469E-OB  1.074E-OB tt tfW       3.006E-06      8.520E-07     3.063E-07 1.673E-07     1.111E-07  5.422E-OB    5.369E-OB 3.830E-OB 2 '42E-08  1.867E-OB ff      3.671E-06      1.031E-06     3.696E-07 2.024E-07     1.332E-07  1.044E-07    7.996E-OG 4.291E-OO 2.852E-OB  2.096E-OB tftfE        2.427E-06      6.624E-07     2.303E-07 1.23OE-07     8.050E-OO  9.700E-OB    5.336E-OS 2.812E-OB 1.850E-OO  1.350E-OB ttE        1.307E-06      3.562E-07     1.236E-07 6.753E-OS     7.927E-OO  9.359E-OB    4.343E-OS 2.300E-OB 1. 514E-08 1.104E-OB DIE          1.085E-06      3.371E-07     2.2I7E-07 2.733E-07     2.036E-07  9.737E-OB    3.89OE-OB 2.051E-OB 1.346E-OO  9.792E-09 E       1.216E-06      3.655E-07     2.185E-07 2.563E-07     1.907E-07  9.125E-OB    3.649E-OB 1.928E-OO 1.268E-OB  9.243E-09 ESE          2.406E-06      7.193E-07     4.104E-07 4.408E-07     3.212E-07  1.403E-07    5.420E-OO 3.164E-OB 2.072E-OS  1.506E-OB SE         3.039E-06      8.532E-07     3.093E-07 2.825E-07     3.389E-07   I.655E-07   8.498E-OO 5.050E-OB 3.341E-OB  2.446E-OB SSE          1.839E-06      5.356E-07     1.932E-07 1.055E-07     6.939E-OB  3.414E-OB    6.983E-OO 5.436E-OO 3.60GE-OB  2.646E-OB

Table 3-10 (contd.) c) B.O-Da Deca De leted CHI/O (sec/meter cubed) for each segment Direction From Site Se nt Boundaries in Hiles from the Site 2-S 1.006E-06 2.858E-07 9.773E-OB 5.164E-OB 3.323E-OB 1.572E-OS 5.545E-OB 4.289E-OS 2.735E-OB 1.946E-OB SSM 8.039E-07 2.269E-07 7.694E-OB 4.035E-OB 2.578E-OB 1.221E-OB 4.794E-OB 3.733E-OS 2.383E-OS 1.697E-OB SM 6.775E-07 1.942E-07 6.924E-DB 3.784E-OB 2.462E-OB 3.644E-OB 5.757E-OS 3.520E-OB 2.246E-OB 1.597E-OB MSM 5.169E-07 1.515E-07 5.398E-DS 2.923E-OB 1.886E-OB 2.562E-OB 3.875E-OB 2.347E-OB 1.493E-OB 1.060E-OB M 2.038E-07 6.726E-OB 2.521E-DS 1.407E-OB 9.223E-09 4.395E-09 2.772E-OB 2. 189E-08 1.399E-OB 9.964E-09 MN 2.8I3E-07 9.369E-OB 3.505E-DB 1.938E-DB 1.263E-OB 6.663E-09 2.393E-OB 1.828E-OB 1. 161E-08 8.239E-09 N 8.584E-07 2.450E-07 8.465E-OB 4.468E-DB 2.865E-OS 1.391E-OS 2.688E-OB 1.987E-OS 1.257E-OB 8.893E-09 HN 2.714E-06 7.416E-07 2.547E-07 1.345E-07 8.724E-OS 4.071E-OB 4.583E-OB 3.202E-OB 2.021E-OB 1.427E-OB H 3. 312E-06 8.954E-07 3.060E-07 1. 619E-07 1.037E-07 8.674E-OB 6.796E-OB 3.375E-OB 2. 123E-08 1.495E-OS NNE 2. I96E-06 5.789E-07 1.924E-07 9.939E-OB 6.360E-OB 8.587E-OB 4.544E-OB 2.222E-OB 1 ~ 387E-08 9.7I3E-09 NE 1.186E-06 3.134E-07 1.045E-07 5.570E-OS 7.080E-OB 8.611E-OB 3.736E-OB 1.841E-OS 1.153E-OB 8.096E-09 ENE 9.883E-07 3. 011E-07 1.866E-07 2 '47E-07 1.554E-07 7.012E-OS 2.504E-OS 1.187E-OB 7.208E-09 4.925E-09 E 1.107E-06 3.252E-07 1.832E-07 2.013E-07 1.456E-07 6.569E-OB 2.347E-OB I.115E-OB 6.784E-09 4.643E-09 ESE 2.182E-06 6.364E-07 3.437E-07 3.464E-07 2.451E-07 I.014E-07 3.460E-OB 1.827E-OB I 107E-08

                                                                                                            ~         7.548E-09 SE       2.747E-06    7.450E-07    2.593E-07 2.509E-07     3.134E-07  1.463E-07   6.976E-OB 3.884E-OB  2.434E-OB   1.709E-OB SSE       1.671E-06    4.722E-07    1.634E-07 8.657E-OB     5.559E-OB  2.633E-OB   5.975E-OB 4.484E-OB  2.835E-OB   2.003E-OB

Table 3-10 (contd.) d) Reactor Buildin Stack Re)ative Depos{t)on Rate 0/ ) Per Un) t Area meter Direction From Site Se nt Boundaries in HI)es from the Site 2- 3- 5 5: 2 -YG. S 7.256E-09 1.756E-09 5.194E-)0 2.494E-IO 1.466E-10 5.865E-11 3.375E-)1 1. 798E-11 9.603E-12 5.944E-)2 SSM 5.752E-09 1.380E-09 4. 082E-10 1. 959E-10 1 ~ 150E-10 4.626E-I) 2.608f-ll 1. 395E- 1 1 7. 448E-12 4.610E-12 SM 3.'I76E-09 7.513E-10 2.191E- IO 1.035E-10 6.028E-)l 3.817E-I ) 2. 414E-11 9.646E-)2 5.151E-12 3.188E-)2 MSM 2.757E-09 6.889E-)0 2. 061E-10 9.796E-11 5. 718E-11 3.358E-11 1.980E-I) 7 ~ 927E-12 4.233E-')2 2.620E-I2 M 1. 601E-09 4.334E-10 1. 358E-10 6.565E-)1 3. 861E-11 1.524E-11 1.094E-11 6.220E-12 3. 321E-12 2.056E-12 MN 2.2)5E-05t 5.797E-10 1.816E-10 8.829E-)1 5.204E-II '.856E-11 1.697E-11 7.175E-12 3.83)E-12 2. 372E-12 tQ 4.90)E-09 1.2IBE-09 3.728E-10 1. 813E-10 1.068E-IO 4.328E-11 2.419E-)1 1.286E-11 6.869E-12 4.252E-12 ttttM 1.235E-OB 2.845E-09 8.198E-10 3.873E-IO 2.303E-)0 9.105E-11 4.558E-I) 2.363E-11 1.262E-11 7. 811E-12 M 1.914E-OB 4.304E-09 1.2I3E-09 5 '60E-10 3.273E-)0 1.707E-IO 7.090E-11 2.810E-) I 1.501E-II 9. 288E-12 RtlE 2.034E-OB 4.577E-09 1.284E-09 5. 961E-10 3. 471E-10 1.810E-10 6.374E-11 2.526E-11 1.349E-11 8.350E-)2 tlE 1.338E-OB 2.986E-09 8. 34)E-10 3. 918E-10 2.819E-10 1 ~ 483E-10 4.323f-ll 1.713E- I I 9. 150E-12 5.663E-)2 EtlE 9.298E-09 2.169E-09 7.730E-IO 4 '79E-10 2. 604E-10 1. 00)E-10 2.897E-11 1.148E- I 1 6.132E-)2 3.795E-12 E 1.017E-OB 2.355E-09 8.239E-10 4.749E-IO 2.699E-IO 1.038E-10 3.003E-11 1.190E-11 6 ~ 355E-12 3.934E-12 ESE 1.832E-OB 4.190E-09 1.440E-09 .8. 177E-)0 4.647E-10 1.780E-10 5. 136E-11 2.049E-.1) 1.094E-I1 6.773E-12 SE 2.006E-OB 4.525E-09 1.262E-09 7. 531E-10 6. 421E-10 2.467E-10 7.197E-11 2.872E-II 1.534E-11 9.492E-)2 SSE 9.32)E-09 2.265E-09 6.764E-09 3. 250E-10 1. 905E-10 7.633E-)1 4.186E-11 2.224E-)1 1.187f-ll 7.350E-12

Table 3-11 a) No Deca Unde leted Oll/II (sec/meter cubed) for each segment Direction From Site Se nt Boundaries in Hiles from the Site 2-3 S 1.791E-05 5.032E-06 1.836E-06 1.019E-06 6.765E-07 3.337E-07 1.405E-07 7.800E-OG 5.333E-OG 4.018E-OG SSW 1.513E-05 4 '82E-06 l. 568E-06 8.729E-07 5.803E-07 2.781E-07 1 ~ 214E-07 6.758E-OG 4.627E-OG 3.489E-OG SW 1. 419E-05 4.080E-06 'I.513E-06 8.468E-07 5.651E-07 2.811E-07 1.198E-07 6.690E-OG 4.584E-OB 3.457E-OG IISM 1.004E-05 2.847E-06 1.044E-06 5.811E-07 3.862E-07 1.909E-07 8.059E-OB 4.481E-OG 3.066E-DB 2.311E-OB W 8.834E-06 2.512E-06 9.240E-07 5.149E-07 3.426E-07 1 ~ 695E-07 7. 171E-08 3.988E-OB 2.728E-OG 2.056E-OG WNM 8.324E-06 2. 320E-06 8.416E-07 4.654E-07 3.080E-07 1.511E-07 6.317E-OB 3.489E-OG 2.380E-OG 1.791E-OG NM 9.578E-06 2.620E-06 9.367E-07 5.135E-07 3.377E-07 1.639E-07 6.739E-OG 3.687E-OB 2.506E-OB 1.881E-OB NNW 1.520E-OS 4.196E-06 1.494E-06 8.198E-07 5.393E-07 2.620E-07 1.078E-07 5 '05E-08 4 '07E-08 3.015E-OB N 1.661E-OS 4.558E-06 1.636E-06 8.987E-07 6.918E-07 2.881E-07 1.189E-07 6.5IGE-OB 4.435E-OB 3,.329E-OB NNE 1.259E-05 3.378E-06 1.189E-06 6.456E-07 4.217E-07 2.025E-07 8.191E-OG 4.445E-OB 3.015E-OG 2.260E-OB NE 1.019E-05 2.764E-06 9.837E-07 5.377E-07 3.528E-07 1.707E-07 6.978E-OB 3.804E-OG 2.581E-OB 1.935E-OG ENE 9. 328E-06 2.528E-06 8.989E-07 4.907E-07 3.215E-07 1.550E-07 6.302E-OG 3.426E-OG 2.322E-OB 1.739E-OB E 8.659E-06 2.344E-06 8.336E-07 4.553E-07 2.985E-07 1. 441E-07 5.868E-OB 3. 191E-08 2.162E-OG 1. 619E-08 ESE 1.452E-05 3.919E-06 1. 391E-06 7.573E-07 4.950E-07 2.375E-07 9.577E-OG 5.173E-OB 3.494E-OB 2.611E-OG SE 2.052E-OS = 5.657E-06 2.038E-06 1.121E-06 7.387E-07 3.595E-07 1.482E-07 8.123E-OG 5. 519E-08 4. 14 I E-08 SSE 2 '28E-05 5.940E-06 2.156E-06 1.193E-06 7.895E-07 3.875E-07 1.619E-07 8.949E-OG 6. 108E-08 4.596E-OB

• Ground level release, Desert Sigmas. All stability classes A through G

Table 3-11 (Contd) b) 2.26-Da Deca Unde leted CNI/O (sec/meter cubed) for each segment Direction From Site Se nt Boundaries in Miles from the Site 1-2 2- -5 -0 S 1.783E-05 4.991E-06 1.809E-06 9.984E-07 6.586E-07 3.195E-07 1.287E-07 6.725E-OB 4.334E-OG 3.079E-OB SSW 1.506E-05 4.246E-06 1.545E-06 8.547E-07 5.647E-07 2.746E-07 1.110E-07 5.810E-OB 3.745E-OB 2 '60E-08 SW 1. 413E-05 4.046E-06 1.490E-06 8.292E-07 5.500E-07 2.689E-07 '.095E-07 5.754E-OG 3.712E-OB 2.637E-OB WSW 9.992E-06 2.823E-06 1.029E-06 5.689E-07 3.75&E-07 1.825E-07 7.359E-OB 3.846E-OB 2.475E-OB 1.756E-OG W 8.792E-06 2.489E-06 9.089E-07 5.030E-07 3.324E-07 1.614E-07 6.487E-O& 3.36&E-OG 2.152E-OG 1.515E-OB WtQ 8.286E-06 2.300E-06 8.282E-07 4.549E-07 2.990E-07 1.441E-07 5.731E-OB 2.961E-OB 1.891E-OG 1.332E-DG tQ 9.550E-06 2.600E-06 "9.244E-07 5.040E-07 3.295E-07 1.576E-07 6.21&E-OG 3.220E-DB 2.073E-OB 1.475E-OB iNQ 1.515E-05 4.145E-06 1.479E-06 8.OBOE-07 5.293E-07 2.541E-07 1.013E-07 5.321E-OB 3.473E-OG 2.503E-OB N 1.656E-OS 4.532E-06 1. 619E-06 8.85&E-07 5.80&E-07 2.794E-07 1.117E-07 5.87&E-OB 3.839E-OB 2.769E-OB NNE 1.255E-05 3.356E-06 1.175E-06 6.350E-07 4.12&E-07 1.956E-07 7.62&E-OB 3.941E-OG 2.54&E-OB 1.82)E-OB NE 1.015E-05 '.743E-06 9.705E-07 5.274E-07 3.441E-07 1.63&E-07 6.419E-OG 3.303E-OB 2.117E-OB 1.500E-OG ENE 9.291E-06 2.50&E-06 8.865E-07 4.810E-07 3.133E-07 1.487E-07 5.7&BE-OB 2.966E-OB 1.897E-OB 1.342E-OG E 8.626E-06 2.326E-06 8.225E-07 4.467E-07 2.912E-07 1.384E-07 5.403E-OB 2.774E-OB 1.777E-OG 1.259E-OB ESE 1.446E-05 3.891E-06 1.373E-06 7.435E-07 4.834E-07 2.285E-07 8.846E-OB 4.521E-OG 2.893E-OB 2.049E-OB SE 2.045E-05 5.61&E-06 2.013E-06 1 ~ 102E-06 7.222E-07 3.446E-07 1.376E-07 7.159E-OB 4.625E-OB 3.30IE-OB SSE 2.120E-05 5.895E-06 2.127E-06 1.170E-06 7.700E-07 3.721E-07 1.491E-07 7.790E-OB 5.030E-OG 3.583E-OB

Table 3-11 (Contd) c) 8.0-0a Deca De leted OII/II (sec/meter cubed) for each seynent Direction From Slee Se nt Boundaries in Hiles from the Site

                              -2         2-         3-                                                                   4TH S       1.602E-05    4.299E-06   1. 468E-06 7.920E-07    5.081E-07   2.342E-07   8.603E-OG 4. 155E-08 2.544E-OB   1.742E-OG SSW       1.353E-OS    3.657E-06   1.269E-06  6.782E-07    4.358E-07   2.014E-07   7.428E-OS 3.596E-OS  2.205E-OB   1.510E-OB SW       I.269E-05    3.485E-06   1.224E-06  6.579E-07    4.244E-07   1.972E-07   7.328E-OB 3.561E-OS  2. 184E-08  1.497E-OB MSM       8.976E-06    2.432E-06   8.448E-07  4.515E-07    2.901E-07   1 ~ 339E-07 4.930E-OB 2.384E-OS  1.460E-OG   9.995E-09 W       7.901E-06    2.145E-06   7.473E-07  3.998E-07    2.571E-07   1.188E-07   4.375E-OB 2.112E-OS  1.291E-OS   8.8'19E-09 WNW       7 '46E-06    1.982E-06   6.808E-07  3.614E-07    2.312E-07   1.060E-07   3.858E-OS 1.850E-OB  1. 129E-08  7.701E-09 NM       8.579E-06    2 '39E-06   7.584E-07  3.993E-07    2.538E-07   1.152E-07   4.137E-OG 1.971E-OS  1.201E-OO   8.205E-09 NN        1.360E-05    3.564E-06   1.211E-06  6.382E-07    4.061E-07   1.847E-07   6.652E-OB 3.185E-OS  1.950E-OS   1.337E-OG N       1.487E-05    3.897E-06   1.326E-06  6.996E-07    4.456E-07   2.030E-07   7.334E-OB 3.516E-OB  2.153E-OB   1.477E-OB NNE       1.127E-05    2.GOBE-06   9.630E-07  5 '23E-07    3.173E-07   1.426E-07   5.042E-OS 2.386E-OB  1 ~ 454E-08 9.933E-09 NE       9.117E-06    2.362E-06   7.964E-07  4.180E-07    2.651E-07   1.199E-07   4.280E-OS 2.030E-OG  1.235E-OB   8.415E-09 ENE       8.348E-06    2.16OE-06   7.277E-07  3.814E-07    2.416E-07   1.089E-07   3.864E-OB 1.827E-DS  1. 109E-08  7.752E-09 E       7.750E-06    2.003E-06   6.749E-07  3.539E-07    2.243E-07   1.013E-07   3.601E-OO 1.704E-OG  1.035E-OG   7.046E-09 ESE       1.299E-05    3.349E-06   1.126E-06  5.889E-07    3.722E-07   1.671E-07   5.883E-OS 2.766E-OB  1.676E-OS   1.139E-OG SE       1.836E-05    4.834E-06   1.650E-06  8.720E-07    5.555E-07   2.529E-07   9.116E-OB 4.353E-OB  2.656E-OG   1.815E-OB SSE       1.904E-OS    5.075E-06 . 1.745E-06  9.273E-07    5.933E-07   2.723E-07   9.932E-OB 4.779E-OB  2.925E-OG   2.002E-OG

I Table 3-11 (contd.) d) Turbine Buildin Relative De osition Rate 0/0) Per Unit Area (meter Direction From Site Se nt Boundaries in Hiles from the Site 1-2 2-3 3-4 -5 -0 0- - 0 S 2.244E-OB 4.597E-09 1. 200E-09 5.390E-10 3.049E-IO 1. 'I 73E-10 3.392E-11 1.344E-11 7 ~ 1&DE-12 4.444E-12

 'SSII       1.749E-OB    3.583E-09     9. 353E-10 4.201E-IO      2.376E-10  9. 138E-11    3.644E-11  1.04&E-,11  5 ~ 595E-12 3.463E-12 SW        1.21&E-OB    2.496E-09     6. 515E-10 2.926E-10      1.655E-10   6.366E-11    1.842E-11  7.299E-12   3. 89&E-12  2. 413E-12 MSW        1.0IOE-OB    2.069E-09     5.402E-10  2.426E-10      1.372E-10   5. 17&E-11   1.527E-I 1 6.051E-I2   3. 231E-12  2.000E-12 W       7.46&E-09    1.530E-09     3.993E-10  1.794E-10      1.015E-IO   3.902E-11    1.129E-11  4.474E-12   2. 389E-12  1.479E-12 WIiM       8.961E-09    1.836E-09     4.792E-IO  2.152E-10      1. 21&E-10  4.682E-11    1.355E-11  5.36&E-12   2. 867E-12  1. 774E-12 IQ        1. 615E-08   3.309E-09     8.63&E-IO  3.8&DE-ID      2.195E-10   8.440E-11    2.442E-ll  9.677E-12   5.16&E-12   3.199E-12 NN   ~

3.066E-OB 6.2&DE-09 1.639E-09 7.363E-IO 4.165E-10 1.602E-IO 4.634E-ll 1.837E-11 9.80&E-12 6. 070E-12 N 3.891E-OB 7.970E-09 2.081E-09 9.345E-10 5 '87E-10 2.033E-10 5.881E-11 2. 331E-11 1.245E-ll 7.705E-11 NNE 3.647E-OB 7.471E-09 1.950E-09 8.760E-10 4.956E-IO 1.906E-10 5.513E-11 2. I BSE-11 1.167E-11 7.222E-12 NE 2.492E-OB ~ 5.104E-09 1.333E-09 5.985E-10 3.386E-10 1.302E-10 3.766E-11 1.493E-11 7.972E-12 4.934E-12 ENE 1.906E-OB 3.905E-09 1.019E-09 4.57&E-10 2.590E-10 9.960E-II 2.881E-11 1.142E-I I 6.09&E-12 3.775E-12 E 1.977E-OB 4.050E-09 1.057E-09 4.748E-10 2. 686E-10 1.033E-IO 2.9&BE-10 1.184E-11 6.325E-12 3. 915E-12 ESE 3.404E-O& 6.972E-09 1.820E-09 8.175E-10 4. 624E-10 1.77&E-IO 5.145E-11 2.039E-II 1.089E-11 6. 740E-12 SE 4.158E<<OB 8.518E-09 2.224E-09 9.987E-10 5.650E-10 2.173E-10 6.285E-11 2.491E-11 1.330E-11 8.234E-12 SSE 2.983E-OG 6.111E-09 1.595E-09 7.165E-IO 4. 053E-10 1.559E-10 4.509E-11 1.787E-11 9.543E-12 5.907E-12

Table 3-12 RADWASTE BUILDING X/0 AttD D/ VALUES* a) No Deca Unde leted Cltl/Q (sec/meter cubed) for each segment Direction From Site Se nt Boundaries in Hiles from the Site 1-2 2-3 3- - 0 S 1. 791E-05 5.032E-06 1.836E-06 1.019E-06 6.765E-07 3.337E-07 1.405E-07 7.800E-OB 5.333E-OB 4.018E-OB SSM 2.513E-05 4.282E-06 1.568E-06 8.729E-07 5.765E-07 2.871E-07 1.214E-07 6.758E-OB 4.627E-OB 3.489E-OS SW 1. 419E-05 4.080E-06 1. 513E-06 8.46SE-07 5.651E-07 2.811E-07 1.198E-07 6.690E-OB 4.584E-OB 3.457E-OB WSM 1.004E-05 2 '47E-06 1.044E-06 5.811E-07 3.862E-07 1.909E-07 8.059E-OS 4.481E-DS 3.066E-OB 2.31'IE-08 M 8.834E-06 2.512E-06 9.240E-07 5.149E-07 3.426E-07 1.695E-07 7. 171E-08 3.988f-08 2.72BE-OB 2.056E-OS MttM 8.324E-06 2.320E-06 8.416E-07 4 '54E-07 3.080E-07 1.511E-07 6.317E-OS 3.489E-OB 2.380E-OB 1.791E-OB ttM 9.587E-06 2.620E-06 9.367E-07 5.135E-07 3.377E-07 1.639E-07 6.739E-OS 3.687E-OB 2.506E-OB 1.881E-OS NNM 1.520E-05 4.169E-06 1.494E-06 8.198E-07 5.393E-07 2.620E-07 1.078E-07 5.905E-OB 4.017E-OB 3.015E-OB N 1.661E-OS 4.558E-06 1.636E-06 8.987E-07 5.918E-07 2.881E-07 1.198E-07 6.518E-OS 4.435E-OS 3.329E-OB ttttE 1.259E-05 3.378E-06 1.189E-06 6 '56E-07 4.217E-07 2.025E-07 8. 191E-08 4.445E<<OB 3.015E-OB 2.260E-OB Nf 1.019E-05 2.764E-06 9.837E-07 5.377E-07 3.52SE-07 1.707E-07 6.978E-OS 3.804E-OB 2.581E-OB '1.935E-OB Ettf 9.328E-06 2.528E-06 8.989E-07 4.907E-07 3.215E-07 1.550E-07 6.302E-OB 3.426E-OS 2.322E-OB 1.739E-OB E 8.659E-06 2.344E-06 8.336E-07 4 '53E-07 2.985E-07 1.441E-07 5.868E-OB 3. 191E-08 2. 162E-08 1. 619E-08 ESE 1.452E-05 3.919E-06 1.391E-06 7.573E-07 4.950E-07 2.375E-07 9.577E-OS 5.173E-OB 3.494E-DB 2.611E-OS SE 2.052E-05 5 '57E-06 2.038E-06 1.121E-06 7.387E-07 3.595E-07 1.483E-07 8.123E-OB 5.519E-OB 4. 141E-08 SSE 2.128E-05 5.940E-06 2.156E-06 1. 193E-06 7.895E-07 3.875E-07 1.619E-07 8.949E-OB 6. 108E-08 4.596E-OB <<Ground Level release. Desert sigmas. All stability classes A through G.

Table 3-12 (Cont'd) b) 2.26-Da Deca Unde leted CIII/II (sec/meter cubed) for each seynent Direction From Site Se nt Boundaries in Hiles from the Site

                                                                                                                 - 0 S       1.783E-05     4.991E-06 1.809E-06  9.984E-07      6.586E-07 3.195E-07   1.287E-07  6.725E-OS  4.334E-OB   3.079E-OB SSM        1.506E-05     4.246E-06 1.545E-06  8.547E-07      5.647E-07  2.746E-07  1.110E-07  5.810E-OG  3.745E-OB   2.660E-OB SM        1.413E-05     4.056E-06 1.490E-06  8.292E-07      5.500E-07 2.689E-07   1.095E-07  5.754E-OS .3. 7 I2E-08 2.637E-OG MSM        9.992E-06     2.823E-06 1.029E-06  5.689E-07      3.75&E-07  1.825E-07  7.359E-OB  3.846E-OB  2.475E-OB   1.756E-OB M       8.792E-OG     2.489E-06 9.089E-07  5.030E-07     3.324E-07   1.614 E-07 6.487E-OG  3.36&E-OG  2.152E-OB   1. 515E-08 MIIM       8.286E-06     2.300E-06 8.282E-07  4.549E-07      2.990E-07  1.441E-07  5.731E-07  2.961E-OB  1.891E-OG   1.332E-OO N         9.550E-06     2.600E-06 9.244E-07  5.040E-07     3.295E-07   1.576E-07  6.218E-OB  3.220E-OB  2.073E-OB   1.475E-OB HtlM       1.515E-05     4.145E-06 1.479E-06  B.OBOE-07     5.293E-07  2.541E-07   1.013E-07  5.321E-OB  3.473E-OB   2.503E-OB ll      1.656E-05     4.532E-06 1.619E-06  8.858E-07     5.BOSE-07  2.794E-07   1.117E-07  5.87&E-OB  3.839E-OS   2 '69E-08 HtlE       1.255E-05     3.356E-06 1.175E-06  6.350E-07     4. 12BE-07  1.956E-07  7. 628E-08 3.941E-OB  2.548E-OB   1.821E-OG tlE       1.015E-05     2.743E-06 9.705E-07. 5.274E-07     3. 441E-07  I.63BE-07  6.419E-OB  3.303E-OO  2.117E-OB   1.500E-OB ENE        9.291E-06     2.50&E-06 8.865E-07  4.810E-07     '3.133E-07  1.487E-07  5.78&E-OB  2.966E-OB  1.897E-OB   1.342E-OB E       8.626E-06     2 '26E-06 8.225E-07  4.467E-07     2.912E-07   1.384E-07  5.403E-OB  2.774E-OB  1.777E-O&   1.259E-OB ESE        1.446E-05     3.891E-06 1.383E-06  7.435E-07     4.834E-07  2.285E-07   8.846E-OB  4.521E-OB  2.893E-OB   2.049E-OB SE        2.045E-05     5.618E-06 2.013E-06  1.103E-OG      7.222E-07 3.466E-07   1.376E-07  7.159E-OB  4.625E-OB   3.30IE-OB SSE        2.120E-05     5.895E-06 2.127E-06  1. 170E-06    7.700E-07  3.721E-07   1. 491E-07 7.790E-OB  5.030E-GO   3.583E-OB

Table 3-12 (Cont'd) c) 8.0 Oa Oeca Oe leted (Corrected for Open Terrain Recircu1ation) Ctlt/g (sec/meter cubed) for each segnent Direction From Site .Seqment Boundaries in tliles from the Site

                              -2                                                                0-                                 - 0 S      1.602E-05    4.299E-06    1.486E-06    7.920E-07     5.081E-07     2.342E-07    8.603E-OB  4.155E-OB  2.544E-OS  1.742E-OB SSW       1.353E-05    3.657E-06    1.269E-06    6.782E-07     4.358E-07     2.014E-07    7.428E-OB  3.596E-OB  2.205E-OB  1.510E-OB Sll      1.269E-05    3.485E-06    1.224E-06    6.579E-07     4.244E-07     1.972E-07    7.328E-OS  3.561E-OB  1.460E-OB  1.497E-OB MSM       8.976E-06    2.432E-06    8.448E-07    4.515E-07     2.90IE-07     1.339E-07    4.930E-OB  2.384E-OB  1. 460E-08 9.995E-09 H      7.901E-06    2.145E-06    7.473E-07    3.998E-07     2.571E-07     1.188E-07    4.375E-OB  2. 112E-08 1.291E-OB  8.819E-09 llN       7.446E-06    1.982E-06    6.BOSE-07    3.614E-07     2.312E-07     1.060E-07    3.858E-OB  1.850E-OB  1. 129E-08 7.701E-09 tQ      8.579E-06    2.239E-06    7.584E-07    3.993E-07     2.538E-07     1.152E-07    4. 137E-08 1.971E-OS  1.201E-OB  8.205E-09 tlN       1.360E-05    3.564E-06    1. 211E-06   6.382E-07     4.061E-07     1.847E-07    6.652E-OB  3.185E-OB  1.950E-OS  1.337E-OB N      1.487E-05    3.897E-06    1.326E-06    6.996E-07     4.456E-07     2.030E-07    7.334E-OB  3. 516E-08 2. 153E-08 1.477E-OS HttE      1 '27E-05    2.88BE-06    9.630E-07    5.023E-07     3.173E-07     1.426E-07    5.042E-OS  2.386E-OB  1.454E-OB  9.933E-09 tlE      9.117E-06    2.362E-06    7.964E-07    4.180E-07     2.651E-07     1.199E-07    4.280E-OB  2.030E-OB  1.235E-OS  8.415E-09 EtlE      8.348E-06    2.160E-06    7.277E-07    3.814E-07     2. 416E-07    1.089E-07    3.864E-OS  1.827E-OB  1.109E-OB  7.552E-09 E      7.750E-06    2.003E-06    6.749E-07    3.539E-07     2.243E-07     1.013E-07    3.60IE-OB  1.704E-OB  1.035E-OS  7.046E-09 ESE       1.299E-05    3.349E-06    1.126E-06    5.889E-07     3.722E-07     1.671E-07    5.883E-OB  2.766E-OB  1.676E-OB  1. 139E-08 SE       1.836E-OS    4.834E-06    1.650E-06    8.720E-07     5.555E-07     2.529E-07    9.116E-OB  4.353E-OS  2.656E-OB  1 ~ 815E-08 SSE       1.904E-05    5.075E-06    1.745E-06    9.273E-07     5.933E-07     2.723E-07    9.932E-OB  4.779E-OB  2.925E-OB  2.002E-OB

Table 3-12 (Cont'd) d) Radwaste Buildin Relative Oe sition Rate 0/ Per Unit Area Heter-2) 0 I rect ion From Site Se nt Boundaries in Biles from the Site 5 2.244E-OB 4.597E-09 1.200E-09 5.390E-10 3.049E-IO 1.173E-10 3.392E-11 1.344E-ll 7 '80E-12 4. 444E-12 SSM 1.749E-OB 3.583E-09 9.353E-IO 4 ~ 201E-10 2. 376E-10 9. 138E-11 2.644E-11 1.048E-II 5.595E-12 3. 463E-12 SM 1.218E-OB 2.496E-09 6.515E-10 2.926E-10 1. 655E-10 6.366E-11 1.842E-11 7.299E-12 3.898E-12 2. 413E-12 MSW 1.010E-OB 2.069E-09 5. 402E-10 2.426E-10 1.372E-10 5.278E-11 1.527E-11 6. 051E-12 3. 231E-12 2. OOOE-12 M 7 '68E-09 I ~ 530E-09 3.993E-10 1.794E-10 1.015E-10 3.902E-11 1.129E-11 4.474E-12 2. 389E-12 1.479E-12 WN 8.961E-09 1.836E-09 4.792E-10 2.152E-10 1. 218E-10 4.682E-11 1.355E-11 5.368E-12 2.867E-12 1. 774E-12 N l. 615E-08 3.309E-09 8.638E-10 3.880E-10 2. 195E-10 8.440E-11 2.442E-II 9.677E-12 5.168E-12 3.199E-12 HN 3.066E-OB 6.280E-09 1.639E-09 7.363E-10 4.165E-10 1.602E-10 4.634E-11 1.837E-11 9.808E-12 6.070E-12 ti 3.891E-OB 7.970E-09 2.081E-09 9 '45E-10 5.287E-IO 2.033E-10 5.881E-11 2.331E-ll 1.245E-11 7.705E-12 HttE 3.647E-OB 7.471E-09 1.950E-09 8.760E-IO 4.956E-10 1. 906E-.10 5.513E-11 2.185E-11 1.I67E-11 7.222E-12 ttE 2.492E-OB 5.104E-09 1.333E-09 5.985E-IO 3.386E-IO 1.302E-10 3.766E-11 1.493E-11 7;972E-12 4 '34E-12 EttE 1.906E-OB 3.905E-09 1.019E-09 4.578E-10 2. 590E-10 9.960E-ll 2.881E-11 1. 142E-11 6.098E-12 3.775E-12 E 1.977E-OB 4.050E-09 1.057E-09 4.74BE-10 2.686E-10 1.033E-10 2.988E-11 1.184E-11 6.325E-12 3. 915E-12 ESE 3.404E-OB 6.972E-09 1.820E-09 8.175E-10 4. 624E-10 1.778E-10 5.145E-11 2. 039E-11 1.089E-11 6 740E- I2

                                                                                                                             ~

SE 4.158E-OB 8.518E-09 2.224E-09 9.987E-10 5.650E-10 2.173E-10 6.285E-11 2.491E-ll 1.330E-11 8.234E-12 SSE 2.983E-OB 6.111E-09 1.595E-09 7.165E-IO 4.053E-10 1.559E-IO 4.509E-116 1.787E-11 9.543E-12 5.907E-12

Table 3-13 CHARACTERISTICS OF WNP-2 GASEOUS EFFLUENT RELEASE POINTS Reactor Radwaste Turbine Buil din Buildin Buildin Height of release point above ground level (m) 70.6m 31.1 27.7 Annual average rate of air flow from release point (m3/sec) 44.8 38.7 125.6 Annual average heat flow from release point (cal/sec) 1.06 x 106 2.9 x 106 9.1 x 105 Type and s i ze o f re 1 ease Ouct 3 Louver houses 4 Exhaust fans 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 (m2) 3. 48 2x 2.7 3 x 2.91 Vent velocity (m/sec)* 12.9 2 x 525 cfm** 14.4 Effective diameter (m) 1.0 ( ~r2 = 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 Orawing 6-41, 525 cfm x 2 out of 3, will run at any one time.

85

0 ~ " r ~ i ~. ~ ' or,>> aaa .'wa w~ J r . (...~: ~ . i'~ . " w - ~ .v-., - ~ AMENDMENT NO. 7 December 1989 Table 3-14 REFERENCES FOR VALUES LISTED IN TABLE 3-9 Reference 1 U.S. Hap Reference 2 Site Specific Reference 3 Regulatory Guide 1.109, Revision 1, Table E-15 Reference 4 Section 2.3, MNP-2 FSAR, Table 2.3-1 Reference 5 Section 2.3, WNP-2 FSAR, page 2.3-3 Reference 6 MNP-1 8 MNP-2 Emergency Preparedness Plan Table 12.1, Permanent Population Distribution, Rev 5, Feb. 88 Reference 7 1986 50-Nile Land Use Census, WPPSS RENP Reference 8 WNP-2 Effluent Analysis for Applicable Time Period Reference 9 Radiological Programs Calculation Log No. 88-3 Reference 10 NUREG/CR-2919 XOQOOQ: Computer Program For The Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations, September 1982. 86

Table 3-15 OESIGN BASE PERCENT NOBLE GAS 30-MINUTE OECAY)* Isoto e Percent of Total Activit Kr-83M 2.9 Kr-85M 5.6 Kr-85 0 Kr-87 15 Kr-88 18

                    'I Kr-89                           0.2 Xe-131M                         0. 02 Xe-133M                         0.3 Xe-133                          8.2 Xe-135M                         6.9 Xe-135                         22 Xe-137                          0.7 Xe-138                         21

• From Table 11.3-1 WNP-2 FSAR 87

AMENDMENT NO. 7 December 1989 TABLE 3-16 ANNUAL DOSES AT TYPICAL LOCATIONS Source: MNP-2 Gaseous Ef fluent Whole Thyroid Distance Occupancy Body Oose Oose Location Miles hrs/ r mrem/ r mrem/ r BPA Ashe Substation 0.5 N 2080 1.1E+00 1. 7E+00 OOE Train 0.5 SE* 78 6.7E-02 1.0E-01 Mye Burial Site 0.5 MNM 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-O') 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/l} values (within 0-0.5 mile radius) was used.

   **Closest residental 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.

88

AMENDMENT NO. 7 December 1989 TABLE 3-17 ANNUAL OCCUPIED AIR DOSE AT TYPICAL LOCATIONS Annual Annual Beta Air dose Gamma Air Dose Location mrad mrad BPA Ashe Substation 8.9E-01 1.5E+00 DOE Train 5.3E-02 9.2E-02 Mye Burial Site 3.2E-03 5.7E-03 MNP-1 3.3E-02 2.8E-02 NNP-4 5.3E-02 8.5E-02 NNP-2 Yisitor 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.

89

4 4K

                                       -.8 g                   X Q
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O u C Q eO Q g Vl u E 1L Q. 0., Site Boundary for Radioactive Gaseous and Liquid Eff1uents Figure 3-1 NUCLEAR UNIT 2 90'ASHINGTON

AMENDMENT NO. 6 November 1988 Plenum Noni tor s ~Rx Bld . Refueling Exhaust To Pump Rooms Plenum Elevated Main Steam Tunnels Release Plenum Monitors Turbine Bld . Cond. Pump Rm. Exhaust To Turb. Opr. Deck Plenum Atmosphere Condenser Area Radwaste Bld . Hot Mach. Shop Filter To Chem. Labs Units Atmosphere Demin. Room Radwaste Proc. Area RM Control Room SIMPLIFIED BLOCK DIAGRAM OF GASEOUS HASTE SYSTEM Figure 3-2 90a

I L

AMENDMENT NO 6 November 1988 Main Condenser Air Ejector, Off-Gas Condensers Preheaters Hater Off-Gas Off-Gas Separator Condenser Recombiners Gycol Off-Gas Off-Gas Cooler Moisture Pre-Filter s Condenser Separator s Charcoal Coolers Dryers Absorbers Post Treatment Monitor, After Filters Elevated Release SIMPI IFIED BLOCK DIAGRAM OF OFF"GAS TREATMENT SYSTEM Figure 3-3 90b

AMENDMENT NO. 3 February 1986 4.0 COMPLIANCE WITH 40 CFR 190 4.1 Technical S ecification Re uirement Technical Specification 3.11.4 states, "The annual (calendar year) dose or dose commitment to any Member 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 ODCM Methodolo for Determinin Dose and Dose Commitment from Uranium Fuel Cycle 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. c) Air doses due to noble gases released in gaseous effluents. d) Dose to any or gan 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. 91

0 AMENDMENT NO. 4 August 1986 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 Technical Specification 3. 11.1.2a, 3. 11.1.2b, 3.11.2.2a, 3.11.2.2b, 3.11.2.3a, or 3.11.2.3b. Direct radiation measurements will also be made to determine if the limits of Specification 3.11.4 have been exceeded. 4.2.1 Total Dose from Li uid Effluents The annual dose to a Member 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 dose contribution pathways to a 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 Member 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 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 sample 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. 91a

AMENDMENT NO. 1 May 1984 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 (REMP) 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. The WNP-2 REMP 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. 92

AMENDMENT NO. 3 February 1986 Radiological environmental monitoring activities implemented by PPM 1.11.1 "Radiological Environmental Monitoring Program (REMP) Implementation Proce-dure", as detailed in the following sections, meet or exceed the criteria of the REMP plan as specified by Plant Technical Specifications, 3/4.12. 5.1 Radiolo ical Environmental Monitoring Program (REMP) Environmental samples for the REMP are collected in accordance with Table 5-1. This table provides a detailed outline of the environmental sampling plan including both Technical Specification and non Technical Specification items by sample type, sample location code, sampling and collection frequency, and type and frequency of analysis of samples collected within exposure path-way. 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, 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 speci-men. All deviations from the sampling plan Technical Specification items detailed in Table 5-1 shall be documented and reported in the Annual Radio-logical Environmental Operating Report in accordance with PPM 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 sub-stitution made shall be implemented in the sampling program within 30 days of identification of the problem. All changes implemented in the sampling pro-gram due to unavailability of samples shall be fully documented in the next Semiannual Radioactive Effluent Release Report and ODCM per PPM 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. 93

AMENDMENT NO. 7 December 1989 WNP-2 sampling stations are described in Table 5-2. Each station is iden-tified by an assigned number or alphanumeric designation, meteorological sector (16 different, 22-1/2 compass 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 Technical Specification 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 the Plant Technical Specifications. It shall identify within a distance of 8 km (5 miles) in each of the 16 meteorological sectors, the location of the nearest milk animal, the nearest residence and the nearest garden of greater than 150m 2 (500 ft2 ) 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. The information obtained during the field survey is used along with other demographic data to assess population changes in the unrestricted area that might require mod-ifications 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 percent greater estimated dose or dose commitment than a currently sampled location, 94

AMENDMENT NO. 7 December 1989 will be reported in the next Semiannual Radioactive Effluent Report in accor-dance with PPM 1.10.2 and Technical Specification 3/4.12.2. The REMP plan, ODCM, will be changed to reflect new sampling location(s). The new sampling location(s) wi 11 be added to the REMP 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 PPM 1.10,2 and Technical Specification requirements. 5.3 Laborator Intercom arison Pro ram Analysis of REMP samples is contracted to a provider of radiological analyti-cal services. By contract, this analytical service vendor is required to con-duct 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 participa-tion in the Environmental Protection Agency's (EPA) Environmental Radioactiv-ity 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. The results of the contractor's analysis of Crosscheck samples shall be included in the Annual Radiological Environmental Operating Report in accordance with PPM 1.10.2 and Technical Specification. Besides the vendor's required participation in the EPA's Crosscheck Program, the Department of Social and Health Services (DSHS) 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/DSHS 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 95

h AMENDMENT NO. 4 August 1986 o4her 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 REMP Annual Report, when available. 5.4 Reportin Re uirements WNP-2 radiological environmental surveillance program activities are presented annually per PPM 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, DSHS, by May 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 tabu-lated summary; interpretations and analyses of trends for results of radio-logical environmental surveillance activities for the report period, including comoarisons with operational controls, preoperational studies, and previous environmental surveillance reports as appropriate; an assessment of the observed impacts of plant operation on the environment; a brief description 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 sum-mary shall be presented in a format represented in Table 5-3. A supple-mentary report is required if all analytical results are not available for 96

J AMENDMENT NO. 4 August 1986 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 (1) + Concentration (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 indi-vidual 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. When it can be demonstrated that the anomolous sample result(s) exceeding reporting levels is not the result of plant effluents, a nonroutine report does not have to be submitted. A full discussion of the sample result and subsequent evaluation or investigation of the anomolous result will be included in the Annual Radiological Environmental Operational Report. 97

TA 5-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PLAN 11 Sampling and Type and Frequency Sam le T e Sam le Location Code Collection Fre uenc 1 of Anal sis 1

1. AIRBORNE

a. Particulates and 1, 4-9A, 21, 23, 40, Continuous sampling Particulate: Gross radioiodine (5/12) 48 and 57 Weekly collection ii gamma isotopic3, quarterly composite (by location)

Radioiddine: I-131 analysis, weekly

b. Soil10 9A, 1, 7, 21, and 23 Annually Gamma isotopic3, (0/5) annually strontium-90 when requested 10

2. DIRECT RADIATION TLD4 1-9A, 10-25, 40-47, Quarterly, annually TLD converted to (34/57) 49-51, 53-56, 61, exposure quarterly, 1S-16S annual processing

3. WATERBORNE

a. Surface/ 26, 27, 28 and 29 Composite aliquots5, Gamma isotopic>,

Drinking6 monthly gross beta, monthly; (3/4) tritium, quarterly CD % composite CD O Pl strontium-90, CD ~ CD iodine-131, when CX M CD Pl requested6 s m

b. Ground water 31, 32, and 52 Quarterly CD ~

Gamma isotopic3 00 CD IQ ~ (2/3) and tritium, quarterly

TABLE (contd.) Sampling and Type and Frequency Sam le T e Sam le Location Code* Collection Fre uenc 1 of Anal sis WATERBORNE (contd.)

c. Sediment from 33 and 34 Semiannually Gamma isotopic>,

shoreline semiannually (1/2)

4. INGESTION

a. Hi 1k> 9B, 36, 40, 59 Semimonthly during Gamma isotopic3 and (4/5) and 96 grazing season, iodine-131, monthly/

monthly at other semimonthly times strontium-90, when requested~ b ~ FishB 30, 38, or 39 Seasonal or Gamma isotopic3, when (2/2) Semiannually sampled

c. -

Garden produce9 9C, 37 and 91 Honthly during. grow- Gamma isotopic>, when (2/3) ing season in the sampled Riverview area of Pasco and a control near Grandview. Annually for the apple sample collection at Station 91

     *Sample    locations are graphically depicted in Figures 5-1      and  5-2.

1Deviations are permitted if samples are unobtainable due to hazardous conditions, seasonal avail-ability, malfunction of automatic sampling equipment, or other legitimate reasons. All deviations will be documented in the Annual Radiological Environmental Honitoring Report. 2Particulate sample filters will be analyzed for gross beta after at least 24-hour decay. If gross beta activity is greater than 10 times the yearly mean of the control sample, gamma isotopic analysis shall be performed on the individual sample. Gamma isotopic means identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents of the facility.

TABLE (contd.) 4TLD refers to thermoluminescent dosimeter. For purposes of WNP-2 RBlP, 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 REHP 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 lS-16S are special interest stations and are not included amongst the 34 routine TLD stations required by Plant Technical Specifi-cation, Table 3.12-1. 5Composite samples will be collected with equipment which is capable of collecting an aliquot at time intervals which are short relative to the compositing period. 6Station 26, WNP-2 makeup water intake from the Columbia River, satisfies the Technical Specifica-tion criteria for upstream surface water and drinking water control samples. The discharge water (Station

27) samples are used to fulfill the Technical Specification 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 than 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.

7Hilk samples will be obtained from farms or individual milk animals which are located in sectors with high calculated annual average ground-level D/Qs 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. BThere are no commercially important species in the Hanford reach of the Columbia River. Host recreationally important species in the area are anadromous, primarily salminoids. Four fish specimen will normally be collected by electroshock technique in the vicinity of the plant discharge (Station 30). If electroshocking produces insufficient fish samples, anadromous species may be obtained from Ringold Fish Hatchery (Station 39). Control samples are normally collected in the vicinity of Ice Harbor Dam (salminoids may be obtained through the National t3arine Fisheries Service at Lower Granite Dam). 9Garden produce will routinely be obtained from farms or gardens using Columbia River water for irrigation. One sample of a root crop, leafy vegetable, and a fruit should be collected each sample period if available; The variety of the produce sample will be dependent on seasonal availability. 10Soil samples are collected to satisfy the requirements of the Site Certification Agreement (SCA), WNP-2. If gamma isotopic results for a sample are greater than ten times the mean of the control station (station 9) data, the sample shall be analyzed for Sr-90.

TABLE 5-1 (contd.) The fraction in parenthesis under each sample type gives the ratio of the number of Technical Specificaton 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.

E 5-2 WNP-2 REMP LOCATIONS Station Sector Radial Nil esa TLD AP/AI SW DW GW SE MI FI GP SO b 1 S 1.3 0 X 2 NNE 1.8 0 3 SE 2.0 X 4 SSE 9.3 0 0 5 ESE 7.7 0 X 6 S 7.7 0 X 7 WNW 2.7 0 X 8 ESE 4.7 0 0 9A* MSW 30.0 0 0 9B* WSW 35.0 9C WSW 33.0 10 E 3.1 ll ENE 3.1 12 NHW 6.1 13 SW 1.4 14 WSW 1.4 15 W 1.4 C Z EA C 2: 16 MNW 1.4 Eh c+ M Pl 17 NNW 1.2 lO W C) Ol K CD

TABLE 5-2 (Continued) Station Sector Radial Miles TLD AP/AI SII DM GII SE NI ..FI GP SOD 18 N 0 19 HE 1.8 0 20 ENE 1.9 0 21 ENE 1.5 X X 22 E 2.1 0 23 ESE 3.0 X X 24 SE 1.9 . 0 25 SSE 1.6 0 26* E 3.2 0 0 27 E 3.2 X 28 SSE 7.4 0 0 29 SSE 11.0 0 30 E 3.5 31 E 32 E 1.2 33* ENE 3.3 34 ESE 3.3 35 ENE 10. 5 36 ESE 7.2 37A SSE 17.0 37B SSE 16.0 38* E 26.5 (95.0)

8 LE5-2 (Continued) Station Sector Radial Hiles TLD AP/AI SW DW GW Sf HI FI GP SOb 39 NE 4.3 40 SE 6.4 0 0 41 SE 5.8 0 42 ESE 5.6 0 43 E 5.7 0 44 ENE 5.7 0 45 ENE 4.2 0 46 NE 4.7 0 47 N 0.5 48 Nf 4.3 49 NW 1.2 50 SSW 1.2 51 ESE 2.1 52 N 0.1 53 N 7.5 54 NNE 6.5 55 SSE 7.0 56 SSW 7.0 57 N 0.7 59 SE 9.6 '0 91 ~ ESE 4.5 96 WSW 36.0

TABLE 5-2 (Continued) Station Sector Radial Miles TLD AP/AI SW DW GW SE tlI FI GP SOb ls H 0.3 X 2S HHE 0.4 X 3S NE 0.5 X 4S ENE 0.4 X 5S E 0.4 X 6S ESE 0 X 7S SE 0.5 X 8S SSE 0.7 X 9S S 0.7 X 10S SSW 0.8 X 11S SW 0.7 X 12S WSW 0.5 X 13S'4S W 0.5 X WHW 0.5 X 15S NW 0.5 X 16S HHW 0.4 X Contro ocation. X-Sample collected at station (non-RETS) 0-Radiological Environmental Technical Specification (RETS) sample collected at station. aEstimated from center of MNP-2 Containment from map positions. Included in sampling program to satisfy requirements for Site Certification Agreement with the State of Washington. AP/AI = Air Particulate and Iodine SW = Surface Mater (River Water) DW = Drinking Mater GW = Ground Mater 'SE = Shoreline Sediment HI = t1i1 k FI = Fish GP = Garden Produce SO = Soil

AMENDMENT HO. 6 November 1988 (I I OOttvot

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1 4J Slt'PPLY SYSTEM 880334 RADIOLOGICAL ENVIRONMENTAL MONITORING SAMPLE LOCATIONS INSIDE OF 10 MILE RADIUS Figure 5-1 106

Othe ao NASHINGTON I L 30B Lower I Connell errata Suer Grenilo Dam

                                                                                                                                                                    !   IDAHO Priest Rapirls        llenlord Dem              Rose/var/on                                                                                                                               I Lower           Lirde Goose Aronumonlel            Dam           Pomeroy                                t IVNP 2                                            Dent 0

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TABLE 5-3 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM ANNUAL SUHMARYa Name of Facility Docket No. Location of Facility Reporting Period County, tate Location with Highest Hedium or Type and All Indicator Annual Hean Number of Pathway Sampled Total Number Lower Limit Locations Name Haan t Control Locattons Nonroutine (Unit of of Analyses of Detection Hean (f) Distance and Hean (f) Reported Measurement Performed Oirect <on ~Ran e Measurements Air particulates Gross 416 0.01 0.08 (200/312) Hiddletown 0.10 (5/52) 0.08 (8/104) (pCi/m3) (0.05-2.0) 5 mi. 340'0.08-2.0) (1.05-1.40)

                       -Spec 32 137Cs                 0.01         0.05 (4/24)         Smithville      0.08 (2/4)      LLO (0.03-0.13)         2.5 Nil. 160'0.03-2.0) 1311                  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/kq) -Spec. 8 (wet weight) 137Cs 130 LLD LLO 90 (I/4) 134Cs 130 LLD LLD LLD 130 180 (3/4) River Mile 35 See Column 4 LLD (150-225) aSuamary Table is taken from the NRC's Branch Technical Position, Rev. 1, Nov. 1979, and provided for illustrative purposes only. cHean and range based upon detectable measurements only. Fraction of detectable measurements at specified locations is indicated in parentheses (f).

TABLE 5-4 REPORTING LEVELS FOR NONROUTINE OPERATING REPORTS eport>ng Leve RL Airborne Particulate Broad Leaf

     ~Anal sis      Water            or Gases                 Fish             ~V (pCi/1)            (pCi/m3)           (pCi/kg, wet) (pCi/1) (pCi/Kg, wet)

H-3 2 x 104* Mn-54 1 x 103 3 x 104 Fe-59 4 x 102 1 x 104 Co-58 1 x 103 3 x 104 Co-60 3 x 102 1 x 104 Zn-65 3 x 102 2 x 104 Zr-Nb-95 4 x 102 I-131 0.9 1 x 102 Cs-134 30 10 1 x 103 60 1 x 103 Cs-137 50 20 2 x 103 70 2' 103 Ba-La-140 2 x 102 3x 102

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

AMENDMENT NO. 3 February 1986 6.0 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT Routine Radioactive Effluent Release Reports covering the operation of MNP-2 during the previous 6 months of operation are submitted within 60 days after January 1 and July 1 of each year. These reports shall include a summary of the quantities of radioactive liquid and gaseous effluents released from the unit (MNP-2). Reports shall include each class of soild waste (as defined by 10 CFR 61) shipped offsite during the reporting period with the following information; container volume, total curie quantity, principal radionuclides, source of waste and processing employed, container type, and solidification agent or absorbent. 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 made during the reporting period. The Radioactive Effluent Release Reports include any changes made during the reporting period to the Process Control Program and to the ODCM pursuant to Technical Specification 6.13 and 6.14, respectively, as well as any major change to Liquid, Gaseous, or Solid Radwaste Treatment System, pursuant to Technical Specification 6.15. It also includes a listing of new locations for dose calculations and or environmental monitoring identified by the Land Use Census pursuant to Technical Specification 3.12.2. The Radioactive Effluent Release Reports also include an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Technical Specification 3.3.7.11 or 3.3.7.12, respectively; and a description of the events leading to liquid holdup tanks exceeding the limits of Technical Specification 3.11.1.4. 110

AMENDMENT NO. 7 December 1989 The Radioactive Effluent Report to be submitted within 60 days after January 1 of each year includes an annual summary of meteorological data collected over the previous year. This annual summary will be in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. This same report includes an assessment of the radiation doses due to the radio-active liquid and gaseous effluents released from the unit during the previous calendar year. This same report also includes, an assessment of the radiation doses from radioactive liquid and gaseous effluents to Members of the Public due to their activities inside the Site Boundary during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, are included in these reports. The assessment of radiation doses will be performed in accordance with the methodology and parameters in the OOCM, using the NRC computer programs LADTAP II and GASPAR II for the actual dose determinations. The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year also includes, as required by Technical Specification 3.11.4, an assessment of radiation doses to the likely most exposed Member of the Public from NNP-2 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 190, "Environ-mental Radiation Protection Standards for Nuclear Power Operation".

Reference:

10CFR50.36a(a)(2)

      .8908280226, WNP-2 SEMIANNUAL EFFLUENT REPORT REPORTING PERIOD JANUARY THROUGH JUNE 1989 WASHINGTON PUBLIC POWER SUPPLY SYSTEM LICENSE NO. NPF-21

' (" j I

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TABLE OF CONTENTS KGDM

1.0 INTRODUCTION

.                                      1 2.0  LIQUID EFFLUENTS                                    1 3.0  GASEOUS EFFLUENTS                                   5 4,0  RADHASTE .                                         16 5.0  METEOROLOGICAL DATA                                22 6.0  DOSE ASSESSMENT  IMPACT ON MAN                    23 7.0  REVISIONS TO THE  ODCM                             24 8.0  REVISIONS TO THE PROCESS CONTROL PROGRAM (PCP)     25 9.0  NEH OR DELETED LOCATIONS FOR DOSE ASSESSMENTS AND/OR ENVIRONMENTAL MONITORING   LOCATIONS....... 26 10.0 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS AND SOLID HASTE TREATMENT SYSTEMS.                 27

LIST OF TABLES 2-1 WNP-2 LIQUID EFFLUENTS SUMMATION OF ALL RELEASES JANUARY JUNE 1989 . 2-2 WNP-2 LIQUID EFFLUENTS SOURCE TERHS-JANUARY JUNE 1989. 3-1 HNP-2 GASEOUS EFFLUENTS SOURCE TERHS-HIXED MODE RELEASES HAIN PLANT VENT JANUARY JUNE 1989. 3-2 NNP-2 GASEOUS EFFLUENTS SOURCE TERMS GROUND LEVEL RELEASES TURBINE BUILDING JANUARY JUNE 1989 10 3-3 HNP-2 GASEOUS EFFLUENTS SOURCE TERMS GROUND LEVEL RELEASES RADHASTE BUILDING JANUARY JUNE 1989 ~ ~ 0 1 3-4 NNP-2 GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES JANUARY JUNE 1989 14 3-5 NNP-2 GASEOUS EFFLUENTS BATCH RELEASES JANUARY JUNE 1989 15 4-1 SCALING FACTORS FOR REQUIRED NUCLIDES ~ ~ o

                                                     ]9 4-2 SCALING FACTORS FOR CONDITIONAL NUCLIDES          19 4-3 HNP-2 SOLID HASTE SHIPMENTS JANUARY  JUNE 1989 20

1 1.0 This report is submitted in compliance with Technical Specification 6.9.1.11. It includes a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from NNP-2 during the previous six months of operation, with data summarized on a quarterly basis. 2.0 The radwaste liquid effluents were released in "batch mode" during the reporting period. No liquid releases occurred during the first calendar quarter and 51 batch releases occurred during the second calendar quarter. The total time period for the batch releases was 98.4 hours, with the maximum, minimum and average time periods for a release being

3. 17, 0.05 and 1. 93 hours respectively. The volume of dilution water considered is assumed to be the total volume of recirculating cooling tower blowdown flow for the period. The average flow rate of the Columbia River during January through June 1989 was 1.14E+05 cubic feet per second.

Periodic LADTAP II computer runs were performed to verify compliance with Technical Specification limits. There were no liquid releases during the first quarter. The second quarter calculated dose for the maximum indi-vidual (adult age group) was 1,8E-03 mrem whole body and 4.3E-03 mrem for the maximum organ. No abnormal liquid releases occurred during this reporting period. The liquid batch releases were recirculated prior to sampling. A repre-sentative sample was obtained and analyzed for each batch release, A composite of the batch samples for each quarter was analyzed for strontium and iron analyses. The methods used for measuring the total radioactivity were gamma spectroscopy, liquid scintillation and propor-tional counting. Table 2-1 provides a summation of all liquid releases during this reporting period. The percent of MPC limit is based on the total of the MPC fractions using the nuclides in Table 2-2 and the concentrations listed in 10CFR20, Appendix B, Table 2, Column 2. Estimated total errors are listed in Table 2-1, and are propagated from individual error estimates of sample activity, sample volume, tank volume, and tank homogeneity. The estimated total errors were calculated by obtaining the square root of the sum of the squares of the individual error contributions and multiplying by 1.96 for a 951. confidence level.

Table 2-1 HNP-2 LIQUID EFFLUENTS SUMMATION OF ALL RELEASES Report Period: January June 1989 1st 2nd Est. Quarter Quarter Total E~~A A. Fission and activation products

1. Total release (not including A ~K-g2

2. Average diluted concentration ml B. Tritium

2. Average diluted concentration r

C. Dissolved and entrained gases

2. Average diluted concentration D. Gross alpha radioactivity E. Volume of waste (prior to F, Volume of dilution water

• At 951. confidence level
  • There were no liquid releases during the first quarter of 1989.

Table 2-2 HNP-2 LIQUID EFFLUENTS SOURCE TERMS Report Period: January June 1989 BATCH MODE 1st 2nd Nuclides Released Unit Quarter Quarter

                                                       .4
                                                            -0 1.4 e-5                                       2.1 E-4 Te hn                                                    -0 B  i m-                                           2.7    -  4

l r

TABLE 2-2 (Continued) E-E- Tr i " There were no liquid releases during the first quarter of 1989. 50K: Less than (<) values are not included in the Total For Period values.

3.0 The gaseous radwaste effluents from WNP-2 were released from three (3) release points:

1. Main Plant Vent mixed mode release

2. Turbine Building ground level release

3. Radwaste Building ground level release The gaseous source terms from each release point are listed in Tables 3-1, 3-2, and 3-3. Table 3-4 provides a summation of the total activity released, the average release rate, the percent of Technical Specifica-tion limit, gross alpha radioactivity and the estimated total error associated with the measurements of radioactivity in the gaseous eff1 vents.

Radioactivity measurements for gaseous effluent releases are performed for fission and activation gases by collecting the samples on charcoal traps and analyzing them using gamma spectroscopy. Tritium is sampled by

   . freeze trapping and analyzed by liquid scintillation counting. Particu-lates and iodines are sampled using particulate filters and charcoal cartridges and are analyzed using gamma spectroscopy.

The percent of Technical Specification limit used in Table 3-4 is the most restrictive value based on the quarterly dose evaluations. The fission and activation gases (air dose) calculation is based on quarterly limits of ten (10) milli rads for beta and five (5) milli rads for gamma. The percent of Technical Specification limit calculations for iodines, particulates with half-lives greater than eight (8) days and tritium are based on the quarterly limit of 7.5 mrem to any organ. Locations 3 through 7 listed below as identified by the latest Land Use Census (LUC) were used to determine the most restrictive value for each quarter to be used in Table 3-4. Total error estimates are propagated from individual error estimate's of sample volume, sample activity and effluent flow rate measurements. The overriding uncertainty in all cases is in the measurement of the effluent and sample volumes. The estimated error was determined to be 36% at the 95'/ confidence level. 1 I Calculations were performed for releases using the GASPAR II computer program and parameters as outlined,in the ODCM. Quarterly doses were determined at 'the 'following locations: Site Boundary; 1.2 miles al*Jto Beta

                    ~~          l  MQt 2.7 E-03          0,03            1.4 E-02          0. 14 Gamma            4.6 E-03          0.09            1.1 E-02          0.22

Beyond Site Boundary; 4.2 miles ESE AIM ~'~ Beta 1MQt 8.1 E-03 0.08 1.5 E-02 0.15 Gamma 9.6 E-03 0.19 1.0 E-02 0.20 inhalation pathways)

          ~D   4.8 miles      (ground, vegetables n  3:              SE                          and IUIII 1st Qtr mern 4.6 E-03
                                                ~~m  0.06 2nd Qtr               2.3 E-02             0.31 6.4 miles SE (ground, vegetables,        meat, cow   milk, and  inhala-tion pathways)

RlU 1st Qtr IKQ!6 3.0 E-03 L3e~m. 0.04 2nd Qtr 4.2 E-02 ~ 0.56 4.2 miles ESE (ground, vegetables and inhalation pathways) HiaheM~~hm m~ LZe~zm 1st Qtr 4.5 E-03 0,06 2nd Qtr 1,9 E-02 0,25 LkgU~L5: 4.3 miles NE (ground and inhalation pathways) mr~ 1st Qtr 6.6 E-04 0.01 2nd Qtr 1.0 E-03 0.01 4.1 miles ENE (ground, vegetables and inhalation pathways) HiahuMQzg~ma,, mrsE-03 1st Qtr 3.3 0.04 2nd Qtr 1. 9 E-02 0.25 t In addition to the reactor site, WNP-2 has a permanent laundry facility located approximately 0.75 miles from the site. Its ventilation system contains HEPA filters on the discharge and is continuously monitored for particulates. Also near this location is a backup chemistry laboratory within the Emergency Operations Faci'lity (EOF). The radiochemical hood within the chemistry lab contains HEPA filters and is monitored for radioactive releases when in operation. Gamma spectrometry indicated no radioactive material present other than that attributable to natural background. There were no abnormal releases of gaseous effluent during the first and second quarters of 1989.

Table 3-1 NNP-2 GASEOUS EFFLUENTS SOURCE TERMS MIXED MODE RELEASES MAIN PLANT VENT Report Period January June 1989 CONTINUOUS MODE 1st 2nd r r

1. Fission gases

4. +

l. +1 2.4 +

e m .1 E+0 n- 35 nn- 5

2. Iodines 3

I in- 2 2. -0 1. E-

Table 3-1 (Continued)

3. Particulates B~rg~

                                   ~ 2

~.abtuaM 9 w hT-

4. Tri i m

~N : Less than (<) values are not included in the Total For Period values.

Table 3-2 NNP-2 GASEOUS EFFLUENTS SOURCE TERMS GROUND LEVEL RELEASES TURBINE BUILDING Report Period January June 1989 CONTINUOUS MODE 1st 2nd

1. Fission gases r n-5 E-E- 1 r n-7

                                                   +0 n  n-1                                     f-   1      E 0

n-1 5 E- 1 E- 1

2. Iodines I in -1 2. E- 4 I n-1 E- E- 4 E-

~" Table 3-2 (Continued)

3. Particulates r i 2. E- 4 r n s - 3 E- 4 M 1 .OE-3 E-r m-14 E-E-

I -5 E- 4 2. E- 4 Zin -6 4 E- 4 ri im 1. E+ 1 1 e + ~T : Less than (<) values are not included in the Total For Period values, with exception being Iodines for the First Quarter. 11

Table 3-3 WNP-2 GASEOUS EFFLUENTS SOURCE TERMS GROUND LEVEL RELEASES RADWASTE BUILDING Report Period January June 1989 CONTINUOUS MODE 1st 2nd

1. Fission gases n-1 n-1 E- 1 E-

2. Iodines E-12

Table 3-3 (Continued)

3. Particulates

2. E-

                         -L  n -1                            ~~~E-     5

l. -4 eri m-Ir -5 2. E- 5 2. E-Zi .0 E-T 1 f E-Tr i m E-T 1 1 5QTT,: Less than (<) values are not included in the Total For Period values.

I Table 3-4 HNP-2 GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES Report Period January June 1989 1st 2nd Est. Total

                                                                                 'I'.

Fission 5 activation gases 1.4 + 2

2. Average release r i d aiSaax

3. Percent of Tech.

                                                          -0    2. E- 1 B. Iodines

1. T al I ine 3. E-0 .6 E+ 1

2. Average release 3.

I Percent of Tech.

4. 4 K3 E9 0

0 E- 1 C. Particulates

1. Particulates s E- 2 ~~E+ 1

2. Average release 5.

3. Percent of Tech.

4. 1 Gross alpha

                                                          -  2  '~E~

11 II D. Tritium

2. Average release r ri E- 1

3. Percent of Tech.

0 C. 0 E- 1 " At 95% confidence level

Table 3-5 NNP-2 GASEOUS EFFLUENTS BATCH RELEASES Report Period January June 1989 Total Maximum Minimum Mean i e hr h Purge 35.95 13.61 3.42 5.99 Ven 50': Batch releases were performed through the Hain Plant Vent mixed mode release. 15

/ 4.0 A total volume of 6164.5 ft (174.58 m ) of solid waste was transported in 14 shipments during the January through June, 1989 reporting period. The total activity of the waste shipped was 313.69 Ci; 312 Ci were con-tained in dewatered spent resins, 1.69 Ci were contained in Dry Active Haste (DAH). A. Dewatered resins accounted for 2,464,5 ft (69.79m ) of the radio-active wastes shipped during the reporting period. The burial con-tainers were ES-190 and ES-142 liners provided by NUPAC Services, Inc. The total activity of the resins shipped during the reporting period was 312 Ci. The principle nuclides and their percent contri-bution to the total activity are listed in Table 4-3. The solid wastes were shipped to the U.S. Ecology, Hanford burial site using flat bed trailers and NUPAC 10-142, or LN-14-170 casks. The counting error associated with the total activity has been found to be less than 1.0% at one standard deviation in previous effluent reports and to decrease with increasing activity. The statistical counting error is assumed to be 1% for the purpose of this error evaluation. Other parameters considered in estimating the total error of the activity shipped included the error in measuring the absolute volume, the weight of the waste in the liners, the representative-ness of the sample taken, the homogeneity of the nuclide distribu-tion within a batch or liner and the geometry error in the gamma spectroscopy analysis. The gamma spectroscopy calibration error was approximately 5%. The best estimate of the total error in the activity of spent resin shipped was assumed to be less than or equal to 20%. B. D v D A total of 3700 ft (104.78 m ) of DAW was shipped in 40 Container Products Corporation, B-25 steel boxes. The total activity of the DAW shipped was 1.69 Ci. The values for the activities shipped were determined by using dose rate-to-curie conversion factors. The con-version factors were based on nuclide'distribution taken from analy-sis of contamination found in each of the major DAH production areas. The nuclide distribution is updated monthly. Short-lived nuclides were eliminated based on decay of the DAW prior to shipment. A meaningful counting error cannot be'enerated for the DAH; however, the total error may be assumed to be less than or equal to 20%, since DAH would be subjected to similar error contributions as the spent resins. 16

lt t t p '1

   ;!       j

i F Scaling factors are based on outside laboratory (SAIC) analysis of hard-to-measure nuclides. The process of updating scaling factors has been initiated. For those waste streams where the scaling or the scaled nuclide concentration is not sufficient to provide a viable scaling factor, the final EPRI Report "Updated Scaling Factors in Low Level Radwaste", NP-5077, March 1987 has been used as a basis for the determination of a scaling factor. Sampling of individual waste streams was performed with analyses performed by an outside lab. The H-3 concentration was measured per gram of waste material. This value was compared to the Reactor Coolant System H-3 concentration. The scaling factor is derived from the ratio of the H-3 concentration in the waste stream to RCS H-3 concentration.

 -4   T        -1 Sampling    of the individual waste stream was performed with 'analysis by  off-site   lab to determine isotopic concentration. Ratios were developed between the scaled nuclide to the scaling nuclide concen-tration determined by analysis. In those cases where the scaling nuclide is not available in large enough quantities to develop reliable (viable) scaling factors, the recommendations made in Section 7 of the referenced EPRI report for the plant in the initial stages of operation are used.

T r- N-TRU nuclides would be scaled to Ce-144. As recommended by the AIF report "Methodolgies for Classification of Low 'Level Radioactive Haste from Nuclear Power Plants". These nuclides are not considered to be present if the scaled values are less than: 1 nCi/g for TRU, 35 nCi/g for Pu-241 or 200 nCi/g for Cf-242. TRU nuclides will be reported if the scaling nuclide (Ce-144) is reliably detected and Cs-137 is also present. Sampling of individual waste streams has been performed with analyses by an outside laboratory. Cs-137 and Sr-90 concentrations were measured in each waste stream except waste oil. The ratio of Cs-137 to Sr-90 has been determined and is used as the scaling factor for Sr-90 from Cs-137. For waste oil, the values from the referrenced EPRI Report will be used for scaling factors. Co-60 and Ni-63 con-centrations were measured in each of the'samp'led waste streams. The ratio of Co-60 to Ni-63 has been determined and is used as the scal-ing factor for Ni-63 from Co-60. 17

Table 4-1 lists scaling factors by waste stream for those nuclides that are required to be reported, Table 4-2 lists scaling factors for the conditional nuclides that are reported only when the scaling nuclide is found to be present. 4.2 Pr The Process Control Program (PCP) used to control solidification at HNP-2 will be provided by the vendor waste processor, Pacific Nuclear Inc. in accordance with Contract C-20452, and will be subjected to POC review prior to any solidification of radwaste. As an alter-native, approved High Integrity Containers (HIC's) could be used for the transport of wastes requiring stabilization. Other portions of the radwaste program are controlled by the WNP-2 procedures PPH 1.12.1, "Radwaste Management Program", PPM 1.12.2, "Radwaste Process Control Program", and l. 12.3, "Contract (Vendor) Haste Processing". There were no significant changes during the reporting period. 4.3 T Y I On duly 11, 1989 Science Applications International Company informed the Supply system of an analytical error on sample analysis used to develop scaling factors for Dry Active Haste. The first and second Semiannual Radioactive Effluent Reports for 1988 will be ammended to reflect the changes in scaling factor data and resulting radio-nuclide activities. The amendment for 1988 changes will be submit-ted under a separate letter. 18

SCALING FACTORS RWCU CFD EDR/FDR EDR/FDR POWDER POWDER POWDER BEAD gQIN ~I JU~ BULBS> H-3/Rx Coolant 3.5E-l 4.55E-1 4.55E-1 4.55E-l 3.56E-l 3. 10E-1 4.0E-5+ C-14/Co-60 4.78E-4 1.69E-5 6.2E-4++++ 1. 18E-3 5.81E-2 8.81E-5 1.3E-2+ Tc-99/Cs-137 4.6E-4+ 1.94E-6 9.3E-5+ 9.3E-5+ 9.3E-S+ 9.3E-5+ 4.ZE-5+ I-,.129/Cs-137 4.6E-4+ 2.23E-S 3.9E-S+ 3.9E-S+ 3.9E-5+ 3.9E-5+ 6.3E-5+ T b Ni-63/Co-60 6.05E-2 7. 73E-3 1.3E-2 4.53E-2 6.36E-2 1.5E-2+++ 1 ~ 2EO+ Fe-55/Co-60 8.52E-1 2.37E-1 5.16E-1 6.03E-1 1.9E-2 4. 10E-1 1.5EO+ Sr-90/Cs-137 2.6E-3+ 1 ~ 19E-4 3.88E-Z 2.92E-3 1. 11E-3 2.67E-5 3.3E-l+ Pu-239/Ce-144 4.5E-3+ 9. 1E-3+ 9.7E-3+ 9.7E-3+ 9.7E-3+ 8.7E-4+ 1.1E-2+ Pu-238/Pu-239 1 ~ 5EO+ 1. 3EO+ 1.7EO+ 1.7EO+ 1.7EO+ 1.7EO+ 1.6EO+ Pu-241/Pu-239 1.1E2+ B.BE1+ 9.6El+ 9.6El+ 9.6El+ 9. 1E 1+ 1.2E2+ Am-241/Pu-239 9. 1E-1+ 9. OE-1+ 6.6E-l+ 6.6E-1+ 6.6E-1+ 1.7EO+ 4.7E-1+ Cm-242/Pu-239 9. 5E-1+ 1.0EO+ 9.7E-l+ 9.7E-1+ 9.7E-1+ 5.7E-1+ 3. 1E-1+ Cm-244/Pu-239 7.2E-1+ 8 'E-1+ 7.6E-l+ 7.6E-1+ 7.6E-1+ 7.8E-1+ 2.9E-1+

  +   Scaling or scaled nuclide not present in enough concentration to make determination of scaling factor. In these cases the scaling factor obtained from the "Updated Scaling Factors in Low-Level Radwaste" EPRI HP-5077 Final Harch 1987.,

t+ The report from SAIC, showed the H-3 concentration in RWCU equal to Reactor Coolant concen-tration. The resin mix used in RWCU and CFD are the same. The reactor coolant and conden-sate H-3 concentration are approximately the same ~ The Scaling Factor for CFD is 4.55E-1 which is more representative of H-3 retention on the two resin streams. +++ The report from SAIC, showed the Ni-63 concentration of sludges at 4.03E-3 uCi/gm which compares to the Co-60 concentration of 3.52E-2 uCi/gm. This comparison would yield a Scaling Factor of 1.14E-l. The above mentioned EPRI Report recommends a Scaling Factor of 1.5E-2. Because of the long period of time between the generation of the waste and the counting of the sample (approximately 1 year) the EPRI Number is considered more accurate. ++w The report from SAIC showed the C-14 concentration in CFD of l. 12E-2 uCi/gm which compares to the Co-60 concentration of 1.08E-2 uCi/gm. This comparison would yield a Scaling Factor of 1.02EO. The above mentioned EPRI report reconmends a Scaling Factor of 6.2E-4. It is felt that there was cross contamination of the sample at the lab resulting in high concen-tration of C-14. The recommended EPRI number will be used. I

     /

0

Table 4-3 WNP-2 SOLID HASTE SHIPMENTS January - June 1989 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL

1. Type of Haste Unit Est. Total Haste Stream 6-month P ri F~~

a. Spent resins, filter sludges, m 69.79 v r 12

b. Dry active waste, contaminated 104.78

     ~rc. Irradiated    components,  control          m   No   Ship-

d. Other, (absorbed aqueous liquid) m No Ship-

2. Estimate of major nuclide composition (by type of waste):

a. Dewatered Spent Resins 1 E2 2 6 1

1. E

                                                              .1  El "Indicates scaled nuclide

b. Dry Active Wastes (DAW)

Nu ide 1 E-C - 1 l. E-2.2

                                                                       -2 2  1 H  -54                                          E-2

1. E-2

c. Irradiated Components None

d. Other Absorbed Liquids Hone

3. Solid Waste Disposition ILIA d f r D~i~n~in 14 Flat bed trailer (6) US Ecology 10-142 Cask (1) Richland, WA 14-170 Cask (7)

B. IRRADIATED FUEL SHIPMENTS (Disposition) Hone 'Indicates scaled nuclide 21

5. 0 HUlIEtLOOl The meteorological data for the first half of calendar year 1989 will be included in the Semiannual Effluent Report due 60 days after January 1, 1990 and will include data covering the full calendar year 1989.

6.0 D E E I T The dose impact on man for the calendar year 1989 will be included in the Semiannual Effluent Report due 60 days after January 1, 1990. 23

7 0 V T T 0 No amendments were made to the HNP-2 Offsite Dose Calculations Manual (ODCM) during this reporting period. rj ) 0

8.0 V N T T E No changes were made to the Process Control Program (PCP) during this reporting period which required POC approval.

9 0 KEMdunKLEIN~ TI NI I 9.1 Locations for dose calculations to be performed as identified by the 1988 Land Use Census (LUC) are:

9. 1 . 1 4. 8 miles southeast (SE) for the highest organ dose using ground, inhalation and vegetation pathways.

9.1.2 6.4 miles southeast (SE) for the highest organ dose using ground, inhalation, vegetables, cow milk and meat pathways. 9.1.3 4.2 miles east southeast (ESE) for the highest organ dose using ground, inhalation and vegetable pathways. 9.1.4 4.3 miles northeast (NE) for the highest organ dose using ground and inhalation pathways. 9.1.5 4,1 miles east northeast (ENE) for the highest organ dose using ground, inhalation and vegetable pathways. 9,2 A new direct radiation (TLD) station was established at a location 6.5 miles southeast (SE) . This sampling location was initiated as a public relations gesture and is in close proximity to the Pettett farm (6.4 miles SE) where milk, air and direct radiation (TLD) are routinely sampled or measured. 9.3 No sampling locations were deleted during this reporting period. 1 p

10.0 0 H N E T ADI TIVE LI ID SE D L D NA T T EATNENT

    ~SY TENS No  major changes were made to the radioactive waste systems (liquid, gaseous, or solid) during this reporting period.

h 4 U

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