Text

Rev 13 to Odcm
	ML20063H768
Person / Time
Site:	Farley
Issue date:	01/01/1994
From:	Hill R; SOUTHERN NUCLEAR OPERATING CO.
To:	;
Shared Package
ML20063H766	List: ML20063H763; ML20063H768;
References
	FNP--M-11, FNP-0-M-011, NUDOCS 9402220186
	Download: ML20063H768 (221)
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_. . _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ -

O O

t Offsite Dose Calculation Manual Revision 13 O

O 9402220186 DR 940211 ~,14 ADOCK 05000348 95 PDR "#

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-FNP-0-M-011

'ecember 12, 1993 Revision 13 4

SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT FNP-0-M-011 S

A.

F E

T Y.

OFFSITE DOSE CALCULATION MANUAL

.R E

L A

T E

D Approved:

N'uclear Plant '- General Manager Date. Issued: /"[" fM*

List of Effective Paoes Page Rev.

All 13.

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i FNP-0-M-011 DISTRIBUTION LIST b)

() For information pertaining to distribution of the ODCM, contact Farley Nuclear Plant Document Control.

V v

L Cen. pay, I3

-~~ - ~ - - _ _

FNP-O-M-011 TABLE OF CONTENTS p

V PAGE DISTRIBUTION LIST .

. . . . . . . . . . . . . . . . . . . . . . . . . . . i TABLE OF CONTENTS e

............................ 11 LIST OF TABLES

............................. v LIST OF FIGURES

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Vill CHAPTER 1: INTRODUCTION

. . . . . . . . . . . . . . . . . . . . . . . . 1-1 CHAPTER 2: LIQUID EFFLUENTS

. . . . . . . . . . . . . . . . . . . . . . 2-1 2.1 LIMITS OF OPERATION 2-1 2.1.1 Liouid Effluent Monitorina Instrumentation Control 2-1 2.1.2 Lieuid Effluent Concentration Control 2-7 2.1.3 Liouid Effluent Dose Centrol 2.1.4 2-11 Licuid Radwaste Treatment Svetem Control 2-13 2.1.5 Maior Chances to Liould Radioactive Waste Treatment Systems 2-14 2.2 LIQUID RADWASTE TREATMENT SYSTEM 2.3 LIQUID EFFLUENT MONITOR SETPOINTS 2-15 2.3.1 2-19 General Provisions Raoardino Setooints 2-19 2.3.2 Setnoints for Radwaste System Discharam Monitors 2.3.3 2-21 Setnoints for Monitors on Normally Low-Radioactivity Stre==- 2-29 2.4' LIQUID EFFLUENT DOSE CALCULATIONS 2.4.1 Calculation of Dose 2-30 2-30 2.4.2 Calculation of A 3 2-31 2.4.3 Calculation of CF;y 2-32 2.5 LIQUID EFFLUENT DOSE PROJECTIONS 2-42 2.5.1 Mirty-One Dav Dese Proiections 2-42 2.5.2 Dame Proiections for Soecifie Releases 2-42 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS 2-43 CHAPTER 3: GASEOUS EFFLUENTS 3.1

. . . . . . . . . . . . . . . . . . . . . 3-1 LIMITS OF OPERATION 3-1 3.1.1 cameous Effluent Monitorina Instr - ntation Control 3-1 3.1.2 Gaseous Effluent Dose Rate Control 3-6 3.1.3 caseous Effluent Air Dese Control 3-10 3.1.4 Control on caseous Effluent Dose to a " " =r of the Public 3-12 11 Gen. Rev. 13

l FNP-0-M-Ol1 TABLE OF CONTENTS (Continued)

E.b9.In 3.1.5 Caseous Radwaste Treatment System Control 3-14 6 3.1.6 MAJOR CHANCES to the GASEOUS RADIOACTIVE WASTE TREATMirrr SYSTEM and the VENTIIATION EXHAUST TREATMENT SYSTEM 3-16 GASEOUS RADWASTE TREATMENT SYSTEM 3 '

{ 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS 3-19 3.3.1 General Provisions Reaardina Noble Gas Monitor Setnoints 3-19 3.3.2 Setooint for the Final Noble Gas Monitor on Each Release Pathway 3-21 3.3.3 Setooints for Noble Gas Monitors on Effluent Source Str=--- 3-25 3.3.4 Determination of Allocation Factors. AG. 3-28 3.3.5 Setooints for Noble Gas Monitors with Snecial Raouir-~ nts 3~31 3.3.6 Setooints for Particulate and Iodine Monitors 3-31 3.4 GASEOUS IFFLUENT COMPLIANCE CALCULATIONS 3-32 3.4.1 Dese Rates at and Beyond the Site Boundarv 3-32.

3.4.2 Noble Gas Air Dome at or Revond Site Boundarv 3-33 3.4.3 Dose to a Mm=%r of the Public at or Bevcnd Site 11oundary 3-37 3.4.4 Dome Calculations to Sunnort Other Raouir- nts ~

,- 40 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS. 3-46 3.5.1 Thirty-One Day Dome Proiections 3-46 3.5.2 Dose Proiections for Snecific Releases 3-47 3.6 DFFINITIONS OF GASEOUS EFFLUENT TERMS 3 CHAPTER 4:

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM . . . . . . . 4-1' 4.1 LIMITS OF OPERATION 4-1 4.1.3 Radioloalcal Fnvironmental Monitorina 4 4.1.E tand Use canaus 4-8 4.1.3 Interlaboratory h arison Proaram 4-10 4.2 RADIOLOGICAL INVIRONMENTAL MONITORING LOCATIONS 4-11 CHAPTER 5: TOTAL DOSI DETERMINATIONS . . . . . . . . . . . . . . . . . 5-1 5.1 LIMIT OF OPERATION 5-1 5.1.1 Anglicability 5-1 5.1.2 Actions 5-1 5.1.3 eamv.111.. a .iir-- nta 5-2 5.1.4 33313 5-2 5.2 DEMONSTRATION OF COMPLIANCE 5-3 CHAPTER 6:

POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY . . . . . . . . . . . . 6-1 6.1 REQUIREMENT FOR CALCULATION 6-1 111 Gen. Rev. 13 '

FNP-O-M-Oll TABLE OF CONTENTS (Continued)_

FAGE

, 6.2 CALCULATIONAL METHOD 6-1 CHAPTER 7: REPORTS . . . . . .

7.1

.................... 7-1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 7-1 '

7.1.1 Reauirement for Report 7-1 7.1.2 Reoort Contents 7-1 7.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 7-3 7.2.1 Raouir- nt for Report 7-3 7.2.2 Recort contents 7-3 7.3 MONTHLY OPERATING REPORT 7-7 7.4 SPECIAL REPORTS 7-7 CHAPTER 8: METEOROLOGICAL .'40DELS 8.1

................... 8-1 ATMOSPHERIC DISPEP.SION 8-1 8.1.1 Ground-Level Releases 8-1 8.1.2 Elevated Releases 8-3 8.1.3 Mixed-Mode Releases 8-5 8.2 RELATIVE DEPOSITION 8-7 8.2.1 Ground-Level Releases 8-7 8.2.2 Elevated Releases 8-7 8.2.3 Mixed-Mode Releases j 8.3 ELEVATED PLUME DOSE FACTORS 8-8 8-8 8.4 METEOROLOGICAL

SUMMARY

8-8 CHAPTER 9:

METHODS AND PARAMETERS FOR CALCULATION OF GASEOUS EFFLUENT PATHWAY DOSE FACTORS, Pg . . . . . . . . . . . . . . . . . . C-1 9.1 INHALATION PATHWAY FACTOR 9.2 9-1 GROUND PLANE PATHWAY FACTOR 9.3 9-2 GARDEN VEGETATION PATHWAY FACTOR 9.4 9-3 GRASS-COW-MILK PATEWAY FACTOR 9.5 9-6 GRASS-OQkT-MILE PATHWAY FACTOR 9.6 9-10 GRASS MEAT FATRWAY FACTOR 9-14 CHAPTER 10:

DhFINITIONS OF EFFLUENT CONTROL TERMS' . . . . . . . . . . . 10-1 10.1 TERMS SPECIFIC TO THE ODCM 10-1 10.2 TERMS DEFINED IN THE TECHNICAL SPECIFICATIONS 10-5 ,

O iv Gen. Rev. 13

FMP-0-M-011 LIST OF TABLES

\j Table 2-1. PAGE Radioactive Liquid Effluent Monitoring Instrumentation 2-3 Table 2-2. Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 2-5 Table 2-3. Radioactive Liquid Waste Sampling and Analysie Program 2-9 Table 2-4. Applicability of Liquid Monitor Satpoint Methodologies Table 2-5.

2-20 Parameters for Calculation of Doses Due to Liquid Effluent Releases Table 2-6.

2-35 Element Transfer Factors 2-36 Table 2-7. Adult Ingestion Dose Factors 2-37 Table 2-8.

Site-Related Ingestion Dose Factors, Aj7 2-40 Table 3-1.

Radioactive Caseous Effluent Monitoring Instrumentation 3-3 Table 3-2.

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

3-5 Radioactive Gaseous Waste Sampling and Analysis Program 3-8 Table 3-4.

Applicability of caseous Monitor Setpoint Methodologies 3-20 Table 3-5.

Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases Table 3-6. 3-35 Dose Factors for Exposure to Direct Radiatio. from Noble cases in an Elevated Finite Plume 3-36 Table 3-7.

Attributes of-the Controlling Receptor 3-39 Table 3-8. R aipj f or Ground Plane Pathway, All Age Groups 3-42 Table 3-9. R aipj for Inhalation Pathway, Child Age Group 3-43 Table 3-10. Rapj j for Cow Meat Pathway, Child Age Group 3-44 Table 3-11. R aipj for Garden Vegetation Pathway, Child Age Group 3-45 Table 4-1.

Radiological Environmental Monitoring Program 4-4 Table 4-2.

Reporting Levels for Radioactivity Concentrations in Environmental Samples 4-6 Table 4-3.

Values for the Minimum Detectable concentration 4-7 Table 4-4.

Radiological Environmental Monitoring Locations 4-12 Table 6-1.

Attributes of Member of the Public Receptor Locations Inside the SITE BOUNDARY Table 8-1. 6-3 Terrain Elevation Above Plant Site Grade 8-9 Table 8-2. Annual Average (176) for Mixed Mode Releases Tabis 8-3. 8-10 Annual Average (176) for Ground-Level Releases 8-11 Table 8-4. Annual Average (67b) for Mixed Mode Releases Table 8-5. 8-12 Annual Average (676) for Ground-Level Releases 8-13 Table 9-1. Miscellaneous Parameters for the Garden Vegetation "

Pathway 9-5 Table 9-2.

Miscellaneous Parameters for the Grass-Cow-Milk Pathway 9-9 s

v Gen. Rev. 13

FNP-0-M-011 LIST OF TABLES (Continued)

PAGE Table 9-3.

Miscellaneous Parameters for the Grass-Goat-Milk Pathway 9-13 Table 9-4.

Miscellaneous Parameters for the Grass-Cow-Meat Pathway 9-17 Table 9-5. Individual Usage Factors Table 9-6. 9-18 Stable Element Transfer Data 9-19 '

Table 9-7.

Inhalation Dose' Factors for the Infant Age Group- 9-20 Table 9-8.

Inhalation Dose Factors for the Child Age Group 9-23 Table 9-9.

Inhalation Dose Factors for the Teenager Age Group 9-26 Table 9-10. Inhalation Dose Factors for the Adult Age Group 9-29 Table 9-11. Ingestion Dose Factors for the Infant Age Group 9-32 Table 9-12. Ingestion Dose Factors for the Child Age Group 9-35 Table 9-13. Ingestion Dose Factors for the Teenager Age Group 9-38 Table 9-14. Ingestion Dose Factors for the Adult Age Group Table 9-15. 9-41 External Dose Factors for Standing on Contaminated Ground 9-44 O

49

I vi Gen. Rev. 13

FNP-0-M-011 LIST OF FIGURES Figure 2-1. PA9I Liquid Radwaste Treatment System (Typical of Both Units) 2-16 Figure 2-2.

Steam Generator Blowdown System (Typical of Both Units). 2-17 Figure 2-3. Liquid Discharge Pathways Figure 3-1.

2-18 Schematic Diagram of Routine Release Sources and Release '

Points (Typical of Both Units) 3-18 Figure 4-1. Airborne Sampling Locations, 0-5000 feet Figure 4-2. 4-15 Indicator II (community) Sampling Locations for Direct Radiation Figure 4-3.

4-16 Airborne Sampling Locations, 0-20 miles 4-17 Figure 4-4. Water Sampling Locations 4-18 Figure 8-1.

Vertical Standard Deviation of Material in a Plume (ag) 8 .',4 Figure 8-2.

Terrain Recirculation Factor (K r) 8-15 Figure 8-3.

Plume Depletion Effect for Ground Level Releases 8-16 Figure 8-4. Plume Depletion Effect for 30-Meter Releases Figure 8-5.

8-17 Plume Depletion Effect for 60-Meter Releases 8-18' Figure 8-6. Plume Depletion Effect for 100-Meter Releases Figure 8-7.

8-19 Relative Deposition for Ground-Level Releases 8-20 Figure 8-8. Relative Deposition for 30-Meter Releases Figure 8-9. 8-21 Relative Deposition for 60-Meter Releases 8-22 Figure 8-10.

Relative Deposition for 100-Meter.(or Greater) Releases 8-23

( Figure 10-1. Site Map for Effluent controls 10-8 vii Gen. Rev. 13

FNP-O-M-011 REFERENCES

1. J.S. Boogli, R.R.

Bellamy, W.L. Britz, and R.L. Waterfield, " Preparation or Radiological Effluent Technical 1978. Specifications for Nuclear Power

[ Plents," NURIG-1033, October 2.

"Calculstion of Annual Doses to Man from Routine Releases of Reactor Effluents I,"

Appendix for the Purpose of Evaluating Compliance with 10 U.S. NRC Reculatorv Guide 1.109, March 1976.

CFR 50,

3. <

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents I,"

Appendix for the Purpose of Evaluating Compliance with 10 CFR 50, U.S.

NRC Reculatorv Guide 1.109. Revision 1, October 1977.

4.

" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents Raoulatorv Guide 1.111, March 1976.

in Routine Releases from Light-Water-Cooled Reactors," U.S. NRC 5.

" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous

- Effluents in Routine Releases from Light-Water-cooled Reactors," U.S. NRC.

Raoulatory Guide 1.111, Revision 1, July 1977..

6.

" Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," U.S. NRC' Raoulatorv Guide 1.113, April 1977.

7. Jonach M. Farlev Nuclear Plant Units 1 and 2 Final Safety Analysis Renort, Alabama Power Company.

8. Jemanh M.

Parlev Nuclear Plant Units 1 and 2 Environmental Renort -

Onoratino License stace, Alabama Power Company.

9. T.E. Young, T.S. Bohn, and W. Serrano, " Technical Evaluation Report for the Evaluation Units 1 and 2,"ofEGC-PHY.8674, ODCM Revision 7 for Joseph M. Farley Nuclear Plant, letter dated November 9, 1989. dated August 1989, transmitted by NRC 10.

W.M. Jackson, " Survey Report of Chattahoochee River Water Use Downstream of Farley Nuclear Plant Liquid Effluent Discharge," dated July 19, 1990.

11.

J.E. Till and H.R.

NUREG/CR-3332, Meyer, 1983. eds., Radiolooleal Assessment, U.S. NRC Report 12.

L.A. Currie, Lower Limit of Detection: Definition and Elaboration of a Pronomad Position of Radiolooical Effluent and Environs _ntal Measurements, U.S. NRC Report NUREG/CR-4007, 1984.

13.

" Radiological Assessment Branch Technical Position", U.S. Nuclear Regulatory Cosuaission, Revision 1, November 1979.

14. U.S. DOR Report FNL-5484.

15.

D.C.

11026, KocherI 1981. " Radioactive Decay Data Tables," U.S. DOE Report DOE / TIC-

-1

16. Internal -

1randu=-

Company, June 4, 1990.J.E. Carlinoton to D.N. Morev, Alabama Power viii Gen. Rev. 13

I FNP-0-M-011 CHAPTER 1 INTRODUCTION O The Offsite Dose Calculation Manual is a supporting document of the Technical Specifications. As such, it describes the methodology and parameters to be used '

in the calculation of offsite doses due to radioactive liquid and gaseous '

effluents, and in the calculation of liquid and gaseous effluent monitoring instrumentation alarm setpoints. In addition, it contains the following:

e

) The controle required by the Technical Specifications, governing the

} radioactive effluent and radiological environmental monitoring programs.

e Schematics of liquid and gaseous radwasta effluent treatment systems, which include designation of release points to UNRESTRICTED AREAS.

s 4

e A list and maps indicating the specific sample locations for the Radio-logical Environmental Monitoring Program.

1 i i

Specifications and descriptions of the information that must be included ,

in the Annual Radiological Environmental Surveillance Report and the i

Annual Radioactive Effluent Release Report ' required by the Technical Specifications. '

The ODCM will be maintained at the plant for use as a reference guide and training document of accepted methodologies and calculations. Changes in the calculational methods or parameters will be incorporated into the ODCM in order to ensure that it represents current methodology in all applicable areas. Any computer software used to perform the calculations described will be maintained current with the ODCM.

Equations and methods used in the ODCM are based on those presented in NUREG-0133 (Reference 1), in Regulatory Guide 1.109 (References 2 and 3), in Regulatory Guide 1.111 (Referencee 4 and 5), and in Regulatory Guide 1.113 (Reference 6).

, , [,4 -

O l-1 Gen. Rev. 13

FNP-0-M-011 CHAPTER 2 I LIOUID EFFLUENTS I O 2.1 LIMITS OF OPERATION The following Liquid Effluent Controls implement requirements established by Technical Specifications section 6.0.

Terms printed in all capital letters are defined in Chapter 10.

2.1.1 Liould Effluent Monitorino Instrumentation control In accordance with Technical Specification 6.8.3.e(i), the radioactive liquid ef fluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits specified in ,

Section 2.1.2 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with Section 2.3. '

l 2.1.1.1 Applicability 1

This limit applies at all times. l 2.1.1.2 Actions With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above control, immediately ;

suspend the release of radioactive liquid affluents monitored by the affected '

channel, declare the channel inoperable, or change the setpoint to a conservative value.

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

This control does not affect shutdown requirements or MODE changes.

2.1.1.3 Surveillance Requirements Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 2-2. ,

1 I

l l

( 2-1 Gen. Rev. 13 ,

FNP-0-M-011 2.1.1.4 Bnaio

/~"N The radioactive liquid effluent instrumentation is provided to monitor and t

control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. l The Alarm / Trip j Setpoints for these instruments chall be calculated and adjusted in accordance {

with the methodology and parameters in Section 2.3 to ensure that the alarm / trip '

will occur prior to exceeding the limits of Section 2.1.2. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

/T

\ ,) \

ll 1 2-2 Gen. Rev. 13

_ _ _ _ _ _ _ _ _ ____m

I FNP-O-M-011 Tablo 2-1.

Radioactivo Liquid Efflusnt Monitoring Instrumentation I

(\

\_,/

I l

l OPERABILITY Requirements a ,

l Instrument Minimum Channels Operable ACTION '

1.

Gross Radioactivity Monitors Providing Automatic Termination of Release i I

a.

Liquid Radwaste Ef fluent Line (RE-18) 1 28

b. {

Steam Generator Blowdown Effluent Line (RE-233) 1 '29 2.

Flowrate Measurement Devices

a. Liquid Radwaste Effluent Line

1) Waste Monitor Tank No. I 1 30

)

2) Waste Monitor Tank No. 2 1 30

b. Discharge Canal Dilution Line (Service Water) 1 30 c.

Line Steam Generator Blowdown Effluent 1 30 a.

All requirements in this table apply to each unit.

OP

(/

\s 2-3 Gen. Rev. 13

1 h

a FNP-0-M-011 j Tablo 2-1 (contd). Notation for Table 2 ACTION Statements i

j f

ACTION 28 - With the number of channels OPERABLE less than required by the 1 Minimam Channels OPERABLE requirement, effluent releases may continue for up to 14 days provided that prior to initiating a releasen j

a. At least two independent samples are analyzed in accordance with Section 2.1.2.3, and '

i i b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving and (1) Verify the manual portion of the computer input for the i

release rate calculations performed on the computer, or l

(2) Verify the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive' effluents via this Pathway.

1

ACTION 29 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for i

gross radioactivity (beya or gamma) at a MINIMUM DETECTABLE CONCENTRATION no greater than 1 x 10- yCi/mL:

a. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the specific activity of the secondary coolant is greater than 0.01 pC1/ gram DOSE EQUIVALENT I-131.

b.

i At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of . the secondary coolant is less than or equal to 0.01 pCi/ gram DOSE ;

EQUIVALENT I-131.

ACTION 30 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE r6quirement, effluent releases via this pathway may continue for up to 30 days provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during re ctual releases.

i estimate flow.

Pump curves may be used to i

\

2-4 Gen. Rev. 13

FNP-0-M-011 Teble 2-2.

Rcdioactive Liquid Ef fluent Monitoring Instrumentation Surveillance Requirements O

Surveillance Requiremented Instrument CHANNEL CHANNEL CHANNEL FUNCTIONAL CHECK SOURCE CHECK CALIBRATION TEST

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Liquid Radwaste Effluent Line (RE-18) D P Rb Qa

b. Steam Generator Blowdown Effluont Line (RE-23B) D M Rb Qa

2. Flowrate Measurement Devices

a. Liquid Radwaste Effluent Line

1) Waste Monitor Tank No. 1 DC NA R NA

2) Waste Monitor Tank No. 2 Dc NA R NA

b. Discharge Canal Dilution Line (Service Water) DC NA R Q

c. Steam Generator Blowdown Effluent Line DC NA R NA x

O 2-5 Gen. Rev. 13

FNP-0-M-011 Tabla 2-2 (contd). Notation for Table 2-2 i l

i O a.

In addition to the basic functions of a CHANNEL FUNCTIONAL TEST (Section' 10.2):

i fi (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists: ,

(a) Instrument indicates measured levels above the alarm / trip i j

setpoint; (b) Loss of control power; or (c) Instrument controls loss of instrument power.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that. control room alarm annunciation occurs if any of the following conditions exists (a) Instrument indicates a downscale failure; or (b) Instrument controls not set in operate mode.

b.

The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards I

{

and Technology or using standards that have been obtained from suppliers that participate in measurements assurance activities with NIST. For !

subsequent CHANNEL CALIBRATION, sources that have been related to the-initial calibration shall be used.

0, c.

CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on l

)

i days on which continuous, periodic, or batch releases are made. 1 I

d. All requirements in this table apply to each unit.

J 2-6 Gen. Rev. 13

i FNP-0-M-011 2.1.2 Licuid Effluent concentration control

\ In accordance with Technical Specifications 6.8.3.e(ii) and 6.8.3.e(iii), the concentration of radioactive material released in liquid effluents to ;1

' UNRESTRICTED AREAS (see Figure 10-1) shall be limited at all times to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or l entrained noble gases, the concentration shall be limited to 1 x 10-4 pCi/mL total activity.

8 l 2.1.2.1 Applicability 1 l

I \

This limit applies at all times.

2.1.2.2 Actions

+

with the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the limits stated in Section 2.1.2, immediately restore the concentration to within the stated limits.

This control does not affect shutdown requirements or MODE changes.

2.1.2.3 Surveillance Requirements The radioactivity content of each batch of radioactive liquid waste shall be

{ ,

determined by sampling and analysis in accordance with Table 2-3. The results

) of radioactive analyses shall be used with the calculational methods in Section 3

2.3 to assure that the concentration at the point of release is maintained within

1 the limits of Section 2.1.2.

4

2.1.2.4 Basis I

This control is provided to ensure that the concentration of radioactive I j

materials released in liquid waste affluents to UNRESTRICTED AREAS will be less than ten tisse ther concentration levels specified in 10 CFR 20, Appendix B, Table !

1 2, Column 2.

Thie limitation provides additional assurance that the levels of radioactive mistorials in bodies of water in UNRESTRICTED ARIAS will result in exposures within (1) the Section II. A design objectives of Appendix I,10 CFR 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.1301 to the population.

The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC j

in air (submersion) was converted to an equivalent concentration in water using.

the methods described in International Commission on Radiological Protection

\

2-7 Gen. Rev. 13 l

1; l

FNP-0-M-011

)

(ICRP) Publication 2 (1959). The resulting concentration of 2 x 10~4 was then multiplied by the ratio of the effluent concentration limit for xe-135, stated in Appendix B, Table 2, Column 1 of 10 CFR 20 (paragraphs 20 1001 to 20.2401),

to the MPC for Xe-135, stated in Appendix B, Table II, Column 1 of 10 CFR 20 (paragraphs 20.1 to 20.601), to obtain the limiting concentration of 1 x 10'4 j pCL/mL.

I l

l l l

l l

O i I

i

)

l l l

l O 2-8 Gen. Rev. 13

4 FNP-0-M-011 Tcblo 2-3. Radioactive Liquid Waste Sampling and Analysis Program i

T 4

k Sampling and Analysis Requirementsa,b t

l MINIMUM i

Liquid DETECTABLE Minimum CONCENTRATION l Release Sampling Analysis Type of Activity Type (MDC)

FREQUENCY FREQUENCY Analysis (pC1/mL) 3 A. Waste Tanks Producing BATCH RELEASES

~

PRINCIPAL CAMMA 5 E-7 P P EMITTERS Each BATCH Each BATCH I-131 1 E-6 Dissolved and 1 E-5

One BATCH /M All (Cama Emitters) p g H-3 1 E-5 i

Each BATCH COMPOSITE Gross Alpha 1 E-7 I

5 p g Sr-89, Sr-90 5 E-8 Each BATCH COMPOSITE Fe-55 1 E-6 B. CONTINUOUS RELEASESC

~ ,

4 PRINCIPAL GAMMA 5 E-7

' D W EMITTERS l Grab Sample COMPOSITE I-131 1 E-6 M Dissolved and 1 E-5 Steam Grab Sample M Entrained Gases Generator (Cama Emitters)

Blowdown ,

D M H-3 1 E-5 '

Grab sample COMPOSITE i

Gross Alpha 1 E-7 D Sr-89, Sr-90 5 E-8 Q i Grab sample COMPOSITE Fe-55 1 E-6 j

hrbine s PRINCIPAL GAMMA 5 E-7 Building- P W EMIMERS 3 ump Orab sample COMPOSITE H-3 1 E-5 O 2-9 Gen. Rev. 13

FNP-0-M-011 Tablo 2-3 (contd). Notation for Table 2-3

a. All requirements in this table apply to each unit. Deviation from the MDC irements of this table shall be reported in accordance with Section b.

Terms printed in all capital letters are defined in Chapter 10.

c.

Sampling will be performed only if the ef fluent will be discharged to the environment.

I O

. r.

+

0 2-10 Gen. Rev. 13 I

____-___-__----- ------------ A

4 FNP-0-M-011 f

l 2.1.3 Licuid Ef fluent Dese Control

\

In accordance with Technical Specifications 6.8.3.e(iv) and 6.8.3.e(v), the dose l

)

or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid I ef fluents released, from each unit, to UNRESTRICTED AREAS (see Figure 10-1) shall be limited:

a a.

During any calendar quarter to less than or equal to 1.5 mrom to the total body and to less than or equal to 5 mrom to any organ, and b.

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

2.1.3.1 Applicability I

These limits apply at all times.

l 2.1.3.2 Actions With the calculated dose from the release of radioactive materials in liquid i

ef fluents exceeding any of the limits of Section 2.1.3, prepare and submit to the Nuclear Regulatory Comnission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s);

p defines the corrective actions to be taken to reduce the releases; and defines l(

i the proposed corrective actions to be taken to assure that subsequent releases j

will be in compliance with the limits of Section 2.1.3.

This control does not affect shutdown requirements or MODE changes.

2.1.3.3 Surveillance Requirements At least once per 31 days, cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined, for each unit, in accordance with Section 2.4.

l 2.1.3.4 Basis This control is provided to implement the requirements of Sections II. A, III.A and IV. A of Appendix I, 10 CFR Part 50.

i The limits stated in Section 2.1.3 implement the guides set forth in Section II. A of Appendix I. The ACTIONS stated in Section 2.1.3.2 provide the required operating flexibility and at the same l

time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as 2-11 Gen. Rev. 13

FNP-O-M-Oli is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is I reasonable assurance that the operation of the facility will not result in

%g radionuclide concentrations in the finished drinking water that are in excess of 4 the requirements of 40 CFR Part 141.

The dose calculations in Section 2.4 implement the requirements in Section III.A of Appendix I, which state that

, l conformance with the guides of Appendix I be shown by calculational procedures {

based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in Section 2.4 for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3) and Regulatory Guide 1.113 (Reference 6).

i 7,

This control applies to the release of liquid effluents from e&ch unit at the site.

The liquid ef fluents from shared LIQUID RADWASTE TREATMENT SYSTEMS are to

! be proportioned between the units. !

l

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2-12 Gen. Rev. 13 4

, , ~ ,-n ----

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FNP-0-M-011 {

2.1.4 Licuid Radwaste Treatment System control In accordance with Technical Specification 6.8.3.e(vi), the LIQUID RADWASTE

( TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the system shall be used to reduce radioactivity in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Figure 10-1) would exceed 0.06 mrom to the total body or 0.2 mrom to r J

any organ of a MEMBER OF THE PUBLIC in 31 days.

i

2.1.4.1 Applicability This limit applies at all times. !

I

/ 2.1.4.2 Actions 1 ;

With radioactive liquid waste being discharged without treatment and in excess of the above limits and appropriate portions of the LIQUID RADWASTE TREATMENT 1

SYSTEM not in operation, prepare and submit to the Nuclear Regulatory Commission within 30 days pursuant to Technical Specification 6.9.2 a special Report which incluGee tr.& fs12owing information:

\

< I a.

Explanation of why liquid radwaste was being discharged without treatment, i

identification of any inoperable equipment or subsystems, and the reason- j i('- for the inoperability, 1

i b.

Action (s) taken to restore the inoperable equipment to OPERABLE status, !

j and 4

1 c.

} Summary description of action (s) taken to prevent a recurrence.

J j

i This control does not affect shutdown requirements or MODE changes.

i 2.1.4.3 Surveillance Requirements I i i j

Doses due to liquid releases to UNRESTRICTED AREAS shall be projected at least 1

once per 31 days, in accordance with Section 2.5, during periods in which the !

LIQ 0ID RADWASTE j fpBATMENT SYSTEMS are not being fully utilized.

I 4

The LIQUID RADWASTE TREATMENT SYSTEM shall be demonstrated OPERABLE:

a. by meeting the controle of Sections 2.1.2 and 2.1.3, or i

2-13 Gen. Rev. 13 m - -s ,. , , . , - . + , . , a .w.,- , . . . , . e .y v m.,w y , ,

. _ . _ _ . . . . m ._ __ _ .- _ _ _ _ . . _ _ _ _ _ _ _ _ _ . _ _ __

h

n 4

FNP-0-M-011 i- b.

by operating thS LIQUID:RADWASTE TREATMENT SYSTEM cquipmen't for-at'least-l 15 minutes at least once per 92 days unless the LIQUID RADWASTE TREATMENT SYSTEM equipment has been utilized to process radioactivw liquid ef fluents during the previous 92 days.

2.1.4.4 Basis e-The OPERABILITY of the LIQUID RADWASTE TREATMENT SYSTEM ensures that this system ;

will be available for use whenever liquid effluents require treatment prior to release to the UNRESTRICTED AREAS. The requirement that the appropriate portions of this' system.be used when specified provides assurance that the-releases of radioactive materials in liquid effluents will be kept'"as low as is reasonably.

achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design ' Criterion 60 of Appendix A to 10 CFR - Part 50, and the design:

objective given in Section II.D of Appendix I to 10 CFR Part 50. The.specified limits governing the use of appropriate portions of the LIQUID RADWASTE TREATMENT SYSTEM were specified as a suitable fraction of the does de0ign objectives set' forth in Section.II.A of Appendix I, 10 CFR Part 50, for liquJd efflaents.

2.1.5 Maier chanans to Lieuld Radioactive Waste Treatment Syst--

Licensee initiated MAJOR CHANGES TO LIQUID RADIOACTIVE WASTE TREATMENT SYSTEMS:

a.

Shall be reported to the Nuclear Regulatory Commission' in the Annual-Radioactive Effluents Release Report for the period'in which the change was implemented, in accordance.with.Section 7.2.2.7.

b.

Shall become effective upon review and approval: in ' accordance with Technical Specification 6.5.3.1.

1 2-14 Gen..Rev. 13 4

(

l 1

i FNP-0-M-011 2.2 LIQUID RADWASTE TREATMENT SYSTEM

' The Farley Nuclear Plant is located on the west bank of the Chattahoochee River approximately 35 river a.ilse above the point where it empties into Lake Seminole.

There are two pressurized water reactors on the site. Each unit is served by a completely separate LIQUID RADWASTE TREATMENT SYSTEM that 'Js illustrated schematically in Figure 2-1. However, both units share a common domineralizar '

bed system for processing liquids prior to release from the site. As shown in Figure 2-2, the Steam Generator Blowdown System is a separate entity. Liquid discharge pathways are shown in Figure 2-3.

t 3

All liquid radwastes treated by the LIQUID RADWASTE TREATMENT SYSTEM ' are i

collected in 5,000-gallon Waste Monitor Tanks for sampling and analysis prior to {

release. Prior.to sampling, each waste monitor tank is recirculated for a minimum of two tank content volumes, to ensure that a representative sample can be taken from the tank. Releases from the waste monitor tanks are routed to the l Service Water discharge line (which provides dilution prior to release to the i UNRESTRICTED AREA), and thence to the Chattahoochee River. The Service Water discharge line also receives input from the Cooling Tower Blowdown and the Turbine Building Sump.

Although no significant quantitles of radioactivity are, expected in the steam generator blowdown processing system, this effluent pathway is monitored as a precautionary measure. The monitors serving this pathway provide for automatic t A

termination of release in the event that radioactivity is detected above predetermined levels.

Like the LIQUID RADWASTE TREATMENT SYSTEMS, the Steam '

Generator Blowdown Systems discharge to the Service Water discharge line. I one potential release pathway, the Turbine Building Sump discharge, is not monitored during release, but is sampled regularly during' discharges. Sampling and analysis of releases via this pathway must be sufficient to ensure that the liquid effluent dose limits specified in Section 2.1.3 are not exceeded.

l I

l 2-15 Gen. Rev. 13

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-l FNP-0-M-011 i

e Unit 1 Unit 2 Service Water Service Water Train A Train 3 Train A Train 5 4 Cooling Tower Slowdown 7

Radvaste Discharse 11Z018 2EE018 Steam Generator Slowdown 13Z233 O, 21E235 I Turbine Suilding Sump 7 V

Te River l

Figure 2-3. Liquid Discharge Pathways O

2-18 Gen. Rev. 13

FNP-0-M-011 2.3 LIQUID EFFLUENT MONITOR SETPOINTS 2.3.1 General Provisions Recardino Setooints O Liquid monitor setpoints calculated in accordance with the methodology presented in this section will be regarded as upper bounds for the actual high alarm setpoints. That is, a lower value for the high alarm setpoint may be established #

or retained on the monitor, if desired. Intermediate level setpoints should be.

established at an appropriate level to give sufficient warning prior to reaching the high alarm setpoint. If no release is planned for a particular pathway, or if there is no detectable activity in the planned release, the monitor setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should an inadvertent release occur.

Two basic setpoint methodologies are presented below. For radwaste system ;

discharge monitors, setpoints are determined to assure that the limits of Section 2.1.2 are not exceeded. For monitors on streams that are not expected to contain significant radioactivity, the purpose of the monitor setpoints-is to cause an alarm on low levels of radioactivity, and to terminate the release where this is possible.

Section 2.1.1 establishes the requirements fos liquid effluent ,

monitoring instrumentation. Table 2-4 lists the monitors for whiah each of the setpoint methodologies is applicable.

4 I

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2-19 Gen. Rev. 13 L'

1 FNP-0-M-011 Tablo 2-4.

Applictbility of Liquid Monitor Setpoint Methodologies {

CJ Setpoint Method Section 2.3.2 Unit 1 or Unit 2 Waste Monitor Tanks Effluent Release Type: BATCH Monitor: IRE-Ol8 / 2RE-Ol8 ,

Unit 1 or Unit 2 Steam Generator Blowdown Effl g Release Type: CONTINUOUS Monitors 1RE-023 8 / 2RE-023 B Normally Low-Radioactivity streams with Termination or Diversion upon Alarm Farley Nuclear Plant has no liquid effluent streams in this category.

Normally Low-Radioactivity Streams with Alarm only Farley Nuclear Plant has no liquid effluent streams in this category.

i 1

l l

I O 2-20 Gen. Rev. 13 l

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I

^

FNP-0-M-011 2.3.2 Setroints for Radwaste system Discharoe Monitors 2.3.2.1 Overview of Method LIQUID RADWASTE TREATMENT SYSTEM effluent line radioactivity monitors are intended to provide alarm and automatic termination of release prior to exceeding the limits specified in Section 2.1.2 at the point of release of the diluted <

effluent into the UNRESTRICTED AREA. Therefore, their alarm / trip setpoints are established to ensure compliance with the following equation (equation adapted j from Addendum to Reference 1):

C*I s TE CECL (2.1)

T+f wheres C ECL =

the Effluent Concentration Limit corresponding to the mix of radionuclides in the effluent being considered for discharge, in yCi/mL.

e=

the setpoint, in yCi/mL, of the radioactivity monitor measuring the concentration of radioactivity in the effluent line prior to dilution and subsequent release. The setpoint represents a concen-tration which, if exceeded, could result in concentrations exceeding the limits of Section 2.1.2 in the UNRESTRICTED AREA.

f=

the effluent flowrate at the location of the radioactivity monitor, in gpm.

F= ,

the dilution stream flowrate which can be assured prior to the !

release point to the UNRESTRICTED AREA, in gym. A predetermined dilution flowrate must be assured for use in the calculation of the radioactivity monitor setpoint. !

TF =

the tolerance factor selected to allow flexibility in the establishment of a practical monitor setpoint which could accommodate affluent releason at concentrations higher than the ECL

{

values stated in 10 CFR 20, Appendix B, Table 2, Column 2; the '

tolerance factor must not exceed a value of 10.

While equation (2.1) shows the relationships of the critical parameters that determine the setpoint, it cannot be applied practically to a mixture of radio-2-21 Gen. Rev. 13 l

I i FNP-0-M-Oll nuclidas with different Effluent Concentration Limits (ECLs). For a mixture of radionuclides, equation (2.1) is satisfied in a practicable manner based on the calculated ECL fraction of the radionuclide mixture and the dilution stream flowrate that can be assured for the duration of the release (Fd ), by calculating the maximum permissible affluent flowrate (fm) and the radioactivity monitor setpoint (c).

i The setpoint method presented below is applicable to the release of only one tank of liquid radwaste per reactor unit at a given time. Liquid releases must be controlled administratively to ensure that this condition is met; otherwise, the setpoint method may not ensure that the limits of Section 2.1.2 are not exceeded, i !

1 2.3.2.2 Setpoint Calculation Steps I i Stoo 1:

1 Determine the radionuclide concentrations in the liquid waste being a

considered for release in accordance with the sampling and analysis requirements of Section 2.1.2.

To ensure that sample analyses are based on samples that are representative of the waste being sampled, the liquid volume must be mixed thoroughly prior to sampling. Mixing may be accomplished by any method that has been demonstrated to achieve sufficient mixing for representative sampling. The Waste Monitor Tanks are recirculated for a minimum of two tank content volumes prior to sampling.

The Service Water discharge line is assumed to be well mixed, so that no additional mixing is required prior to sampling.

4 The total concentration of the liquid waste is determined by the results of all required analyses on the collected sample, as follows:

l [Ci " Ca+[Cs+Cf+Ct + ECg i s g (2.2) 1

! where:

4 C, =

the gross concentration of alpha emitters in the liquid waste, not

. lese than that measured in the most recent applicable composite I sample.

C, =

the concentration of strontium radioisotope s (Sr-89 or Sr-90) in the liquid waste, not less than that measured in the most recent applicable composite sample.

O i b 2-22 Gen. Rev. 13 s

'i FNP-0-M-011 Cg = the concentration of Fe-55 in the liquid waste, not less than that measured in the most recent applicable composite sample.

Ct= the concentration of H-3 in the liquid waste, not less than that measured in the most recent applicable composite sample.

C g= the concentration of gamma emitter g in the 11guld waste as

]

measured by gamma ray spectroscopy performed on the sample for the release under consideration.

The Cg term will be included in the analysis of each waste sample; terms for gross concentrations of alpha emitters, Sr-89, Sr-90, Fe-55, and tritium will be 1 included in accordance with the sampling and analysis program required for the waste ,ctrisam (see Section 2.1.2). For eEch analysis, only radionuclides identified and detected above background for the given measurement should be included in the calculation. When using the alternate setpoint methodology of step 5.b, the historical maximum values of C,, C,, cf, and Cg shall be used.

sten 2:

Determine the required dilution factor for the mix of radionuclides detected in the waste.

Measured radionuclide concentrations are used .o calculate ECL fractions. The O ECL fractions are used along with a safety itetor to calculate the required dilution factor; this is the minimum ratio of dilution flowrate to waste flowrate that must be maintained throughout the release to ensure that the Effluent Con-centration Limits of Section 2.1.2 are not exceeded at the point of discharge into the UNRESTRICTED AREA. The required dilution factor, RDF, is calculated as the sum of the dilution factors required for gamma emitters (RDFy ) and for non-gamm&-emitters (RDFny)8 C; i RDF "

Ei scLg

I(#') (*')I (2.3) !

- JUF7 + RDry '

c E set RDr = .I #' (2.4)

T (st) (fr) e 2-23 Gen. Rev.-13 e

l FNP-0-M-011 Ua Cs C Cg

, , f.

ECL a s ECL, ECL g RDF = ECLj , gg,gy U

j (SF) (TE) '

where: '

Cg = the measured concentration of radionuclide i as defined in step'1, in yCi/mL. The C a, C,, cf, and Ct terms will be included in the calculation as appropriate.

ECL =

the Effluent Concentration Limit for radionuclide i from 10'CFR Part 20, Appendix B, Table 2, Column.2 (except for noble gases as discussed below). In-the absence-of information regarding the solubility classification of a given radionuclide in the . waste stream, the solubility class with the lowest ECL shall be assumed.

For dissolved or entrained noble gases, the concentration shall be limited to lx104 pC1/mL. For gross alpha, the ECL shall be 2x10'9 yci/mL; if specific alpha-emitting radionuclides are measured, the ECL for the specific radionuclide(s) should be used.

SF = the safety factor selected O

to compensate for statistical fluctuations and errors of measurement. The value for the safety factor must be between 0 and 1. A value of 0.5 is reasonable for liquid releases; a more precise value may be developed if desired.

TF =

the tolerance factor (as defined in Section 2.3.2.1).

1 Step 3: Determine the release-specific assured dilution stream flowrate. i Determine the dilution stream flowrate that can be assured during the release period, designated F d*

If simultaneous radioactive releases are planned from the same reactor unit, the unit's dilutama stream must be allocated among all the simultaneous releases, whether or not they are monitored during release. Normally, only the Waste Monitor Tank and Steam Generator Blowdown effluents need be considered, unless there is detectable radioactivity in one of the normally low-radioactivity-streams (see Table 2-4), or in the Turbine Building Sump.. Allocation of the dilution stream to multiple release paths is accomplished as follows:

where:

O 2-24 Gen. Rev. 13

i FNP-O-M-011 Fdp

Ed ( AFp ) (2.6)

F dp = the dilution flowrate allocated to release pathway p, in gpm.

AFp =

the dilution allocation factor for release pathway P. AFp may be ,

assigned any value between 0 and 1 for each active release pathway, j under the condition that the sum of the AF p for all active release pathways for each unic does not exceed 1. [ Note: Because the two units have separate dilution streams, the two units do not affect each other with respect to dilution allocation.]

Fd= the assured minimum dilution flowrate for the unit, in gpm.

If more precise allocation factor values are desired, they may be determined based on the relative radiological impact of each active release pathway; this may be approximated by multiplying the RDF of each affluent stream by its respective planned release flowrate, and comparing these values. If only one release pathway for a given reactor unit contains detectable radioactivity, its AFp may be assigned the value of 1, making F dp equal to Fd '

For the case where RDF s 1, the planned release meets the limits of Section 2.1.2 O~ without dilution, and may be released with any desired affluent flowrate and dilution flowrate.

Sten 4:

Determine the maximum allowable waste discharge flowrate.

For the case where RDF > 1, the maximum permissible affluent discharge flowrate for this release pathway, 2', (in gpm), is calculated as follows:

f = dP 9

(RDF - 1)

(2.7)

)

For the case RDF s 1, equation (2.7) is not valid. However, as discussed above, when RDF s 1, the r,elease may be made at full discharge pump capacity; the radio-activity monitor'setpoint must still be calculated in accordance with step S ,

below.

NOTE la Discharge flowrates are actually limited by the discharge pump '

capacity. When the calculated maximum permissible release flowrate exceeds the pump capacity, the release may be made at full 4 capacity. Discharge flowrates less than the pump capacity must be '

-) ,

2-25 cen. Rev. la

\

1 ;

1 FNP-0-M-011 achieved by throttling if this is available; if throttling is not available, the release may not be made as planned.

NOTE 2:

If, at the time of the planned release, there is detectable radio-activity due to plant operations in the dilution stream, the diluting capacity of the dilution stream is diminished. (In e

addition, sampling and analysis of the other radioactive effluents ,

affecting the dilution stream must be suf ficient to ensure that the

liquid effluent dose limits specified in the controls of section ;

2.1.3 are not exct-eded.) Under these conditions, equation (2.7) must be modified to account for the radioactivity present in the dilution stream prior to the introduction of the planned release:

t , j

' r i F f C I mp *

(RDF - 1) ,

1-E r

{

Ed i , ECLb (2.8) where:

i j

Cir.= the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the dilution stream.

i i

4 O f r =

the effluent discharge flowrate of release pathway r.

U If the entire dilution stream contains detectable activity due to

[ plant operations, whether or not its source is identified, fr=F' d

{

and Cif is the concentration in the total dilution system. This note does not apply: a) if the RDF of the planned release is 5 1; or b) if the release. contributing radioactivity to-the dilution

' stream has been accounted for by the assignment of an allocation factor.

a 3tssL1 Determine the maximum radioactivity monitor setpoint concentration.

i Based on the values determined in previous steps, the radioactivity monitor setpoint for the planned release is calculated to ensure that the - limits of section 2.1.2 will not be exceeded. Because the radioactivity monitor responds primarily to gaema radiation, the monitor setpoint cp for release pathway p (in yC1/mL) is based on the concentration of gamma emitters in the waste stream, as follows:

0 2-26 Gen. Rev. 13 o

-. , _ . , n

a FNP-0-M-011 I I

cp = A p[C g gg,9y 8

where: l

'l

, A p= an adjustment factor which will allow the setpoint to be < i established in a practical manner to prevent spurious alarms while i l

allowing a margin between measured concentrations and the limits of Section 2.1.2. '

l sten 5.a. If the concentration of gamma emitters in the effluent to

, be released is sufficient that the high alarm setpoint can be established at a level that will prevent spurious alarms, Ap should be calculated as follows:

i Ap e x ADF

\

,' RDF 3

, 1 , (fdp + fap) (2.10) 4 nr fy 2

I i

where

l ADF = the assured dilution factor. 1 i

i '

) f,p = the anticipated actual discharge flowrate for the

) planned release (in gpm), a value less than fg.

i

The release must then be controlled so that the I actual effluent discharge flowrate does not j

exceed f,p at any time.

sten 5.h. i 4

Alternatively, A pmay be calculated as follows:

l ADT -RDT Ap = 9 MT y (2.11)

Sten 5.e.

Evaluate the compted value of A p as follows:

If A pa 1, calculate the monitor setpoint, c. p However, if c is p

within about 10 percent of Cg , It may be impractical to d 2-27 Gen. Rev. 13

4I i

FNP-0-M-011 use this value of e. p This situation indicates that measured concentrations are approaching values which

i

< \

would cause the limits of Section 2.1.2 to be exceeded.

Therefore, steps should be taken to reduce potential con-centrations at the point of discharge; these steps may include, decreasing the planned effluent discharge <

flowrate, increasing the dilution stream flowrate, postponing simultaneous releases, and/or decreasing the effluent concentrations by further processing the liquid planned for release. Alternatively, allocation factors for the active liquid release pathways may be reassigned. ;

When one or more of these actions has been taken, repeat I steps 1-5 to calculate a new radioactivity monitor l setpoint. -

If Ap < 1, the release may not be made as planned. Consider the alternatives discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.

)

i 2.3.2.3 Use of the Calculated setpoint The setpoint calculated above is in the units ci/mL. The monitor actually measures a count rate that includes background, so that the calculated setpoint must be converted accordingly:

j c = cp *E p+Bp (2.Sa}

where:

I ch= the monitor setpoint as a count rate.

l E p= the monitor calibration factor, in count rate /(gci/mL). Monitor calibration data

{

for conversion between count rate and {

concentration may include operational data obtained from j determining the monitor response to stream concentrations measured

{

by liquid semple analysis.

B p= the monitor background count rate. In all cases, monitor )

background must be controlled so that th0 monitor is capable of i responding to concentrations in the range ot' the setpoint value.

Eh g ,

2-28 Cen. Rev. 13

l l

FNP-0-M-011 i The count rate units of e p, E ,p and B in p equation (2.8a) must be the same (cpm or eps).

J 2.3.3 Setooints for Monitors on Normally Low-Radioactivity Streams Radioactivity in these streams (listed in Table 2-4 above) is expected to be at very low levels, generally below detection limits. Accordingly, the purpose of these monitors is to alarm upon the occurrence of significant radioactivity in these streams, and to terminate or divert the release where this is possible.

2.3.3.1 Normal Conditions i

l When radioactivity in one of these streams is at its normal low level, its radio-activity monitor setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should an inadvertent release occur.

j 2.3.3.2 Conditions Requiring an Elevated Setpoint Under the following conditions, radionuclide concentrations must be determined and an elevated radioactivity monitor setpoint determined for these pathways:

e For streams that can be diverted or isolated, a new monitor setpoint must be established when it is desired to discharge the stream directly to the dilution water even though the radioactivity in the stream exceeds the level which would normally be diverted or isolated.

l e

For streams that cannot be diverted or isolated, a new monitor setpoint must be established whenever: the radioactivity in the stream becomes

]

detectable above the background levels of the applicable laboratory analyses; or the associated radioactivity monitor detects activity in the I stream at levels above the established alarm setpoint.

When an elevated monitor eetpoint is required for any of these effluent streams, it should be determined in the same manner as ' described in Section 2.3.2.

^ However, special consideration must be given to Step 3. An allocation factor must be assigned to the normally low-radioactivity release pathway under consideration, and allocation factors for other release pathways discharging simultaneously must be adjusted downward (if necessary) to ensure that the sum of the allocation factors does not exceed 1. Sampling and analysis of the normally low-radioactivity streams must be sufficient to ensure that the liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.

O 2-29 Gen. Rev. 13

i FMP-0-M-011 2.4 LIQUID EFFLUENT DOSE CALCULATIONS I O The following sub-sections present the methods required for liquid ef fluent dose calculations, in deepening levels of detail. Applicable site-specific pathways and parameter values for the calculation of D *r Air, and CFiy are summarized in l Table 2-5.

4 2.4.1 Calculation of Dose 1 1

The dose limits for a MEMBER OF THE PUBLIC specified in Section 2.1.3 are on a per-unit basis. Therefore, the doses calculated in accordance with this section must be determined and recorded on a per-unit basis, including apportionment of releases shared between the two units.

For the purpose of implementing Section 2.1.3, the dose to the maximum exposed individual due to radionuclides identified in liquid ef fluents released from each unit to UNRESTRICTED AREAS will be calculated as follows (equation from Ref-erence 1, page 15):

m D7 =

[ Ajf [ (Atj Cff f)f (2 12) i . l= 1 O !

where:

D7 =

the cumulative dose commitment to the total body or to any organ 7, in mrem, due to radioactivity in liquid effluents released during .i' the total of the m time periods Atg. i

, A 7= the site-related adult ingestion dose commitment factor, for the total body or for any organ r, due to identified radionuclide i, in (arememL)/(h*yci). Methods for the calculation of Ajy are presented below in section 2.4.2.

The values of Ajf to be used in does calculations for releases from the plant site are listed in Table 2-8.

Atg =

the length of time period 1, over which Cgg and Fg are averaged for liquid releases, in h.

Cg=

i the average concentration of radionuclide i in undiluted liquid effluent during time period 1, in yC1/mL. Only radionuclides O 2-30 Gen. Rev. 13

I

j. ., !

i- Fi4P-0-M-011

)

j identified and detected above background in their respective 1

{ samples should be included in the calculation.

I 1

Fg = the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA:

i fg

fj = \

Eg x2 (2.13) wheres ft= the average undiluted liquid waste flowrate actually observed during the period of radioactivity release, in-9Pm.

Fg= the average dilution stream flowrate actually observed during the period of radioactivity release, in gpm.

Z=

the applicable dilution factor for '.he receiving water body, in the near field of the discharge structure, during the period of radioactivity release, from Table 2-5.

NOTE:

In equation (2.13), the product (Fg x Z) is limited to O 1000 cfs (= 448,000 gpm) or less.

4.3.)

(Reference 1, Section ;

4 i 1

2.4.2 Calculation of A;7 3

The site-related adult ingestion dose commitment factor, A[7, is calculated as follows (equation adapted from Reference 1, page 16, by addition of the irrigated !

garden vegetation pathway):

Ag = 1.14 x 10 5 _

,4 'w 1

gf gf ,-N if , g y gfj, (2.14) pfjy .

l where:

l 1.14 x 10 5 = a units conversion factor, determined by:

100 pCi/pci x 103 mL/L + 8760 h/y.

l p

l G 2-31 Gen. Rev. 13 l

i FNP-0-M-011 Uw= tho cdult drinking water consumption rate applicable to the plant

site (L/y).

D, = the dilution factor from the near field of the discharge structure for the plant site to the potable water intake location.

t Ig = the decay constant for radionuclide 1 (h-I).

' Values of li used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 15.

A tw= the transit time from release to receptor for potable water-j consumption (h).

Ur = the-adult rate of fish consumption applicable to the plant site-1 (kg/y).

BFj = the bioaccumulation factor for radionuclide i applicable to-j freshwater fish in the receiving water body for the plant site, in (pci/kg)/(pci/L) = (L/kg). For specific values applicable to the plant site, see Table 2-6.

4 tg = the transit time from release to receptor for fish consumption (h).

iO 4,

Uy =

the adult consumption rate for irrigated garden vegetation i applicable to the plant site (kg/y). '

CFjy = the concentration factor for radionuclide i in irrigated garden vegetation, as applicable to ths wi:inity of the plant site, in j (pci/kg)/(pci/L).

Methods for calculation of CFjy are presented

} below in Section 2.4.3.

i

=

}

i DF;7 the dose conversion factor for radionuclide i for adults, in j organ f (mres/pci). For specific values, see Table 2-7.

2.4.3 cateulation of CFjy The concentration factor for radionuclide i in irrigated garden vegetation, CFiy in (L/kg), is calculated as follows:

i

'v 2-32 Gen. Rev. 13 i

)

4

<. , ,- f -e L

.II i . I FNP-0-M-011 c o For radionuclidos other then tritium (equation adapted from Reference 3, equations A-8 and A-9):

I r (1 - e Ei 'e ff Bjy (1 - e )

City = M I 3 +

e -A tfh (2.15) l Yykgg P kj <

\

o For tritium (equation adapted from Reference 3, equations A-9 and A-10):

CEjy =M*L y (2.16) where:

M=

the additional river dilution factor from the near field of the discharge structure for the plant site to the point of irrigation water usage.

I I=

the average irrigation rate during the growing season (L)/(m2 .h).

I r= the fraction of irrigation-deposited activity retained on the l

edible portions of leafy garden vegetation.

Yy =

the areal density (agricultural productivity) of leafy garden vegetation (kg/m 2)

{ i fg = the fraction of the year that garden vegetation is irrigated.

B y= the crop to soil concentation factor applicable to radionuclide 1,

from Table 2-6 (pci/kg garden vegetation)/(pci/kg soil).

P= the effective surface density of soil (kg/m2 ).

1 =

the decay constant for radionuclide 1 (h-I). Values of 1; used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 15.

1, = the rate constant for removal of activity from plant leaves by j weathering (h-I) .

l l .

2-33 Gen. Rev. 13 l

i

-)

FNP-0-M-011 AEi = the effective removal rate for activity. deposited on crop leaves (h-I) calculated as: -AEi * ~ Ai+A w' q

)

t, = the period of leafy garden vegetation exposure during the growing season (h).

tb= the period of long-term buildup of. activity in. soil _(h).

th= the time between harvest of garden vegetation and human consumption (h).

1 L, = the water content of leafy garden-vegetation edible. parts-(L/kg)'.

i l

4

-1

-l l

O

~

N U

2-34 Gen. Rev. 13

FNP-0-M-011 Tcble 2-5.

Parameters for Calculation of Doses Due to Liquid Ef fluent Releases Dose Calculation Receptor Locations:

Eight Vicinity of plant discharge Drinkino Water: None (Ref. 10)

Irricated Carden Vecetation: Farms at River Mile 26 (Ref. 10) <

Numerical Parameters:

Parameter Value Reference Z 5 Ref. 2, Table A-1 Uw 0 L/y

Ref. 10 D, 1.0

Based on Ref. 1, Section 4.3.1 tw 12 h

Ref. 3, Sec. A.2 U, 21 kg/y Ref. 3, Table E-5 tg 24 h Ref. 3, Sec. A.2 Uy 64 kg/y Ref. 3, Table E-5 1

M O.04 Ref. 16

/

I O.126 L/(m2 h) Ref. 10, using pump capacity, k

s garden size, and irrigation 10% of the time.during growing season.

r 0.25 Ref. 3, Table E-15 Yy 2.0 kg/m2 Ref. 3, Table E-15 fg 0.1 Ref. 10 l i

P 240 kg/m 2 Ref. 3, Table E-15 1, 0.0021 h-I (i.e., half- Ref. 3, Table E-15 life = 14 d) l t, 1440 h (= 60 d) Ref. 3, Table E-15 tb 1.31 x 105 h (= 15 y) Ref. 3, Table E-15 th 24 h Ref. 3, Table E-15 Ly 0.92 L/kg Based on Ref. 11, Table 5.16 (for lettuce, cabbage, etc.)

Because there is no drinking water pathway downstream of the plant site, the consumption of drinking water is set to zero, and the default values of tw and D, are used.

N.)

2-35 Gen. Rev. 13

1 I

FNP-O-M-011 Tablo 2-6. Elcm:nt Trenofer Factors '

Freshwater Fish Leafy Garden Vegetation f Element BFi

Djy +

H 9.0 E-01 4.8 E+00' C 4.6 E+03 5.5 E+00 Na 1.0 E+02 5.2 E-02 P 3.0 E+03 1.1 E+00 Cr 2.0 E+02 2.5 E-04 Mn 2.0 E+01 2.9 E-02

(

Fe 1.0 E+03 6.6 E-04 Co 1.0 E+02 9.4 E-03 Ni 1.0 E+02 1.9 E-02 Cu 1.5 E+02 Zn 1.2 E-01 1.0 E+02 4.0 E-01 Br 4.2 E+02 7.6 E-01 Rb 2.0 E+03 1.3'E-01 i i

Sr 3.0 E+01 1.7 E-02 Y 2.5 E+01 !

2.6 E-03 i Zr 2.0 E+02 1.7 E-04 1 Nb 1.0 E+02 i Mo 9.4 E-03. j 1.0 E+02 1.2 E-01 i Tc 1.5 E+01 2.5 E-01 Ru 1.0 E+01 5.0 E-02 Rh 1.0 E+01 Ag 1.3 E+012 2.3 E+00 1.5 E-01 '

Sb 3.0 E+02 Te 1.1 E-02 2.0 E+03 1.3 E+00 I 2.0 E+01 2.0 E-02 Cs 2.0 E+02 Ba 1.0 E-02 4.0 E+01 5.0 E-03 La 2.5 E+01 '2.5 E-03 Ce 2.0 E+02 Pr 2.5 E-03 2.5 E+01 2.5 E-03 -

i Nd 2.5 E+01 2.4 E-03 W 1.2 E+03 1.8 E-02 Np 1.0 E+01 2.5 E-03 I

t Bioaccumulation Factors for freshwater fish, in (pCi/kg)/(pci/L).

They are obtained from Reference 3 (Table A-1), except as follows:

Reference 9 for P; Reference 2 (Table A-8) for Ag; Reference 8 for Mn, Fe, Co, Cu, En, Mo, Sb, Te, I, Cs, Ba, and Ce; and Reference 14 for Zr and Nb.

+ t Crop to soil concentration factors, in (pci/kg garden vegetation) per (pci/kg soil). They are obtained from Reference 3 (Table E-1),

except as follows: Reference 2 (Table C-5) for Br and Sb.

1 i

O- 2-36 Gen. Rev. 13 I

i

FNP-0-M-011 3

Tr.blo 2-7. Adult Ingestion Dose Factors Nuclide Bone Liver Body T. Thyroid Kidney Lung GI-LLI i

i H-3 No Data 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 i j Cr-51 No Data No Data 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 j Mn-54 No Data 4.57E-06 8.72E-07 No Data 1.36E-06 No Data 1.40E-05 Mn-56 No Data 1.15E-07 2.04E-08 No Data 1.46E-07 4 No Data 3.67E-06

, Fe-55 2.75E-06 1.90E-06 4.43E-07 No Data No Data 1.06E-06 4 1.09E-06 Fe-59 4.34E-06 1.02E-05 3.91E-06 No Data No Data 2.85E-06 3.40E-05

. Co-58 tio Data 7.45E-07 1.67E-06 i

No Data No Data No Data 1.51E-05 Co-60 No Deta 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-05 Ni-63 1.431-04 9.01E-06 4.36E-06 No Data

! No Data No Data 1.88E-06 Ni-65 5.23E-07 6.86E-08 3.13E-08 No Data No Data No Data 1.74E-06 Cu-64 No Deta 8.33E-08 3.91E-08 No Data 2.10E-07 No Data 7.10E-06 f~s 2n-65 4.84E-06 1.54E-05 6.96E-06 No Data 1.03E-05 No Data 9.70E-06 Zn-69 1.03E-08 1.97E-08 1.37E-09 No Data l 1.28E-08 No Data 2.96E-09

] Br-83 No Data No Data 4.02E-08 No Data No Data No Data 5.79E-08

, Br-84 No Data No Data 5.21E-08 No Data No Data i

l No Data 4.09E-13 Br-85 No Data No Data 2.14E-09 No Data No Data No Data No Data Rb-86 No Data 2.11E-05 9.83E-06 i

No Data No Data No Data 4.16E-06 Rb-88 No Data 6.05E-08 3.21E-08 No Data No Data No Data 8.36E-19 Rb-89 No Data 4.01E-08 2.82E-08 No Data No Data No Data 2.33E-21 Sr-89 3.08E-04 No Data 8.84E-06 No Data No Data No Data 4.94E-05 Sr-90 7.58E-03 No Data 1.86E-03 No Data No Data No Data 2.19E-04 Sr-91 5.67E-06 No Data 2.29E-07 No Datt No Data No Data 2.70E-05 i

All values are in (mrom/pci ingested) . They are obtained from Reference 3 (Table E-11), except as follows: Reference 2 (Table A-3) for Rh-105, sb-124, and Sb-125.

4 j -

2-37 Gen. Rev. 13 1

i n * '

i i

FNP-0-M-011 Teblo 2-7 (contd). Adult Ingostion Dose Factors m

Nuclide Bone Ltver T. Body Thyroid Kidney Lung GI-LLI Sr-92 2.15E-06 No Data 9.30E-08 No Data No Data No Data 4.26E-05 Y-90 9.62E-09 No Data 2.58E-10 No Data No Data '

No Data 1.02E-04 Y-91m 9.09E-11 No Data 3.52E-12 No Data No Data No Data 2.67E-10 Y-91 1.41E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-05 Y-92 8.45E-10 No Data 2.47E-11 No Data No Data No Data 1.48E-05 Y-93 2.68E-09 No Data 7.40E-11 No Data No Data No Data 8.50E-05 Zr-95 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 Zr-97 1.68E-09 3.39E-10 1.55E-10 No Data 5.12E-10 No Data 1.05E-04 Nb-95 6.22E-09 3.46E-09 1.86E-09 No Data 3.42E-09 No Data 2.10E-05 Mo-99 No Data 4.31E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 Tc-99m 2.47E-10 6.98E-10 8.89E-09 No Data 1.06E-08 3.42E-10 4.13E-07 Tc-101 2.54E-10 3.66E-10 3.59E-09 No Data 6.59E-09 1.87E-10 1.10E-21 Ru-103 1.85E-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 Ru-105 1.54E-08 No Data 6.08E-09 No Data 1.99E-07 No Data 9.42E-06 Ru-106 2.75E-06 No Data 3.48E-07 No Data 5,31E-06 No Data 1.78E-04 Rh-105 1.22E-07 8.86E-08 5.83E-08 No Data 3.76E-07 No Data 1.41E-05 Ag-110m 1.60E-07 1.48E-07 8.79E-08 No Data 2.91E-07 No Data 6.04E-05 Sb-124 2.81E-06 5.30E-08 1.11E-06 6.79E-09 No Data 2.18E-06 7.95E-05 Sb-125 2.23E-06 2.40E-08 4.48E-07 1.98E-09 No Data 2.33E-04 1.97E-05 Ta-125m 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E-05 No Data 1.07E-05 Te-127m 6.77E-06 2.42E-06 8.25E-07 1.73E-06 2.75E-05 No Data 2.27E-05 Te-127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 No Data 8.68E-06 Te-129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 No Data 5.79E-05 Te-129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 No Data 2.37E-08 Te-131m 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 No Data 8.40E-05 Te-131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 No Data 2.79E-09 4

./ 2-38 Gen. Rev. 13

FMP-0-M-011 Teblo 2-7 (contd). Adult Ingestion Dose Factors i

O Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 No Data 7.71E-05 e I-130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 No Data 1.92E-06 I-131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 No Data 1.57E-06 I-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 No Data 1.02E-07 I-133 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31E-06 No Data 2.22E-06 I-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 No Data t 2.51E-10 I-135 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 No Data 1.31E-06 i Cs-134 6.22E-05 1.48E-04 1.21E-04 No Data 4.79E-05 1.59E-05 !

2.59E-06 Cs-136 6.51E-06 2.57E-05 1.85E-05 No Data 1.43E-05 1.96E-06 2.92E-06 Cs-137 7.97E-05 1.09E-04 7.14E-05 No Data 3.70E-05 1.23E-05 i

2.11E-06 Cs-138 5.52E-08 1.09E-07 5.40E-08 No Data 8.01E-08 7.91E-09 4.65E-13 l Ba-139 9.70E-08 6.91E-11 2.84E-09 No Data I 6.46E-11 3.92E-11 1.72E-07 Ba-140 2.03E-05 2.55E-08 1.33E-06 No Data 8.67E-09 1.46E-08 4.18E-05 Ba-141 4.71E-08 3.56E-11 1.59E-09 No Data 3.31E-11 2.02E-11 2.22E-17 ;

Ba-142 2.13E-08 2.19E-11 1.34E-09 No Data 1.85E-11 1.24E-11 3.00E-26 La-140 2.50E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.25E-05 La-142 1.28E-10 5.82E-11 1.45E-11 No Data I No Data No Data 4.25E-07 1 co-141 9.36E-09 6.33E-09 7.18E-10 No Data 2.94E-09 l

No Data 2.42E-05 l Ce-143 1.65E-09 1.22E-06 1.35E-10 No Data 5.37E-10 No Data 4.56E-05 I Ce-144 4.88E-07 2.04E-07 2.62E-08 No Data 1.21E-07 No Data 1.65E-04 Pr-143 9.20E-09 3.69E-09 4.56E-10 No Data 2.13E-09 No Data 4.03E-05 Pr-144 3.01E-11 1.25E-11 1.53E-12 No Data 7.05E-12 No Data 4.33E-18 Nd-147 6.29E-09 7.275-09 4.35E-10 No Data 4.25E-09 No Data 3.49E-05 W-187 1.03E-07 8.61E-08 3.01E-08 No Data No Data No Data 2.82E-05 Np-239 1.19E-09 1.17E-10 6.45E-11 No Data 3.65E-10 No Data 2.40E-05 2-39 Gen. Rev. 13

f

! FNP-0-M-011 Table 2-8.

Site-Related Ingestion Dose Factors, Ag Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 2.54E-01 2.54E-01 2.54d-01 2.54E-01 2.54E-01 2.54E-01 Na-24 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 <

Cr-51 0.00 0.00 1.25E+00 7.45E-01 2.74E-01 1.65E+00 3.13E+02 Mn-54 0.00 2.28E+02 4.34E+01 0.00 6.77E+01 0.00 6.97E+02 Mn-56 0.00 8.69E-03 1.54E-03 0.00 1.10E-02 0.00 2.77E-01 Fe-55 6.58E+03 4.55E+03 1.06E+03 0.00 0.00 2.54E+03 2.61E+03 Fe-59 1.02E+04 2.41E+04 9.22E+03 0.00 0.00 6.72E+03 8.02E+04 Co-58 0.00 1.78E+02 3.99E+02 0.00 0.00 0.00 3.61E+03 Co-60 0.00 5.17E+02 1.14E+03 0.00 0.00 0.00 9.71E+01 Ni-63 3.14E+04 2.18E+03 1.05E+03 {

0.00 0.00 0.00 4.54E+02

{

Ni-65 1.72E-01 2.23E-02 1.02E-02 0.00 0.00 0.00 5.66E-01 Cu-64 0.00 8.07E+00 3.79E+00 0.00 2.04E+01 0.00 6.88E+02 Zn-65 1.17E+03 3.71E+03 1.68E+03 0.00 2.48E+03 0.00 2.34E+03 Zn-69 3.94E-08 7.54E-08 5.24E-09 0.00 4.90E-08

' 0.00 1.13E-08 Br-83 0.00 0.00 3.83E-02 0.00 0.00 0.00 5.52E-02 Br-84 0.00 0.00 1.22E-12 0.00 0.00 0.00 9.61E-18 Br-85 0.00 0.00 0.00 0.00 0.00 0.00 0.00 l Rb-86 0.00 9.74E+04 4.54E+04 0.00 0.00 0.00 1.92E+04 Rb-88 0.00 0.00 0.00 0.00 0.00 0.00 0'

0.00 Rb-89 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sr-89 2.23E+04 0.00 6.41E+02 0.00 0.00 0.00 3.58E+03 Sr-90 5.61E+05 0.00 1.3PE+05 0.00 0.00 i

0.00 1.62E+04 Sr-91 7.07E+01 0.00 2.86E+00 0.00 0.00 0.00 3.37E+02 Sr-92 3.33E-01 0.00 1.44E-02 0.00 0.00 0.00 6.60E+00 Y-90 4.47E-01 0.00 1.20E-02 0.00 0.00 0.00 4.74E+03 Y-91m 1.04E-11 0.00 4.01E-13 0.00 0.00 0.00 3.04E-11 4

Y-91 8.58E+00 0.00 2.30E-01 0.00 0.00 0.00 4.72E+03 Y-92 4.60E-04 0.00 1.35E-05 0.00 0.00 0.00 8.07E+00 Y-93 3.09E-02 0.00 8.54E-04 0.00 0.00 0.00 9.81E+02 Zr-95 1.45E+01 4.64E+00 3.14E+00 0.00 7.27E+00 0.00 1.47E+04 Zr-97 3.01E-01 6.07E-02 2.77E-02 0.00 9.16E-02 0.00 1.88E+04 Nb-95 1.47E+00 8.17E-01 4.39E-01 0.00 8.08E-01 0.00 4.96E+03 ;

Mo-99 0.00 8.03E+02 1.53E+02 0.00 1.82E+03 0.00 1.86E+03 4

Tc-99m 5.60E-04 1.58E-03 2.02E-02 0.00 2.40E-02 7.76E-04 9.37E-01 i

All values are in (mrememL)/(h*yC1). They are calculated using equation (2.14), and data from Table 2-5, Table 2-6, and Table 2-7.

When "No Deta" is shown for a radionuclide-organ combination in Table 2-7, Ag factors in this table are presented as zero.

O 2-40 Gen. Rev. 13 4

d

FNP-0-M-011 1 Teblo 2-8 (contd). Site-Related Ingestion Dose Factors, Ag.

l O Nuclide Tc-101 Bone 0.00 Liver 0.00 T. Body 0.00 Thyroid 0.00 Kidney 0.00 Lung 0.00 GI-LLI 0.00 Ru-103 4.65E+00 0.00 2.00E+00 0.00 1.77E+01 0.00 5.42E+02 Ru-105 8.71E-03 0.00 3.44E-03 i

0.00 1.13E-01 0.00 5.33E+00 !

Ru-106 7.14E+01 0.00 9.03E+00 0.00 1.38E+02 0.00 4.62E+03 Rh-105 1.84E+00 1.34E+00 8.80E-01 0.00 5.68E+00 0.00 2.13E+02 Ag-110m 1.20E+00 1.11E+00 6.61E-01 0.00 2.19E+00 0.00 4.54E+02 Sb-124 2.00E+03 3.77E+01 7.90E+02 4.83E+00 0.00 1.55E+03 5.66E+04 Sb-125 1.61E+03 1.73E+01 3.22E+02 1.43E+00 0.00 1.68E+05 1.42E+04 Te-125m 1.27E+04 4.60E+03 1.70E+03 3.81E+03 5.16E+04 0.00 5.06E+04 Te-127m 3.22E+04 1.15E+04 3.93E+03 8.23E+03 1,31E+05 0.00 1.08E+05 Te-127 8.89E+01 3.19E+01 1.92E+01 6.59E+01 3.62E+02 0.00 7.01E+03 Te-129m 5.40E+04 2.01E+04 8.54E+03 1.85E+04 2.25E+05 0.00 2.72E+05 To-129 8.89E-05 3.34E-05 2.17E-05 6.82E-05 3.74E-04 0.00 6.71E-05 Te-131m 4.76E+03 2.33E+03 1.94E+03 3.69E+03 2.36E+04 0.00 2.31E+05 Te-131 4.32E-16 1.80E-16 1.36E-16 3.55E-16 1.89E-15 0.00 6.12E-17 Te-132 9.75E+03 6.31E+03 5.92E+03 6.97E+03 6.08E+04 ,

0.00 2.98E+05 '

I-130 9.44E+00 2.78E+01 1.10E+01 2.36E+03 4.34E+01 0.00 2.40E+01 I-131 1.86E+02 2.66E+02 1.52E+02 8.71E+04 4.56E+02 0.00 7.01E+01 ;

I-132 7.02E-03 1.88E-02 6.57E-03 6.57E-01 2.99E-02 0.00 3.53E-03 I-133 3.06E+01 5.33E+01' 1.62E+01 7.83E+03 9.30E+01 0.00 4.79E+01

)

I-134 2.91E-08 7.92E-08 2.83E-08 1.37E-06 1.26E-07

' 0.00 6.90E-11 I-135 1.71E+00 4.49E+00 1.66E+00 2.96E+02 7.20E+00 0.00 5.07E+00 Cs-134 2.99E+04 7.11E+04 5.81E+04 0.00 2.30E+04 7.64E+03 1.24E+03 t Cs-136 2.96E+03 1.17E+04 8.42E+03 0.00 6.51E+03 8.92E+02 1.33E+03 Cs-137 3.83E+04 5.24E+04 3.43E+04 0.00 1.78E+04 5.92E+03 1.01E+03 Cs-138 9.12E-13 1.80E-12 8.92E-13 i 0.00 1.32E-12 1.31E-13 7.68E-18 !

Ba-139 5.64E-05 4.02E-08 1.65E-06 0.00 3.76E-08 2.287-08 1.00E-04 Ba-140 I 1.86E+03 2.34E+00 1.22E+02 0.00 7.95E-01 1.34E+00 3.83E+03 i l

Ba-141 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Ba-142 0.00 0.00 0.00 0.00 0.00 0.00 0.00 La-140 9.93E-02 5.01E-02 1.32E-02 0.00 0.00 0.00 3.68E+03 La-142 2.19E-07 9.96E-08 2.48E-08 0.00 0.00 0.00 7.27E-04 Co-141 4.40E+00 2.98E+00 3.38E-01 0.00 1.38E+00 0.00 1.14E+04 Co-143 4.77E-01 3.53E+02 3.91E-02 0.00 1.55E-01 0.00 1.32E+04 Co-144 2.34E+02 9.79E+01 1.26E+01 0.00 5.80E+01 0.00 7.91E+04 Pr-143 5.33E-01 2.14E-01 2.64E-02 0.00 1.23E-01 0.00 2.33E+03 Pr-144 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Nd-147 3.59E-01 4.15E-01 2.48E-02 0.00 2.43E-01 0.00 1.99E+03 W-187 1.47E+02 1.23E+02 4.30E+01 0.00 0.00 0.00 4.03E+04 Np-239 2.15E-02 2.112-03 1.17E-03 0.00 6.60E-03 0.00 4.34E+02 O -

2-41 Gen. Rev. 13

FNP-0-M-011 2.5 LIQUID EFFLUENT DOSE PROJECTIONS J

2.5.1 Thirty-one Day Dose Proiections i

j 1 In order to meet the requirements for operation of the LIQUID RADWASTE TREATMENT SYSTEM (see Section 2.1.4), dose projections must be made at least once each 31 ,

days; this applies during periods in which a discharge to UNRESTRICTED AREAS of liquid effluents containing radioactive materials occurs or is expected. .

I I Projected 31-day doses to individuals due to liquid effluents may be determined as follows:

a f 1 0,p = x 31 +D ra I*D a

i i

where

1; D

9= the projected dose to the total body or organ r, for the next 31 e

days of liquid releases.

Dye = the cumulative dose to the total body or organ 4 r, - for liquid i

releases that have occurred in the elapsed portion of the current-quarter, plus the release under consideration.

i t t=

the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues 'into the next quarter).

4 i

, D,=

7 the anticipated dose contribution to the total body or any organ r,

-)

! due to any planned activities during the next 31-day period, if those activities will result in liquid releases that are in j addition to routine liquid effluents. If only routine -liquid effluents are anticipated, D ra may be set to zero.

1

) 2.5.2 Dona Preisetions for Specific Releases i Dose projectione may be performed for a particular release by performing a pre- .

I release dose calculation assuming that the planned release will proceed as anticipated. For individual dose projections due to liquid releases, follow the methodology of Section 2.4, using sample analysis results for the source to be released, and parameter values expected to exist during the release period..

iO 2-42. Gen. Rev. .13 i

4

I i )

1 FNP-0-M-011 l 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS The following symbolic terms are used in the presentation of liquid effluent calculations in the sub-sections above. I M Definition Section of !

Initial Use Ap =

the adjustment factor used in calculating the effluent monitor setpoint for liquid release pathway p the ratio of the assured dilution to the required dilution [unitiess). 2.3.2.2 ADF = the assured dilution factor for a planned release

! [unitiess). , 2.3.2.2 1 AFp=

the dilution allocation factor for liquid release pathway p (unitless).

2.3.2.2 A =

r the site-related adult ingestion dose commitment factor, for the total body or for any organ r, due to identified radionuclide 1 ((mrem mL)/(hayci)). The values of Air are listed in Table 2-8. 2.4.1 i B;y =

O the crop to soil concentration factor applicable to radionuclide 1,

[(pci/kg garden vegetation)/(pci/kg soil)). Values are listed in Table 2-6. 2.4.3 BFj =

the bicaccumulation factor for radionuclide i for freshwater fish [(pcl/kg)/(pci/L)). Values are listed in Table 2-6.

2.4.2 e=

the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent line, prior to dilution and subsequent release (pci/mL).

2.3.2.1 e p= the calculated effluent radioactivity monitor setpoint for liquid release pathway p (pci/mL). 2.3.2.2 c, = the gross concentration of alpha emitters in the liquid waste as measured in the applicable composite sample (yci/mL).

2.3.2.2

,O i

2-43 Gen. Rev. 13 l

P

~. _ .- . _ _ _ _ _ _ _

l FNP-0-M-011 J

Section of I.1Im Definition Initial Use s l CECL = the Ef fluent concentration Limit stated in 10 CFR 20, Appendix B, Table 2, Column 2 (pci/mL]. 2.3.2.1 Cf= the concentration of Fe-55 in the liquid waste as 1 measured in the applicable composite- sample

[yci/mL).

2.3.2.2 I Cg= the concentration of gamma emitter g in the liquid i i

waste as measured by gamma ray spectroscopy performed on the applicable pre-release waste sample (yci/mL). 2.3.2.2 C =

the messured concentration of radionuclide i in a sample of liquid effluent (pci/mL). 2.3.2.2 l

l Cd= the average concentration of radionuclide. i in l

undiluted liquid effluent during time period 1 (yci/mL).

2.4.1

Cj, = the measured concentration of radionuclide i in release pathway r that is contributing to radio-activity in the dilution stream (yci/mL). 2.3.2.2 I C, = the concentration of strontium radioisotope s (Sr-89 or Sr-90) in the liquid waste as . measured in the I applicable composite sample (yci/mL). 2.3.2.2 Ct= the concentration of H-3 in the liquid waste as measured in the applicable composite sample (pci/mL).

2.3.2.2

=

CF;y the concentration factor for radionuclide i in irrigated garden vegetation ((pci/kg)/(pci/L)]. 2.4.2 D, = the dilution factor from the near field of the ..

discharge structure to the potable water intake location [unitiess). 2.4.2 Q/

2-44 Gen. Rev. 13

l l, FNP-0-M-011 i

! IAIm Definition Section of Initial Use D7 =

the cumulative dose commitment to the total body or to any organ r, due to radioactivity in liquid affluents released during a given time period (mrem). '

2.4.1 Drs = the anticipated dose contribution to the total body or any organ r, due to any planned activities during the next 31-day period (mram). 2.5.1 Dg =

the cumulative dose to the total body or organ r, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release i under consideration (mrom). 2.5.1 ;

j D) q the projected dose to the total body or organ 7, for the next 31 days of liquid releases (mrem). 2.5.1

=

DFj7 the dose conversion factor for radionuclide i for adults, in organ f [mrom/pci). Values are 11:sted in

{ Table 2-7.

2.4.2 ECLj = the liquid Effluent concentration Limit for-radio-nuclide i from 10 CFR Part 20, Appendix B, Table 2, Column 2 [yci/mL).

2.3.2.2 f=

the effluent flowrate at the location of the radio-activity monitor (gpm].

2.3.2.1 f,p = the anticipated actual discharge flowrate for a planned release from liquid release pathway p (gym].

2.3.2.2 fg = the fraction of the year that garden vegetation is irrigated (unitiess). 2.4.3 f mp = the maximum permissible effluent discharge flowrate for release pathway p [gpm).

2.3.2.2 O. 2-45 Gen. Rev. 13

l l

FNP-0-M-011 Igrm Section of Definition Initial Use

[

( f, = the affluent discharge flowrate of release pathway r

[gpm).

2.3.2.2 ft= the average undiluted. liquid waste flowrate actually observed during the period of a liquid release

[gpm). ,

! 2.4.1 I F=

the dilution stream flowrate which can be assured i prior to the release point to the UNRESTRICTED AREA !

[gpm).

2.3.2.1 Fd= the entire assured dilution flowrate for the plant i site during the release period [gpm). 2.3.2.2 l

Fdp = the dilution flowrate allocated to release pathway p l [gpm).

l 2.3.2.2 Fg = the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA

[unitiess). -

2.4.1 Ft= the average dilution stream flowrate actually observed during the period of a liquid release

[gpm).

2.4.1 I=

the average irrigation rate during the growing season

[L/(m2 h)). 2.4.3 I.y =

the water content of leafy garden vegetation edible parts [L/kg).

2.4.3 M=

the additional river dilution factor from the near field of the discharge structure for the plant site to the point of irrigation water usage [unitiess). 2.4.3 P=

the effective surface density of soil [kg/m2 ). 2.4.3 O

2-46 Gen. Rev. 13

I O FNP-0-M-011 IAIm Definition Section of Initial Use l-r= the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation.

2.4.3 i

RDF = the required dilution factors the minimum ratio by which liquid effluent must be diluted before reaching the UNRESTRICTED AREA, in order to ensure that the limits of Section 2.1.2 are not exceeded l

[unitiess). 2.3.2.2 RDF y=

j the RDF for a liquid release due only to its concen- 1 l

l tration of gamma-omitting radionuclides [unitless). 2.3.2.2 RDF ny = the RDF for a 11guld release due only to its concen-tration of non-gamma-emitting radionuclides

[unitless). 2.3.2.2 SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement j

[unitiess). 2.3.2.2 !

t=

the number of whole or partial dayu elapsed into the t current quarter, including the time to the end of the t

l l

release under consideration. 2.5.1 1 tb= the period of long-term buildup of activity in soil

[h). 2.4.3 l

t, =

the period of leafy garden vegetation exposure during j

the growing season [h).

2.4.3 tg = the transit time from release to receptor for fish I consumption [h).

2.4.2 th= the time between harvest of garden vegetation and l l human consumption [h).

2.4.3 t, = the transit time from release to receptor for potable water consumption [h). 2.4.2 I

0 2-47 Gen. Rev. 13 l

FNP-O-M-011 M Section of Qsfinition Inittal Use TF = the tolerance it.ctor selected to allow flexibility in the estsblishment of a practical monitor setpoint which could accommodate effluent releases

~

at concentrations higher than the ECL values stated in ,

10 CFR 20, Appendix B, Table 2, Column 2 [unitiess);

the tolerance factor must not exceed a value of 10. 3.3.2.2 Ug = the adult rate of fish consumption [kg/y). 2.4.2 Uy = the adult consumption rate for irrigated gardan vegetation (kg/y). -

2.4.2 U, = the adult drinking water consumption rate applicable to the plir.t site [L/y). 2.4.2 Y, = the areal density (agricultural productivity) of leafy garden vegetation [kg/m2 ). 2.4.3 2=

the applicable dilution factor for the receiving j water body, in the near fisld of the discharge-N structure, during the period of radioactivity release. !

[unitiess). 2.4.1 Atg = the length of time period .1, over which C ij and Fg are averaged for liquid releases [h). 2.4.1 l 1Ei = the effective removal rate for activity deposited on crop leaves [h-I). 2.4.3 l

l j=

the decay constant for radionuclide 1 [h-I). 2.4.2

{

1, = the rate constant for removal of activity from plant leaves by weathering [h-I). 2.4.3 2-48 Gen. Rev. 13

i FNP-0-M-011 !

CHAPTER 3 i

{

GASEOUS EFFLUENTS 3.1 LIMITS OF OPERATION

. I The following Limits of Operation implement requirements established by Technical Spec.'.fications section 6.0.

Terms printed in all capital letters are defined in Chapt's: 10. ,

3.1.1 Caseous Effluent Monitorino Instrumentation Control 1

In accordance with Technical Specification 6.8.3.e(i), the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3-1 shall be OPERABLE .

with their alarm / trip setpoints set to ensure that the limits of Section 3.1.2.a i

are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with section 3.3.

3.1.1.1 Applicability i

These limits apply as shown in Table 3-1.

J l i

3.1.1.2 Actions 1

] With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip j

setpoint less conservative than required by the above control, immediately i suspend the release of radioactive gaseous effluents monitored by the affected channel, declare the channel inoperable, or restore the setpoint to a value that will ensure that the limits of Section 3.1.2.a are met.

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

l This control does not affect shutdown requirements or MODE changes.

i 3.1.1.3 surveillance Requirements Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of thu CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST cperations at the frequencies shown in Table 3-2.

3-1 Gen. Rev. 13

FNP-0-M-011 ,

3.1.1.4 Basis !

l The radioactive gaseous effluent instrumentation is provided to monitor and i N

control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The i Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in ,

accordance with the methodology and parameters in Section 3.3 to ensure that the '

alarm / trip will occus prior to exceeding the limits of Section 3.1.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64-of Appendix A to 10 CFR Part 50.

b i

i I

t k

i f

i l

t

/* l b}

-2 Gen. Rev. 13 1

\

FMP-0-M-011 Table 3-1. Radioactive Gaseous Effluent Monitoring Instrumentation OPERABILITY Requirementsb Minimum Channels Instrument OPERABLE Applicability ACTION

1. Steam Jet Air Ejector Noble Cas Activity Monitor (RE-15) 1 MODES 1,2,3,4 37

2. Plant Vent Stack

a. Noble Gas Activity Monitor (RE-14 or RE-22) 1 At all times 37"

b. Iodine Sampler 1 At all times 39

c. Particulate Sampler 1 At all times 39

d. Flowrate Monitor 1 At all times 36 !

3. CASEOUS RADWASTE l TREATMENT SYSTEM Noble Gas Activity 1 Monitor (RE-14), with Alarm and Automatic

) Termination of Release 1 At all times 35

)

a. For conti.e,uous releases.

b.

All requirements in this table apply to each unit.

O 3-3 can. Rev. 13

s FNP-0-M-011 {

Table 3-1 (contd). Notation for Table 3 ACTION Statements l

i i

'l ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be i released to the environment for up to 14 days provided that prior to initiating the release: '

l 1

a. At least two independent samples of the tank's contents are 1 analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving, and (1) Verify _the manual portion of the computer input for the release rate calculations performed on the computer, or (2) Verify the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . Normal building ventilation may continue provided the flowrate is estimated once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . i ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are-taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Normal building ventilation may continue provided grab samples of this pathway once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . are taken once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed for gross activity ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, ef fluent releases via the affected l> pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 3-3.

l l

l

O I 3-4 Gen. Rev. 13 l

1 FNP-0-M-011 Table 3-2. Radioactive Gaseous Effluent Monitoring Surveillance Requirements Instrumentation Surveillance Requirementsd Instrument CHANNEL CHANNEL CHANNEL SOURCE CALIBRA- FUNCTIONAL

' CHECK CHECK TION TEST MODESC

l. Steam Jet Air Ejector a

Noble Gas Activity i Monitor (RE-15) D M Rb Qa(2) j 1,2,3,4

2. Plant Vent Stack i

' a. Noble Gas Activity Monitor RE-14 D M Rb ga(1,2) gyy RE-22 o M Rb ga(2) 371

b. Iodine Sampler W NA NA NA All

c. Particulate Sampler W NA NA NA All

d. Flowrate Monitor D NA R Q All d

a.

In addition to the basic functions of a CHANNEL FUNCTIONAL TEST (Section 1

4 10.2): !

(1) l The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic p isolation any of theoffollowing this pathway and control conditions exists:room annunciation occur if (a)

Instrument setpoint; indicates measured levels above the alarm / trip (b) Loss of control powers or 1

(c) Loss of instrument power.

(2)

The CHANNEL FUNCTIONAL TEST shall also demonstrate that control J

room annunciation occurs if any of the following conditions exists:

(a) Instrument indicates a downscale failure; or (b) t Instrument controls not set in the OPERATE mode.

b.

The initial CHANNEL CALIBRATION shall be performed using one or more of

the reference standards certified by the National Institute of Standards ar4d Technology, or using standards that have been obtained from suppliers tnat participate in measurement assurance activities with NIST. For any subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

c.

MODES in which surveillance is required. "All" means "At all times."

d.

All requirements in this table apply to each unit.

1 l

i

!O 3-5 Gen. Rev. 13

FNP-0-M-011 3.1.2 Gaseous Effluent Dose Rate Control

/~'N In accordance with Technical Specifications 6.8.3.e(iii) and 6.8.3.e(vii), the s-licensee shall conduct operations so that the dose rates due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 10-1) are limited as follows:

f a.

For noble gases: Less than or equal to a dose rate of 500 mrom/y to the total body and less than or equal to a dose rate of 3000 mrom/y to the skin, and b.

Jor Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem /y to any organ.

3.1.2.1 Applicability This limit applies at all times.

3.1.2.2 Actions With a dose rate due to radioactive material released in gaseous effluents exceeding the limit stated in Section 3.1.2, immediately decrease the release l

() rate to within the stated limit.

This control does not affect shutdown requirements or MODE changes.

3.1.2.3 Surveillance Requirements The dose rates due to radioactive materials in areas at or beyond the SITE BOUNDARY due to releases of gaseous effluents shall be determined to be within the above li'aits, in accordance with the methods and procedures in Section 3.4.1,

] by obtaining representative samples and performing analyses in accordance with l

1 the sampling and analysis program specified in Table 3-3.

3.1.2.4 Basis This control is provided to ensure that gaseous effluent dose rates will be l

maintained within the limits that historically have provided reasonable assurance l that radioactive material discharged in gaseous effluents will not result in a dose to a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside l the SITE BOUNDARY, exceeding the limits specified in Appendix I of 10 CFR Part 50, while allowing operational flexibility for effluent releases. For MEMBERS

!O 3-6 Gen. Rev. 13

$I FNP-0-M-011 OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the KEMBER OF THE PUBLIC will be suf ficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY.

The dose rate LLait for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days specifically applies to dose rates to a child via the inhalation pathway. f j

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

i I

l l

1 l

l I

i i

3-7 Gen. Rev. 13

I FNP-0-M-011 Table 3-3.

Radioactive Gaseous Waste Sampling and Analysis Program f

(

Sampling and Analysis Requirements a,b MINIMUM Gaseous DETECTABLE Minimum Type of Release Sampling Analysis CONCENTRA-Type Activity TION (MDC)

FREQUENCY FREQUENCY Analysis (uci/mL)

P PRINCIPAL GAMMA 1 E-4 Decay Tank Each Tank EMIMERS 1 Grab Sample Each Tank )

l c PRINCIPAL GAMMA Containment PC 1 E-4 Ea h Pu go EMITTERS Purge g Each Purge i

Condenser Steam Jet PRINCIPAL GAMMA 1 E-4 Air EMITTERS Mc.d.f Ejector, Grab Sample H-3 1 E-6 Plant Vent Stack i

CONTINUOUSI W8 I-131 Charcoal or 1 E-12 Charcoal or l

Silver Silver Zeolite I-133 1 E-lO Zeolite Sample l W8 PRINCIPAL GAMMA 1 E-11 CONTINUOUSS Particulate EMITTERS Sample 1

Plant Vent Stack, M Gross Alpha 1 E-11 Containment CONTINUOUSg COMPOSITE Purge Particulate Sample Q Sr-89, Sr-90 1 E-ll CONTINUOUSg COMPOSITE Particulate Sample CONTINUOUSg Noble Gas Noble Gases 1 E-6 Monitor (Gross Beta and j Gamma) i i

i l l 3-8 Gen. Rev. 13 l

l

, __ _ _ _ . . . _ . . .21

(

FNP-0-M-011 Table 3-3 (contd). Notation for Table 3-3 F

i i

a. All requirements in this table apply to each unit. Deviation from the MDC requirements 7.2.

of this table shall be reported in accordance with Section Deviation from the composite sampling requirements of this table shall be reported in accordance with Section 7.2.

t b.

3 Terms printed in all capital letters are defined in Chapter 10.

i c.

Analyses shall also be performed following shutdown from greater than or equal to 15% RATED THERMAL POWER, startup to greater than or equal to 15%

1 i

RATED THERMAL POWER, or a THERMAL POWER change exceeding 15% of the RATED

THERMAL POWER within a one-hour period.

i

d.

Tritium grab samples shall be taken from the plant vent stack at least

} once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling cKnal is flooded.

i f e.

i, Samples shall be changed at least once per 7 days and analyses shall be

' completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler).

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least.

2 days following each shutdown from greater than or equal to 15% RATED THERMAL POWER, startup to greater than or equal to 15% RATED THERMAL POWER, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in one hour, and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When

< samples collected increased for 24 by a factor ofhours

10. are analyzed, the corresponding MDC may be e

1 f.

i Tritium grab samples shall be taken at least once per 7 days from the l ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

I g.

i The ratio of the sample flowrate to the sampled stream flowrate shall be known for the time period covered by each dose or dose rate calculation J made

'3.1.4.

in accordance with controls specified in Sections 3.1.2, 3.l.3, and i

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3-9 Gen. Rev. 13 i

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d 1 FNP-0-M-011 3.1.3 Gaseous Effluent Air Dose Control In accordance with Technical Specifications 6.8.3.e(v) and 6.8.3.e(viii), the air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) shall be limited to

! the following:

a. During any calendar quarter: Less than or equal to 5 mrad for gamma !

radiation and less than or equal to 10 mrad for beta radiation, and i

b. During any calendar year Less than or equal to 10 mrad for gamma -

radiation and less than or equal to 20 mrad for beta radiation.

3.1.3.1 hpplicability This limit applies at all times.

i

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3.1.3 2 Actions i'

t With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory commission within 30 days, pursuant to Technical-Specification 6.9.2, a Special frs Report which identities the cause(s) for exceeding the limit (s); defines the

! ( ,/ corrective actions that have been taken to reduce the releases; and defines the j

proposed corrective actions to be taken to assure that subsequent releases of radioactive noble gases in gaseous effluents will be in compliance with the

1Leits of Section 3.1.3.

This control does not affect shutdown requirements or MODE changes.

i j 3.1.3.3 Surveillance Requirements '

4 Cumulative air dose contributions from noble gas radionuclides released in gaseous affluents from each unit to areas at and beyond the SITE BOUNDARY, for !

j the current calendar quarter and current calendar year, shall be determined in

] accordance with section 3.4.2 at least once per 31 days. !

4 3.1.3.4 Basis e

i This control is provided to implement the requirements of Sections II.3, III.A and IV. A of Appendix I,10 CFR Part 50. Section 3.1.3 implements the guides set forth in Section II.B of Appendix I.

i The ACTION statements in Section 3.1.3.2 provide the required operating flexibility and at the same time implement the O

3-10 4

Gen. Rev. 13

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FNP-O-M-011 guides set forth in Section IV.A of Appendix I, assuring that the releases of f radioactive material in gaseous af fluents to UNRESTRICTED AREAS will be kept "as

{ low as is reasonably achievable. " The Surveillance requirements in Section 3.1.3.3 implement the requirements in Section III. A of Appendix I, which require that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF <

THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in Section 3.4.2 for calculating the doses due to the actual releases of noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3), and Regulatory Guide 1.111 (Reference 5). The equations in Section 3.4.2 provided for determining the air doses at the SITE BOCINDARY are based upon the historical annual average atmospheric conditions.

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3-11 Gen. Rev. 13 l

FNP-0-M-011 3.1.4 Control on caseous Effluent Dese to a Member of the Public A In accordance with Technical specifications 6.8.3.e(v) and 6.8.3.e(ix), the dose

() to a KEMBER OF THE PUBLIC from I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) shall be limited to the following: '

a.

During any calendar quarter: Less than or equal to 7.5 mrom to any organ, and b.

During any calendar years Less than or equal to 15 mrom to any organ.

3.1.4.1 Applicability This limit applies at all times.

3.1.4.2 Actions with the calculated dose from the release of I-131, I-133, tritium, or radio-nuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission.within 30 days, pursuant to Technical specification 6.9.2, O a special Report which identifies the cause(s) for exceeding the limit; defines the corrective actions that have been taken to reduce the releases of radiciodines and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents; and defines proposed corrective actions to assure that subsequent releases will be in compliance with the limits stated in section 3.1.4.

This control does not affect shutdown requirements or MODE changes.

3.1.4.3 Surveillance Requirements cumulative organ dose contributions to a MEMBER OF THE PUBLIC from I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.3 at least once per 31 days.

O 3-12 Gen. Rev. 13

1 4

1 FNP-0-M-011-3.1.4.4 Basis J i

k This control is provided to implement the requirements of Section II.C, III. A and i

IV.A of Appendix I, 10 CFR Part 50.

The limits stated in Section 3.1.4 are the guides set forth in Section II.C of Appendix I. The ACTION statements in Section

, j 3.1.4.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept j "as low as is reasonably achievable." The calculational methods specified in the Surveillance Requirements of Section 3.1.4.3 implement the requirements in.

Section III. A of Appendiat I that conformance with the guides of Appendix I be '

shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestinated. The calculational methods in Section 3.4.3 for calculating the doses due to the actual releases of the subject materials are

' consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3),

and Regulatory Guide 1.131 (Reference 5). These equations provide for determining the actual doses based upon the historical annual average atmospheric

, conditions. The release specifications for radiciodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the areas at and beyond the SITE I BOUNDARY.

The pathways which wers examined in the development of these l calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy garden vegetation with subsequent j consumption by man, 3) deposition onto grassy areas where milk animals and meat

{

I producing animals graze with consumption of the milk and meat by man, and 4) {

deposition on the ground with subsequent exposure of man. '

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3-13 Gen. Rev. 13 l

k FNP-0-M-011 3.1.5 Gaseous Radwaste Treatment system Control In accordance with Technical Specification .6.8.3.e(vi), the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE.

The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used ,

to reduce radioactive materials in gaseous wastes prior to their discharge when l

the projected air doses due to gaseous effluent releases, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) would exceed 0.2 mrad for gamma radiation or O.4 mrad for beta radiation in 31 days.' The appropriate l

portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce +

radioactive materials in gaseous wastes prior to their discharge when the ,

projected doses due to gaseous effluent releases, from each reactor unit, to areas beyond the SITE BOUNDARY (see Figure 10-1) would exceed 0.3 mrom to any organ of a MEMBER OF THE PUBLIC in 31 days.

9 3.1.5.1 Applicability i

These limits apply at all times.

3.1.5.2 Actions I With gaseous waste being discharged without treatment and in excess of the limits O in Section 3.1.5, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special Report which includes the following informations i I

a.

' Identification of the inoperable equipment or subsystem and the reason for

~

inoperability, b.

Action (s) taken to restore the inoperable equipment to OPTRABLE status, and i

c.

summary description of action (s) taken to prevent a recurrence.

This control does not affect shutdown requirements or MODE changes.

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3-14 Gen. Rev. 13

FNP-0-M-011 3.1.5.3 Surveillance Requirements Doses due to gaseous releases from each unit to areas at and beyond the SITE

\ BOUNDARY shall be projected at least once per 31 days, in accordance with Section 3.5.1, when the CASEOUS RADWASTE TREATMENT SYSTEM or the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.

The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE:

a. by meeting the controls of Sections 3.1.2, and either 3.1.3 (for the GASEOUS RADWASTE TREATMENT SYSTEM) or 3.1.4 (for the VENTILATION EXHAUST TREATMENT SYSTEM), or b.

by operating the GASEOUS RADWASTE TREATMENT SYSTEM equipment and the VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 15 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

3.1.5.4 Basis The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous affluents require treatment prior to release to the environment.

The requirement that the appropriate portions of these syste.as be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."

This control implements the requirements of 10 CFR Part 50.36a, General Design

]

criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in l

Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of these systems were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.c of Appendix I, 10 CFR Part 50, for gaseous effluents.

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3.1.6 MAJOR CHANGES to the GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM Licensee-initiated MAJOR CHANGES to the GASEOUS RADICACTIVE WASTE TREATMENT SYSTEM or the VENTILATION EXHAUST TREATMENT SYSTEM: ;

a.

Shall be reported to the Nuclear Regulatory Comnission in the Annual i Radioactive Effluents Release Report for the' period in which the change was implemented, in accordance with Section 7.2.2.7.

b.

Shall become effective upon review and approval in accordance with Technical Specification 6.5.3.1.

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,1 FNP-0-M-011 3.2 GASEOUS RADWASTE TREATMENT SYSTEM C)g At the Farley Nuclear Plant, there are six designated points where radioactivity may be released to the atmosphere in gaseous discharges

the Unit 1 ani Unit 2 Plant Vent Stacks; the Unit 1 and Unit 2 Turbine Building. Vents (steam jet air ejectors); and the Unit 1 and Unit 2 Integrated Leak Rate Test (ILRT) Vents. Of these six, only four are routine release pathways, since ILRT Vent releases are performed only infrequently.

Figure 3-1 gives schematic diagrams of the Waste Gas Treatment Systems and the Ventilation Systems (Reference 7). Discharges from the two reactor units are separated, with no shared systems. In each unit, containment Purge and Waste Gas Decay Tank effluents are discharged through the respective Plant Vent, and are treated as contributions to the on-going Plant Vent CONTINUOUS release. Although Waste Gas Decay Tank effluents are released via the Plant Vent Stack, they are tracked separately and accounted for as BATCH releases.

Table 3-4 summarizes the release height and release type characteristics of the various release pathways and source streams. Chapter 8 discusses the calculation of atmospheric dispersion parameters using the ground-level and mixed-mode (i.e.,

split-wake) models.

As established in Section 3.1.1, gaseous effluent monitor setpoints are required O for the noble gas monitors on the two Plant Vents.and the two Turbine Building Vents (steam jet air ejectors). Wasta Gas Treatment System discharges are not monitored separately during release, but are sampled prior to release and are monitored by the downstream Plant Vent monitors during release. ILRT discharges are net monitored during release, but are sampled prior to release; the ILRT Vent may be assigned an appropriate allocation factor during the release period, and dose calculations may be based on estimates of the activity concentration and the ;

volume of air released. Sampling and analysis of both these release pathways must be sufficient to ensure that the gaseous effluent dose l.'mits specified in Section 3.1.3 and Section 3.1.4 are not exceeded.

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3-17 Gen. Rev. 13

l 9 FNP-0-M-011 O Waste Gas Decay Tanks i

)( 3( ) <

X 1: :: H 1;

:: n 7 8 l 1 -2 3 4 5 6 I

1r 1r 1r 1r i r ir 1r 1r Jk dk JL Jk Jh JL Jk dk U 9P if da di db

]l Waste Gas RE-13 MONITOR Compressor E PLANT VENT STACK U -1 ILRT *- - - - - - -^- Rz- 2 MONITORS

VENT

O CONTAINMENT AUX CONTAIN. SLOd.

MENT T EXHAUST ty? PLENUM RE-2 MONITOR AUXILIARY g BUILDING RE-15 MONITOR TURBINE TURRINE BUILDING BUILDING VENT (STEAM JET AIR EJECTOR) .

O Figure 3-1.

Schematic Diagram of Routine Release Sources and Release Points (Typical of Both Units) 3-18 Gen. Rev. 13

I FNP-O-M-011 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS I i

3.3.1 General Provisions Recardino Noble Gas Monitor Setooints

!O Noble gas radioactivity monitor setpoints calculated . in accordance with the

, methodology presented in this section are intended to ensure that the limits of l'

} Section 3.1.2.a are not exceeded. They will be regarded as upper bounds for the .'

} actual high alarm setpoints. That is,.a lower high alarm setpoint may be established or retained on the monitor, if desired. Intermediate level setpoints should be established at an appropriate level to give sufficient warning prior 1

to reaching the high alarm setpoint. '

4 If no release is planned for a given pathway, or if there is no detectable i 4

activity in the gaseous stream being evaluated for release, the setpoint should !

f be established as close to background as prac'tical to prevent spurious alarms,

}

t and yet alarm should a significant inadvertent release occur.

a .

Section 3.1.1 establishes the requirements for gaseous effluent monitoring instrumentation, and Section 3.2 describes . the VENTIIlTION EXHAUST TREATMENT 4

SYSTEM and the GASEOUS RADWASTE TREATMENT SYSTEM. From those sections, it can j

be seen that certain monitors are located on final release pathways, that is, i

streams that are being monitored immediately before being discharged from the plant; the setpoint methodology for these monitors is presented in Section 3.3.2.

{ other monitors are located on source streams, that is, streams that merge with i

other streams prior to passing a final monitor and being discharged; the setpoint i

methodology for these monitors is presented in Section 3.3.3.- Table 3-4 identifies which of these setpoint methodologies applies to each monitor. Some additional monitors with special setpoint requirements are discussed in Section 3.3.5, i

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, 3-19 Gen. Rev. 13

L FNP-O-M-011 Table 3-4.

Applicability of Gaseous Monitor Setpoint Methodologies Final Release. Pathways with no Monitored Source Streams Release Elevation: Ground-level Unit 1 or Unit 2 Turbine Buildino Vent '

j Release Type: CONTINUOUS j Monitor:

! Setpoint Method 1RE-15 / 2RE-15 ,

! Section 3.3.2 Maximum Flowrate: 1060 cfm (5.00 E+05 mL/s)

Unit 1 or Unit 2 TLRT Vent i Release Typer BATCH 4

Monitor None Setpoint Method: None

Maximum Flowrate Release-dependent i

i

Final Release Pathways with One or More Monitored Source Streams j Release Elevation: Mixed-Mode Unit 1 or Unit 2 Plant Vent Stack Release Type: CONTINUOUS Monitor: 1RE-14 / 2RE-14, and 4

Setpoint Method 1RE-22 / 2RE-22 Section 3.3.2 ;

Maximum Flowrate: 150,000 cfm 1 (7.08 E+07 mL/s) '

Source Stream Unit I or Unit 2 containment Purce 'I a

Release Type: CONTINUOUS Monitor: 1RE-24 / 2RE-24 j Setpoint Method:

Section 3.3.3 is optional. See Section 3.3.5.

Maximum Flowrates Release-dependent

Source Stream Release TyperUnit 1 or Unit 2 Wasta Gas Decav Tanks BATCM Monitor: None 3 Satpoint Methods None f

Maximum Flowrates Release-dependent ri75),b val e nor use to s.tpoint Calculations Ground-Level Releases: 4.87 x 100 s/m3 [S Sector] !

Mixed-Moda Releases: 1.08 x 104 s/m3 (SSE Sector) l i

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3-20 Gen. Rev. 13

4 i FNP-0-M-011 a

3.3.2 Setooint for the Final Noble Cas Monitor on Each Release Pathway 3.3.2.1 overview of Method i

Gaseous effluent radioactivity monitors are intended to alarm prior to exceeding the limits of Section 3.1.2.a. Therefore, their alarm setpoints are established '

to ensure compliance with the following equations i

4 c = the lesser of (3*1)

AG

ST

X

Rg where:

c=

the setpoint, in yci/mL, of the radioactivity monitor measuring the ,

concentration of radioactivity in the affluent line prior to release. The setpoint represents a concentration which, if i

exceeded, could result in dose rates exceeding the limits ' of Section 3.1.2.a at or beyond the SITE BOUNDARY.

1 AG =

an administrative allocation factor applied to divide the release limit among all the gaseous release pathways at the site.

SF = the safety factor selected to compensate 1

for statistical fluctuations and errors of measurement.

4 X= the noble gas concentration for the release under consideration.

l Rt= the ratio of the dose rate limit for'the total body, 500 mese/y, to the dose rate to the total body for the conditions. of the

release under consideration.

Rk= the ratio of the dose rate limit for the skin, 3000 mrea/y, to the dose rate to the skin for the conditions of the release under consideration.

Equation (3.1) shows the relationships of the critical parameters that determine the setpoint.

However, in order to apply the methodology presented in the ;

equation to a mixture of noble gas radionuclides, radionuclide-specific concentrations and dose factors must be taken into account under conditions of maximum flowrate for the release point and annual average meteorology.

s 3-21 Gen. Rev. 13

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FNP-O-M-011 2

The basic setpoint method presented below is applicable to the radioactivity monitor nearest the point of release for the release pathway. For monitors 4 k measuring the radioactivity in source streams that merge with other streams prior to subsequent monitoring and release, the modifications presented in Section

!, 3.3.3 must be applied.

r 4

3.3.2.2 Setpoint Calculation Steps Sten 1:

Determine the concentration, Xjy, of each noble gas radionuclide i in the gaseous stream v being considered for release, in accordance with the sampling and analysis requirements of Section 3.1.2. Then sum these concentrations to determine the total noble gas concentration, E X;y.

I Stoo 2:

Determine R , the ratio of the dose rate limit for the total body, g

500 mrem /y, to the total body dose rate due to noble gases detected in the release under consideration, as follows:

Rg = # !

(775)vb b [Ki

Ojy] (3.2) 2 7

)

where:

500 =

the dose rate limit for the total body, 500 mrom/y. I (176)vb = the highest annual average relative concentration at'the SITE BOUNDARY for the discharge point of release pathway v. Table 3-4 includes an indication of what release elevation is applicable to each release pathways release elevation determines the appropriate !

value of (17%)vb-Kg =

the total-body dose factor due to gamma emissions from noble gas l i

radionuclide i, in (mrom/y)/(yci/m3 ), from Table 3-5.

Q;y = the release rate of noble gas radionuclide i from the release pathway under consideration, in yci/s, calculated as the product of Xjy and f,y, where:

X;y =

the concentration of noble gas radionuclide i for the O particular release, in pCi/mL.

3-22 Gen. Rev. 13

FNP-0-M-011 f,y =

the maximum anticipated flowrate for release pathway v during the period of the release under consideration, in f mL/s.

Stoo 3:

I Determine kM , the ratio of the dose rate limit for the skin, 3000 ,

l mrom/y, to the skin dose rate due to noble gases detected in the release under consideration, as follows:

, 3000 (T[6)vb bi (ILI + 1'1d l) ' O v}

i (3*3) where:

3000 = the dose rate limit for the skin, 3000 mram/y. .

Lj = the skin dose factor due to beta emissions from noble gas radio-nuclide 1, in (mrem /y)/(yci/m3 ), from Table 3-5.

M =

the air dose factor due to ganuna emissions from noble gas radio-nuclide i, in (mrad /y)/(pci/m3 ), from Table 3-$.

i 1.1 =

the factor to convert air dose in~ mrad to skin dose in mrom.

All other terms were defined previously.

4 sten 4:

Determine the maximum noble gas radioactivity monitor setpoint con-centration.

Based on the values determined in previous steps, the radioactivity monitor setpoint for the planned release is calculated to ensure that the limits of section 3.1.2.a will not kie exceeded. Because the radioactivity monitor responds primarily to-radiatica?from noble gas radionuclides, the monitor setpoint cav is based on the concentration of all noble gases in the waste stream, as follows:

t lO 3-23 can. Rev. 13 I

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FNP-O-M-011 i ,

AG y SF * [ Xiy

R g i

AG y . sr [ X;y . Rg i

4 l

where:

e ny =

the calculated setpoint, in pci/mL, for the noble gas monitor i serving gaseous release pathway v. l t

1 AG y =

the administrative allocation factor for gasecus release pathway v,

applied to divide the rele'ase limit among all the gaseous release pathways at the site. The allocation factor may be assigned any value between 0 and 1, under the condition that the

{

sum of the allocation factors for all simultaneously active final.

release pathways at the entire plant site does not exceed 1.

Alternative methods for determination of AGy. are presented in Section 3.3.4.

SF = the safety factor selected to compensate for statistical f

o I fluctuations and errors of measurement.- The value for the safety factor must be between 0 and 1. A value of 0.5 is reasonable for gaseous releases; a more precise value may be developed if desired.

Xjy = the measured concentration of noble gas radionuclide 1 in gaseous stream v, as defined in Step 1, in pci/mL.

The values of Rt and Rk to be used in the calculation are those which were determined in Steps 2 and 3 above.

Sten 5:

Determine whether the release is permissible, as follows:

If c,y a Ei Xgy, the release is permissible. However, if env is within about 10 percent of E X it may be impractical to use this-value of c,y.

l y,

1 i

This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a -

to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to 3-24 Gen. Rev. 13

i FNP-O-M-011 adjust the allocation of the limits among the active release

[ points.

The setpoint calculations (steps 1-4) must then be repeated with parameters that reflect the modified conditions.

If env < I X y, l

the release may not be made as planned. Consider the <

alternatives discussed in che paragraph above, and calculate a new setpoint based on the results of the actions taken.

3.3.2.3 Use of the Calculated Satpoint The setpoint calculated above is in the units pci/mL. The monitor actually measures a count rate that includes background, so that the calculated setpoint must be converted accordingly:

e,, , =

( c,,,

Ey ) + B y (3.5) where:

en =

the monitor setpoint as a count rate.

Ey =

the monitor calibration factor, in count rate /(pci/mL). Monitor w calibration data for conversion between count rate and concentration may include operational data obtained from determining the monitor response to ef fluent stream concentrations measured by sample analysis.

By= the monitor background count rate. In all cases, monitor j

background must be controlled so that the monitor is capable of !

responding to concentrations in the range of the setpoint value.

Contributions to the monitor background may include any or all of the following factors: ambient background radiation, plant-related ,

radiation levels at the monitor location (which may change between i shutdown and power conditions), and internal background due to contamination of the monitor's sample chamber.

The count rate units for c,*, Ey , and By in equation (3.5) must be the same (cpm se eps).

3.3.3 Setnoints for Noble Gas Monitors on Effluent Source Strs===

3-25 Gen. Rev. 13 F

, a.. ,, , ,_-.r -i er-

FNP-0-M-011 Tcblo 3-4 licto certain goccoup rolcaco pathways as being source streams. As may be seen in the figures of Section 3.2, these are streams that merge with other streams, prior to passing a final radioactivity monitor and being released.

Unlike the final monitors, the source stream monitors measure radioactivity in affluent streams for which flow can be terminated; therefore, the source stream monitors have control logic to terminate the source stream release at the alarm setpoint. #

3.3.3.1 Setpoint of the Monitor on the Source Stream Sten 1:

Determine the concentration X;,of each noble gas radionuclide i in source stream s (in pci/mL) according to the results of its required sample analyses (see section 3.1.2].

Sten 2:

Determine r , the ratio of the dose rate limit for tha total body, g

500 mrom/y, to the total body dose rate due to noble gases detected in the source stream under consideration. Use the Xis values and c

the maximum anticipated source stream flow rate f, in equation (3.2) to determine the total body dose rate for the source stream, substituting rg for R*t i

The SITE BOUNDARY relative dispersion value used in Steps 2 and 3 for the 1( source stream is the same as the (i76)vb that applies to the respective merged stream.

4 This is because the (i?6) value is determined by the meteorology of the plant site and the physical attributes of the release 1

d point, and is unaffected by whether or not a given source stream is operating.

d a

. Cten 3:

Determine ek, the ratio of the dose rate limit for the skin, 3000 aren/y, to the skin dose rate due to noble gases detected in the source stroom under consideration. Use the X;, values and the maxismsa an'.icipated source stream flow rate f,, in equation (3.3) to determiner the skin dose rate for the source stream, substituting rk for Ng.

Stoo 4:

Determine the maximum noble gas radioactivity monitor setpoint con-centration, as follows:

\d )

3-26 Gen. Rev. 13

FNP-O-M-011 AG,

SF * [ Xg, rg c ns

the lesser of (3**)

AG, ST = [ X), rg

! /

1 where:

1 1

c as =

the calculated setpoint (in pC1/mL) fo the noble gas monitor serving gaseous source stremn s.

AG, = the administrative allocation factor applied to gaseous source stream s. For a given final release point v, the sum of all the AG s values for source streams contributing to the final release d

point must not exceed the release point's allocation factor Any, X;, = the measured concentration of noble gas radionuclide i in gaseous 1

source stream s, as defined in Step 1, in yC1/mL.

I

The values of r t and rk to be used in the calculation are those which were

determined in Steps 2 and 3 above. The safety factor, SF, was defined

( previously.

Stoo 5:

l

, Determine whether the release is permissible, as follows: l i If e ns 2 fXis, the release is permissible. However, if cas is within about 8

l l

10 percent of F X;,, it may be impractical to use this value of ca ,. i This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a 1

, to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radLoactive material, or to adjust the allocation of the limits among the active release

)

J points. The setpoint calculations (steps 1-4) must then be

repeated with parameters that reflect the modified conditions.

I If ens < p X;,, the release may not be made as planned. Consider the i

I alternatives discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.

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3-27 Gen. Rev. 13 J

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i FNP-0-M-011, 3.3.3.2 Effect on the Setpoint of the Monitor on the Merged Stream Before beginning a release from a monitored source stream, a_setpoint must be determined for the source stream monitor as presented in Section 3.3.3.1. In addition, whether or not the source stream has its own effluent monitor, the '

previously-determined maximum allowable setpoint for the downstream final monitor on the merged stream must be redetermined. This is accomplished by repeating the steps of Section 3.3.2, with the following modifications.

Modification 1: The new maximum anticipated flowrate of the merged stream is the sum of the old merged stream maximum flowrate

((f,y)old), and the maximum flowrate of the source stream being considered for release (fu)*

(fav)new (fav)old + f as (3.7)

Modification 2:

The new concentration of noble gas radionuclide i in the merged stream includes both the contribution of the merged stream without the source stream, and the source stream being censidered for release.

O (Xfy) 3,, = (fav)ola * (Xiv)old

fu*Xis (fav)new (3.8) 3.3.4 Determination of Allocation Factors, AG When simultaneous gaseous releases are conducted, an administrative allocation factor must be applied to divide the release limit among the active gaseous release pathways. This is to assure that the dose rate limit for areas at and beyond the SITE BOUNDARY (see Section 3.1.2) will not be exceeded by simultaneous releases.

The allocation factor for any pathway may be assigned any value between 0 and 1, under the following two conditions:

1.

The sumw( the allocation factors for all simultaneously-active final release paths'at the plant site may not exceed 1, 2.

The sua of the allocation factors for all simultaneously-active source screams merging into a given final release pathway may not exceed the allocation factor of that final release pathway.

O 3-28 Gen. Rev. 13

.A

FNP-0-M-011 Any of tho following throo mathods may be used to assign the allocation factors !

to the active gaseous release pathways: 1

( 1.

For ease of implementation, AGy may be equal for all release pathways 1

AG v =

(3.9) ;

where ,

I N= the number of simultaneously active gaseous release pathways. 3

2. AG y for a given release pathway raay be selected based on an estimate of 3

the portion of the total SITE BOUNDARY dose rate (from all simultaneous releases) that is contributed by the release pathway. During periods when-l a given building or release pathway is not subject to' gaseous radioactive 1

releases, it may be assigned an allocation factor of zero. )

3. AGy for a given release pathway may be selected based on a calculation of l the portion of the total SITE BOUNDARY dose rate that is contributed by-the release pathway, as follows:

i i

(X/0)p6 )C(K lOlv) 1(~'

RG, .

N (3.10)

~

[

r=1 l

( M )rb b (Kl Or) 1 where: I i

i

, =

(176)vb the annual average SITE BOUNDARY relative concentration applicable to the gaseous rslease pathway v- for which the l

allocation factor is being determined, in s/m3.

4 Kg = the total-body dose factor due to gamma emissions from noble gas radionuclide i, in (mram/y)/(pci/m3 ), from Table 3-5.

4 Q;y = the release rate of noble gas radionuclide i from release pathway v, in pci/s, calculated as the product of X;y and f,y, where:

N i

3-29 Gen. Rev. 13 i

i

.(

FNP-0-M-011 Xiy =

the concentration of noble gas radionuclide i applicable to the gaseous release pathway v for which the allocation factor is being determined, in yci/mL.

f,y =

the discharge flowrate . applicable to gaseous release pathway v for which the allocation factor. is - being '

determined, in mL/s.

=

(X76)rb the annual average SITE BOUNDARY relative concentration applicable to active. gaseous release pathway'r,'in s/m3 .

Q;, = the release rate of noble gas radionuclide i applicable to active-release pathway r, in pCL/s, calculated as'the prc, duct of X ir and far, where: ,

Kir = the concentration of noble gas radionuclide'i applicable to active gaseous release pathway r, in pCi/mL. l i

i far = - the discharge flowrate applicable to act'ive gaseous '

release pathway r, in mL/s.

O N=

the number of simultaneously active gaseous release pathways (including pathway.v that is of interest).

l NOTE: i Although equations (3.9) and (3,10) are written to illustrate the i ,

assignment of the allocation factors for. final release pathways, l {

they may also be used to assign allocation factors to the source ,

streams that merge into a given final release pathway.. -i l

e l

1 3-30 Gen. Rev. 13-h m, . _- _-. , . - . . , . , , - _ , . . _ . . . ~ ,,,4. . - , , . ,nr..- y ., ~

i FNP-0-M-011 3.3.5 SetDoints for Noble Gas Monitors with Scecial Recuirements D

The Farley Nuclear Plant operating philosophy treats the Waste Gas Decay Tank supply monitors (1/2 RE-013) and the Containment Purge monitors (1/2 RE-024) as process monitors, not effluent monitors. However, as a matter of information, e the following may be noted regarding their setpoints:

0 For 1/2 RE-Ol3, the alarm setpoint should be based on a concentration equivalent to no more than the Technical Specification limit for the maximum curie content of a waste Gas Decay Tank, In converting the curie limit to an equivalent concentration at the location of RE-Ol3, the maximum allowable Waste Gas Decay Tank pressure should be used.

O For 1/2 RE-024, the alarm setpoint concentration may be arrived at in either of two ways. In the first method, the maximum .setpoint concentration established by the Technical Specifications may be used.

Alternatively, to provide early detection and termination of an abnormally high containment purge release, the [ lower] detpoint concentration calculated according to Section 3.3.3 may be used.

3.3.6 Setooints for Particulate and Iodine Monitors

( In accordance with Section 5.1.1 of NRC NUREG-0133 (Reference 1), the effluent controls of Section 3.1.1 do not require that the CDCM establish setpoint calculation methods for particulate and iodine monitors.

it ..

3-31 Gen. Rev. 13

I FNP-O-M-011 3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS i

l i r 3.4.1 Dose Rates at and Beyond the Site Boundarv ~

l Because the done rate limits for areas at and beyond the SITE specified in Section 3.1.2 are site limits applicable at any instant in time, the summations extend over all simultaneously active gaseous final release pathways at the plant '

site.

Table 3-4 identifies the gaseous final release pathways at the plant site, and indicates the (X7Q)vb value for each.

3.4.1.1 Dose Rates Due to Noble Gases For the purpose of implementing the controls of Section 3.1.2.a, the, dose rates due to noble gas radionuclides in areas at or beyond the SITE BOUNDARY, due to releases of gaseous affluents, shall be calculated as follows:

For total body dose rates:

DR g =

{vf(Y[Q)vh t

b(A i

l Ol v) i (3.11)

For skin dose rates:

i DRk b (X/Q)vb b [(Li + 1 1M i) D iv) (3.12) i where:

l s

DRt= the total body dose rate at the time of the release, in mrem /y.

DRk= the skin dose rate at the time of the release, in mram/y. L i

Q;y = the release rate of noble gas radionuclide i, in yCi/s, equal to the product of f ev and Xjy, where:

ftv = the actual average flowrate for release pathway v during the period of the release, in mL/s.

All other terms were defined previously.

I t

3-32 Gen. Rev. 13 !

l

! j

FNP-0-M-011 3.4.1.2 I Do30 Ratos Due to Iodine-131, Iodine-133, Tritium, and l Radionuclides in Particulate Form with Half-Lives Greater than 8 Days For the purpose of implementing the controls of Section 3.1.2.b, the dose rates due to Iodine-131, Iodine-133, tritium .d all radionuclides in particulate form with half-lives greater than 8 days, in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows:

DR o =

Ev tRTC)vb E Pjo 0,$y (3.13) i '

where \

l DRo= the dose rate to organ o at the time of the release, in mram/y. ;

I Pjo = the site-specific dose factor for radionuclide i and organ o, 4

in 3

(mrom/y) /(yci/m ) . Since the dose rate limits specified in Section 3.1.2.b apply only to the child age group exposed to the inhalation i pathway, the values of Pjo may be obtained from Table 3-9, "R f0#

aid Inhalation Pathway, Child Age Group." !

( Q fy=

the release rate of radionuclide i from gaseous release pathway v, in pC1/s. For the purpose of implementing the controls of Section 3.1.2.b, only I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be I included in this calculation.

All other terms were defined previously.

3.4.2 Noble Can Air Dose at or Beyond Site Boundarv For the purpose of implementing the controle of Section 3.1.3, air doses in areas i at or beyond' the SITE BOUNDARY due to releases of noble gases from each unit '

shall be calculated as follows (adapted from Reference 1, page 28, by including only long-term releases):

P Dg = 3.17 x 10'I {v (y7g)4 { Ng . djy' (3.14) i i \

3-33 Gen. Rev. 13 i

1 i

i FNP-0-M-011 0 =

7 3.17 x 10-0 { ()(fd) vb [ #1*dv i (3.15). l i

v l where: r 3.17 x 10'8 = a units conversion ' f actor: 1 y/(3.15 x 10 7 s).

l Dg = the air dose due to beta emissions from noble _ gas radionuclides, l t

in mrad.

Dy =

the air dose due to gamma emissions from noble gas radionuclides, in mrad.

N =

the air dose factor due to beta emissions from noble gas radio- I nuclide i, in (mrad /y)/(pci/m 3), from Table 3-5.

Mi= the air dose factor due to gamma emissions from noble gas radio-l l nuclide 1, in (mrad /y)/(yci/m3 ), from Table 3-5.

1 6;y =

the cumulative release of noble gas radionuclide i from release pathway v, in yci, during the period of interest.

l All other terms were defined previously.

Because the air dose limit is on a per-reactor-unit basis, the summations extend <

over all gaseous final release pathways for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned to the two units in any reasonable manner, provided that all activity released via the particular shared release pathway is apportioned to one or the other unit.

The gaseous final release pathways at the plant site, and the (X'/6)vb for each, 1 are idesntified in Table 3-4.

l l

! l l

l l

3-34 Gen. Rev. 13 l i

FNP-0-M-011 Table 3-5.

Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Cases l

l y - Body (K) $ - Skin (L) y - Air (H) $ - Air (N)

(mrom/y) per (mrem /y) per (mrad /y) per (mrad /y) per (yci/m 3) (yci/m 3) (yci/m 3) (yci/m 3)

{

l Kr-83m 7.56 E-02 0.00 E+00 1.93 E+01 2.89 E+02 Kr-85m 1.17 E+03 1.46 E+03 1.23 E+03 1.97 E+03 j Kr-85 1.61 E+01 1.34 E+03 1.72 E+01 1.95 E+03 Kr-87 5.92 E+03 9.73 E+03 l 6.17 E+03 1.03 E+04 j Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 Kr-89' 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 l Kr-90 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03 Xa-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xe-133m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 Xe-133 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03 Xe-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+02 Xe-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 Xe-137 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04 Xe-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 Ar-41 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 All withvalues in this table were obtained from Reference 3 units converted. (Table B-1),

O 3-35 Gen. Rev. 13

I J FNP-0-M-011

Table 3-6.

i Dose Factors for Exposure to Direct Radiation from Noble Gases in l

an Elevated Finite Plume i

i 1

i i

i 1

If i

j ,

i The contents of this table are not applicable to the Farley Nuclear Plant, i

l l

i i

1, i

1 i

i 1

1 i

l i

a i'

l l

1 I .

t uses -

3-36 Gen. Rev. 13

i l

FNP-0-M-011 3.4.3 Dose to a Member of the Public at or Bevond Site Boundarv s

The dose received by an individual due to gaseous releases from each reactor

\ .

unit, to areas at or beyond the SITE BOUNDARY, depends on the individual's location, age group, and exposure pathways. The HEMBER OF THE PUBLIC expected to receive the highest dose in the plant vicinity is referred to as the controlling receptor. The dosimetrically-significant attributes of the currently-defined controlling receptor are presented in Table 3-7.

Domes to a member of the public due to gaseous releases of I-131, I-133, tritium, and all radionuclides in particulate form from each unit shall be calculated as follows (equation adapted from Reference 1, page 29, by considering only long-term releases):

D;a

= 3.17 x 10~E [ [Raip) E H vip * $sv '

(

EO p i v where:

Dja = the dose to organ j of an individual in age group a, due to gaseous releases of I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than 8' days, in mrom.

3.17 x 10-8 = a units conversion factor: 1 y/(3.15 x 107 s).

Raipj = the site-specific dose factor for age group a, radionuclide 1, e exposure pathway p, and organ j. For.the purpose of implementing the controls of Section 3.1.4, the exposure pathways applicable to calculating the dose to the currently-defined controlling receptor are included in Table 3-7; values of R,;p; for each exposure pathway and radionuclide applicable to calculations of dose to the controlling receptor are listed in Table 3-8 through Table 3-11.

A detailed discussion of the methods and parameters used for calculating Rapj j for the plant site is presented in chapter 9.

l That information may be used for recalculating the R,jp j values if l

the underlying parameters change, or for calculating Raipj values

for special radionuclides and age groups when performing the assessments discussed in Section 3.4.4 below.

3-37 Gen. Rev. 13

i

.C FNP-0-M-011..

t.

j. W yp= -the annual average relative '. dispersion or . ' deposition ; at- thei location of the controlling receptor, for release pathway v, as.

j t appropriate to exposure pathway p and radionuclide 1.

. j For all tritium pathways, ~ and for 1 the '-inhalation . of . any radio-

[ nuclide Wy ;p is (176)yp,;;the annual average' relat'ive dispersion =

f -factor'for release pathway.v,' at the location of'the controlling.

i 3

receptor (a/m ) . - Forjthe ground-plane exposure pathway, and-for all ' ingestion-related. pathways for'l radionuclides. otherJ:than $

f tritium:~ - W y ;p 'i s '(676)yp, the annual average relative deposition !

l factor for release pathway v,"at theflocation of the controlling i receptor -(m-2).

Values'Lof".(176fyp and 157_6)yp - for : use : in -

i calculating the dose to the currently-defined controlling receptor -

j are included in Table 3-7.

i i

. i 6fy= the cumulative. release of radionuclide'i~from. release pathway 1v,. '

i j during ,the period of- interest (yci). - For ' the : purpose .of-

~

j implementing the controls of H Section -3.1.4, f.only:. I-131, 'I-133, 1

! tritium, and all radionuclides in particulate form with half-lives - I l

greater than 8 days should be included in this calculation. In-any l dose assessment using the methods of this sub-sections only radio-nuclides detectable above background in their respective samples '

l should be included in the calculation.

3 ,

e ,

4 l

)

{ Because the member of the public dose limit is on a .. per-unit ' basis, the' '

j summations extend over all gaseous final release pathways for .a given 'unle.

l j For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned between the two j

units in any reasonable manner, provided that all activity released from 'the j plant site is apportioned to one unit or the other.

i 1, . Tne gaseous final release pathways at the plant site, and the release elevation j for each, are identified in Table 3-4.

i 'I i

1 l i i

i !

i 1

I

] 3-38 Gen. Rev. 13 ?

4 s

. i i

+ e - .. ,. a.a,, , m me.,,,a , - , , , , ..-e- p .p -w 4 a,g, I

b FMP-0-M-011 Table 3-7. Attributes of the controlling Receptor '

D The locations of members of the public in the vicinity of the plant site, and the exposure pathways associated with those locations, are determined in the Annual Land Use census. Dispersion and deposition values were calculated based on site meteorological data collected for the years 1971 through 1975.

Based on the Land Use Census of June 7, 1991, the current controlling receptor for the plant site is described as follows.

Sector: SW Distance: 1.2 miles Ace Groues Child Exoosure Pathways: Ground Plane Inhalation .

Garden Vegetation Grass / Cow / Heat Discersion Factors (X76) yp s Ground-Level discharge points: 8.74 x 10'0 s/m 3 l Mixed-Mode discharge points: 8.03 x 10'7 s/m 3 Decosition Factors (676) :

f Ground-Level discharge points: 2. 64 x 10-8 m*2 Mixed-Mode discharge points: 1.05 x 10-8 m -2 l

This D/Q location factorsrepresents the residence in the vicinity of the FNP.with the highest annual average X/Q and The referenced Land Use Census identified the plant no locations site; thus, where animals are maintained for milk within 5 miles of it is very unlikely that any real dairy location (which would be beyond 5 miles) would have a higher potential dose impact than the real residence location selected.

O 3-39 Gen. Rev. 13

FNP-O-M-011 3.4.4 Dose calculations to Suecort other Reauirements Case 1: Under Technical Specification 6.6.1, a radiological impact assass-w ment may be required to support evaluation of a reportable event.

Dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the dispersion and deposition parameters ((X/Q) and (D/Q)) for the period covered by the report, and using the appropriate pathway dose factors (Raipj) for the receptor of interest. Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8.

The values of R aipj presented in Table 3-8 through' Table 3-11 are applicable only to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor, Rjj ap values applicable to that receptor must first be calculated. ~ Methods and parameters for calculating R,j p; for radionuclides and age groups other than those required in Section 3.4.3 are presented in Chapter 9. When calculating Raipj for evaluation of an event, pathway and usage factors specific to the receptor involved in the event may me used in place of the values in Chapter 9, if the specific values arv known.

case 2:

A dose calculation is required to evaluate the results of the Land Use census, under the provisions of Section 4.1.2.

In the event that the Land Use Census reveals that exposure pathways have changed at previously-identified locations, - or if new locations are identified, it may be necessary to calculate doses at two or more locations to determine which should be designated as the controlling receptor. Such dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the annual average dispersion and depositi'.,n values [(X7D) and (676)} for the locations of interest, and using the agropriate pathway dose factors (Raipj) for the receptors of interes::.

9 Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8. The values of Raipj Presented in Table 3-8 through Table 3-11 are applicable only to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor, Raipj values applicable to that receptor must first be calculated.

Methods and parameters for calculating Raipj for radionuclides and age 3-40 Gen. Rev. 13

J l

FNP-0-M-011 1

groups other than those required in Section 3.4.3 are presented in chapter 9.

J D

j V came 3: Under Section I i

5.2, a dose calculation is required to support determination of total dose to a receptor of age group other than that currently defined as the controlling receptor. j i

5 t Dose calculations shall be performed using the equations in Section 3.4.3, J j

using the dispersion and deposition parameters defined in Table 3-7 for-

[

the controlling receptor, but substituting the appropriate pathway dose -

factors (Raipj) for the receptor age group'of interest.

.i j

j The values of R,jp j presented in Table 3-8 through Table 3-11 are applicable ,

' only to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor age group, Raipj values applicable to that receptor must first be calculated. Methode and parameters for calculating R,j pj for radionuclides and age groups other than those required in Section 3.4.3 are presented-in Chapter 9.

4 t

k O}

NJ J

(

4 i

i 1

1 e

4 g$.

f.

, : A

l 1

1 l

i 5

a 3o j

i 3-41 Gen. Rev. 13 j

l i

1 E

FNP-0-M-011 Table 3-8. Rapj j for Grour.d Plane Pathway, All Age Groups

-c Nuclide T. Body Skin 3 H-3 0.00 0.00 i

Cr-51 4.66E+06 5.51E+06 2

Mn-54 1.39E+09 1.63E+09 Fe-55 0.00 0.00 '

} Fe-59 2.73E+08 3.21E+08

Co-58 3.79E+08 4.44E+08 co-60 2.15E+10 2.53E+10

Ni-63 0.00 0.00 Zn-65 7.47E+08 8.59E+08 Rb-86 8.99E+06 1.03E+07 Sr-89 2.16E+04 2.51E+04 l Sr-90 0.00 0.00 Y-91 1.07E+06 1.21E+06 Zr-95 2.45E+08 2.84E+08 Nb-95 1.37E+08 1.61E<08 Ru-103 1.08E+08 1.26E+08

, Ru-106 4.22E+08 5.07E+08 Ag-110m 3.44E+09 4.01E+09 Sb-124 5.98E+08 6.90E+d8 1

Sb-125 2.34E+09 2.64E+09 4

Te-125m 1.55E+06 2.13E+06 Te-127m 4 9.16E+04 1.08E+05 4

Te-129m 1.98E+07 2.31E+07 1 I-131 1.72E+07 2.09E+07 4

I-133 2.45E+06 2.98E+06 Cs-134 6.86E+09 8.00E+09

co-136 1.51E+08 1.71E+08 i

Co-137 1.03E+10 1.20E+10 Ba-140 2.05E+07 2.35E+07 Co-141 1.37E+07 1.54E+07

Co-144 6.95E+07 8.04E+07 j Pr-143 0.00 0.00 Nd-147 4

8.39E+06 1.01E+07 i 1. Units are m2 .(arem/yr)/(pci/s).

2. The values in the Total Body column also apply to the Bone, 1 3.

Liver, Thyroid, Kidney, Lung, and GI-LLI organs.

This table also supports the calculations of Section 6.2.

1 3-42 Gen. Rev. 13 4

l

)

FNP-O-M-011 Tcble 3-9. Rapj j for Inhalation Pathway, Child Age Group a

) Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI l

H-3 0.00 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 i 3 Cr-51 0.00 0.00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 l

Mn-54 0.00 4.29E+04 9.51E+03 0.00 1.00E+04 1.58E+06 2.29E+04 Fe-55 4.74E+04 2.52E+04 7.77E+03 0.00 0.00 1.11E+05 2.87E+03 Fe-59 2.07E+04 3.34E+04 1.67E+04 0.00 0.00 1.27E+06 7.07E+04 co-58 0.00 1.77E+03 3.16E+03 0.00 0.00 1.11E+06 3.44E+04 Co-60 0.00 1.31E+04 2.26E+04 0.00 0.00 7.07E+06 9.62E+04 Ni-63 8.21E+05 4.63E+04 2.80E+04 0.00 0.00 2.75E+05 6.33E+03

, Zn-65 4.26E+04 1.13E+05 7.03E+04 0.00 7.14E+04 9.95E+05

' 1.63E+04 Rb-86 0.00 1.98E+05 1.14E+05 0.00 0.00 0.00 7.99E+03 Sr-89 5.99E+05 0.00 1.72E+04 0.00 0.00 2.16E+06 1.67E+05 Sr-90 1.01E+08 0.00 6.44E+06 0.00 0.00 1.48E+07

)

3.43E+05 i Y-91 9.14E+05 0.00 2.44E+04 0.00 0.00 2.63E+06

{

1.84E+05 {

] Zr-95 1.90E+05 4.18E+04 3.70E+04 0.00 5.96E+04 2.23E+06 6.11E+04

Nb-95 2.35E+04 9.18E+03 6.55E+03 0.00 8.62E+03 6.14E+05 {

" 3.70E+04 ;

Ru-103 2.79E+03 0.00 1.07E+03 0.00 7.03E+03 6.62E+05 4.48E+04

) Ru-106 1.36E+05 0.00 1.69E+04 0.00 1.84E+05 1.43E+07

' 4.29E+05 Ag-110m 1.69E+04 1.14E+04 9.14E+03 0.00 2.12E+04 5.48E+06 1.00E+05

$b-124 0.00 0.00 0.00 0.00 0.00 i

0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 6.73E+03 2.33E+03 9.14E+02 1.92E+03 0.00 4.77E+05 3.38E+04 Te-127m 2.49E+04 8.55E+03 3.02E+03 6.07E+03 6.36E+04 1.48E+06 7.14E+04

] Te-129m 1.92E+04 6.85E+03 3.04E+03 6.33E+03 5.03E+04 1.76E+06 1.82E+05 I-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 0.00 2.84E+03

) 1-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 0.00 5.48E+03 Cs-134 6.51E+05 1.01E+06 2.25E+05 0.00 3.30E+05 1.21E+05 3.85E+03 !

Cs-136 6.51E+04 1.71E+05 1.16E+05 0.00 9.55E+04

' 1.45E+04 4.18E+03 Cs-137 9.07E+05 8.25E+05 1.28E+05 0.00 2.82E+05 1.04E+05 3.62E+03 Ba-140 7.40E+04 6.48E+01 4.33E+03 0.00 2.11E+01 1.74E+06 1.02E+05

, co-141 3.92B+04 1.95E+04 2.90E+03 0.00 8.55E+03 5.44E+05 5.66E+04 co-144 6. 773+06 2.12E+06 3.61E+05 0.00 1.17E+06 1.20E+07 3.89E+05

Pr-143 1.85E+04 5.55E+03 9.14E+02 0.00 3.00E+03 4.33E+05 9.73E+04 Nd-147 1.083+04 4.73E+03 6.81E+02 0.00 4.81E+03 3.28E+05 8.21E+04 i

Units are (mram/yr)/(yci/m 3 ) for all radionuclides.

4 4

4

/'. - !

i 3-43 Gen. Rev. 13

FNP-O-M-011 Tcblo 3-10. R aid for Cow Heat Pathway, Child Age Group I j

l i

) Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 Cr-51 0.00 0.00 8.79E+03 4.88E+03 1.33E+03 8.91E+03 4.66E+05 Mn-54 0.00 8.01E+06 2.13E+06 0.00 2.25E+06 0.00 6.72E+06 Fe-55 4.57E+08 2.42E+08 7.51E+07 0.00 i

0.00 1.37E+08 4.49E+07 '

Fe-59 3.76E+08 6.09E+08 3.03E+08 0.00 0.00 1.77E+08 6.34E+08 Co-58 0.00 1.64E+07 5.02E+07 0.00 0.00 0.00 9.58E+07 Co-60 0.00 6.93E+07 2.04E+08 0.00 0.00 0.00 3.84E+08 Ni-63 2.91E+10 1.56E+09 9.91E+08 0.00 0.00 0.00 1.05E+08 Zn-65 3.75E+08 1.00E+09 6.22E+08 0,.00 6.30E+08 0.00 1.76E+08 Rb-86 0.00 5.77E+08 3.55E+08 0.00 0.00- 0.00 3.71E+07 Sr-89 4.82E+08 0.00 1.38E+07 0.00 0.00 0.00 1.87E+07 Sr-90 1.04E+10 0.00 2.64E+09 0.00 0.00 )

0.00 1.40E+08 Y-91 1.80E+06 0.00 4.82E+04 0.00 0.00 0.00 2.40E+08 Zr.-95 2.66E+06 5.85E+05 5.21E+05 0.00 8.38E+05 Nb-95 0.00 6.11E+08 3.10E+06 1.21E+06 8.62E+05 0.00 1.13E+06 0.00 2.23E+09 Ru-103 1.55E+08 0.00 5.96E+07 0.00 3.90E+08 0.00 4.01E+09 Ru-106 j 4.44E+09 0.00 5.54E+08 0.00 5.99E+09 Ag-110m 0.00 6.90E+10 !

8.39E+06 5.67E+06 4.53E+06 0.00 1.06E+07 Sb-124 0.00 6.74E+08 0.00 0.00 0.00

(' Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00

\--

Te-125m 5.69E+08 1.54E+08 7.59E+07 0.00 0.00 1.60E+08 0.00 0.30 5.49E+08 Te-127m 1.77E+09 4.78E+08 2.11E+08 4.24E+08 5.06E+09 0.03 1.44E+09 Te-129m 1.79E+09 5.00E+08 2.78E+08 5.77E+08 5.26E+09 0.00 2.18E+09 I-131 1.65E+07 1.66E+07 9.46E+06 5.50E+09 2.73E+07 0.00 1.48E+06 I-133 5.67E-01 7.02E-01 2.66E-01 1.30E+02 1.17E+00 0.00 2.83E-01 cs-134 9.22E+08 1.51E+09 3.19E+08 0.00 4.69E+08 Cs-136 1.68E+08 8.16E+06 !

1.62E+07 4.46E+07 2.88E+07 0.00 2.37E+07 3.54E+06 1.57E+06 Cs-137 1.33E+09 1.28E+09 1.88E+08 0.00 4.16E+08 1.50E+08 7.99E+06 Ba-140 4.38E+07 j 3.84E+04 2.56E+06 0.00 1.25E+04 2.29E+04 co-141 2.22E+07 2.22E+04 1.11E+04 1.64E+03 0.00 4.86E+03 Ce-144 0.00 1.3BE+07 2.32E+06 7.26E+05 1.24E+05 0.00 4.02E+05 0.00 1.89E+08 Pr-143 3.34E+04 1.00E+04 1.66E+03 0.00 5.43E+03 Nd-147 0.00 3.60E+07 1.17E+04 9.47E+03 7.33E+02 0.00 5.19E+03 0.00 1.50E+07 Units are (mrom/yr)/(yci/m3 ) for tritium, and m2-(mrom/yr)/(pci/s) for all other radionuclides. i

)

(~'/

s-3-44 Gen. Rev. 13 1

s s

FNP-0-M-011 Table 3-11. Rgpj for Garden Vegetation Pathway, Child Age Group O Nuclide Bone Liver T. Body Thyroid Kidney- Lung GI-LLI H-3 0.00 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 Cr-51 0.00 0.00 1.17E+05 6.50E+04 1.78E+04 1.19E+05 '

6.21E+06 Mn-54 0.00 6.65E+08 1.77E+08 0.00 1.86E+08 0.00 5.58E+08 Fe-55 8.01E+08 4.25E+08 1.32E+08 0.00 0.00 2.40E+08 7.87E+07 Fe-59 3.98E+08 6.43E+08 3.20E+08 0.00 0.00 1.86E+08 6.70E+08 Co-58 0.00 6.44E+07 1.97E+08 0.00 0.00 3

0.00 3.76E+08 '

Co-60 0.00 3.78E+08 1.12E+09 0.00 0.00 0.00 2.10E+09

Ni-63 3.95E+10 2.11E+09 1.34E+09 0.00 0.00 0.00 1.42E+08 i Zn-65 8.13E+08 2.16E+09 1.35E+09 0.00 1.36E+09 '

0.00 3.80E+08 Rb-86 0.00 4.52E+08 2.78E+08 0.00 0.00 0.00 2.91E+07 Sr-89 3.60E+10 0.00 1.03E+09 0.00 0.00 0.00 1.39E+09 Sr-90 1.24E+12 0.00 3.15E+11 0.00 0.00 0.00 1.67E+10 Y-91 1.86E+07 0.00 4.99E+05 0.00 0.00 0.00 2.48E+09 Zr-95 3.86E+06 8.48E+05 7.55E+05 0.00 1.21E+06 0.00 8.85E+08 1

Nb-95 4.10E+05 1.60E+05 1.14E+05 0.00 1.50E+05 0.00 2.965+08 ;

Ru-103 1.53E+07 0.00 5.90E+06 0.00 3.86E+07 0.00 3.97E+08 Ru-106 7.45E+08 0.00 9.30E+07 0.00 1.01E+09 0.00 1.16E+10 Ag-110m 3.21E+07 2.17E+07 1.73E+07 O Sb-124 sb-125 0.00 0.00 0.00 0.00 4.04E+07 0.00 0.00 0.00 2.58E+09 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 3.51E+08 9.50E+07 4.67E+07 9.84E+07 0.00 0.00 3.38E+08 Te-127m 1.32E+09 3.56E+08 1.57E+08 3.16E+08 3.77E+09 0.00 1.07E+09 Te-129m 8.41E+08 2.35E+08 1.31E+08 2.71E+08 2.47E+09 0.00 1.03E+09 I-131 1.43E+08 1.44E+08 8.17E+07 4.75E+10 2.36E+08 0.00 1.28E+07 I-133 3.53E+06 4.373+06 1.65E+06 8.11E+08 7.28E+06 0.00 1.76E+06 Co-134 1.60E+10 2.63E+10 5.55E+09 0.00 8.15E+09 2.93E+09 1.42E+08 Cs-136 8.24E+07 2.275+08 1.47E+08 0.00 1.21E+08 1.80E+07 7.96E+06 Co-137 2.395+10 2.293+10 3.38E+09 0.00 7.46E+09 2.68E+09 1.43E+08 Ba-140 2.773+08 2.42E+05 1.61E+07 0.00 7.89E+04 1.45E+05 1.40E+08 Co-141 6.565+05 3.273+05 4.86E+04 0.00 1.43E+05 0.00 4.08E+08 Co-144 1.275+08 3.983+07 6.78E+06 0.00 2.21E+07 0.00 1.04E+10 Pr-143 1.465+05 4.373+04 7.23E+03 -0.00 2.373+04 0.00 1.57E+08 Nd-147 7.15E+04 5.793+04 4.48E+03 0.00 3.185+04 0.00 9.17E+07 I

,1 3

Units are (ares /yr)/(yC1/m ) for tritium, and m2 -(ares /yr)/(yci/s) for all other radionuclides.

O 3-45 - Gen. Rev. 13 a--_a -

.vw-aa 1,n.,w ,w-,,,,,.---.s.,v-,,,,a%-.,-,,,n,w .w,w..,, y , ---g,.i,#my,,,.f.y..,,._,,%., ,,p,..-. mpy,7s_, 9 5,ag e a y.--y, .e 9 p .rsg, ,g., ., oy9

l FNP-O-M-011

3.5 GASEOUS EFFLUENT DOSE PROJECTIONS

( 3.5.1 Thirtv-One Day Dese Proiections 1

In order to meet the requirements of the limit for operation of the gaseous l j

radweste treatment system (see Section 3.1.5), dose projections must be made at j least once each 31 days; this applies during periods in which a discharge to {

areas at or beyond the SITE BOUNDARY of gaseous effluents containing radioactive i materials occurs or is expected. {

Projected 31-day air doses and doses to individuals due to gaseous effluents may be determined as follows:

)

i For air doses:

)

1 i

= x 31 + Dg, Dgp

'D ' (3.17) i D yp =

x 31 + D, 7

j l

For individual doses Dop = x 31 + D, (3*18) o l

where: I Dgp = the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases.

Dge = the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current q

,uarter, plus the release under consideration.

Dg, = sthe ~ anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, Dg, may be set to zero.

D 9= the projected air dose due to gamma emissions from noble gases for s the next 31 days of gaseous releases.

3-46 Gen. Rev. 13

1 FNP-0-M-011 D p= the cumulative air dose due to gamma emissions from noble -gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

1 D,=

y the anticipated air dose due to gamma emissions from noble gas i

releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine' ,

gaseous effluents are anticipated, Dy, may be set to zero.

a e

b

D op = the projected dose to the total body or organ o, due to releases of I-131, I-133, tritium, and particulates for the next 31 days of gaseous releases, i

Doc = the cumulative dose to the total body or organ o, due to releases-of I-131, I-133, tritium, and particulates that have occurred in '

the elapsed portion of the current quarter, plus the release under -

consideration.

s D,=

o the anticipated dose to the total. body or organ o, due to releases of I-131, I-133, tritium, and particulates, contributed by any-i

(

planned activities during the next 31-day period, if those '

. activities will result in gaseous releases that are in addition to- i routine gaseous affluents. If only routine gaseous affluents are t anticipated, Dos may be set to zero.

i i t= the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under I consideration (even if the release continues into the next i 1

quarter).

i

, 3.5.2 Dose Proisetions for Snecific Releases 1

Dose projections may be performed for a particular release by performing ~a pre-release dose calculation assuming that the planned release will proceed as anticipated.

I For air dose and individual dose projections due to gaseous effluent releases, follow.the methodology of Section 3.4, using sample analysis results for the gaseous stream to be released, and parameter values expected to exist during the release period.

i I

4 3-47 Gen. Rev. 13 i

- - w v9

FNP-O-M-Oli 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS l

IRIID Definition Section of Initial Use

{

AG =

the administrative allocation factor for gaseous streams, applied to divide the gaseous release limit among all the release pathways [unitiess). 3.3.2.1 AG, = the administrative allocation factor for gaseous i source stream s, applied to divide the gaseous release limit among all the release pathways I

[unitiess). 3.3.3 AGy = the administrative allocation ' f actor for gaseous !

release pathway v, applied to divide the gaseous ,

release limit among all the release pathways '

[unitiess). 3.3.2.2 e=

the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the affluent line prior to release [pci/mL). 3.3.2.1 cns = the calculated noble gas effluent monitor setpoint for gaseous source stream s [pci/mL). 3.3.3 e ny =

the calculated noble gas effluent monitor setpoint for release pathway v [ Ci/mL). 3.3.2.2 Dj, = the dose to organ j of an individual in age group a, due to gaseous releases of I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days [mrom). 3.4.3 i

1 l

D,=

o the anticipated dose to organ o due to releases of i non-noble-gae radionuclides, contributed by any planned activities during the next 31-day period (area). 3.5.1 D og =

the cumulative dose to organ o due to releases of non-noble-gas radionuclides that have occurred in the l-elspeed portion of the current quarters plus the release under consideration (mrsm). 3.5.1 m

3-48 Gen. Rev. 13 i

l

k FNP-0-M-011 IgIm Section of Definitien Initial Use

' O D op = the projected dose to organ o due to the next 31 days of gaseous releases of non-noble-gas radionuclides (mrem). 3.5.1 Dg = the air dose due to beta emissions from noble gas radionuclides [ mrad). 3.4.2 Dg, = the anticipated air dose due to beta emissions from noble gas releases, . contributed by any planned activities during the next 31-day period (mrad). 3.5.1 Dge = the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [ mrad). 3.5.1 Dgp = the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases O Dy =

[ mrad). 3.5.1 the air dose due to gamma emissions from noble gas radionuclides (mrad). 3.4.2 ,

D7 , =. the anticipated air dose due to gamma emissions from noble gas releases, contributed by any planned activities during the next 31-day period (mrad). 3.5.1 D p= the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release ;

under consideration [ mrad). 3.5.1 '

D !

9= the projected air dose due to gamma emissions from !

noble gases, for the next 31 days of gaseous releases

[ mrad). 3.5.1 1

l O i 3-49 Gen. Rev. 13 l

FNP-O-M-Oli Igm section of Definition Initial Use (D76)yp = the annual average relative deposition factor for release pathway v, at the location of the controlling receptor, from Table 3-7 (m'2]. 3.4.3 DRk= the skin dose rate at the time of the release (mrom/y].

3.4.1.1 DRo= the dose rate to organ o at the time of the release (mrom/y]. 3.4.1.2 DRt= the total body dose rate at the time of the release (mrom/y).

3.4.1.1 f,y = the maximum anticipated actual discharge flowrate for release pathway v during the period of the planned release (mL/s]. 3.3.2.2 f,, = the maximum anticipated actual discharge flowrate for gaseous source stream s during the period of the.

planned release (mL/s). 3.3.3 Kg = the total body dose factor due to gamma emissions from noble gas radionuclide 1, from Table 3-5

((mrom/y)/(pci/m3) }. 3.3.2.2 L; = .

the skin dose factor due to beta emissions from noble gas radionuclide i, from Table 3-5 L(ares /y)/(pci/m ) }.

3 3.3.2.2 Mj = the air dose factor due to gamma emissions from noble

{

i gas radionuclide i, from Table '3-5

((arad/y)/(pci/m 3)]. 3.4.2 N= the number of simultaneously active gaseous release !

J pathways (unitless). 3.3.4 O

3-50 Gen. Rev. 13

I PNP-0-M-011 4

i 1923 Definition Section of Initial Use 4

i Nj = the air dose factor due to beta emissions from noble

>\

gas radionuclide 1, from Table 3-5

{ ((mrad /y)/(pci/m3 )). 3.4.2 Pjo = the site-specific dose factor for radionuclide i

! (I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) and organ o. The values of Pjo are equal to the site-specific Raipj values presented in Table 3-9 j ((mrom/y)/( Ci/m 3)].

3.4.1.2

}

Q;y =

the release rate of. noble gas radionuclide i from

!, release pathway v during the period of interest

[pci/s).

3.3.2.2 Q fy=

the release rate of radionuclide 1 (I-131, I-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) from gaseous release pathway v during the period of interest [pci/s).

hO 3.4.1.2 V 6;y =

the cumulative release of noble gas radionuclide i i

from release pathway v during the period of interest (pci).

i 3.4.2

~

Q fy= the cumulative release of non-noble-gas radionuclide 1

i from release pathway v, during the period of interest (pci).

3.4.3 Raipj =

the site-specific dose factor for age group a, radio-f nuclide i, exposure pathway p, and organ j. Values !

( and units of R,;g for each exposure pathway, age

! group,- and radionuclide that .may arise in

] calculatione for implementing section 3.1.4 are listed in Table 3-8 through Table 3-11.

e l I

1 -3.4.3 1 i l Rk= the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in !

l the release under consideration (unitiess). 3.3.2.1

(

t i

i 3-51 Gen. Rev. 13 e

4 1

i FNP-0-M-011 n Definition Section of.

lb

'V Initial Use l Rt= the ratio of the total body dose rate limit for noble l l

gases, to the total body dose rate due to- noble gases i

in the release under consideration (unitiess). 3.3.2.1 l

the ratio of the skin dose rate limit for noble rk =

1 gases, to the skin dose rate'due to noble gases'in !

the source stream under consideration (unitiess).

! 3.3.3.1 !

l u

1 rg = the ratio of the total body dose rate limit for noble l 1

gases, to the total body dose rate due to noble gases in the source stream under ' consideration (unitless). 3.3.3.1 SF = l the safety factor used in gaseous setpoint J calculations to compensate for statistical I fluctuations and errors of measurement (unitless). '3.3.2.2 t= the number of whole or partial days elapsed in the -

current quarter, including the period of the release under consideration. '

3.5.1-Wjyp=

the annual average relative dispersion [(X/6)yp) or

{

deposition [(D7D)yp) at the location of the l controlling receptor, for release pathway v, as appropriate to exposure pathway p and radio-nuclide 1.

3.4.3 X= the noble gas concentration for the release under consideration (yci/mL). 3.3.2.1 Xg = the concentration of radionuclide i applicable to 1

active gaseous release pathway r (yci/mL). 3.3.4 X;, = the measured concentration of radionuclide i in gaseous source stream s (yci/mL). 3.3.3 X;y =

the measured concentration of radionuclide 1 in gaseous stream v (yci/mL). 3.3.2.2 0

3-52 Gen. Rev. 13

FNP-0-M-011 T.3Im Definition Stetion of Initial Use (X/Q) =

the highest relative concentration at any point at or

]

beyond the SITE BOUNDARY [s/m3~]. 3.3.2.1

)

5 (17Dhb = the annual average SITE BOUNDARY relative concen-tration applicable to active gaseous release pathway {

r [s/m3]. 3,3.4 I (176)vb = the highest annual average relative concentration at the SITE BOUNDARY for the discharge point of release pathway v, from Table 3-4 [s/m3 ).

3.3.2.2 (176)yp = annual average relative dispersion factor for release '

l pathway v, at the' location of the controlling receptor, from Table 3-7 [s/m 3]. 3.4.3-i tf lb II O

F 3-53 Gen. Rev. 13 ;

I ! ,

I' FMP-0-M-011 CHAPTER 4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGEM

a. .

4.1 LIMITS OF OPERATION The following limits are the.same for both units at the site. Thus, a single l

1 program including monitoring,' land use survey, and quality assurance serves both

units.

4.1.1 Radiolooleal Environmental Monitorina In accordance with Technical Specification 6.8.3.f(i), the Radiological 4

Environmental Monitoring Program (REMP) shall be conducted as specified in Table 4-1. -

I 4.1.1.1 Applicability This control applies at all times.

i 4.1.1.2 Actions 4

4.1.1.2.1 With the REMP not being conducted as specified in Table 4-1, submit to the Nuclear Regulatory Commission (NRC), in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting i

' the program as required and the plans for preventing a recurrence. Deviations from the required sampling schedule are permitted if specimens are unobtainable a

due to hazardous conditions, unavailability, inclement weather, equipment malfunction, or other just reasons. If deviations are due to equipment l malfu'n ction, efforts shall be made to complete corrective action prior to the end of the next sampling period.

d l 4.1.1.2.2 with the confirmedI measured level of radioactivity as a result of plant effluents in an environmental sampling medium specified in Table 4-1

' exceeding the reporting levels of Table 4-2 when averaged over any calendar )

{

quarter, submit within 30 days a special Report to the NRC pursuant to Technical Specification 6.9.2.

i The Special Report shall identify the cause(s) for exceeding the' limit (s) and define the corrective action (s) to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC i

e 1

Defined as confirmed by reanalysis of the original sample, or analysis of a duplicate or new sample, as appropriate. The results of the confirm- !

atory analysis shall be completed at the earliest time consistent with the analysis.

i

% )

4-1 Gen. Rev. 13 I

i

FNP-0-M-011 is less than the calendar year limits of Sections 2.1.3, 3.1.3, and 3.1.4. The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in the special Report.

When more than one of the radionuclides in Table 4-2 are detected in the sampling medium, this report shall be submitted if concentracion (1) , concentracion (2) , , , , y 3,n reporting level (1) reporting level (2)

When radionuclides other than those in Table 4-2 are detected and are the result l of plant effluents, this Special Report shall be submitted if the potential t

annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits stated in Sections 2.1.3, 3.1.3, and 3.1.4. This Special Report is l not required if the measured level of radioactivity was not the result of plant l effluents; however, in such an event, the condition shall be described in the Annual Radiological Environmental Operating Report. The levels of naturally-l occurring radionuclides which are not included in the plant's effluent releases l need not be reported.

4.1.1.2.3 If adequate samples of milk, or during the growing season, forage or fresh leafy vegetation, can no longer be obtained from one or more of the sample locations required by Table 4-1, or if the availability is frequently or

(A; t

persistently wanting, efforts shall be made: to identify specific locations for obtaining suitable replacement samples; and to add any replacement locations to the REMP given in the ODCM within 30 days. The specific locations from which samples became unavailable may be deleted from the REMP. Pursuant to Technical Specification 6.14, documentation shall be submitted in the next Annual Radioactive Effluent Release Report for the change (s) in the ODCM, including revised figure (s) and table (s) reflecting the changes to the location (s), with supporting information identifying the cause of the unavailability of samples and justifying the selection of any new location (s). ,

4.1.1.2.4 This control does not affect shutdown requirements or MODE changes. 1 i

l l

i i

1

/ 4-2 Gen. Rev. 13 I

i I

l

, .~

4 FNP-0-M-011

, 4.1.1.3. Surveillance Requirements The REMP samples shall be collected pursuant to Table 4-1 from the locations

( described in Section 4.2, and shall be analyzed pursuant to the requirements of Table 4-1 and Table 4-3.

Program changes may be initiated based on operational experience.

Analyses shall be performed in such a manner that the stated MINIMUM DETECTABLE CONCENTRATIONS (MDCs) will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering radionuclides, or other uncontrollable circumstances may render these MDCs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

4.1.1.4 Basis The REMP required by this control provides representative measurements of radiation and of radioactive materials in those exposure pathways, and for those radionuclides, which lead to the highest potential radiation exposures of MEMBERS "

OF THE PUBLIC resulting from the plant operation. The REMP implementsSection IV.B.2, Appendix I,10 CFR 50, and thereby supplements the radiological effluent monitoring program by measuring concentrations of radioactive materials and

. avels of radiation, which may then be compared with those expected on the basis of tne effluent measurements and modeling of the environmental exposure pathways.

The dotection capabilities required by Table 4-3 are within state-of-the-art for routir.e environmental measurements in industrial laboratories.

l 1

i i

O 4-3 Gen. Rev. 13

O O

a tr Number of "

Pat y an /or Samples and Sampling and Collection Sample Sample Frequency Type and Frequency of Analysis a, Locations * "

1. AIRBORNE Particulates 2:

Continuous operation of Particulate sampler. Analyze for Indicator 3 sampler with sample gross beta radioactivity 224 hours0.00259 days 0.0622 hours 3.703704e-4 weeks 8.5232e-5 months control 2 collection weekly, following filter change. Perform g

e gansna isotopic analysis on each j sample when gross beta activity is y

>10 times the yearly mean of control se samples. Perform gamma isotopic "

analysis on composite (by location) y sample quarterly. <

e-Radiciodine Indicator 3 Radioiodine canister. Analyze o weekly for I-131.

control 2 @

e

ft

2. DIRECT RADIATION f, TLD Quarterly.

Indicator I 16 Gamma dose quarterly. d Indicator II 16 %

(community) et Control 3 O y

42

3. WATERBORNE

.a Surface n Composite ** sample collected Gamma isotopic analysis monthly.

Indicator 1 monthly. S Control 1 Tritium analysis of composite (by "-

location) sample quarterly. $

Ground Quarterly.

Indicator 1 Gamma isotopic and tritium analysis of each sample.

Control 1 e Sediment Semiannually.

D Indicator 1 Gasuna isotopic analysis yearly. ..:

33 %,

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G E

e Number of 8*"Ples and i Pat y nd/or g p. g Sampling and collection Frequency Type and ' "

Sample "

gnalye a Locations

8 E,

4. INGESTION Milk

' 3 semimonthly when animals are on Gamma isotopic end I-131 Indicator 3*** pastures monthly at other times.

Control 1

analysis of each sample.

Fish :o -

One sample in season, or semiannually Gamma isotopic analysis $,

Indicator 1 if not seasonal. One sample of each of on edible portions. p control 1 the following species: -

1. Game Fish

2. Botton Feeding Fish S p

Forage or Grab sample cut from green forage or I.eafy 1 vegetation monthly.

Gamma isotopic analysis a Vegetation 1 which includes I-131 en E Indicator analysis of each sample. E control y o

5 rt Sample locations are shown in Table 4-4, and in Figure 4-1 through Figure 4-4. "

- x Composite hours. samples shall be collected by collecting an aliquot at intervals not exceeding 2 8 p

- - o Up to three sampling locations within 5 miles in different sectors with the highest dose potential will be used as available.

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Reporting Level I

3 Airborne Particulate or Water Game Forage or Leafy la y Fish Milk Vegetation Analysis (pC1/L) (PC1/m ) $m (pC1/kg, wet) (pci/L) (pci/kg, wet) 30 H-3 2 E+4" 20 ,

o Mn-54 1 E+3 3 E+4 n e

Fe-59 4 E+2 1 E+4 e w

Co-58 1 E+3 "

3 E+4 Co-60 3 E+2 o 1 E+4 "

o En-65 3 E+2 2 E+4 $.

Er-95 4 E+2 Mb-95 7 E+2 I-131 2 E+0 h 7

9 E-1 3 E+D 1 E+2 N N

_Co-134 3 E+1 1 E+1 1 E+3 6 E+1 1 E+3 Cs-137 5 E+1 2 E+1 2 E+3 7 E+1 2 E+3 $

Ba-140 2 E+2 u 3 E+2 E La-140 D 1 E+2 4 E+2 $

0 e

a. This is the 40 CFR 141 value for drinking water samples. If no drinking water pathway y exists, a value of 3 E+4 pC1/L may be used.

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$ b. u If no drinking water pathway exists, a value of 20 pCi/L may be used. $

3 M 2 ?

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- _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ m ,_._ -- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _

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o 3 O H

Minimum Detectable Concentration (MDC)" It e

Forage or .

Airborne Leafy 8 Patciculate Fish Vegetation P Water or Gasgo (pci/kg, Milk (pci/kg, Sediment Analysis (sci /L) (PCi/m ) wet) (pci/L) wet) (pci/kg, dry) gross beta 4 R+0 1 E-2 e "c

H-3 2 E+3 h E m

Mn-54 1.5 E+1 1.3 E+2 Fe-59 3 E+1 2.6 E+2 $

e Co-58, co-60 1.5 E+1 1.3 E+2 En-65 3 E+1 2.6 E+2 "

9 Er-95 3 E+1 g

, Nb-95 1.5 E+1 s E I-131 1 E+0C 7 E-2 1 E+0 6 E+1 $

Co-134 1.5 E+1 5 E-2 1.3 E+2 $&

1.5 E+1 6 E+1 1.5 E+2 co-137 1.8 E+1 6 E-2 e 1.5 E+2 1.8 E+1 8 E+1 1.8 E+2 o

Ba-140 6 E+1 o 6 E+1 D La-140 1.5 E+1 7e 1.5 E+1 D E

n

a. See the definition of MINIMUM DETECTABLE CONCENTRATION in Section 10.1. 'Other peaks which- o are measurable and identifiable as plant effluents, together with the radionuclides in this D table, shall be analyzed and reported in accordance with Section 7.1.

o

b. If no drinking water pathway exists, a value of 3 E+3 pC1/L may be used.

$ c. If no drinking water pathway exists, a value of 1.5 E+1 pC1/L may,be used. E 3

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FNP-0-M-011 l 4.1.2 Land Use census 0

[ In accordance with Technical Specification 6.8.3.f(ii), a land use census shall be conducted and shall identify the location of the nearest milk animal l .nd s the nearest permanent residence, in each of the 16 meteorological sectors, 'sithin a distance of 5 miles.

4.1.2.1 Applicability This control applies at all times.

4.1.2.2 Actions 4.1.2.2.1 With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than values currer.tly being calculated in accordance with section 3.4.3, identify the new location (s) in the next Annual Radioactive Effluent Release Report.

4.1.2.2.2 With a land use census identifying a incation(s) which yields a calculated dose or dose comunitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Section 4.1.1, add the new location (s) to the REMP within 30 days if samples are available. The sampling location, excluding control station I location (s), having the lowest calculated dose or dose commitment (via the same l

exposure pathway) may be deleted from the REMP if new sampling locations are added. Pursuant to Technical Specification 6.14 subunit in the next Annual l

Radioactive Effluent Release Report any change (s) in the ODCM, including the revised figure (s) and table (s) reflecting any new location (s) and in' formation supporting the change (s).

4.1.2.2.3 l This control does not affect shutdown requirements or MODE changes.

4.1.2.3 surveillance Requirements i

The land use census shall be conducted annually,,using that information which !

will provide good results, such as a door-to-door census, a visual census from l

automobile or aircraft, consultation with local agriculture authorities, or some combination of those methods, as feasible. Results of the land use census shall be included in the Annual Radiological Environmental Operating Report.

l 1

b Defined as a cow or goat that is producing milk for human consumption.

4-8 Gen. Rev. 13

FNP-O-M-011 ;

4.1.2.4 Basis I

This control is provided to ensure that changes in the use of UNRESTRICTED AREAS '

are identified and that modifications to the REMP are made if required by'the l results of this census. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

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FNP-O-M-011 4.1.3 Interlaboratory Comoarison Procram l

l l In accordance with Technical specification 6.8.3.f(iii), analyses shall be

g performed on radioactive materials supplied as part of an Interlaboratory t

I comparison Program which has been approved by the NRC. Analyses are required to be performed only in cases in which the sample type and analysis are the same as the sample type and analysis included in Table 4-1.

4.1.3.1 Applicability This control applies at all times.

4.1.3.2 Actions With analyses not being performed as required by section 4.1.3, report the corrective actions taken to prevent a recurrence in the Annual Radiological Environmental Operating Report.

This control does not affect shutdown requirements or MODE changes.

4.1.3.3 surveillance Requirements Either a summary of the results obtained as part of the required Interlaboratory Coreparison Program shall be included in the Annual Radiological Environmental Operating Report, or participants in the EPA cross-check program shall provide the EPA program code designation for the plant in the Annual Radiological Environmental Operating Report.

l l 4.1.3.4 Basis The requirement for participation in an approved Interlaboratory comparison ;

Program is provided to ensure that independent enecks on the precision and {

accuracy of the esasurements of radioactive matarial in environmental sample matrices are performed as part of the quality aest rance program for environmental )

j monitoring, in order to demonstrate that the resu).ts are reasonably valid for the

! purposes of Section IV.E.2, Appendix I, 10 CFR 50.

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4.2 RADIOL.OGICAL ENVIRONMENTAL MONITORING LOCATIONS a ,

l Table 4-4, and Figure 4-1 through Figure 4-4 specify the locations at which the measurements and samples are taken for the REMP required by Section 4.1.1'.

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FNP-0-M-011 Tablo 4-4. Radiological Environmental Monitoring Locations i 1

Exposure l

\ Path ay Sampling Locations

3, 3, )

' , p Identifi- (

Sample cation l l

1. AIRBORNE Partic- Indicator Stations: l l.

ulates

River Intake Structure (ESE-0.8 miles)I PI-0501 South Perimeter (SSE-1.0 miles) PI-0701 Plant Entrance (WSW-0.9 miles) PI-1101 North Perimeter (N-0.8 miles) PI-1601 Control Stations

Blakely, GA (NE-15 miles) PB-0215 :

Dothan, AL (W-18 miles) PB-1218 '

j Neals Landing, FL (SSE-18 miles)I PB-0718 8

, Community Stations:

1 Georgia Pacific Paper Co. (SSE-3 miles) PC-0703 i

Ashford, AL (WSW-8 miles) PC-1108 Columbia, AL (N-5 miles) PC-1605 l

Radiciodine Indicator Stations: l i

River Intake Structure (ESE-0.8 miles)I I -O'01 {

South Perimeter (SSE-1.0 miles) II-0701 '

Plant Entrance (WSW-0.9 miles) II-1101 North Perimeter (N-0.8 miles) II-1601 Control Stations:

Blakely, CA (NE-15 miles) IB-0215 Dothan, AL (W-18 miles) IB-1218

("' Neals Landing, FL (SSE-18 miles)I IB-0718 '

\w Community Stations:

Georgia Pacific Paper Co. (SSE-3 miles)2 gc.o7o3

2. DIRECT RADIATION i TLD Indicator I Stations:

Plant Perimeter (NNE-0.9 miles) RI-0101

! (NE- 1.0 miles) RI-0201 (ENE-0.9 miles) RI-0301 i

(E- 0.8 miles) RI-0401 (ESE-0.8 miles) RI-0501 (SE- 1.1 miles) RI-0601 i (SSE-1.0 miles) RI-0701 !

(S- 1.0 miles) RI-0801 (ssW-1.0 miles) RI-0901 ;

(SW- 0.9 miles) RI-1001 (WSW-0.9 miles) RI-1101 (W- 0.8 miles) RI-1201 (WNW-0.8 miles) RI-1301 (NW- 1.1 miles) RI-1401 (NNW-0.9 miles) RI-1501 (N- 0.8 miles) RI-1601 4-12 Gen. Rev. 13

i 8

FNP-O-M-011

Table 4-4 (contd). Radiological Environmental Monitoring Locations

O Exposure i

PathwaI Sampling Locations

3,p g, pr Identifi-

Sample #"Di*"

{ TLD (contd) Control Stations:

4 Blakely, GA (NE-15 miles) RB-0215 Neals Landing, FL (SSE-18 miles) RB-0718 Dothan, AL (W-15 miles) RB-1215

+

Dothan, AL (W-18 miles) RB-1218 l

a Webb, AL (WNW-11 miles) RB-1311 Haleburg, AL (N-12 miles) RB-1612 Indicator II fCommunity) Stations:

i (NNE-4 miles) RC-0104 (NE- 4 miles) RC-0204 (ENE-4 miles)

J RC-0304 5

(E- 5 miles) RC-0405

! (ESE-5 miles) RC-0505 l

(SE- 5 miles) RC-0605

! (SSE-3 miles) RC-0703 i (5- 5 miles) RC-0805 1

' (SSW-4 miles) RC-0904 l (SW- 1.2 miles) RC-1001 )

] (SW- 5 miles) RC-1005 1

(WSW-4 miles) RC-1104 j (WSW-8 miles) RC-1108 i

(W- 4 miles) RC-1204 (WNW-4 miles) RC-1304 (NW- 4 miles) f]g RC-1404 g' (NNW-4 miles) RC-1504 J (N- 5 miles) RC-1605

3. WATERBORNE Surface Indicator Stations Georgia Pacific Paper Co. Intake Structure WRI -

3 (River Mile - 40)

Control Station: '

Andrew Lock & Dam Upper Pier (River Mile - 47) WRB

, Ground Indicator Stations Georgia Pacific Paper Co. Well (SSE-4 miles) WGI-07 control station:

Whatley Well (SW-1.2 miles) WGB-10 Sediment Tndteator stations Se&th's Bond (River Mile - 41) RSI caatrol stations Andrews Lock & Dam Reservoir (River Mile - 47) RSE I 4-13 Gen. Rev. 13 i

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FNP-0-M-011 Table 4-4 (contd). Radiological Environmental Monitoring Locations

()s i-s Exposure Pathway 3,,p g ,

Sampling Locations

Identifi-Sample **D1 "

4. INGESTION Milk Indicator Station:

None (There are no milk animals within 5 miles per the current land use survey)

Control Station:

Ray Lewis Dairy MB-1114 Ashford, AL (WSW-14 miles)

Fish Indicator Station:

Smith Bend (River Mile - 41)

Game Fish FGI Bottom Feeding Fish FBI 1

Control Station Andrews Lock & Dam Reservoir (River Mile - 47)

Game Fish FGB Bottom Feeding Fish FBB Forage or Indicator Stations:

Leafy i Vegetation South Southeast Perimeter (SSE-1.0 miles) FI-0701 '

North FI-1606 SouthPerimeter Perimeter_(N-0.8 miles)3 i (S-1.0 miles) FI-0801 j Northeast Perimeter (NE-1.0 miles)3 FI-0201

(")

( j gentrol StatiED8 1

\

Dothan, AL (W-18 miles) FB-1218 1

i l

Distance and direction as measured from the centerpoint between Unit 1 and Unit 2 plant vent stacks.

1. Not required by Section 4.1.1. Used as a spare station.

2. Not required by Section 4.1.1.

State of GA EPD. Use for comparison purposes with

3. Alternate forage plots.

O 4-14 Gen. Rev. 13

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4 Indicator II (Commsunity) sampling Locations for Direct Radiation 1

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FNP-0-M-011-CHAPTER 5 p TOTAL DOSE DETERMINATIONS b

5.1 LIMIT OF OPERATION In accordance with Technical Specification 6.8.3.e(x), the ' dose or ' dose commitment to any MEMBER OF THE PUBLIC over a calendar year, due to releases of radioactivity and to radiation from uranium fuel cycle sources, shall be limited

.to less than or equal to 25 mrom to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrom.

5.1.1 Aeolicability This limit applies at all times.

5.1.2 Actions With the calculated. doses from the release of radioactive materials in 7.ipid or gaseous effluents exceeding twice the limits of Section 2.1.3, 3.1.3, o;;3.1.4,-

calculations shall be made according to Section 5.2 methods to determine whether the limits of Section 5.1 have been exceeded. If .these limits have been 0

/

exceeded, prepare and submit a .special Report . to the Nuclear Regulatory Commission, pursuant to Technical Specification ~ 6.9.2, within 30' days, which defines the corrective actions to be taken to reduce subsequent releasesLto prevent recurrence of exceeding the limits of Section 5.1 and includes the-schedule for achieving conformance with the limits of Section 5.1. This Special Report, as defined in 10 CFR 20.2203, shall-also include an analysis which estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources (including all effluent pathways and direct radiation) for the calendar year that includes the release (s) covered by this report. This Special Report shall also describe the levels of radiation and concentrations 'of radioactive material involved, and the cause of the exposure- levels or concentrations. If the estimated dose (s) exceeds the limits of Section 5.1, and if the release condition resulting in violation of the provisions of 40 CFR 190'-

has not already been corrected, the Special Report shall include a request for variance in accordance with the provisions of 40 CFR 190 and including the specified information of 40 CFR 190.11(b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request-is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of-10 CFR Part.20, as addressed in other sections of this ODCM. !

O i 5-1 Gen. Rev. 13

I' i

FNP-O-M-011' This control does not affect shutdown requirements or MODE changes.

I 5.1.3 Surveillance Recuirements 5 Cumulative dose contributions from liquid and gaseous effluents and from direct i

radiation shall be determined in accordance with Section 5.2. This requirement is applicable only under the conditions set forth above in Section 5.1.2. '

5.1.4 Basis This control is provided to meet the dose limitations of 40 CFR 190.- The control' i

) requires the preparation and submittal of a Special - Report whenever the !

calculated doses from plant radioactive effluents combined with doses' due to 3 direct radiation from the plant exceed the limits . of 40 CFR 190. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a i

i MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will 5 describe a course of action which should result in the limitation of dose to a MEMBER OF THE PUBLIC for a calendar year to within the4 0 CFR 190 limits. For j the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible with the exception that dose contributions from . other uranium fuel cycle facilities at the same site or within a radius of 5 miles must be considered.

1 i

If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been -

I

' corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.2203(a)(4), ic considered to be a timely request and fulfills the requirements 3

of 40 CFR 190 until NRC staff action is coepleted. An individual ia not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged j

in carrying out any operation which is part of the nuclear fuel cycle. {

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FNP-O-M-011 {

5.2 DEMONSTRATION OF COMPLIANCE V There are no other uranium fuel cycle facilities within 5 miles of the plant <

site.

Therefore, for the purpose of demonstrating compliance with the limits of-Section 5.1, 'f the total dose to a MEMBER OF THE PUBLIC in the vicinity of the ;

plant site due to uranium fuel cycle sources shall be determined as follows:

)

Dy7

=

DL+Da+Dp+Dy (5.1}

wheres DTk = the total dose or dose commitment to the total body or organ k, in mrom.

I DL= the dose to the same organ due to radioactivity discharged from the plant site in liquid effluents, calculated in accordance with 1 Section 2.4.1,'in mrom.

l DG= the dose to the - same organ due to . non-noble-gas radionuclides discharged from the plant site in gaseous effluents, calculated for the controlling receptor in accordance with section 3.4.3, in mrom.

DD= the direct radiation dose to the whole body of an ' individual at the controlling receptor location, due to radioactive materials retained within the plant site, in mrom. Values of direct ;

radiation dose may be determined by measurement, calculation, or a combination of the two.

DN= the external whole body dose to an individual at the controlling receptor location, due to gamma ray emissions from noble gas radio-nuclides discharged from the plant site in gaseous affluents, in ares. DN is calculated as follows (equation adapted from Raforence 1, page 22, by re-casting in cumulative dose form):

Dy = 3.17 x 10'I {' (y75)9 hg . df , '

(5.2)

V t where

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5-3 Gen. Rev. 13 l

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4 i

FNP-0-M-011 3.17 x 10'8 = a units conversion f actor: 1 y/(3.15 x 10 7 s).

t' 6;y =

the cumulative release of noble gas radionuclide i from ,

release pathway v (pci), during the period of interest. +

Kj = the total-body dose factor due to gamma emissions from noble gas radionuclide 1 (mrom/y)/(pci/m 3), from Table 3-5.

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(176)yp = annual average relative dispersion factor for release i 4

pathway v, at the location of the controlling receptor, i

from Table 3-7 (s/m3 ]. '

j i As defined above, DL and Dg are for dif ferent age groups, while DD and Dg are !

not age group specific. When a more precise determination of DTk is desired, I values of DL and Da may be calculated for all four age groups, and those values used in equation (5.1) to determine age group specific values of DTk; the largest ;

value of DTk for any age group may then be compared to the limits of section 5.1. .

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FNP-0-M-011 i

CHAPTER 6 '

I POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO i

)

THEIR ACTIVITIES INSIDE THE SITE BOUNDARY !

i 6.1 REQUIREMENT FOR CALCULATION Current FNP effluent controls as established by this ODCM do not require assessment of the radiation doses from radioactive liquid and gaseous ef fluents 1 to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY I (Figure 10-1). However, when such an assessment is desired, it should be performed in accordance with section 6.2.

6.2 CALCULATIONAL METHOD I For the purpose of perf orming the calculations required in Section 6.1, the dose to a member of the public inside the SITE BOUNDARY shall be determined at the locations, and for the receptor age groups, defined in Table 6-1. The dose to such a receptor at any one of the defined locations shall be determined as follows:

Dik O {DA+DS*DP}'To (6.1) where !

l Dg = the total dose to the total body or organ k, in area. ~

DA= the dose to the same organ due to-inhalation of non-noble-gas i radionuclides discharged from the plant site in gaseous effluents, ,

calculated in accordance with section 3.4.3, in area. The (i76)

{

value to be used is given for each receptor location in Table 6-1; '

depleted (176) values may be used in calculations for non-noble-gas radionuclides. ,

D3= the dose to the same organ due to around elane deoesition of non-l L

noble-gas radionuclides discharged from the plant site in gaseous effluents, calculated in accordance with section 3.4.3, in area.

The (D76) value to be used is given for each receptor location in Table 6-1.

l l

s 6-1 Gen. Rev. 13 l

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r FNP-0-M-011 Dp = the external whole body dose due to gamma ray emissions from noble gas radionuclides discharged from the . plant ' site in gaseous l effluents, calculated using equation (5.2), in mrem. The - (i?Q) I

.i values that are to be used are given for'each receptor location in Table 6-1. j Fa =

the occupancy factor for the given location,-which is the fraction of the year that one individual MEMBER OF THE PUBLIC is assumed to-l be present at the receptor location (unitiess), values of F o for )

each receptor location are included in Table 6-1.

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4 FMP-0-M-011 i

Tablo 6-1.

l Attributoo SITE BOUNDARY of M:mbor of the Public Raceptor. Locations Inside the i

?

Location: Visitor Center, WSW at 0.19 miles 1

Ace Grouo Child i

Occuoancy Factor 1.37 E-03 (based on 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per year) 1 Discorsion and Decosition Parameters:

}

e Parameter Ground-Level Mixed-Mode (176), s/m3 1.04 E-04 8.80 E-06 (D/Q), m*2 4.80 E-07 6.20 E-08 1

A Y

i j

Locations Service Water Pond, SSW at 0.60 miles 4

i Ace crount Child i

Occuoancy Factor a

7.53 E-03 (based on 66 hours7.638889e-4 days 0.0183 hours 1.09127e-4 weeks 2.5113e-5 months per year) 4 Disoarsion and Deoosition Parameters:

4 Parameter Cround-Level Mixed-Mode >

j (176), s/m3 4.74 E-05 9.75 E-07 i

(676), m*2 1.31 E-07 i 2.78 E-08

1 i

i 4 %

l I

i 1

i

I j

6-3 . Gen. Rev. 13 4

+

., i f

FNP-O-M-011 Table 6-1 (contd). Attributes of Member of the Public Receptor Locations Insids the SITE BOUNDARY t

, v j Location River Water Discharge, SE at 1.02 miles f Ace Groues . Child

(; Occucancy Factor: 1.14 E-02 (based on 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months per year)

I l Disoarsion and Deoosition Parametern:

i Parameter Ground-Level M.'xed-Mode i

s/m 3

(X/Q), 1.63 E-05 7.05 E-07 (67Q), m -2 4.55 E-08 1.39 E-08 l .

I i

1 N

l a

l

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4 4

i 1

1 i

i 1

7 1

1 J

14 4

i 6-4 Gen. Rev. 13

l FNP-O-M-011 4 CHAPTER 7 REPORTS a

7.1

?

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 4

7.1.1 Reauirement for Reoort i

i In accordance with Technical Specifications 6.9.1.6 and 6.9.1.7, the Annual Radiological Environmental Operating Report covering the REMP activities during i

the previous calendar year shall be submitted before May 1 of each year. (A single report fulfills the requirements for both units.) The material provided shall be consistent with the objectives outlined in Section 4.1 and Section 7.1.2 v of the ODCH, and in Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR 4

Part 50.

1 1

7.1.2 Reoort Contents 1

1 The materials specified in the following sub-sections she.11 be included in each i Annual Radiological Environmental Operating Report 7.1.2.1 Data

O The report shall include summarized and tabulated results of all REMP samples required by Table 4-1 taken during the report period, in a format similar to that contained in Table 3 of the Radiological Assessment Branch Technical Position !

(Reference 13); the results for any additional samples shall also be included.

j In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results; the missing data shall be submitted as soon as possible in a

! supplementary report. The results for naturally-occurring radionuclides not 4

included in plant effluents need not be reported.

i 7.1.2.2 Evaluations J

Interpretatione and analyses of trends of the results shall be included in the report, including the followings (as appropriate) comparisons with pre-I operational studies, operational controls, and previous environmental operating reports; and an assessment of any observed impacts of the plant operation on the environment. If the measured level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 4-2 is not the result of plant affluents, the condition shall be described as required by section 4.1.1.2.2.

O 3

7-1 Gen. Rev. 13 4

, , "N**

M

, 0 t

FNP-0-M-011 7.1.2.3 Programmatic Information Also to be included in each report are the followings a summary description of the REMP; a map (s) of all sampling locations keyed to a table giving distances and directions from the center point between the Unit 1 and Unit 2 plant vent stacks; the results of land use censuses required by Section 4.1.2; and the results of licensee participation in the Interlaboratory Comparison Program required by Section 4.1.3.

[The report shall include either a summary of the results obtained as; part of the required Interlabc ratory comparison Program or, for licensees participating in the EPA cross-check program, the EPA program code designations for the plant.]

7.1.2.4 Descriptions of Program Deviations Discussions of deviations from the established program must be included in each report, as follows:

l 7.1.2.4.1 If the REMP is not e7nducted as required in Table 4-1, a

~'

description of the reasons for not conducting the program as required, and the 1

plans for preventing a recurrence, must be included in the report.

7.1.2.4.2 !

If the NDCs required by Table 4-3 are not achieved, the contributing factors must be identified and described in the report.

'( 7.1.2.4.3 If Interlaboratory Comparison Program analyses are not performed as required by Section 4.1.3, the corrective actions taken to prevent a recurrence must be included in the report. ;

I j

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)

s/ 7-2 Gen. Rev. 13

4 I

![ FNP-O-M-011 2

7.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT j

- 7.2.1 Recuirement for Reoort A

l a

In accordance with Technical Specifications 6.9.1'.8 and 6.9.1.9, the Annual' I Radioactive Effluent Release Report covering the operation of the units during ;

the previous calendar year of operation shall be submitted before May 1 of each j year. ( A single submittal may be ,

for Units 1 and 2. However, the submittal j shall specify the releases of active material in liquid . and gaseous ef fluents from each unit and solid radioactive waste from the site.) The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units. The material provided shall i be consistent with the objectives outlined thr ghout this ODCM and the Process 1

Control Program (PCP) and . in conformance wif CFR Part 50.36a and section i

IV.S.1 of Appendix I to 10 CFR Part 50 t

j 7.2.2 Reoort Content.g I

?

The materials specified in the following sub-sections shall be included in each j Annual Radioactive Effluent Release Reports ;

i ,

f

7.2.2.1 Quantities of Radioa6 s Materials Released j '

i The report shall include a summary of the quantities of radioactive liquid and ;

{ gaseous effluents and solid waste released from the units as outlined in NRC !

Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous l

Effluents from Light-Water-cooled Nuclear Power Plants," Revision 1, June 1974, I

with liquid and gaseous effluent data summarized on a quarterly basis and solid radioactive waste data summarized on a semiannual basis following the format of i

Appendix B thereof. - Unplanned releases of radioactive materials in gaseous and liquid effluents from the site to UNRESTRICTED AREAS shall be included in the report, tabulated either by quarter or by event. For gamma emitters released in

! liquid and gaseous effluents, in addition to the principal gamma emitters for j

which MDCe are specifically established in Table 2-3 and Table 3-3, other peaks which are measurable and identifiable also shall be identified and reported.

3 7.2.2.2 Meteorological Data I

The report shall include an annual summary of hourly meteorological data (

collected over the previous year. This annual summary may be either in the form

] of an hour-by-hour listing of wind speed, wind direction, and atmospheric stability, and precipitation (if measured).on magnetic tape; or in the form of l r

2 7-3 Gen. Rev. 13 a

,.-.s,. p --gig..,. y -gi7,,..q.sg ,,,,,g -w ww- y9 wy 9 i- 9-- owng-

FNP-0-M-011 joint frequency distributions of wind speed, t 'ind direction, and atmospheric stability.

Os In lieu of submission with the Annual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

7.2.2.3 Dose Assessments The report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from each unit during the previous calendar year. Historical annual average meteorology or the meteorological conditions concurrent with the time of release of radioactive materials in gaseous affluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway dose. This assessment of radiation doses shall be performed in accordance with Sections 2.1.3, 2.4, 3.1.3, 3.1.4, 3.4.2, 3.4.3, 5.1, and 5.2.

If a determination is required by Section 5.1.2, the report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary affluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation g Protection Standards for Nuclear Power operation; this dose assessment must be performed in accordance with Chapter S.

[

7.2.2.4 Solid Radwaste Data For each type of solid waste shipped offsite during the report period, the following information shall be included:

a. Container volume, b.

Total curie quantity (specify whether determined by measurement or estimate),

c.

Principal radionuclides (specify whether determined by measurement er estimate),

d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),

e.

Type of container (e.g., LSA, Type A, Type B, Large Quantity), and f.

Solidification agent (e.g., cement, urea formaldehyde.)

{

U 7-4 Gen. Rev. 13

5 FNP-0-M-Oll 7.2.2.5 Licensee Initiated Document Changes Licensee initiated changes shall be submitted to the Nuclear Regulatory commission as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period.in which any changes were made. Such changes to the ODCM shall be submitted ' pursuant to Technical Specification 6.14. This requirement includes:

7.2.2.5.1 Any changes to the sampling locations in the radiological environmental monitoring program, including any changes made pursuant to Section 4.1.1.2.3. Documentation of changes made pursuant to section 4.1.1.2.3 shall include supporting information identifying the cause of the unavailability of samples. -

h 7.2.2.5.2 Any changes to dose calculation locations or pathways, including 4

any changes made pursuant to Section 4.1.2.2.2.

7.2.2.6 Descriptions of Program Deviations Discussions of deviations from the established program shall be included in each report, as follows:

O)

( 7.2.2.6.1 The report shall include deviations from composite sampling 4

requirements included in Table 2-3 and Table 3-3.

7.2.2.6.2 The report shall include deviations from Minimum Detectable Concentration (MDC) requirements included in Table 2-3 and Table 3-3.

7.2.2.7 Major changes to Radioactive Waste Treatment Systems As required by Sections 2.1.5 and 3.1.6, licensee initiate 6 MAJOR CHANGES TO RADICACTIVE MASTE TREATMENT SYSTEMS (liquid and gaseous) shall be reported to the Nuclear Regulatory Commission in the Annual Radioactive Effluents Release Report covering the period in ,which the change was reviewed and accepted for implementation.I I

In lieu of inclusion in the Annual Radioactive Effluents Release Report,

(*~ this same information may be submitted as part of the annual FSAR update.

7-5 Gen. Rev. 13

FNP-0-M-011 The discussion of each change shall contains 7.2.2.7.1 A summary of the evaluation that led to the determination that the '

change could be made in accordance with 10 CFR 50.59; l

7.2.2.7.2 Sufficient detailed information to totally support the reason for the change without benefit of additional or supplementa.'. information; 7.2.2.7.3 A detailed description of the equipment, componer.cs and processes l involved and the interfaces with other plant systems; I 7.2.2.7.4 An evaluation of the change which shows'the predicted releases of radioactive materials in liquid and gaseous effluents that differ from those previously predicted in the license application and hmendments thereto; 7.2.2.7.5 An evaluation of the change which shows the expected maximum exposures to MEMBERS OF THE PUBLIC in the UNRESTRICTED AREAS and to the general population that differ from those previously estimated in the license application and amendments thereto; 7.2.2.7.6 A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents, to the actual releases for the period prior to when the changes are to be made; 7.2.2.7.7 An estimate of the exposure to plant operating persennel as a result of the change; and ;

~

7.2.2.7.8 Documentation of the fact that the change was reviewed and found acceptable by the PORC.

1 O 7-6 Gen. Rev. 13

F FNP-0-M-011 j i

7.3 MONTHLY OPERATING REPORT This ODCM establishes no requirements pertaining to the Monthly operating Report.

>.4 SPECIAL REPCMTS i

l Special reports shall be submitted to the Nuclear Regulatory Commission . in f accordance with Technical specification 6.9.2, as required by sections 2.1.3.2, 1 2.1.4.2, 3.1.3.2, 3.1.4.2, 3.1.5.2, 4.1.1.2.2, and 5.1.2. !

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FNP-0-M-011 I CHAPTER 8 h METEOROLOGICAL MODELS i

I The models presented in this chapter are those which were used to compute the r'

-j specific values of meteorology-related parameters that are referenced throughout this ODCM.

These models should also be used whenever it is necessary to calculate values of these parameters for new locations of interest.

NOTE: Although Plant Farley has no pure elevated releases, the sections on elevated-mode calculations (8.1.2 and 8.2.2) are included for converiencet in calculating mixed-mode values, and to preserve section. nut @er e

compatibility with the ODCMs of the other plants in the Southern Nuclear System. '

8.1 ATMOSPHERIC DISPERSION Atmospheric dispersion may be calculated using the appropriate form of the sector-averaged Gaussian model. Gaseous release elevations may be considered to be either at ground-level, elevated, or mixed-mode. Facility release elevatione for each gaseous release point a:e as indicated in Table 3-4.

I 8.1.1 cround-Level Releases Relative concentration calculations for ground-level releases, or for the ground-level portion of mixed-mode releases, shall be made as follows:

l 2.032 8 K,, afk (x/0)g =

{ (s.1)

N r . Uj _ Q

.,g l where: h 1

(X/Q)O .= the ground-level sector-averaged relative concentration' for a given wind direction (sector) and distance (s/m3 ).

1 2.032 = (2/r)1/2 divided by the width in radians of a 22.5' sector, which is 0.3927 radians.

i 8= the plume depletion factor for all radionuclides other than noble gases at a distance r shown in Figure 8-3. For noble gases, the depletion factor is unity. If an undepleted relative concentration a

I 8-1 !

Gen. Rev. 13 e

l* 5

. w, . . -

. . - .- . . . i

.1 FNP-0-M-011 .I 10 dmoired, the depletion factor is unity, only depletion by t

deposition is considered since depletion by radioactive decay would t O be of little significance at the distances considered. l V

g= the terrain recirculation factor corresponding ' to a distance r, ;

taken from Figure 8-2.

l njk = the number of hours that wind of wind speed class j is directed into the given sector during the time atmospheric stability category k existed.

N=

the total hours of valid meteorological data recorded throughouti !

the period of interest for all sectors, wind speed classes, and stability categories.

uj = the wind speed (mid-point of wind speed class j) at ground level l (m/s).

r= the distance from release point to location of interest (a).

l Eg = the vertical standard deviation of the plume concentration i j

distribution considering the initial dispersion within the building '

wake, calculated as follows:

< V r 2 b2 M E Zg = the lesser of a og 0*h i t

{*dh od = the vertical standard deviation of the plume concentration !

l patributie4(m) for a given distance and stability category k.as !

31a 8-1. The stability. category is determined by the l

, temperature gradient AT/As ('C/100 m or 'F/100 ft). Plant

.[ , Mfts-values must be adjusted for As of 165 ft.

% %. .2 l w= 3.1415 b= the maximum height of adjacent plant structure, which is the l

containment building (40 m).

r i l l

B-2 Gen. Rev. 13

S FNP-0-M-011 l 8.1.2 F)..a,yated Releases Relative dispersion calculations for elevated releases, or for the elevated V portion of mixed-mode releases, shall be made as follows:

-h 2 2.032 x y 6r njk exP 2 (X/0)E "

- E , 2 ozk. (3.3) jk

) *:k wheres (X/Q)E = the elevated release sector-averaged relative concentration for a given wind direction (sector) and distc.nce (s/m3 ),

6k= the plume depletion factor for all radionuclides other than noble gases at a distance r for elevated releases, as shown in Figure 8-4, Figure 8-5, and Figure 8-6. For an elevated release, this factor is stability dependent. For noble gases, the depletion factor is unity. If an undepleted relative concentration is desired, the depletion factor is unity. Only depletion by deposition is considered since depletion by radioactive decay would be of little significance at the distances considered.

njk = the number of hours that wind of wind speed class j is directed into the given sector during the time atmospheric stability category k existed.

uj = the wind speed (mid-point of wind speed class j) at the effective release height h (m/s).

m<

h= tho3 effective height of the release (m), which is calculated as

,tollower

~ *

,,s h=hy+hpr - hg-cy (s.4}

hy = the height of the release point (m).

O v

8-3 Gen. Rev. 13

t k

FNP-0 M-011 4 ht= the maximum terrain height between the release point and th<s point .

) s of interest (m), from Figure 2.3-26 of Reference 7.

i

)

h,=

p the additional height due to plume rise (m) which is calculated as

?

follows and limited by hp g ):

, ,2 1 W# - -

hpf = 1.44 d 3- 3 (8.5)-

)<

r i

l 4

i W

) 3 A *d

.t; . "J s i

q hpr(max) =

the lesser of: OR (8*0) 4 i1 1 F

1.5 2 3 3 0

[

J

"] s l i

t d=

the inside diameter of the vent (m).

Wo= the exit velocity of the plume (m/s).

I cy = the correction for low vent exit velocity (m), which is' calculated 3 as follower W# ~w 3 1.5 .d for # < 1. 5 i

]s Uj '

cy =

3 1.

og (8.7) ;

i r I

4

' y l 4

O for # 2 1.5-l

}

r, = the momentum flux parameter 4(m2/s ), which is calculated as follows t

(under the assumption that the effluent air and the. ambient air l

have the same density):

!O 4 V -

8-4 Gen. Rev. 13 l

i

$~

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

l l

l FNP-0-M-011 Tm * *

(No) (8*8) s= the stability parameter, which is calculated as follows:

3 = [ 9. 8 ) , ' $ + 9. 8 x 10-3

\ TI > 42 {g,,)

T= the ambient air temperature (*K).

(AT/As) =

the rate of increase of the ambient air temperature with increasing height above the ground (*K/m).

All other symbols are as previously defined in section 8.1.1.

8.1.3 Mixed-Mode Releases Relative dispersion calculations for mixed-mode releases - shall be made as follows:

(X/0)y =

( 1 -E) * (X/0)g ~ + E - (X/Q)g (s.10) where i

(X/Q)y = the mixed-mode release sector-averaged relative concentration for j a given wind direction (sector) and distance (s/m3 ).

'E=

the fraction of hours during which releases are considered as ground-level releases, calculated as follows:

l I

1 l

I I

i i

O 1 8-5 Gen. Rev. 13 l

i i

. . . . _ . . . . - ~ - . - - - . . ..

FNP-0-M-011 I

-W 1.0 for o 51.0 "1

i i

f 5

.l W' .w' <

2.58 - 1.58 - ._# for 1. 0 < ' # 's 1. 5 !

, ") , ")

-E = .]

(3.11). .

W, W !

0.3 - 0.06 - - for 1. 5 < ~, s 5. 0 ;

"J . ")

W .

O for o > 5.0 ~;

uj '

.i, All other symbols are as previously defined. _

t

-? .

r a

t i

E 1

l 2

a f

i r

P i

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8-6 Gen.,Rev. 13

.I 1

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. -. ~ . _ . . - _ . . .

FNP-0-M-011 8.2 REI.ATIVE DEPOSITION Plume depletion may be calculated using the appropriate form of the sector-Q averaged Gaussian model. Gaseous release elevations may be considered to be either at ground-level, elevated, or mixed-mode. Facility release elevations for each gaseous release points are as indicated in Table 3-4.

l l

8.2.1 Ground-Level Releases Relative deposition calculations for ground-level releases, or for the ground-level portion of mixed-mode releases, shall be made as follows:

2.55 O g K r (D/0)G N r Eg nk (s.12) where:

(D/Q)g = the ground level sector-averaged relative deposition for a given wind direction (sector) and distance (m-23 ,

2.55 = the inverse of the number of radians in a 22.5* sector

[= (2 vr/16)-I].

['\

Q D g= the deposition rate at distance r, taken from Figure ' 8-7 for ground-level releases (m-I).

1 nk = the number of hours in which the wind is directed into the sector of interest, and during which stability category k exisi:s.

All other symbols are as defined previously in Section 8.1.

8.2.2 Elevated Raleases Relative deposition c,alculations for elevated releases, or for the elevated portion of mixed-mode releases, shall be made as follows:

2.55 K (D/0)E y , E ("k Dd) (s.13) k where:

g i

\ j l

8-7 Gen. Rev._13 i

! 'j FNP-0-H-011

( D/Q) p, =

tho olevated-pluma ecctor-averaged relative deposition for a given wind direction (sector) and distance (m-2) .

Dd= the elevated plume deposition rate at <iistance r, taken from Figure 8-8, Figure 8-9, or Figure 8-10, as appropriate to the plume effective release height h defined in Section 8.1.2, for stability class k (m-l) .

All other symbols are as defined previously.

8.2.3 Mixed-Mode Releases Relativa deposition calculations for mixed-mode releases shall be made as-follows:

(D/0)y =

( 1 -E) * (D/Q)E +

E*(D/0)g (3.14) where:

(D/Q)g = the- mixed-mode release sector-averaged relative deposition for a given wind direction (sector) and distance (m-2) .

E=

l i ] the fraction of hours during which releases are considered as s

f ground-level releases, defined in Section 8.1.3.

All other symbols are as previously defined.

8.3 ELEVATED PLUME DOSE FACTORS These factors are not required in effluent dose calculations for FNP due to the fact that all gaseous effluent releases are either ground-level or mixed-mode.

8.4 METEOROLOGICAL

SUMMARY

A sununary of meteorological data for the years 1971 through 1975 is presented in Table 8-2 through Table-8-5.

t 8-8 Gen. Rev. 13 y y-. ,+y . + n -

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J l

4 i ..

1 FNP-0-M-011

, Table 8-1. \

Terrain Elevation Above Plant Site Grade l

1 1

i 5 l I I

)

4 s

i a

t 1

i

)

i

1 4

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1 4 1

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This table intentionally left blank.

j I A.

4 i

= l i

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i .

4 I i

^

! 1

1 i

! l 9 .

1 i

4 d

l d,

4 e

4 8-9 4

Gen. Rev. 13 I 1 1

! l 1

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

FNP-0-M-011 Teblo 8-2. Annual Average (X7Q) for Mixed Mode Releases Distance to Location, in miles

( O.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 N 2.16 E-06 9.21 E-07 5.92 E-07 3.83 E-07 2.42 E-07 NNE 2.35 E-06 1.02 E-06 6.18 E-07 3.82 E-07 2.34 E-07 NE 2.23 E-06 9.61 E-07 6.06 E-07 3.86 E-07 2.40 E-07 ENE 1.12 E-06 5.03 E-07 3.76 E-07 2.65 E-07 1.76 E-07 E 1.20 E-06 5.21 E-07 3.57 E-07 2.45 E-07 1.60 E-07 ESE 1.55 E-06 6.43 E-07 3.83 E-07 2.44 E-07 1.55 E-07 SE 2.47 E-06 9.69 E-07 1

5.52 E-07 3.47 E-07 2.19 E-07 SSE 2.77 E-OG 1.08 E-06 6.57 E-07 4.34 E-07 2.83 E-07 5 2.50 E-06 9.37 E-07 5.90 E-07 4.09 E-07 2.74 E-07 l SSW 2.02 E-06 8.29 E-07 6.30 E-07 4.16 E-07 2.66 E-07 SW 2.05 E-06 8.34 E-07 8.0) E-07 5.07 E-07 3.16 E-07 WSW 1.89 E-06 7.41 E-07 7.33 E-07 4.66 E-07 2.88 E-07 W 1.67 E-06 6.74 E-07 5.81 E-07 4.12 E-07 2.53 E-07 WNW 1.43 E-06 5.97 E-07

, 4.11 E-07 3.13 E-07 2.17 E-07

} NW 1.32 E-06 )

5.65 E-07 3.88 E-07 2.68 E-07 1.77 E-07 NNW 1.66 E-06 7.21 E-07 4.85 E-07 3.23 E-073 2.07 E-07 l

l Distance to Location, in miles i

2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 4.5-4.99 N 1.65 E-07 1.24 E-07 1.01 E-07 9.11 E-08 8.27 E-08 !

NNE 1.55 E-07 1.15 E-07 9.23 E-08 a

O' NE 1.61 E-07 1.19 E-07 9.62 E-08 8.28 E-08 8.63 E-08 7.48 E-08 7.79 E-08 s ENE 1.22 E-07 9.28 E-08 7.61 E-08 6.88 E-08 6.24 E-08

} E 1.12 E-07 8.54 E-08 7.09 E-08 6.43 E-08

' 5.86 E-08 ESE 1.07 E-07 8.13 E-08 6.75 E-08 6.12 E-08 5.58 E-08 SE 1.51 E-07 1.14 E-07 9.50 E-08 8.61 E-08 7.88 E-08

'SSE 1.96 E-07 1.50 E-07 1.26 E-07 1.15 E-07 1.05 E-07

! 1.96 E-07 S 1.52 E-07 1.29 E-07 1.18 E-07 1.09 E-07 SSW 1.84 E-07 1.39 E-07 1.22 E-07 1.18 E-07 1.08 E-07 SW 2.13 E-07 1.60 E-07 1.30 E-07 1.27 E-07 1.15 E-07 WSW 1.92 E-07 1.57 E-07 1.26 E-07 1.13 E-07 1.02 E-07 W 1.68 E-07 1.69 E-07 1.34 E-07

' 1.19 E-07 1.08 E-07 j WNW 1.74 E-07 1.72 E-07 1.35 E-07 1.21 E-07 1.09 E-07 NW 1.37 E-07 1.24 E-07 1.18 E-07 1.06 E-07 9.60 E-08 i NNW 1.42 E-07 1.07 E-07 1.04 E-07 9.36 E-08 8.50 E-08 1

values are in s/m 3, extracted from Reference 7.

h\~ ! 8-10 Gen. Rev. 13 t

..n. .

,er- ~w

l

! 1 FNP-0-M-011 1

Table 8-3. Annual Average (x7Q) for Ground-Level Releases

's

~

) Distance to Location, in miles

[ 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 N 7.25 E-05 2.38 E-05 8.63 E-06 4.02 E-06 2.05 E-06 NNE 6.16 E-05 2.02 E-05 7.32 E-06 3.39 E-06 1.73 E-06 !

NE 5.86 E-05 1.94 E-05 7.04 E-06 3.24 E-06 1.65 E-06 ENE 5.27 E-05 1.74 E-05 6.32 E-06 2.92 E-06 1.49 E-06 E 6.28 E-05 2.02 E-05 7.27 E-06 3.40 E-06 1.75 E-06 ESE 6.18 E-05 1.97 E-05 7.09 E-06 3.33 E-06 1.72 E-06 SE 9.48 E-05 3.01 E-05 1.07 E-05 5.06 E-06 2.63 E-06 SSE 1.44 E-04 4.55 E-05 1.61 E-05 7.65 E-06 3.99 E-06 S 1.55 E-04 4.87 E-05 1.72 E-05 8.20 E-06 4.28 E-06 SSW 9.78 E-05 3.12 E-05 1.11 E-05 5.23 E-06 2.71 E-06 SW 7.40 E-05 2.40 E-05 8.74 E-06 4.05 E-06 2.07 E-06 WSW 6.01 E-05 1.97 E-05 l 7.18 E-06 3.31 E-06 1.68 E-06 W 5.76 E-05 1.88 E-05 6.79 E-06 3.14 E-06 1.60 E-06 WNW 5.55 E-05 1.82 E-05 6.55 E-06 3.03 E-06 1.55 E-06 NW 5.67 E-05 1.86 E-05 6.76 E-06 3.14 E-06 1.60 E-06 4

NNW 6.60 E-05 2.16 E-05 7.85 E-06 3.65 E-06 1.87 E-06 Distance to Location, in miles 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 4.5-4.99 O N NNE 1.19 E-06 1.00 E-06 8.24 E-07 6.94 E-07 6.09 E-07 5.13 E-07 5.35 E-07 4.50 E-07 4.71 E-07 3.96 E-07 NE 9.47 E-07 6.54 E-07 4.82 E-07 4.23 E-07 3.71 E-07 ENE 8.56 E-07 5.92 E-07 4.37 E-07 3.82 E-07 3.37 E-07 E 1.02 E-06 7.08 E-07 5.24 E-07 4.61 E-07 4.06 E-07 ESE 1.02 E-06 6.99 E-07 5.18 E-07 4.56 E-07 4.02 E-07 SE 1.54 E-06 1.07 E-06 7.99 E-07 7.04 E-07 6.20 E-07 SSE 2.34 E-06 1.64 E-06 1.22 E-06 1.08 E-06 9.49 E-07 S 2.51 E-06 1.76 E-06 1.31 E-06 1.16 E-06 1.02 E-06 i

SSW 1.58 E-06 1.10 E-06 8.17 E-07 7.19 E-07 6.33 E-07 SW 1.20 E-06 8.30 E-07 6.12 E-07 5.38 E-07 4.73 E-07 WSW 9.65 E-07 6.67 E-07 4.91 E-07 4.31 E-07 3.79 E-07

, W 9.20 E-07 6.37 E-07 4.71 E-07 4.13 E-07 3.63 E-07 WNW 8.92 E-07 6.18 E-07 4.56 E-07 4.01 E-07 3.52 E-07 i NW 9.25 E-07 6.41 E-07 4.73 E-07 4.16 E-07 3.65 E-07 NNW 1.10 E-06 7.50 E-07 5.54 E-07 4.87 E-07 4.28 E-07 Values are in s/m3 , extracted from Psference 7.

I 1

(D

. su b

8-11 Gen. Rev. 13 l

FMP-0-M-011 Table 8-4. Annual Average (67Q) for Mixed Mode Releases.

i

# Distance to Location. in miles 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 :2.0-2.49 N 3.82 E-08 1.78 E-08 7.53 E-09 3.39 E-09 NNE 1.62 E-09 i 4.57 E-08 2.08 E-08 8.69 E 3.88 E-09 1.85 E-09.

NE 4.78 E-08 2.20 E-08 9.08 E-09 4.03 E-09 ENE 1.92 E-09 2.67 E-08 1.32 E-08 5.63 E-09 2.54 E-09' 1.22 E-09.

E 2.87 E 1.40 E-08 5.77 E-09 2.55 E-09 1.22 E-09 ESE 3.29 E-08 1.53 E-08 6.17 E-09 2.70 E-09 1.28 E-09 SE 5.30 E-08 2.37 E-08 9.31 E-09 4.01 E-09 1.90 E-09.

SSE 5.07 E-08 2.35 E-08 9.53 E-09 4.19 E-09 1.99 E-09 ;

S 4.86 E-08 2.29 E-08 .9.16'E-09 4.00 E-09 1.90-E-09 SSW 4.29 E-08 2.10 E-08 9.09 E-09 3.97 E-09' 1.88 E-09 l SW 4.70 E-08 2.28 E-08 '1.05 E-08 4.39 E-09 2.04 E-09; ;

W5W 4.46 E-08 2.17 E-08 9.88 E-09 4.12 E-09 '1.92 E-09: -!

.W 3.96 E-08 1.94 E-08 8.39 E-09' 3.63 E 1.70 E-09 WNW 3.22 E-08 1.56 E-08 6.35 E-09 2.85 E L1.37 E-09 ,

NW 2.83 E-08 1.35 E-08 5.55 E-09 2.46 E-09 1.18.E-09 NNW 3.24 E-08 1.55 E-08 6.59 E-09 -2.97 E '1;42 E-09

.i

-Distance to Location, in miles 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 -4.5-4.99' N 8.71 E-10 )

5.64 E-10 3.10 E-10 3.37 E-10 2.91 E-10 NNE 9.91 E-10 6.43 E-10 4.44 E-10 3.82 E-10 -3.30 E-10 NE 1.03 E-09 6.65 E-10 4.62 E-10 3.98 E-10 3.43 E-10.

ENE 6.57 E-10 4.22 E-10' 2.96 E-10 2.55 E-10 2.20 E-10 E 6.57 E-10 4.20 E-10 2.96 E '2.55-E-10 2.20 E-10 ESE 6.88 E-10 4.40 E-10 3.09 E-10: 2.66 E-10 2.29 E-10

SE 2.01 E-09 6.48 E-10 4.55 E-10 ~ 3.90 E 3.36 E-10 SSE 1.07 E-09 6.85 E-10 4.79 E-10 4.12 E-10 3.55 E-10 -

5 1.02 E-09 6.49 E-10 4.59 E-10 -3.94 E-10 '3.40 E-10 SSW 1.00 E-09 6.41 E-10 4.50 E-10 3.86 E-10 3.32 E-10 l SW 1.08 E-09 6.90 E-10 4.81 E-10 4.12 E-10: 3.53 E-10' Wsw 1.02 3-09 6.51 E-10 4.53 E -3.87 E-10 3.32 E .;

W 9.00 3-10 5.92 E-10 4.13 E-10 3.54 E-10 3.04 E-10 '

WNW 7.33 E-10 4.95 E-10 3.52 E-10 3.05 E-10 2.65 E-10~

NW G.37 E-10 4.11 E-10 2.91 E-10 2.50 E-10 2.14 E !

NNW 7s56 E-10 4.95 E-10 -3.45 E-10 2.97 E 2.56 E-10 'I Values are in m*2, extracted from Reference 7.

) i l

8-12 Gen. Rev. 13-i a: ...-. - a

FNP-0-M-011 Table 8-5. Annual Average (676) for Ground-Level Releases O Sector N

0.25-0.5 2.50 E-07 Distance to Location, in miles 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 7.84 E-08 2.53 E-08 9.61 E-09 4.28 E-09 NNE 2.48 E-07 7.77 E-08 2.51 E-08 9.53 E-09 NE 4.24 E-09 2.49 E-07 7.80 E-08 2.52 E-08 9.57 E-09 4.26 E-09 ENE 1.69 E-07 5.29 E-08 1.71 E-08 6.48 E-09 2.88 E-09 E 1.69 E-07 5.28 E-08 1.71 E-08 6.48 E-09 2.88 E-09 ESE 1.80 E-07 5.54 E-08 1,79 E-08 6.80 E-09 3.02 E-09 SE 2.75 E-07 8.63 E-08 2.79 E-08 1.06 E-08 SSE 4.71 E-09 3.66 E-07 1.15 E-07 3.71 E-08 1.41 E-08 6.25 E-09 S 3.70 E-07 1.16 E-07 3.75 E-08 1.42 E-08 6.33 E-09 SSW 2.75 E-07 8.62 E-08 2.79 E-08 1.06 E-08 4.70 E-09 SW 2.60 E-07 8.15 E-08 2.64 E-08 1.00 E-08 4.45 E-09 WSW 2.31 E-07 7.24 E-08 2.34 E-08 8.88 E-09 3.95 E-09 W 2.11 E-07 6.61 E-08 2.14 E-08 8.11 E-09 3.61 E-09 WNW 1.83 E-07 5.73 E-08 1.85 E-08 7.02 E-09 3.12 E-09 NW 1.74 E-07 5.45 E-08 1.76 E-08 6.68 E-09 2.97 E-09 NNW 2.13 E-07 6.67 E-08 2.16 E-08 8.19 E-09 3.64 E-09 Sector Distance to Location, in miles 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 4.5-4.99 N 2.22 E-09 1.45 E-09 9.79 E-10 8.27 E-10 6.99 E-10 O NNE NE 2.20 E-09 2.21 E-09 1.43 E-09 1.44 E-09 9.71 E 9.75 E-10 8.20 E-10 8.23 E-10 6.93 E-10 6.96 E-10 ENE 1.50 E-09 9.76 E-10 6.60 E-10 5.58 E-10 4.72 E-10 E 1.50 E-09 9.75 E-10 6.60 E-10 5.57 E-10 4.71 E-10 l ESE 1.57 E-09 1.02 E-09 6.72 E-10 5.85 E-10 4.94 E-10 SE 2.44 E-09 1.59 E-09 1.08 E-10 9.11 E-10

{

7.70 E-10 SSE 3.25 E-09 2.12 E-09 1.43 E-10 1.21 E-10 1.02 E-10 S 3.29 E-09 2.14 E-09 1.45 E-10 1.22 E-09 1.04 E-10 SSW 2.44 E-09 1.59 E-09 1.08 E-10 9.10 E-10 7.69 E-10 SW 2.31 E-09 1.51 E-09 1.02 E-10 8.60 E-10 7.27 E-10 WSW 2.05 E-09 1.34 E-09 9.04 E-10 'i.04 E-10 6.46 E-10 l W 1.87 E-09 1.22 E-09 8.25 E-10 6.97 E .10 5.90 E-10 WNW 1.62 E-09 1.06 E-09 7.15 E-10 6.04 E-10 5.11 E-10 NW 1.54 E-09 1.01 E-09 6.80 E-10 5.75 E-10 4.86 E-10 NNW '1.89 E-09 1.23 E-09 8.34 E-10 7.04 E-10 5.95 E-10 Values are in ac2, extracted from Reference 7.

O 8-13 Gen. Rev. 13 i

l 1!

i' l

' I FNP-0-N-011 i

l i

i 1

seco . . .

1 l , 1 g I / )~

j / / /

l l / / ..

l / / '

/ f s

, f p f'

! / / / /

,.' /

A

p soo l ,

/ --

/ ---

o . . 1 1 c -

) J .V >F A J' J"

{ .f a ,s' / a f /"

1

/ / 7 / / / -

I / // cff f /

^

/// /k f / , , , /

/ / / , / , /

/

// /,f , .

" ;/ ; o l a -

/ 2Y A A 1 s s .r >

i / / /

\ / / '

,/' /

i t

/

! ' O.1 1A 10 100 PLubit TRAVEL DISTANCE (KILOest71158 i

i 1

Range of Vertical Range of vertical-l category Temperature eradient Temperature eradient-a

(*C/100 m) (*F/100 ft) 1 A. AT/A8 < -1.9 ' AT/AS 4 -1.0 j

i 8 .

-1.9 s AT/AI < -1.7 -1.0 s AT/A5 4 -0.9 e

{

/

C ~1.7 s AT/A5 < -1.5 -0.9 s AT/A5 < -0.8 i # :

1 D. -1. 5 ' s AT/A 8 ' < -0. 5 -0.8 s AT/As < -0.3 l l 3 ' 0.5 s AT/A8 < 1.5-

~

=0.3 s AT/AS < 0.0 4

I F 1.5 s AT/A5 < 4.0 0.8 s AT/A8 4 2.2

) e 4.0 s AT/AS 2.2 s AT/AS J

j This graph is reproduced from Reference S (Figure 1).

j Figure 8-1.

Vertical Standard Deviation of Material is a Plume (es.)

I j 8-14 Gen.' Rev. 13 4

l .. . , _ , . .. c - --

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

?

a.

{ -FNP-O-M-011 i

a i

4 1

t 4

l

! l 4

i i !

< l a ~I

$ )

1 to ;

l i . , , , ,,

' * + ,,, ' i 4 g 1

$ I e t ll I l

\

N l i

8

~ l \

j N

h s k 10 i

j 6 ' 6

' i 4

i l i

I

! lll l 1

i l

)

j 0.1 i s.'

i u

DW7Angs 4KIL0tetT3ng h

i a

i i

i l

4 i i

l 4

a This graph is reproduced from Reference 4.

Figure S-2.

i Terrain Recirculation Factor (q) j i

8-15 Gen.-Rev. 13 i.

- - , . . . . - . _ m,.-,-3, . . . . . , ,,.-,....m..~, -

pe ,- , - - , ,

e a

9 1

l FNP-0-M-011 i

i 1

i, i

i

]

i

ta k -% % ,

- - m j es j '

. wu '%

5 %

g %

y u %

i h \

! u 1

e.e '

i

u (

1 u

e.:

T

, e.1 -

J J

j a.i sa gas seu auss )

n.ums vnavat eenssos sounnsvans

, l

i J

+

1 5

1 4

't i

I 1

i ,

This graph is reproduced from Reference 5 (Figure 2). j j Figure 8-3. Flues Depletion Effect for Ground Level Rolesses 3 8-16 Gen. Rev. 13 I

d

.i 4

FNP-0-M-011 l l

~

4 l

1 i .l j .

i i

r 4

I 1A g3 A NSl/TR.i fH E.PA) 4

^q

- DQ w 1u i

%s\s

! _ t w :

, u up a

N f5

\, \

sA '

1 3 4

j e4 (

i

,, \N E3 f

4 e.1 4

) i

! est 1A tea tesA' asea.

. PLussa TRAVEL Des?ANCE subeusTsam i

ef

.f .

I t

j

.i I

i This graph is reproduced from Reference 5 (Figure 3).

Figure 8-4. Flume Depletion Effect for 30-Meter Releases !

s 1, 4-17 Gen. Rev. 13 4

j t i

i FNP-0-M-011 1

}-

i i

]

k i

(

j i

f i

f 3

1 4

1A staate',

)

3 .I NQ as N: T u h,,..

j T3 8" i

E D hk ussstaat, y l 5 4t

'# k g l

u \

e4 h(

e.:

> - 3 i

i s.1 l<

e.1 u - "' "

' Ptunes TaAvat ausrAsses l

f l'

j . -

This graph is reproduced from Reference 5 (Figure 4).

Figure 8-S. Plume Depletion Sffect for 60-Isoter Releases 8-18 Gen. Rev. 13

l I

j rup-o-u-011 1

I l

j -% % %

'% % *=

=,_msvenat al !

es -

g=- \ \

, e3 I

na as.' s %

7 '

$*@, \ \.

lu (FRACTMII RNAANNNS E*1 q

O u

(

\

eJ u

e.i u sa - tea tema sees pumasTRavet ourAmes snaustana a

~

.o g This graph is reproduced fees Reference 5 (Figure 5).

Figure S-6. Plume DepletLon Effoct ior 100-Meter Releases s

8-19 een. Rev. 13

i I 3

J i

J t FNP-0-M-011 i

i IIP 3 1 i 1 i 1e

% w ,

i 1

g s I '

h.

% i N .!

! s j E .N . l 2 E T\

! W sH j s l

M \

1 a

g N, 1 I

X i w mg 3 O i

\

x5 4 5 1 i E 10 4

's i

3 4

i

=

1.

2 y7

l AI IS ISA 1NA 30A PLUISE TRAVEL DISTANCE (KILORIETERS

]

a I

This graph is reproduced from Reference 5 (Figure 6).

Figure S-7. Relative Deposition for Ground-Levei Releases 8-20 Gen. Rev. 13-

I i

I 1,,.--,~. . . , . . , . ' - - - - - - _ - _ , , . . . _ _ _ , _ _ , _ , , _ _ _ _ _ _ _ _ , _ , , _ _ , .

FNP-0-M-011

!O M

J a

UNETW.E (A.SlCl io 4 s%

q '

3 l

\m g ( W NEUTRAL e / \\

a r NEUTRALD) y g A sTAats lse.s ;

. - n i N '

t 3 c w O

n r

I , VA '

N s / f \ $ n

!* /

/ N \\

I I

/. Tam. ... , <

[ ,

r b

/

I to-7 ;

8.1 1A - 18A . SesA 3DOJ PLutIE TRAVEL DISTANCE (K8toteETEMB)

This graph is reproduced froan Reference S (Figure ~ 7).

O Figure 8-8. Relative Deposition for 30-Meter Releases 8-21 Gen. Rev. 13

} I q !

1 i

l m -o-x-cii l

t l

1 tir4 d

I i

j #% UNSTABLE (A.Lt3 f \

/

i %-A F '

j , [-

) 10 8 N.UTRAL (0) Nhr 1 : ;.

y I i

I a x W

4 m

f n i e xw k / \ M %. L911'ana g s J \ 'N ls r

/

.. ors \ \

x 3 l- 1H .

E a 1 1 r

I J 1 O I I l 4 m ; '

I k

A I 5 w

[ .3 l 8 i

'I l 10-7 IITA M (E,F,0) l I i l

l I i i i 1

l l

i t

i 1H i 0.1 1A '18 8' 1984 200A -

PLUR$6 TRAVEL DISTANCE (KIL430ETERS l

4 3

This graph is reproduced from Reference 5 (Figure 8).

j Figure 8-9. Relative Deposition for 60-Meter Releases i

i i

4-22 Gen. Rev. 13-i 4

1 - - ...,,5.1_ .~ , , , , _ , , , , .- ..mm,w., , ,-- ,___,3 9.,n,.~,...-y ,,,,94,.

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

4 j rup-0-m-011

!O i

i i I

l 10-4 I i !

d l i UNETABLE (AA,CI I f

b 2 / h w l s

I 3 1 N l ig-a "

/ --h-

)

- - u,

- , s ,

m f

[

f NEUTRAL (D) N s I I \ '

a j / ' '5 i

l E

/ i N N

! /

l, I i

[ I O

I i r i E I J I

E (

a ;

g / STABLE (E.FJN 'I J NO DEPLETION i 10-7 b

I s

I f

)

- I S.1 1A tea ~ tesA assa PLUIEE TRAVEL DISTANCE 9tILOMETEf4 l O This graph is reproduced from Reference 5 (Figure 9).

Figure 8-10. Relative Deposition for 100-Meter (or areater) Releases 4-23 Gen. Rev. 13_

e

{

~ FNP-0-M-011

i. .

CHAPTER 9 a METHODS AND PARAMETERS FOR CALCULATION OF i

GASEOUS EFFLUENT PATHWAY DOSE FACTORS, R 2 L

( aipj 9.1 INHALATION PATHWAY FACTOR !

i ,

i For the inhalation pathway, R,jp ; in (mrem /y) per (yci/m3 ) 'is calculated as t f follows (Reference 1, Section 5.3.1.1):

.i i .!

R alpj "'K 1 * (BR), * (DFA)ay t

(9.1)

, where:

l

.l Kg = the units conversion factor: 106 pCi/yci. '

J (BR)a = the breathing rate of receptor age' group a, in m3 /y, from

, Table 9-5.

4 h (DFA),jj = the inhalation dose factor for receptor age group a, j radionuclide 1, and organ j, in mres/pci, from Table 9-7 '

through Table 9-10.

i 3

. i

'l 1

t N

1 i

4

l i

- i l

l 9-1 Gen. Rev. 13

3 FNP-0-M-011 9.2 GROUND PLANE PATHWAY FACTOR For the ground plane external exposure pathway, Raipj in (m2 .mrom/y) per (yci/s) is calculated as follows (Reference 1, Section 5.3.1.2):

-A f t

~* l R ap jj =

Kg . K2 * (SNE) *

(DEG)sj .

(9.2) g u

i ,

where:

Kg = the units conversion factor: 106 pci/pci.

K2= the units conversion factor 8760 h/y.

t

=

(SHF) the shielding factor due to structure (dimensionless).

The value used for (SHF) is 0.7, from (Reference 3, Table E-15).

(DFG);j = the ground plane dose f actor for radionuclide i and organ '

j, in (mrom/h) per (pci/m ),2 from Table 9-15. Dose factors are.the same for all age groups, and those for ,,

the total body also apply to all organs other than skin. I 1 =

the radioactive decay constant for radionuclide i, in a -I. Values of 1; used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 15.

t= the exposure time, in s. The value used for t is 4.73 x 108 s (= 15 y), from (Reference 1, Section 5.3.1.2).

l 9-2 Gen. Rev. 13 l

)

l i i

FNP-0-M-011-9.3 GARDEN VEGETATION PATHWAY FACTOR For radionuclides other than tritium in the garden vegetation consumption pathway, R,jp ; in (m 2mrom/y) per (yci/s). .is calculated as follows (Reference 1, Section 5.3.1.5):

R aipj *

=

Kg *

(M)aif Yy (11 + Aw) ,

(9.3) {

Ual fL* IL!

Usfg*

a

l'hv i

l.

i where i

Kg = the units conversion factor: 100 pci/yci. !

r= the fraction of deposi*:ed activity retained on the edible -

parts of garden vegetation (dimensionless). The'value used for r is 1.0 for radiciodines and -0.2 for particulates, from (Reference 3, Table E-1).

D Yy =

the areal density (agricultural productivity) of growing leafy garden vegetation, in kg/m2 , from Table 9-1.

lj= the radioactive decay constant for radionuclide i, in j s -I. Values of 1 used in' effluent calculations should be I based on decay data from a recognized and current source, j such as Reference 15.

(= the rate constant for removal of activity on leaf and j plant surfaces by weathering, in s-I, from Table 9-1.

l l

(DFL)q = the ingestion dose factor for receptor age group a, radionuclide 1, and organ j, in mrem /pci, from Table 9-11 through Table 9-14. '

l U t, = the consumption rate of fresh leafy garden vegetation by a receptor in age group a, in kg/y, from Table 9-5.

9-3 Gen. Rev. 13

_ , _ _ _ , _. -, . -- I

.i 1:;

2 l

FNP-0-M-011' -t Us=

g the consumption rate of stored garden vegetation by a. j i

receptor in age group a, in kg/y, from Table 9-5.

{

ft= the fraction of the annual intake of. fresh leafy garden f vegetation that is grown locally (dimensionless), .from j Table 9-1. '

= 6 f

g the fraction of the -annual intake of stored garden-vegetation that is grown locally (dimensionless), from Table 9-1. t f!

tt= the average time between harvest of fresh leafy garden ,

1 I vegetation and its consumption, in s, from Table 9-1.

thy = h the average time between harvest of stored garden '

vegetation and its consumption, in s, from Table 9-1.

1

. e ih For tritium in the garden vegetation consumption pathway, Rg p; in (mrom/y) . ;

l per (pci/m3 ) is calculated as follows (Reference 1, Section 5. 3.1. 5 ) ,

based on the concentration in air rather than deposition onto the grounds f

Rg; p =

Kg

K3 * (DTL)g; * { Ug fg +

Ua yfg )

0.15 1 OI 1 (9.4}'

-t where: ,

I

'l K3= the units cenversion factor: 103 g/kg.

R=

the absciute humidity of atmospheric air,-in g/m3 , from, Table 9-1.

u q0.7qp = ..the fraction of the mass of total garden vegetation that is water (dimensionless).

O.5 =

the ratio of the specific activity of tritium'in garden vegetation water: to that in atmospheric water:

(dimensionless).

and other parameters are as defined above.

O 9-4 Gen. Rev. 13;

-l i

FNP-O-M-011 Table 9-1.

Miscellaneous Parameters for the Garden Vegetation Pathway The following parameter values are for use in calculating Rapj j for the garden vegetation pathway only. The terms themselves are defined in '

section 9.3.

Parame*.or Value Reference Yy 2.0 kg/m2 Ref. 3, Table E-15 t

1, 5.73 x 10*7 s'I stof. 1, page 33 (14-day half-life) fg 1.0 Ref. 1, page 36 f 0.76 g Ref. 1, page 33 tg 8.6 x 104 s Ref. 3, Table E-15 '

(1 day) thy 5.18 x 106 s Ref. 3, Table E-15 (60 days)

O H 8 g/m 3 Ref. 3 I

l l

O 9-5 Gen. Rev. 13

i e

4 FNP-O-M-011 f 9.4 GRASS-COW-MILK PATHWAY FACTOR

For radionuclides other than tritium in the grass-cow-milk pathway, Rjj 2 ap in (m *mram/y) per (pCL/s) is calculated as follows (Reference 1, Section 5.3.1.3):

j 1

Raip) " Kg* *Op*Uap *Fmi * (DEL)ag; ,

( Ag + Aw) 4

-Aj tg (9.5) !

fp f, + (1 - fp f,) e . . Xg tf

~

i Yp Ys i

j where l j Kg = the units conversion factor: 106 pCi/yci.

e= the fraction of deposited activity retained on the edible parts of vegetation (dimensionless). The value used for ;

j r is 1.0 for radiciodines and 0.2 for particulates, from

)(

i (Reference 3, Table E-1). i 1; = the radioactive decay constant for radionuclide 1, in s -I. Values of 1 used in affluent calculations should be i based on decay data from a recognized and current source, '

such as Reference 15.

d

.i

(= the rate constant for removal of activity on leaf and

, plant surfaces by weathering, in s-I, from Table 9-2.

'I 9p = the cow's consumption rate of feed, in kg/d, from !

p Table 9-2.

Unp =

the crAsumption rate of cow milk by a receptor in age y group a, in L/y, from Table 9-5.

Fg = the stable element transfer coefficient applicable to radionuclide i, for cow's milk, in d/L, from Table 9-6.

l 9-6 Gen. Rev. 13' P

e - 4 '

J i

J I

i FNP-0-M-011 !

(DFL)gj = the ingestion dose factor for receptor age group a, l radionuelide 1, and organ j, in mrem /pci, from Table 9-11 through Table 9-14. '

i f p= the fraction of the year that the cow is on pasture (dimensionless), from Table 9-2.

4 ;

f, = the fraction of the cow's feed that is pasture grass while the cow is on pasture (dimensionless),- from Table 9-2.

Yp= the areal density (agricultural productivity) of growing pasture feed grass, in kg/m 2, from Table 9-2.

Y, = the ' areal density (agricultural productivity) of growing stored feed, in kg/m 2, from Table 9-2.

tg = the transport time from harvest of stored feed to its consumption by the cow, in a, from Table 9-2.

tg = the transport time from consumption of feed by the cow, to consumption of milk b'/ the receptor, in s, from Table 9-2.

For tritium in the grass-cow-milk pathway, Raipj in (mram/Y) Per (pci/m 3) !

' is calculated as follows (Reference 1, Section 5.3.1.5), based on the concentration in air rather than deposition onto the grounds i l

,e,,,

=,,.,,.,,.,,,.,_,.(m>,,,.o.25.(of) (9..)

l where i K3= the units conversion factors lo3 g/kg.

H= the absolute' humidity of atmospheric air, in g/m3 , from Table 9-2.

G 9-7 Gen. Rev. 13

l

'l i- .

i

, FNP-0-M-011 'j

[.

0.75 =

tha fraction of the mass-of total' vcgstation ~ that ' is '

l j i water.(dimensionless).

'l 0.5 =

the ratio of the specific ~ activity .. of ' tritium ' in vegetation water to. 'that in ' atmospheric water I (dimensionless).

and other parameters are as defined above.

.i

-l l

'i

~t t

, .i l- !

l l

f l !

t l

5 l

i t

1 I

9-8 .g,,n. Rev. 13

.j

4 1

l i

FNP-0-M-011

Table 9-2.

Miscellaneous Parameters for the Grass-Cow-Milk Pathway

D O The following parameter values are for use in calculating Rg pj for ' the grass-cow-milk pathway only. The terms themselves are defined in Section

) 9.4.

i Parameter Value Reference

{

r j

1,7 5.73 x 10*I s'I Ref. 1, page 33 (14-day half-life) l Op 50 kg/d Ref. 3, Table E-3 i

f 1.0 p Ref.'1, page 33 1

} f, 1.0 Ref. 1, page 33 i

l Yp O.7 kg/m 2 Ref. 3, Table E-15 i

1 Y, 2.0 kg/m2 Ref. 3, Table E-15 thm 7.78 x 106 s Ref. 3, Table E-15 (90 days) i tg 1,73 x 105 s Ref. 3, Table E-15 (2 days) !

H 8 g/m 3 Ref. 3

^

l t

e l

i

. 4r 9

, 'r r;

e j u

a i

1 1 l

.i

("'

\

9-9 Gen. Rev. 13 1

i .

l.

l FNP-0-M-011 9.5 GRASS-GOAT-MILK PATHWAY FACTOR t

For radionuclides 2

other than tritium in the grass goat-milk pathway,. R,jpj

{ in (m mrom/y) per (yci/s) is calculated as follows (Reference 1, Section 5.3.1.3): i R aipj '

" Kg *

Oy = U ap *I

. mi* (DEL)at]

(?<7) s+ (1 ~ pS sI )

xj ,g . , ~hi If fp i Yp Y, where Kg = the units conversion factor .106 peijpeg, r= the fraction of deposited activity retained on the edible

~

parts of vegetation (dimensionless). The value used for ,.

r is 1.0 for radioiodines and 0.2 for particulates, from d s (Asference 3, Table E-1). i 1 =

the radioactive decay constant for radionuclide i, in s -I. Values of Ag used in effluent calculations should be based on decay data from a recognized and current bource, such as Reference 15.

1, = the rate constant for removal of activity en leaf and plant surfaces by weathering, in e-I,.fcom Table 9-3.

Op = the goat's rate of feed, in kg/d, from Table 9-3.

U,p = the consumption er.te of goat milk by a receptor in age-group a, in L/,, from Table 9-5.

Fg = the L,imole element transfer coefficient applicable' to radionuclide i, for goat's milk, in d/L, from Table 9-6.-

9-10 Gen. Rev. 13

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

.1

.i FNP-0-M-011 (DFL)gj = the ingestion dose factor for receptor age group a, radionuclide i, and organ j, in mrem /pci, from Table 9-11 I through Table 9-14.

=

f p the fraction of the year that the goat'is on pasture (dimensionless), from Table 9-3.

f, = tho ' fraction of the goat's feed that is pasture grass while the goat is on pasture (dimensionless), from Tabl's 9-3.

Y p= the areal density (agricultural productivity) of growing-pasture feed grass, in kg/m 2, from Table 9-3.

Y, = the areal density (agricultural productivity) of growing stored feed, in kg 'at2, from Table,9-3.

tg = the transport time from harvest of stored feed to its consumption by the goat, in e,~from Table 9-3.

tg = the transport time from consumption of feed by the' goat, '

to consumption of milk by the receptor, in s, from ,

Table 9-3.

For tritium in the grass-goat-milk pathway, Rgp; in (mrom/y) per'(pci/m3) is calculated ae follows (Reference 1, Section 5.3.1' 5), based on.the concentration in air rather than deposition onto the grounds l

Rapjj = Kg

K3 Qy

Uap = Emi * (DFL)ajj*0.75*fO*5 (9.8) i#d ,

i where K3= the units conversion factor: 103 g/kg.

H= the absolute humidity of atmospheric air, in g/m3 , from Table 9-3.

O i

-J 9-11 Gen. Rev. 13

I FNP-0-M-011 ;

O.75 = the fraction of the mass of total vegetation that is water (dimensionless).

0.5 = the ratio of the specific -activity of . tritium in ;

vegetation water. to. that in atmospheric l water .j (dimensionless),

and other parameters are as defined above.

1

)

l l

l 1

l

.I I

l I

l i

1 I

J I

i i

9-12 Gen. Rev.-13 l

l' . - . - ., . . _ . . - . . - -

FNP-0-H-011 Table 9-3.

Miscellaneous Parameters for the Grass-Goat-Milk Pathway

( I t%

The following parameter values are for use in calculating Raipj fo.* the grass-goat-milk pathway only. The terms themselves are defined in section 9.5.

1 l

Parameter Value Reference l 1, 5.73 x 10~7 s'I Ref. 1, page 33 (14-day half-life)

Qp 6 kg/d Ref. 3, Table E-3 i

f 1.0 p Ref. 1, page 33 f, 1.0 Ref. 1, page 33 Y

p O.7 kg/m 2 Ref. 3, Table E-15 Y, 2.0 kg/m2 Ref. 3, Tabla E-15 tg 7.78 x 106 s Ref. 3, Table E-15 I n (90 days)

\

tg 1.73 x 105 s. Ref. 3, Table E-15 (2 days)

H 8 g/m3 Ref. 3

\

9-13 can. Rev. 13

1 4

FMP-0-M-011 '

9.6 GRASS-COW-MEAT PATHWAY FACTGR For radionuclides other than tritium in the grass-cow-neat

2 pathway, Rap jj 1

in (m *mrom/y) per (pci/s) is calculated as follows (Re19rence 1, Section

5.3.1.4)

l R aip) #

~ Kg * .

Qy

Uap

Fjg ( M )ay

('*8) fp i s+

(1

pf Is)

xj ,g

.e ~hi ff 1

Yp Ys -

1 i

, where:

1 Kg = the units conversion factor: 106 petjpet, 1,

r= the fraction of deposited acti'rity retained on the edible parts of vegetation (dimensionless). The value used for f

j -

r is 1.0 for radiciodines and 0.2 for particulates, from (Reference 3, Table E-1).

1 =

the radioactive decay constant for radionuclide 1, in 1 s -I. Values of lj used in ef fluent calculations should be 1

based on decay data from a recognized and current source, such as Reference 15. {

1 1, = the rate constant for removal of activity on leaf and plant surf aces by weathering, in s'I, from Table . 9-4.

Op = the- cow's consumption rate of feed, in kg/d, frren l Table 9-4.

)

U,p = the consumption rate of meat by a receptor in age group a, in kg/y, from Table 9-5.

FS= the stable element transfer coefficient applichle. to radionuclide i, for meat, in d/kg, from Table 'J-6.

O 9-14 Gen. Rev. 13-

. u-

1

! FNP-0-M-011

=

~

(DFL)ajj the ingestion dose factor for receptor age group a, 1

radionuclide i, a.nd organ j, in mrem /pci, from Table 9 through Table 9-14.

i

=

f p the fraction of the year that the cow is en pasture (dimensionless), from Table 9-4.

2

) f, = the fraction of the cow's feed that is pasture grass

~

while- the cow is on pasture (dimensionless), from Table 9-4.

i j Y p= the areal density (agricultural productivity) of growing pasture feed grass, in kg/m 2, from Table 9-4.

Y, = the areal density (agricultural productivity) of growing 1

4 stored feed, in kg/m 2, from Table 9-4.

l i

tg = the transport time from harvest of stored feed to its 4

consumption by the cow, in s, from Table 9-4.

tg = the transport time from consumption of feed by the cow,

~

i 1

to consumption of meat by the receptor, in s, from Table 9-4. ,

For tritium in the grass-cow-meat pathway, R,;pj in (mrom/y) per (pci/m3) is calculated as follows (Reference 1, Section 5.3.1. 4) , based on the concentration in air rather than deposition onto the grounds i

Ralp)

M1'K3

Oy

Uap

Ty * (DFL)ag;

0. 75

l 0 * ' Y (9.10}'

>n>

where:

i 4

the units conversion factors K3= 103 g/kg.

H=

the absolute humidity of atmospheric air, in g/m3, from Table 9-4.

t

%J i

9-15 Gen. Rev. 13 i

m _ . -

i ,

FNP-0-M-011 0.75 = the fraction of the mass of total vegetation that'is water (dimensionless).

0.5 = the ' ratio of the specific activity of tritium in vegetation water to that in atmospheric. water (dimensionless). l

- and other parameters are as defined above.

I l

l l

l i

i 1

-I I

1 I

i l

4 t

l

(

'l

?

l l

I i

l l I

l 9-16 Gen. Rev.:13 i,

, , , -~ -- -~ ._ .- ,- . _ . _ , . . _ _.-,..m,,

i 1

4 FNP-0-M-011 Table 9-4.

Miscellaneous Parameters for the Grass-Cow-Meat Pathway

!O The following parameter values are for use in calculating Raipj . for the .

f grass-cow-meat pathway only. The terms themselves are defined in Section 9.6.

d 3

4 Parameter Value Reference 1, 5.73 x 10"7 s*I Ref. 1, page 33 (14-day half-life) i' QF 50 kg/d -

Ref. 3, Table E-3 1

fp 1.0 Ref. 1, page 33

) f, 1.0 Ref. 1, page 33

Yp O.7 kg/m 2 Ref. 3, Table E-15 Y, 2.0 kg/m 2 Ref. 3, Table E-15 thm 7.78 x 100 s Ref. 3, Table E-15 (90 days) tg 1.73 x 106 s Ref. 3, Table E-15 (20 days)

H 8 g/m3 ' Ref. 3 -

)

l l

l O

9-17 Gen. Rev. 13 l

1

FNP-0-M-011 Tablo 9-5. Individual Ucaqs Factore

(,,/

Receptor Age Group Usage Factor Infant Child Teenager Adult Milk Consumption Rate, U,P 330 330 400 310 (L/Y)

Meat consumption Rats, U aP 0 41 65 110 (kg/Y)

Fresh Leafy Garden Vegetation Consumption Rate, UgL 0 26 42 64 (kg/Y)

Stored Carden Vegetation Consumption Rate, U gs 0 520 630 520 (kg/y)

Breathing Rate, (BR)a (m 3 /y) 1400 3700 8000 8000 O

1 1

l 1

All values are from Reference 3, Table E-5.

I 6

\

9-18 Gen. Rev. 13

FNP-O-M-011 Table 9-6. Stable Element Transfer Data b

h Cow Milk Goat Milk Meat Element y m (d/L)* Fm (d/L)* Fg (d/kg)*

H 1.0 E-02 1.7 E-01 1.2 E-02 C 1.2 E-02 1.0 E-01 3.1 E-02 Na 4.0 E-02 4.0 E-02 3.0 E-02 P 2.5 E-02 2.5 E-01 4.6 E-02 Cr 2.2 E-03 2.2 E-03 2.4 E-03 Mn 2.5 E-04 2.5 E-04 8.0 E-04 Fe 1.2 E-03 1.3 E-04 4.0 E-02 Co 1.0 E-03 1.0 E-03 1.3 E-02 Ni 6.7 E-03 6.7 E-03 5.3 E-02 Cu 1.4 E-02 1.3 E-02 8.0 E-03 Zn 3.9 E-02 .3.9 E-02 3.0 E-02 Br 5.0 E-02 5.0 E-02 2.6 E-02 Rb 3.0 E-02 3.0 E-02 3.1 E-02 Sr 8.0 E-04 1.4 E-02 6.0 E-04 Y 1.0 E-05 1.0 E-05 4.6 E-03 Zr 5.0 E-06 5.0 E-06 3.4 E-02 Nb 2.5 E*03 2.5 E-03 2.8 E-01 Mo 7.5 E-03 7.5 E-03 8.0 E-03 Tc 2.5 E-02 2.5 E-02 4.0 E-01 Ru 1.0 E 1.0 E-06 4.0 E-01 Rh 1.0 E-02 1.0 E-02 1.5 E-03 Ag 5.0 E-02 5.0 E-02 Sb 1.5 E-03 1.7 E-02 1.5 E-03 4.0 E-03 To 1.0 E-03 1.0 E-03 7.7 E-02 I 6.0 E-03 6.0 E-02 2.9 E-03 Cs 1.2 E-02 3.0 E-01 Ba 4.0 E-03 4.0 E-04 4.0 E-04 3.2 E-03 La 5.0 E-06 5.0 E-06 Ca 2.0 E-04 1.0 E-04 1.0 E-04 1.2 E-03 Pr 5.0 E-06 5.0 E-06 4.7 E-03 Nd 5.0 E-06 5.0 E-06 W 3.3 E-03 5.0 E-04 5.0 E-04 1.3 E-03 Wp 5.0 E-06 5.0 E-06 2.0 E-04 Values from Reference 3 (Table E-1) except as follows:

Reference 2 (Table C-5) for Br and Sb.

+

Values from Reference 3, Table E-2 for H, C, P, Fe, Cu, Sr, I, and Cs in goat milk, and Table E-1 for all other elements in cow milk, except as_ follows:

Reference 2 (Table C-5) for Br and Sb in cow stilk.

9-19 Gen. Rev. 13

l 4

FNP-0-M-011 Table 9-7.

Inhalation Dose Factors for the Infant Age Group Nuclida 8cne T. Body

v Liver Thyroid Kidney Lung GI-LLI H-3 No Data 4.62E-07 4.62E-07 4.62E-07 4.62E-07 4.62E-07 4.62E-07 C-14 1.89E-05 3.79E-06 3.79E-06 3.79E-06 3.79E-06 3.79E-06 3.79E-06 Na-24 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 7.54E-06 P-32 1.45E-03 8.03E-05 5.53E-05 No Data No Data No Data 1.15E-05 Cr-51 No Data No Data 6.39E-08 4.11E-08 9.45E-09 9.17E-06 2.55E-07 Mn-54 No Data 1.81E-05 3.56E-06 No Data 3.56E-06 7.14E-04 5.04E-06 Mn-56 No Data 1.10E-09 1.58E-10 No Data 7.86E-10 8.95E-06 5.12E-05 Fe-55 1.41E-05 8.39E-06 2.3BE-06 No Data No Data 6.21E-05 7.82E-07 Fe-59 9.69E-06 1.68E-05 6.77E-06 No Data No Data 7.25E-04 1.77E-05 Co-58 No Data 8.71E-07 1.30E-06 No Data No Data 5.55E-04 7.95E-06 Co-60 No Data 5.73E-06 8.41E-06 No Data No Data 3.22E-03 2.28E-05 Ni-63 2.42E-04 1.46E-05 8.29E-06 No Data No Data 1.49E-04 1.73E-06 Ni-65 1.71E-09 2.03E-10 8.79E-11 No Data No Data 5.80E-06 3.58E-05 Cu-64 No Data 1.'34E-09 5.53E-10 No Data 2.84E-09 6.64E-06 1.07E-05 Zn-65 1.38E-05 4.47E-05 2.22E-05 N No Data 2.32E-05 4.62E-04 3.67E-05 Zn-69 3.85E-11 6.91E-11 5.13E-12 No Data 2.87E-11 1.05E-06 9.44E-06 Be-83 No Data No Data 2.72E-07 No Data No Data No Data No Data Br-84 No Data No Data 2.86E-07 No Data No Data No Data No Data Br-85 No Data No Data 1.46E-08 No Data No Data No Data No Data Rb-86 No Data 1.36E-04 6.30E-05 No Data No Data No Data 2.17E-06

)

Rb-88 No Data 3.98E-07 2.05E-07 No Data No Data No Data 2.42E-07 l

Rb-89 No Data 2.29E-07 1.47E-07 No Data No Data No Data 4.87E-08 Sr-89 2.84E-04 No Data 8.15E-06 No Data No Data 1.45E-03 4.57E-05 Sr-90 2.92E-02 {

No Data 1.85E-03 No Data No Data 8.03E-03 9.36E-05 !

Sr-91 6.83E-08 No Data 2.47E-09 No Data No Data 3.76E-05 5.24E-05

All values are in (mrom/pc1 inhaled). They are obtained from Reference 3 (Table E-10). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-125.

!/O t) 9-20 Gen. Rev. 13

FNP-0-M-011 j

Table 9-7 (contd). Inhalation Dose Factors for the Infant Age Group b

4V 1

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI l Sr-92 7.50E-09 No Data 2.79E-10 No Data No Data 1.70E-05 1.00E-04 Y-90 2.35E-06 No Data 6.30E-08 No Data No Data 1.92E-04 7.43E-05 Y-91m 2.91E-10 No Data 9.90E-12 No Data No Data 1.99E-06 1.68E-06 j Y-91 4.20E-04 No Data 1.12E-05 No Data No Data 1.75E-03 5.02E-05 Y-92 1.17E-08 No Data 3.29E-10 No Data No Data 1.75E-05 9.04E-05 Y-93 1.07E-07 No Data 2.91E-09 No Data No Data 5.46E-05 4 1.19E-04 Zr-95 8.24E-05 1.99E-05 1.45E-05 No Data 2.22E-05 1.25E-03 1.55E-05

! Zr-97 1.07E-07 1.83E-08 8.36E-09 No Data 1.85E-08 7.88E-05 1.00E-04 Nb-95 1.12E-05 4.59E-06 2.70E-06 No Data 3.37E-06 3.42E-04 9.05E-06 Mo-99 No Data 1.18E-07 2.31E-08 No Data 1.89E-07 9.63E-05 3.48E-05 Tc-99m 9.98E-13 2.06E-12 2.66E-11 No Data 2.22E-11 5.79E-07 1.45E-06 Tc-101 4.65E-14 5.88E-14 5.80E-13 No Data 6.99E-13 4.~17E-07 6.03E-07 Ru-103 1.44E-06 No Data 4.85E-07 No Data 3.03E-06 3.94E-04 1.15E-05 Ru-105 8.74E-10 No Data 2.93E-10 No Data 6.42E-10 1.12E-05 3.46E-05 Ru-106 6.20E-05 No Data 7.77E-06 No Data 7.61E-05 8.26E-03 1.17E-04 Rh-105 No Data No Data No Data No Data No Data No Dais No Data Ag-110m 7.13E-06 5.16E-06 3.57E-06 No Data 7.80E-06 2.62E-03 2.36E-05 Sb-124 No Data No Data No Data No Data No Data l

No Data No Data j Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 3.40E-06 1.42E-06 4.70E-07 1.16E-06 No Data 3.19E-04 9.22E-06 Te-127m 1 195-05 4.935-06 1.48E-06 3.48E-06 2.68E-05 9.37E-04 1.95f-05 Te-127 1.593Mp 6.81E-10 3.49E-10 1.32E-09 3.47E-09 7.39E-06 1.74E-0 Te-129m 1.015-05 4.355-06 1.59E-06 3.91E-06 2.27E-05 1.20E-03 4.93E-05 Te-129 5.63E-11 2.483-11 1.34E-11 4.82E-11 1.25E-10

. 2.14E-06 1.88E-05 Te-131m 7.62E-08 3.93E-08 2.59E-08 6.38E-08 1.89E-07 1.42E-04 8.51E-05 Te-131 1.24E-11 5.87E-12 3.57E-12 1.13E-11 2.85E-11 1.47E-06 5.87E-06 9-21 Gen. Rev. 13

.- - w

i 4

FNP-0-M-011 Table 9-7 (contd). Inhalation Dose Factors for the Infant Age Group 1

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.66E-07 1.69E-07 1.26E-07 1.99E-07 7.39E-07 2.43E-04 3.15E-05 I-130 4.54E-06 9.91E-06 3.98E-06 1.14E-03 1.09E-05 No Data 1.42E-06 I-131 2.71E-05 3.17E-05 1.40E-05 1.06E-02 3.70E-05 No Data 7.56E-07 7

I-132 1.21E-06 2.53E-06 8.99E-07 1.21E-04 2.82E-06 No Data 1.36E-06 I-133 9.46E-06 1.37E-05 4.00E-06 2.54E-03 1.60E-05 No Data 1.54E-06 I-134 6.58E-07 1.34E-06 4.75E-07 3.18E-0,5 1.49E-06 No Data 9.21E-07 1-135 2.76E-06 5.43E-06 1.98E-06 4.97E-04 6.05E-06 No Data 1.31E-06

! Cs-134 2.83E-04 5.02E-04 5.32E-05 No Data 1.36E-04 5.69E-05 9.53E-07 Cs-136 3.45E-05 9.61E-05 3.78E-05 No Data 4.03E-05 8.40E-06 1.02E-06 Cs-137 3.92E-04 4.37E-04 3.25E-05 No Data 1.23E-04 5.09E-05 9.53E-07 Cs-138 3.61E-07 5.58E-07 2.84E-07 No Data 2.93E-07 4.67E-08 6.26E-07 Ba-139 1.06E-09 7.03E-13 3.07E-11 No Data 4.23E-13 4.25E-06 3.64E-05

[ Ba-140 4.00E-05 4.00E-08 2.07E-06 No Data 9.59E-09 1.14E-03 2.74E-05

\

Ba-141 1.12E-10 7.70E-14 3.55E-12 No Data 4.64E-14 2.12E-06 3.39E-06 Ba-142 2.84E-11 2.36E-14 1.40E-12 No Data 1.36E-14 1.11E-06 4.95E-07 La-140 3.61E-07 1.43E-07 3.68E-08 No Data No Data 1.20E-04 6.06E-05 La-142 7.36E-10 2.69E-10 6.46E-11 No Data No Data 5.87E-06 4.25E-05 Co-141 1.98E-05 1.19E-05 1.42E-06 No Data 3.75E-06 3.69E-04 1.54E-05 i

co-143 2.09E-07 1.38E-07 1.58E-08 No Data 4.03E-08 8.30E-05 3.55E-05

Co-144 2.28E-03 8.65E-04 1.26E-04 No Data 3.84E-04 7.03E-03 1.06E-04 i

Pr-143 1.005-05 3. 74 E-(J 6 4.99E-07 No Data 1.41E-06 3.09E-04 2.66E-05 Pr-144 3.42E-11 1.32E-11 1.72E-12 No Data 4.80E-12 1.15E-06 3.06E-06 Nd-147 5.67E-06 5.812-06 3.57E-07 No Data 2.25E-06 2.30E-04 2.23E-05 W-187 9.26E-09 6.44E-09 2.23E-09 No Data No Data 2.83E-05 2.54E-05 Np-239 2.65E-07 2.37E-08 1.34E-08 No Data 4.73E-08 4.25E-05 1.78E-05 9-22 Gen. Rev. 13

FNP-0-M-011 Table 9-8. Inhalation Dose Factors for the Child Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07 3.04E-07 C-14 9.70E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 1.82E-06 Na-24 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 4.35E-06 P-32 7.04E-04 3.09E-05 2.67E-05 No Data No Data No Data 1.14E-05 Cr-51 No Data No Data 4.17E-08 2.31E-08 6.57E-09 4.59E-06 2.93E-07 Mn-54 No Data 1.16E-05 2.57E-06 No Data 2.71E-06 4.26E-04 6.19E-06 Mn-56 No Data 4.48E-10 8.43E-11 No Data 4.52E-10 3.55E-06 3.33E-05 Fe-55 1.28E-05 6.80E-06 2.10E-06 No Data No Data 3.00E-05 7.75E Fe-59 5.59E-06 9.04E-06 4.51E-06 No Data No Data 3.43E-04 1.91E-05 Co-58 No Data 4.79E-07 8.55E-07 No Data No Data 2.99E-04 9.29E-06 Co-60 No Data 3.55E-06 6.12E-06 No Data No Data 1.91E-03 2.60E-05 Ni-63 2.22E-04 1.25E-05 7.56E-06 No Data No Data 7.43E-05 1.71E-06 Ni-65 8.08E-10 7.99E-11 4.44E-11 No Data No Data 2.21E-06 2.27E-05 Cu-64 No Data 5.39E-10 2.90E-10 No Data 1.63E-09 2.59E-06 9.92E-06

'E Zn-65 1.15E-05 3.06E-05 1.90E-05 No Data 1.93E-05 2.69E-04 4.41E-06 Zn-69 1.81E-11 2.61E-11 2.41E-12 No Data 1.58E-11 3.84E-07 2.75E-06 Br-83 No Data No Data 1.28E-07 No Data No Data No Data No Data Br-84 No Data No Data 1.48E-07 No Data No Data No Data No Data Br-85 No Data No Data 6.84E-09 No Data No Data No Data No Data Rb-86 No Data 5.36E-05 3.09E-05 No Data No Data No Data 2.16E-06 Rb-88 No Data 1.52E-07 9.90E-08 No Data No Data No Data 4.66E-09 Rb-89 No Data 9.33E-08 7.83E-08 No Data No Data No Data 5.11E-10 Sr-89 1.62E-04 No Data 4.66E-06 No Data No Data 5.83E-04 4.52E-05 Sr-90 2.73E-02 No Data 1.74E-03 No Data No Data 3.99E-03 9.2BE-05 se-91 3.28E-08 No Data 1.24E-09 No Data No Data 1.44E-05 4.70E-05 All values are in (mrom/pci inhaled) . They are obtained from Reference 3 (Table E-9). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-125.

\

.9-23 Gen. Rev. 13

FNP-0-M-011 Table 9-8 (contd). Inhalation Dose Factors for the Child Age Group O Nuclide Bone Liver T. Body Thyroid Kidney Lung OI-LLI Sr-92 3.54E-09 No Data 1.42E-10 No Data No Data 6.49E-06 6.55E-05 Y-90 1.11E-06 No Data 2.99E-08 No Data No Data 7.07E-05 7.24E-05 Y-91m 1.37E-10 No Data 4.98E-12 No Data No Data 7.60E-07 4.64E-07 Y-91 2.47E-04 No Data 6.59E-06 No Data No Data 7.10E-04 4.97E-05 Y-92 5.50E-09 No Date 1.57E-10 No Data No Data 6.46E-06 6.46E-05 Y-93 5.04E-08 No Data 1.38E-09 No Data No Data 2.01E-05 1.05E-04 Zr-95 5.13E-05 1.13E-05 1.OOE-05 No Data 1.61E-05 6.03E-04 1.65E-05 Zr-97 5.07E-08 7.34E-09 4.32E-09 No Data 1.05E-08 3.06E-05 9.49E-05 Nb-95 6.35E-06 2.48E-06 1.77E-06 No Data 2.33E-06 1.66E-04 1.OOE-05 Mo-99 No Data 4.66E-08 1.15E-08 No Data 1.06E-07 3.66E-05 3.42E-05 Tc-99m 4.81E-13 9.41E-13 1.56E-11 No Data 1.37E-11 2.57E-07 1.30E-06 Tc-101 2.19E-14 2.30E-14 2.91E-13 No Data 3.92E-13 1.58E-07 4.41E-09 Ru-103 7.55E-07 No Data 2.90E-07 No Data 1.90E-06 1.79E-04 1.21E-05 Ru-105 4.13E-10 No Data 1.50E-10 No Data 3.63E-10 4.30E-06 2.69E-05 Q)

Ru-106 3.68E-05 No Data 4.57E-06 No Data 4.97E-05 3.87E-03 1.16E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 4.56E-06 3.08E-06 2.47E-06 No Data 5.74E-06 1.48E-03 2.71E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 1.82E-06 6.29E-07 2.47E-07 5.20E-07 No Data 1.29E-04 9.13E-06 Te-127m 6.72E-06 2.31E-06 8.16E-07 1.64E-06 1.72E-05 4.OOE-04 1.93E-05 Te-127 7.495-10 2.573-10 1.65E-10 5.30E-10 1.91E-09 2.71E-06 1.52E-05 Te-129s 5.195-06 1.85E-06 8.22E-07 1.71E-06 1.36E-05 4.76E-04 4.912-05 Te-129 2.64E-11 9.45E-12 6.44E-12 1.93E-11 6.94E-11 7.93E-07 6.89E-06 Te-131m 3.63E-08 1.60E-OB 1.37E-08 2.64E-08 1.08E-07 5.56E-05 8.32E-05 Te-131 5.87E-12 2.28E-12 1.78E-12 4.59E-12 1.59E-11 5.55E-07 3.60E-07 O

9-24 Gen. Rev. 13

3 FNP-0-M-011 Table 9-8 (contd). Inhalation Dose Factors for the Child Age Group p

d l Nuclide Bone Liver T. Body Thyroid Kidney Lung CI-LLI Te-132 1.30E-07 7.36E-08 7.12E-08 8.58E-08 4.79E-07 1.02E-04 3.72E-05 I-130 2.21E-06 4.43E-06 2.28E-06 4.99E-04 6.61E-06 No Data 1.38E-06 j I-131 1.30E-05 1.30E-05 7.37E-06 4.39E-03 2.13E-05 No Data 7.68E-07 I-132 5.72E-07 1.10E-06 5.07E-07 5.23E-05 1.69E-06 No Data 8.65E-07 I-133 4.4BE-06 5.49E-06 2.08E-06 1.04E-03 9.13E-06 No Data l'.48E-06 3.17E-07 5.84E-07 2.69E-07 1.37E-05 8.92E-07 l

I-134 No Data 2.58E-07 I-135 1.33E-06 2.36E-06 1.12E-06 2.14E-04 3.62E-06 No Data !

1.20E-06 i Cs-134 1.76E-04 2.74E-04 6.07E-05 No Data 8.93E-05 3.27E-05 1.04E-06 Cs-136 1. 7 6E -05 4.62E-05 3.14E-05 No Data 2.58E-05 1 3.93E-06 1.13E-06 Cs-137 2.45E-04 2.23E-04 3.47E-05 No Data 7.63E-05 2.81E-05 9.78E-07

} Cs-138 1.71E-07 2.27E-07 1.50E-07 No Data 1.68E-07 1.84E-08 7.29E-08 Ba-139 4.98E-10 2.66E-13 1.45E-11 No Data 2.33E-13 1.56E-06 1.56E-05 Ba-140 2.00E-05 1.75E-08 1.17E-06 No Data 5.71E-09 4.71E-04 2.75E-05 Ba-141 5.29E-11 2.95E-14 1.72E-12 No Data 2.56E-14 7.89E-07 7.44E-08 Ba-142 1.35E-11 9.73E-15 7.54E-13 No Data 7.87E-15 4.44E-07 7.41E-10 La-140 1.74E-07 6.08E-08 2.04E-08 No Data No Data 4.94E-05 6.10E-05 La-142 3.50E-10 1.11E-10 3.49E-11 No Data No Data 2.35E-06 2.05E-05 Co-141 1.06E-05 5.28E-06 7.83E-07 No Data 2.31E-06 1.47E-04 1.53E-05 Co-143 9.89E-08 5.37E-08 7.77E-09 No Data 2.26E-08 3.12E-05 3.44E-05 Co-144 1.83E-03 5.72E-04 9.77E-05 No Data 3.17E-04 3.23E-03 1.05E-04 Pr-143 4.99E-06 1.50E-06 2.47E-07 No Data 8.11E-07 1.17E-04 2.63E-05 Pr-144 1.61E-11 4.99E-12 8.10E-13 No Data 2.64E-12 4.23E-07 5.32E-08 Nd-147 2.92E-06 2.36E-06 1.84E-07 No Data 1.30E-06 8.87E-05 2.22E-05 W-187 4.41E-09 2.61E-09 1.17E-09 No Data No Data 1.11E-05 2.46E-05 Np-239 1.26E-07 9.04E-09 6.35E-09 No Data 2.63E-08 1.57E-05 1.73E-05 m

V 9-25 Gen. Rev. 13

1 FNP-0-M-011 1

Table 9-9. Inhalation Does Factors for the Teenager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.59E-07 1.59E-07 1.59E-07 1.59E-07 1.59E-07 1.59E-07 C-14 3.25E-06 6.09E-07 6.09E-07 6.09E-07 6.09E-07 6.09E-07 6.09E-07 Na-24 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 P-32 2.36E-04 1.37E-08 8.95E-06 No Data No Data No Data 1.16E-05 Cr-51 No Data No Data 1.69E-08 9.37E-09 3.84E-09 2.62E-06 3.75E-07 a

Mn-54 No Data 6.39E-06 1.05E-06 No Data 1.59E-06 2.48E-04 8.35E-06 Mn-56 No Data 2.12E-10 3.15E-11 No Data 2.24E-10 1.90E-06 7.18E-06 Fe-55 4.18E-06 2.98E-06 6.93E-07 No Data No Data 1.55E-05 7.99E-07 Fe-59 1.99E-06 4.62E-06 1.79E-06 No Data No Data 1.91E-04 2.23E-05 Co-58 No Data 2.59E-07 3.47E-07 No Data No Data 1.68E-04 1.19E-05 Co-60 No Data 1.89E-06 2.48E-06 No Data No Data 1.09E-03 3.24E-05 NL-63 7.25E-05 5.43E-06 2.47E-06 No Data No Data 3.64E-05 1.77E-06 Ni-65 2.73E-10 3.66E-11 1.59E-11 No Data No Data 1.17E-06 4.59E-06

[~'

4 Cu-64 No Data 2.54E-10 1.06E-10 No Data 8.01E-10 1.39E-06 7.68E-06 g

Zn-65 4.82E-06 1.67E-05 7.80E-06 No Data 1.08E-05 1.55E-04 5.83E-06 1

Zn-69 6.04E-12 1.15E-11 8.07E-13 No Data 7.53E-12 1.98E-07 3.56E-08 Br-83 No Data No Data 4.30E-08 No Data No Data No Data No Data

Br-84 No Data No Data 5.41E-08 No Data No Data No Data No Data

Br-85 No Data No Data 2.29E-09 No Data No Data No Data No Data Rb-86 No Data 2.38E-05 1.05E-05 No Data No Data No Data 2.21E-06 Rb-88 No Data 6.825-08 3.40E-08 No Data i

No Data No Data 3.65E-15 Rb-89 30 Data 4.405-08 2.91E-08 No Data No Data No Data 4.22E-17 Sr-89 5.43E-05 No Data 1.56E-06 No Data No Data 3.02E-04 4.64E-05 Sr-90 1.353-02 No Data 8.35E-04 No Data No Data 2.06E-03 9.56E-05 Sr-91 1.10E-08 No Data 4.39E-10 No Data No Data 7.59E-06 3.24E-05 All values are in (mrom/pci inhaled). They are obtained from Reference 3 (Table E-8). Neither Reference 2 nor Reference 3 ,

contains data for Rh-105, sb-124, or Sb-125.

O 9-26 Gen. Rev. 13

(

FNP-0-M-011 Table 9-9 (contd). Inhalation Dose Factors for the Teenager Age Group O

V Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI j Sr-92 1.19E-09 No Data 5.08E-11 No Data No Data 3.43E-06 1.49E-05 i Y-90 3.73E-07 No Data 1.00E-08 No Data No Data 3.66E-05 6.99E-05 '

Y-91m 4.63E-11 No Data 1.77E-12 No Data No Data 4.00E-07 3.77E-09 Y-91 8.26E-05 No Data 2.21E-06 No Data No Data 3.67E-04 5.11E-05 Y-92 1.84E-09 No Data 5.36E-11 No Data No Data 3.35E-06 2.06E-05

, Y-93 1.69E-08 No Data 4.65E-10 No Data i

No Data 1.04E-05 7.24E-05 Zr-95 1.82E-05 5.73E-06 3.94E-06 No Data 8.42E-06 3.36E-04 1.86E-05 Zr-97 1.72E-08 3.40E-09 1.57E-09 No Da'ta 5.15E-09 1.62E-05 7.88E-05 !

Hb-95 2.32E-06 1.29E-06 7.08E-07 No Data 1.25E-06 9.39E-05 i

1.212-05 Mo-99 No Data 2.11E-08 4.03E-09 No Data 5.14E-08 1.92E-05 3.36E-05 Tc-99m 1.73E-13 4.83E-13 6.24E-12 No Data 7.20E-12 1.44E-07 1

7.66E-07 Tc-101 7.40E-15 1.05E-14 1.03E-13 No Data 1.90E-13 8.34E-08 1.09E-16 Ru-103 2.63E-07 No Data 1.12E-07 No Dae- 9.29E-07 9.79E-05 1.36E-05 i Ru-105 1.40E-10 No Data 5.42E-11 No Data 1.76E-10 2.27E-06 1.13E-05 Ru-106 1.23E-05 No Data 1.55E-06 No Data 2.38E-05 2.01E-03 1.20E-04 i Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 1.73E-06 1.64E-06 9.991-07 No Data 3.13E-06 8.44E-04 3.41E-05 Sb-124 No Data No Data No Data No Data No Data 1

No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 6.10E-07 2.80E-07 8.34E-08 1.75E-07 No Data 6.70E-05 9.38E-06 !

Te-127m 2.25E-06 1.02E-06 2.73E-07 5.48E-07 8.17E-06 2.07E-04 1.99E-05 Te-127 2.51E-10 1.145-10 5.52E-11 1.77E-10 9.10E-10 1.40E-06 1.01E-05 Te-129m 1.74E-06 8.23E-07 2.81E-07 5.72E-07 6.49E-06 2.47E-04 5.06E-05 Te-129 8.87E-12 4.22E-12 2.20E-12 6.485-12 3.322-11 4.12E-07 2.02E-07 Te-131m 1.23E-08 7.51E-09 5.03E-09 9.06E-09 5.49E-08 2.97E-05 7.76E-05 Te-131 1.97E-12 1.04E-12 6.30E-13 1.55E-12 7.72E-12 2.92E-07 1.89E-09 L

9-27 Gen. Rev. 13

FNP-0-M-011 Table 9-9 (contd). Inhalation Dose Factors for the Teenager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 4.50E-OB 3.63E-08 2.74E-08 3.07E-08 2.44E-07 5.61E-05 5.79E-05 I-130 7.80E-07 2.24E-06 8.96E-07 1.86E-04 3.44E-06 No Data 1.14E-06 I-131 4.43E-06 6.14E-06 3.30E-06 1.82E-02'i.05E-05 No Data 8.11E-07 I-132 1.99E-07 5.47E-07 1.97E-07 1.89E-05 8.65E-07 No Data 1.59E-07 I-133- 1.52E-06 2.56E-06 7.78E-07 3.65E-04 4.49E-06 No Data 1.29E-06 I-134 1.11E-07 2.90E-07 1.05E-07 4.94E-06 4.58E-07 No Data 2.55E-09 I-135 4.62E-07 1.18E-06 4.36E-07 7.76E-05 1.86E-06 No Data 8.69E-07 Cn-134 6.28E-05 1.41E-04 6.86E-05 No Data 4.69E-05 1.83E-05 1.22E-06 Cs-136 6.44E-06 2.42E-05 1.71E-05 No Data 1.38E-05 2.22E-06 . 361 ,$

Cs-137 8.38E-05 1.06E-04 3.89E-05 No Data 3.80E-05 1.51E-05 1.Yd-06 Cs-138 5.82E-08 1.07E-07 5.58E-08 No Data 8.28E-08 9.84E-09 3.38E-11 Ba-139 1.67E-10 1.18E-13 4.87E-12 No Data 1.11E-13 8.08E-07 8.06E-07 Ba-140 6.84E-06 8.38E-09 4.40E-07 O Ba-141 1.78E-11 1.32E-14 5.93E-13 No Data No Data 2.85E-09 2.54E-04 1.23E-14 4.11E-07 2.86E-05 9.33E-14 Ba-142 4.62E-12 4.63E-15 2.84E-13 No Data 3.92E-15 2.39E-07 5.99E-20 La-140 5.99E-08 2.95E-08 7.82E-09 No Data No Data 2.68E-05 6.09E-05 La-142 1.20E-10 5.31E-11 1.32E-11 No Data No Data 1.27E-06 1.50E-06 Co-141 3.55E-06 2.37E-06 2.71E-07 No Data 1.11E-06 7.67E-05 1.58E-05 Co-143 3.32E-08 2.42E-08 2.70E-09 No Data 1.08E-08 1.63E-05 3.19E-05 Co-144 6.11E-04 2.53E-04 3.28E-05 No Data 1.51E-04 1.67E-03 1.08E-04 i Pr-143 1.67E-06 6.64E-07 8.28E-08 No Data 3.86E-07 6.04E-05 2.67E-05 !

Pr-144 5.37E-12 2.20E-12 2.72E-13 No Data 1.26E-12 2.19E-07 2.94E-14 Nd-147 9.833-07 1.07E-06 6.41E-08 No Data 6.28E-07 4.65E-05 2.28E-05 W-187 1.50E-09 1.22E-09 4.29E-10 No Data No Data 5.92E-06 2.21E-05 Np-239 4.23E-08 3.99E-09 2.21E-09 No Data 1.25E-08 8.11E-06 1.65E-05 l

i O

N) !

I i

9-28 Gen. Rev. 13

FNP-0-M-011 Table 9-10. Inhalation Dose Factors for the Adult Age Group r"N

( )

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.58E-07 1.58E-07 1.58E-07 1.58E-07 1.58E-07 1.58E-07 C-14 2.27E-06 4.26E-07 4.26E-07 4.26E-07 4.26E-07 4.26E-07 4.26E-07 Na-24 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 1.28E-06 P-32 1.65E-04 9.64E-06 6.26E-06 No Data No Data No Data 1.08E-05 Cr-51 No Data No Data 1.25E-08 7.44E-09 2.85E-09 1.80E-06 4.15E-07 Mn-54 No Data 4.95E-06 7.87E-07 No Data 1.23E-06 1.75E-04 9.67E-06 Mn-56 No Data 1.55E-10 2.29E-11 No Data 1.63E-10 1.18E-06 2.53E-06 Fe-55 3.07E-06 2.12E-06 4.93E-07 No Data No Data 9.01E-06 7.54E-07 Fe-59 1.47E-06 3.47E-06 1.32E-06 No Data No Data 1.27E-04 2.35E-05 Co-58 No Data 1.98E-07 2.59E-07 No Data No Data 1.16E-04 1.33E-05 Co-60 No Data 1.44E-06 1.85E-06 No Data No Data 7.46E-04 3.56E-05 Ni-63 5.40E-05 3.93E-06 1.81E-06 No Data No Data 2.23E-05 1.67E-06 Ni-65 1.92E-10 2.62E-11 1.14E-11 No Data No Data 7.00E-07 1.54E-06

/'Nt Cu-64 No Data 1.83E-10 7.69E-11 e No Data 5.78E-10 8.48E-07 6.12E-06 Zn-65 4.05E-06 1.29E-05 5.82E-06 No Data 8.62E-06 1.08E-04 6.68E-06 Zn-69 4.23E-12 8.14E-12 5.65E-13 No Data 5.27E-12 1.15E-07 2.04E-09 Br-83 No Data No Data 3.01E-08 No Data No Data No Data 2.90E-08 Be-84 No Data No Data 3.91E-08 No Data No Data No Data 2.05E-13 Br-85 No Data No Data 1.60E-09 No Data No Data No Data No Data Rb-86 No Data 1.69E-05 7.37E-06 No Data No Data No Data 2.08E-06 Rb-88 No Data 4.84E-08 2.41E-08 No Data No Data No Data 4.18E-19 Rb-89 No Data 3.20E-08 2.12E-08 No Data No Data No Data 1.16E-21 Sr-89 3.80E-05 No Data 1.09E-06 No Data No Data 1.75E-04 4.37E-05 Sr-90 1.24E-02 No Data 7.62E-04 No Data No Data 1.20E-03 9.02E-05 Sr-91 7.74E-09 No Data 3.13E-10 No Data No Data 4.56E-06 2.39E-05 All values are in (mrom/pci inhaled). They are obtained from Reference for Rh.105,3Sb-124, (Table and E-7),Sb-125.

except as follows: Reference 2 (Table C-1) gO b

9-29 Gen. Rev. 13

1 FNP-0-M-011 Table 9-10 (contd). Inhalation Dose Factors for the Adult Age Grotir J

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI

Sr-92 8.43E-10 No Data 3.64E-11 No Data No Data 2.Deg-06 5.38E-06 Y-90 2.61E-07 No Data 7.01E-09 No Data No Data 2.12E 05 6.32E-05 Y-91m 3.26E-11 No Data 1.27E-12 No Data No Data 2.40E-07 1.66E-10 Y-91 5.78E-05 No Data 1.55E-06 No Data No Data 2.13E-04 4.81E-05 Y-92 1.29E-09 No Data 3.77E-11 No Data No Data 1.96E-06 9.10E-06 Y-93 1.18E-08 No Data 3.26E-10 No Data No Data 6.06E-06 5.27E-05 Zr-95 1.34E-05 4.30E-06 2.91E-06 No Data 6.77E-06 2.21E-04 1.88E-05 Zr-97 1.21E-08 2.45E-09 1.13E-09 No Data 3.71E-09 9.84E-06 6.54E-05 Nb-95 1.76E-06 9.77E-07 5.26E-07 No Data 9.67E-07 6.31E-05 1.30E-05 No-99 No Data 1.51E-08 2.87E-09 No Data 3.64E-08 1.14E-05 3.10E-05 Tc-99m 1.29E-13 3.64E-13 4.63E-12 No Data 5.52E-12 9.55E-08 5.20E-07 Tc-101 5.22E-15 7.52E-15 7.38E-14 No Data 1.35E-13 4.99E-08 1.36E-21 Ru-103 1.91E-07 No Data 8.23E-08 No Data 7.29E-07 6.31E-05 1.38E-05 Ru-105 9.88E-11 No Data 3.89E-11 No Data 1.27E-10 1.37E-06 6.02E-06 Ru-106 8.64E-06 No Data 1.09E-06 No Data 1.67E-05 1.17E-03 1.14E-04 Rh-105 9.24E-10 6.73E-10 4.43E-10 No Data 2.86E-09 2.41E-06 1.09E-05 Ag-110m 1.35E-06 1.25E-06 7.43E-07 No Data 2.46E-06 5.79E-04 3.78E-05 Sb-124 3.90E-06 7.36E-08 1.55E-06 9.44E-09 No Data 3.10E-04 5.08E-05 Sb-125 8.26E-06 8.91E-08 1.66E-06 7.34E-09 No Data 2.75E-04 1.26E-05 Te-125m 4.27E-07 1.98E-07 5.84E-08 1.31E-07 1.55E-06 3.92E-05 8.83E-06 Te-127m 1.58E-06 7.21E-07 1.96E-07 4.11E-07 5.72E-06 1.20E-04 1.87E-05 Te-127 1.755-10 8.03E-11 3.87E-11 1.32E-10 6.37E-10 8.14E-Cl 7.17E-06 Te-129m 1.22E-06 5.84E-07 1.98E-07 4.30E-07 4.57E-06 1.45C-04 4.79E-05 Te-129 6.22E-12 2.99E-12 1.55E-12 4.87E-12 2.34E-11 2.42E-07 1.96E-08 Te-131m 8.74E-09 5.45E-09 3.63E-09 6.88E-09 3.86E-08 1.82E-05 6.95E-05 Te-131 1.39E-12 7.44E-13 4.49E-13 1.17E-12 5.46E-12 1.74E-07 2.30E-09 !

i 9 \

l l

9-30 Gen. Rev. 13 '

i FNP-0-M-011 Table 9-10 (contd). Inhalation Dose Factors for the Adult Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI

. Te-132 3.25E-08 2.69E-08 2.02E-08 2.37E-08 1.82E-07 3.60E-05 e

6.37E-05 i I-130 5.72E-07 1.68E-06 6.60E-07 1.42E-04 2.61E-06 No Data 9.61E-07 I-131 3.15E-06 4.47E-06 2.56E-06 1.49E-03 7.66E-06 i

No Data 7.85E-07 I-132 1.45E-07 4.07E-07 1.45E-07 1.43E-05 6.48E-07 No Data 5.08E-08 I-133 1.08E-06 1.85E-06 5.65E-07 2.69E-04 3.23E-06 No Data 1.12E-06 I-134 8.05E-08 2.16E-07 7.69E-08 3.73E-06 3.44E-07 No Data 1.26E-10

I-135 3.35E-07 8.73E-07 3.21E-07 5.60E-05 1.39E-06 No Data 6.56E-07 i

cs-134- 4.66E-05 1.06E-04 9.10E-05 No Data 3.59E-05 1.22E-05 1.30E-06 Cs-136 4.88E-06 1.83E-05 1.38E-05 No Data 1.07E-05 1.50E-06 1.46E-06 Cs-137 5.98E-05 7.76E-05 5.35E-05 No Data 2.78E-05 9.40E-06 1.05E-06 co-138 4.14E-08 7.76E-08 4.05E-08 No Data 6.00E-08 6.07E-09 l 2.33E-13 Ba-139 1.17E-10 8.32E-14 3.42E-12 No Data 7.78E-14 4.70E-07 1.12E-07 1 -

Ba-140 4.88E-06 6.13E-09 3.21E-07 No Data 2.09E-09 1.59E-04 2.73E-05 Ba-141 1.25E-11 9.41E-15 4.20E-13 No Data

- 8.75E-15 2.42E-07 1.45E-17 Ba-142 3.29E-12 3.38E-15 2.07E-13 No Data 2.86E-15 1.49E-07 1.96E-26 La-140 4.30E-08 2.17E-08 5.73E-09 No Data No Data 1.70E-05 5.73E-05 2

La-142 8.54E-11 3.88E-11 9.65E-12 No Data No Data 7.91E-07 2.64E-07 !

Co-141 2.49E-06 1.69E-06 1.91E-07 No Data 7.83E-07 4.52E-05 1.50E-05 Co-143 2.33E-08 1.72E-08 1.91E-09 No Data 7.60E-09 9.97E-06 2.83E-05 Co-144 4.29E-04 1.793-04 2.30E-05 No Data 1.06E-04 9.72E-04 1.02E-04 Pr-143 1.173-06 4.693-07 5.80E-08 No Data 2.70E-07 3.51E-05 4 2.50E-05 Pr-144 3.76E-12 1.56E-12 1.91E-13 No Data 8.81E-13 1.27E-07 2.69E-18 Nd-147 6.59E-07 7.623-07 4.56E-08 No Data 4.45E-07 2.76E-05 2.16E-05 W-187 1.06E-09 8.85E-10 3.10E-10 No Data No Data 3.63E-06 1.94E-05 Np-239 2.87E-OS 2.82E-09 1.55E-09 No Data 8.75E-09 4.70E-06 1.49E-05 0 -

9-31 Gen. Rev. 13 1

FNP-O-M-011 Tcblo 9-11. Ingaction Doeo Factors for the Infant Age Group p Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 3.08E-07 3.08E-07 3.08E-07 3.08E-07 3.08E-07 3.08E-07 C-14 2.37E-05 5.06E-06 5.06E-06 5.06E-06 5.06E-06 5.06E-06 5.06E-06 Na-24 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 1.01E-05 P-32 1.70E-03 1.00E-04 6.59E-05 No Data No Data No Data 2.30E-05 Cr-51 No Data No Data 1.41E-08 9.20E-09 2.01E-09 1.79E-08 4.11E-07 Mn-54 No Data 1.99E-05 4.51E-06 No Data 4.41E-06 No Data 7.31E-06 Mn-56 No Data 8.18E-07 1.41E-07 No Data 7.03E-07 No Data 7.43E-05 Fe-55 1.39E-05 8.98E-06 2.40E-06 No D,ata No Data 4.39E-06 1.14E-06 Fe-59 3.08E-05 5.38E-05 2.12E-05 No Data No Data 1.59E-05 2.57E-05 Co-58 No Data 3.60E-06 8.98E-06 No Data No Data No Data 8.97E-06 Co-60 No Data 1.08E-05 2.55E-05 No Data No Data No Data 2.57E-05 Ni-63 6.34E-04 3.92E-05 2.20E-05 No Data No Data No Data 1.95E-06 NL-65 4.70E-06 5.32E-07 2.42E-07 No Data no Data No Data 4.05E-05 Cu=64 No Data 6.09E-07 2.82E-07 No Data 1.03E-06 No Data 1.25E-05 Zn-65 1.84E-05 6.31E-05 2.91E-05 No Data 3.06E-05 No Data 5.33E-05 Zn-69 9.33E-08 1.68E-07 1.25E-08 No Data 6.98E-08 No Data 1.37E-05 Br-83 No Data No Data 3.63E-07 No Data No Data No Data No Data Br-84 No Data No Data 3.82E-07 No Data No Data No Data No Data Br-85 No Data No Data 1.94E-08 No Data No Data No Data No Data Rb-86 No Data 1.70E-04 8.40E-05 No Data No Data No Data 4.35E-06 Rb-88 No Data 4.98E-07 2.73E-07 No Data No Data No Data 4.85E-07 Rb-89 No Data 2.86E-07 1.97E-07 No Data No Data No Data 9.74E-08 Sr-89 2.51E-03 No Data 7.20E-05 No Data No Data No Data 5.16E-05 sr-90 1.85E-02 No Data 4.71E-03 No Data No Data No Data 2.31E-04 Sr-91 5.005-05 No Data 1.81E-06 No Data No Data No Data 5.92E-05 All values are in (mrom/pc1 ingested). They are obtained from Reference 3 (Table E-14). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-125.

9-32 Gen. Rev. 13

i I

i FNP-0-M-011 Table 9-11 (cont'd). Ingestion Dose Factors for the Infant Age Group

(}

v Nuclide Bone Liver T. Body Thyroid Kidney Lung CI-LLI Sr-92 1.92E-05 No Data 7.13E-07 No Data No Data No Data 2.07E-04 Y-90 8.69E-08 No Data 2.33E-09 No Data No Data No Data 1.20E-04 Y-91m 8.103 *0 No Data 2.76E-11 No Data No Data No Data 2.70E-06 i Y-91 1.13E-06 No Data 3.01E-08 No Data No Data No Data 8.10E-05 '

Y-92 7.65E-09 No Data 2.15E-10 No Data No Data No Data 1.46E-04 Y-93 2.43E-08 No Data 6.62E-10 No Data No Data No Data 1.92E-04 !

Zr-95 2.06E-07 5.02E-08 3.56E-08 I No Data 5.41E-08 No Data 2.50E-05 Zr-97 1,48E-08 2.54E-09 1.16E-09 No Data 2.56E-09 No Data 1.62E-04 Nb-95 4.20E-08 1.73E-08 1.00E-08 No Data 1.24E-08 No Data 1.46E-05 Mo-99 No Data 3.40E-05 6.63E-06 No Data 5.08E-05 No Data 1.12E-05 Tc-99m 1.92E-09 3.96E-09 5.10E-08 No cita 4.26E-08 2.07E-09 1.15E-06 Tc-101 2.27E-09 2.86E-09 2.83E-08 No Data 3.40E-08 1.56E-09 4.86E-07 Ru-103 1.4BE-06 No Data 4.95E-07 No Data 3.08E-06 No Data 1.80E-05 Ru-105 1.36E-07 No Data 4.58E-08 No Data 1.00E-06 No Data 5.41E-05 V Ru-106 2.41E-05 No Data 3.01E-06 No Data 2.85E-05 No Data 1.83E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 9.96E-07 7.27E-07 4.81E-07 No Data 1.04E-06 No Data 3.77E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 2.33E-05 7.79E-06 3.15E-06 7.84E-06 No Data No Data 1.11E-05 Te-127m 5.85E-05 1.94E-05 7.08E-06 1.69E-05 1.44E-04 No Data 2.36E-05 Te-127 1.00E-06 3.35E-07 2.15E-07 8.14E-07 2.44E-06 No Data 2.10E-05 Te-129m 1.005-04 3.43E-05 1.54E-05 3.84E-05 2.50E-04 No Data 5.97E-05 Te-129 2.84E-07 9.79E-08 6.63E-08 2.38E-07 7.07E-07 :

No Data 2.27E-05 l Te-131m 1.52E-05 6.12E-06 5.05E-06 1.24E-05 4.21E-05 t No Data 1.03E-04 Ta-131 1.76E-07 6.50E-08 4.94E-08 1.57E-07 4.50E-07 No Data 7.11E-06 ,

l l

l O

9-33 Gen. Rev. 13 l

i FNP-0-M-011 Table 9-11 (contd). Ingestion Dose Factors for the Infant Age Group i

{

)

J Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI To-132 2.08E-05 1.03E-05 9.61E-06 1.52E-05 6.44E-05 No Data 3.81E-05 I-130 6.00E-06 1.32E-05 5.30E-06 1.48E-03 1.45E-05 No Data 2.83E-06 I-131 3.59E-05 4.23E-05 1.86E-05 1.39E-02 4.94E-05 No Data 1.51E-06 I-132 1.66E-06 3.37E-06 1.2JE-06 1.58E-04 3.76E-06 No Data 2.73E-06 I-133 1.25E-05 1.82E-05 5.33E-06 3.31E-03 2.14E-05 No Data 3.08E-06 I-134 8.69E-07 1.78E-06 6.33E-07 4.15E-05 1.99E-06 No Data 1.84E-06 I-135 3.64E-06 7.24E-06 2.64E-06 6.49E-04 8.07E-06 No Data 2.62E-06 Cs-134 3.77E-04 7.03E-04 7.10E-05 No Data 1.81E-04 7.42E-05 1.91E-06 Cs-136 4.59E-05 1.35E-04 5.04E-05 No Data 5.38E-05 1.10E-05 2.05E-06 Cs-137 5.22E-04 6.11E-04 4.33E-05' No Data 1.64E-04 6.64E-05 1.91E-06 Cs-138 4.81E-07 7.82E-07 3.79E-07 No Data 3.90E-07 6.09E-08 1.25E-06 Ba-139 8.81E-07 5.84E-10 2.55E-08 ra Data 3.51E-10 3.54E-10 5.58E-05 Ba-140 1.71E-04 1.71E-07 8.81E-06 No Data 4.06E-08 1.05E-07 4.20E-05 Ba-141 4.25E-07 2.91E-10 1.34E-08 No Data 1.75E-10 1.77E-10 5.19E-06 Ba-142 1.84E-07 1.53E-10 9.06E-09 No Data 8.81E-11 9.26E-11 7.59E-07 !

La-140 2.11E-08 8.32E-09 2.14E-09 No Data No Data No Data 9.77E-05 1

, La-142 1.10E-09 4.04E-10 9.67E-11 No Data No Data No Data 6.86E-05 Co-141 7.87E-08 4.80E-08 5.65E-09 No Data 1.48E-08 No Data 2.48E-05 Co-143 1.48E-08 9.82E-06 1.12E-09 No Data 2.86E-09 No Data 5.73E-05 Co-144 2.98E-06 1.22E-06 1.67E-07 No Data 4.93E-07 No Data 1.71E-04 Pr-143 8.135-08 3.04E-08 4.03E-09 No Data 1.13E-08 No Data 4.29E-05 Pr-144 2.745-10 1.06E-10 1.38E-11 No Data 3.84E-11 No Data 4.93E-06 Nd-147 5.53E-08 5.68E-08 3.48E-09 No Data 2.19E-08 No Data 3.60E-05 W-187 9.03E-07 6.28E-07 2.17E-07 No Data No Data No Data 3.69E-05 Np-239 1.11E-08 9.93E-10 5.61E-10 No Data 1.98E-09 2.87E-05 No Data l C

9-34 Gen. Rev. 13

FNP-0-M-011 n Table 9-12. Ingestion Dose Factors for the Child Age Group U Nuclide Bone Liver T. Body

~

Thyroid Kidney Lung GI-LLI umummes j

H-3 No Data 2.03E-07 2.03E-07 2.03E-07 2.03E-07 2.03F 07 2.03E-07 C-14 1.21E-05 2.42E-06 2.42E-06 2.42E-06 2.42E-06 2.42E-06 2.42E-06 Na-24 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 5.80E-06 P-32 8.25E-04 3.86E-05 3.18E-05 No Data No Data No Data 2.28E-05 Cr-51 No Data No Data 8.90E-09 4.94E-09 1.35E-09 9.02E-09 4.72E-07 Kn-54 No Data 1.07E-05 2.85E-06 No Data 3.00E-06 No Data 8.98E-06 Mn-56 No Data 3.34E-07 7.54E-08 No Data 4.04E-07 No Data 4.84E-05 Fe-55 1.15E-05 6.10E-06 1.89E-06 No Data No Data 3.45E-06 1.13E-06 Fe-59 1.65E-05 2.67E-05 1.33E-05 No Data No Data 7.74E-06 2.78E-05 Co-58 No Data 1.80E-06 5.51E-06 No Data No Data No Data 1.05E-05 Co-60 No Data 5.29E-06 1.56E-05 No Data No Data No Data 2.93E-05 Ni-63 5.38E-04 2.88E-05 1.83E-05 No Data No Data No Data 1.94E-06 Ni-65 2.22E-06 2.09E-07 1.22E-07 No Dara No Data No Data 2.56E-05 Cu-64 No Data 2.45E-07 1.48E-07 a No Data 5.92E-07 No Data 1.15E-05 2n-65 1.37E-05 3.65E-05 2.27E-05 No Data 2.30E-05 No Data 6.41E-06 Zn-69 4.38E-08 6.33E-08 5.85E-09 No Data 3.84E-08 No Data 3.99E-06 Br-83 No Data No Data 1.71E-07 No Data No Data No Data No Data Br-84 No Data No Data 1.98E-07 No Data No Data No Data No Data Br-85 No Data No Data 9.12E-09 No Data No Data No Data No Data Rb-86 No Data 6.70E-05 4.12E-05 Ne Data No Data No Data 4.31E-06 Rb-88 No Data 1.90E-07 1.32E-07 flo Data No Data No Data 9.32E-09 Rb-89 No Data 1.17E-07 1.04E-07 ho Data No Data No Data 1.02E-09 Sr-89 1.32E-03 No Data 3.77E-05 No Data No Data No Data 5.11E-05 Sr-90 1.70E-02 No Data 4.31E-03 No Data No Data No Data 2.29E-04 Sr-91 2.40E-05 No Data 9.06E-07 No Data No Data No Data 5.30E-05 All values are in (mrom/pci ingested) . They are obtained from Reference 3 (Table E-13). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-12!.

O 9-35 Gen. Rev. 13

FNP-0-M-011 Table 9-12 (contd). Ingestion Dose Factors for the Child Age Group O Nuclida Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 1.03E-06 No Data 3.62E-07 No Data No Data No Data 1.71E-04 Y-90 4.11E-08 No Data 1.10E-09 No Data No Data No Data 1.17E-04 Y-91m 3.82E-10 No Data 1.39E-11 No Data No Data No Data 7.40E-07 Y-91 6.02E-07 No Data 1.61E-08 No Data No Data No Data 8.02E-05 Y-92 3.60s-09 No Data 1.03E-10 No Data No Data No Data 1.04E-04 Y-93 1.14E-08 No Data 3.13E-10 No Data No Data No Data 1.70E-04 Zr-95 1.16E-07 2.55E-08 2.27E-08 No Data 3.65E-08 No Data 2.66E-05 Zr-97 6.99E-09 1.01E-09 5.96E-10 No Data 1.45E-09 No Data 1.53E-04 Nb-95 2.25E-08 8.76E-09 6.26E-09 No Data 8.23E-09 No Data 1.62E-05 Mo-99 No Data 1.33E-05 3.29E-06 No Data 2.84E-05 No Data 1.10E-05 Tc-99m 9.23E-10 1.81E-09 3.00E-08 No Data 2.63E-08 9.19E-10 1.03E-06 Tc-101 1.07E-09 1.12E-09 1.42E-08 No Data 1.91E-08 5.92E-10 3.56E-09 Ru-103 7.31E-07 No Data 2.81E-07 No Data 1.84E-06 No Data 1.89E-05 Ru-105 6.45E-08 No Data 2.34E-08 No Data 5.67E-07

\J No Data 4.21E-05 Ru-106 1.17E-05 No Data 1.46E-06 No Data 1.58E-05 No Data 1.82E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 5.39E-07 3.64E-07 2.91E-07 No Data 6.78E-07 No Data 4.33E-05 Sb-124 No Data No Data No Data No Data No Data No Data ~ No Data sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 1.14E-05 3.09E-06 1.52E-06 3.20E-06 No Data No Data 1.10E-05 Te-127m 2.89E-05 7.785-06 3.43E-06 6.91E-06 8.24E-05 No Data 2.34E-05 Te-127 4.71E-07 1.27E-07 1.01E-07 3.26E-07 1.34E-06 No Data 1.84E-05 Te-129m 4.87E-05 1.36E-05 7.56E-06 1.57E-05 1.43E-04 No Data 5.94E-05 Te-129 1.34E-07 3.743-08 3.18E-08 9.56E-08 3.92E-07 No Data 8.34E-06 Te-131m 7.20E-06 2.49E-06 2.65E-06 5.12E-06 2.41E-05 No Data 1.01E-04 Te-131 8.30E-08 2.53E-08 2.47E-08 6.35E-08 2.51E-07 No Data 4.36E-07 O

9-36 Gen. Rev. 13

1 I

FNP-0-M-011 Table 9-12 (contd). Ingestion Dose Factors for the Child Age Group i

i Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 1.01E-05 4.47E-06 5.40E-06 6.51E-06 4.15E-05 No Data 4.50E-05 I-130 2.92E-06 5.90E-06 3.04E-06 6.50E-04 8.82E-06 No Data 2.76E-06 I-131 1.72E-05 1.73E-05 9.83E-06 5.72E-03 2.84E-05 No Data 1.54E-06 I-132 8.00E-07 1.47E-06 6.76E-07 6.82E-05 2.25E-06 No Data 1.73E-06 I-133 5.92E-06 7.32E-06 2.77E-06 1.36E-03 1.22E-05 No Data 2.95E-06 I-134 4.19E-07 7.78E-07 3.58E-07 1.79E-05 1.19E-06 No Data 5.16E-07 I-135 1.75E-06 3.15E-06 1.49E-06 2.79E-04 4.83E-06 No Data 4

2.40E-06 Cs-134 2.34E-04 3.84E-04 8.10E-05 No Data 1.19E-04 4.27E-05 2.07E-06 Cs-136 2.35E-05 6.46E-05 4.18E-05 No Data 3.44E-05 5.13E-06 2.27E-06 Cs-137 3.27E-04 3.13E-04 4.62E-05 No Data 1.02E-04 3.67E-05 1.96E-06 Cs-138 2.28E-07 3.17E-07 2.01E-07 No Data 2.23E-07 2.40E-08 1.46E-07 Ba-139 4.14E-07 2.21E-10 1.20E-08 No Data 1.93E-10 1.30E-10 2.39E-05 Ba-140 8.31E-05 7.28E-08 4.85E-06 No Data 2.37E-08 4.34E-08 4.21E-05 l

Ba-141 2.00E-07 1.12E-10 6.51E-09 No Data 9.69E-11 6.58E-10 1.14E-07 Ba-142 8.74E-08 6.29E-11 4.88E-09 No Data i 5.09E-11 3.70E-11 1.14E-09 l La-140 1.01E-08 3.53E-09 1.19E-09 No Data No Data No Data 9.84E-05 ;

La-142 5.24E-10 1.67E-10 5.23E-11 No Data No Data

~

No Data 3.31E-05 co-141 3.97E-08 1.98E-08 2.94E-09 4

No Data 8.68E-09 No Data 2.47E-05 Co-143 6.99E-09 3.79E-06 5.49E-10 No Data 1.59E-09 No Data 5.55E-05 Co-144 2.08E-06 6.523-07 1.11E-07 No Data 3.61E-07 No Data 1.70E-04 Pr-143 3.93E-08 1.18E-08 1.95E-09 No Data 6.39E-09 No Data 4.24E-05 Pr-144 1.295-10 3.993-11 6.49E-12 No Data 2.11E-11 No Data 8.59E-08 Nd-147 2.79E-08 2.26E-08 1.75E-09 No Data 1.24E-08 No Data 3.58E-05 L

W-187 4.29E-07 2.54E-07 1.14E-07 No Data No Data No Data 3.57E-05 Np-239 5.25E-09 3.77E-10 2.65E-10 No Data 1.09E-09 No Data 2.79E-05 O

9-37 Gen. Rev. 13

i l

FNP-0-M-011 Table 9-13. Ingestion Dose Factors for the Teenager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.06E-07 1.06E-07 1.06E-07 1.06E-07 1.06E-07 2.06E-07 )

C-14 4.06E-06 8.12E-07 8.12E-07 8.12E-07 8.12E-07 8.12E-07 ' 8.1: ,,F-07 l Na-24 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-05 2.30E-05 a

2.302-06 P-32 2.76E-04 1.71E-05 1.07E-05 No Data No Data No Data 2.32E-05 Cr-51 No Data No Data 3.60E-09 2.00E-09 7.89'-10 5.14E-09 6.05E-07 Mn-54 No Data 5.90E-06 1.17E-06 No Data 1.76E-06 t

No Data 1.21E-05 Kn-56 No Data 1.58E-07 2.81E-08 No Data 2.00E-07 No Data 1.04E-05 Fe-55 3.78E-06 2.68E-06 6.25E-07 No Data No Data 1.70E-06 1.16E-06 Fe-59 5.87E-06 1.37E-05 5.29E-06 No Data No Data 4.32E-06 3.24E-05 Co-58 No Data 9.72E-07 2.24E-J6 No Data No Data No Data 1.34E-05 Co-60 No Data 2.81E-06 6.33E-06 No Data No Data No Data 3.66E-05 Ni-63 1.77E-04 1.2L/-05 6.00E-06 No Data No Data No Data 1

1.99E-06 1 NL-65 7.49E-07 9.57E-08 4.36E-OB No Data No Data No Data 5.19E-06 Cu-64 No Data 1.15E-07 5.41E-08 No Data 2.91E-07 No Data 8.92E-06 O Zn-65 5.76E-06 2.00E-05 9.33E-06 No Data 1.28E-05 No Data 8.47E-06 Zn-69 1.47E-08 2.80E-08 1.96E-09 No Data 1.83E-08 No Data 5.16E-08 Br-83 No Data No Data 5.74E-08 No Data No Data No Data No Data Br-84 No Data No Data 7.22E-08 No Data No Data No Data No Data

) Br-85 No Data No Data 3.05E-09 No Data No Data No Data No Data Rb-86 No Data 2.98E-05 1.40E-05 No Data No Data No Data 4.41E-06 Rb-88 No Data 8.52E-08 4.54E-08 No Data No Data No Data 7.30E-15 Rb-89 No Data 5.503-08 3.89E-08 No Data No Data 1

No Data 8.43E-17 Sr-89 4.40E-04 No Data 1.26E-05 No Data No Data No Data 5.24E-05 Sr-90 S.30E-03 No Data 2.0$E-03 No Data No Data No Data 2.33E-04 :

Sr-91 8.07E-06 No Data 3.21E-07 No Data No Data No Data 3.66E-05 l

All values are in (mrem /pci ingested) . They are obtained from Reference 3 (Table E-12). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-125.

l

(

V 9-38 Gen. Rev. 13

FNP-0-M-011 Table 9-13 (contd). Ingestion Dose Factors for the Teenager Age Group i O l

\

I Nuclide Bone Liver T.Sody Thyroid Kidney Lung GI-LLI Sr-92 3.05E-06 No Data 1.30E-07 No Data No Data No Data

! 7.77E-05 Y-90 1.37E-08 No Data 3.69E-10 No Data No Data No Data 1.13E-04 4

Y-91m 1.29E-10 No Data 4.93E-12 No Data No Data

No Data 6.09E-09 Y-91 2.01E-07 No Data 5.39E-09 No Data No Data No Data 8.24E-05 Y-92 1.21E-09 No Data 3.50E-11 No Data No Data No Data 3.32E-05.

Y-93 3.83E-09 No Data 1.05E-10 No Data No Data No Data 1.17E-04 Zr-95 4.12E-08 1.30E-08 8.94E-09 No Data 1.91E-08 No Data 3.00E-05 Zr-97 2.37E-09 4.69E-10 2.16E-10 No Data 7.11E-10 No Data 1.27E-04 Nb-95 8.22E-09 4.56E-09 2.51E-09 No Data 4.42E-09 No Data 1.95E-05 l Mo-99 No Data 6.03E-06 1.15E-06 No Data 1.38E-05 No Data 1.08E-05 Tc-99m 3.32E-10 9.26E-10 1.20E-08 i No Data 1.38E-08 5.14E-10 6.08E-07 I Tc-101 3.60E-10 5.12E-10 5.03E-09 No Data 9.26E-09 3.12E-10 8.75E-17 i

Ru-103 2.55E-07 No Data 1.09E-07 No Data 8.99E-07 No Data 2.13E-05 l Ru-105 2.18E-08 No Data 8.46E-09 No Data l

2.75E-07 No Data 1.76E-05 Ru-106 3.92E-06 No Data 4.94E-07 No Data 7.56E-06 No Data 1.88E-04 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 2.05E-07 1.94E-07 1.18E-07 i No Data 3.70E-07 No Data 5.45E-05 Sb-124 No Data No Data No Data No Data No Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data Te-125m 3.83E-06 1.38E-06 5.12E-07 1.07E-06 No Data No Data 1.13E-05 Te-127m 9.67E-06 3.43E-06 1.15E-06 2.30E-06 3.92E-05 No Data 2.41E-05 Te-127 1.58E-07 5.60E-08 3.40E-08 1.09E-07 6.40E-07 No Data

.. 1.22E-05 j

Te-129m 1.43E-05 6.05E-06 2.58E-06 5.26E-06 6.82E-05 No Data 6.12E-05 {

Te-129 4.48E-08 1.67E-08 1.09E-08 3.20E-08 1.88E-07 No Data 2.45E-07 1 Te-131m 2.44E-06 1.17E-06 9.76E-07 1.76E-06 1.22E-05 No Data 1 9.39E-05 Te-131 2.79E-08 1.15E-08 8.72E-09 2.15E-08 1.22E-07 No Data 2.29E-09 O

9-39 Gen. Rev. 13

FNP-0-M-011 Table 9-13 (contd). Ingestien Dose Factors for the Teenager Age Group

(

Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 3.49E-06 2.21E-06 2.08E-06 2.33E-06 2.12E-05 No Data 7.00E-05 I-130 1.03E-06 2.98E-06 1.19E-06 2.43E-04 4.59E-06 No Data 2.29E-06 I-131 5.85E-06 8.19E-06 4.40F-n6 2.39E-03 1.41E-05 No Data 1.62E-06 I-132 2.79E-07 7.30E-07 2.62E-07 2.46E-05 1.15E-06 No Data 3.18E-07 I-133 2.01E-06 3.41E-06 1.04E-06 4.76E-04 5.98E-06 No Data 2.58E-06 I-134 1.46E-07 3.87E-07 1.39E-07 6.45E-06 6.10E-07 No Data 5.10E-09 I-135 6.10E-07 1.57E-06 5.82E-07 1.01E-04 2.48E-06 No Data 1.74E-06 Cs-134 8.37E-05 1.97E-04 9.14E-05 No Data 6.26E-05 2.39E-05 2.45E-06 Cs-136 8.59E-06 3.38E-05 2.27E-05 No Data 1.04E-05 2.90E-06 2.72E-06 Cs-137 1.12E-04 1.49E-04 5.19E-05 No Data 5.0?E-05 1.97E-05 2.125-06 Cs-138 7.76E-08 1.49E-07 7.45E-08 No Data 1.10E-07 1.28E-08 6.76E-11 Ba-139 1.39E-07 9.78E-11 4.05E-09 No Data 9.22E-11 6.74E-11 1.24E-06 Ba-140 2.84E-05 3.48E-08 1.83E-06 No Data 1.18E-08 2.34E-08 4.38E-05 Ba-141 6.71E-08 5.01E-11 2.24E-09 No Data 4.65E-11 3.43E-11 1.43E-13 Ba-142 2.99E-08 2.99E-11 1.84E-09 No Data 2.53E-11 1.99E-11 9.18E-20 La-140 3.48E-09 1.71E-09 4.55E-10 No Data No Data No Data 9.82E-05 La-142 1.79E-10 7.95E-11 1.98E-11 No Data No Data No Data 2.42E-06 Co-141 1.33E-08 8.88E-09 1.02E-09 No Data 4.18E-09 No Data 2.54E-05 Co-143 2.35E-09 1.713-06 1.91E-10 No Data 7.67E-10 No Data 5.14E-05 l Co-144 6.96E-07 2.88E-07 3.74E-08 No Data 1.72E-07 No Data 1.75E-04 Pr-143 1.31E-08 5.233-09 6.52E-10 No Data 3.04E-09 No Data 4.31E-05 Pr-144 4.30E-11 1.763-11 2.18E-12 No Data 1.01E-11 No Data 4.74E-14 Nd-147 9.38E-09 1.02E-08 6.11E-10 No Data i 5.99E-09 No Data 3.68E-05 W-187 1.46E-07 1.19E-07 4.17E-08 No Data No Data No Data 3.22E-05 Np-239 1.76E-09 1.66E-10 9.22E-11 No Data 5.21E-10 No Data 2.67E-05 l

l 0- 1 9-40 Gen. Rev. 13

FNP-0-M-011 Table 9-14. Ingestion Dose Factors for the Adult Age Group e mmmmmmmmmmmmme Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 C-14 2.84E-06 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 5.68E-07 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 P-32 1.93E-04 1.20E-05 7.46E-06 No Data No Data No Da'4 2.17E-05 Cr-51 No Data No Data 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 Mn-54 No Data 4.57E-06 8.72E-07 No Data 1.36E-06 No Data 1.40E-05 Mn-56 No Data 1.15E-07 2.04E-08 No Data 1.46E-07 No Data 3.67E-06 Fe-55 2.75E-06 1.90E-06 4.43E-07 No Data No Data 1.06E-06 1.09E-06 Fe-59 4.34E-06 1.02E-05 3.91L-06 No Data No Data 2.85E-06 3.40E-05 Co-58 No Data 7.45E-07 1.67E-06 No Data No Data No Data 1.51E-05 Co-60 No Data 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-05 N L-63 1.30E-04 9.01E-06 4.36E-06 No Data No Data No Data 1.88E-06 Ni-65 5.28E-07 6.86E-08 3.13E-08 No Data No Data No Data 1.74E-06 Cu-64 No Data 8.33E-08 3.91E-08 No Data 2.10E-07 No Data 7.10E-06 Zn-65 4.84E-06 1.54E-05 6.96E-06 No Data 1.03E-05 No Data 9.70E-06 Zn-69 1.03E-08 1.97E-08 1.37E-09 No Data 1.28E-08 No Data 2.96E-09 Br-83 No Data No Data 4.02E-08 No Data No Data No Data 5.79E-08 Be-84 No Data No Data 5.21E-08 No Data No Data No Data 4.09E-13 Br-85 No Data No Data 2.14E-09 No Data No Data No Data No Data Rb-86 No Data 2.11E-05 9.83E-06 No Data No Data No Data 4.16E-06 Rb-88 No Data 6.05E-08 3.212-08 No Data No Data No Data 8.36E-19 Rb-89 No Data 4.013-08 2.82"-08 No Data No Data No Data 2.33E-21 Sr-89 3.08E-04 No Data 8.84E-06 No Data No Data No Data 4.94E-05 Sr-90 7.585-03 No Data 1.86E-03 No Data No Data No Data 2.19E-04 Sr-91 5.675-06 No Data 2.29E-07 No Data No Data No Data 2.70E-05 All values are in (mrom/pci ingested) . They are obtained from Reference 3 (Table E-11), except as follows: Reference 2 (Table A-3) for Rh-105, Sb-124, and Sb-125.

O 9-41 Gen. Rev. 13 v - - - _ - - - - - . . _ . _ _ . - - _ _ . . _ . - - - - - _ . _ _ - _ _ _

FNP-O-M-011 Table 9-14 (contd). Ingestion Dose Factors for the Adult Age Group 4O Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 2.15E-06 No Data. 9.30E-08 No Data No Data No Data 4.26E-05 Y-90 9.62E-09 No Data 2.58E-10 No Data No Data No Data 1.02E-04 Y-91m 9.09E-11 No Data 3.52E-12 No Data No Data No Data 2.67E-10 Y-91 1.41E-07 No Data 3.77E-09 No Data No Data

! No Data 7.76E-05 Y-92 8.45E-10 No Data 2.47E-11 No Data No Data 4 No Data 1.48E-05 Y-93 2.6BE-09 No Data 7.40E-11 No Data No Data No Data 8.50E-05

~

Zr-95 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 Zr-97 1.68E-09 3.39E-10 1.55E-10 No Data 5.12E-10 No Data 1.05E-04 j Nb-95 6.22E-09 3.46E-09 1.86E-09 No Data 3.42E-09 No Data 2.10E-05 i Mo-99 No Data 4.31E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06

Tc-99m 2.47E-10 6.98E-10 8.89E-09 No Data 1.06E-08 3.42E-10 4.13E-07 Tc-101 2.54E-10 3.66E-10 3.59E-09 No Data 6.59E-09 1.87E-10 1.10E-21 Ru-103 1.85E-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 j Ru-105 1.54E-08 No Data 6.08E-09 No Data 1.99E-07 No Data 9.42E-06 Ru-106 2.75E-06 No Data 3.48E-07 No Data 5.31E-06 i No Data 1.78E-04 Rh-105 1.22E-07 8.86E-08 5.83E-08 No Data 3.76E-07 No Data 1.41E-05 Ag-110m 1.60E-07 1.48E-07 8.79E-08 No Data 2.91E-07 No Data 6.04E-05

Sb-124 2.81E-06 5.30E-08 1.11E 06 6.79E-09 No Data 2.18E-06 7.95E-05 Sb-125 2.23E-06 2.40E-04 4.48E-07 1.98E-09 No Data i 2.33E-04 1.97E-05 Te-125m 2.68E-06 9.71E-07 3.5sh-d7 8.06E-07 1.09E-05 No Data 1.07E-05 3

Te-127m 6.77E-06 2.423-06 8.25E-07 1.73E-06 2.75E-05 No Data 2.27E-05

Te-127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4.48E-07 No Data

8.68E-06 Te-129m 1.15E-05 4.293-06 1.82E-06 3.95E-06 4.80E-05 No Data 5.79E-05 To-129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 No Data 2.37E-08 Te-131m 1.73E-06 8.46E-07 7.05E-07 1.34E-06 8.57E-06 No Data 8.40E-05 Te-131 1.97E-08 8.23E-09 6.22E-09 1.62E-08 8.63E-08 No Data 2.79E-09 i

O 9-42 Gen. Rev. 13

j_ FNP-0-M-011 Tabis 9-14 (contd). Ingestion Done Factors for the Adult Age Group 4

i

==

/ Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.52E-06 1.63E-06 1.53E-06 1.80E-06 1.57E-05 No Data 7.71E-05 I-130 7.56E-07 2.23E-06 8.80E-07 1.89E-04 3.48E-06 No Data 1.92E-06 I-131 4.16E-06 5.95E-06 3.41E-06 1.95E-03 1.02E-05 No Data 1.57E-06 I-132 2.03E-07 5.43E-07 1.90E-07 1.90E-05 8.65E-07 No Data 1.02E-07 I-133 1.42E-06 2.47E-06 7.53E-07 3.63E-04 4.31E-06 No Data 2.22E-06 I-134 1.06E-07 2.88E-07 1.03E-07 4.99E-06 4.58E-07 No Data 2.51E-10 '

I-135 4.43E-07 1.16E-06 4.28E-07 7.65E-05 1.86E-06 No Data 1.31E-06 Co-134 6.22E-05 1.48E-04 1.21E-04 No Data 4.79E-05 1.59E-05 2.59E-06 Cs-136 6.51E-06 2.57E-05 1.85E-05 No Data 1.43E-05 1.96E-06 2.92E-06 Cs-137 7.97E-05 1.09E-04 7.14E-05 No Data 3.70E-05 1.23E-05 2.11E-06 Co-138 5.52E-08 1.09E-07 5.40E-08 No Data 8.01E-08 7.91E-09 4.65E-13 Ba-139 9.70E-08 6.91E-11 2.84E-09 No Data 6.46E-11 3.92E-11 1.72E-07 Ba-140 2.03E-05 2.55E-08 1.33E-06 No Data 8.67E-09 1.46E-08 4.18E-05 Ba-141 4.71E-08 3.56E-11 1.59E-09 No Data 3.31E-11 2.02E-11 2.22E-17 Ba-142 2.13E-08 2.19E-11 1.34E-09 No Data 1.85E-11 1.24E-11 3.00E-26 La-140 2.50E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.25E-05 La-142 1.28E-10 5.82E-11 1.45E-11 No Data No Data No Data 4.25E-07 !

co-141 9.36E-09 6.33E-09 7.18E-10 No Data 2.94E-09 I No Data 2.42E-05 Co-143 1.65E-09 1.22E-06 1.35E-10 No Data 5.37E-10 No Data 4.56E-05 Ce-144 4.88E-07 2.04E-07 2.62E-08 No Data 1.21E-07 No Data 1.65E-04 Pr-143 9.20E-09 3.69E-09 4.56E-10 No Data 2.13E-09 No Data 4.03E-05 Pr-144 3.013-11 1.253-11 1.53E-12 No Data 7.05E-12 No Data 4.33E-18 Nd-147 6.29E-09 7.27E-09 4.35E-10 No Data 4.25E-09 No Data 3.49E-05 W-187 1.035-07 8.61E-08 3.01E-08 No Data No Data No Data 2.82E-05 Np-239 1.19E-09 1.17E-10 6.45E-11 No Data 3.65E-10 No Data 2.40E-05 l l

O 9-43 Gen. Rev. 13 l

l

FNP-0-M-011 l

Table 9-15. External Dose Factors for Standing en Contaminated Ground O Nuclide T. Body Skin Nuclide T.{cg Skin H-3 0.00 0.00 Sr-91 7.10E-09 8.30E-09 C-14 0.00 0.00 Sr-92 9.00E-09 1.00E-08 Na-24 2.50E-08 2.90E-08 Y-90 2.20E-12 2.60E-12 P-32 0.00 0.00 Y-91m 3.80E-09 4.40E-09 Cr-51 2.20E-10 2.60E-10 Y-91 2.40E-11 2.70E-11 Mn-34 5.80E-09 6.80E-09 Y-92 1.60E-09 1.90E-09 Kn-56 1.10E-08 1.30E-08 Y-93 5.70E-10 7.80E-10 Fe-55 0.00 0.00 Zr-95 5.00E-09 5.80E-09 Fe-59 8.00E-09 9.40E-09 Zr-97 5.50E-09 6.40E-09 Co-58 7.00E-09 8.20E-09 Nb-95 5.10E-09 6.00E-09 Co-60 1.70E-08 2.00E-09 Mo-99 1.90E-09 2.20E-09 Ni-63 0.00 0.00 Tc-99m 9.60E-10 1.10E-09 NL-65 3.70E-09 4.30E-09 Tc-101 2.70E-09 3.00E-09 Cu-64 1.50E-09 1.702-09 Ru-103 3.60E-09 4.20E-09 Zn-65 4.00E-09 4.60E-09 Ru-105 4.50E-09 5.10E-09 Zn-69 0.00 0.00 Ru-106 1.50E-09 1.80E-09 Br-83 6.40E-11 9.30E-11 Rh-105 6.60E-10 7.70E-10 Br-84 1.20E-08 1.40E-08 Ag-110m 1.80E-08 2.10E-08 Br-85 0.00 0.00 Sb-124 1.30E-08 1.50E-08 Rb-86 6.30E-10 7.20E-10 Sb-125 3.10E-09 3.50E-09 Rb-88 3.50E-09 4.00E-09 Te-125m 3.50E-11 4.80E-11 Rb-89 1.50E-08 1.80E-08 Te-127m 1.10E-12 1.30E-12 Sr-89 5.60E-13 6.50E-13 Te-127 1.00E-11 1.10E-11 3r-90 0.00 0.00 Te-129m 7.70E-10 9.00E-10 A

All valties ,are in- (arem/h) per (pci/m2 ). They are obtained from Reference-4 (Table R-6),

for Rh-10$l 54-124, except ao followes ' Reference 2 (Table A-7) and Sb-125.

(

k.

9-44 Gen. Rev. 13 i

l FNP-0-M-011 Table 9-15 (contd). External Dose Factors for Standing on Contaminated Ground Nuclide T. Body Skin Te-129 7.10E-10 8.40E-10 Te-131m 8.40E-09 9.90E-09 '

Te-131 2.20E-09 2.60E-06 Te-132 1.70E-09 2.00E-09 I-130 1.40E-08 1.70E-08 I-131 2.80E-09 3.40E-09 1

I I-132 1.70E-08 2.00E-08 I-133 3.70E-09 4.50E-09 I-134 1.60E-08 1.90E-08 I-135 1.20E-08 1.40E-08 Cs-134 1.20E-08 1.40E-08 Cs-136 1.50E-08 1.70E-08 Cs-137 4.20E-09 4.90E-09 Cs-138 2.10E-08 2.40E-08 Ba-139 )

2.40E-09 2.70E-09 Ba-140 2.10E-09 2.40E-09 Ba-141 4.30E-09 4.90E-09 Ba-142 7.90E-09 9.00E-09 La-140 1.50E-08 1.70E-08 i La-142 1.50E-08 1.80E-08 co-141 5.50E-10 6.20E-10 Co-143 2.20E-09 2.50E-09 Co-144 3.20E-10 3.70E-10 j Pr-143 0.00 0.00 Pr-144 2.005-10 2.30E-10 Nd-147 1.005 1.20E-09 W-187 3.10E-09 3.60E-09 Np-239 9.50E-10 1.10E-09 I

O U

9-45 Gen. Rev. 13 l - -.

]

FNP-0-M-011 i

j CHAPTER 10 1 DEFINITIONS OF EFFLUENT CONTROL TERMS I

[

The terms defined in this chapter are used in the presentation of the above i

{

chapters. These terms are shown in all capital letters to indicate that they are !

specifically defined.

j l j 10.1 TERMS SPECIFIC TO THE ODCM I i

i

'l ,

} The following terms are used in the ODCM, but are not found in the Technical ,

4 Specifications:

I BATCH RELEASE i'

i A BATCH RELEASE is the discharge of wastes of a discrete volume. Prior to

) sampling for analyses, each liquid batch shall be isolated and then f thoroughly mixed by a method described in the ODCM to assure

representative sampling.

j 1

COMPOSITE SAMPLE

} A COMPOSITE SAMPLE is one which contains material from multiple waste releases, in which the quantity of sample is proportional to the quantity of waste discharged, and.in which the method of sampling employed resulta j i

in a specimen that is representative of the wastes released.- Prior to O

D analyses, all liquid samples that are to be aliquotted for a COMPOSITE SAMPLE shall be mixed thoroughly, in order for the COMPOSITE SAMPLE to be representative of the affluent release.

When assessing the consequences of a waste release at the pre-release or l post-release stage, the most recent available COMPOSITE SAMPLE results for ~ l the applicable release pathway may be used.

i CONTINUOUS RELEASE A CONTINUOUS RELEASE is the discharge of wastes of a non-discrete volume, e.g.,

from,a plume within a system that has an input flow during the contiasmus respose. To be representative of~the quantities and concen-tratioedoli radiometive materials ' in CONTINUOUS RELEASES of liquid effluente, semp!"eir*shall be collected in proportion to the rate of flow of the efflueet' stream, or to the quantity of. waste discharged.

cAstous mAnwasTa TanATuenT system A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup O 10-1 Gen. Rev. 13

1 i

i.

! FNP-0-M-011 for the purpose of reducing the total radioactivity-prior to release to the env!.ronment. This system consists of at least one gas compressor,

's waste gas decay tanks, and associated components providing for. treatment flow and functional control.

.i i

LIQUID RADWASTE TREATMENT SYSTEM

.{ A LIQUID RADWASTE TREATMENT SYSTEM is any system designed and installed to !

j reduce radioactive materials in 4

liquid effluents by systematic collection, retention, and processing through filtration, evaporation, i separation and/or ion exchange treatment. This system consists of at i

least one collection tank, one evaporator or domineralizer system, one post-treatment tank and associated components providing for treatment flow .

3 and functional control. .

i 1

$ MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS For the purposes of the ODCM, MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS include the following changes to such systems:

5 j (1) Major changes in process equipment, components, structures, or effluent monitoring instrumentation as described in the Final Safety Analysis Report (FSAR) or as evaluated in the Nuclear Regulatory Coasnission staff's safety Evaluation Report (SER) (e.g.,

deletion of evaporators and installation of domineralizers);

I (2) 1 Changes in the design of radwaste - treatment systems that .could significantly increase quantities of effluents released from those previously considered in_the FSAR and SER; 1

i (3) Changes in system design which may invalidate the accident analysis -

as described in the SER (e.g., changes in tank capacity that-would

} alter the curies released); or' k

(4) Changes in system design that could potentially result in a significant increase in occupational exposure of operating i

1 personnel (e.g., use of temporary equipment without adequate j shielding provisions). ,

4 l

MEwarmiS) OF THE PUBLICI I A MEMBER OF THE PUBLIC shall be an individual in a controlled area or an UNRESTRICTED AREA. However, an individual is not a MEMBER OF THE PUBLIC !

1 e

i 1 The italicized terms in this definition, which are not otherwise used.in this ODCM, shall have the definitions assigned to them by 10 CFR 20.1003.

10-2 Gen. Rev. 13 l

~

n- - ,

2 i

l FNP-0-M-011 l

during any period in which the individual receives an occupational dose.

! s This category may include persons who use portions of the site for ,

recreational, occupational, or other purposes not associated with the plant.

MINIMUM DETECTABLE CONCENTRATION The MINIMUM DETECTABLE CONCENTRATION (MDC) is defined, for purposes of the controls in this ODCM, as the smallest concentration of radioactive material in a sample that will yield a not count above system background and that will be detected with 95-percent probability, with only 5-percent probability of falsely concluding that a blank observation represents a-

"real" signal.

For a particular measurement system, which may include _ radiochemical -

separation, the MDC for a given radionuclide is determined as follows (Reference 12): ,

2. 71 + 3. 29 Jth - - + - -

t, y , c, cy, MDC =

E*Va 2.22 x 106 , y , , -A At lO l where: I i

MDC =

the a priord MINIMUM DETECTABLE CONCENTRATION-(yci per unit mass or volume).

2.71 =

the square of the standard normal variate (1.645) for the 95 percent confidence level (Ref. 12,Section II.D).

3.29 =

Two times the standard normal variate (1.645) for the 95 percent confidence level (Ref. 12,Section II.C).

Rb= the background counting rate, or the counting rate of a blank sample, as appropriate (counts per minut'e).

t, =

the length of the sample counting period (minutes).

tb.= the length of the background counting period'(minutes).

B=

the counting efficiency-(counts per disintegration)-

V=

the sample size (units of mass or volume).

2.22 x 10 6 = the number of disintegration's per minute per yci.

Y=

the fractional' radiochemical yield, when applicable.

1= the radioactive decay constant for the given radionuclide (h'I) . Values of A used in effluent calculations should be based on decay data from a recognized and- current source, such as Reference 15.

10-3 Gen. Rev.-13

FNP-0-M-011 At =

for effluent samples, the elapsed time between the midpoint of sample collection and the time of counting O

(h);

for environmental samples, the elapsed time between the end of sample collection and the time of counting (h).

Typical values of E, V, Y, and At should be used in the calculation.

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

PRINCIPAL GAMMA EMITTERS The PRINCIPAL GAMMA EMITTERS for which the MINIMUM- DETECTABLE CONCENTRATION (MDC) limit applies include exclusively the following radio-nuclides:

e For liquid radioactive affluents: 'Mn-54, Fe-59, co-58, co-60, i Zn-65, Mo-99, Cs-134, Cs-137,.and Co-141. Co-144 shall' also be measured, but with an MDC of 5 x 10~0 pCi/mL.

e For gaseous radioactive affluents: In noble gas releases, Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, Xe-138; and in particulate releases, Mn-54, Fe-59, Co-58, co-60, 2n-65, Mo-99, Cs-134, Cs-137, O

D Co-141, and Co-144.

e For environmental media: The gamma emitters specifically listed in Table 4-3.

These lists do not mean that only these nuclides are to be considered.

Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report, the Annual Radiological Environmental Operating Report, or other applicable report (s).

~r SITE BQ4M" For ty .].)eggoes of effluent controla defined in the ODCM, the SITE BOUNDAMB shall be as shown in Figure 10-1.

UNRESTRICTED AREA The UNRESTRICTED AREA shall be any area access to which is neither limited nor controlled by the licensee or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

! f 10-4 Gen. Rev. 13

FNP-0-M-011 '

10.2 TERMS DEFINED IN THE TECHNICAL SPECIFICATIONS O The following terms are defined in the Technical Specifications, Section 1.0.

Because they are used throughout the Limits of Operation sections of the ODCM, ,

4:

they are presented here for convenience. In the event of discrepancies between t*

the definitions below and those in the Technical Specifications, the Technical Specification definitions shall take precedence.

ACTIONt5)

An ACTION shall be that part of a control that prescribes remedial measures required under designated conditions.

Q]Q M EL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel, such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock, and/or trip functione and may be performed by any series of sequential, overlarping, or total .

channel steps, such that the entire channel is calibrated.

CHANNEL CHECK !

, 4-A CHANNEL CHECK shall be the qualita'.ive assessment of channel behavior during operation by observation. Thi s determination shall include, where O- possible, comparison of the channel indicacion and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

l CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall bei e

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

Bistable Channels - the injection of a simulated signal into the

. sensor to ' verify OPERABILITY including alarm, and/or trip functions.f/

3:

DOSE ROUIVA MWT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (pci/g) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, 1977.

A 10-5 Gen. Rev. 13 l

'?

4 FNP-O-M-011 8

, FREQUENCY NOTATION i

The FREQUENCY NOTATION specified for the performance of surveillance requirements shall corraspond to the intervals defined below, with a +

j maximum allowable extension not to exceed 25% of the surveillance b l interval, i

\: NOTATION FREQUENCY l 5 (Once per shift) At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

W (Weekly) At least once per 7 days. ,

M (Monthly) At least once per 31 days.

Q (Quarterly) At least once per 92 days.

, SA (Semi-annually) At least once per 184 days. i R (Refueling) At least once per 18 months.

S/U (Startup) Prior to each reactor startup.

1 NA Not applicable.

P (Prior) Completed prior to each release.

i MODE for OPERATIONAL MODE)

An OPERATIONAL MODE shall correspond to any one inclusive coabination of 3

core reactivity condition, power level, and average reactor coolant

, temperature specified in section 1.0 of the Technical specifications.

OPEIULBLE f or OPERABILITYl OPERABILITY exists when a system, subsystem, train, component or device is capable of performing its specified function (s), and when all necessary !

t l attendant instrumentation, controls, electrical power, cooling or seal j

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

1 4

RATED THERMAL POWER RATED

@ shall be a total reactor core heat transfer rate to the ef 2652 MWt.

+ -y-

'I

~

'A 'l be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. '

THERMAL POWER THIRMAL POWER shall be the total reactor core heat transfer rate to the

{, reactor coolant.

O 10-6 Gen. Rev. 13 i

.e

,- - n --sm

3 i

FNP-0-N-011 VENTILATION EXMAUST TREATMENT SYSTEM The VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine or radioactive -material in particulate form in affluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous . exhaust stream prior to the release to the environment (such a system is not considered to have any effect on any noble gas effluents). Engineered Safety Feature (ESP) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. This system consists of the radwaste filtration unit, fuel pool exhaust filtration units and associated '

components providing for treatment flow and functional control. '

f

. 1

[

i i

Wh - ry :9f*; %

t O -

10-7 Gen. Rev. 13

~- , ,-

y PNP-0-N-011 '

O '

I N NS.000 , PROG (RTYLtgt,@gy!#ganvfg@MMpq N. - .

- 6

\,

e,i UNIT No.2-CONTAINhtENT AND AUXILIARY BUILDING WASTE LINS UNITNO.3

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,g }

m g wAmtes ,

t UNIT NO.1 e f

i 4 UNIT NO.1 - '

CONTAINhltNT '

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TRAININS/ lop CSNTER,,,,,,,

ANG E i RIVER ,

t i INTAKE f i

' $8RVIC8 WAftR IT8UETU"I 1

\ INTAE8 STRUCTURE t I

)

,7 DISCNARGE.

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L, mA POIIET

~ )

9 ,

==

,s

. e,+ : -

y= gg*f y3 ** ~ ~ ** i R. * '

Figture 20-1.

Site Map for Effluent controls 10-8 Gen. Rev. 13:

-

ML20063H768

Text

SUMMARY

SUMMARY

==

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