ML20205T008
| ML20205T008 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/23/1998 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML20205S953 | List: |
| References | |
| FNP--M-11, FNP-0-M-011, NUDOCS 9904270110 | |
| Download: ML20205T008 (200) | |
Text
SHARED
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FIGURE 1 h PROCEDURE REQUEST FORM (PRF)
(o') (1) FNP-0-M-011 Procedure Number 17 current nevhion g g Procedure
Title:
Offsite Dose Calculation Manual _ _ . . . _ . _ . .
UAU i IUN; This copyis notmaintained O New E Revision # 18 O Deleti8f*"' "'""'"" * ***d #
1 E Safety Related O Non- Safety Related O Infrequently Performed Test or Evolution (2) Description of Change (s) and Page Number (s) Affected: Table 4-4 p.4-14: chanced the milk sampline location from Bruce Ivey Dairy in Webb. Alabama to Robert Weir Dairy in Donalsonville. Georeia: Finure 4-3 p. 4-17: deleted Bruce Ivey Dairy as the milk sampline location and added in its place Robert Weir Dairy. !
Reason for change: Bruce Ivey Dairy has eone out of business and therefore milk samoles are no loneer available from this location. A new sample location has been identified (Robert Weir Dairv) and has been added to the ODCM in olace of Bruce Ivey Dairv.
(3) Prepared by: %dM it.frileg
[s su. . o.i, (4) 10CFR 50.59 and if applicable 50.54(q), or 50.54(p) Review Form attached: E Yes l Emironmental Evaluation Determination was Required (if yes, attach Figure 4): 0 Yes E No Commitment Update Required (if yes, attach Figure 5): 0 Yes E No PORC review required (AP-2 or Safety Evaluation) E Yes O No Licensing Document Chan Request,R uired (if yes, complete AP-98): E Yes O No Reviewed By: /2 /7-17
/ / si '.im o.i, (5) Cross-Disciplinary Review:
Group Sienature Title Date M I fam amm m m- ,
PORC Resie / /217 fd' (6) Final Approval O Group Supervisor /
O Manager /
O MSAER /
O Vice President -Project /
O - > /S_ / i I Nuclear Plant General Manager ~8 8 C / /2 M 3/f~
( sign Gre
/p. ,/
Nge 1 f1 Revision 38 9904270110 990421 PDR ADOCK 05000348 PDR QC(_ \ R
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SHARED RTYPE A6.35 MGURE1
[ FARLEY NUCLEAR PLANT LICENSING DOCUMENT CHANGE REQUEST LDCR NO:__ _ 1 TOTAL A'ITACHMENTS SWFT 1 OF:_ 3 SECTIONI:
Preparer: L H. Davis '
Date:__ December 15.1998 Licensing Document:
Offsite Dohe Calculation Manual Change
Description:
See LDCR Continuation Sheet l
Reason LDCR for Change:
Continuation _The Sheet for details.l chances involve chances to the location of a radiolorica) envirer .
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h SECTIONII: Does this change:
1.
Yes W No Require a change to the: (Identify whieh)
O ohratingLicense O TchhnicalSpecifications ,
j Q EnhonmentalProtection Plan If the answer to question (1)is yqs, perform a 10 CFR 50.92 evaluation. NRC approval will be implementation of this change. B. asis:
Although mentioned in the FSAR and Technical Specifications, the ODCM is a stand alone document. The ypoposed changes to the ODCM (described on the LDCR Con q not impact any of the programmdic controls for radioactive effluents contained in Section 6.8.3.e o
) Specifications. The changes involve changing the location of a radiological environmental m station.
- 2. R Yes M No Require a change to the Quality Assurance Program (FSAR 17.2)7 If the answer to question (2)is ye , perform a 10 CFR 50.54(a) evaluation. NRC approval m implementation of this change. Changes to the PSAR 17.2 also require a 10CFR50.59 evaluation items in FSAR Section 3A may also be considered as part of the QAP. Basis: The ODCM establishe' over routine radiological effluent!{ and does not alter or affect the Quality Assurance Progrt.m.
3.
M Yes @ No Require a change to any of the following: (Identify which)
Q Seettrity Plan (contains Contingency Plan) l Q Sect.rity Personnel Training and Qualification Plan lf the answer implementation to change.
of this question Ba (3) is yes( perform a 10 CFR 50,54(p) evaluation. NRC approva not alter or affect the Security Piar sis: The ODCM establishes controls over routine radiological effluents and does
- of the Security PersonnelTraining and Qualification Plan. ;
- 4. R Yes M No Require a change to the Emergency Pian?
l If the answcr to question (4) is yes, perform a 10 CFR 50.54(q) evaluation. NRC approval m implementation of this change. Basis: The ODCM establishes controls over routine radiological e not alter or affect the Emergency Plan.
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FNP-0-AP-98 RTYPE A6.35
__ FIGURE 1 5.
@ Yes ] No i Require s change to: (Identify which)
O O the p ant as described in the FSAR or requiros a revision or an addition to the PSAR O TechnicalSpecificationBases Q Technical Requirements Manual O to procedures as described in the PSAR Q involve a test or experunent not described (TRM)(whenissued)
Q PressureTemperatureLimits in the!FSAR Report (PTLR)(whenissued)
Q offsif Dose Calculation Manual (ODCM) O Core Operating Limits Repon U Process ControlProgram(PCP)
(COLR)
Q Fire IIar.ard Analysis (FHA)(part of FSAR)
If the answer to question (5)is y :s, perform: (Identify which)
Q a comolete 10 CFR 50.59 cvsluation.
O a 10 CFR 50.59 screening fa( editorial changes. (Refer to PNP-0-AP-88) 6.
Constitute a matter which could r a.
asult in adverre environmental impact (either direct or indirect)? Check (a) or (b
@ No The nature o
- this matter is such that it will not produce conditions which could result in adverse esvironmentalimpact.
- b. C Possibly (Explain B 'efly):-
i Ifitem (b) is checked, this matter must be referred to Southern Nuclear Environmental Services for preparation of an Environmental Evaluation.
SECTIONIII j
Based on the evaluations performeh for this change, is NRC approval required prior to {
implernentation of this change? Yes x No O = implementation of this han e7 Based on the evaluations perforn# for b Yes No LDCR Preparer: K[, O eg[s.
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Date: /J ~ /s T LDCR Review: [ Date: /2. V 6 , P Cross Disciplinary Review _ N72'I.
- Group: _bES Date: 3.D /t !ff Cross Disciplinary Review:
Group: Date; FNP Responsible Dept. Manager:
L i Date: /2d7" N LDCR Coordinator: j _i[
Date:
SECTIONIV j PORC Review: Date: '
_ PORC Meeting No. 3273 SECITON V '
Follow-up Action:(Ifrequired)
Change Implernented: Document No. & Rev.:
Closed By:-
Date:
O !
I Revision 2
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RTYPE A6.35 I FIGURE 2 LDCR CONTINUATION SHEET LDCR NO. Sheet __ 3 of __ 3 l
Explanation of response to Item Namber 5: !
involve changes to the location of al radiological environmental monitanng station.T i i !
Changes:
i
- 1) On page 4-14, Table 4-4 (contd', change the milk sampling control location from "Bmee Ivey (W 12 miles) to " Robert Weir Dair ; Donaldsonville, GA (SSE 14 miles)" ,
I 2)
On page 4-17 Figure 4-3, deldte the location location of the Robert Weir Dairy m ik sampling location.
of the Bruce Ivry Dairy milk sampling location a I
Justifications: I I i 1)
The Bruce Ivey Dairy which ser'ved as the milk sampling control location went out of business samples are no longer available at this location. A new milk sampling control location has been identif l Robert Weir Dairy; Donaldsonville, pA.
2)
The Bruce Ivey Dairy was delbted and location of the new milk sampling con' trollocation.
the Robert Weir Dairy added to Figure 4-3 to correctly indic i
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GO-NG-42 FIGURE 1 FARLEY NUCLEAR 10 CFR 50.59 EVALUATION SUPPORT SHEET 1OF 4 __
A Unit () One [] Two V [x] Shared: Evaluation Revision Number:
Number:
Document Number: Offsite Dose Calculation Manual Revision or TCN Number: 18
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FNP-0 M-011 B 10 CFR 50.59 SCREENING Does the document to which this evt.luation applies represent:
- 1. [] Yes [x] No A change to the plant as described in the FSAR, or will this change require a revision er .an addition to some portion of the FSAR7 Basis for Answer: The proposed activity consists of changes to the ODCM by which the radiologica environmental monitoring program is changed by relocating the milk sampling control station. Therefore, chan;-s are administrative in nature and do r et require any physical changes to the plant as described in the FSAR.
- 2. [x] Yes [] No A change to the procedures described in the FSAR7 Bas.4 for Answer The ODOM is referenced in the FSAR and thereby is regarded indirectly as a proce described in the FSAR. Tha proposed activity involves changing the ODCM by changing the radiologi environmental monitoring program by relocating the milk sampling control station.
- 3. [] Yes [x] No A test or experiment not described in the FSAR7 Basis for Answer: The proposed activity consists of changes to the ODCM that are administrative in nature.
The proposed activity does not in.roduce any abnormal method of effluent control and therefore does not involve any non-previously analyzed test or expenment.
- 4. [] Yes [x] No
- A change to the Technical Spee2fications and/or Environmental Protection Plan incorporated in the operating license?
Basis for Answer: Although the technical specifications do mention and briefly describe the ODCM, these proposed changes to the OECM do not change the programmatie controls for radioactive effluents contained in technical specifications. The proposed changes do not affect effluent monitor serpoint calculations or offsite doses. The proposed changes to the ODCM, which establishes controls over radioactive effluents, do not affect the Environmental Protection Plan. which establishes controls over non-radiological effluents.
If ANY of the four questions in Section B src answered "Yes", then PORC review of the safety evaluation is req implementation and a change to a licensed d6eument is indicated. Refer to FNP-0 AP 98.
Preparer- -- //. l)E Date:4-4-& Reviewed By: [ Date: '2-/6 &
Reviewer: N, Date:/2 /c. ff Reviewed By:
- Date: E/W Reviewed By: Date:
//
Approved By: Date:
Revicwed By: Date: FNP Approved: ,hLbh Date:O Reviewed By: __ Date: PORC Revie Date: /2*M'Y Reviewed By: Date: NORB Review: _
Date:
O
GO-NG-42 i
. FIGURE 2 I FARLEY NUCLEAR SUPPORT i Os SAFETY EVAL.UATION SHEET 2 OF 4 EAEFU EVALUATION:
D 1. [] Yes [x No May the proposed activity increase the probability of occurrence of an acendent previously evaluated in the FSAR?
B6 sis for Answer; h prgposed activityinvolves changes to the ODCM that are administrative in nature proposed activity will not riegrade the ability of any system, structure, or component to perform its d function. Therefore, the proposed activity will not increase the probability of occurrence (if an accident previously evaluated in the PSAR.
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- 2. [] Yes (x) No May the proposed activity increase the consequences of an accident previously evaluated in ttw FSAR7 Basis for Answer: The proposed changes to the ODCM are administrative in nature and do not ,
in the processing of radioactive effluems. The proposed changes will not aher any conditions o !
on which the PSAR accident analyses are based nor will they affect a structure or component wh to mitigate the consequencos of an accident. Therefore, the proposed changes will not inercase the consequences of an accider.t previously evaluated in the FSAR. ;
- 3. [] Yes [x] No f
U May the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR7 Basis for Answer: The proposed changes to the ODCM are administrative in nature and do not ch safety-related equipment is operated or maintained. The proposed activity will not introduce any new sys interaction or adversely impact the reliability of any equspment important to safety. The proposed a not degrade the ability of ar;y system, structure, or cornponent to perform the safety functions described in the FSAR. Therefore, the propssed changes will not increase the probability of occurrence of a malfunction of equipment important to sofoty previously described in the PSAR. '
- 4. [] Yes [x] No May the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the FSAR?
Basis for Answer:
The proposed changes are administrative in nature and do not change the processing of radiological effluents. The)roposed changes do not change any conditions or assumptions paviously evaluating the consequence: of a malfunction of equipment important to safety as discussed in the FSAR nor will they affect any structum, system, or component which is required to mitigate the radiological consequences of an acciden:. Thus the proposed changes to the ODCM will not increase the consequences of a malfunction of safety-related equipment previously evaluated in the FSAR.
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GO-NG-42 '
FARLEY NUCLEAR SUPPORT SAFETY EVALUATION SHEET 3 OF 4 SAFETT EVALUATION:
D 5. I) Yes [x] No May the proposed activity create the possibility of an accident of a different type than any previously evaluated in the FSAR?
Basis for Answer: The proposed changes to the ODCM are administrative in nature and do physical changes to the plant or to the way in which any system, structure or component is opera maintained. Consequently, the proposed changes do not introduce a different type of failure or existing failure probability more credible. Therefore, the proposed changes will not create the po accident of a different type than any previously evaluated in the PSAR,
- 6. [] Yes [x] No May the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the PSAR?
Basis for Answer:
The proposed changes to the ODCM are administrative in nature and do not result in new system interactions nor: change the way in which any system, structure or component is opera maintained. The proposed changes do not introduce any new failure modes or limiting single failures.
Therefore, the proposed chariges will not create the possibility of a malfunction of equipment of a different type than any previously evaluated in the FSAR.
- 7. [] Yes [x] No ,
Does the proposed activity reduce the margin of safety as defined in the basis for any Technica l Specification?
D y Basis for Answer: The proposed changes to the ODCM are administrative in natute and will not impact the svailability or capability of any system, structure or component that is governed by e operational in the technical specifications. Therefore, the proposed changes will not reduce the as defined in the basis of any technical specification.
If the answer to ANY of the seven questfons in Section D is "Yes", an unreviewed safety qtwstion ma indicated. Approval from the NRC is swluired krfsg the document / activity may be implemented, e
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GO-NG-42 FIGURE 3 10 CFR 50,59 SAFETY EVALUATION WRTITEN DESCRIPTION SHEET 4 OF- 4 E Unit Number:
() One [] Two [x] Shared Document Number: Offsite Done Calculation Manual Revision or TCN Number IR
Background:
Section 4.2 of the ODCM requires that Tabic 4-4 and Figure 4-3 specify the locations at which measurements and samples are taken for the radiological environmental monitoring program.
The Bruce ivey Dairy Webb, Alabatt a, at which control milk samples have been collected, went out o has been identified as a replacement controllocation for collecting milk .
The proposed changes to the ODCM ievise page 4-14, Table 4-4 (contd), and page 4 17, Figur milk sampling Donaldsonville, control location at the liruce Ivey Dairy, Webb, Alabama, and replacing at with Georgia.
O Rev.5 t
p-FNP-0-M-011 p December 17, 1998
( Revision 18 SOUTHERN NUCLEAR OPERATING COMPANY JOSEPH M. FARLEY NUCLEAR PLANT FNP-0-M-011 1
OFFSITE DOSE CALCULATION MANUAL l 1
S A
PROCEDURE USAGE REQUIREMENTS-per FNP-0-AP-6 SECTIONS E Continuous Use f .
Reference Use R
[ Information Use ;ALL E
( L A
T UNCONTROLLED COPY Approved CAUTION. ppyis notmsntamed j cunent. Do not use in a W NW W-Nu61efaf plaft1"-General Manager i
Date Issued: /M*NN List of Effective Pages Page Rev.
i to vii, 2-1 to 2-8, 2-11 to 4-3, 4-5 to 4-9, 4-11 to 4-12, 4-15 to 4-16, 4-18 to 7-1, 7-3 to 10-1, 10-0 13 14 l
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viii, 4-10, 7-2 15 1-1, 4-4, 4-13, 10-2 to 10-7 16 l
2-9 to 2-10 17 4-14, 4-17 18 l
O('
i FNP-O M M DISTRIBUTION LIST i
For information pertaining to distribution of the ODCM, contact Farley Nuclear l Plant Document Control.
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t FNP-0-M-Og TABLE OF CONTENTS m
RAGE DISTRIBUTION LIST . . . . .
....................... 1 TABLE OF CONTENTS . . . . . .
...................... ii LIST OF TABLES . . . . . . . ...................... v LIST OF FIGURES . . . . . . .
...................... vii REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . viii CHAPTER 1: INTRODUCTION . .
...................... 1-1 CHAPTER 2: LIQUID EFFLUENTS ...................... 2-1 2.1 LIMITS OF OPERATION 2-1 2.1.1 Licuid Effluent Monitorina Instrn==ntation Control 2-1 2.1.2 Licuid Effluent concentration Control 2-7 2.1.3 Licuid Effluent-Dose Control' 2-11 2.1.4 Licuid Radwaste Treatment System Control 2-13 2.1.5 Maior Chances to Liould Radioactive Waste Treatment Systems 2-14 2.2 LIQUID RADWASTE TREATMENT SYSTEM 2-15 2.3 LIQUID EFFLUENT MONITOR SETPOINTS 2-19 2.3.1 General Provisions Reaard4na Setnoints 2-19 2.3.2 Setnoints for Radwaste Svatem Discharoe Monitors 2-21 2.3.3 Setnoints for Monitors on Normally Low-Radioactivity Str==== 2-29 2.4 LIQUID EFFLUENT DOSE CALCULATIONS 2-30 2.4.1 Calculation of Dome 2-30 2.4.2 Calculation of A9 2-31 2.4.3 Calculation of CFw 2-32 2.5 LIQUID EFFLUENT DOSE PROJECTIONS 2-42 2.5.1 Thirty-One Day Dose Proiections 2-42 2.5.2 Dose Proiections for Soecific Releases 2-42 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS 2-43
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CHAPTER 3:- GASEOUS EFFLUENTS . . . . . . . . . . . . . . . . . . ... 3-1 3.1 LIMITS OF OPERATION 3-1 3.1.1 Gaseous Effluent Monitorina Tnatrn==ntation Control 3-1 3.1.2 Gaseous Effluent Dose Rate Control 3-6 3.1.3 Gaseous Effluent Air Dose Control 3-10
[- 3.1 4 Control on Gaseous Effluent Dose to a Member of the Public 3-12 11 Gen. Rev. 13
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FNP-0-M-011 TABLE OF CGiur.dia (Continued)
RAGE 3.1.5 Gaseous Radwaste Treatment System Control 3-14 3.1.6 MA670R CHANGES to the GASEOUS RADIOACTIVE WASTE TREATMENT SYsima4 and the VENTILATION RYM2 LUST TREATMENT SYSTEM 3-16 3.2 GASEOUS RADWASTE TREATMENT SYSTEM 3-17 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS 3-19 3.3.1. General Provisions Recardina Noble Gas Monitor Setnoints 3-19 3.3.2 Setnoint for the Final Noble Gas Monitor on Each Release Pathway 3-21 3.3.3 Setnoints for Noble Gas Monitors on Effluent Source Streams 3-25 3.3.4 Determination of Allocation Factors. AG 3-28 3.3.5 Setnoints for Noble Gas Monitors with Snecial Reauirements 3-31 3.3.6 Setnoints for Particulate and Iodine Monitors 3-31 3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3-32 3.4.1 Dose Rates at and Bevond the Site Boundarv 3-32 3.4.2 Noble Gas Air Dose at or Bevond Site Boundarv 3-33 3.4.3 Dose to a Member of the Public at or Beyond Site Boundarv 3-37 3.4.4 Dose Calculations to Sunnort Other ReauiE nts 3-40 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS 3-46 3.5.1 Thirty-One Day Dose Proiections 3-46 3.5.2 Dose Proiections for Snecific Releases 3-47 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS 3-48 CHAPTER 4: RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM . . . . . . . 4-1 4.1 LIMITS OF OPERATION 4-1
'4.1.1 Radiolocical Envirnnmantal Monitorino 4-1 4.1.2 fand Use Canmus 4-8 4.1.3 Interlaboratory Comnarison Proaram 4-10 4.2 RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS 4-11 CHAPTER 5: TOTAL DCSE DETERMINATIONS . . . . . . . . . . . . . . . . . 5-1 5.1 . LIMIT OF OPERATION 5-1 5.1.1 Annlicability 5-1 5.1.2 Actions 5-1 I 5.1.3 Surveillance Reauiramants 5-2 S.1.4 B3313 -
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 i iii Gen. Rev. 13 1
FNP-0-M-011 TABLE OF COlum. dis (Continued) 2.MiE 6.2 CALCULATIONE METHOD 6-1
. CHAPTER 7: TdPORTS . . . . . . . ................... 7-1 7.1 ANNUAL RADIOLOGICE'ENVIRONMENTE OPERATING REPORT 7-1 7.1.1 Reauirement for Reoort 7-1 7.1.2 Reoort Contents 7-1 7.2 ANNUE RADIOACTIVE EFFLUENT RELEASE REPORT 7-3 7.2.1 Reauirement for Reoort 7-3 7.2.2 Reoort contents 7-3 7.3 MONTHLY OPERATING REPORT 7-7 7.4 SPECIAL REPORTS 7-7 j CHAPTER 8: METEOROLOGICAL MODELS ................... 8-1 8.1 ATMOSPHERIC DISPERSION 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 O
V 8.2.2 8.2.3 Elevated Releases Mixed-Mode Releases 8-7 8-8 8.3 ELEVATED PLUME DGSE FACTORS 8-8 8.4 METEOROLOGICAL
SUMMARY
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CHAPTER 9: METHODS AND PARAMETERS FOR CECULATION OF GASEOUS EFFLUENT PATHWAY DOSE FACTORS, Raid * * * * * * * * * * * * * * * *
- 8-1 S.1 INHALATION PATHWAY FACTOR 9-1 9.2 GROUND PLANE PATHWAY FACTOR 9-2 9.3 GARDEN VEGETATION PATHWAY FACTOR 9-3 9.4 GRASS-COW-MILK PATHWAY FACTOR 9-6 9.5 GRASS-GOAT-MILK PATHWAY FACTOR 9-10 9.6 GRASS-COW-MEAT PATHWAY FACTOR 9-14 i
CHAPTER 10: DEFINITIONS 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
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FNP-0-M-011 LIST OF TABLES
- 2. AGE Table 2-1. 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 Analysis Program 2-9 Table 2-4. Applicability of Liquid Monitor Setpoint Methodologies 2-20 Table 2-5. Parameters for Calculation of Doses Due to Liquid Effluent Releases 2-35 Table 2-6. Element Transfer Factors 2-36 Table 2-7. ' Adult Ingestion Dose Factors 2-37 Table 2-8. Site-Related Ingestion Dose Factors, Ajy . 2-40 Table 3-1. Radioactive Gaseous Effluent' Monitoring Instrumentation 3-3 Table 3-2. Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3-5 Table 3-3. Radioactive Gaseous Waste Sampling and Analysis Program 3-8 Table 3-4. Applicability of Gaseous Monitor Setpoint Methodologies 20 Table 3-5. Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases 3-35 Table 3-6. Dose. Factors for Exposure to Direct Radiation from Noble Gases in~an Elevated Finite Plume 3-36 Table 3-7. Attributes of the Controlling Receptor- 3-39 Table 3-8, R,;pj for Ground Plane Pathway, All Age Groups 3-42 Table 3-9. Raipj for Inhalation Pathway, Child Age' Group 3-43 Table 3-10. Raipj for Cow Meat Pathway, Child Age Group 3-44 Table 3-11. Raipj for Garden Vegetation Pathway, Child Age Group 3-45
. Table 4-1. Radiological Environmental Monitoring Program 4-4
-Table 4-2. Reporting I4vels for Radioactivity Concentrations in Environmental Samples 4-6 i
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 i Inside the SITE BOUNDARY 6-3 Table 8-1. Terrain Elevation Above Plant Site Grade 8-9 {
Table 8-2. Annual Average (I76) for Mixed Mode Releases 8-10 Table 8-3. Annual Average. (176) for Ground-Level Releases ~
8-11 Table 8-4. Annual Average. (676) for Mixed Mode Releases 8-12 Table 8-5. ' Annual. Average (B76) for Ground-Level Releases 8-13 Table 9-1. {
Miscellaneous Parameters for the Garden Vegetation i Pathway 9-5 Table 9-2. Miscellaneous Parameters for the Grass-Cow-Milk Pathway 9-9 l
v Gen. Rev. 13 l
FNP-0-M-011 LIST OF TABLES (Continued)
[ Eh95 \
Table 9-3. i Miscellaneous Parameters for the Crass-Goat-Milk Pathway 9-13 Table 9-4. {
Miscellaneous Parameters for the Grass-Cow-Meat Pathway 9-17 Table 9-5. Individual Usage Factors 9-18 Table 9-6. Stable Element Transfer Data 9-19 Table 9-7. Inhalation Dose Factors for the Infant Age Group 9-20 Table 9-a. 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 9-41 Table'9-15. External Dose Factors for Standing on Contaminated Ground 9-44 l
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FNP-0-M-011 LIST OF FIGURES l
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E Figure 2-1. 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 2-18 Figure 3-1. Schematic Diagram of Routine Release Sources and Release Points (Typical of Both Units) 3-18 Figure 4-1. Airborne Sampling Locations, 0-5000 feet 4-15 Figure 4-2. Indicator II (Community) Sampling Locations for Direct Radiation 4-16 1 Figure 4-3. 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 (a z) 8-14 Figure 8-2. Terrain Recirculation Factor (K r) 6-15 Figure 8-3. Plume Depletion Effect for Ground Level Releases 8-16 Figure 8-4. Plume Depletion Effect for 30-Meter Releases 8-17 Figure 8-5. Plume Depletion Effect for 60-Meter Releases 8-18 Figure 8-6. Plume Depletion Effect for 100-Meter Releases 8-19 Figure 8-7. Relative Deposition for Ground-Level Releases 7 8-20 Figure 8-8. Relative Deposition for 30-Meter Releases 8-21 g- Figure 8-9. Relative Deposition for 60-Meter Releases 8-22 Pigure 8 10. Relative Deposition for 100-Meter (or Greater) Releases 8-23 Figure 10-1. Site Map for Effluent Controls 10-8 !
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d FNP-0-M-011 REFERENCES
- 1. J.S. Boegli, R.R. Bellamy, W.L. Britz, and R.L. Waterfield, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," NUREG-0133, October 1978 l
2.
" Calculation of Annual Doses to Man froin Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with.10 CFR 50, Appendix I," U.S. NRC Reaulatorv Guide 1.109, March 1976.
' 3. " Calculation of Annual Doses to Man from Routine Releases of Recctor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," U.S. NRC Reaulatory Guide 1.109; Revision 1, October 1977
- 4. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. NRC Reaulatory Guide l'.111, March 1976.
- 5. " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. NRC Reculatorv 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 Reaulatory Guide 1.113, April.1977.
- 7. Joseeh M. Parlev Nuclear Plant tinits 1 and 2 Final Safety Analysis ReDort, Alabama Power Company.
- 8. Josech M. Parlev Nuclear Plant tin 4 ts 1 and 2 E'avironmental Recort -
Operatina Licanne Staae, Alabama Power Company.
- 9. .T.E. Young, T.S. Bohn, and W. Serrano, " Technical Evaluation Report for the Evaluation of ODCM Revision 7 for Joseph M. Farley Nuclear Plant, Units 1 and 2," EGG-PHY.8674, dated August 1989, transmitted by NRC letter dated November 9, 1989.
- 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. Meyer, eds., Radioloaical Asses ===nt, U.S. NRC Report M G/CR-3332, 1983.
- 12. -L.A. Currie, Lower Limit of Detection Definition and Elaboration of a Pronosed Position of Radioloaical Effluent and Enviran==ntal Measurements, U.S. NRC Report NUREG/CR-4007, 1984.
- 13. " Radiological Assessment Branch Technical Position", U.S. Nuclear Regulatory Commission, Revision 1, November 1979.
- 14. U.S. DOE Report PNL-5484.
- 16. Internal Memorandnm. J.E. Garlinaton to D.N. Morev, Alabama Power -
Company, June 4, 1990.
O viii Rev. 15
FNP-0-M-011 CHAPTER 1 INTRODUCTION 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: f e
The controls required by the Technical Specifications, governing the radioactive effluent and radiological environmental monitoring programs. 9 I
Schematics of liquid and gaseous radwaste effluent treatment systems, which include designation of release points to UNRESTRICTED AREAS.
- A list and maps indicating the specific sample locations for the Radio-logical Environmental Monitoring Program, e Specifications and descriptions of the information that must be included in the Annual Radiological Etvironmental Operating Report and the Annual l 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 (References 4 and 5), and in Regulatory Guide 1.113 (Reference 6).
l 1-1 Rev. 16 i
j
r FNP-0-M-011 CHAPTER 2 LIOUID EFFLUENTS 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 Licuid Effluent Monitorina Instrumentation Centrol In accordance with Technical Specification 6.8.3.e(i), the radioactive liquid effluent 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.
2.1.1.1 Applicability This limit applies at all times.
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 effluents monitored by the affected channel, declare the channel inoperable, or change the setpoint to a conservative value.
With less than the minimum number of radioactive liquid affluent 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.
O 2-1 Gen. Rev. 13
FNP-0-M-011 2.1.1.4 Basis y/ The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in 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.
1 1
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2-2 Gen. Rev. 13
FNP-0-M-011 Table 2-1. Radioactive Liquid Effluent Monitoring Instrumentation
[
OPERABILITY Requirementsa natrument Minimum Channels Operable ACTION
- 1. Gross Radioactivity Monitors Providing Automatic Termination of Release
- a. Liquid Radwaste Effluent Line (RE-18) 1 28
- b. Steam Generator Blowdown Effluent Line (RE-23B) 1 29
- 2. Flowrate Measurement Devices
- a. Liquid Radwaste Effluent Line
- 1) Waste Monitor Tank No. 1 1 30 l
- 2) Waste Monitor Tank No. 2 1 30
- b. Discharge Canal Dilution Line (Service Water) 1 30
- c. Steam Generator Blowdown Effluent Line 1 30
- a. All requirements in this table apply to each unit.
l
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2-3 Gen. Rev. 13
FNP-0-M-011 Table 2-1 (contd). Notation for Table 2 ACTION Statements U
ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Chanriels OPERABLE requirement, effluent releases may continue for up to 14 days provided that prior to initiating a releases
- a. At least two independent samples are analyzed in accordance with Section 2.1.2.3, 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 it such calculations are performed manually.
Otherwise, suspend release of radioactive effluents via this Pathway.
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 gross radioactivity (be no greater than 1 x 10 ga or gamma) pCi/mL:at a MINIMUM DETECTABLE CONCENTRATION
- a. At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> when the specific activity of the A I secondary coolant is greater than 0.01 pCi/ gram DOSE EQUIVALENT '
( I-131.
- b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves may be used to estimate flow.
O j
't - 4 Gen. Rev. 13 l l
FNP-0-M-011 Table 2-2. Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements
\
Surveillance Requirementsd Instrument CHANNEL CHANNEL CHANNEL FUNCTIONAL CHECK SOURCE CHECK CALIBRATION TEST l
- 1. Gross Radioactivity Monitors Providing Automatic Temination of Release
- a. Liquid Radwaste Effluent Line (RE-18) D P RD Qa
- b. Steam Generator Blowdown Effluent Line (RE-23B) D M RD 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 D* NA R NA
- b. Discharge Canal Dilution Line (Service Water) DC NA R O
- c. Steam Generator Blowdown Effluent Line DC NA R NA l U :
2-5 Gen. Rev. 13 l
FNP-0-M-011 Table 2-2 (contd). Notation for Table 2-2
/ O V) a. In addition to the basic functions of a CHANNEL FUNCTIONAL TEST (Section 10.2):
(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 setpointi (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 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 p initial calibration shall be used.
\
- 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made,
- d. All requirements in this table apply to mach unit. l 1
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2-6 Gen. Rev. 13
]
r 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 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 entrained noble gases, the concentration shall be limited to 1 x 10 4 pCi/mL total activity.
2.1.2.1 Applicability This limit applies at all times. l
)
2.1.2.2 Actions with the concentration of radioactive material releared 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.
O Q 2.1.2.3 Surveillance Requirements l
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 2.3 to assure that the concentration at the point of release is maintained within the limits of Section 2.1.2.
2.1.2.4 Basis This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in
~
exposures within (1) the Section II. A design objectives of Appendix I,10 CFR 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 in air (submersion) was converted to an equivalent concentration in water using p) g the methods described in International Commission on Radiological Protection V
2-7 Gen. Rev. 13
FNP-0-M-011 (ICRP) Publication 2 (1959). The resulting concentration of 2 x 104 was then multiplied by the ratio of the effluent concentration limit for Xe-135, stated O- 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
- Ci/mL.
(j T
h i
O 2-8 Gen. Rev. 13
FNP-0-M-011 Table 2-3. Radioactive Liquid Waste Sampling and Analysis Program O
LJ Sampling and Analysis Rcquirements a,b MINIMUM Liquid DETECTABLE Minimum CONCENTRATION Release Sampling Analysis Type of Activity Type (MDC)
FREQUENCY FREQUENCY Analysis (uci/mL)
A. Waste Tanks Producing BATCH RELEASES PRINCIPAL GAMMA 5 E-7 P P EMITTERS l Each BATCH Each BATCH I-131 f 1 E-6 p Dissolved and 1 E-5 One BATCH /M Entrained Gases All (Gamma Emitters) i P M 1 E-5 Each BATCH COMPOSITE Gross Alpha 1 E-7 p g Sr-89, Sr-90 5 E-8 Each BATCH COMPOSITE Fe-55 1 E-6 O' B. CONTINUOUS RELEASESC PRINCIPAL GAMMA 5 E-7 D W EMITTERS Grab Sample COMPOSITE I-131 1 E-6 Dissolved and 1 E-5 l Steam Grab ample M Entrained Gases i Generator (Gamma Emitters)
Blowdown 3 g H-3 1 E-5 Grab Sample COMPOSITE Gross Alpha 1 E-7 D Q ~ #~' ~
Grab Sample COMPOSITE Fe-55 1 E-6 Turbine PRINCIPAL GAMMA 5 E-7 pd W EMITTERS Building l Sump Grab Sample COMPOSITE H-3 1 E-5 C
b' 29 Rev. 17
FNP-0-M-011 Table 2-3 (contd). Notation for Table 2-3
- a. All requirements in this table apply to each unit. Deviation from the MDC requirements of this table shall be reported in accordance with Section 7.2.
- b. Terms printed in all capital letters are defined in Chapter 10.
c.
Sampling environment, will be performed only if the effluent will be discharged to the
- d. Samples will be taken prior to or during each discharge. l l ,
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2-10 Rev. 17 l
FNP-0-M-011 2 .1. 3' Liauid Effluent Dose Control In accordance with Technical Specifications 6.8.3.e (iv) and 6.8.3.e (v) , the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 10-1) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ, and
- b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.
2.1.3.1 Applicability These limits apply at all times.
2.1.3.2 Actions With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the limits of Section 2.1.3, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification
('
6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s);
defines the corrective actions to be taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases 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 l determined, for each unit, in accordance with Section 2.4.
2.1.3.4 Basis This control is provided to implement the requirements of Sections II.A7 III.A and IV.A of Appendix I, 10 CFR Part 50. 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 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 s
11 Gen. Rev. 13 I I
FNP-0-M-011 is reasonably achievable. " Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is
/
reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculations in Section 2.4 implement the requirements in Section III. A of Appendix I, which state 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 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).
This control applies to the release of liquid effluents from each unit at the site. The liquid effluents from shared LIQUID RADWASTE TREATMENT SYSTEMS are to be proportioned between the units.
O
-k 2-12 Gen. Rev. 13
I 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 I 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 mrem to the total body or 0.2 mrem to any organ of a MEMBER OF THE PUBLIC in 31 days.
2.1.4.1 Applicability i 1
i This limit applies at all times.
2.1.4.2 Actions i
With radioactive liquid waste being discharged without treatment and in excess of the above limits and appropriate portions of the LIQUID RADWASTE TREATMENT 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 includes the following information:
1
- a. Explanation of why liquid radwaste was being discharged without treatment,
() identification of any inoperable equipment or subsystems, and the reason for the inoperability, l
l
- b. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- c. Summary description of action (s) taken to prevent a recurrence.
This control does not affect shutdown requirements or MODE changes.
2.1.4.3 Surveillance Requirements Doses due to liquid releases to UNRESTRICTED AREAS shall be projected at least once per 31 days, in accordance with Section 2.5, during periods in which the LIQUID RADWASTE TREATMENT SYSTEMS are not being fully utilized.
The LIQUID RADWASTE TREATMENT SYSTEM shall be demonstrated OPERABLE:
- a. by meeting the controls of Sections 2.1.2 and 2.1.3, or
/
2-13 Gen. Rev. 13
FNP-0-M-011
- b. by operating the LIQUID RADWASTE TREATMENT SYSTEM equipment for at least t
15 minutes at least once per 92 days unless the LIQUID RADWASTE TREAMENT g SYSTEM equipment has been utilized to process radioactive liquid effluents during the previous 92 days.
2.1.4.4 Basis The OPERABILITY of the LIQUID RADWASTE TREAMENT 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 w(re specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents. I j
1 l
2.1.5 Maior chances to Licuid Radioactive Waste Treatment Systems n
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.
O' 2-14 Gen. Rev. 13 1
'Tr-0-M-011 2.2~ LIQUID RADWASTE TREA'INENT SYSTEM The Parley Nuclear Plant is located on the west bank of the Chattahoochee River approximately 35 river miles 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 is illustrated schematically in Figure 2-1. However, both units share a common domineralizer 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.
All liquid radwastes treated by the LIQUID RADWASTE TREATMENT SYSTEM are 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 reprisentative sample can be taken from the tank. Releases from the waste monitor tanks are routed to the Service Water discharge line (which provides dilution prior to release to the UNRESTRICTED AREA), and thence to the Chattahoochee River. The Service Water discharge line also receives anput from the Cooling Tower Blowdown and the Turbine Building Sump.
Although no significant quantities of radioactivity are expected in the steam generator bl.swdown processing system, this effluent pathway is monitored as a precautionary measure. The monitors serving this pathway provide for automatic j termination of release in the event that radioactivity is detected above i predetermined levels. Like the LIQUID RADWASTE TREA'INENT SYSTEMS, the Steam Generator Blowdown Systems discharge to the Service Water discharge line.
One potential release pathway, the Turbine Building Sump discharge, is not i 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.
~
2-15 Gen. Rev. 13
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s:ms O Figure 2 2. Steam Generator Blowdown System (Typical of Both Units) 2-17 Gen. Rev. 13
FNP-0-M-011
\
O Unit 1 Unit 2 Service Water Service Water Train A Train B Train A Train B 4 Cooling Tower 31owdown >
Radwaste Discharge 11E018 2EE018 staan Generator Blowdown O 11E235 2RE235 4 Turbine Suilding Sep 7 V
To River O Figure 2-3. Liquid Discharge Pathways 2-18 Gen. Rev. i3 1
FE FNP-0-M-011 ;
2.3 LIQUID EFFLUENT MONITOR SETPOINTS' 2.3.1 - General Provisions Recardina Setoc Lat,g Liquid monitor setpoints calculated in accordanc e with the methodology presented in this section will be regarded as upper be inds for the ac.tual high alarm setpoints. That is, a lower value for the high 4 larm setpoint may be established or retained on the monitor, if desired. Internadiate level setpoints should be established at an appropriate level to give suf icient warning prior to reaching the high alarm setpoint. If no release is pla.ned for a particular pathway, or
-if there is no detectable activity in the plar ned release, the monitor setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should an inadvertent ;elease occur.
Two basic setpoint methodologies are prest ated 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 str .ams that are not expected to contain sign 1ficant radioactivity, the purpose of :he monitor setpoints is to cause an alarm on low levels of radioactivity, and ' .o terminate the release where this is possible. Section 2.1.1 establishes fne requirements for liquid effluent monitoring instrumentation. Table 2-4 7 tats the monitors for which each of the setpoint methodologies is applicable.
i 1
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l 2-19 Gen. Rev. 13
1 l
l FNP-0-M-011 j l
Table 2-4. Applicability of Liquid Monitor Setpoint Methodologies
(
- Liquid Radwaste Discharge Monitors Setpoint Method: Section 2.3.2 Unit 1 or Unit 2 Waste Monitor Tanks Effluent Release Type BATCH Monitor: 1RE-018 / 2RE-018 Unit 1 or Unit 2 bteam Gener'ator Blowdown Effluent Release Type: CONTINUOUS Monitor: 1RE-023 B / 2RE-023 B l
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.
O 2-20 Gen. Rev. 13
FNP-0-M-011
,2.3.2 Setnoints for Radwaste Svatem Discharoe Monitors 2.3.2.1 Overview of Method j i
i LIQUID RADWASTE TREATMENT SYSTEM effluent line radioactivity monitors are intended to provide alarm and automatic temination 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 from Addendum to Reference 1):
- I s TF CECZ, (2.1)
F+f where:
Cg - the Effluent Concentration Limit corresponding to the mix j of radionuclides in the effluent being considered for i i
discharge, in pCi/mL.
c= the setpoint, in pCi/mL, of the radioactivity monitor measuring the g 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 gpm. 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 effluent releases 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-V 2-21 Gen. Rev. 13
FNP-0-M-011 nuclides with different Effluent Concentration Limits (ECLs). For a mixture of p 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 assuud for the duration of the release (Fd ), by calculating the maximum permissible effluent flowrate (fm) and the radioactivity monitor setpoint (c).
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.
2.3.2.2 Setpoint Calculation Steps Sten 1: Determine the radionuclide concentrations in the liquid waste being 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 fm 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 dischrrge line is assumed to be well mixed, so that no additional mixing is required prior to sampling.
The total concentration of the liquid waste is determined by the results of all required analyses on the collected sample, as follows:
bCl "
C#+bCs+C+Cl f
- bC8 (2.2) i 3 8 where:
c-a the gross concentration of alpha emitters in the liquid waste, not !
less than that measured in the most recent applicable composite sample. -
C, = the concentration of st2contium radioisotope s (Sr-89 or Sr-90) in the liquid waste, not less than that measured in the most recent applicable composite sample, f~
t i 2-22 Gen. Rev. 13
FNP-0-M-011 Cf= the concentration of Fe-55 in the liquid waste, not less than that !
measured in the most recent applicable composite sample. I Ct= the concentration of H-3 in the liquid waste, not less than that measured in'the most recent applicable composite sample.
I Cg- the concentration of gamma emitter g in the liquid waste as measured by gamma ray spectroscopy performed on the sample for the release under consideration.
The C g 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 included in accordance with the sampling and analysis program required for the waste stream (see Section 2.1.2). For each 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,, Cg, and Ct shall be used.
Sten 2: Determine the required dilution factor for the mix of radionuclides detected in the waste. I Measured radionuclide concentrations are used to calculate ECL fractions. The ECL fractions are used along with a safety factor to calculate the required i 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- !
gamma-emitters (RDFnyl3 RDF =
bi -
IJ
+ [(SF) (TF))
(2.3)
=
RDFy + RDFm C
E KCL (2.4) 7 (SF) (TF) I e).
2-23 Gen. Rev. 13
FNP-0-M-011 Ca c, c,
,3 t
ECLa
,g . cf .
RDF =
D (SF) ( TF) where C = the measured concentration of radionuclide i as defined in step 1 in pCi/mL. The C a, C,, cf, and Ct terms will be included in the calculation as appropriate.
ECLj = 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 1x104 pCi/mL. For gross alpha, the ECL shall be 2x10-9 pCi/mL; if specific alpha-emitting radionuclides are measured, the ECL for the specific radionuclide(s) should be used.
SF = the safety factor selected 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).
Steo 3: Determine the release-specific assured dilution stream flowrate.
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 dilution 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:
2-24 Gen. Rev. 13 l
l
E FNP-0-M-011 Fgp
= Fg (ATp) (2.6)
\
Fdp = 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, under the condition that the sum of the AF p for all active release pathways for each unit 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 effluent 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 Fdp equal to Fd-1 For the case where RDF s 1, the planned release meets the limits of Section 2.1.4 O- without dilution, and may be released with any desired effluent flowrate and i
dilution flowrate.
Sten 4: Determine the maximum allowable waste discharge flowrate.
{
For the case where RDF > 1, the maximum permissible effluent discharge flowrate for this release pathway, fg (in gpm), is calculated as follows:
f = FdP (2.7)
(RDF - 1)
For the case RDF s 1, equation (2.7) is not valid. However, as discussed above, when RDF s 1, the release may be made at full discharge pump capacity; the radio-activity monitor setpoint must still be calculated in accordance with , Step 5 below.
NOTE 1: 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 O capacity. Discharge flowrates less than the pump c jacity must be 2-25 Gen. Rev. 13
FNP-0-M-011 achieved by throttling if this is available; if throttling is not
,q available, the release may not be made as planned.
! }
%J 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 addiHon, sampling and analysis of the other radioactive ef fluents affecting the dilution stream must be sufficient to ensure that the liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded. ) 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:
Y %
dp Q ci, tmp "
3_g (2.8}
(RDF - 1) , r Fg g , ECLj, ,
where:
Cir - the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the dilution stream.
(
(' '/ f r -
the effluent discharge flowrate of release pathway r.
If the entire dilution stream contains detectable activity due to plant operations, whether or not its source is identified, fr=F' d and Cr is the concentration in the total dilution system. This note does not apply: a) if the RDF of the planned release is s 1; or b) if the release contributing radioactivity to the dilution stream has been accounted for by the assignment of an allocation factor.
Steo 5: Determine the maximum radioactivity monitor setpoint concentration.
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. Becarse the radioactivity monitor responds primarily to gamma radiation, the monitor setpoint cp for release pathway p (in pCi/mL) is based on the concentration of gamma emitters in the waste stream, as follows:
/^\
i i N/
2-26 Gen. Rev. 13
FNP-0-M-011 cp = A p[cg gg,,y k
where Ap= an adjustment factor which will allow the setpoint to be established in a practical manner to prevent spurious alarms while
{
allowing a margin between measured concentrations and the limits I 1
of Section 2.1.2.
l Steo 5.a. If the concentration of gamma emitters in the effluent to j 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:
Ap = 1 x EF RDF l
, 1 , I#p+fap) d (2.10)
RDF fap l k where:
ADF = the assured dilution factor.
f,p = the anticipated actual discharge flowrate for the planned release (in gpm), a value less than f g.
The release must then be controlled so that the ;
actual effluent discharge flowrate does not exceed f,p at any time.
Steo 5.b2 Alternatively, Ap may be calculated as follows:
ADF - RDFm Ap= RDFy (2.11)
Steo 5.c. Evaluate the computed value of A as follows:
p If Ap = 1, calculate the monitor setpoint, c. p However, if cp is within about 10 percent of cg , it may be impractical to 2-27 Gen. Rev. 13
FNP-0-M-011 use this value of e. p This situation indicates that O measured concentrations are approaching values which 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 discharges these steps may include decreasing the planned effluent discharge flowrate, increasing the dilution stream flowrate, postponing simultaneous releases, and/or decreasing the affluent 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 steps 1-5 to calculate a new radioactivity monitor 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.
2.3.2.3 Use of the Calculated setpoint 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:
c; =
cEp+Bp p (2.Sa) where:
c = the monitor setpoint as a count rate.
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 determining the monitor response to stream concentrations measured by liquid sample analysis.
B p= the monitor background count rate. In all cases, monitor background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value. '
O O
2-28 Gen. Rev. 13 I
j J
FNP-0-M-011 The count rate units of c , E ,p and B in p equation (2.8a) must be the same (epm Lor eps),
V 2.3.3 Setooints for Monitors on Normally Low-Radioactivity Streams Radio &ctivity 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 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.
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:
- 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.
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 analysesi or the associated radioactivity monitor detects activity in the stream at levels above the established alarm setpoint.
When an elevated monitor setpoint 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 l 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 )
O effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.
(d !
2-29 Gen. Rev. 13
FNP-0-M-0;1,,
2.4 LIQUID EFFLUENT DOSE CALCULATIONS (qj The following sub-sections present the methods required for liquid effluent dose calculations, in deepening levels of detail. Applicable site-specific pathways and parameter values for the calculation of D 7, A 7, and CFjy are summarized in f
Table 2-5.
2.4.1 Calculation of Dose l
l 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 l 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 effluents released from each unit to UNRESTRICTED AREAS will be calculated as follows (equation from Ref-l erence 1, page 15):
m D7 =
{Aff { (Atf Cff F)f (2.12 )
i .1-1 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 the total of the m time periods Atg.
Ag7 = the site-related adult ingestion dose commitment factor, for the total body or for any organ 7, due to identified radionuclide i, in (mrem mL) / (h pci) . Methods for the calculation of A 7 are presented below in Section 2.4.2.
The values of Air to be used in dose calculations for releases from the plant site are listed in Table 2-8.
Atj = the length of time period 1, over which C gi and Fg are averaged for liquid releases, in h.
C 3= the average concentration of radionuclide i in undiluted liquid effluent during time period 1, in pCi/mL. Only radionuclid.es b
2-30 Gen. Rev. 13 l
FNP-0-M-011 identified and detected above background in their respective samples should be included in the calculation.
Fg = the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA:
fg Fl - (2.13)
Fg x Z where:
ft" the average undiluted liquid waste flowrste actually observed during the period of radioactivity release, in 9Pm.
Pg= the average dilution' stream flowrate actually observed during the period of radioactivity release, in gym.
Z= the applicable dilution factor for the receiving water body, in the near field of the discharge structure, during the- period of radioactivity release, from Table 2-E.
NOTE: In equation (2.13), the product (Ftx Z) is limited to 1000 cfs (u 448,000 gpm) or less. (Reference 1, Section 4.3.)
2.4.2 calculation of A: 7 I
I The site-related adult ingestion dose commitment factor, Aj,, is calculated as follows (equation adapted from Reference 1, page 16, by addition of the irrigated garden vegetation pathway) :
Aj7 = 1.14 x 10 5 , Al 'w . ,-Al 'f + . gy er op, j (2.14) where: ,
l l'.14 x 105 = a units conversion factor, determined by:
100 pCi/pci x- 103 mL/L + 8760 h/y. )
l l
i l
2-31 Gen. Rev. 13 i
l 1
l i
FNP-0-M-0(1 U, . the adult drinking water consumption rate applicable to the plant i
r"'Si site (L/y).
)
v 1 D, . the dilution factor from the near field of the discharge structure for the plant site to the potable water intake location.
Aj - the decay constant for radionuclide i (h-I). Values of 1 used in effluent calculations should be based on decay data from a J recognized and current source, such as Reference 15.
t, = the transit time from release to receptor for potable water l
consumption (h) .
i Ug - the adult rate of fish consumption applicable to the plant site (kg/y).
BFj = the bioaccumulation factor for radionuclide i applicable to i
{
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.
[g tg - the transit time from release to receptor for fish consumption (h) . !
l 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 l vegetation, as applicable to the vicinity of the plant site, in (pci/kg) / (pci/L) . Methods for calculation of CFjy are presented below in Section 2.4.3.
DFg7 = the dose conversion factor for radionuclide i for adults, in organ 7 (mrem /pci). For specific values, see Table 2-7.
2.4.3 Calculation of CF h' The concentration factor for radionuclide i in irrigated garden vegetation, CFjy in (L/kg), is calculated as follows:
m 2-32 Gen. Rev. 13
FNP-0-M-011 o
For radionuclides other than tritium (equation adapted from Reference 3, equations A-8 and A-9):
$ql v
e -Alh
- II - # "
CFjy = M*I + (2.15)
Yv AEi pal 0 For tritium (equation adapted from Reference 3, equations A-9 and A-10) :
CFjy = M Ly (2.16)
(
l where: I 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= the average irrigation rate during the growing season (L)/(m2 h) .
1 r= the fraction of irrigation-deposited activity retained on the f] edible portions of leafy garden vegetation.
'% )
Yy =
the areal density (agricultural productivity) of leafy garden vegetation (kg/m 2) ,
l fg = the fraction of the year that garden vegetation is irrigated.
Bjy = the crop to soil concentration factor applicable to radionuclide I i, from Table 2-6 (pci/kg garden vegetation)/(pci/kg soil).
P= the effective surface density of soil (kg/m )2 .
Ag = the decay constant for radionuclide i (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 weathering (h-I) .
V 2-33 Gen. Rev. 13
r; PNP-0-M-011 Ag = the effective removal rate for activity deposited on crop leaves (h-I) calculated as: Ag = Ag + 1,.
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).
Ly- the water content of leafy garden vegetation edible parts (L/kg).
v l
2-34 Gen. Rev. 13
FNP-0-M-011 Table 2-5. Parameters for Calculation of Doses Due to Liquid Effluent Releases
/ Dose Calculation Receptor Locations:
E.ish: Vicinity of plant discharge Drinkino Water: None (Ref. 10)
Irricated Garden Vecetation: Farms at River Mile 26 (Ref . 10)
Numerical Parameters parameter Value Reference
)
Z 5 Ref. 2, Table A-1 U, O L/y
- Ref. 10 D, 1.0 Based on Ref. 1, Section 4.3.1 tw 12 h
- Ref. 3, Sec. A.2 Uf 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 M 0.04 Ref. 16 D\
V I 0.126 L/ (m2 h) Ref. 10, using pump capacity, '
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 P 240 kg/m2 Ref. 3, Table E-15 1, 0.0021 h'I (i.e., half- Ref. 3, Table E-15 life = 14 d) tg 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.) I 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.
2-35 Gen. Rev. 13
FNP-0-M-011 Table 2-6. Element Transfer Factors Freshwater Fish Leafy Garden
. Vegetation Element BFg
+
BV I
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 1.2 E-01 Zn 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 Sr 3.0 E+01 1.7 E-02 Y 2.5 E+01 2.6 E-03 Zr 2.0 E+02 1.7 E-04 Nb 1.0 E+02 9.4 E-03 Mo 1.0 E+02 1.2 E-01 Tc 1.5 E+01 2.5 E-01 Ru 1.0 E+01 5.0 E-02 Os Rh Ag 1.O 2.3 E+01 E+00 1.3 1.5 E+01 E-01 Sb 3.0 E+02 1.1 E-02 Te 2.0 E+03 1.3 E+00 I 2.0 E+01 2.0 E-02 Cs 2.0 E+02 1.0 E-02 Ba 4.0 E+01 5.0 E-03 )
I La 2.5 E+01 2.5 E-07, Ce 2.0 E+02 2.5 E-03 Pr 2.5 E+01 2.5 E-03 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 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, Zn, Mo, Sb, Te, I, Cs, Ba, and Ce; and Referehce 14 for Zr and Nb.
+ Crop to soil concentration factors, in (pci/kg garden vegetation) per (pCi/kg soil) . They are obtained from Reference 3 (Table 1:-1),
except as follows: Reference 2 (Table C-5) for Br and Sb. ,
ry b
2-36 Gen. Rev. 13 l
~
FNP-0-M-011 Table 2-7. Adult Ingestion Dose Factors p
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 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 Cr-51 No Data No Data 2.66L-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 3.91E-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 Ni-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 Br-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 3.21E-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 Dat4 4.94E-05 Sr-90 7.58E-03 No Data 1.66E-03 No Data No Data No Data 2.19E-04 Sr-91 5.67E-06 No Data 2.29E-07 No Data No Data No Data 2.7QE-05 I l
All values are in (mrem /pCi ingested). They are obtained from i Reference 3 (Table E-11) , except as follows: Reference 2 (Table i A-3) for Rh-105, Sb-124, and Sb-125. I j
2-37 Gen. Rev. 13 I l
1 i
FNP-0-M-011 Table 2-7 (contd?. Adult Ingestion Dose Factors V
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 3.51E-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 I 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 l 1
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 O 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 b 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 Te-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 2-38 Gen. Rev. 13 i
FNP-0-M-011 Table 2-7 (centd). Adult Ingestion Dose Factors N
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 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 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 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 1.2BE-10 5.82E-11 1.45E-11 No Data No Data No Data 4.25E-07 Ce-141 9.36E-09 6.33E-09 7.18E-10 No Data 2.94E-09 No Data 2.42E-05 Ce-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 3.01E-11 1.25E-11 1.53E-12 No Data 7.05E-12 No D&ta 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 1.03E-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-12 No Data 3.65E-10 No Data 2.40E-05 2-39 Gen. Rev. 13
m FNP-0-M-011 Table 2-8. Site-Related Ingestion Dose Factors, Aq,
/~~T f )
U Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 2.54E-01 2.54E-01 2.54E-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+03 Ni-63 3.14E+04 2.18E+03 1.05E+03 0.00 0.00 0.00 4.54E+02 l Ni-65 1.72.5-01 2.23E-02 1.02E-02 0.00 'O.00 0.00 5.66E-01 Cu-64 0.00 8.07E+00 3.79E+C0 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 l 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 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 0.00
[Ns 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 1.38E+05 0.00 0.00 0.00 1.62E+04 Sr-S1 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 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 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 (mrem mL) / (h pCi) . They are calculated using i l
equation (2.14) , and data from Table 2-5, Table 2-6, and Table 2-7 When "No Data" is shown for a radionuclide-organ combination in Table 2-7, 79r factors in this table are presented as zero.
2-40 Gen. Rev. 13 1
FNP-0-M-011 Table 2-8 (contd). Site-Related Ingestion Dose Factors, Jyr
(" \
Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Tc-101 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Ru-103 4.65E+00 0.00 2.00E+00 0.00 1.772+01 0.00 5.42E+02 Ru-105 8.71E-03 0.00 3.44E-03 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 2.2SE+05 0.00 2.72E+05 Te-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-331 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 1.8BE-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 O
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 u7E+00 Cs-134 2.99E+04 7.11E+04 5.81E+04 0.00 2.30E+04 7.64E+03 1.24E+03 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 0.00 1.???-12 1.31E-13 7.68E-18 Ba-139 5.64E-05 4.02E-08 1.65E-06 0.00 3.76E-08 2.28E-08 1.00E-04 Ba-140 1.86E+03 2.34E+00 1.22E+02 0.00 7.95E-01 1.34E+00 3.83E+03 !
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.Ou 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 i Ce-141 4.40E+00 2.98E+00 3.38E-01 0.00 1.38E+00 0.00 1.14E+04 Ce-143 4.77E-01 3.53E+02 3.91E-02 0.00 1.55E-01 0.00 1.32E+04 Ce-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 2.11E-03 1.17E-03 0.00 6.60E-03 0.00 4.34E+02 A
d 2-41 Gen. Rev. 13
FNP-0-M-011 2.5 LIQUID EFFLUENT DOSE PROJECTIONS f',
2.5.1 Thirtv-One Day Dose Proiections 4 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.
1 Projected 31-day doses to individuals due to liquid effluents may be determined as follows:
r p ,
Dyp = x 31 + Dra (2.17) where:
D,p = the projected dose to the total body or organ 7, for the next 31 days of liquid releases.
Dyg = the cumulative dose to the total body or organ 7, for liquid
[N
\
releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration, t= the number of whole or partial days elapsed into the current quarter, including the time to the end of the releaae under consideration (even if the release continues into the next quarter).
D ra = the anticipated dose contribution to the total body or any organ 7, due to any planned activities during the next 31-day period, if those activities will result in liquid releases that are in addition to routine liquid effluents. If only routine liquid effluents are anticipated, Dra may be set to zero.
2.5.2 Dose Proiections for Soecific Releases 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. 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.
2-42 Gen. Rev. 13
1 FNP-0-M-011 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS The following symbolic terms are used in the presentation of liquid effluent calculations in the sub-sect *.ons above.
Term Section of Definition Initial Use 1
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
?.Fp= the dilution allocation factor for liquid release pathway p [unitless). 2.3.2.2 Air = the site-related adult ingestion dose commitment factor, for the total body or for any organ 7, due to identified radionuclide 1 [(mrem mL) /(h pCi)] . The values of Aj7 are listed in Table 2-8. 2.4.1
.Oa Bjy =
the crop to soil concentration factor applicable to i
{
radionuclide i, [(pCd /kg garden vegetation)/(pci/kg soil)]. Values are listed in Table 2-6. 2.4.3 BFg = the bioaccumulation factor for radionuclide i for freshwater fish [(pCi/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 l line, prior to dilution and subsequent release
[pci/mL). 2.3.2.1 cp= the calculated effluent radioactivity monitor -
setpoint for liquid release pathway p [yci/mL) . 2.3.2.2 C, = the gross concentration of alpha emitters in the liquid waste as measured in the applicable composite i sample [pci/mL). 2.3.2.2 2-43 Gen. Rev. 13
m FNP-0-M-011 I, gym Section of Definition Initial Use Cn = the Effluent Concentration Limit stated in 10 CFR 20, Appendix B, Table 2, Column 2 [pC1/mL). 2.3.2.1 1
C{ = the concentration of Fe-55 in the liquid waste as measured in the applicable composite sample
{
[pci/mL). 2.3.2.2 l Cg= the concentration of gamma emitter g in the liquid waste as measured by gamma ray spectroscopy performed on the applicable pre-release waste sample
[pci/mL).
2.3.2.2 Cg = the measured concentration of radionuclide i in a sample of liquid effluent [pci/mL). 2.3.2.2 Cjg = the average concentration of radionuclide i in undiluted liquid effluent during time period 1
[pci/mL). 2.4.1 O
U Cir - the measured concentration of radionuclide i in release pathway r that is contributing to radio-activity in the dilution stream [pci/mL). 2.3.2.2 Cs= the concentration of strontium radioisotope s (Sr-8 9 or Sr-90) in the liquid waste as measured in the applicable composite sample [pci/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 CFgy =
the concentration factor for radionuclide i in irrigated garden vegetation [(pci/kg)/(pci/L)) . 2.4.2 l D, = the dilution factor from the near field of the )
discharge structure to the potable water intake location [unitless). 2.4.2
.)
2-44 Gen. Rev. 13
FNP-0-M-011 Section of M Definition Initial Use O D7= the cumulative dose commitment to the total body or to any organ 7, due to radioactivity in liquid effluents released during a given time period l
[ mrem). 2.4.1 Dra = the anticipated dose contribution to the total body i or any organ 7, due to any planned activities.during the next 31-day period [ mrem]. 2.5.1 D,e = the cumulative dose to the total body or organ 7, for t
i liquid releases that have occurred in the elapsed j portion of the current quarter, plus the release under consideration [ mrem]. 2.5.1 D,p - the projected dose to the total body or organ 7, for the next 31 days of liquid releases-[ mrem]. 2.5.1 i i
DFg, = the dose conversion factor for radionuclide i for adults, in organ 7 [ mrem /pci). Values are listed in ,
Table 2-7. 2.4.2 !
ECLj = the liquid Effluent Concentration Limit for radio-1 nuclide i from 10 CFR Part 20, Appendix B, Table 2, !
Column 2 [pCi/mL). 2.3.2.2 f= the effluent flowrate at the locacion of the radio-activit'y n.onitor . [gpm] . 2.3.2.1 f, = the anticipated actual discharge flowrate for a planned release from liquid release pathway p
[gpm]. 2.3.2.2 l fg = the fraction of the year that garden vegetation is
~
irrigated [unitiess). 2.4.3 i f g= the maximum permissible effluent discharge flowrate for release
, pathway p [gpm] . 2.3.2.2 l
2-45 Gen. Rev. 13
FNP-0-M-011 Section of Igrg Definition Initial Use p3 i
9 )- f r= the effluent discharge flowrate of release pathway r
[gpm). 2.3.2.2 fg= the average undiluted liquid waste flowrate actually observed during the period of a liquid release
[gpm). 2.4.1 F= the dilution stream flowrate which can be assured prior to the release point to the UNRESTRICTED AREA !
[gpm). 2.3.2.1 I 1
Fd= the entire assured dilution flowrate for the plant I site during the release period [gpm). 2.3.2.2 !
I Fdp = the dilution flowrate allocated to release pathway p I
[gpm). 2.3.2.2 Fg - the near-field average dilution factor in the
[
\
receiving water of the UNRESTRICTED AREA
\- [unitless). 2.4.1 Fg= the average dilution stream flowrate actually I 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 Ly - 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 [unitleus). 2.4.3 P= the effective surface density of soil [kg/m1. 2 2.4.3 O
V 2-46 Gen. Rev. 13
- FNP-0-M-011 Section of Igrm Definition G IDitial Use
/ i r= the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation. 2.4.3 RDF = the required dilution factor: 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
[unitless). 2.3.2.2 )
l RDF 7= the RDF for a liquid release due only to its concen-tration of gamma-emitting radionuclides [unitiess). 2.3.2.2 RDFny = the RDF for a liquid 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 i
,_ l statistical fluctuations and errors of measurement (G [unitless). 2.3.2.2 t= the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration. 2.5.1 tb= the period of long-term buildup of activity in soil
[h) . 2.4.3 te= the period of leafy garden vegetation exposure during the growing season [h) . 2.4.3 tg - the transit time from release to receptor for fish consumption [h). 2.4.2 th= the time between harvest of garden vegetation and human consumption [h). 2.4.3 t, = the transit time from release to receptor for potable
[ \ water consumption [h) . 2.4.2
\
2-47 Gen. Rev. 13
FNP-0-M-011 Section of h Definition Initial Use
(~
t
\
TF = the tolerance factor selected to allow flexibility in the establishment of a practical monitor aetpoint which could accommodate effluent releases at concentrations higher than the ECL values stated in 10 CFR 20, Appendix B, Table 2, Column 2 lunitless) ;
the tolerance factor must not exceed a value of 10. 3.3.2.2 Uf= the adult rate of fish consumption [kg/y). 2.4.2 Uy= the adult consumption rate for irrigated garden vegetation (kg/y). 2.4.2 U, = the adult drinking water consumption rate applicable to the plant site [L/y). 2.4.2 Yy- the areal density (agricultural productivity) of leafy garden vegetation [kg/m 2). 2.4.3 Z= the applicable dilution factor for the receiving d water body, in the near field of the discharge structure, during the period of radioactivity release
[unitless). 2.4.1 Atj = the length of time period 1, over which Cgt and Fg are averaged for liquid releases [h) . 2.4.1 lEl - the effective removal rate for activity deposited on crop leaves [h-I) . 2.4.3 Ig = the decay constant for radionuclide i [h-I). 2.4.2 1, = the rate constant for removal of activity from plant leaves by weathering [h-I) . - 2.4.3 b('s 2-46 Gen. Rev. 13 1
FT l FNP-0-M-011 CHAPTER 3 GASEOUS EFFLUENTS 3.1 LIMITS OF OPERATION The following Limits of operation implement requirements established by Technical Specifications Section 6.0. Terms printed in all capital letters are defined in Chapter 10.
3.1.1 Gaseous Effluent Monitorine Instrumentation Centml 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 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with Section 3.3.
3.1.1.1 Applicability These limits apply as shown in Table 3-1.
3.1.1.2 Actions With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above control, immediately 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.
This control does not affect shutdown requirements or MODE changes. ;
3.1.1.3 Surveillance Requirements l
Each radioactive gaseous effluent monitoring instrumentation channel sfiall be l j
demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 3-2.
O 3-1 Gen. Rev. 13
r-FNP-0-M-011 3.1.1.4 Basis The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of geseous effluents. The 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 occur 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 l
s
)
O(N.
3-2 Gen. Rev. 13
FNP-0-M-011 Table 3-1. Radioactive Gaseous Effluent Monitoring Instrumentation O'
OPERABILITY RequirementsD Minimum Channels Instrument OPERABLE Applicability ACTION
- 1. Steam Jet Air Ejector Noble Gas 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 A" all times 37a
- 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. GASEOUS RADWASTE TREATMENT SYSTEM Noble Gas Activity Monitor (RE-14), with i
Alarm and Automatic Termination of Release 1 At all times 35
- a. For continuous releases,
- b. All requirements in this table apply to each unit.
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3-3 Gen. Rev. 13 l
FNP-0-M-011 Table 3-1 (contd). Notation for Table 3 ACTION Statements ACTION 35 - With the number of channels OPERABLE less than required by the I Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to initiating the releases
- a. At least two independent samples of the tank's contents are analyzed, and j l
- b. At least two technically qualified members of the Facility Staff e
independently verify the discharge line valving, and (1) verify the manual portion of the computer input for the 1 J
release rate calevlations 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 - W. th 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 rwr 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Normal building ventilation may continue provided the flowrate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
V 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Normal building ventilation may continue provided grab samples of this pathway are taken once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and analyzed for gross activity once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continueusly collected with auxiliary sampling equipment as required in Table 3-3.
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O 3-4 Gen. Rev. 13
e FNP-0-M-011 Table 3-2. Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Surveillance Requirementsd Instrument CHANNEL CHANNEL CHANNEL SOURCE CALIBRA- FUNCTIONAL CHECK CHECK TION TEST MODESC
- 1. Steam Jet Air Ejector Noble Gas Activity Monitor (RE-15) D M RD Qa(2) 1,2,3,4
- 2. Plant Vent Stack
- a. Noble Gas Activity Monitor RE-14 D M RD QaO,2) gyy RE-22 p M RD Qa(2) All
- b. Iodine Sampler W NA NA NA All
- c. Particulate Sampler W NA NA NA All
- d. Flowrate Monitor D NA R u All
- a. In addition to the basic functions of a CHANNEL FUNCTIONAL TEST (Section 10.2)
( (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic 1 e
D isolation of this pathway and control room annunciation occur if any of the following conditions exists:
(a) Instrument indicates measured levels above the alarm / trip setpoint; (b) Loss of control power; or (c) Loss of instrument power.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room annunciation occurs if any of the following conditions exists:
(a) Instrument indicates a downscale failure; or (b) 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 and Technology, or using standards that have been obtained from suppliers that 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.
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U) 3-5 Gen. Rev. 13
FNP-0-M-011 3.1.2 Gaseous Effluent Dose Rate Control In accordance with Technical Specifications 6.8.3.e (iii) and 6. 0,3.e (vii) , the 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:
- a. For noble gases: Less than or equal to a dose rate of 500 mrem /y to the total body and less than or equal to a dose rate of 3000 mrem /y to the skin, and
- b. For 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 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 limits, in accordance with the methods and procedures in Section 3.4.1, by obtaining representative samples and performing analyses in accordance with 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 maintained within the limits that historically have provided reasonable assurance 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 the SITE BOUNDARY, exceeding the limits specified in Appendix I of 10 CFR Part 50, while allowing operational flexibility for effluent releases. For MEMBERS 3-6 Gen. Rev. 13
FNP-0-M-011 OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in O the atmospheric diffusion factor.above that for the SITE BOUNDARY.
-The dose rate limit for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than a days specifically applies to dose rates to a child via the inhalation pathway.
This control applies to the release of gaseous affluents from all reactors at the site.
O 3-7 Gen. Rev. 13
FNP-0-M-011 Table 3-3. Radioactive Gaseous Waste Sampling and Analysis Program j
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Sampling and Analysis Requirementa a,b l
MINIMUM l DETECTABLE 1 Gaseous Minimum Type of CONCENTRA-Release Sampling Analysis Activity TION (MDC)
Type FREQUENCY FREQUENCY Analysis ( Ci/mL)
P PRINCIPAL GAMMA 1 E-4 Decay Ta Each Tank Grab Sample Each Tank c PRINCIPAL GAMMA 1 E-4 Containment Each Purge 9*
Grab Sample Each Purge Condenser PRINCIPAL GAMMA 1 E-4 Steam Jet EMITTERS Air Mcd.f Ejector, Grab Sample H-3 1 E-6 Plant Vent Stack CONTINUOUS 8 We I-131 1 E-12 Charcoal or Charcoal or Silver Silver Zeolite I-133 1 E-10 O- Zeolite Sample WC PRINCIPAL GAMMA 1 E-11 CONTINUOUSS Particulate EMITTERS Sample Plant Vent M Gross Alpha Stack, 1 E-11 CONTINUOUSg COMPOSITE Containment p, Purge Sample Q Sr-89, Sr-90 1 E-11 CONTINDOUSg COMPOSITE Particulate Sample CONTINUOUSg N Gas Noble Gases 1 E-6 (Gross Beta and Gamma)
O 3-8 Gen. Rev. 13
FNP-0-M-011 Table 3-3 (contd). Notation for Table 3-3 O
V a. All requirements in this table apply to each unit. Deviation from the MDC requirements of this table shall be reported in accordance with Section 7.2. Deviation from the composite sampling requirements of this table shall be reported in accordance with Section 7.2.
- b. Terms printed in all capital letters are defined in Chapter 10.
- c. Analyses shall also be performed following shutdown from greater than or equal to 15% RATED THERMAL POWER, startup to great ar than or equal to 15%
RATED THERMAL POWER, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a one-hour period,
- d. Tritium grab samples shall be taken from the plant vent stack at least '
once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.
- e. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or after removal from sampler).
Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding MDC may be increased by a factor of 10.
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% - f. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.
- g. 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 i made in accordance with controls specified in Sections 3.1.2, 3.1.3, and 3.1.4.
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3-9 Gen. Rev. 13
I FNP-0-M-011
'3.1.3 Gammous 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 followings
- 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
- 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 Applicability i
This limit applies at all times.
3.1.3.2 Actions 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 J Commission within 30 days, pursuant to Technical Specification 6.9.2, a Special e Report which identifies the cause(s) for exceeding the limit (s); defines the corrective actions that have been taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases of radioactive noble gases in gaseous effluents will be in compliance with the limits of Section 3.1.3.
This control does not affect shutdown requirements or MODE changes.
3.1.3.3 Surveillance Requirements
. Cumulative air dose contributions from noble gas radionuclides 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 3.4.2 at least once per 31 days.
3.1.3.4 Basis -
l This control is provided to implement the requirements of Sections II.B, 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. The ACTION statements in Section 3.1.3.2
)
provide'the required operating flexibility and at the same time implement the 3-10 Gen. Rev. 13 l l
FNP-0-M-011 !
guides set-forth in Section IV.A of Appendix I, assuring that the releases of m radioactive material in gaseous effluents 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 conforrr.ance with the guides of Appendix I be shown by calculational I 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 BOUNDARY are based upon the historical annual average atmospheric conditions.
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O i 3-11 Gen. Rev. 13
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FNP-0-M-011 3.1.4 control on (1==eous Effinant' Dose to a " W r of the Public In accordance with Technical. Specifications 6.8.3.e (v) and 6.8.3.e (ix) , the dose to a MEMBER 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 ereas at and beyond the SITE BOUNDARY.(see l Figure 10-1) shall be limited to the following:
j a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ, !
and
- b. During any calendar years Less than or equal to 15 mrem to any organ.
3.1.4.1 Applicability This limit applies at all times.
3.1.4.2 Actions-l 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, prepara and submit to the Nuclear Regulatory Commission within 30 days, pursuant to Technical Specification 6.9.2, 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, s6all be determined in accordance with section 3.4.3 at least once per 31 days.
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FNP-0-M-011 3.1.4.4 Basis This control is provided to implement the requirements of Section II.C, III. A and 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 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 "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 Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to i be substantially underestimated. 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),
snd Regulatory Guide 1.111 (Reference 5). These equations provide for determining the actual doses based upon the historical annual average atmospheric conditions. The release specifications for radioiodines, 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 EOUNDARY. The pathways which were examined in the development of these '
calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy garden vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man, i
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's 3-13 Gen. Rev. 13
FNP-0-M-011 3.1.5 Gameous Radwaste Treatment System Control In accordance with Technical Specification 6.8.3.e (vi) , the GASBOUS RADWASTE TREAMENT SYSTEM and the VENTILATION EXHAUST TREAMENT 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 the projected air doses due to gaseous effluent releases, from each reactor unit,
. to areas at and brayond the SITE BOUNDARY (see Figure 10 1) would exceed 0.2 mrad for gamma radiation or 0.4 mrad for beta radiation in 31 days. The appropriate 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 mrem to any organ of a MEMBER OF THE PUBLIC in 31 days.
3.1.5.1 Applicability These limits apply at all times.
3.1.5.2 Actions With gaseous waste being discharged without treatment and in excess of the limits 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 information:
- a. Identification of the inoperable equipment or subsystem and the reason for inoperability,
- b. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- c. Summary description of action (s) taken to prevent a recurrence.
This control does not affect shutdown requirements or MODE changes.
O 3-14 Gen. Rev. 13
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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 GASEOUS RADWASTE TREATMENT SYSTEM or the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.
The GASEOUS RADWASTE TREAmENT SYSTEM and the VENTILATION EXHAUST TREATMENT l
SYSTEM shall be demonstrated OPERABLE: l
- a. by meeting the controls of Sections 3.1.2, and either 3.1.3 (for the GASEOUS RADWASTE TREAMENT SYSTEM) or 3.1.4 (for the VENTILATION EXHAUST TREAMENT SYSTEM), or
- b. by operating the GASE5US RADWASTE TREAWENT SYSTEM equipment and the VENTILATION EXHAUST TREAMENT 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 effluents require treatment prior to release to the environment.
The requirement thest the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."
This 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 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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3-15 Gen. Rev. 13
r-FNP-0-M-011 3.1.6 MAJOR CHANNE to the GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEM and the VENTILATION WYuAUST TREATMENT SYSTEM O Licensee-initiated MAJOR CHANGES to the GASEOUS RADIOACTIVE WASTE TREA*INENT 1
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-SYSTEM or the VENTILATION EXHAUST TREA7NENT SYSTEM:
- a. Shall be reported to the Nuclear Regulatory Conunission 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 I
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l O l 3-16 Gen. Rev. 13 !
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FNP-0-M-011 3.2 GASEOUS RADWASTE TREATMENT SYSTEM j At the Farley Nuclear Plant, there are six designated points where radioactivity may be released to the atmosphere in gaseous discharges: the Unit 1 and 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 perforned 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 for the noble gas monitors on the two Plant Vents and the two Turbine Building vents (steam jet air ejectors) . Waste 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 not 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 limits specified in Section 3.1.3 and Section 3.1.4 are not exceeded.
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3-17 Gen. Rev. 13
FNP-0-M-011 O' Waste cas Decay Tanks N E E
- E E H H :: S X 7 8 1 2 3 4 5 6
<r ir 3r 1r i r sr $r r n n n n n n a n 1P 1P 1r dL db db
]l Waste Gas RE-13 MONITOR Compressor O
E PLANT VENT STACK C,7 -l ILRT - RE-22 MONITORS
- VENT CONTAINMENT AUX.
PURGE " SLDG
, CONTAIN- :
EXHAUST MENT PLENUM 4
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RE-2 MONITOR AUXILIARY ,
BUILDING f i
I RE-15 MONITOR j TURBINE !
WR8tNE 5- BUILDING BUILDING VENT
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j (STEAM JET l AIR EJECTOR) l I
( Figure 3-1. Schematic Diagram of Rout:.ne Release Sources and Release Points l
\ (Typical of Both Units) I 3-1B Gen. Rev. 13 !
PNP-0-M-011 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS 3.3.1 General Provisions Recardino Noble Gas Monitor Setooints 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 setpoir.ts. 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 to reaching the high slarm setpoint.
If no release is planned for a given pathway, or if there is no detectable activity in the gaseous stream being evaluated for release, the setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should a significant inadvertent release occur.
Section 3.1.1 establishes the requirements for gaseous effluent monitoring
! instrumentation, and Section 3.2 describes the VENTILATION EXHAUST TREATMENT SYSTEM and the GASEOUS RADWASTE TREATMENT SYSTEM. From those sections, it can be seen that certain monitors are located on final release pathways, that is, 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 '
other streams prior to passing a final monitor and being discharged; the setpoint 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.
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y FNP-0-M-011 Table 3-4. Applicability of Gaseous Monitor Setpoint Methodologies
/'N k_,) Final Release Pathways with no Monitored Source Streams Release Elevation: Ground-level Unit 1 or Unit 2 Turbine Buildino Vent Release Type: CONTINUOUS Monitor: 1RE-15 / 2RE-15 Setpoint Method: Section 3.3.2 Maximum Flowrate: 1060 cfm (5.00 E+05 mL/s)
Unit 1 or Unit 2 ILRT Vent Release Type: BATCH Monitor: None Setpoint Method None Maximum Flowrate: Relrape-dependent Final Release Pathways with One or More Monitored Source Streams Release Elevation: Mixed-Mode Unit 1 or Unit 2 Plant Vent Stack Re1eaae Type: CONTINUOUS Monitor: 1RE-14 / 2RE-14, and 1RE-22 / 2RE-22 Setpoint Method: Section 3.3.2 Maximum Flowrate: 150,000 cfm (7.08 E+07 mL/s)
Source Stream; Unit 1 or Unit 2 Containment Purce
('~T
'- ) Release Type: CONTINUOUS Monitor: 1RE-24 / 2RE-24 Setpoint Method: Section 3.3.3 is optional. See Section 3.3.5.
Maximum Flowrates Release-dependent Source Streamt Unit 1 or Unit 2 Waste Gas Decav Tanks i Release Type: BATCH Monitor: None i
Setpoint 'Aethod: None j Maximum Flowrate: Release-dependent j (i?6)tb Values for Use in Setpoint Calculations Ground-Level Releases: 4. 87 x 10 5 s/m3 (S Sector)
Mixed-Mode Releasgg 1.08 x 10 4 s/m3 [SSE Sector) f (3
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3-20 Gen. Rev. 13
FNP-0-M-011 3.3.2 Setroint for the Final Noble Gas Monitor on Each Release Pathway Overview of Method 3.3.2.1 I
Gaseous ef fluent 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 equation:
P c = the lesser of < ( **}
AG
- SF X Rg where c- the setpoint, in pCi/mL, of the radioactivity monitor measuring the concentration of radioactivity in the offluent line prior to release. The setpoint represents a concentration which, if exceeded, could result in dose rates exceeding the limits of Section 3.1.2.a at or beyond the SITE BOUNDARY.
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 for statistical fluctuations and errors of measurement.
X= the noble gas concentration for the release under consideration.
Rt= the ratio of the dose rate limit fcr the total body, 500 mrem /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 mrem /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
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the setpoint. However, in order to apply the methodology presented in the equation to a mixture of noble gas radionuclides, radionuclide-specific concentratiens and dose factors must be taken into account under conditions of i
maximum flowrate for the release point and annual average meteorology.
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FNP-0-M-011 The basic setpoint method presented below is applicable to the radioactivity monitor nearest the point of release for the release pathway. For monitors
() measuring the radioactivity in source streams that merge with other streams prior to subsegaent monitoring and release, the modifications precented in Section 3.3.3 must be applied.
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 Xjy.
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Sten 2: Determinet R , the ratio of the dose rate limit for the total body, 500 mrem /y, to the total body dose rate due to noble gases detected I in the release under consideration, as follows:
i
, 500 (TTC)v6 b (Ki
- O lv] (3.2) i i !
O where:
500 = the dose rate limit for the total body, 500 mrem /y.
(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 te each release pathway; release elevation determines the appropriate value of (176)vb-K =
the total-body dose factor due to gamma emissions from noble gas radionuclide i, in (mrem /y)/(pci/m 3), from Table 3-5.
Ojy =
the release rate of noble gas radionuclide i from the telease pathway under consideration, in pC1/s, calculated as the product of X y and fav, where:
Xy - the concentration of noble gas radionuclide i for the particular release, in pCi/mL.
3-22 Gen kev. 13
r i
FNP-0-M-011 f,y = the maximum anticipated flowrate for release pathway v .
during the period of the release under consideration, in mL/s.
1 Sten 3: Determine kR , the ratio of the dose rate limit for the skin, 3000 mrem /y, to the skin dose rate due to noble gases detected in the release under consideration, as follows:
1 Rg = 000
("f7D)vb b (ILI + l lM)l
- O y) l (3*3)
I where:
3000 = the dose rate limit for the skin, 3000 mrsm/y.
I Lj = the skin dose factor due to beta emissions from noble gas radio-nuclide i, in (mrem /y)/(pci/m 3), from Table 3-5.
M =
the air dose factor due to gamma emissions from noble gas radio-nuclide i, in (mrad /y) / (pci/m )3 , from Table 3-5.
1.1 = the factor to convert air dose in mrad to skin dose in mrem.
All other terms were defined previously.
Sten 4: Determine the maximum noble gas radioactivity monitor setpoint con-centration.
4 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 be exceeded. Because the radioactivity monitor responds primarily to radiation from noble gas radionuclides, the monitor setpoint eny is based on the concentration of all noble gases in the waste stream, as follows:
3-23 Gen. Rev. 13
C FNP-0-M-011
-1 ' AGy = SF = Xiy Rg, Q cnv
- the lesser of , (3.4)
AGy SF
- Xiy
- Rk where:
cny - the calculated setpoint, in pCi/mL, for the noble gas monitor serving gaseous release pathway v.
AGy- the administrative allocation. factor for gaseous release pathway v, applied no divide the - release limit among all the gaseous release pathveys at the site. The allocation' factor may be assigned any value between 0 and 1, under the condition that the sum of the allo :ation factors for all simultaneously active final l release pathways 3t the entire plant site does not exceed 1.
I Alternative methods for asterndnation of AGy - are presented in Section 3.3.4.
SF m ' the' safety factor selected to compansate for statistical fluctuations and errors of measurement. ne 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 ba developed if desired.
i Xjy = -the measured concustration of noble gas radionealide i in gaseous stream v, as defined in Step 1, in pCi/mL.
1 The - values of tR and Rk to be used in the calculation are those which were determined in Steps 2 and.3 above.
l 4
sten 5: Determine whether the release is permissible, as follows:
If env = EXjy, . the release is permissible. However, if e nv is within about i
10 percent of E Xjy, it may be impractical to use this value of c av- !
This situation indicates that measured concentrations are l approaching values which would cause the limits of Section 3.1.2.a I to be exceeded. Therefore, steps should be taken to reduce l
contributing source terms of gaseous radioactive material, or to I
3-24 Gen. Rev. 13
FNP-0-M-011 adjust the allocation of the limits among the active release 3 points. The setpoint calculations (steps 1-4) must then be t
b) repeated with parameters that reflect the modified conditions.
If env c r Xjy, the release may not be made as planned.
i Consider the alternatives discussed in the paragraph above, and calculate a now setpoint based on the results of the actions taken.
3.3.2.3 Use of the Calculated Setpoint 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:
l c,y = (cnv ' Ev ) + 8 v (3.5) where:
cnv = the monitor setpoint as a count rate.
/3 Ey- the monitor calibration factor, in count rate / ( Ci/mL) , Monitor calibration data for conversion between count rate and concentration may include operational data obtained from determining the monitor response to effluent stream concentrations measured by sample analysis.
By = the monitor background count rate. In all cases, monitor background must be controlled so that the moni' r 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 shutdown and power conditions) , and internal background due to contamination of the monitsr's sample chamber.
The count rate units for cny, Ey , and By in equation (3. 5) must be the same (epm or cps).
3.3.3 SetDoints for Noble Gas Monitors on Effluent Source Streams r8 k
v
/
3-25 Gen. Rev. 13
V FNP-0-M-011 Table '3-4 lists certain gaseous release 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 effluent 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 Step 1: : Determine the concentration X is 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].
Steo 2: Determine rg, the ratio of the dose rate limit for the total body, 500 mrem /y, to the total body dose rate due to noble gases detected in the source stream under consideration. Use the Xis values and 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. g
)
! The SITE BOUNDARY relative dispersion value used in Steps 2 and 3 for the source stream is the same as the _(176)vb that applies to the respective merged stream. This is because the (176) value is determined by the meteorology of the plant site and the physical attributes of the release point, and,is unaffected by whether or not a given source stream is operating, sten 3: Determine rka.the ratio of the dose rate limit for the skin, 3000 mrem /y, to the skin dose rate due to noble gases detected in the source stream under consideration. Use the Xis values and the maximum anticipated source stream flow rate af , in equation (3.3) to determine the skin dose rate for the source stream, substituting rk for Rk* ~
steo 4: Determine the maximum noble gas radioactivity nonitor setpoint con-centration, as follows:
s 3-26 Gen. Rev. 13
N FNP-0-M-011 cm= the lesser of (3.6)
AGs SF - Xis
- Tk l
l l
where: i l
i i
cm= the calculated setpoint (in Ci/mL) for the noble gas monitor serving gaseous source stream 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 point must not exceed the release point's allocation factor AGy.
Xis = the measured concentration of noble gas radionuclide i in gaseous source stream s, as defined in Step 1, in pC1/mL.
The values of rt 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, sten 5: Determine whether the release is permissible, as follows:
If c EX m*i is, the release is permissible. However, if cm is within about 10 percent of IXis, it may be impractical to use this value of c m-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 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 cm < r Xg,, the release may not be made as planned. Consider the i
alternatives discussed in the paragraph above, and calculate a new j setpoint based on the results of the actions taken. l 3-27 Gen. Rev. 13
FNP-0-M-011 l
3.3.3.2 Effect on the Setpoint of the Monitor on the Merged '
-~3 Stream
(
%j
)
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 )
((fav) old) , and the maximum flowrate of the source stream 1 i
being considered fcc release (fu).
(fav)new (fav) old + fas (3.7) l i
Modification 2: The new concentration of noble gas radionuclide i in the j merged stream includes both the contribution of the !
merged stream without the source stream, and the source j
, stream being considered for release. J
( )'
L/ (fav)old ' (Xiv)old
- fas ' Xis (Xjy) new (3*8)
(fav)new 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 sum of the allocation factors for all simultaneously-active final release paths at the plant site may not exceed 1. .
- 2. The sum of the allocation factors for all simultaneously-active source streams merging into a given final release pathway may not exceed the allocation factor of that final release pathway.
l }
%J 3-28 Gen. Rev. 13
~
[ ;
l l
PNP-0-M-011 Any of the following three methods may be used to assign the allocation factors
-~ to the active gaseous release pathways:
(ms/ '
1.
For ease of implementation, AG y may be equal for all release pathways:
AGy = 3-N (3.9) where: l i
N= the number of simultaneously Ective gaseous release pathways.
1
- 2. AG y for a given release pathway may be selected based on an estimate of the portion of the total SITE BOUNDARY dose rate (from all simultaneous I releases) that is contributed by the release pathway. During periods when a given building or release pathway is not subject to gaseous radioactive I releases, it may be assigned an allocation factor of zero.
- 3. AG y for a given release pathway may be selected based on a calculation of the portion of the total SITE BOUNDARY dose rate that is contributed by the release pathway, as follows: '
i O (YlD)vb E{Kg Qjy}
AG y = 5 N (3.10)
E ,(YlD)tb E(K iO tr),
rol I where:
(176)vb =
the annual average SITE BOUNDARY relative concentration applicable to the gaseous release pathway v for which the allocation factor is being determined, in s/m 3.
Kj = the total-body dose factor due to gamma emissions from noble gas radionuclide i, in (mrem /y)/(pci/m 3), from Table 3-5.
Ojy - the release rate of noble gas radionuclide i from release pathway v,
in pCi/s, calculated as the product of Xjy and f,y, where:
(%
\~-).
3-29 Gen. Rev. 13
FNP-0-M-011
]
\-
Kjy - the concentration of noble gas radionuclide i applicable l to the gaseous release pathway v for which the allocation s factor is being determined, in pCi/mL. 1 f,y -
the discharge flowrate applicable to gaseous release l
pathway v for which the allocation factor is being i determined, in mL/s.
I
\
(x76)rb =
the annual average SITE BOUNDARY relative concentration !
applicable to active gaseous release pathway r, in s/m3 ,
Qr=
i the release rate of noble gas radionuclide i applicable to active release pathway r, in pCi/s, calculated as the product of Xr and f g, where X r- the concentration of noble gas radionuclide i applicable to active gaseous release pathway r, in pCi/mL.
f ,-- the discharge flowrate applicable to active gaseous release pathway r, in mL/s.
O N= the number of simultaneously active gaseous release pathways (including pathway v that is of interest).
NOTE: Although equations (3. 9) and (3.10) are written to illustrate the assignment of the allocation factors for final release pathways, they may also be used to assign allocation factors to the source streams that merge into a given final release pathway.
d 3-30 Gen. Rev. 13 l
l
y FNP-0-M-011 I
l 3.3.5 Setooints for Noble Gas Monitors with Soecial Recuirements l
The Parley 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 l process monitors, not effluent monitors. However, as a matter of information,
- i. the following may be noted regarding their setpoints:
o For 1/2 RE-013, 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-013, 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) setpoint 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 ODCM establish setpoint l calculation methods for particulate and iodine monitors.
?\
3-31 Gen. Rev. 13 l
l
FNP-0-M-011 3.4 ~ GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3~.4.1 Dose Rates at an'9 Beyond the site Boundarv Because the dose rate limits for areas at and beyond the SITE specified in Section 3.1.2 are sf ee limits applicable at any instant in time, the summations extend over all simultaneously active gaseous final rolease pathways at the plant site. Table 3-4 identifies the gaseous final release pathways at the plant site, and indicates the (x~/D)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 nObie gas radionuclides in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows:
For total body dose rates:
DRf =
{v (M)g [ {K Of fy] (3,11)
, i For skin dose rates:
DRg =
{v (M)g [ { (I,f + 1.1Nj) O fy] (3,12)
I i l
where:
DRg= the total body dose rate at the time of the release, in mrem /y.
DRk= the skin dose rate at the time of the .elease, in mrem /y.
Ojy = the release rate of noble gas radionuclide i, in pCi/s, equal to the product of fty and Xgy, where fty = the actual average flowrate for release pathway v during the period of the release, in mL/s. -
All other terms were defined previously.
k 3 32 Gen. Rev. 13 I
FNP-0-M-011 l 3.4.1.2 Dose Rates Due to Iodine-131, Iodine-133, Tritium, and
,, Radionuclides in Particulate Form with Half-Lives Greater (q) 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, and 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 ={v (7/D)yg { Pjo Ofp (3.13) i '
where:
DRo= the dose rate to organ o at the time of the release, in mrem /y.
Pjo = the site-specific dose factor for radionuclide i and organ o, in 3
(mrem /y) / (pci/m ) . Since the dose rate limits specified in Section 3.1.2.b apply only to the child age group exposed to the inhalation pathway, the values of P omay be obtained from Table 3-9, "R I aid O#
(~} Inhalation Pathway, Child Age Group."
U Qjy- the release rate of radionuclide i from gaseous release pathway v, in pCi/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 included in this calculation.
All other terms were defined previously.
3.4.2 Noble Gas Air Dose at or Beyond Site Boundary For the purpose of implementing the controls of Section 3.1.3, air doses in areas 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): -
Dg = 3.17 x 10-0 { (Y/D)vb b ,E, iv (3'14) v i V
3-33 Gen. Rev. 13 l
PNP-0-M-011 D
7 = 3.17 x 10'I {< (Y/D)g { gj . Ojy >
(3.15) v i where:
3.17 x 10-8 = a units conversion factor: 1 y/ (3.15 x 10 7 s).
Dg - the air dose due to beta emissions from noble gas radionuclides, in mrad.
Dy= .the air dose due to gamma emissions from noble gas radionuclider, in mrad. l 1
Ng = .the air dose factor due to beta emissions from noble gas radio-nuclide 1, in (mrad /y) /(pci/m 3), from Table 3-5.
Mj = the air dose factor due to gamma emissions from noble gas radio-nuclide 1, in (mrad /y)/(pci/m 3), from Table 3-5.
6y-i the cumulative-release of noble gas radionuclide i from release pathway v, in pCi, during the period of interest.
All other terms were defined previously.
i
\
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 l 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.
l The gaseous final release pathways at the plant site, and the (X76)vb for each, are identified in Table 3-4.
l (e-V 3-34 Gen. Rev. 13
FNP-0-M-011 Table 3-5. Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases d
y - Body (K) $ - Skin (L) y - Air (M) $ - Air (N)
(mrem /y) per (mrem /y) per (mrad /y) per (mrad /y) per (pCi/m3) (pCi/m 3) (pCi/m 3)
(pCi/m 3)
Kr-83m 7.56 E-02 0.00 E+00 1.93 E+01 2.88 E+02 Kr-85m 1.17 E+03 1.46 E+03 1.23 E+03 1.97 E+03 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 6.17 E+03 1.03 E+04 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 Kr-90 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03 Xe-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 values in this table were obtained from Reference 3 (Table B-1) ,
with unita converted.
O 3-35 Gen. Rev. 13
r.
FNP-0-M-011 Table 3-6. Dose Factors for Exposure to Direct Radiation from Noble Gases in an Elevated Finite Plume l l l
The contents of this table are not applicable to the Farley Nuclear Plant.
l I
O 3-36 Gen. Rev. 13
I FNP-O-M-0Il i
3.4.3 Dose to a Member of the Public at or Beyond Site Boundarv n
.()
(
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 MEMBER 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.
Doses 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):
Dja =
3.17 x 10~E { {Rafpj { Wjyp 0,fy (3.16) p i v 1
i i
where:
Dja = the dose to organ j of an individual in age group a, due to gaseous b
releases of I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greateL than 8 days, in mrem.
3.17 x 10-8 = a units conversion factor: 1 y/ (3.15 x 107 ,), l Rgpj =
the site-specific dose factor for age group a, radionuclide i, 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 Rg 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 ussd for calculating Rgp j for the plant site is presented in Chapter 9.
That information may be used for recalculating the Rg p j values if j
the underlying parameters change, or for calculating Rg p; values for special radionuclides and age groups when performing the
(
assessments discussed in Section 3.4.4 below.
t
(
v) 3-37 Gen. Rev. 13
FNP-0-M-011 Wjyp= the annual average relative dispersion or deposition at the
(~^g location of the controlling receptor, for release pathway v, as t b
\_,/ appropriate to exposure pathway p and radionuclide i.
For all tritium pathways, and for the inhalation of any radio-nuclide: W yp is (17D) yp, the annual average relative dispersion factor for release pathway v, at the location of the controlling receptor (s/m )3 . For the ground-plane exposure pathway, and for all ingestion-related pathways for radionuclides other than tritium: Wy ;p is (676)yp, the annual average relative depositien factor for release pathway v, at the location of the controlling receptor (m-2), values of (176)yp and (576)yp for use in calculating the dose to the currently-defined controlling receptor are included in Table 3-7.
61y- the cumulative release of radionuclide i from release pathway v, during the period of interest (pci) . For the purpose of implementing the controls of section 3.1.4, only I-131, I-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation. In any f3 dose assessment using the methods of this sub-section, only radio-(w)I nuclides detectable above background in their respective samples should be included in the calculation.
Because the member of the public dose limit is on a per-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 between the two units in any reasonable manner, provided that all activity released from the plant site is apportioned to one unit or the other.
The gaseous final release pathways at the plant site, and the release elevation for each, are identified in Table 3-4.
1
~
/
! s\
\m /
3-38 Gen. Rev. 13
FNP-0-M-011 Table 3-7. Attributes of the controlling Receptor i
\ The locations of members of the public in the vicinity :,f the plant site, and the exposure pathways associated with those locations, ar> determir.ed 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 Arte Grouo Child Exoosure Pathways: Ground Plane Inhalation Garden Vegetation Grass / Cow / Meat Discersion Factors (X761 9:
Ground-Level discharge points: 8.74 x 10 4 s/m3 Mixed-Mode discharge points: 8.03 x 10*7 s/m3 Deoosition Factors (D76)yp t D Ground-Level discharge points: 2.64 x 10~0 m-2 Mixed-Mode discharge points: 1.05 x 10-8 m-2 This location represents thu residence with the highest annual average X/Q and D/0 factors in the vicinity of the FNP. The referenced Land Use Census identified no locations where animals are maintained for milk within 5 miles of the plant sites thus, 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.
3-39 Gen. Rev. 13
FNP-0-M-011 3.4.4 Dose Calculations to Suncort other Recuirements y Case 1: Under Technical Specification 6.6.1, a radiological impact assess-ment may be required to support evaluation of a reportable event.
Dose calculations may be performed using the equations in Se: tion 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 (Rgj) p for the receptor of interest. Methods for calculating (x/Q) and (D/Q) from meteorological data are presented in Chapter 8.
The values of Rgpj presented in Table 3-8 through Table 3-11 are applicable only to the currently-defined controlling receptor, so that when dose i calculations must be performed for a different receptor, Rgp ; values l applicable to that receptor must first be calculated. Methods and parameters for calculating Rg pj for radionuclides and age groups other than those required in Section 3.4.3 are presented in Chapter 9. When calculating Rgp j for evaluation of an event, pathway and usage factors specific to the receptor involved in the event may be used in place of the values in Chapter 9, if the specific values are known.
O Case 2: A dose calculation is required to evaluate the results of the Land i
Use Census, under the provisions of Section 4.1.2.
)
In the event that the Land Use census re,eals 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 deposition values ((X76) and (D76)] for the locations of interest, and using the appropriate pathway dose factors (Rgj) p for the receptors of interest.
Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8. The values of Rgp ; 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, Rgp j values app?.icable to that receptor must first be calculated.
Methods and parameters for calculating Rgp j for radionuclides and age
)
3-40 Gen. Rev. 13
PNP-0-M-011 groups other than those required in Section 3.4.3 are presented in Chapter O case 3: Under Section 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.
Dose calculations shall be performed using the equations in Section 3.4.3, using the dispersion and deposition parameters defined in Table 3-7 for the controlling receptor, but substituting the appropriate pathway' dose factors (Rgj) p for the receptor age group of interest.
The values of Rgp j presented in Table 3-8 through Table 3-11 are applicable I only to the currently-defined controlling receptor, so that when dose calculations must be perfomed for a different receptor age group, Rgp; values applicable to that receptor must first be calculated. Methods and parameters for calculating Rgpj for radionuclides and age groups other than those required in Section 3.4.3 are presented in Chapter 9.
O 3-41 Gen. Rev. 13
FNP-0-M-011 Table 3-8. Raipj for Ground Plane Pathway, All Age Groups
%.J Nuclide T. Body Skin H-3 0.00 0.00 Cr-51 4.66E+06 5.51E+06 Mn-54 1.39E+09 1.63E+09 Fe-55 0.00 0.00 Fe-59 2.73E+08 3.21E+0B Co-58 3.79E+08 4.44E+0B Co-60 2.15E+10 2.53E+10 Ni-63 0.00 0.00 Zn-65 7.47E+0B 8.59E+08 Rb-86 8.99E+06 1.03E+07 Sr-89 2.16E+04 2.F1E+04 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 k Sb-124 5.98E+08 6.90E+08 Sb-125 2.34E+09 2.64E+09 Te-125m 1.55E+06 2.13E+06 Te-127m 9.16E+04 1.08E+05 Te-129m 1.98E+07 2.31E+07 I-131 1.72E+07 2.09E+07 I-133 2.45E+06 2.98E+06 Cs-134 6.86E+09 8.00E+09 Cs-136 1.51E+08 1.71E+08 Cs-137 1.03E+10 1.20E+10 Ba-140 2.05E+07 2.35E+07 Ce-141 1.37E+07 1.54E+07 Ce-144 6.95E+07 8.04E+07 Pr-143 0.00 0.00 Nd-147 8.39E+06 1.01E+07
- 1. Units are m 2- (mrem /yr)/ (pci/s) .
- 2. The values in the Total Body column also apply to the Bone, Liver, Thyroid, Kidney, Lung, and GI-LLI organs.
- 3. This table also supports the calculations of Section 6.2.
O 3-42 Gen. Rev. 13
FNP-0-M-011 Table 3-9. Rgg for Inhalation Pathway, Child Age Group O
N.)
Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E403 l Cr-51 0.00 0.00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 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 1.9BE+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 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 C,,T
/ Sb-124 0.00 0.00 0.00 0.00 0.00 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 l
I-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 0.00 2.84E+03 I-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E404 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 1
Cs-137 9.07E+05 8.25E+05 1.28E+05 0.00 2.82E+05 1.04E+05 3.62E+03 Ba-140 l
7.40E+04 6.48E+01 4.33E+03 0.00 2.11E+01 1.74E+06 1.02E+05 Ce-141 3.92E+04 1.95E+04 2.90E+03 0.00 8.55E+03 5.44E+05 5.66E+04 Ce-144 6.77E+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 1.08E+04 8.73E+03 6.81E+02 0.00 4.81E+03 3.28E+05 8.21E+04 i 1
i Units are (mrem /yr)/(pci/m3 ) for all radionuclides. !
i h
G('~
3-43 Gen. Rev. 13
F 1 1 \
l 1
FNP-0-M-011 Table 3-10. Rgg for Cow Meat Pathway, Child Age Group
(-
k_
% 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 I
Cr-51 i 0.00 0.00 8.79E+03 4.88E+03 1.33E+03 8.91E+03 4.66E+05 i
Mn-54 0.00 8.01E+06 2.13E+06 0.00 2.2CE+06 0.00 6.72E+06 Fe-55 4.57E+08 2.42E+08 7.51E+07 0.00 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 0.00 6.11E+08 Nb-95 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 4.44E+09 0.00 5.54E+08 0.00 5.99E+09 0.00 6.90E+10 e Ag-110m 8.39E+06 5.67E+06 4.53E+06 0.00 1.06E+07 0.00 6.74E+08 4 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 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 1.60E+08 0.00 0.00 5.49E+08 Te-127m 1.77E+09 4.78E+08 2.11E+08 4.24E+08 5.06E+09 0.00 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 1.4BE+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 1.6BE+08 8.16E+06 Cs-136 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 3.84E+04 2.56E+06 0.00 1.25E+04 2.29E+04 2.22E+07 Ce-141 2.22E+04 1.11E+04 1.64E+03 0.00 4.86E+03 0.00 1.38E+07 Ce-144 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 0.00 3.60E+07 Nd-147 1.17E+04 9.47E+03 7.33E+02 0.00 5.19E+03 0.00 1.50E+07 3
Units are (mrem /yr) / (pCi/m ) for tritium, and m - 2(mrem /yr) / (pCi/s) for all other radionuclides.
3-44 Gen. Rev. 13
r-l FNP-0-M-011 Table 3-11. Raipj for Garden Vegetation Pathway, Child Age Group 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 l
Cr-51 0.00 0.00 1.17E+05 6.50E+04 1.78E+04 1.19E+05 6.21E+C6 Mn-54 0.00 6.65E+08 1.77E+08 0.00 1.86E+0B 0.00 5.58E+08 Fe-55 8.01E+08 4.25E+0B 1.32E+08 0.00 0.00 2.40E+08 7.67E+07 i 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 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 Zn-65 8.13E+08 2.16E+09 1.35E+09 0.00 1.36E+09 0.00 3.80E+08 f Rb-86 0.00 4.52E+08 2.78E+08 0.00 0.00 0.00 2.91E+07 Br-89 3.60E+10 0.00 1.03E+09 0.00 0.00 0.00 1.39E+09 j 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 Nb-95 4.10E+05 1.60E+05 1.14E+05 0.00 1.50E+05 0.00 2.96E+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 0.00 4.04E+07 0.00 2.58E+09 Sh-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 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 1.57F,+08 3.16E+08 3.77E+09 0.00 1.07E+09 i
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-123 3.53E+06 4.37E+06 1.65E+06 8.11E+08 7.28E+06 0.00 1.76E+06 Cs-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 2.27E+08 1.47E+08 0.00 1.21E+08 1.80E+07 7.96E+06 Cs-137 2.39E+10 2.29E+10 3.38E+09 0.00 7.46E+09 2.68E+09 1.43E+08
)
Ba-140 2.77E+08 2.42E405 1.61E+07 0.00 7.89E+04 1.45E+05 1.40E+08 Ce-141 l 6.56E+05 3.27E+05 4.86E+04 0.00 1.43E+05 0.00 4.08E+08 l
Ce-144 1.27E+08 3.98E+07 6.78E+06 0.00 2.21E+07 0.00 1.04E+10 Pr-143 1.46E+05 4.37E+04 7.23E+03 0.00 2.37E+04 0.00 1.57E+08 Nd-147 7.15E+04 5.79E+04 4.48E+03 0.00 3.18E+04 0.00 9.17E+07 (
Units are (mrem /yr)/(pci/m3 ) for tritium, and m - 2(mrem /yr) / (pci/s) for all other radionuclides.
3-4b Gen. Rev. 13 i
j
PNP-0-M-011 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS
/^
( 3.5.1 Thirty-One Day Dose Prolections In order to meet the requirements of the limit for operation of the gaseous radwaste treatment system (see Section 3.1.5), dose projections must be made at 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 materials occurs or is expected.
Projected 31-day air doses and doses to individuals due to gaseous effluents may be determined as follows:
For air doses:
=
'D6c' Dgp x 31 + Dga (3.17)
'Dyc' Dyp = x 31 + D ya For individual doses:
Dop = x 31 + Du (3.18) where:
Dgp = the projected air dose due to beta emissiors 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 quarter, plus the release under consideration.
D
- a = the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during he 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, D$a may be set to zero.
D the projected air dose due to gamma emissions from noble gases for 9=
the next 31 days of gaseous releases.
O 3-46 Gen. Rev. 13
m FNP-0-M-011 Dye - the cumulative air dose due to gamma emissions from noble gas
(~'s releases that have occurred in the elapsed portion of the current h quarter, plus the release under consideration.
Dya = the anticipated air dose due to gamma 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 ef fluents. If only routine gaseous effluents are anticipated, Dya may be set to zero.
Dop = the projected dose to the total body or organ o, due to releases of I-131, I-133, trii and particulates for the next 31 days of gaseous releases.
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.
Doa = the anticipated dose to the total body or organ o, due to releases i
of I-131, I-133, tritium, and particulates, contributed by any
( planned activities during the next 31-day period, if those t) activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, Doa may be set to zero.
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) .
3.5.2 Dose Proiections for Soecific 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. 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 te exist during the release period.
U,-
3-47 Gen. Rev. 13 L
FNP-0-M-011 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS Section of i
Term Definition Initial Use l
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 fact.or for gaseous source atte&c s, applied to divina the gaseous release limit among all the release pathwayr
[unitless). 3.3.3 AGy= the administrative allocation factor for gaseous release pathway v, applied to divide the gaseous release limit among all the release pathways
[unitless). 3.3.2.2 e= the setpoint of the radioactivity monitor measuring i the concentration of radioactivity in the effluent line prior to release [pCi/mL). 3.3.2.1 m '
ens = the calculated noble gas effluent monitor setpoint for gaseous source stream s [pci/mL). 3.3.3 cnv = the calculated noble gas effluent monitor setpoint for release pathway v [pci/mL). 3.3.2.2 Dja = 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 [ mrem). 3.4.3 Doa = the anticipated dose to organ o due to releases of non-noble-gas radionuclides, contributed by any planned activities during the next 31-day period
[ mrem). 3.5.1 Doc = the cumulative dose to organ o due to releases of l non-noble-gas radionuclides that have occurred in the elapsed portion of the current quarter, plus the release under consideration [ mrem). 3.5.1 l
3-48 Gen. Rev. 13
.c l PNP-0-M-011 Section of IEE Definition Initial Use D the projected dose to organ o due to the next 31 days 9-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 l activities during the next 31-day period [ mrad] . 3.5.1 ;
i l
Dg - the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed !
l portion of the current quarter, plus the release under consideration [ mrad]. 3.5.1 Dg - the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases
[ mrad). 3.5.1 Dy -- the air dose due to gamma emissions from noble gas radionuclides [ mrad]. 3.4.2 1
Dyg = 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 ;
1 Dy= the cumulative air dose due to garmna emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release i 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 i
i O
3-49 Gen. Rev. 13
FNP-0-M-011 Section of M Definition Initial Use A
(D70)9 = 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
[ mrem /y]. 3.4.1.1 DRo= the dose rate to organ o at the time of the release
[ mrem /y].
3.4.1.2 DRt= the total body dose rate at the time of the release
[ mrem /y). 3.4.1.1 fav - the maximum anticipated actual discharge flowrate for release pathway v during the period of the planned release [mL/s]. 3.3.2.2 fu= the maximum anticipated actual discharge flowrate for gaseous source stream s during the period of the 1 planned release [mL/s]. 3.3.3 K =
the total body dose factor due to gamma emissions from noble gas radionuclide i, from Table 3-5
[(mrem /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 I
[(mrem /y)/(pci/m 3)) . 3.3.2.2 Mj = the air dose factor due to gamma emissions from noble gas radionuclide i, from Table 3-5
[(mrad /y)/(yCi/m )) .
3 3.4.2 N= the number of simultaneously active gaseous release i pathways [unitless). 3.3.4 i
O '
3-50 Gen. Rev. 13
FNP-0-M-011 Section of g Definition Initial Use Ng = the air dose factor due to beta emissions from noble gas radionuclide i, from Table 3-5
[ (mrad /y) / (pci/m3) } . 3.4.2 Pjo = the site-specific dose factor for radionuclide 1 (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
[ (mrem /y) / (pCi/m3) }. 3.4.1.2 Ojy = the release rate of noble gas radionuclide i from release pathway v during the period of interest
[pCi/s) . 3.3.2.2 Qjy= 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 [yci/s). 3.4.1.2 6jy - the cumulative release of noble gas radionuclide i from release pathway v during the period of interest
[pci) . 3.4.2 6jy= the cumulative release of non-noble-gas radionuclide 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-nuclide i, exposure pathway p, and organ j. Values and units of Raipj for each exposure pathway, age group, and radionuclide that may arise in
~
calculations for implementing Section 3.1.4 are listed in Table 3-8 through Table 3-11. 3.4.3 Rk= the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in b
v the release under consideration [unitiess) . 3.3.2.1 3-51 Gen. Rev. 13
FNP-0-M-011 Section of M Definition Initial Use Rg= the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the release under consideration [unitiess) . 3.3.2.1 rk = the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in the source stream under consideration [unitiess). 3.3.3.1 rg = . the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the source stream under consideration
[unitiess). 3.3.3.1 SF = the safety factor used in gaseous setpoint calculations to compensate for statistical 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 Q under consideration. 3.5.1 wjyp= the annual average relative dispersion [$76)yp] or deposition [(D76)yp] at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and radio-nuclide i. 3.4.3 X= the noble gas concentration for the release under consideration [ Ci/mL) . 3.3.2.1 X
ir - the concentration of radionuclide i applicable to active gaseous release pathway r [pci/mL). 3.3.4 X;g = .the measured concentration of radionuclide i in -
gaseous source stream s [pCi/mL). 3.3.3 Xjy - the measured concentration of radionuclide i in gaseous stream v [pci/mL]. 3.3.2.2 3-52 Gen. Rev. 13
PNP-0-M-011 Section of IRIE Definition Initial Use s
(x/Q) -
the highest relative concentration at any point at or beyond the SITE BOUNDARY [s/m3 ]. 3.3.2.1 (17D)g = the annual average SITE BOUNDARY relative concen-tration applicable to active gaseous release pathway r ' [s/m3) . 3.3.4 (17D)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 (17D)yp - annual average relative dispersion factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [s/m 3]. 3.4.3 I
i 1
l
\ 1 l
3-53 Gen. Rev. 13 i
FNP-O-M-011 CHAPTER 4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.1 LIMITS OF OPERATION The following limits are the same for both units at the site. Thus, a single program including monitoring, land use survey, and quality assurance serves both units.
4.1.1 Radiolocical Environmental Monitorino In accordance with Technical Specification 6. 8. 3. f (i) , the Radiological l Environmental ' Monitoring Program (REMP) shall be - conducted as specified in Table 4-1.
4.1.1.1 Applicability This control applies at all times.
4.1.1.2 Actions 4.1.1.2.1 With the REMP not being conducted as specified in Table 4-1, submit j to the Nuclear Regulatory Commission (NRC), in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. Deviations from the required sampling schedule are permitted if specimens are unobtainable due to hazardous conditions, unavailability, inclement weather, equipment ;
malfunction, or other just reasons. If deviations are due to equipment malfunction, efforts shall be made to complete corrective action prior to the and of'the next sampling period.
4.1.1.2.2 With the confirmedl 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. 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 1
Defined as confirmed by reanalysis of the original sample, or analysis of a duplicate or new sample, ss appropriate. The results of the confirm-atory analysis shall be completed at the earliest time consistent with the O
analysis.
4-1 Gen. Rev. 13 l
PNP-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 ifs concentration (1) , concentration (3) , , , , , 1,n l
reporting 1evel i1) reporting 1evel (2)
)
When radionuclides other than those in Table 4-2 are detected and are the result of plant effluents, this special Report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits stated in Sections 2.1.3, 3.1.3, and 3.1.4. This Special Report is not required 11 the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be described in the
-Annual Radiological Environmental Operating Report. The levels of naturally-occurring radionuclides which are not included in the plant's effluent releases 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 O, sample locations required by Table 4-1, or if the availability is frequently or 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 seleccion of any new location (s) . '
4.1.1.2.4 This control does not affect shutdown requirements or MODE changes. l i
4-2 Gen. Rev. 13
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 unachieval'e. 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 levels of radiation, which may then be compared with those expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.
The detection capabilities required by Table 4-3 are within state-of-the-art for routine environmental measurements in industrial laboratories.
O 4-3 Gen. Rev. 13
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r FNP-0-M-011 4.1.2 Land Use Census
{ 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 anima 1I and the nearest permanent residence, in each of the 16 meteorological sectors, within a distance of 5 miles.
4.1.2.1 Applicability This centrol 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 currently 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 location (s) which yields a calculated dose or dose commitment (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 location (s),' having the lowest calculated dose or dose commitment (via.the same exposure pathway) may be deleted from the REMP if new sampling locations are added. Pursuant to Technical Specification 6.14 submit in . the next Annual Radioactive Effluent Release Report any change (s) in the ODCM, including the revised figure (s) and table (s) reflecting any new location (s) and information supporting the change (s).
4.1.2.2.3 This control does not affect shutdown requirements or MODE changes.
4.1.2.3 Surveillance Requirements 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 automobile or aircraft, consultation with local agriculture authorities, or some combination of these methods, as feasible. Results of the land use censui shall be included in the Annual Radiological Environmental Operating Report.
1 Defined as a cow or goat that is producing milk for human consumption.
I 4-8 Gen. Rev. 13 !
t FNP-0-M-011 4.1.2.4 Basis 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 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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1 49- Gen. Rev. 13
b FNP-0-M-011 4.1.3 Interlaboratory canmarison Proaram In accordance with Technical Specification 6.8.3.f(iii), analyses shall be j performed _ on radioactive materials supplied as part of an Interlaboratory
! s 'atisfies the requirements of Regulatory Guide 4.15, Comparison Program which Revision 1, February 1979.
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.
l 4.1.3.3 Surveillance Requirements A summary of the results obtained as part of the required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.
4.1.3.4 Basis
- The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and l accuracy of the measurements of radioactive material in environmental sample 1 matrices are perfonned as part of the quality assurance program for environmental monitoring, in order to demonstrate that the results are reasonably valid for the ,
purposes of Section IV.B.2, Appendix I, 10 CFR 50. I i
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4-10 Rev. 15
i FNP-0-M-011 4.2 -RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONE 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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4-11 Gen. Rev. 13 I
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FNP-0-M-011 Table 4-4. Radiological Environmental Monitoring Locations O Exposure Path aY
, dfor Sampling Locations
- Iden ifi-Sample cation
- 1. AIRBORNE Partic- Indicator Stations:
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 Neals Landing, FL (SSE-18 miles)I PB-0718 Community Stations:
Georgia Pacific Paper Co. (SSE-3 miles) PC-0703 Ashford, AL (WSW-8 miles) PC-1100 Columbia, AL (N-5 miles) l PC-1605
~
Radiciodine Indicator Stations: !
River Intake Structure (ESE-0.8 miles)I II-0501 l 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:
O Blakely, GA (NE-15 miles)
Dothan, AL (W-18 miles)
Neals Landing, FL (SSE-18 miles)I IB-0215 IB-1218 IB-0718 Community Stations:
Georgia Pacific Paper Co. (SSE-3 miles)2 re_o7o3
- 2. DIRECT RADIATION TLD Indicator I Stations:
Plant Perimeter (NNE-0.9 miles) RI-0101 (NE- 1.0 miles) RI-0201 (ENE-0.9 miles) RI-0301 (E- 0.8 miles) RI-0401 (ESE-0.8 miles) RI-0501 (SE- 1.1 miles) RI-0601 (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 O
4-12 Gen. Rev. 13
FNP-0-M-011 Table 4-4 (contd) . Radiological Environmental Monitoring Locations O Exposure Pathway Sample and/or Sampling Locations, Identifi-Sample cation TLD (contd) Control Stations:
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 Webb, AL (WNW-11 miles) RB-1311 Haleburg, AL (N-12 miles) RB-1612 Indicator II (Community) Stations:
(NNE-4 miles) RC-0104 (NE- 4 miles) RC-0204 (ENE-4 miles) RC-0304 (E- 5 miles) RC-0405 (ESE-5 miles) RC-0505 (SE- 5 miles) RC-0605 (SSE-3 miles) RC-0703 (S- 5 miles) RC-0805 (SSW-4 miles) RC-0904 (SW- 1.2 miles) RC-1001 (SW- 5 miles) RC-1005 (WSW-4 miles) RC-1104 (WSW-8 miles) RC-1108 s
(W- 4 miles) RC-1204 g (WNW-4 miles) RC-1304
) (NW- 4 miles) 2C-1404 k_/ (NNW-4 miles) RC-1504
(;N - 5 miles) RC-1605
- 3. WATERBORNE Surface Indicator Station:
Georgia Pacific Paper Co. Intake Struccure WRI (River Mile - 40)
Control Station:
Andrew Lock & Dam Upper Pier (River Mile - 47) WRB Ground Indicator Station:
Georgia Pacific Paper Co. Well (SSE-4 miles) WGI-07 Control Station:
Whatley Well (SW-1.2 miles) WGB-10 Sediment Indicator Station:
Smith's Bend (River Mile - 41)3 RSI Control Station:
Andrews Lock & Dam Reservoir (River Mile - 48)3 RSB _
O 1
4-13 Rev. 16
FNP-0-M-011 t
Table 4-4 (contd). Ra11ological Environmental Monitoring Locatione f
Exposure Pat ay 3,,p y ,
, , Sempling Locations *
' Identifi-Sample cati n
___ mar 4 INGESTION Milk Indicator' Station None (There are no milk animale within 5 miles per the current land use survey)
Control Station:
Robert Weir Dairy MB-C?le Donaldsonville, GA (SSE-14 miles) l Fieh Indienter station:
Smith Band (River Mile - 41)4 Game Fia'h FGI Bottom Feeding Fish FBI Centrol StN.ian*
Andrews Lock & Dam Reservoir (River Mile - 46)4 Game Pialb FGB Bottom Feeding Fish FBB Forage or Leafy Indiester Stations:
Vegetation south Sout'neast Perimeter (SSE-1.0 miles) FI-0701 North FI-lG01 South Perimeter Perimeter (S-1.0 I
(N-0.8m11es) miles)5 FI-oe01 Northemat Perimeter (13-1.0 miles)5 FI.0101
() " "'" ' '"**
Dothan, AL (W-18 miles)
FB-1218 Distance and di; ection as measured from the centerpoint between Unit 1 and Unit 2 plant vent etacks.
- 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. These collectiocs are normally made at river mile 4? 3 for the indicator station and mils 47.8 for the control stations however, due to river botton. sediment shifting caused by high flows, dredging, etc., collections may be made from river mile to to 42 for the indicator station and from river mile 47 to 49 for the control station.
4.
Since a few milus of river water may be needed to obtain adequate fish samples, these river mile positions represent the approximate 1 Locations about which the catches are taken.
collectione for the indicator station should te from river _
mile 47 to37.5
- 52. to(CAR 42.E and for the control etetion from river mile
,2203)
- 5. Alternate forage plots.
/~~N 4-14 Rev. 1B
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FNP-0-M-011 O
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4-16 Gen. Rev. 13
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FNP-0-M-011 0
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1 FNP-0-M-011 CHAPTER 5 i g~. E TAL DOSE DETERMINATIONS b
I 5.1 LIMIT OF OPERATION In accordance with Technical Specification 6. 8. 3. e (x) , the dose or dose 1
commitment to any MEMBER OF THE PUBLIC over a calendar year, due to releases of I radioactivity and to radiation from uranium fuel cycle sources, shall be limited to less than or equal to 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to ~15 mrem. j 5.1.1 Aeolicability i
This limit applies at all times. '
5.1.2 Actions With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Section 2.1. 3, 3.1. 3, or 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 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 tacen to reduce subsequent releases to 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 <iirect 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 iteport 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 5-1 Gen. Rev. 13
FNP-0-M-011 This control does not affect shutdown requirements or MODE changes.
D 5.1.3 Surveillance Recuirements Cumulative dose contributions from liquid and gaseous ef fluents and from direct 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 requires the preparation and submittal of a Special Report whenever the calculated doses-from plant radioactive effluents combined with doses due to 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 MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors reme.in within the reporting requirement level. The Special Report will 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 the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible with the exception that dose contributions from other uranium fuel cycle facilities at the same site or within a radius of 5 miles must be considered.
If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.2203 (a) (4), is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is 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. l l
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5-2 Gen. Rev. 13
FNP-0-M-011 5.2 DEMONSTRATION OF COMPLIANCE O
There are'no other uranium fuel cycle facilities within 5 miles of the plant site. Therefore, for the purpose of demonstrating complisnee with the limits of Section 5.1,'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:
Dn = DL*DG*ID+Dy '(5,1) where:
g= the total dose or dose commitment to the total body or organ k, in mrom.
DL= the dose to the same organ due to radioactivity discharged from the plant site in liquid effluents, calculated in accordance with Section 2.4.1, in mrem.
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.
O Dp . -- the direct radiation dose to the whole body of an individual.at the I
controlling receptor location, due to radioactive materials retained within the plant site, in mrem. Values of direct radiation dose may be determined by measurement, calculation, or a combination of the two.
Dy - tho' 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 effluents, in mrem. Dy is calculated as follows (equation adapted from Reference 1, page 22, by re-casting in cumulative dose form) :
Dy = 3.17 x 10 {v Kg
- Ojy (5.2)
(YTC)9 {
, I where:
5-3 Gen. Rev. 13
PNP-0-M-011 3.17 x 10 a units conversion factor: 1 y/ (3.15 x 10 7 s).
6;y - the cumulative release of noble gas radionuclide i from release pathway v (pci), during the period of interest.
Kg = the total-body dose factor due to gamma emissions from noble gas radionuclide i (mrem /y)/(pci/m3 ), from Table 3-5.
(176)yp - annual average relative dispersion factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [s/m3] .
1 As defined above, Dg and Dg are for different age groups, while DD and Dg are not age group specific. When a more precise determination of Dn is desired, values of DL and DG may be calculated for all four age groups, and those values used in equation (5.1) to determine age group specific values of Dg; the largest value of Dn for any age group may then be compared to the limits of Section 5.1.
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5-4 Gen. Re.v. 13
FNP-0-M-011 CHAPTER 6 f POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO
's THEIR ACTIVITIES INSIDE THE SITE BOUNDARY 6.1 REQUIREMENT FOR CALCUIATION Current FNP effluent controls as established by this ODCM do not require assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 10-1). However, when such an assessment is desired, it should be performed in accordance with Section 6.2.
6.2 CALCULATIONAL METHOD For the purpose of performing 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:
/%
& DJk ={DA+DS*DP}'Fo (6.1) l where:
Dg - the total dose to the total body or organ k, in mrem.
DA= the dose to the same organ due to inhalation of non-noble-gas l radionuclides dischargod from the plant site in gaseous effluents, calculated in accordance with Section 3.4.3, in mrem. The (176) value to be used is given for each receptor location in Table 6-1; depleted (27D) values may be used in calculations for non-noble-gas radionuclides.
D3= the dose to the same organ due to around niane deoosition ~f non-noble-gas radionuclides discharged from the plant site in gaseous effluents, calculated in accordance with Section 3.4.3, in mrem.
The (576) value to be used is given for each receptor location in m Table 6-1.
6-1 Gen. Rev. 13
r FNP-0-M-011 Dp = the external whole body dose due to gamma ray emissions from noble f' gas radionuclides discharged from the plant site in gaseous t effluents, calculated using equation (5.2), in mrem. The (176) values that are to be used are given for each receptor location in "
Table 6-1.
Fo= the occupancy factor for the given location, which is the fraction of the year that one individual MEMBER OF THE PUBLIC is assumed to be present at the receptor location (unitiess). Values of F foro each receptor location are included in Table 6-1.
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FNP-0-M-011 I Table 6-1. Attributes of Member of the Public Receptor Locations Inside the l SITE BOUNDARY
,s i Locations visitor Center, WSW at 0.19 miles l
) Ace GrouD: Child Occupancy Factor: 1.37 E-03 (based on 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per year)
Dispersion and Deoosition Parameters:
Parameter Ground-Level Mixed-Mode ,
(17D), s/m3 1.04 E-04 8.80 E-06 (57D), m-2 4.80 E-07 6.20 E-08 Location: Service Water Pond, SSW at 0.60 miles Ace Groun: Child Occupancy Factor 7.53 E-03 (based on 66 hours7.638889e-4 days <br />0.0183 hours <br />1.09127e-4 weeks <br />2.5113e-5 months <br /> per year)
DisDersion and DeDosition Parameters:
Parameter Ground-Level Mixed-Mode (57D), s/m3 4.74 E-05 9.75 E-07 (676), ac2 1.31 E-07 2.78 E-08 l
6-3 Gen. Rev. 13 l
FNP-0-M-011 l Table 6-1 (contd). Attributes of Member of the Public Receptor Locations
"' Inside the SITE BOUNDARY Locations River Water Discharge, SE at 1.02 miles Ace Groun: Child Occuoancy Factor: 1.14 E-02 (based on 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year) .
l Dispersion and DeDosition Parameters- l l
Parameter Ground-Level Mixed-Mode (176), s/m3 1.63 E-05 7.05 E-07 (D76), m"2 4.55 E-08 1.39 E-08 i
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6-4 Gen. Rev. 13
PNP-0-M-011 CHAPTER 7 REPORTS 7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 7.1.1 Reauirement for ReDort In accordance with Technical Specifiestions 6.9.1.6 and 6. 9.1. 7, the Annual Radiological Environmental Operating Report covering the REMP activities during 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 of the ODCM, and in Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. i 7.1.2 ReDort Contents The materials specified in the following sub-sections shall be included in each Annual Radiological Environmental Operating Reportt 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.
In the event that some results are not available for inclusion with the repcrt, 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 l
included in plant effluents need not be reported. )
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7.1.2.2 Evaluations Interpretations and analyses of trends of the results shall be included in the !
report, including the following: (as appropriate) comparisons with pre-operational studies, operational controls, and previous environmental operating reports; and an assessment of any observed impacts of the p3 r.nt operation on the environment. If the measured level of radioactivity in an environmental sampling medium exceeding the reporting levels of ..le 4-2 is not the result of plant effluents, the condition shall be described as required by Section 4.1.1.2.2.
O 7-1 Gen. Rev. 13
I FNP-0-M-011 7.1.2.3 Programmatic Information Also to be included in each report are the following: 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.
7.1.2.4 Descriptions of Program Deviations Discussions of deviations from the established program must be included in each report, as follows:
7.1.2.4.1 If the REMP is not conducted as required in Table 4-1, a description of the reasons for not conducting the program as required, and the plans for preventing a recurrence, must be included in the report.
7.1.2.4.2 If the MDCs required by Table '4-3 are not achieved, the !
contributing factors must be identified and described in the report. l 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, l
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m FNP-0-M-Og 7.2 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT
(
( 7.2.1 Reapirement for Reoort In accordance with Technical Specifications 6.9.1.8 and 6. 9.1. 9, the Annual Radioactive Effluent Release Report covering the operation of the unito during the previous calendar year of operation shall be submitted before May 1 of each year. (A single submittal may be made for Units 1 and 2. However, the submittal shall specify the releases of radioactive material in liquid and gaseous effluents 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 be consistent with the objectives outlined throughout this ODCM and the Process Control Program (PCP) and in conformance with 10 CFR Part 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
7.2.2 ReDort Contents The materials specified in the following sub-sections shall be included in each Annual Radioactive Effluent Release Report:
7.2.2.1 Quantities of Radioactive Materials Released
()
fh The report shall include a summary of the quantities of radioactive liquid and i
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 I Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, l with liquid and gaseous effluent data summarized on a quarterly basis and solid j
radioactive waste data summarized on a semiannual basis following the format of 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 l liquid and gaseous effluents, in addition to the principal gamma emitters for which MDCs are specifically established in Table 2-3 and Table 3-3, other peaks !
which are measurable and identifiable also shall be identified and reported. {
7.2.2.2 Meteorological Data ~
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 l of an hour-by-hour listing of wind speed, wind direction, and atmospheric j
/G stability, and precipitation (if measured) on magnetic tape; or in the form of G 1 7-3 Gen. Rev. 13 3
PNP-0-M-011 joint frequency distributions of wind speed, wind direction, and atmospheric stability.
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 meteorclogy or the meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway 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 effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation; this dose assessment must be performed in accordance with Chapter 5.
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 or 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), anB
- f. Solidification agent (e.g., cement, urea formaldehyde.)
7-4 Gen. Rev. 13
FNP-0-M-011 7.2.2.5 Licensee Initiated Document Changes O
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.
7.2.2.5.2 Any changes to dose calculation locations or pathways, including any changes nade 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 ;
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 initiated MAJOR CHANGES TO RADIOACTIVE WASTE TREA'INENT 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 l l
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In lieu of inclusion in the Annual Radioactive Effluents Release Report, m/
) this same information may be submitted as part of the annual FSAR update.
7-5 Gen. Rev. 13
FNP-0-M-011 s
.,_ The discussion of each change shall contain:
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; 7.2.2.7.2 Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; 7.2.2.7.3 A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; 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 amendments 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 rac"'4ctive materials, in liquid and gaseous effluents, to the actual releases for th period prior to when the changes are to be made; 7.2.2.7.7 An estimate of the exposure to plant operating personnel 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.
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l 7.3 MOFFTHLY OPERATING REPORT l
( This ODCM establishes no requirements pertaining to' the Monthly Operating Report.
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7.4 SPECIAL REPORTS Special reports shall be submitted to the Nuclear Regulatory Commission in accordance with Technical Specification 6.9.2, as required by Sections 2.1.3.2, 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 CHAPTER 8 METEOROLOGICAL MODELS The models presented in this chapter are those which were used to compute the 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 convenience in calculating mixed-mode values, and to preserve -section number 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 elevations for each gaseous release point are as indicated in Table 3-4.
b 8.1.1 Ground-Level Releases Relative concentration calculations for ground-level releases, or for the ground-level portion of mixed-mode releases, shall be made as follows:
2.032 6 K' Djk (x/0)g =
{ (8.1)
Nr *) Ezk jg where:
(X/Q)G = the ground-level sector-averaged rclative concentration for a given wind direction (sector) and distance (s/m3),
2.032 = (2/w) l/2 divided by the width in radians of a 22.5* sectort which is 0.3927 radians.
6= the plume depletion factor for all radionuclides other than noble gases at a distance r shown in Figure 8-3. For noble gases, the g depletion factor is unity. If an undepleted relative concentration
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U 8-1 Gen. Rev. 13
FNP-0-M-011 is desired, the depletion factor is unity. Only depletion by s
deposition is considered since depletion by radioactive decay would be of little significance at the distances considered.
Kr= the terrain recirculation factor corresponding to a distance r, taken from Figure B-2.
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 throughout 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 (m/s).
r= the distance from release point to location of interest (m).
Ed= the vertical standard deviation of the plume concentration f distribution considering the initial dispersion within the building wake, calculated as follows:
'2 b2'1/2 i
[d 2E Ed" Ch' l'88'# Of: OR (s*2}
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I ak=
z the vertical standard deviation of the plume concentration distribution (m) for a given distance and stability category k as shown in Figure 8-1. The stability category is determined by the vertical temperature gradient AT/Az (*c/100 m or *F/100 f t) . Plant Farley AT/Az values must be adjusted for Az of 165 ft.
w= 3.1416 b= the maximum height of adjacent plant structure, which is the containment building (40 m) .
O 8-2 Gen. Rev. 13
FNP-0-M-011 8.1.2 Elevated Releases 1 O
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Relative dispersion calculations for elevated releases, or for the elevated portion of mixed-mode releases, shall be made as follows:
' _y2' (x/0)E " ,2ah I8'3I "J 'zk where:
(X/Q)E = the elevated release sector-averaged relative concentration for a given wind direction (sector) and distance (s/m 3) .
6g= 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 B-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 O 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 j into the given sector during the time atmospheric stability j category k existed. l u; = the wind speed (mid-point of wind speed class j) at the effective release height h (m/s) .
h= the effective height of the release (m), which is calculated as follows:
h=h y + hp , - hg - c y (8.4) hy= the height of the release point (m).
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r FNP-0-M-011 ht= the maximum terrain height between the release point and the point of interest (m), from Figure 2.3-26 of Reference 7.
hpr = the additional height due to plume rise (m) which is calculated as follows and limited byp h g):
i2 1 1 hpr = 1.44 d 3-l 1 3 (8.5) f 9 Wo 3 - d
. "J .
hp,(max) =
the lesser of: OR (***)
' y* ' $ b 1.5 3S 6
. "J.
d= the inside diameter of the vent (m).
Wo= the exit velocity of the plume (m/s).
cy = the correction for low vent exit velocity (m), which is calculated as follows:
Wo W 3 1.5- d for o c 1. 5
. u). u) j 1
cy =
og (S*U i
W O for o a 1. 5 i u; '
F=
m the momentum flux parameter 4(m2/s ), which is calculated as follows (under the assumption that the effluent air and the ambient air have the same density):
I 8-4 Gen. Rev. 13
FNP-0-M-011 Fm "
(8 8)
(Wo )
- l l S =' the stability parameter, which is calculated as follows: ;
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, E + 9. 8 x 10 ~ (3.3) i T= the ambient air temperature (*K).
(AT/Az) =
the rate of increase of the ambient air temperature with increasing height above the ground (*K/m) . 3 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 - r) - (x/0)E + E * (X/0)G (3.10) 1 d
where: i (X/Q)M = the mixed-mode release sector-averaged relative concentration for 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:
i 4-5 Gen. Rev. 13
FNP-0-M-011 w
1.0 for' 0 s 1. 0
")
r , l i
WO W 2.58 - 1.58 * - for 1. 0 < O s 1. 5
. ")- ")
E =
(8.11) l No W 0.3 - 0.06 - - for 1. 5 < o s 5.0
,UJ, "}
w 0 for # > 5. 0 "1
All other symbols are as previously defined.
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r FNP-0-M-011 8.2 RELATIVE DEPOSITION Plume depletion 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 elevations for each gaseous release points are as indicated in Table 3-4.
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:
l 2.55 D# K 7 (D/0)G " y g. E "k (8.12) k where (D/Q)G = the ground-level sector-averager' elative deposition for a given wind direction (sector) and distance (m-2),
2.55 = the inverse of the number of radians in a 22.5' sector p [. (2 w/16)-I) .
U Dg= the deposition rate at distance r, taken from Figure 8-7 for ground-level releases (m-I).
nk = the number of hours in which the wind is clirected into the sector of interest, and during which stability category k exists.
All other symbols are as defined previously in Section 8.1.
8.2.2 Elevated Releases Relative deposition calculations for elevated releases, or for the elevated portion of mixed-mode releases, shall be made as follows:
2.55 K -
(D/0)E " Nr Eg ("k Dd) (8.13) where O
8-7 Gen. Rev. 13
FNP-0-M-011 (D/Q)E = the elevated-plume sector-averaged relative deposition for a given wind direction (sector) and distance (m-2) .
Dek - the elevated plume deposition rate at distance 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-I) .
All other symbols are as defined previously.
8.2.3 Mixed-Mode Releases Relative deposition calculations for mixed-mode releases shall be made as follows:
(D/0)y = (1 - E) - (D/0)E + E " (D/0)G (8 14) where (D/Q)g = the mixed-mode release sector-averaged relative deposition for a given wind direction (sector) and distance (m-2),
,l E= the fraction of hours during which releases are considered as ground-level releases, defined in Section 8.1.3.
All other symbols are as previously defined.
B.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 summary of meteorological data for the years 1971 through 1975 is presented in p Table 8-2 through Table 8-5.
B-8 Gen. Rev. 13
WP-0-M-011 Table 8-1. Terrain Elevation Above Plant Site Grade ;
O This table intentionally left blank.
O 8-9 Gen. Rev. 13
FNP-0-M-011 Table 8-2. Annual Average (17D) for Mixed Mode Releases 1
(
+
Distance to Location, in miles 1
\
Sector 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 9.60 E-07 5.52 E-07 3.47 E-07 2.19 E-07 SSE 2.77 E-06 1.08 E-06 6.57 E-07 4.34 E-07 2.81 E-07 l S 2.50 E-06 9.37 E-07 5.90 E-07 4.09 E-07 2.74 E-07 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 8.03 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 l 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 Dista1ce 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 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 8.28 E-08 7.48 E-08 NE 1.61 E-07 1.19 E-07 9.62 E-08 8.63 E-08 7.79 E-08 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 S 1.96 E-07 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 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 NNW 1.42 E-07 1.07 E-07 1.04 E-07 9.36 E-08 8.50 E-08 Values are in s/m 3 , extracted from Reference 7.
8-10 Gen. Rev. 13
l FNP-0-M-011 i
Table 8-3. Annual Average (X7D) for Ground-Level Releases
% Distance to Location, in miles Sector 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 j 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 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 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 Sector 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 G 4.5-4.99 k) N 1.19 E-06 8.24 E-07 6.09 E-07 5.35 E-07 4.71 E-07 NNE 1.00 E-06 6.94 E-07 5.13 E-07 4.50 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-C6 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 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 S.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 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/m 3 , extracted from Reference 7.
O 8-11 Gen. Rev. 13
1 i
FNP-0-M-011 Table 8-4. Annual Average (D76) for Mixed Mode Releases Distance to Locatien, in miles Sector 0.25-0.5 0. 5- 0. S.4 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 1.62 E-09 NNE 4.57 E-08 2.08 E-08 8.69 E-09 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 1.92 E-09 ENE 2.67 E-08 1.32 E-08 5.63 E-09 2.54 E-09 1.22 E-09 E 2.87 E-08 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-00 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 WSW 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-09 1.70 E 09 WNW 3.22 E-08 1.56 E-08 6.35 E-09 2.85 E-09 1.37 E-09 NW 2.83 E-08 1.35 E-08 5.55 E-0) 2.46 E-09 1.18 E-09 NNW 3.24 E-08 1.55 E-08 6.59 E-09 2.97 E-09 1.42 E-09 l 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 l NE 1.03 E-09 6.65 E-10 4.62 E-10 3.98 E-10 3.43 E-10 l ENE 6.57 E-10 4.22 E-10 2.96 E-10 2.55 E-10 2.20 E-10 I E 6.57 E-10 4.20 E-10 2.96 E-10 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 1.01 E-09 6.48 E-10 4.55 E-10 3.90 E-10 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 S 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 SW 1.08 E-09 6.90 E-10 4.81 E-10 4.12 E-10 3.53 E-10 WSW 1.02 E-09 6.51 E-10 4.53 E-10 3.87 E-10 3.32 E-10 W 9.00 E-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 6.37 E-10 4.11 E-10 2.91 E-10 2.50 E-10 2.14 E-10 NNW 7.66 E-10 4.95 E-10 3.45 E-10 2.97 E-10 2.56 E-10 l
Values are in m-2, extracted from Reference 7.
(
(
8-12 Gen. Rev. 13
,,FNP-0-M-011 Table 8-5. Annual Average (fi/D) for Ground-Level Releases
/ Distance to Location, in miles Sector 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 N 2.50 E-07 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 4.24 E-09 NE 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-OB 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 4.71 E-09 SSE 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 6.6~ E-08 2.14 E-08 8.11 E-09 3.61 E-09 i 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 Distance to Location, in miles Sector 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 NNE 2.20 E-09 1.43 E-09 9.71 E-10 8.20 E-10 6.93 E-10 NE 2.21 E-09 1.44 E-09 9.75 E-10 8.23 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 ESE 1.57 E-09 1.02 E-09 6.72 E-10 5.85 E-10 1.94 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 l 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 l 2.05 E-09 1.34 E-09 9.04 E-10 7.64 E-10 6.46 E-10 !
W 1.87 E-09 1.22 E-09 8.25 E-10 6.97 E-10 5.90 E-10 l 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 l NNW 1.89 E-09 1.23 E-09 8.34 E-10 7.04 E-10 5.95 E-10 Values are in mO, extracted from Reference 7.
O 8-13 Gen. Rev. 13
FNP-0-M-011 I A f I f }'
I / /
, / / .* l l s / /
l l ,f , f too A
/ ,
[ ,
/
.r # e s
[ .f n. A / a f Y
/ * "
) f 7 / , , -- --=
k / / / E// m
/
" ~
l / ///
~
// / . --
y / // ,f /
z
'/ / / / af 1
s / / /
/ / af }'
/ / / '
r
/ /
(, f
/
0.1 1.8 18 188 PLutet TRAVEL DesTANCE (KIL0e8ETEllBI I
l l
Range of Vertical Range of Vertical l Category Temperature Gradient Temperature Gradient '
(*C/100 m) (*F/100 ft) l A AT/Az < -1.9 AT/AZ < -1.0 B -1.9 s AT/AZ < -1.7 -1.0 s AT/AZ < -0.9 C -1.7 s AT/AZ < -1.5 -0.9 s AT/AZ < -0.8 i D -1.5 s AT/AZ < -0.f. -0.8 s AT/AZ < -0.3 E -0.5 s AT/AZ < 1.5 -0.3 s AT/AZ < 0.8 F 1.5 s AT/AZ < 4.0 0.8 s AT/AZ < 2.2 G 4.0 s AT/AZ 2.2 s AT/AZ This graph is reproduced from Reference 5 (Figure 1).
Figure 8-1. Vertical Standard Deviation of Material in a Plume (a g) 6-14 Gen. Rev. 13
FNP-0-M-011
[use former Figure 3-3 or comparable)
%/
to i . ..i, ,
. , .i ,
s ,
6e\ \
g l \
lI I I
llll \ l N
8 I \
e- ,
V Ns s
% 'A f 1.c
' l ',
A g U
5 l l l '. ', l
" i I I '
i 4 iil l !: l lll ll 0.1 __
0.1 1.0 le 100 OtBTANCE (KILOMETE11ts)
T This graph is reproduced from Reference 4.
Figure 8-2. Terrain Recirculation Factor (Q) 8-15 Gen. Rev. 13
FNP-0-M-011 0
ta g ,,
N %
u I g 9%
h A u
'N ,
u N y
O I- u N
u e.1 -
I a.1 ta tea tems sees PLusse vnavet eieraseos samassersne 1
This graph is reproduced from Reference 5 (Figure 2). !
O- Figure 8-3. Plume Depletion Effect for Ground Level Releases 8-16 Gen. Rev. 13
- FNP-0-M-011 O
I
-4Q; = a yfAs ts e3 A NftFTRAL f)) I,P,( )
^
82 Ek
' s.7 h
I .,
- =?$k
,, *NN s eA NN
., \
s.1 i
4.1 ts tea sess' .goom j Ktass TRAVEL INSTANCE lEILateETEnst This graph is reproduced from Reference 5 (Figure 3).
(/ Figure 8-4. Plume Depletion Effect for 30 Meter Releases 8-17 Gen. Rev. 13
FFP-0-M-011 o .
18 .TAats'.
y % %
Iu h"V'T
,gY mW lu umm T
%g i
\ N Iuu \
h(
\
O lu., x
.. T e.1 ,
l 6.1 ts ,gs m gen PwmatnaveLaseramentumanettsam g This graph is reproduced from Reference 5 (Figure 4).
Figure 8-5. Plume Depletion Effect for 60-Meter Releases 8-18 Gen. Rev. 13
FN1 J-M-011 O
j t
12 --
% N
'~
,, % -, _Neumt m)
' *= , N N N
! ,, umsTAsts %%
s' n WAM g ETABLElt.F A %g No DEPLETot, q \
I ,e,s IFRACTMiss RWIAAsNING = tm K
\
(h s
V e IeAeJ u
l I
s.1 e1 1A tea tesa sana FLunit TRAVEL DETANCE (EIL0estTERei i l
This graph is reproduced from Reference 5 (Figure 5).
( Figure 8-6. Plume Depletion Effect for 100-Meter Releases B-19 Gen. Rev. 13 L ._
r l
FNP-0-M-011 l N !
l 1
l M .
N 10 4 N -
a S ,
g '
t I \
r T m
$ TA E Ns a s 104 (
E
's s
E x y '
\
E 5 h '
E
- 104
's tr7 0,1 1A 10A 100A 200A PLUnst TRAVEL DISTANCE (KILOMETERS)
O This graph is reproduced from Reference 5 (Figure G).
Figure 8-7. Relative Deposition for Ground-Level Releases 8-20 Gen. Rev. 13
)
I FNP-0-M-011 l
O )
e 1
)
e u-r- um
=
r a' '
w
't
\
'hm E # '\
['
4 g h NEUTRAL l e
< / TA I r
NEUTRALEN g g i
5 srAaLE i te4 1
/ A '
< l 8 c c w' , <
l I
('**== m 7
- g O 2 /
/
N h EU.BLE (E,FA l
r l 1 i !
I l
I f
1V1 0.1 1A 10A 100A 3DOA PLuteE TRAVEL DISTANCE (KILOeAETERSI l
1 i
iI 1
i O Figure 8-8.
This graph is reproduced from Reference 5 (Figure 7).
Relative Deposition for 30-Meter Releases j
8-21 Gen. Rev. 13
FNP-0-N-011 0
104 f , UNSTABLE (A,0,C)
[ \
/ N
$g4 l
/ /
f NEUTRAL %(D) k 4(
E i a l ux w I 1 W
$ / / NA 3 / \ ^ v UNSTABLE g / '
g N g w ! NEUTRAL \ \
l f . N T 5iod ,
/ i=
e
. - ~
m 1 o
'O e E
l l I
f l
E w i 5 i I E
10-7 I f STABLE (E,F,G) l '
I i
I I
I I
r 104 1
}
0.1 1A 10A- 100.0 200A PLUME TRAVEL DISTANCE (KILOMETERS)
This graph is reproduced from Reference 5 (Figure 8).
g Figure 8-9 Relative Deposition for 60-Meter Releases i
8-22 Gen. Rev. 13
FNP-0-M-011 O
.04 UhSTABLE (AA,C)
[ g 104 j
/ -
N '%
~~
2 l 'Rx g g' NEUTRAL 10) f f xs I f I N '
E ^
- A l '%. m E
e
/ / I N s. X j s 3
$ 104 ,
u : .
I I O l i i
l s !
2 J d / STABLE (E,F,0) 10-7 f NO DEPLET8ON I
y i
I l
}
,- /
0.1 1A 10A 100.0 200A PLUt0E TRAVEL DISTANCE (KILOMETERS)
This graph is reproduced from Reference 5 (Figure 9).
O Figure 8-10. Relative Deposition for 100-Meter (or Greater) Releases 8-23 Gen. Rev. 13
F.
FNP-0-M-011 CHAPTER 9
! METHODS AND PARAMETERS FOR CALCULATION OF
(_ GASEOUS EFFLUENT PATHWAY DOSE FACTORS, R.jp j i
9.1 INHALATION PATHWAY FACTOR 3
For the inhalation pathway, Raipj in (mrem /y) per (pCi/m ) is calculkted as follows (Reference 1, Section 5.3.1.1):
Rapj) = Kg (BR)a * (DFA)ajj (9.1) where: I Kg = the units conversion factor: 106 pCi/pC1.
(BR)a = the breathing rate of receptor age group a, in m3 /y, from Table 9-5. ;
(DFA)aij = the inhalation dose factor for receptor age group a, radionuclide 1, and organ j , in mrem /pci, from Table 9-7 through Table 9-10. l I
1 1
l 1
9-1 Gen. Rev. 13
r )
FNP-0-M-011 9.2 GROUND PLANE PATHWAY FACTOR l
For the ground plane external exposure pathway, Raipj in (m
- mrem /y) per 1
(pCi/s) is calculated as follows (Reference 1, Section 5.3.1.2):
-Af t 1~* (9.3)
Raip) " Kg = K2 * (SHF) (DFG) g) where l
l Kg = the units conversion factor: 100 pCi/pci.
-K2= the units conversion factor: 8760 h/y, (SHF) = the shielding factor due to structure (dimensionless).
The value used for (SHF) is 0.7, from (Reference 3, Table E-1E).
(DFG){j= the ground plane dose factor for radionuclide i and organ O j, in (mrem /h) per (pci/m 2- 3, 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. l 1 = the radioactive decay constant for radionuclide i, in l s -I. Values of Ag used in effluent calculations should be based on decey data from a recognized and current source, I such as Reference 15.
I t= the exposure time, in s. The value used for t is 4.73 x 10 0 s (= 15 y), from (Reference 1, Section 5.3.1.2).
O 9-2 Gen. Rev. 13 l
1 i
l
FNP-0-M-011 ;
9.3 GARDEN VEGETATION PATHWAY FACTOR For radionuclides other than tritium in the garden vegetation consumption ,
pathway, R aipj in (m 2mrem /y) per (pCi/s) is calculated as follows (Reference 1, Section 5.3.1.5) :
l Rapj) = Kg * - (DFL)ajj Yy (Af + Aw) ,
l (9.3)
-Xg tL -h l ' '
vL a fL*
- UaS fg* ihl l
where:
Kg = the unite conversion factor: 106 pCi/pC1.
1 r= the fraction of deposited activity retained on the edible parts of garden vegetation (dimensionless). The value used for r is 1.0 for radioiodines and 0.2 for l particulares, from (Reference 3, Table E-1).
) Yy =
the areal density (agricultural productivity) of growing leafy garden vegetation, in kg/m2 , from Table 9-1, 13- the radioactive decay constant for radionuclide i, in s -I. Values of 1; used in effluent calculations should be based on decay data from a recognized and current source, j such as Reference 15.
1, - the rate constant for removal of activity on leaf and plant. surfaces by weathering, in s-I, from Table 9-1.
l (DFL)aij = the ingestion dose factor for receptor age group a, !
radionuclide i, and organ j , in mrem /pci, from Table 9-11 through Table 9-14. -
i Ud= 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
b i'
FNP-0-M-011 Ug=
g the consumption rate of stored garden vegetation by a receptor in age group a, in kg/y, from Table 9-5.
ft= the fraction of the annual intake of fresh leafy garden vegetation that is grown locally (dimensionless), from Table 9-1.
f g= the fraction of the annual intake of stored garden vegetation that is grown locally (dimensionless), from Table 9-1.
tt= the average time between harvest of fresh leafy garden vegetation and its consumption, in s, from Table 9-1.
thy -
the average. time between harvest of stored garden vegetation and its consunption, -.in s, from Table 9-1.
For tritium in the garden vegetation consumption pathway, Raid ni (mrem /y) per (pci/m3 ) is calculated as follows (Reference li Section 5.3.1.5),
based on the concentration in air rather than deposition onto the ground:
Rj)=Kg*K3 ap * (DFL)4 . { Ug fL
- Ua5 fg )
- O '5
- l ;1 (9 4) where:
K3= the units conversion factor: 103 g/kg.
H= the absolute humidity of atmospheric air, in g/m3, from Table 9-1, 0.75 = the fraction of the mass of total garden vegetation that is water (dimensionless).
0.5 = .the ratio of the specific activity of tritium in garden vegetation water to that in atmospheric water (dimensionless).
O and other parameters are as defined above.
9-4 Gen. Rev. 13 j
FNP-0-M-011 Table 9-1. Miscellaneous Parameters for the Garden Vegetation Pathway The following parameter values are for use in calculating Raipj for the garden vegetation pathway only. The terms themselves are defined in Section 9.3.
Parameter Value Reference Yy 2.0 kg/m2 Ref. 3, Table E-15 1, 5.73 x 10~7 s *I Ref. 1, page 33 (14-day half-life) ft 1.0 Ref. 1, page 36 fg 0.76 Ref. 1, page 33 tt 8.6 x 104 s Ref. 3, Table E-15 (1 day) 6 thy 5 18
- go s Ref. 3, Table E-15 (60 days)
H 8 g/m3 Ref. 3
+ l l
O 9-5 Gen. Rev. 13
j 1
FNP-0-M-011 9.4 GRASS-COW-MILK PATHWAY FACTOR For radionuclides other than tritium in the grass-cow-mi3k pathway, R aipj in (m2mrem /y) per (pci/s) is calculated as follows (Reference 1, Section ;
5.3.1.3):
Raip) ~ Kg *
- Oy Uap
- Fg (DFL)ajj l
, , (9.5) ;
fp i (1 ~ pf sf)*
s+ .e "hl If i Y Ys P
l where:
Kg = the units conversion factor: 106 pCi/pci.
r= the fraction of deposited activity retained on the edible parts of vegetation (dimensionless). The value used for r is 1.0 for radiciodines and 0.2 for particulates, from (Reference 3, Table E-1).
1 = the radioactive decay constant for radionuclide i, in i s -I. Values of Ag used in effluent calculations should be based on decay data from a recognized and current source, I such as Reference 15.
1, = the rate constant for removal of activity on leaf and plant surfaces by weathering, in s-I, from Table 9-2.
Op = the ccw's consumption rate of feed, in kg/d, from Table 9-2.
U,p = the consumption rate of cow milk by a receptor in age
~
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.
9-6 Gen. Rev. 13
FNP-0-M-011 (DFL)aij = the ingestion dose factor for receptor age group a, radionuclide i, and organ j, in mrem /pci, from Table 9-11
(/ through Table 9-14.
l f p= the . fraction of the year that the cow is on pasture j (dimensionless), from Table 9-2. !
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.
l thm = the transport time from harvest of stored feed to its consumption by the cow, in s, from Table 9-2.
{ *g=
the transport time from consumption of feed by the cow, to consumption of milk by the receptor, in s, from Table 9-2.
For tritium in the grass-cow-milk pathway, Raipj in (mrem /y) per (pCi/m 3) is calculated as follows (Reference 1, Section 5. 3.1. 5) , based on the i concentration in air rather than deposition onto the ground: l i
i Rj]=Kg ap K3 07 U m Fg - (DFL)aij
- O 15
- l (9*6) where:
K 3= the units conversion factor: 103 g/kg. ,
H= the absolute humidity of atmospheric air, in g/m3 , from Table 9-2.
\
J 9-7 Gen. Rev. 13
FNP-0-M-011
.0.75 = the fraction of the mass of total vegetation that is water (dimensionless).
0.S = the ratio of the specific activity of tritium in vegetation water to that in atmospheric water 1
(dimensionless).
and other parameters are as defined above.
l 1.
l L
O 9-8 Gen. Rev. 13
FNP-0-M-011 Table 9-2. Miscellaneous Parameters for the Grass-Cow-Milk Pathway ammmmmmmmmmme The following parameter values are for use in calculating Raipj for the grass-cow-milk pathway only. The terms themselves are defined in Section 9.4.
Parameter Value Reference 1, 5.73 x 10*7 s *I Ref. 1, page 33 (14-day half-life)
Op 50 kg/d Ref. 3, Table E-3 f 1.0 Ref. 1, page 33 p
fs 1.0 Ref. 1, page 33 Yp 0.7 kg/m 2 Ref. 3, Table E-15 Ys 2.0 kg/m2 Ref. 3, Table E-15 tg 7.78 x 106 s Ref. 3, Table E-15 (90 days) tg 1.73 x 105 s Raf. 3, Table E-15 (2 days)
H 8 g/m3 Ref. 3 O
9-9 Gen. Rev. 13
i FNP-0-M-011 I 9.5 GRASS-GOAT-MILK PATHWAY FACTOR f*
( For radionuclides other than tritium in the grass-goat-milk pathway,.R 2 aid )
in (m mrem /y) per (pci/s) is calculated as follows (Reference 1, Section '
5.3.1.3):
Raip) = Kg Qy = Uap
- Fmi - (DFL)ajj fp f, (1 - fp f,) e
-hi thm _g ,
Yp Ys I
where:
I K] =
the units conversion factor: 106 pCi/pci.
r- the fraction of deposited activity retained on the edible parts of vegetation (dimensionless). The value used for I r is 1.0 for radioiodines and 0.2 for particulates, from (Reference 3, Table E-1).
Aj = the radioactive decay constant for radionuclide 1, in s -I. Values of Aj 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 on leaf and plant surf aces by weathering, in a-I, from Table 9-3.
Op = the goat's consumption rate of feed, in kg/d, from Table 9-3.
U the consumption rate of goat milk by a receptor in age 9-group a, in L/y, from Table 9-5.
Fg = the stable element transfer coefficient applicable to radionuclide i, for goat's milk, in d/L, from Table 9-6.
fi G
9-10 Gen. Rev. 13
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
(/ through Table 9-14 f p= the fraction of the year that the goat is on pasture (dimensionless), from Table 9-3.
fg= the fraction of the goat's feed that is pasture grass while the goat is on pasture (dimensionless), from Table 9-3.
Yp= the areal density (agricultural productivity) of growing pasture feed grass, in kg/m 2, from Table 9-3.
Yg- the areal density (agricultural productivity) of growing stored feed, in kg/m 2, from Table 9-3.
I thrn = the transport time from harvest of stored feed to its consumption by the goat, in s, 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. I For tritium' in the grass-goat-milk pathway, Rg jp in' (mrem /y) 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 ground:
Rapj; =
Kg
- K3
- Op Up F,pgl - (DFL)g 0.15
- O**
(9.8) 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 9-11 Gen. Rev. 13
I FNP-0-M-011 0.75 = the fraction' of the mass of total vegetation that is I water (dimensionleas).
0.5 = the- ratio of the specific activity of tritium in vegetation water to that in atmospheric water (dimensioniess).
and other parameters are as defined above.
O l
I l
l 4
9-12 Gen..Rev. 13
WP-0-M-011 Table 9-3. Miscellaneous Parameters for the Grass-Goat-Milk Pathway )
O V
The following parameter values are for use in calculating Rgp j for the grass-goat-milk pathway only. The terms themselves are defined in l Section 9.5.
i Parameter Value Reference 1, 5.73 x 10~7 s -I Ref. 1, page 33 (14-day half-life)
Op 6 kg/d Ref. 3, Table E-3 f 1.0 Ref. 1, page 33 p
f 1.0 Ref. 1, page 33 s
Yp 0.7 kg/m2 Ref. 3, Table E-15 Yg 2.0 kg/m 2 Ref. 3, Table E-15 tg 7.78 x 106 s Ref. 3, Table E-15 (90 days) tg 1.73 x 105 s Ref. 3, Table E-15 (2 days)
H 8 g/m3 Ref. 3
(
9-13 Gen. Rev. 13
FNP M
~9.6 GRASS-COW-MEAT PATHWAY FACTOR For radionuclides other than tritium in the grass-cow-meet pathway, Rjg g 2
in (m mrem /y) per (pci/s) is calculated as follows (Reference 1, Gection _
5.3.1.4):
i Rgp ; = i; *
- Qp . Liq
- Fp - (DFL)g)
(9.9) fp is (1 ~ fp i s)
- kgthm
, ,~h Iif Yp Ys where:
Kg = the units conversion factor: 106 pCi/pC1.
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 tO (Reference 3, Table E-1).
b lg - the radioactive decay constant for radionuclide 1, in s -I. Values of Ag 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 on leaf and plant surfaces by weathering, in s-I, from Table 9-4.
Op = the cow's consumption rate. of feed, in kg/d, from Table 9-4.
U the consumptien rate of meat by a receptor in age group 9=
a, in kg/y, from Table 9-5.
Fg=
f the stable element transfer coefficient applicable to radionuclide i, for meat, in d/kg, from Table 9-6.
O 9-14 Gen. Eev. 13
r PNP-0-M-011 (
i (DFI,)aij = the ingestion dose factor for receptor age group a,
' "" ' ""' ' ' ' " ' ' ' " ' ' ' ' ' ' " ' ' ' ' ' ' ' " ' ' ' ~ ' '
(J through Table 9-14.
f p= the fraction of the year that the cow is on pasture (dimensionless), from Table 9-4. l l
f, - the fraction of the cow's feed that is pasture grass w>ile the cow is on pasture (dimensionless), from Table 9-4.
l Yp= 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 stored feed, in kg/m 2 , from Table 9-4.
i thm = the transport tima, from harvest of stored f a -d to its consumption by the cow, in s, from Table L .
tg = the transport time from consumption of feed by the cow, to consumption of meat by the receptor, in s, from Table 9-4.
For tritium in the grass-cow-meat pathway, Raipj in (mrem /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 ground:
OU Rgp] = Kg = K3
- Op
- Uq
- Fp * (DFL) ajj
- 0. 75
- 1 (9.10) i where:
1 K3= the units conversion factor: 103 g/kg. ,
i l
H= the absolute humidity of atmospheric air, in g/m 3, from Table 9-4. I i
l 9-15 Gen. Rev. 13
I' PNP-0-M-011 0.75 = the fraction of the mass of total vegetation that is water (dimensionless).
0.5 = the ratio of the rpecific activity of tritium in vegetation water to that in atmospheric water (dimensionless).
and other parameters are as defined above.
l
> Gen. Rev. 13
FNP-0-M-011 Table 9-4. Misce.'.laneous Parameters for the Grass-Cow-Meat Pathway
)
v The following parameter values are for use in calculating Raipj for the grass-ec,w-meat pathway only. The terms themselves are defined in Section 9.6.
Parameter Value Reference 1, 5. 73 x 10-7 3-1 Ref. 1, page 33 (14-day half-life) op 50 kg/ Ref. 3, Table E-3 f 1.0 Ref. 1, pago 33 p
i f
s 1.0 Ref. 1, page 33 Yp 0.7 kg/m 2 Ref. 3, Table E-15 Ys 2.0 kg/M Ref. 3, Table E-15 tg 7.78 x 106 s Ref. 3, Table E-15 A '
(90 days) tg 1.73 x 106 s Ref. 3, Table E-15 (20 days)
H B g/m3 Ref. 3 o
O t
9-17 Gen. Rev. 13 L
FNP-0-M-011 Table 9-5. Individual Usage Factors O
Receptor Age Group Usage Factor Infant Chile: Teenager Adult Milk Consumption Rate, U"E 330 330 400 310 (L/y)
Meat Consumption Rate, U 8P o 41 65 110 (kg/y)
Fresh Leafy Garden Vegetation Consumption Rate, Ual 0 26 42 64 (kg/y)
Stored Garden Vegetation Consumption Rate, UaS 0 520 630 520 (kg/y)
Breathing Rate, (BR)A '
1400 3700 -2000 8000 (m3 /y)
O All values are from Reference 3, Tab'e i E-5.
O 9-18 Gen. Rev. 13
FNP-0-M-011 Table 9-6. Stable Element Transfer Data h
J Cow Milk Goat Milk Meat Element p m (d/L)
- Fm (d/L) + Fg (d/kri)
- 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 (j]
f'~ Tc Ru Rh 2.5 E-02 1.0 E-06 1.0 E-02 2.5 E-02 1.0 E-06 1.0 E-02 4.0 E-01 4.0 E-01 1.5 E-03 Ag 5.0 E-02 5.0 E-02 1.7 E-02 Sb 1.5 E-03 1.5 E-03 4.0 E-03 Te 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 4.0 E-03 Ba 4.0 E-04 4.0 E-04 3.2 E-03 La 5.0 E-06 5.0 E-06 2.0 E-04 Ce 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 3.3 E-03 W 5.0 E-04 5.0 E-04 1.3 E-03 Np 5.0 E-06 5.0 E-06 2.0 E-04 Values from Reference 3 (Table E-1) except as followse 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 mille, 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 milk.
O -
9-19 Gen. Rev. 13
FNP-0-M-011 Table 9-7. Inhalation Dose Factors for the Infant Age Group
, \
V Nuclide Bone Liver T.Bady 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 1.5BE-10 No Data 7.86E-10 8.95E-06 5.12E-05 (
1 Fe-55 1.41E-05 8.39E-06 2.38E-06 No Data No Data 6.21E-05 7.82E-07 l 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 3.5BE-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 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 Br-83 No Data No Data 2.72E-07 No Data No Data No Data No Data
%r-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 5.30E-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 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. 2 4 E- 05 All values are in (mrem /pci 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 O
9-20 Gen. Rev. 13
FNP-0-M-011 Table S-7 (contd). Inhalation Dose Factors for the Infant Age Group i
-? =
Nuclide Bone Liver T. Body Thyroid Kidney Lang GI-LLI 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 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 1.19E-04 Zr-95 8.24E-05 1.99E-05 1.45E-05 No Data 2.22E-05 1.2SE-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-03 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 4.650-14 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 '. 15E-05 Ru-105 8.74E-10 No Data 2.93E-10 No Data 6.42E-10 1.12E-05 s.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 Data No Data Ag-110ta 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 No Data No Data 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 l Te-127m 1.19E-05 4.93E-06 1.48E-06 3.48E-06 2.68E-05 9.37E-04 1.95E-05 I
1 Te-127 1.59E-09 6.81E-10 3.49E-10 1.32E-09 3.47E-09 7.39E-06 1.74E-05 l
Te-129m 1.01E-05 4.35E-06 1.59E-06 3.91E-06 2.27E-05 1.20E-03 4.93E-05 Te-129 5,63E-11 2.48E-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"D5 Te-131 1.24E-11 5.87E-12 3.57E-12 1.13E-11 2.85E-11 1.473-06 5.87E-06 1
I 9-21 Gen. Rev. 13
FNP-0-M-011 Table 9-7 (contd). Inhalation Dose Factors for the Infant Age Group Nuclide Bone Liver T. Body Thyrofd Kidney Lung GI-LLI l I
Te-132 2.66E-07 1.69E-07 1.26E-07 1.99E-07 7.39E-07 2.43E-04 3.15E-05 j I-130 l 4.54E-06 9.91E-06 3.98E-06 1.14E-03 1.09E-05 No Data 1.42E-06 l
I-131 2.71E-05 3.17E-05 1. 4 0 L'- 0 5 1.06E-02 3.70E-05 No Data 7.56E-07 1
l 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-05 1.49E-06 No Data 9.21E-07 I-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-0(.I 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 2.91E-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 f% Ba-140 4.00E-05 4.00E-08 2.07E-06 No Data (d 9.59E-09 1.14E-03 2."4E-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 Ce-141 1.98E-05 1.19E-05 1.42E-06 No Data 3.75E-06 3.69E-04 1.54E-05 Ce-143 2.09E-07 1.38E-07 1.58E-08 No Data 4.03E-08 8.30E-05 3.55E-05 Ce-144 2.28E-03 8.65E-04 1.26E-04 No Data 3.84E-04 7.03E-03 1.06E-04 Pr-143 1.00E-05 3.74E-06 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 5.81E-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 s-t 9-22 Gen. Rev. 13
FNP-0-M-011 Table 9-8. Inhalation Dose Factors for the Child Age Group (3
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-07 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 Zn-65 1.15E-05 s, *
,E-05 1.90E-05 No Data 1.s3E-05 2.69E-04 4.41E-06 Zn-69 1.81E-11 2.61U-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 I I
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 9.28E-05 Sr-91 3.28E-08 No Data 1.24E-09 No Data No Data 1.44E-05 4.70E.D5 All values are in (mrem /pci inhaled). They are obtained from ,
Reference 3 (Table E-3). Neither Reference 2 nor Reference 3 !
contains data for Rh-105, Sb-124, or Sb-125. I f
v . . .
)
I 9-23 Gen. Rev. 13 i
r l
FUP-0-H-011.
Table 9-8 (contd). Inhalation Dose Factors for the Child Age Group
/~T U
Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-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-9?m 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 Data 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 1.00E-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 1.00E-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 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 3.0BE-06 2.47E-06 No Data 5.74E-06 1.4BE-03 2.71E-05 Sb-124 No Data No Data No Data No Data Ko Data No Data No Data Sb-125 No Data No Data No Data No Data No Data No Data No Data 7c-125m 1.82E-06 6.29E-07 2.47E-07 5.20E-07 Ho Data 1.29E-04 9.13E-06 Te-127m 6.72E-06 2.31E-06 8.16E-07 1.64E-06 1.72E-05 4.00E-04 1.93E-05 Te-127 7.49E-10 2.57E-10 1.65E-10 5.30E-10 1.91E-09 2.71E-06 1.52E-05 Te-129m 5.19E-06 1.85E-06 8.22E-07 1.71E-06 1.36E-05 4.76E-04 4.91E-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-08 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 9-24 Gen. Rev. 13
FNP-0-M-011 Table 9-8 (contd). Inhalation Dose Factors for the Child Age Group t
b .
Nuclide Bone . Liver T. Body Thyroid Kidney Lung GI-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 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 4.48E-06 5.49E-06 2.08E-06 1.04E-03 9.13E-06 No Data 1.48E-06 I-134 3.17E-07 5.84E-07 2.69E-07 1.37E-05 8.92E-07 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 Cs-134 1.76E-04 2.74E-04 6.07E-05 No Data 8.93E-05 3.27E-05 1.04E-06 Cs-136 1.76E-05 4.62E-05 3.14E-05 No Data 2.58E-05 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 S.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
's_ / 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 l
Ce-141 1.06E-05 5.28E-06 7.83E-07 No Data 2.31E-06 1.47E-04 1.53E-05 Ce-143 9.89E-08 5.37E-08 7.77E-09 No Data 2.26E-09 3.12E-05 3.44E-05 Ce-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 2.22"-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 O
9-25 Gen. Rev. 13
FNP-0-M-011 Table 9-9. Inhalation Dose Factors for the Teenager Age Group
(
t i Nuclide Bone T. Body Liver 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 l 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 1.91E-04 No Data 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 Dcta 1.09E-03 3.24E-05 Ni-63 7.25E-05 5.43E-06 2.47E-06 No Data No Data 3.84E-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
/N Cu-64 No Data 2.54E-10 1.06E-10 No Data 8.01E-10 1.39E-06 7.68E-06 Zn-65 4.82E-06 1.67E-05 7.80E-06 No Data 1.08E-05 1.55E-04 5.83E-06 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 Date 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 1.0SE-05 No Data No Data No Data 2.21E-06 Rb-88 No Data 6.82E-08 3.40E-08 No Data No Data No Data 3.65E-15 Rb-89 No Data 4.40E-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 1.35E-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 3.2tE-05 All values are in (mrem /pci inhaled) . They are obtained f::om Reference 3 (Table E-8). Neither Reference 2 nor Reference 3 contains data for Rh-105, Sb-124, or Sb-125.
p 97 26 Gen. Rev. 13
FNP-0-M-011 Table 9-9 (contd). Inhalation Dose Factors t'or the Teenager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 1.19E-09 No Data 5.08E-11 No Data No Data 3.43E-06 1.49E-05 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 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 Data 5.15E-09 1.62E-05 7.88E-05 Nb-95 2.32E-06 1.29E-06 7.0BE-07 No Data 1.25E-06 9.39E-05 1.21E-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 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 Data 9.29E-07 9.79E-05 1.36E-05 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 Rh-105 No Data No Data No Data No Data No Data No Data No Data Ag-110m 1.73E-06 1.64E-06 9.99E-07 No Data 3.13E-06 6.44E-04 3.41E-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 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 1.14E-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 6.48E-12 3.32E-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
%*e-131 1. 97E-12 l 1. 04E-12 6.30E-13 1.55E-12 7.72E-12 2.92E-07 1.89E-09 '
O V
9-27 Gen. Rev. 13
f FNP-0-M-011 Table 9-9 (contd). Inhalation Dose Factors for the Teenager Age Group U
Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 4.50E-08 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 1.83E-03 1.05E-05 No Datt 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 Cs-134 6.2BE-05 1.41E-04 6.86E-05 No Data 4.69E-05 1.83E-05 1.22E-06 i Cs-136 6.44E-06 2.42E-05 1.71E-05 No Data 1.38E-05 2.22E-06 1.36E-06 Cs-137 8.38E-05 1.06E-04 3.89E-05 No Data 3.80E-05 1.51E-05 1.06E-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 No Data 2.85E-09 2.54E-04 2.8EE-05
(
N Ba-141 1.78E-11 1.32E-14 5.93E-13 No Data 1.23E-14 4.11E-07 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 Ce-141 3.55E-06 2.37E-06 2.71E-07 No Data 1.11E-06 7.67E-05 1.58E-05 Ce-143 3.32E-08 2.42E-06 2.70E-09 No Data 1.08E-08 1.63E-05 3.19E-05 Ce-144 6.11E-04 2.53E-04 3.28E-05 No Data 1.51E-04 1.67E-03 1.08E-04 Pr-143 1.67E-06 6.64E-07 8.28E-08 No Data 3.86E-07 6.04E-05 2.67E-05 l
Pr-144 5.37E-12 2.20E-12 2.72E-13 No Data 1.26E-12 2.19E-07 2.94E-14 Nd-147 9.83E-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
O 9-28 Gen. Rev. 13
F' FNP-0-M-011 Table 9-10. Inhalation Dose Factors for the Adult Age Group V Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 1.58E-07 1.5BE-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 l 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 j l
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
(~~T Cu-64 No Data 1.63E-10 7.69E-11 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-OB No Data No Data No Data 2.90E-08 Br-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 .19E- 05 All values are in (mrem /pci inhaled) . They are obtained from Reference 3 (Tabic E-7) , except as follows : Reference 2 (Table C-1) for Rh-105, Sb-124, and Sb-125.
G 9*)
a Gen. Rev. 13
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 Sr-92 8.43E-10 No Data 3.64E-11 No Data No Data 2.06E-06 5.30E-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 l Y-92 1.29E-09 No Data 3.77E-11 No Data No Data 1.96E-06 9.19E-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 Mo-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 B.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-10$ 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 e
Te-127 1.75E-10 8.03E-11 3.87E-11 1.32E-10 6.37E-10 0.14E-07 7.17E-06 Te-129m 1.22E-06 5.84E-07 1.98E-07 4.30E-07 4.b7E-06 1.45E-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 6.75E-05 Te-131 1.39E-12 7.44E-13 4.49E-13 1.17E-12 5.46E-12 1.74E-07 2.30E-09 O
9-30 Gen. Rev. 13
FNP-0-M-011 Table 9-10 (contd). Inhalation Dose Factors for the Adult Age Group huclide 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 6.37E-05 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 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 1.11E-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 3 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 2.7BE-05 9.40E-06 1.05E-06 Cs-139 4.14E-08 7.76E-08 4.05E-08 No Data 6.00E-08 6.07E-09 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 Ba-140 4.88E-06 6.13E-09 3.21E-07 No Data 2.09E-09 1.59E-04 2.73E-05' I i V 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 La-142 8.54E-11 3.88E-11 9.65E-12 No Data No Data 7.91E-07 2.64E-07 .
Ce-141 2.49E-06 1.69E-06 1.91E-07 l No Data 7.83E-07 4.52E-05 1.50E-05 Ce-143 2.33E-08 1.72E-08 1.91E-09 No Data 7.60E-09 9.97E-06 2.83E-05 Ce-144 4.29E-04 1.79E-04 2.30E-05 No Data 1.06E-04 9.72E-04 1.02E-04 Pr-143 1.17E-06 4.69E-07 5.80E-08 No Data 2.70E-07 3.51E-05 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 7.62E-07 4.56E-08 No Data 4.45E-07 2.76E-05 2.16E-05 W-107 1.06E-09 8.85E-10 3.10E-10 No Data No Data 3.63E-06 1.94E-05 Np-239 2.87E-08 2.82E-09 1.55E-09 No Data 8.75E-09 4.70E-06 1.49E-05
(~'N -
l l
9-31 Gen. Rev. 13 l l
r 1
FNP-0-M-011 Table 9-11. Ingestion Dose Factors for the Infant Age Group iO !
Nuclide Done Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 3.08E-07 3.0BE-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 1.41R-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 8.9BE-06 2.40E-06 No Data 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 1.00E-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 i Ni-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 i I
Zn-69 9.33E-08 1.68E-07 1.25E-08 No Data 6.98E-08 No Data 1.37E-05 l Br-63 No Data No Data 3.63E-07 No Data No Data No Data No Data Er-84 No Data No Data 3.82E-07 No Data No Data No Data No Data
]
i 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 5.00E-05 No Data 1.81E-06 No Data No Data No Data 5.12E-05 All values are in (mrem /pCi 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.
V 9-32 Gen. Rev. 13
m l
l FNP-0-M-011 Table 9-11 (contd). Ingestion Dose Fact. ors for the Infant Age Group O Nuclide Bone Liver T. Body Thyroid Eidney Lung GI-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 8.1CE-10 No Data 2.76E-11 No Data No Data No Data 2.70E-06 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 ,
1 Zr-95 2.06E-07 5.02E-08 3.56E-08 No Data 5.41E-08 No Data 2.50E-05 Zr-97 1.tBE-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 j 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 Data 4.26E-OS 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 1.48E-06 No Data 4.95E-07 No Data 3.08E-06 No Data 1.80E-05 Ru-105 1.36E-07 No Data 4.5BE-08 No Data 1.00E-06 No Data 5.41E-05
%./
Ru-106 2.41E-05 No Dats 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 aca 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 .00E-06 3.35E-07 2.15E-07 8.14E-07 2.44E-06 No Data 2.10E-05 Te-129m 1 00E-04 3.43E-05 1.54E-05 3.84E-05 2.50E-04 No Data 5.97E-05 Te-129 2.94E-07 9.79E-08 6.63E-08 2.38E-07 7.07E-07 No Data 2.27E-05 Te-131m 1.52E-05 6.12E-06 5.05E-06 1.24E-05 4.21E-05 No Data 1.072-04 Te-131 1.76E-07 6.50E-08 4.94E-08 1.57E-07 4.50E-07 No Data 7.11E-06 V
9-33 Gen. Rev. 13
r_
FNP-0-M-011 Table 3-11 (contd). Ingestion Dose Factors for the Infant Age Group p
- Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Te-132 2.08E-05 1.03E-05 9.61E-06 1.52E-05 6.44E-05 Mo 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 1.20E-06 1.58E-04 3.76E-06 No Data 2.73E-06 I-133 1.25E-05 1.32E-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 No 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 l
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 La-142 1.10E-09 4.04E-10 9.67E-11 No Data No Data No Data 6.86E-05 Ce-141 7.87E-08 4.80E-08 5.65E-09 No Data 1.48E-08 No Data 2.48E-05 Ce-343 1.48E-08 9.82E-06 1.12E-09 No Data 2.86E-09 No Data 5.73E-05 ,
i Ce-144 2.98E-06 1.22E-06 1.67E-07 No Data 4.93E-07 No Data 1.71E-04 I Pr-143- 8.13E-08 3.04E-08 4.03E-09 No Data 1.13E-08 No Data 4.29E-05 l
i Pr-144 2.74E-10 1.06E-10 1.38E-11 No Data 3.84E-11 No Data 4.93E-06 l
Nd-147 5.53E-08 5.68E-08 3.48E-09 No Data 2.19E-08 No Data 3.602-05 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 No Data 2.87E-05 i
l
- O 9-34 Gen. Rev. 13 i
{
FNP-0-M-011 Table 9-12. Ingestion Dose Factors for the Child Age Group O, Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI H-3 No Data 2.03E-07 2.03E-07 2.03E-07 2.03E-07 2.03E-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 8.2SE-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 Mn-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 Data No Data No Data 2.56E-05
/~'~ Cu-64 No Data 2.45E-07 1.48E-07 No Data 5.92E-07 No Data 1.15E-05 Zn-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 No Data No Data No Data 4.31E-06 Rb-88 No Data 1.90E-07 1.32E-07 No Data No Data No Data 9.32E-09 Rb-89 No Data 1.17E-07 1.04E-07 No 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 5.3aE-05 All values are in (mrem /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-125.
() z
~~
9-35 Gen. Rev. 13
FNP-0-M-011 Table 9-12 (contd). Ingestion Dose Factors for the Child Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI Sr-92 9.03E-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 7.482-07 Y-91 6.02E-07 No Data 1.61E-08 No Data No Data No Data 8.02E-05 Y-92 3.60E 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 Or-95 1.16E-07 2.55E-08 2.27E-08 No Data 1
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 O Ru-105 6.45E-08 No Data 2.34E-08 No Data 5.67E-07 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 7.78E-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 1.363-05 7.56E-06 1.57E-05 1.43E-04 No Data 5.94E-05 Te-129 1.34E-07 3.74E-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 1.0rE-04 Te-331 8.30E-08 2.53E-08 2.47E-08 6.35E-08 2.51E-07 No Data 4.36E-07 m . .
9-36 Gen. Rev. 13
FNP-0-M-011 Table 9-12 (contd). Ingestion Dose Factors for the Child Age Group O
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 2 92E-05 No Data 2.95E-06 I-134 4.19E-07 7.78E-07 3.58E-07 1.79E-05 3 LYL-06 No Data 5.1GE-07 I-135 1.75E-06 3.15E-06 1.49E-06 2.79E-04 4.83E-06 No Data 2.40E-06 Cs-134 2.34E-04 3.84E-04 8.10E-05 No Data 7.19E-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-13B 2.29E-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-Ob Ba-140 8.31E-05 7.28E-08 4.85E-06 No Data 2.37E-08 4.34E-08 4.21E-05 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 6.292-11 4.88E-09 No Data 5.09E-11 3.70E-11 1.14E-09 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 Ce-141 3.97E-08 1.98E-08 2.94E-09 No Data 8.68E-09 No Data 2.47E-05 Ce-143 6.99E-09 3.79E-06 5.49E-10 No Data 1.59E-09 No Data 5.55E-05 Ce-144 2.08E-06 6.52E-07 1.11E-07 No Data 3.61E-07 No Data 1.70E-04 Pr-143 5.93E-08 1.18E-08 1.95E-09 No Data 6.39E-09 No Data 4.24E-05 Pr-144 1.29E-10 3.99E-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 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
FNP-0-M-011 Table 9-13. Ingestion Dose Factors for the Teenager Age Group p
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 1.06E-07 C-14 4.06E-06 8.12E-07 8.12E-07 8.12E-07 8.12E-07 8.12E-07 8.12E-07 Na-24 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-06 2.30E-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 7.89E-10 5.14E-09 6.05E-07 Mn-54 No Data 5.90E-06 1.17E-06 No Data 1.76E-06 No Data 1.21E-05 Mn-SC 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-06 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 1.25E-05 6.00E-06 No Data No Data No Data 1.99E-06 Ni-65 7.49E-07 9.57E-08 4.36E-08 No Data No Data No Data 5.19E-06
, Cu-64 No Data 2.15E-07 5.41E-08 No Data 2.91E-07 No Data 8.92E-06 J
t
\
Zn-65 5.76E-06 2.00E-05 9.33E-06 No Data 1.28E-05 No Data B.47E-06 Zn-69 1.47E-08 2.80E-08 1.96E-09 No Data 1.83E-08 No Data 5.16E-08 l Br-83 No Data No Data 5.74E-08 No Data No Data No Data No Data Br-04 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 5.50E-08 3.89E-08 No Data No Data 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 8.30E-03 No Data 2.05E-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.
A I 9-38 G2n. Rev. 13
FNP-0-M-011 Table 9-13 (contd). Ingestion Dose Factors for the Teenager Age Group Nuclide Bone Liver T. Body 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 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 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 No Data 1.38E-08 5.14E-10 6.08E-07 Te-101 3.60E-10 5.12E-10 5.03E-09 No Data 9.26E-09 3.12E-10 8.75E-17 Ru-103 2.55E-07 No Data 1.09E-07 No Data 8.99E-07 No Data 2.13E-05 1
Ru-105 2.18E-08 No Data 8.46E-09 No Data 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 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 Te-129m 1.63E-05 6.05E-06 2.5BE-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 Te-131m 2.44E-06 1.17E-06 9.76E-07 1.76E-06 1.22E-05 No Data 9.33E-05 Te-131 2.79E-08 1.15E-08 8.72E-09 2.15E-08 1.22E-07 No Data 2.29E-09 0
9-39 Gen. Rev. 13
r i FNP-0-M-011 Table 9-13 (contd). Ingestion Dose Factors for the Teenager Age Group Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI l l
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 . 9 8 E- 06,1.19E- 06 2.43E-04 4.59E-06 No Data 2.29E-06 I-131 5.85E-06 8.19E-06 4.40E-06 2.39E-03 1.41E-05 No Data 1.62E-06 l 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 2.5BE-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 3.3BE-05 2.27E-05 No Data 1.84E-05 2.90E-06 2.72E-06 Cs-137 1.12E-04 1.49E-04 5.19E-05 No Data 5.07E-05 1.97E-05 2.12E-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 Da-141 6.71E-08 5.01E-11 2.24E-09 No Data 4.65E-1? 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 Ce-141 1.33E-08 8.8BE-09 1.02E-09 No Data 4.18E-09 No Data 2.54E-05 Ce-143 2.35E-09 1.71E-06 1.91E-10 No Data 7.67E-10 No Data 5.14E-05 Ce-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 5.23E-09 6.52E-10 No Data 3.04E-09 No Data 4.31E-05 Pr-144 4.30E-11 1.76E-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 5.99E-09 No Data 3.6BE-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 NJ 9-40 Gen. Rev. 13
FNP-0-M-011 Table 9-14. Ingestion Dose Factors for the Adult Age Group i
D 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-3; 1.93E-04 1.20E-05 7.46E-06 No Data No Data No Data 2.17E-05 Cr-51 No Data No Data 2.66E-09 1.59E-09 5.86E-10 3.53E-09 6.69E-07 l l
l 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 3.91E-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 j Co-60 No Data 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-05 Ni-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 O 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 Br-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 3.21E-08 No Data No Data No Data 8.36E-19 Rb-89 No Data 4. ole-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.G7E-06 No Data 2.29E-07 No Data No Data No Data 2.7aE-05 All values are in (mrem /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.
i 1 L) 9-41 Gen. Rev. 13 l
l
i I
FNP-0-M-011 ,
Table 9-14 (contd). Ingestion Dose Factors for the Adult Age Group i
\.~) Nuclide Bone Liver T. Body Thyroid Kidney Lung GI-LLI l 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 2.5BE-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 i
Y-91 1.41E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-05 1
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 Me-99 No Data 4.31E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 Tc-99m 2.47F-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 6.0eE-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 Te-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.36 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 O
9-42 Gen. Rev. 13
FNP-0-M-011 Table 9-14 (contd). Ingestion Dose Factors for the Adult Age Group f~
(
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 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 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 1
i 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 O 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 Ce-141 9.36E-09 6.33E-09 7.18E-10 No Data 2.94E-09 No Data 2.42E-05 Ce-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 3.01E-11 1.25E-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 1.03E-07 8.61E- 8 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-03 D
\
9-43 Gen. Rev. 13
FNP-0-M-011 Table 9-15. External Dose Factors for Standing on Contaminated Ground v Nuclide T. Body Skin Nuclide T. Body 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 2.502-08 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-54 5.80E-09 6.80E-09 Y-92 1.60E-09 1.90E-09 Mn-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-08 Mo-99 1.90E-09 2.20E-09 Ni-63 0.00 0.00 Tc-99m 9.60E-10 1.10E-09 Ni-65 3.70E-09 4.30E-09 Tc-101 2.70E-09 3.00E-09 Cu-64 1.50E-09 1.70E-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 Sr-90 0.00 0.00 Te-129m 7.70E-10 9.00E-10 ,
All values are in (mrem /h) per (pci/m )2 . They are obtained from Reference 3 (Table E-6), except as follows: Reference 2 (Table A-7) for Rh-105, Sb-124, and Sb-125.
1 I
9-44 Gen. Rev. 13 '
FNP-0-M-011 Table 9-15 (contd). External Dose Factors for Standing on Contaminated Ground I
/
N-Tl
\
Nuclide T. Body Skin 1
Te-129 7.10E-10 8.40E-10 l
Te-131m 8.40E-09 9.90E-09 Te-131 2.20E-09 2.60E-06 i
Te-132 1 70E-09 2.00E-09 I-130 1.40E-08 1.70E-08 I-131 2.80E-09 3.40E-09 I-132 1.70E-08 2.00E-08 I-133 3.70E-09 4.50E-09 l 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 k/ Ba-140 2.10E-09 2.40E-09 l
Ba-141 4.30E-09 4.90E-09 Ba-142 7.90E-09 9.00E-09 l La-140 1.50E-08 1.70E-08
]
La-142 1.50E-08 1.80E-08 l
Ce-141 5.50E-10 6.20E-10 Ce-143 2.20E-09 2.50E-09 Ce-144 3.20E-10 3.70E-10 Pr-143 0.00 0.00 Pr-144 2.00E-10 2.30E-10 Nd-147 1.00E-09 1.20E-09 W-187 3.10E-09 3.60E-09 Np-239 9.50E-10 1.10E-09 -
N 9-45 Gen. Rev. 13
FNP-0-M-011 CHAPTER 10 DEFINITIONS OF EFFLUENT CONTROL TERMS The terms defined in this chapter are used in the presentation of the above chapters. These terms are shown in all capital letters to indicate that they are specifically defined.
10.1 TERMS SPECIFIC TO THE ODCM The following terms are used in the ODCM, but are not found in the Technical Specifications l
BATCH RELEASE A BATCH RELEASE is the discharge of wastes of a discrete volume. Prior to l sampling for analyses, each liquid batch shall be isolated and then thoroughly mixed by a method described in the ODCM to assure representative sampling.
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 results I f
in a specimen that is representative of the wastes released. Prior to analyses, all liquid samples that are to be aliquotted for a COMPOSITE 1
SAMPLE shall be mixed thoroughly, in order for the COMPOSITE SAMPLE to be representative of the effluent release.
1 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 the applicable release pathway may be used.
CONTINUOUS RELEASE A CONTINUOUS RELEASE is the discharge of wastes of a non-discrete volume, e.g., from a volume within a system that has an input flow during the continuous release. To be representative of the quantities and concen-trations of radioactive materials in CONTINUOUS RELEASES of liquid effluents, samples shall be collected in proportion to the rate of flow of the effluent stream, or to the quantity of waste discharged. ~
GASEOUS RADWASTE TREATMENT SYSTEM A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant C\ system offgases from the primary system and providing for delay or holdup 10-1 Gen. Rev. 13
]
FNP-0-M-011 for.the purpose of reducing the total radioactivity prior to release to the environment. This system consists of at least one gas compressor, O waste gas decay tanks, and associated components providing for treatment flow and functional control.
LIOUID RADWASTE TREA'INENT SYSTEM
. A LIQUID RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive materials in liquid effluents by systematic collection, retention, and processing through filtration, evaporation, separation and/or ion exchange treatment. This system consists of ' at least one collection tank, one evaporator or demineralizer system, one post-treatment tank and associated components providing for treatment flow and functional control.
l MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMg For the purposes of the ODCM, MAJOR CHANGES TO RADIOACTIVE WASTE TREATMEN'l SYSTEMS include the following changes to such systems:
(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 Commission staff's Safety Evaluation Report (SER) (e.g.,
deletion of evaporators and installation of domineralizers);
(2) 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; (3) Changes in system design which may inva3idate the accident analysis as described in the SER (e.g. , changes in tank capacity that would alter the curies released); or (4) Changes in system design that could potentially result in a significant increase in occupational exposure of operating (e.g., use of temporary equipment without adequate personnel !
shielding provisions).
MEMBER (S) OF THE PUBLICI ~
j A MEMBER OF THE PUBLIC means any individual except when that individual is l l
1 O 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 Rev. 16
FNP-0-M-011 receiving an occupational dose I.
This category may include persons who use l
,q portions of the site for recreational, occupational, or other purposes not 4
) 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 i material in a sample that will yield a net 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 l 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 Rb - + i Cs
% C's ty, NDC =
E V 2.22 x 100 Y e~A A#
O where:
MDC = the a priord MINIMUM DETECTABLE CONCENTRATION (pCi per unit mass oz volume).
2.71 = the square of the standard normal variate (1.645) for the i 95 percent confidence level (Ref. 12,Section II.D).
3.29 = Two times the standard nomal variate (1.645) for the l 95 percent confidence level (Ref. 12,Section II.C) .
Rh= the background counting rate, or the counting rate of a blank sample, as appropriate (counts per minute). '
ts= the length of tha sample counting period (minutes).
tb= the length of the background counting period (minutes).
E= the counting efficiency (counts per disintegration)
V= the sample size (units of mass or volume).
2.22 x 10 6 = the number of disintegrations per minute per pci.
Y= the fractional radiochemical yield, when applicable.
A l Except as delineated in other parts of 10 CFR Chapter I.
10-3 Rev. 16
r-FNP-0-M-011 1- the radioactive decay constant for the given radionuclide (h-I). Values of 1 used in effluent calculations should b)
\ be based on decay data from a recognized and current source, such as Reference 15.
At - for effluent samples, the elapsed time between the midpoint of sample collection and the time of counting (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 EM7rTEES.
The PRINCIPAL GAMMA EMITTERS for which the MINIMUM DETECTABLE CONCENTRATION (MDC) limit applies include exclusively the following radio-nuclides:
- For liquid radioactive effluents: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an MDC of 5 x 10-6 peij,n, e For gasecus radioactive effluents: 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, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144.
- 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). I SITE BOUND E For the purpose of effluent controls defined in the ODCM, the SITE BOUNDARY shall be as shown in Figure 10-1. )
U 10-4 Rev. 16 l
FNP-0-M-011 UNRESTRICTED AREA The UNRESTRICTED AREA shall be any area access to which is neither limited (n) a 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.
10.2 TERMS DEFINED IN THE TECHNICAL SPECIFICATIONS The following terms are defined in the Technical Specifications, Section 1.0.
Because they are used throughout 1-he Limits of Ope, ation sections of the ODCM, they are presented here for convenience. In the event of discrepancies between the definitions below and the se in the Technical Specifications, t.4e Technical Specification definitions shall take precedence.
ACTION (S)
An ACTION shall be that part of a control that prescribes remedial measures required under designated conditions.
CHANNEL 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 (n)
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channel including the sensors and alarm, interlock, and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps, such that the entire channel is calibrated.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:
- 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.
e Bistable Channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm, and/or trip functions.
(mm) 10-5 Rev. 16 l
FNP-0-M-011 DOSE EQUIVAt8NT 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.
l'REQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of surveillance requirements shall correspond to the intervals defined below, with a maximum allowable extension not to exceed 25% of the sur':d11ance interval.
NOTATIOl' FREOUENCY S (Once per suift) At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D (Daily) At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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 p r 184 days.
R (Refueling) At least once per 18 months.
S/U (Startup) Prior to each reactor startup.
NA Not applicable.
P (Prior) Completed prior to each release.
MODE for OPERATIONAL MODE)
An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Section 1.0 of the Technical Specifications.
OPERABLE for OPERABILITY)
OPERABILITY exists when a system, subsystem, train, component or device is i
capable of performing its specified functior 's), and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are alac capable of performing their related support function (s).
~
i RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2652 Mwt.
10-6 Rev. 16 l 1
PNP-0-M-011 SOURCE CHECK A SOURCE CHECK shall,be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.
1EERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
VENTILATION wrum_tST TREA'DGENT SYAM The VENTILATION EXHAUST TREATMENT - SYSTEM is any system designed and installed to reduce ' gaseous radiciodine or radioacti.a material in particulate form in effluents by passing ventilation or vent exhaust gases
- throutjh charcoal adsorbers and/or HEPA filters for ths purpose of removing lodines ' or particulates from the gaseous exhaust strenm prior to the I
release to the environment (such a system is not considered to have any effect on any noble gas effluents). Engineered Safety Feature (ESP) 1 atmospheric cleanup systems are not considered to be VENTILATION EXHAUST l' 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.
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v 10-7 Rev. 16 l l
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PROPERTY LINE Figure 10-1. Site Map for Effluent Controls 10-8 Gen. Rev. 13
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