ML20059F875
| ML20059F875 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 02/05/1988 |
| From: | NIAGARA MOHAWK POWER CORP. |
| To: | |
| Shared Package | |
| ML17056A978 | List: |
| References | |
| PROC-880205, NUDOCS 9009120041 | |
| Download: ML20059F875 (135) | |
Text
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As N Date Description ages (Itselee ah neiere/reseen er seneras ebensee' O W u 4 1o .ua n und z aJ 4t//i aAes<.gp cash /a.J (w <$o.,. As. SAM :4.J?O aD CLwgea 4 do' e efted Auo-E at d w s,..it.e h w4n. n iIwia ,uveen, ' R SE.4 'T*ME o.1TT14CW44 4~O 4 hE.ZOIW 4lt
. MODIFICATION nFT.ATED CHANCRM YRM F1 No Moni t'nwTanf L No. 7-
= m pena 0ic sEVEW WITN NO Q4ANGES (Prd Rev. NC). USE T>E L.AST P0009E0 REVISION NUMBER AND CONTWEE SEVEW PQOCESS.
INTRAD SCIPLIBARY gIET (minimos et omeg rgeired) Ctl Aro Assal ha.M M. & T. %l,.1_ .2 3 - e r-0 < h# LLs.'l 0 46 . h H o (.!.l2% M ,v/s-M - g,DgIPLIBARY p (if not reestred.gigg isatificationgpt) A)o4. Ea1W - No e44eth im Gkus - efisc.iolds1. - b'a.tmp4 . " i IF NOT IN CONCU53ENCE. 00 NOT SleN SUT RETURN 00CUPENT TO THE AUTHOR WITH COMPWNTS Routed to Quality f reeleur: Yes D. No 91 No, r-AzJ==4.or O 6 et SeLfe Q. A. Repressatative Date & comments are attenbed. D. no e.toA g.u.,g g , g ,o.N. <RN . =
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A.LA.R.A. Representative aa*=
& c-eanea are attached. O.
>>>>>>>>>>>>>>>n>>>>>>>>>>>> Beste to AUTH0R / UNIT SUPV.
5AFETY ANALYSIS REQUDIED NO Fr. YES O (SEE ATTACHED) IF YE5. ANALYSIS ASSIGNED TU: 51TE Q. OR TO INGINEERING O. DATT REVIET OF TEE SUBJECT DOCUMENT HAS BEEE COMPLETED AND APPROYjA. 15 RECOMBEREED. ( Approvers shalt eisaity approvat on the precedere cover sheet) . . c DocUuIsr =n rOR s < Ina . ..........). A P P R o v E D. r is a . n o e r - OYBERSEIP DEPT S . u Nruf d' BATF . 1 FIS 3.0.4 SHERT 8 OF 4
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PNV ;n AP-2.0' -28 hay 1987' ;
l TECBDICAL REVIET ACD CONTROLE EVAL'. A :.!N i.T NEED FOR SATETY ANALYSIS IN ACCORDANCE WITH 10 CMt 30 59 1 (Documents that require General Supt. approval' per Tech Spec. 6.8) FOR DOCUMENT NO. M 1 M REV. DA'IT
# ~~" ~~
The Author ( A) and four 50BC Members (Mialaum - 2 regular members, 2 alterantes) are to roepend to each of the questions below. YE5* 30[' O-Does the documeaurevision result la a change to the fac111ty A- .- or precedures described la the F3AR 7 1 0 2 :D
- 3. :0
- 4. O Does the documeaurevision deviate from compliance to Tech Specs. or is the margia of safety defined la the basis A G[Q reduced ?
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Does the documenvrevision incrosse the probability of A- O ' occurrence, et the ceasequences of as accident, or malfunctica 1 .O-of equipment important to safety (Class 1) evaluated la the 2 Q. F5AR increased ? , 3 0 4 }\' O Does the documenvrevision crests the possibility for sa accident or malfunction of a different type than any evaluated A Y Q-in the F5AR ? 1 d]' O 2 W- O l
= A ' mayst constitutes a Yas' respeese.
SORC MEMBm anmMMENDATIONS TO GENER AL SUPERINTENDENT Recommended Nuclear Engineering or Tech services perform a safety ANALYSIS to present to 50RC (noted by a "YI5* 1 2 3 4 response to any of the above questicas) O O O O Recommended fu115(MC committee review this Evaluation 1 2 3 4 of need for Safety Analysis. O O O O Recommended approval - This document does met involve sa unreviewed safety questism. 1 d f3 3/4 ET A sotc Member Name Dans t O' L' %d 2 k. baai4 M
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CaO h EON
.2./ #2.1 fx SORC nestleg 3 #. hA(" -
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Figure 3.0.4 W 3 SF 4 l AP-2.0 -29. August 1966
TECHBICAL REVIEW HDD CODTHOL REFERENCE DOCUMENTS 1 1 The itene entered beter hope been lasteded 'la the preparatlea and/or review at the attached i reference docuesat and are presented la place of a specific check shoot for the documeet., The felleries pereene were Preceders le la compliance with ceasetted about this precedure ' the fellerlag Testaisal Speelfications NAME TIT'2 "Y .JEC?t0M AMENDMENT BY LL%E U-t L W.Jo IW 4-L LNo M.t wf Compliance with: CPR / US NRC. Ceeplianse with - N : ANSI STANDARD (e) DATED: -M RERlLATORY GJ]ES(e) ' DATED i II K J f / n A_
- '=v gj g vg Ceepilease with: ASb5 Beller and le ceanistent with the fellering Staties '
Pressure Vessel Code (s) er Site precederee:
=-TnGhi DATE ADF hund By MU- - any.
, - gy L A }hL l p!f W r irs l
OTHER INFORMATION SOURCES CONSULTED BY uaN 1 vvu w+ 1 l l l
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l AuTuos ...M. .... . .h.k.1.f4'hg. naviewnD w ............ .o4,s. ,, ;;,;;;;;;; . oar .. manmrrt i FIGUEE 3.8.4 SEEET 3 W 4 AP-2.0 -30 August 1986
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u i TECHDICAL REVIEW allD C0llTHOL ; REVIEW l CHECK LIST - To as peEPARED ST AWTWDS 1 L- CusCx usT rom oocumsNT N0dd.dd...$.$Ed asy. .....I......... oars.....d.....
* /* ONLY B055 TEAY APPLY YES - NA;
- . All references aseded to imptomaat the precedure are clearly identitled and eval'lable...........[ Q i
i The precedure costalas adequate equipment lists, precautions and limitations,- prereqmioites, graphs, dlagrams or data cheete as required................... ..................................j ur O t Surveillease and Malatensase Procedure utilless PLANT IWACT statessat associated with . , approval /permiss ica for use... ............ .. ........... .. .;.... .. .. .... .... ................ ......... ...................... O As appropriate, presedere addresses use'et MARE - UPe. ...................................................... 01 If appropriate, precedure requires use of fire protestica asseures. [! le. hursias permite etc ..................... ....................'............................ ..................................... O N, it leads are lifted lumpers placus er blasts used la the 1 statement askseeledens such use....... ............................preseders, the PLANT _lWACT
...................................................O! V-i As appropriate, precedure metities other affected departments'such as Q.C.', Operations. >
14 C. Malatensace. Red Protection etc........................................................................................ O If Technical Specifleaties is esseeded, appropriate metien le idestifled.. ............................. [ O The precedure references valve numbers, aster centrol numbers, power supplies. > lastremontaties identilleatien le eleer and eerroet....... .......... ............................................... U When encountered, s.Q. related equipesamt is identified as seet..............................................' O B' - Procedure steps are clear and accurate. They are set unseesseerily difficult to laplement.... [ O The precedure reflects the latest system er component ceafiguraties..... ................................ O
- F The procedure refleets work as it le to be dese at the
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seatles.................................................... # D Procedure removes any jumpers er hiests and restere vert.............. ................. ........ ..................
..s lifted leads used to effectusthe
.............................................................O
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*RsTURN TO SssVIG" asse deshle verificaties and'ideatifies specifies beide verified........ O. #
rer = a e precedores, agrygN TO SERVICE
- either performs a 3057 MAINTENANG TEST er reiareasse a roesired test..........ooo... ................................ ................................... O
\ ' i l MARK - UPs are c leared er surrendered........... ........................ ............................................ d l l l l
- ACCsPTANG CRITERIA" identafles aseamplisheses et spesille goale.................................... O- l[
roan PaspAnso sf. .. !..[4 uph... o47s......!/.d f..I....... . n oems z.e.4 sums? 4 er 4 AP-2.0 -31 August 1986
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, , TEMPORARY CHANGE NOTICE This temporary change shall be documented and approved by LThe General' Superintendent-Nuclear Generation based upos recommendation of 80RC membera within 14 -
days in accordance with Technical specification 6.8.3. l 70: STATION SUPERINTENDENT, UNIT' k 95 The attached Temporary Change was made to l precedure No. Oh C A-f j Rev. # Tl t1e 13x.#- St T1- Da so' CAleazarwJ Anw w W4L. NODIFICATION RELATED CHANGE O YES Dilo
. Reason A un , A A m .l - .. > m M. ./ d' N00 CONTROL NUMBER uL i- a sR7JaA - - :/ -- u m a
(~$ w C/d a - n 91 AJL-M.K/~AEL5~L/N4 And is recommended'to'be: ONE TIME ONLY . . . . ......O PERMANENT CHANGE . . ......V The intent of the original procedure is not altere Author 31gnature d fa) - - 24 Date 1/19/hV The temporary precedure revision was approved by: Dept. Supv. Signature ' ' - Date f' f'~ SRO Signature OMw # Date A -A 9-PP Station Supt. #M2 #A ' ~ Date _J///s! iip ~~ SORC MEMBER RECOMMENDATIONS (Minimum 2 regular members, 2 alternates) 1 2 3 4 Recomunend full SORC comunittee review this temporary change O O O O Recommend Approval - this tempogary change does not change the intent of the original procedu e and gggg, 1 2 3 4 DM.lavolve an unreviewed safety question. '/ d M e -. SORC Member Signatures Date or SORC Meeting 1 W^A ,, 3%-D number (if required) 2 16 1/'tNN 3 W 3 b/YV
. 4 W r ca>- = 3 M rr
(/ C)
' ' i CENERAL SUPERINTENDENT (or designee) APPROV1L The temporary change is approved in accordance with Technical Specification-6.8.3.
Wl0 YA2 AAAD ~ b Signature'j' Date J FIGURE 2.0.5 SH 1 OF 1 AP-2.0 -32 June 1987
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4 :l s , Ufj 2.1.3.2 Service Water and Cooling Tove t 819wdown Eff1 East Monito r i Service Water' A' and B and the Cooling. Tower Blowdown are pumped: to thef,1 discharge tunnel which in turn flows directly to Iake_0ntario. Normal flow :)
- i. rates for each Service Water Pump is 15,000 sps while that for the cooling Tower Blowdown is 10,200 spa. Credit.
~
l is not taken for.'any dilution of , l these. individual effluent streams. . 3 The radiation detector is a sodium; tedida crystal. It. is a: scintillation device. .The~ crystal is sensitive - to, samma and ; beta . radiation. - Howeve r .
- because . of the setal walls = . in - i ts' sample chamber and the absorption.
characteristics of water. the sonitor is not particularly sensitive to beta radiation. .t Detector response ti(Ci*CFi) will-- ' initially . be ' assumed to correspondi 'to g ,, p that calculated by the manufacturer. . However, this will be evaluated..p+4+c , ; d 'y* *y(ef. -so Commercial sample 0peration of Reactor Coolant, (af and' ter atduring every- fuel cycle by placing a d
.2.' s noting its gross count rate.. two; hour decay) in t46a'sonitor. and7) i
' Reactor- Coolant is = chosen . because . i t represents the most likely contaminate of Station Waters.: .. .
g M_ Ad 9g A two hour decay' is chosen by ' judgement of i the staf f of - Niagara Mohawk '3 ! Power Corporation: ' Reactor Coolant . wit:h no' decay: contains . a considerable - S{ 5 amount of very energetic nuclides which' would bias the detector response ters high. However assuming a longer than 2 hour decay is not realistic as :fl kI
.the most likely release sechanism is a ' leak through the Residual Hea t /3 Removal Heat Exchangers which would- contain- Reactor Coolant during: -6 shutdowns. t !
4 :' The initial setpoint calculation is presente'd as both 'an example and for h
~
.the purposes of documenting the calculation.' It will- be recalculated prior y to commercial operation and' during . every ' fuel cycle when a Radiochemi f analysis of Reactor Coolant is completed for E bar determination as N9-required by TS Table 4.4.5-1 or when activity is detected in the respective effluent streas.
IS0 TOPE 2 HR DECAY MPC FRACTION DETECTOR' CPM NAME ACTIVITY OF MFC ~ RESPONSE TOTAL: CONCENTRATION uC1/mi cpa/uci/a1 B/C cpa uCi/mi A B C D ;E F 't (Ci) (MPC1) ( Ci/MPC1) (CFi)' ( CiCFi) H3 1.0E-2 3Ei5 3.3. -- -- F18 1.9E-3 '5E-4 3.8 -- -- - - - - - - NA24 3.7E-3 3E-5 1.2E-2 - - - - P32 7.8E-5 2E-5 3.9 '- - - - - - - - - - - CR51 2.3E-3 2E-3 1.2 - - --- MN54 4.0E-5 1E-4 4. 0 E-1 8.4 2E7 ' 3.4E3 NN56 2.9E-2 1E-4 2.9E-2 1.2E8- 3.5E6 FE55 3.9E-4 8E-4 4.9E-1 --- - - - - - FE59 8.0E-5 5E-5 1.6 8.63E7 6.~ 9E s C058 5.0E-3 9E-5 5. 6 E-1 1.14E8 5.7E5 C060 5.0E-4 3E-5 1.7E 1.65E8 8.3 E 4 NI63 3.9E-7 3E-5 1.3'd-2 ---- - - - - - - l 1 May 1986 I l
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-t WINE MILE POINT NUCLEAR STATION NINE,Mit,E POINT UNIT ?
OFF. 811 r. t>0rE CALCUI.ATION Mt.'!!' A!. (0 Cat) l' ATE AND 7 NIT? t.I.0
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S ui.una r_y, p f Pa r te Revision 4 (Effective b Ob ) W ',[*' 1 ru :. 1-111,1-4,6,8,10-14,17-18, 20-35,37-b3,55-89,92 May 1986 5 August 1986 9 l 15,16 October 1986 May 1987 ~
$4 19 May 1987 (TCN-1) I 7 June'1987 (TCN-2) ,,u . . ,, .
m.,,-.36- :-<-*~ ' June 1987 (TCN-3) ~ . 1 90-91,93-104 ' ' ~~* ~ sept' ember'1987 (TCN-4) ~ .~ _ , " , _ '. February 1988 WJ ACARA MOHAWK POWER CORPORATigl[ 1 1 THIS PROCEDUtt NOT TO BE ! USED AFTet February 1990 l SUBJECT TO PttIODIC ttVIEW. ! l
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i 2.1.3.2 Service Water and Cooling Tower Blowdown Effluent Monitor Service Water A and B and the Cooling Tower Blowdown are pumped to the . discharge tunnel which in turn flows directly to lake Ontario. Normal flow ; rates for each Service Water Pump is 15,000 sps while that for the Cooling ' c Tower Blowdown is 10,200 spa. Credit is not taken for any dilution of i these individual effluent streams. ' The radiation detector is a sodium iodide crystal. It is a scintillation ; device. The crystal is _ sensitive to ganna and beta radiation. Howeve r , > because of the metal walls in i ts ' sample chamber and t h.e absorption i characteristics of water, the monitor is not particularly sensitive to beta ; radiation. Detector response ti(Ci*CFi) will initially be assumed to correspond to that calculated by the manufacturer. However, this will be evaluated prior to Commercial Operation and 'during every fuel cycle by placing a diluted sample of Reactor Coolant (af ter a two hour decay) in the monitor and i noting its gross count rate. Reactor Coolant is chosen because it
- represents the most likely contaminate of Station Waters. i'
' A two hour decay is chosen by judgement of the staff of Niagara Mohawk Power Corporation Reactor Coolant with no decay: contains a considerable amount of very energetic nuclides which would bias the detector response r t era high. However aJeuning a longer than.2 hour decay is not realistic as the most likely release mechanism is a leak through the Residual Heat Removal Heat Exchangers which would contain Reactor Coolant during ' shutdowns. ' , The initial setpoint calculation is presented as both an example and- for i the purposes of documenting the calculation. It will be recalculated prior ' to commercial operation and during every f uel cycle when a Radiochemical analysis of Reactor Coolant is completed for E bar determination as required by TS Table 4.4.5-1 or when activity is detected in the respective effluent stream. , i ISOT0FE 2 HR DECAY MPC FRACTION ' DETECTOR CPM NAME ACTIVITY OF MPC RESPONSE TOTAL l CONCENTRATION uC1/a1 B/C cpa/uci/mi eps uC1/mi A B C D E F (C1) (MPCi) ( C1/MPCi) (CF1) ( CiCFi ) D 1.0E-2 3E-3 3.3 - - ---- -------- F18 1.9E-3 5E-4 3.8 -------- -------- NA24 3.7E-3 3E-5 1.2E-2 -------- -------- P32 7. 8 E-5 2E-5 3.9 -------- -------- CR51 2.3E-3 2E-3 1.2 --- - -- - - - - - - - MN54 4.0E-5 1E-4 4.0E-1 8.4 2E7 3.4E3 MN56 2.9E-2 1E-4 2.9E-2 1.2E8 3.5E6 FE55 3.9E-4 8E-4 4.9E-1 - - - - - - - - - - - - - FE59 8.0E-5 5E-5 1.6 8.63E7 6.9E 3 C058 5.0E-3 9E-5 5. 6 E-1 1.14E8 5.7E5 i C060 5.0E-4 3E-5 1.7E-1 1.65E8 8.3E4 NI63 3.9E-7 3E-5 1.3E-2 --~~ - - - - - - - May 1986
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SUPtliStutu l NINE MILE POINT WUCLEAR STATIO.N._ 7 WINE MILE POINT UNIT 2 ' OFF-SITE DOSE CALCULATION MANUAL (ODCM) ' DATE AND INITIALS APPROVALS SICNATURES i REVISION 4 RU_I U 12!! j, REVISION 6 Chemistry & Radlochemistry } Supervisor ,j J. N. Duell . Nd ti y//sf. C,, s'/ / , Cheelstry & Radiation Management t et , 1 = N8 i Station Superintendent ' NMPNS Unit 2 R. B. Abbott bA, _
' 2k.b f
Ceneral Superintend Nuclcar Generation .
, s T. J. Perkins '
t l' E / /
- e ' 71T6t7 O['4
/
Summary of Paaes Revision 4 (Effective 2/25/88 3 , EAC,g pgIg 1-111,1-4,6,8,11-14.17-18, 28 -3 5,3 7- 5 3,5 5-K! . 8 7-89. 9 2 May 1986 5 August 1986 9 15,16 October 1986 May 1987 - 54 May 1987 (TCN-1) 19 7 June 1987 (TCN-2) 36 June 1987 (TCN-3) September 1987 (TCN-4) 90-91,93-104 February 1988 10
*20-27.83-86 March 1988 (TCN-5)
April 1988 iReissue) NIACARA MO)LWK POWER CORPORATION THIS PROCEDURE NOT TO BE 'I 9 USED AFTER February 1990 SUD3ECT TO PERIODIC REVIEW.
- Changes per Section 11.5 AP-2.0 Signed
/Mk Date i
4 TABLE OF CONTENTS SECTION SUBJECT TS SECTION PAG E or TABLE ~ or TABLE
1.0 INTRODUCTION
1 2.0 LIQUID EFFLUENTS 2 2.1 Liquid Effluent Monitor Alara Setpoints 2 2.1.1 Basis 3.11.1.1 2 2.1. 2 Setpoint Determination Methodology 3.3.7.10 2 2.1.2.1 Liquid Radweste Effluent Radiation. 2-3 Alara Setpoint 2.1. 2. 2 Contaminated Dilution Water Radwaste '4 Effluent Monitor Alara Setpoint Calculations 2.1.2.3 Service Water and Cooling Tower Blowdown 5 Ef fluent Radiation Alara Setpoint 2.1.3 Discussion 5 2.1. 3.1 Liquid Radwaste Effluent 6-9 ! 2.1. 3. 2 Service Water and Cooling Tower Blowdown 10-12 l 2.2 Liquid Effluent Concentration 3.11.1.1 12 Calculation 4.11.1.1.2 2.3 Liquid Effluent Dose Calculation 3.11.1.2 13-14 3.11.1.3 4.11.1.2
'4.11.1.3.1 2.4 Liquid Effluent Dose Factor 14-15 Derivation - Ait 2.5 Liquid Effluent Sampling 4.11-1 15-16 Representativeness note b 2.6 Liquid Radweste System Operation 3.11.1.3 16-17 Table 2-1 Liquid Effluent Detector Response 18 Table 2-2 Ait Liquid Effluent Dose Factor 19 Figure 2-1 Liquid Radwaste Treatment System 3.11.1.3 20-27 thru 2-8 Flow Diagrams 3.11.3 Figure 2-9 Liquid Radiation Monitoring 28 Figure 2-10 Of f-line Liquid Monitor 29 l 3.0 GASEOUS EFFLUENTS 30 3.1 Gaseous Effluent Monitor Alara Setpoints 30 3.1.1 Basis 3.11.2.1 30 3.1.2 Satpoint Determination Methodology 3.3.7.11 30 '
3.1. 2.1 Stack Noble Gas Radiation Alara Setpoint 30 3.1.2.2 Vent Noble Gas Radiation Alara Setpoint 31-32 l 3.1. 2. 3 Offgas Pretreatment Radiation Alara 32 Setpoint 3.1.3 Discussion 33 l 3.1. 3.1 Stack Effluent 34 3.1.3.2 Vent Effluent 35 : 3.1.3.3 Offgas Process 36 I 3.2 Gaseous Effluent Dose Rate Calculation 3.11.2.1 37 3.2.1 Total Body Dose Rate Due to Noble Gases 3.11. 2.1. a 37-38 4.11.2.1.1 3.2.2 Skin Dose Rate Due to Noble Gases 3.11. 2.1. a 38-3 9 4.11.2.1.1 l
-1 May 1986 i l
( r j l f i
- TABLE OF CONTENTS l
SECTION l l SUBJECT TS SECTION PAG E or TABLE [ or TABLE i 3.2.3 Organ Dose Rate Due to 1-131; l '.337 39T Tritium and Particulates with half- 3.11.2.1.b ; lives greater than 8 days 4.11.2.1.2 ! 3.3 Caseous Effluent Dose Calculation 3.11.2.2 41-42 : Methodology 3.11 .2.3. ' 3.11.2.5 i 3.3.1 Ganna Air Dose Due to Noble Cases 3.11. 2. 2. a - 42 f 4.11.2.2 ! 3.3.2 Beta Air Dose Due to Noble Gases 3.11. 2. 2. b 43 3.3.3 Organ Done Due to I-131, I-133 Tritium 43-45 I and Particulates with half-lives ' 3.11.2.3 }
- greater than 8 days. 3.11.2.5 4.11.2.3 4.11.2.5.1 ;
3.4 Gaseous Effluent Dose Factor Definition 45 and Derivation i 3.4.1 Bi- Plume Shine Gamma Air Dose Factor 45-47 Vi- Plume Shine Total Body Dose Factor ' 3.4.2 Ki, Li, Mi and Ni- Immersion Dose Factors 47 i 3.4.3 Pi- Iodine, Particulate and Tritiva 47-51 Organ Dose Rate Factors ! 3.4.4 Ri- Iodine, Particulate and Tritium 51-5 7 i i Organ Dose Factors 3.4.5 X/Q and Wy- Dispersion Factors for Dose Rate 58 l 3.4.6 We and Wy- Dispersion Factors for Dose 59 ) 3.5 Caseous Effluent I-133 Estination 59 i 3.6 Use of Concurrent Meteorological Data vs. .59 ; Historical Data 3.7 Gaseous Radweste Treatment System 3.11.2.4 59 Operation [ 3.8 ( Ventilation Exhaust Treatment System 3.11.2.5 60 Operations i Ta ble 3-1 Offgas Noble Gas Detector Response 61 Table 3-2 Bi and Vi- Plume Shine Dose Factors 62 Table 3-3 Ki, Li, Ni and Ni- Immersion Dose Factors 63 Ta le 3-4 Pi- Ground Plane Dose Rate Factors 64 Table 3-5 Pi- Inhalation Dose Rate Factors 65 Table 3-6 Pi- Food (Cow Milk) Dose Rate Factors 66. Table 3-7 Ri- Inhalation Dose Factors for Infant 67-70
- t o 3-10 Child, Teen and Adult Table 3-11 Ri- Ground Plane Dose Factors 71 Table 3-12 Ri- Cowsilk Ingestion Dose Factors for 72-75 t o 3-15 Infant, Child, Teen and Adult Table 3-16 Ri- Cowmeat Ingestion Dose Factors for 76-7 8 i t o 3-18 Child, Teen and Adult Ta ble 3-19 Ri- Vegetation Ingestion Dose Factors for 79-81 t o 3-21 Child. Teen and Adult
-11 May 1986
l l OFF-SITE DOSE CALCULATION MANUAL (0DCM) l t
1.0 INTRODUCTION
This is the OFFSITE DOSE CALCUIATION MANUAL (ODCM), referenced in the Nine Mile Point - Unit 2 Technical Specification. It describes the i methodology for liquid and gaseous effluent monitor alars setpoint I calculations, the methodology for computing the offsite dose due to i liquid effluents, gaseous effluents, and ~ the uranius. fuel cycle as I 4 well as the radiological environmental monitoring and inte:isboratory comparison programs. The ODCM will be reviewed and approved by the NRC. Changes shall be { provided in the semi annual radioactive effluent release reports submitted to the NRC. i Section 2 establishes methods used to calculate the Liquid Effluent Monitor Alara setpoints and to demonstrate compliance with TS Section L 3.11.1.1 limits on concentration of releases to the ' environment as required in TS Section 3.3.7.10 and 4.11.1.1.2 respectively. ! Additionally, the method used to calculate the cumulative dose I contributions from liquid effluents and the methods used to assure thorough mixing ~and sampling of liquid radioactive waste tanks to be > discharged as required in TS Section 4.11.1.2, 4.11.1.3.1 and Table 4.11-1 note b respectively are presented. Section 3 establishes calculational methods used to calculate th e Gaseous Effluent Monitor Alarm setpoints and to demonstrate I compliance with TS Section 3.11.2.1 limits on dose' rates due to gaseous releases to the environment as required in TS Section 3.3. 7.11, 4.11.2.1.1 and 4.11.2.1.2 respectively. Additionally, the calculational methods used to calculate cumulative dose contributions from gaseous effluents as required in TS Section 4.11.2.2, 4.11 .2.3 and 4.11.2.5 are presented. Section 4 establishes the sethod used to determine cumulative dose l contributions from the Uranium Fuel Cycle as required by TS Section ! 4.11.4.1, 4.11.4.2 and 6.9.1.8. Section 5 establishes the environmental monitoring program as required by TS Section 3.12 and 4.12 including the Interlaboratory Comparison Progran required by TS Section 4.12.3. i Section 6 discusses some of the references contained in TS Table 3.12-1, Radiological Environmental Monitoring Program. May 1986 ,
@?II g CID nun aluummunmuaumus 0%
NINE MILE POINT UNIT 2 OFF-SITE DOSE CALCULATION F3"JAL (ODCM) DATE AND INITIALS j APPROVALS SIGNATURES REVISION 4 REVISION 5 REVISION 6 Chemistry & Radiochemistry . Supervisor 7 "9 f. J. W. Duell / Mt' d. 4/s
</ , ,
Cheelstry & Radiation Management p *
$"'IUt
/$ d D 2hdd Station Superintendent NNPNS Unit 2
, R. B. Abbott ,OA b'L[ b General Superintendent Nuclear Generation T. J. Perkins ,.1'/ e
, N; /4 go _
k29k"f7)/ f 8=- ^ry of Panes Revision 4 (Effective 2/23/00 )
- f. Aft.E PAI.E l 111.2-4.6,8.11-14,17-18, '
2E-33,37 53,35 82.87 89.92 May 1986 5 Au8ust 1986 9 October 1986 15,16 May 1987 - i 54 May 1987 (TCN-1) 19 June 1987 (TCW-2) 7 June 1987 (TCN-3) l 36 September 1987 (TCW-4) 90-91,93-104 Pobruary 1988 10 March 1988 (TCN-5)
*20-27,83-86 April 1988 (Reissue)
*i,11,1 November 1988 (Reissue)
NIAGARA MOHAWK POWER CORPORATION THIS PROCEDUtt NOT TO 88 i 7 USED AFTER February 1990 ; 8 ECT TO PERIODIC REVIEW.
- Changes per Section 11.5 AP-2.0 /f84 M
- Changes per Section 11.5. AP-2.0 d '21
#_ U___.
Date Signed
i i M 2 0F CONTENTS l l SECTION SUBJECT TS SECTION APPLICASLE PACE i j or TABLE or TABLE _ PROCEDURE ! l
1.0 INTRODUCTION
1 l 2.0 LIQUID EFFLUENTS N/A 2 i 2.1 Liquid Effluent Monitor Alarm Setpoints N/A 2 ; 2.1.1 Basis 3.11.1.1 N/A 2 l 2.1.2 Setpoint Determination Methodology 3.3.7.9 N/A 2 i 2.1.2.1 Liquid Radweste Effluent Radiation N2-CSP-4V 2-3 l Alarm Setpoint
- Contaminated Dilution Water Radweste 2.1.2.2 N/A 4 '
Effluent Monitor Alarm Setpoint calculations j 2.1.2.3 Service Water and Cooling Tower Blowdown NP-CSP-13 App. D 5 Effluent Radiation Alarm Setpoint t 2.1.3 Discussion N/A 5-i 3 1.3.1 Liquid Radweste Effluent N/A 6-9 i! 2.1.3.2 Service Water and Cooling Tower 81owdown 10-12 '1 2.2 Liquid Effluent Concentration 3.11.1.1 N2-CSP-4V 12 Calculation 4.11.1.1.2 2.3 Liquid Effluent Dose calculation 3.11.1.2 N2-CSP-4V 13-14 3.11.1.3 $ 4.11.1.2 4.11.1.3.1 l 2.4 Liquid Effluent Dose Factor 14-15 l Derivation - Alt l 2.5 Liquid Effluent Sampling 4.11.1-1 N2-CSP-4V 15-16 Representativeness note b , 2.6 Liquid Radwaste System Operation 3.11.1.3 16-17 Table 2-1 Liquid Effluent Detector Response 18 Table 2-2 Alt Liquid Effluent Dose Factor 19 i t Figure 2-1 Liquid Radweste Treatment System 3.11.1.3 20-271 thru 2-8 Flow Diagrams 3.11.3 .i Figure 2-9 Liquid Radiation Monitoring 28 l ' l Figure 2-10 Off-line Liquid Monitor 29 1 i 3.0 GASEOUS EFFLUENTS 30 3.1 Gaseous Effluent Monitor Alarm Setpoints 30 ! 3.1.1 Basis 3.11.2.1 30 3.1.2 Setpoint Determination Methodology 3.3.7.10 30 3.1.2.1 Stack Noble Gas Radiation Alarm Setpoint N2-CSP-13 App. D 30-31 3.1.2.2 Vent Noble Gas Radiation Alarm Setpoint N2-CSP-13 App. D 31-32 3.1.2.3 Offgas Protreatment Radiation Alara 32 Setpoint 3.1.3 Discussion 33 l 3.1.3.1 Stack Effluent 34 3.1.3.2 Vent Effluent 35 3.1.3.3 offgas Process 36 , 3.2 Gaseous Effluent Dose Rate Calculation 3.11.2.1 N2-CSP-7V 37 3.2.1 Total Body Dose Rate Due to Noble Cases 3.11.2.1.a N2-CSP-7V 37-38 4.11.2.1.1 3.2.2 Skin Dose Rate Due to Noble Cases 3.11.2.1.a N2-CSP-7V 38-39 4.11.2.1.1-
-1 November 1988
TA8LE OF CONTENTS SECTION SU8 JECT TS SECTION APPLICABLE I PAGEf or TABLE or TABLE EggggDURE 3.2.3 Organ Dose Rate Due to I-131. I-133 W2-CSP-7V 39-41 Tritium and Particulates with half- 3.11.2.1.b i lives greater than 8 days 4.11.2.1.2 3.3 Gaseous Effluent Dose Calculation 3.11.2.2 N2-CSP-7V 41-42, Methodology 3.11.2.3 l,' 3.11.2.5 : 3.3.1 Gamma Air Dose Due to Noble Gases 3.11.2.2.a N2-CSP-7V 42 l 4.11.2.2 1 3.3.2 Beta Air Dose Due to Noble Cases 3.11.2.2.b N2-CSP-7V 43 l ' 3.3.3 organ Dose Due to I-131. I-133. Tritium N2-CSP-7V 43-45I and Particulates with half-lives 3.11.2.3 l greater than 8 days. '; 3.11.2.5 ' 4.11.2.3 4.11.2.5.1 3.4 Gaseous Effluent Dose Factor Definition 45 i I and Derivation 3.4.1 81- Plume Shine Gamma Air Dose Factor 45-47i Vi- Plume Shine Total Body Dose Factor ; 3.4.2 Kl. Lt. Mi and N1- Immersion Dose Factors 3'.4.3 Pi- Iodine. Particulate and Tritium 47 l Organ Dose Rate Factors 47-51i 3.4.4 Ri- Iodine. Particulate and Tritium Organ Dose Factors 51-57 l 3.4.5 X/Q and Wy- Dispersion Factors for Dose Rate 58 l 3.4.6 We and Wy- Dispersion Factors for Dose 59 3.5 Gaseous Effluent I-133 Estimation 3.6 N2-CSP-78Q 59 1 Use of Concurrent Meteorological Data vs. 59 I Historical Data ! 3.7 Gaseous Radwaste Treatment System 3.11.2.4 59 i Operation 4 3.8 Ventilation Exhaust Treatment System 3.11.2.5 60 l Operations Table 3-1 Offgas Noble Gas Detector Response 61 Table 3-2 81 and Vl+ Plume Shine Dose Factors 62 Table 3-3 Ki, Lt. Ni and N1- Immersion Dose Factors 63 l Table 3-4 Pi- Ground Plane Dose Rate Factors 64 ! Table 3-5 Pi- Inhalation Dose Rate Factors 65
- Table 3-6 Pi- Food (Cow Milk) Dose Rate Factors 66 Table 3-7 Ri- Inhalation Dose Factors for Infant 67-70 to 3-10 Child. Teen and Adult Table 3-11 Ri- Ground Plane Dose Factors 71 Table 3-12 Ri- Cownlik Ingestion Dose Factors for
, to 3-15 Infant. Child. Teen and Adult 72-75} i Table 3-16 Ri- Cowmeat Ingestion Dose Factors for 76-78 to 3-18 Child. Teen and Adult Table 3-19 Ri- Vegetation Ingestion Dose Factors for 79-81 to 3-21 Child. Teen and Adult
-11 November 1988
I TABLE OF CONTENTS f i S ECTION SUBJECT TS SECTION PAGE or TABLE or
~
TABLE : Table 3-22 X/Q Wy and Ws- Dispersion Factors for 82. f Receptor Incations Figure 3-1 Gaseous Radweste Treatment Systen Flow 3.11.2.4 83-85 l thru 3-3 Diagrams 1 , Figure 3-4 Ventilation Exhaust Treatment System 3.11.2.5 86 l Flow Diagran.s . Figure 3-5 Gaseous Radiation Monitoring 87 l l Figure 3-6 Gaseous Effluent Monitoring Systen 88 l 4.0 URANIUM FUEL CYCLE 3.11.4- 89-90 ! 4.1 Evaluation of Doses From Liquid Effluents 4.11.4.1 90-91 l 4.2 Evaluation of Doses From Gaseous Effluents 4.11.4.1 92 i 4.3 Evaluation of Doses From Direct Radiation 4.11.4.2 92 .! 4.4 Doses to Members of the Public Within 6.9.1.8 93-94 l Site Boundary ! i 5.0 ENVIRONMENTAL MONITORING PROGRAM 3.12 95 ! 4.12 [ 5.1 Sampling Stations 3.12.1 95 ! 4.12.1 ; 5.2 Interlaboratory Comparison Program 4.12.3 95 ! 5.3 Capabilities for Thermoluminescent Dosimeters 97-97 t Used for Environmental Measurements ! Table 5.1 Radiological Environmental Monitoring 3.12.1 98-100 } Program Sampling Iocations 4.12.1 : Table 3.12-1 ! l Note (a) ! 6.0 DISCUSSION OF TECHNICAL SPECIFICATION REFERENCES 101 l' 6.1 Table 3.12-1 note 101 i 6.2 Table 3.12-1 note h 101 ! 6.3 Table 3.12-1 note i 102 l 6.4 Table 3.12-1 note 1 102 ' Figure 5.1-1 Nine Mile Point On-Site Map I l Figure 5.1-2 Nine Mile Point Off-Site Map ! Figure 5.1.3-1 Site Boundaries l i i P I l
-111 May 1986 l
I 0FF-BITE DOSE CALCULATION RANUAL (003) l l , l l I l
1.0 INTRODUCTION
l This is the OFFSITE DOSE CALCULATION MANUAL _(ODCN), referenced in the-Nine Mile Point - Unit 2 Technical Specification. It describes the l methodology for liquid and gaseous effluent monitor alarm setpoint i calculations, the methodology for computing the offsite dose due to ( liquid effluents, gaseous effluents, and the uranium fuel cycle as I
- well as ~ the radiological environmental monitoring -and interlaboratory '
comparison programs. i The ODCN will be reviewed and approved by the NRC.. Changes shall.be i provided in the semi annual- radioactive effluent release reports ; submitted to the NRC.
- i Section 2 establishes methods used to calculate the Liquid Effluent j Monitor Alarm setpoints and to demonstrate compliance with TS Section ,
I 3.11.1.1 limits on concentration of releases to the environment as ! required in TS Section 3.3.7.9 and 4.11.1.1.2 respectively.l* ' Additionally, the method used' to calculate the cumulative dose contributions from liquid effluents and the methods used to assure i thorough mixing and sampling of 11guld radioactive waste tanks to be ! discharged as required in TS Section 4.11.1.2, 4.11.1.3.1 and Table ! 4.11.1-1 note b respectively are presented. l, .j Section 3 establishes calculational methods used - to calculate the ! Gaseous Effluent Monitor Alarm setpoints and to demonstrate J compliance with TS Section 3.11.2.1 limits on dose rates due to { i gaseous releases to the environment as required in TS Section l 3.3.7.10, 4.11.2.1.1 and 4.11.2.1.2 respectively. Additionally, thel
- calculational methods used to calculate cumulative dose contributions from gaseous effluents as required in TS Section 4.11.2.2, 4.11.2.3 i and 4.11.2.5 are presented. ,
i Section 4 establishes the method used to determine cumulative dose ; contributions from the Uranium Fuel Cycle as required by TS Section ' 4.11.4.1, 4.11.4.2 and 6.9.1.8. , Section 5- establishes the environmental monitoring program as i required by TS Section 3.12 and 4.12 including the Interlaboratory Comparison Program required by TS Section 4.12.3. ; i Section 6 discusses some of the references coutained in TS Table + 3.12-1, Radiological Environmental Monitoring Program. t i I
-1 November 1938 ,
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l l 2.0 LIQUID EFFLUENTS Service Water A and B, Cooling Tower Blowdown and the Liquid Radioactive Waste Discharges comprise the Radioactive Liquid Effluents at Unit 2. (See figure 2-9) Presently there are no temporary outdoor tanks containing radioactive wate r capable of affecting the nearest known or future water supply in an unrestricted ! area. NUREG 0133 and Regulatory Guide 1.109, Rev.1 were followed in . the development of this section. j 2.1 Liquid Effluent Monitor Alara Setpoint s i l 2.1.1 Basis ! l Technical Specification 3.11.1.1 provide the basis for the-alars i setpoints The concentration of radioactive material released in l liquid effluents to UNRESTRICTED ARELS (see Figure 5.1.3"1) shall be lialted to the concentrations . specified in 10 CFR 20 3 Appendix B, ; Ta ble II, Column 2, for radionuclides other thaa dissolved' or 1 entrained noble gases. For dissolved or entrained nobles gases, the
- concentration shall be limited by 2 x 10-4 microcurie /mi total activity.
5 2.1.2 Setpoint Determination Methodology ; 2.1.2.1 Liquid Radwaste Effluent Radiation Alara Setpoint This monitors setpoint takes into account the dilution of Radwaste ; Effluents provided by the Service Water and Cooling Tower Blowdown flows. Detector response for the nuclides to be discharged (cpa) is multiplied by the Actual Dilution Factor (dilution flow / waste stream ; flow) and divided by the Required Dilution Factor (total fraction of ' MPC in the waste stream). A safety factor is used to ensure that the i limit is never exceeded. Servica Water and Cooling Tower Blowdown are normally non-radioactive. If they are found to be contaminated prior to a Liquid Radweste discharge then an alternative equation is used to take into account the contamination. If they become contaminated during a Radwaste discharge, then the discharge will be immediately terminated and the situation fi41y assessed. Normal Radwaste Effluent Alara Satpoint Calculation: Alara Setpoint ( [ 0.8*(F/f)* ti(Ci*CFi))/(Ii(C1/MPC1)] + Background . i Where Alara Setpoint = The Radiation Detector Alara Setpoint, cpm 0.8 = Safety Factor, unitiese F = Nonradioactive dilution flow rate, spa. Service May 1986
I Water Flow ranges from 30,000 to 58,000 spa. l Blowdown flow is typically 10,200 spa. i , Cf = Concentration of isotope i in Radweste. ' tank prior to dilution, uCi/mi i CFi = Detector response for isotope 1, . net cpa/uci/a1 See Table 2-1 for a list of nominal values
=
f The permissible Radwaste Effluent Flow rate, spa
=
Symbol to denote multiplication. l MPCi = Concentration limit for isotope i from 10CFR20 Appendix B. Table II, Column 2 - uCi/mi Background = Detector response when sample chamber is filled - i with nonradioactive water, cpa . l ti ( Ci*CFi) = The total detector response when exposed to the , concet.tration of nuclides in the Radwaste tank, i cpm ti(Ci/MPC1) = The total fraction of the 10CFR20, Appendix B,
. Table II, Column 2 limit that is in the Radwaste I tank, unitiese. This is also known as the Required Dilution Factor (RDF)
- i CR*tici = An approximation tori (CiCFi) determined. l at each calibration of tha effluent monitor, by '
recording monitor cpa re9ponse to a typical radwaste tank misture analysed by multichannel analyzer (traceable to NBS). CR is a weighted i summation of CF. F /f = An approximation to (F+f)/f, the Actual Dilution i Factor in effect during a discharge. Pe rmissible effluent flow, f shall be calculated to determina l that h?C will not be exceeded in the discharge canal. ' f = (Dilution Flow) * (1 - Fraction Temperina) l (RDF)
- 1.5 Fraction Tempering =
A diversion of some f raction of discharge flow to { the intake canal for the purpose of temperature == control. N OTE: If Actual Dilution Factor is set equal to the Required Dilution Factor, then the alara points required by the above equations I correspond to a concentration of 80% of the Radweste Tank concentration. No discharge could occur, since the monitor would be in alara as soon as the discharge commenced. To avoid this situation, maximum allowable radweste discharge flow is calculated using a avitiple (usually 1.5 to 2) of the Required Dilution Factor, resulting in discharge canal concentration of 2/3 to 1/2 of MPC. l l May 1986
v i 2.1.2.2 Contaminated Dilution Water Radweste Effluent Monitor Alara Setpoin t Calculation: The silowable discharge flow rate for a Radweste tank, when one of the normal dilution streams (Service Water A, Service Water B, or Cooling Tower Blowdown) is contaminated, will be calculated. by an iterative process. Using Radweste tank concentrations with a nominal radweste effluent flow rate (200 spa, for eansple) the resulting fraction of MPC in the discharge canal will be calculated. FMPC = Ii [Is(Fe*Cis)/(MPCi*ts[Fs])] ' Then the permissible radwaste effluent flow rate is given byt-f= Nominal Flow , FMPC*2 The corresponding Alara Setpoint will then be calculated using the following equation, with f limited as above. . 0.8*Ii(Ci*CFi) Alara Setpoint ( + Background til ts(Fe*Cis)/(MPCi* Is [Fs ])] Wheret - Alara Setpoint = The Radiation Detector Alara Setpoint, cps 0.8 = Safety Factor, Unitiese Fs = An Effluent flow ratc f or strema e, spa , Ci = Concentration of isotope i.in Radwasto ' tank prior to dilution, uCi/mi Cis = Concentration of isotope i in Effluent stress e including the Radwas* Effluent tank undiluted, uCi/mi CF1 = Detector response for istope i, not epa /nci/mi ; See Table 2-1 for a list of nominal values. ' MPCi = Concentration limit for isotope i from 10CFR20 Appendix B Table II, Column 1, _ nC1/mi f = The permissible Radweste Effluent Flow rate, spa Background = Detector response when sample chamber is filled i i with nonradioactive water, cps 11(Ci*CFi) = The total detector response when esposed to the concentration of nuclides in the Radwaste tank, cpa Es (Fa*Cis] = The total activity of nuclide i in all Effluent streams, uCi spa /a1 Es[Fs] = The total Liquid Effluent Flow rate, spa (Service Water & CT Blowdown ~& Radweste) : May 1986 l
"i t 2.1.2.3 Service Water and Cooling Tower Blowdown Effluent Alars Setpoint i
These monitor setpointe do not take any credit for dilution of each ! respective effluent streas. Detector response for the distribution of I nuclides potentially discharged is divided by the total MPC fraction of the : radionuclides potentially in the respective stream. A safety f actor is . used to ensure that the limit is never escoeded. , Service Water and Cooling Tower Blowdown are normally non-radioactive.- If they are found to be contaminated by statistically significant increase in ! ! detector response then grab staples will be obtained and analysis neeting > the 11D requirements of Table 4.11-1 completed so 1.ha t an estimate of ; offsite dose can be made and the situation fully assessed. Service Water and Cooling Tower Blowdown Alara Setpoint Equations f Alara Setpoint < [0.8*Ii (Ci*Cri)]/[It (Ci/MPci)) + Background. I Where: Alara Setpoint = The Radiation Detector Alara Setpoint,
. c pm 0.8 = Safety Factor, unitiess j Ci =
Concentration of isotope l'in potential 7 containment, uC1/mi CFi = Detector response for isotope i, not cps /uCi/a1 See Table 2-1 for a list of nominal values MPCi = Concentration limit for isotope i from 10CFR20 Appendix B, Table II, Column 2, uCi/mi Background = Detector response when sample chamber is filled l with nonradioactive water, cpa , ti(Ci*CFi) = The total detector response when exposed to the concentration of nuclides in the potential contaminant, cps ti(Ci/MPC1) = The total fraction of the 10CFR20, Appendix B. Table II, Column 2 limit that is in the potential containment, unitiess. CR* Iici = An approximation to ti(CiCFi) determined, at each calibration of the effluent monitor, by ! recording monitor cpa response to a typical i contaminant sixture analysed by multichannel l analyser (traceable to NBS). CR is a weighted summation of C?i. [ 2.1.3 Discussion l 1 1 l l l \ August 1986
2.1.3.1 Liquid Radweste Effluent Monitor ! The Liquid Radioactive Waste System Tanks are pumped to the discharge , tunnel which in turn flows directly to Iake Ontario. At the end of the ; discharge tunnel in Lake Ontario, a diffuser structure has been installed. ; Its purpose is to maintain surface water temperatures low enough to meet - l t hermal polletion limits. However, it also assists in the near field ' dilution of any activity released. Service Water and the Cooling Tower i Blowdown are also pumped to the discharge tunnel and will provide r dalution. If the Service Water or the Cooling Tower Blowdown is found to l be contastnated, then its activity will be accounted for when calculating the permissible radwaste effluent flow for a Liquid Radweste discharge. The Liquid Radweste Systen Monitor provides alars and automatic termination of release if radiation levels above its alara setpoint are detected. ! t The radiation detector is a sodium iodide crystal.. It is a scintillation ! device. The crystal is sensitive to ganna and beta radiation. Howeve r , ! because of the metal walls of the sample chamber and the absorptiop characteristics of water, the monitor is not particularly sensitive to beta i radiation. Actual detector response ti(Ci*CFi), cps, will be evaluated
- by placing a unaple of typical radioactive waste into the monitor and };
j recording the gross count rate, cya. A calibration ratio, CR , < cpa/uci/al, will be developed by dividing the noted detector response, , ti(Ci*CF1) cps, by total concentration of activity ti(Ci), uCi/cc..; The quantification of the activity will- be completed with gamma s pectrometry e quipment whose calibration is traceable to NBS. This calibration ratio will be used for subsequent setpoint calculations in the determination of detector responses ti(Cl*CFi) = CR* 11(Ci) Where the factors are all as defined above. ; For the calculation of 11( C1/MPC1) the contribution from non gaana esitting nuclides except tritium will be initially estimated based on the . expected ratios to quantified nuclides as listed in the FSAR Table 11.2.5. . Fe-55, Sr-89 and Sr-90 are 2.5, 0.25 and 0.02 times the concentration of f co-60. Periodic analysis of waste for these non gamma esitting nuclides by offsite analysis will provide a better estimate once sufficient activity is , present. Tritium concentration is assumed to equal the latest concentration detected in the monthly Tritium analysis (performed offsite) on liquid radioactive ' I waste tanks discharged or based on the latest tritius detected in the spent , fuel pool if liquid radioactive waste tank discharges have not been made within the last 6 months. , May 1986
'i Nominal flow rates of the Liquid Radioactive Waste System Tanks discharged in 165 sps while dilution flow from the minimum number of Service Water Pumps always in service is over 30,000 spa. and Cooling '
Tower R10wdown is 10.200 gps. Because of the large amount of dilution i the alarm setpoint could be substantially greater than that which I would correspond to the concentration actually in the I _ tank. Potentially a discharge could continue even if the distributton of nuelldes in the tank we; o substantially different from the grab sample obtained prior to discharge which was used to establish the detector I alarm point. To avoid this possibility of "Non representative ' St+pling" resulting in erronous assumptions about the discharge of a tank, the tank is recirculated for a minimum of 2.5 tank volumes prior to sampling. A setpoint of 355 cpm above background will be used until grab sample snelysis with the ' required LLD sensillvity on TS Table 4.11-1 detects 7pp.3 4 activity at such a level that it cannot be discharged with the nominal L setpoint. 355 cpm is the same nominal setpoint as for the service . water and cooling tower blowndown radiation monitors. These are all identical detectors. -> 4 s 6 June 1987
q A sampla calculation; l's _ presented : lelow ' assuming tank concentrations , equivalent to the diluted concentration' presented in FSAR Table 11.2.5 , which is the expected concentration = of effluent waste' af ter dilution thati are discharged with the design limit for fuel failure-(the- table below is - the undiluted concentration-- corresponding to ai tank -2040 sal per day - discharge with' only cooling tower blowdown dilution of 10,200 spa) . - ISOTOPE ACTIVITY MPC FRACTION DETECTOR CPMt > NAME CONCENTRATION OF MPC . RESPONS E TOTAL 6C1/m1- uCi/mi' (B/C)- cpa/pci/mi cpa , A B. C D E- F (Ci) . (MPC1) ( Ci/MPC1)' (CFi). l ( CiCFi) . 2.8 H3 8.4 E-3 3E-3 - - - - - - - - - - - - - NA24 1.7E-6 3E-5 5.7E-2 ---- - - - - - - - P32' 6.8E-8 2E-5 3.4 E-3 - - - - - - - - - - - - - CR51 2.0E-6 2E-3 1.03-3 ~ ~ - - - - - - - - - l-MN54 2.4 E-8 1E 2.4E-4' .8.42E7 1.98E+0-MNS6 3.2E-7 1E-4 3.2E-3 ' '1.2E7 3.9E +1 - FESS 3.5E-7 8E-4 4.3E-4 ---- -- --= - 1,0E-8 " FE59 SE 2.1E-4' 8.63E7 9.0E -l' C058 6 SE-8 9E-5 7.6E-4 1.14E8 <7.8E+0 C060. 1.4E-7 3E-5 L 4.7E-3 1.65E8 ' 2.4E +1 . NI63 3.5E-10 3E-5 1.1E - -- '- - - - - - NI65 1.8E-9 1E-4 1.8E-5 - - - - - - -------- s CU64 4.3E-6 2E-4 2.1E-2 - - - - - -------- _f ZN65 6.8E-8 1E 6.8E-4. ----- - - - - - - ! BR83 3.3E-8 3E-6 -1.1E-2 - - - - - - - -------- 8.9E-14 1.12E8' 1.0E-5 BR84 ---- - - - - - l SR89 3.6E-8 3E -1.2E-3 7.8E3 2.8E-4 i 2.4E-9 3E-7 7.8E-3' SR90 SR91 4.6E-7 SE-5 9.3E 1.22E8 5.7E+1 3 SR92 7.6E-8 6E-5 1.2E-3 8.17E7 6.1E+0-Y91 1.7E-8 3E-5 5.7E-4 2.47E8 4.2E+0 l Y92 4.6E-7 6E-5 7.8E-3 2.05E7 ' 9.5 Y93 5.1E-7 3E-5 1.7E =- -- - - - - - - -
-- t ZR95 2.7E-9 6E-5 4.5E-5 8.35E 7 2.3E-1 I ZR97 1.0E-9 2E-5 5.2E-5 - -
NE95 2.7E-9 1E-4 2.7E-5' 8.5E7 2.4E-1 M099 6.0E-7 '4E-5 1.6E-2 .2.32E7 , 1.4E+1, TC99M 1.2E-6 3E-3 4.1E-4' . 2.32E 7 2.8E +1 RU103 6.8E-9 8E-5 8.5E-5 - - - - - - - - - - 1 RU105 6.8E-8 1E-4 6.3E-4 -- - - - - - - - RU106 1.0E-13 1E-5 1.0E-4 -- - -- -- _I AG110M 3.5E-10 3E-3 1.1E-5 --- -- --- i TE129M 1.4E-8 2E-5 7.4E - - - - - - - ? l TE131M 2.4E-8 4E 6.0E-4 ------ -- --- 1
-q
))
l May 1986 l l l' , J
g 4a 1 ISOTOPE ACTIVITY MPC FRACTION ~ DETECTOR- CPM NAME CONCENTRATION OF MPC RESPONSE- TOTAL 1 uCi/al uC1/al B/C cpe/ 41/mi cpm A B C-
~
D- E - F . 1 (C1) (MPC1) -(Ci/MPCi)- (CFi) (CiCFi)- TE132 2.9E-9: 2E-5 1.5E-4 1.12E8L 3.3E-1 , 1131 1.4E-6 3E-7 .4.7 1.01E8 1.4E+2-1132 2.5E-7 8E-6 -3.2- 2.63E8 '6.7E+1< " I133 l'.2E-5 1E-6 '. 12.3 9.67E7 51.2E+3 1134 7.2E-10 2E-5 3.6E-5 -2.32E8- 1.7E-1 1135 3.8E-6 4E-6 -9.4E-1 1.17E8! J 4.41+2 - , CS134 5.1E 9E-6 5.7E-2 1.97E8- '1.0E+2 CS136 3.3E-7 6E 5.5E-3 2.89E8: 9.4E+1'- CS137 1.3E-6 2E-5 6.6E 7.32E7- 9.4E-1 t CS138 8.4E-12 --- - --
'1.45E8 1.2E-3 l BA140 1.3E-7 -2E-5 6.6E-2 4~. 99E7- 6.6E+0, LA142 3.2E-9 3E-6 1.0E+3- - - -- ---
.4 1.0E-8 '9E-5 '1.1E-4.
CE141 ---- ----- CE143 7.6E-9 4E-5 1.9E-4: ----
- CE144 7.6E-9 1E-5 1.9E 3 '.03E7 . 1.0E-2 PR143 1.4E-8 '5E '2.8E-4 ND147 1.0E-9 .6E-5 1.7E-9 - - -
W187 6.3E-8 6E-5 1.0E-3 - - - - - ! NP239 2.3E-6 1E-4 2.3E-2 o --- -- , TOTALS 2.14+1 2.4E+3
'l For the example tank, permissf.ble discharge flow to ensure a concentration '
less than MPC in the discharge canal would bes-g . 10,200
- 1 = 324' spa- -
~2.1El
- 1.5 Since maximum obtainable Liquid Radwaste discharge flow 'is 165 spe, this y value would be used for the discharge, Land; for ' calculation of the ala re ,
setpoint. The Liquid Radweste Effluent-Radiation Monitor Alara Setpoint equation ist-A16ra Setpoint = [0.8*F/f*I1(Ci*CFi)]/[Ii(Ci/MPCi)] +. Background. Where the. Alara Setpoint is . in cpe, - F . is 10,200 spa, Ii(Ci*CFi) is 2.4E+3 cym, f is 165 sps and Ii(ci/MPC1) is 2.1E+1 unitiese. These values yield . an Alara Setpoint of 5.7E+3 cpe above background,=while the expected detector respmse is 2.4E+3 cpu. It should be noted that ' the lack of. detector response data for many of the nuclides makes this calculation conservative. Additionally it should be noted that: if grab sample analysis of the tank ' indicates that no activity detectable above the LLD requirements of Table 4.11-1 then the .- Liquid Radweste Effluent Radiation , Monitor Alara Setpoint will be set at less than 355 cpe above. Background, cpm. This is the same as the service Leer monitors initial- alarm , setpoint, see section 2.1.3.2. October 1986 _ _ _ . _. ._. ... .._ -_.________ _ __________t
. - ~b
\;
2.1.3.2 Service Water and Cooling Tower-Blowdown' Effluent Monitor '. Service Water A' and B 'and tho' Cooling Tower . Blowdown are pumped to the I discharge tunnel which;in turn flows directly to Lake Ontario. ' Normal flow rates for each Service Water Pump'is 15,000 gym while that for the Cooling. Tower Blowdown is 10,200 syn. Credit is ~ not . taken for any. dilution l of: , these individual effluent streams.- t The radiation detector is a sodium. iodide crystal. -It' is ~ a- scintillation
' device. The crystal is. sensitive to ' gansaa and beta , radiation. l Howevet', - .
because of the metal walls; in its . sample chamber and' the ' absorption? I characteristics of water, the monitor is not particularly sensitive. to beta , , radiation. . r. Detectorf response El(CitCFi)' will ' initially be assumed tot correspond to 3 that calculated by the manufacturer. However, this will' be': evaluated' within 90 days af ter; Commercial Operation and during. every fuel cycle- by J ! placing a diluted sample of Reactor Coolant (af ter a- two hour decay) in a 'T( ' representative monitor and noting:its gross. count rate. ; Reactor Coolant is
~
chosen because it. represents ~the most likely contaminate of Station Waters. A two hour . decay is chosen by judgement of: the staff of. Niagara ' Mohawk 1 l Power Corporation: Reactor Coolaht with ~ no- decay contains' a ~ considerable ; amount of very energetic nuclides: which would - bias .the detector response
~
term high. However assuming a longer than 2 hour decay.Is s not' realistic'as ' the most likely release' mechanism is a leski through the : Residual- Heat Removal Heat Exchangers which would' contai'i , Reactor = CoolantL during' shutdowns. ' The initial setpoint calculation - is presented :as. both -an example sand ' for. - the purposes of documenting the calculation. It'will be recalculated prior, ; to commercial operation and during every fuel Lcycle whenJ a Radiochemical y analysis of Reactor Coolant is cory:leted for E .bar : determination =as ' required by TS Table'4.4.5-1 or when activity ~is detected in the-respective. effluent stream, f ISOTOPE 2 HR DECAY MPC FRACTION DETECTOR CPM MAME ACTIVITY OF MPC RESPONSE TOTAL-CONCENTRATION pct /al B/C cys/pci/el cpm j pC1/mi A B C D E F (Cl) (MPCI) (C1/MPCI) '(CFi) (CICFi) H3 1.0E-2 3E-3 3.3. -------- -------- F18 1.9E-3 53-4 3.8 -------- -------- WA24 3.7E-3 3E-5 1.2E-2 ~--------- -------- P32 7.8E-5 2E-5 3.9 -------- -------- .i
]
CR51 2.3E-3 2E-3 1.2 -------- -------- MN54 4.0E-5 15-4 4.0E-1 8.42E7 3.4E3 MM56 2.9E-2 1E-4 2.9E-2 1.2E8 3.5E6 FESS 3.95-4 8E-4 ,4.9E-1 -------- -------- FE59 8.0E-5 SE-5 1.6 8.63E7 6.933 C058 5.0E-3 95-5 5.68-1 1.14E8 5.7E5 'l C060 5.0E-4 35-5 1.7E-1 1.65E8 8.3E4 l NI63 3.9E-7 3E-5 1.3E-2 -------- --------
-10 March 1988
l l 1 1 ISOTOPE 2 NR DECAY' MPC FRACTION. DETECTOR CPN- ! NAME ACTIVITY OF MPC RESPONSE - TOTAL . CONCENTRATION uCi/mi B/C c Pm/ uci/al _ c pa - nCi/al: A B . C D E- F- . (C1) (KPCi) (C1/MPC1). (CFi) . ( CiCFi) , 1 NI65- 3.0E-4 1E-4 -3.0 -------- -------- CU64- 1.1E-2 2E-4 5.5El o
- ZN65 7.8E-5 1E-4 7.8 E-1 ------- - --------
d ZN69M 7. 4 E-4 -6E-5 1.2E1 -------- -------- BR83 '1.3E-2 3E-6 ' 4.3E3 -------- ------- - - BR84 2.1E-3 ---- ------ 1.12E8: ,2.4E5! ; RB89- 1.0E-4 ---- ------ -------- - - - - - - - - SR89 3.1E 3E-6 1.0E3: 7.8E3 -2.4E1 3 SR90- 2.3E-4 3E-7 7.7E2 ~- ------- - - - - - - - - - - 5 SR91 6.0E-2 SE-5 1.2E3 -1.22E8- 7.3E6-SR92 6.6E-2 6E-5 1.1E3 8.17E7- 5.4E6 Y91- 1.1E-4 3E-5 3.7 2.47E8 2.7E4 'l Y92 1.3E-2 6E-5 2.2E2 2.05E7 2.7E5 ~! Y93- -1.0E-2 3E-5 3.3E2 ---- - ZR95 4.0E-5 6E-5 6.7E-1 8.35E7 < 3.3E3 ZR97 - 2.9E-5 2E-5 1.5 -------- -- - NB95 4.1E-5 1E-4
- 4.1E-1 8.5E 7 3.5E3 M099 2. 2 E-2 4E-5 5.5E-1 2.32E7- 5.1E5 TC99M 2.2E-1 3E-3 7.3 E1 2.32E7 - 5.1E 6 l RU103 5. 4 E-5 '8E-5 6.8E-1 -------- -- ----
RU105 4.5E-3 1E-4 L 4.5 El -------- ' '- RU106 8. 4 E-6 1E 8.4 E-1 -------- - - -t AG110M 6.0E-5 3E-5' 2.0 -------- - - - TE129M '1.1E-4 '2E-5 5.5 - - - - - - - - - TE131M 7.7E-4' 4E-5 6.8 -------- - - - - - - - - TE132 c.8E-2
~
2E-5 2.4 E3 - 1.12E8' 5.4E6 1131 1.3E-2 3E-7 4.3E4 1.01E8 1.3E6 I132 1.2E-1 8E-6 1.5E4 -2.63E8 3.2E7 ..!
' I133 1.5E-1 '1E-6 1.5E5' 9.67E7 1.45E7 !
l 1134 8.0E-2 2E-5 4.0E3 2.32E8 1.86E7 j 1135 1.4E-1 4E-6 3.5E4 - 1.17E8 1.6E 7 ! CS134 1.6E-4 9E-6 1.8E1- 1.97E8- 3.2E4 CS136 1.1E-4 6E-5 1.8- 2.89E8 ' 3.2E4 CS137 - 2.4 E-4 2E-5 1.2E1 7.32E7 1.8E4 CS138 1.4E-2 ---- ------ 1.45E8 2.0E6 ; BA140 9.0E-3 2E-5 4.5E2 4.99E7 4.5E5 7.1E-3 LA142 3E-6 2.4E3 ----- - ----- CE141 9E-5 CE143 4E-5 CE144 8.1E-5 1E-5' 3.5 1.03E7 3.6E2 PR143 5E-5 ND147 6E-5 ' W187 6E-5 NP239 2.3E-1 1E-4 2.3E3 ------ ----- TOTALS 2.7E5 1.ZE5 > May 1986 ' i,
l-1 i p The Service Water Effluent Radiation Monitor Alara Setpoint equation is l' i Alare Setpoint = [0.8*Ii(CiaCFi)]/[I1(C1/MPC1)] + Background. t I Where the Alara Setpoint- is ' in cps, .Ii(Ci*CFi). is- 1.2E81 cps, 'and E(C1/MPC1)' is 2.7E5 unitiese. These values yield : an Alara Setpoint of ; 3.55E2 cpu above background.- It should be' noted that ' the lack. of detector response data for many of the nucildes makes this' calculation conservative. - ; 2.2 Liquid Effluent Concentratica Calculation This calculation documents compliance with TS Section .3.11.1.1 The concentration of radioactive material released in111guif ef fluents . to ' UNRESTRICTED AREAS .(see Figure 5.1.3-1); shall' be. ' limited: to the . , concentrations specified:in~ 10 CFR 20; Appendix B, . Table II,' Column 2, for .i radionuclides other than dissolved or entrained noble gases. For: dissolved or entrained noble gases, the: concentration shall be. limitedt.to 2 2 10E-L sierocurie/mi total activity. The concentration of radioactivity fros Liquid Radwaste, ServicecWater 4 'I and B and the' Cooling Tower Blowdown are included 'in' the calculation.. The;' calculation is - performed for a specific period..of time. f No credit taken 1 for averaging or totaling. The . limiting 1 concentration "is ' calculated a s . j follows: MPC Fraction = Il [- Is(Cis*Fs)/(MPCi*Is(Fs))) Where: MPC Fraction = The = limiting concentration Jof J 10: CFR 420,~
. Appendix B, ' Table II, Column 2, for's radionuclides other' than- dissolved or:
entrained noble gases. For noblef gases',Jehej concentration: shall' be limited ; to 2 :x 10E-4 ' ? nicrocurie/a1 total. activity, unitiess~ _ l l Ci s =
' The concentration of' nuclide iiin particular 1 effluent: stresa e, aci/al . ,
h =- The; flow rate- of. a particular e f flue'nti stream s, spa' MP Ci = The limiting concentration' of 'a specific' l nuclide i from 110CFR20, . Appendix' b , Table: l L II, Column 2 (noble gas limit is 2E-4), i uCi/mi l Is(Cis*Fs) = The, total activity rate of'nuclide i', in all-the effluent straans e , . uCi/mi
- gps .
Is(Fs) = The total flow rate of- all effluent' streams s, gps. A value of less than one f or MPC f raction is considered; acceptable f for. compliance with TS Section ~3.11.1.1. l 1 1 i May 1986 4
,.-,n,s , - , ..y.. , , ,,
.m -
i 2.3 Liquid Effluent Dose Calculation Methodology i i This calculation' documents. compliance with . TS Section: 4.11.1.2 'and: 4.11.1.3.1' for doses'due to liquidt releases. It is completed once ; per sonth to assure that TS Section 3.11.1.2 and 3.11.1.3 are not exceeded:- i The - dose _ or dose commitment to a : MEMBER OF THE PUBLIC from radioactive i materials in liquid ' effluents released, f rom each unit, to UNRESTRICTED -~ ! AREAS (see' Figure 5.1.3-1)'shall be limited:
}
- a. During any calendar quarter to:less. than or equal tocl.5 area to the - l
~
whole body and:to less than or equal to 5 aren to any organ, and: (
- b. During any calendar year to less! than or equal. to 3 ares lto' the whole.
body and to less than or squal to 10 area to any organ. The liquid radwaste treatment systen shallL be OPERABLE, and appropriate f portions of - the systen shall be used to reduce releases of radioactivity? when the - projected doses due . to' the liquid ef fluent ,' f rom ' the - unit, . to UNRESTRICTED AREAS (see figure 5.1.3-1) would exceed 0.06 area to the whole - body or 0.2 area to any organ in a 31-day period.- ; Doses due to Liquid Effluents are calculatsd-monthly for the fish ingestion s and drinking water pathways from all detected nuclides in liquid effluents-released to the unrestricted areas using the - following expression fron'> - NUREG 0133, Section 4.3. Dt = ' Ii ( Ait* !L*(dT1*C11* F1)] Where: Dt = The cumulative dose commitment to the tota 1L body 'or any organ, t ' from the liquid effluents for the _ total time period; t1(dT1), area dT1 = The length of the 1 th time' period over which Cil and F1 are , averaged for all liquid releases, hours Cil = The average concentration of radionuclide, 1. 'in undiluted' liquid effluents during time period dT1 from any liqui'd release, uC1/a1 I i-Ait = The site related ingestion dose comaltaent factor' to the total I body or any organ t for each identified principal gassa_ or beta i emitter, area /hr per uCi/al. Table 2-2. I F1 = The near field average dilution factor for Cil during any liquid l effluent release. Defined as the ratio of the nazimus undiluted liquid waste flow during release to the product of ~ the average
. flow from the site discharge structure to unrestricted receiving waters times 5.9. (5.9 is the site specifice applicable- f actor for the mixing effect of the discharge structure.) See- the Nine -
Mile Point Unit 2 Environmental- Report - Operating License ; Stage, Table 5.4-2 footnote 1. May 1986 ,
= _-
Example Calculation -_ Thyroid-A sample of a radwaste tank indicates I-131 and H-3 concentrations of 1.5E-6 and -. 8.9E-3 uCi/cc respectively. The tank contains 20,000 gallons - of _ waste to be discharged. The tank is discharged at 165 spa and there .is-30,000 spa of available dilution water: De = ti[ Aitat1(dtleci1*F1)). Where De . area is the dose to organ t. Ait_ area /hr per uCi/al; is the ' t ingestion dose' commitment factor, dT hours is the time interval over which the release occurs, Ci uCi/a1 is the undiluted concentration of nuclide.-i - in the release and F1- unitiess is the dilution factor for the release. From Table 2-2 Ait is 7.21E4 and 3.37E-1 area /hr per _ uCi/al: respectively for I-131 and H-3 dose .to - the thyroid. From the discharge 'and dilution flow rate, F1 unitiess can be calculated j F1l= 165 spa /(30,000gpa *5.9)- = 9.32E-04. - , From the tank volume and discharge rate the length.of time required for the.- , discharge ist dT = 20,000 gal /165 spa = 121.2 min' = 2.02 hr These values will yield 2.04E-4 and S.65E-6 area for I-131 and H-3'I respectively for the thyroid when inserted into-the equation for Dt. Thus the total dose from the tank is 2.06E-4 area to . the thyroid. . The dos e ' limit to the maximum exposed organ is specified. by TS Section 3.'11.1.2 and ~ ' 3.11.1.3. 2.4 Liquid Effluent Dose Factor Derivation Ait Ait area /hr per uC1/mi takes into account the dose from ingestion of fish -l and drinking water. -It;should be noted that the fish ingestion pathway is the most significant pathway for doser from liquid- effluents. The water consumption pathway =is included for consistancy with NUREG 0133.- Drinking water is not' routinely sampled - as part of the Environmental _ Monitoring-Program because of its insignificance. The above equation for calculating dose contributions requires the use o f dosa factor Ait for each nuclide, i, which embodies the dose factors, pathway transfer factors (e.g., bioaccumulation factors) . pathway usage ' f actors, and dilution factors for the points of pathway origin. The adult' ' total body and-organ dose factor for each radionuclide will be used from Table E-11 of Regulatory Guide 1.109. The dose factor equation for a fresh-water site ist Ait = Ko*(Uw/Dw + Uf*BFi)*DFi i May 1986
c > l where: Is the composite dose . parameter for the total? body or organ of
~
Alt = an . adult ' for nuclide, 1. . for all appropriate ' pathways. mres/hr per acl/ml . I l .Ko -= Is the unit _ conversion _ factor, 1.14E5=1E1086pci/pci x ' 1E3 '
- al/kg -
- - 87f0 hr/hr t
Uw = 730 kg/yr, adult water consumption , Uf =- 21 kg/yr, adult fish consumption- > BFi 4 Bioaccumulation. factor for .nuclide, 1, in fish., pCi/kg _perl PC1/1, from Table-A-1 of RG 1.109 ( l' Dose conversion f actor = for- nuclide, _ 1, for, adults in respective ? DFi = organ, t, in mrom/pci, from Table E-11 of.RG 1,109 7 ., Dw . Dilution factor from- the near. fleid' area within one-quarter mils ( of the . release point to the potable water intake for the adult ] l water c on s umpt ion '. . _ Thi's ' is the Metropolitian Wat'or Board,4 Onondaga County intake structure located west of: the City- of ', Oswego. From the NMP-2 ER-OLS Table 5.4-2 footnotei3 this va19e is 463.8. However the near field dilution factor, footnote'l is' 5.9. So as to not take double ' account i of - the near ' fleid dilution the value used for Dw is 463.8/5.9 or 78.6, unitiess.' Inserting the usage factors of RG 1.109 as appropriate 'into the ~ equation gives the following expression: Alt = 1.14E5*(730/Dw + 21*BF1)*DF1. Example Calculation For I-131 Thyroid Dose Factor for exposure from Liquid-Effluents: DFi = 1.95E-3 mrem /pC1 BFi = 1.5El -pC1/Kg per pC1/1 UF = 21 Kg/yr Dw = 78.6 unitiess Ko = 1.14E5 These values will -yloid an Alt Factor of ~ 7 '. 2154 mrem-al por pCl-hr_as listed on Table 2-2. It should ' be noted ' that only a limited number of i nuclidss are listed on ' Table 2-2. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in- a revision to the ODCE. 2.5 sampling Representativeness This section covers TS Table 4.11.1-1 note b concerning thoroughly- mixing each batch of 11guld radwaste prior to sampling. There are four tanks in the radwaste system designed to be discharged to the discharge canal. .These tanks are labeled 4A, AB, 5A, and 58. , Liquid Radwaste Tank 5A and 5B at Nine Mile Point Unit 2 contain a sparger spray ring which assist the mixing of the tank' contents while it is being" recirculated prior to sampling. This sparger effectively sizes the ~ tankj four times faster than simple recirculation. t May 1987
1 Liquid Radwaste Tank 4A and' 48 contain a aizing ring . but no- sparger. 3 No credit is tamen for - the : aixing effects; of the ring.- Normal
. recirculation flow is 150 spa for tank 5A and 58,110- spa for tank; 4A :
and 4B wnile wea'tinn contains up to J 25,000 gallons altnougn the 3-entire . contents . are . not discharged. To J assure that the tanas 'are ,
- adequately aired prior to sampling, it is- a plant requirement that'tne -
tank be recirculated for ene time required to pass 2.5 times the volume of the tanks s Recirculation Time = 2'.5*T/R*M - f 3 daere , Recirculation Time . Is the etniaua time to recirculate the Tank,' ain' 2.5 Is tne plant requirement', unitiess L T- Is tne tank volume, gal R Is the 1. circulation flow rate, spa. '
' t; M
Is the factor that takes into account the-a1xing_of the sparger, unitless, four for tank! 3' l 5A and B, one for tank 4A and B. J q Additionally the Alert Alara setpoint . of the' Liquid Radwaste . Effluent Aadiation Monitor is set at a value corresponding to not' more than - twice its calculated response to the grab sample.- Thus this radiation , monitor will alars if the . grab sample is significantly . lower in activity than any part of the tank contents being' discharged, a Service dater A and B and tne Cooling Tower Blowdown are sampled from the radiation sonitor on eacn respective streas. These monitors continuously withdraw a ' sample and - pump it. back to ene effluent-stream. The length of tubing between the continuously flowing sample and the sample l spigot contains less than 200ml wnica is adequately: purged by requiring a purge of at least 1 liter when_ grabbing 'a sample. 2.6 Liquid Radwaste Systes Operation' Tecnnical Specification 3.11.1.3 requires the Liquid Radwaste Treatment System to be OPERABLE and used when projected doses due to liquid radweste would exceed 0.06 area to the whole body or 0.2 area to any organ in a 31-day period. Cumulative doses will be determined at least once per 31 days (as-indicated in.Section 2.3) and doses will also fully be projected if the radwaste treatment systems are - not being utilized. Full utilization will be determined on tho- basis of utilization of the indicated components of each process streas to-process contents of~the respective system collection tanks:
- 1) Low Conductivity (Waste Collector): Radwaste Filter (see Fig.
2-2) and Radwsste Desin. (see Fig. 2-3)
- 2) Hign Conductivity-(Floor Drains)i Floor Drain Filter (see Fig.
2-5) or Waste Evaporator (see F1j. 2-6) May 1987
-r- e v -- -
<w- ,o
3)- Regenerant Wastat~ Regenerant Evaporator (see Fig. 2-8): . NOTE:- 'Regenerant Evaporato r and Waste ' Evaporator may - be ~ used' interchangeably. The dose- projection indicated above will be performed in accordance- with. the methodology of 'Section 2.3 when ever Liquid Waste:is being discharged. ! without treatment in order to determine that the=above dose limits-are not , exceeded. ' r Ef I l ; i j ^t 1 i
-l t
I~ l l l l l May 1986 l l 1
1 1 L TABLE 2-1 i
- LIQUID EFFLUENT DETECTORS RESPONSES * '
5 NUCLIDE (CPM / uC1/al) x '108 Sr 89 ~ 0.78E-04' ,i S r 91- 1.'22 i' Sr 92 0.817 Y 91 2.47 Y 92 0.205 ~'
- Zr 95 0.835-l- Nb 95 . 0. 8 5 ' ;
Mo 99 0.232 Tc 99a 0.232- , Te 132 1.12 !! Ba 140 -* 0.499 ; ce 144 0.103-Br 84 1.12, .: I 131 1.01 -- ! I 132 2.63 - 1 l I 133 0.967 i ~I 134 12.32' l-1 135 l'.17 ., Cs 134 1.97- I Cs 136 2.89 ! Cs 137 .0.732 'j Cs 138 . 1.45 ' Mn 54 0.842- ; Mn 56 :1.2 :i Fe 59 0.863 l Co 58 1.14 'I Co 60 1.65 I
-1 l- .
-i i
- Values from SWEC purchase specification NMP2-P281F..
1 i t l l -. l L 1 l May 1986
- TABLE.2-2: ,
'Aig' VALUES;- LIQUID * .
ares - m1 ; hr - uC1
]
l; NUCLIDE' T 30DY GI-TRACT BONE. LIVER KIDNEYL .TRYROID ' LUNGE ! H3 3.37E-1 3.37E -- 3.37E-l' 3.37E-1 :3.37E-1 '3.37E-l' Cr 51 1.28 3.21E2' 2.81E-1 7.63E 1.69
~
Cu 64 4.72 TCN-2L 8.57E2 -- l'.01E1. L2.54E1 -- ---- Mn 54 8.36E2' 1.34E4 1
.4.38E3 1.30E3- -- --
Fe 59 9.40E2 8.18E3 1.04E3 2.45E3 '-- -- 6.85E2- - Co 58 2.01E2- 1.82E3 -- 9.00E1 -- -- -- Co 60
~
5.70E2 4.85E3 -- 2.58E2 -- -- -- 7.n 65 3.33E4 4.65E4 2.32E4- 7.38E4 '4.93E4 1-- -- l Sr 89 6.44E2 3.60E3 2.24E4 -- :-- --- -; Sr 90 1.36E5 1.60E4 5.52E5 -- -- -- --- ! Zr 95 5.91E-2 2.77E2 2.72E-1 8.74E-2' '1.37E --- -- Mn 56 1.96El 3.52E3 -- 1.10E2 1.40E2 -- -- TCN-2 t
~
Mo 99 2.05El 2.50E2 --
'l.08E2 '2.44E2 -- --
Na 24 4.09E2 4.09E2 4.09E2 4.09E2 4.09E2 4.09E2 4.09E2 TCN-2 a l I 131 1.26E2- 5.80E1 1.54E2 '2.20E2- 3.77E2 7.21E4 -= l Ni 65 7.53 4.18E2 1.27E2 1.65El- -- -- -- TCN-2 ; I 133 2.78E1 8.21El 5.25El- 9.13E1 1.59E2 1.34E4 -- i Cs 134 5.79E5 1.24E4 2.98E5 7.09E5 2.29ES --
-7.61E4 L
i Cs 136 8.86E4 1.40E4 3.12E4 1.23E5 6.85E4 -- 9.39E3 ! Cs 137 3.42E5 1.01E4 3.82E5 5.22E5 1.77E5 -- 5.89E4' Ba 140 1.41F1 4.45E2 2.16E2 2.71E-1 9.22E-2 -- 1.57E-1 Ce 141 2.48E-3 8.36El 3.23E-2 2.19E-2 1.02E-2 -- -- Nb 95 1.34E2 1.51E6 4.47E2 2.49E2 2.46E2 -- -- La 140 2.03E-2 5.63E3 1.52E-1 7.67E-2 -- --- -- Ce 144 9.05E-2 5.70E2 1.69 7.04E-1 4.18E-1 -- --
- Calculated in accordance with NUREG 0133, Section 4.3.1 June 1987
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f-~~- (DAU) g g,gg, l .I i i , I i S) . I b ( s Of PUMP (11) J NOTES: (1) GLO8E VALVE, ALL OTHER MANUALLY (6) TS. TEMPERATURE SWITCH OPERATED VALVES ARE BALL VALVES (7 CS. CHECK SOURCE (2) REQUIRED ONLY IF SAMPLE FLUID TEMPERATURE EXCEEDS SELLERS (8) Pl. PRESSURE INDICATOR DETECTOA TEMPERATURE REQUIREMENTS (9) Fl. FLOW IN0lC ATOR - (3) N NORMALLY CLOSED (10) PS. FLOW SWITCH (4) CH3 NOFiMAL' Y OPEN (11) DRAIN CONNECTION (S) Tl. TEMPERATURE INDICATlON ODCM Fig. 2-10 FIGURE 11.5 3.
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l NIAGARA MOHAWK POWER CORPORATION NINE' MILE POINT-UNIT:2 FINAL SAFETY ANALYSIS REPORT-
. . . . - _ . . , . . , _ . _ . . _ . - .._ -29 ..May_1986'
_ _ _ ~ __ _._ _ _ _ _ _ __ - - . _ _ _ - . _ _ _ _ __ _ 4y I
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3,0 GASEOUS EFFLUENTS.
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The gaseous : effluent release points are the stack' and the combined: i Radwaste/ Reactor Building. vent. (See Figure 3.5). The stack effluent 1 point includes Turbine Building - ventilation, main condenser = offgas ; (after charcoal - bed ~ holdup), 'and Standby Gas Treatmen t . - Syste n- i exhaust.- . NUREG 0133 and Regulatory Guide 1.109. . Rev.1 were f ollowed- J in the development of this section. , I l' 3.1 Gaseous Effluents Monitor Alara Setpoints ) 3.1.1 Basis ! Technical Specification Section ' 3.11.2.1 and ' 3.11.2.7 provide the' ; basis for the gaseous effluent monitor alara setpoints. TS Section 3.11.2.1 Th e dose rate from radioactive ' materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDRY (see
- Figure 5.1.3-1) shall be limited to the followings
- a. For noble gases less than or equal ~ to 500 ares /yr to the whole body and less than or equal to 3000 ares /yrit o the skin, and
- b. /or iodin e-131, for iodin e-13 3 f o r. tritius, and for all radionuclides with half-lives greater than 8 dayst Less than or l equal to 1500 ares /yr to any organ.
TS Section 3.11.2.7 : i Th e radioactivity rate of noble ' gases measured downstreaa- of the recombiner shall be ' limited to less than or -equal- 350,000 microcuries/second during offgas system operation. 3.1. 2 Setpoint Determination Methodology-The alara setpoint-for Gaseous Effluent Noble Gas Monitors-are based i on a dose rat e' limit of 500 ares /yr to the Whole Body. These i l monitors are sensitive to only noble gases. Because; of this11t. is- l considered impractical to base their alara setpoints on' organ dose ' - rates due to iodines or particulates. Additionally skin dose rate-is 1 never significantly greater than the whole body dose rate. The alarm aetpoint for the Offgas Noble. Gac monitor is based' on a limit of 350,000 L uci/sec . This is the release rate for which a FSAR' accident analysis was completed. At this . rate the Offgas Systen-charcoal beds will not contain enough activity so that their failure and subsequent release of activity will present a significant offsite dose assuming accident meterology. 3.1.2.1 Stack Noble Gas Detector Alara Satpoint Equation: ) 0.8*R* Ii( Ci) l Alara Setpoint < Ii(Ci W Alara Setpoint Is the alara setpoint of the Stack Effluent Monitor, oC1/se c May 1986
~ . . . . . . . - .- --
j 0.8, Is a SOfety Factor, unitiess' , y R; Is a value of.500. ares /yr 'or less depending upon the dose rate f rom other release' points within.the site- ii such that the total rate corresponds to 600 area /yr j l p Ci lIs the concentration of nuclide i, uC1/mi F -Is the Stack effluent flow rate, al/sec ~ t Vi Is the constant ' for each identified noble- gas nuclide ' l accounting for the ;whole body, dose -fronL' the ; elevated finite plume: listed on Table-3-2, stes/yr per uci/sec - l-Ii(C1) Is the total concentration of noble gasi nuclides..in. j L the Stack effluent, uC1/mi ! ti(Ci*Vi') Is- the total . .of.- the - product. of L the each isotope: concentration tines its res constant , aren/yr per al/sec.pective whole Ebodyl plume- 1 It should be noted that the ' flow rate of .the. Sta' ck effluent has been canceled out of the above expression'. . The; equation ratios the. basis, R, to the ' actual ~ dose rate f ros the effluent, ~ F*Il (Ci*Vi'), : and k multiplies the unitiess result by ' the - actual . af fluent : release' rate, F*Ii(C1). Since .the- Stack Effluent Monitor' actually measures release rate in uCi/sec the detector response does not enter'in. 3.1. 2. 2 Vent Noble Gas Detector Alara Setpoint Equationt l 0.8*R*Ii (C1) Alara Setpoint < (X/Q)v* ti( Ci*Ki ) Where: i Alara Setpoint Is the alara setpoint. of - the ' Vent Effluent Monitor, ' uCi/sec. O8 Is a Safety Factor - R Is a value of 500 area /yr or less depending upon the dose rate from other release . points within the site such that the total rate corresponds to < 500 area /yr N Is- the concentration of nuclide 1, uC1/mi > F Is the Vent effluent flow rate.- al/sec - (X/Q)v Is the highest' annual average atmospheric dispersion coefficient at the site boundry as listed in the Final Environmental Statement, NUREG.1085, Table D-2, 2.0E-6 sec/m3 Ki Is the constant for each' identified noble gas nuclide accounting for the whole body dose from the semi-infinite cloud listed on Table 3-3, area /yr per l uci/m3
-31 May 1986
. . . ._ ~ _
- 1;
.j E(C1) Is' the1 totc1 cccc:ntratict1 of noblo 32s nuclidas in !
'the Vent effluent, eCi/mi -l n(Ci*Ki) .Is the total of the product of: the each .~ isotope -
concentration times- its- ~ respective whole :bodyJ immersion constant, area /yr per al/m3 l It should be noted that the flow rate of- the Vent effluent has been canceled out of. the above expression.- The equation ratios the basis, !
-R, to the actual -dose rate from t he = e f fluent , F_* (X/Q)v*Ii(Ci*Ki) - ;
and multiplies the unitiess result by the actual. e f fluent ' release - i rate, F*n(C1). Since the Vent Effluent Monitor actually' measures ) release rate in nCi/see the detector _ response does not ' enter- in'. 3.1. 2. 3 Offgas Pretreatment Noble Gas Detector Alara Setpoint Equationti ' O.8*350,000*2.1E-3*ti(Ci*CF1) .. Alara Setpoint < fati(C1) + Background Where ,; Alara Setpoint Is the alara setpoint- for .'the - offgas pretreatment Noble Gas Detector,-cpa1 ' O.8 Is a Safety Factor, unities s 4 350,000 Is the Technical' specification Limit ' for Offgas Pretreatment, uC1/sec 2.1E-3 Is a unit conversion,'60 sec/ min / 28317 al/CF Ci Is the ' concentration of nuclide, i, in- the Offgas, uCi/a1 CFi Is the Detector response to nuclide 1,- net .; cps / uci/a1 See Table' 3-1 for ailist of nominal . values. See section 3.1.3.3 for discussion 3 I f Is the Offgas Systen Flow rate, CFM i i Background Is the detector response when- its chamber is filled with nonradioactive air, cps: ti(CiCFi) Is the summation of- the. product of the nuclide I concentration .and corresponding _ detector response,_ net cpa ri(C1) Is the summation of the concentration of-nuclides in offgas, uC1/a1 , 1 May 1986
L 3.1.3 Discussion. l The Stack 'at ; Ninei Mile Point Unit 2? receiver the Offgas after 'l l charcoal _ bed delay,- Turbine building ventilation'and the Standby Gas ' Treatment system exhaust. The Standby Gas Treatment._ system - exhaust. the primary containment during normal shutdowns ; and maintains 'aL negative. pressure on' the Reactor Building during' secondaryf 1 containment isolation.- The Standby Gas' Treatment will-- isolat e on-high radiation ? during primary containment purges. - That Stack is' , considered an elevated release because its height, (131a). is more th'an F 2.5 ; ;.iaes the height -of any adjacent buildings' .; Nominal flow rate for the sesek is 10 2,000 CFM. The - Offgas system has .a radiation detector downstream _ of t ho ' recombiners and before the charcoal decay beds. - The : offgas , : af te r - decay, is exhausted to'the main stack. The system will automatically; isolate.. if its pretreataeut radiation nonitor detects levels of. radiation above the alara setpoint. s The' Vent contains the Reactor Building ventilation above / and below .
~
the refuel floor and the Radwaste Building. ventilation- effluents.- The Reactor Building Ventilation will isolate when. radiation monitors l detect high levels of radiation (these- are ~seperate monitors . not - otherwise discussed in the ODCM). It _ is considered .- a combined elevated / ground _ level release because' even though itiis - higher than any adjacent buildings it is not more than . 2.5 times E the ' height . Nominal flow' rate for_ the vent is 237,310 CFM. Nine Mile Point Unit 1 and the James A Fitzpatrick nuclear plants occupy the same site as Nine Mile Point Unit.2. ' Because' 'of ' thel - independance of these plants safety systems. control- rooms ' and operating staffs it is assumed that simultaneous accidents are not likely to occur at the different units. _ However, c there are ,. two release points at Unit 2. It is assumed;that if an accident were to , occur at -Unit 2 that both release points could be involved.- Thus;the f f actor R which is the. basis for the alara setpoint? calculation ' is -. i nominally taken as equal to 250 mRea/yr. If. there are significant releases froa-any gaseous release point on the site ' (>25 area /yr) l for-an extended period of time then the setpoint will be recalculated with an appropriately smaller value for R. Initially, and in accordance with . Specification' 4.3.7.11,L the Germanium multichannel analysis systems of the Stack and' Vent will be calibrated with gas, or with . cartridge standards (traceable - to NBS) in accordance with Table 4.3.7.11-1, note (c) . The quarterly Channel Functional Test will include operability of the 30cc chamber and the dilution- stages to confira monitor high range capability. : (Se e . 1 Figure 3-6). May 1986 i
~ _ _ _ - . _ _ _ _ . . . . , . . _ _ _ _ . 4
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1 3.1. 3.1 Stack Noble Gas Detector Alara Setpoin t ~ This detector is-sade of germanium.; It is sensitive to ohly gassa-radiation. However, because it.is a computer based.- multichannel analysis system it is able: to acurately . quantify 'the activity released in terms of Ci of specific ^ nuclides. -Only pure' alpha,and beta emitters are not detectable, of ~ which there are no conson noble , gases. AE distribution of Noble . Gases corresponding to offgas 3 is , chosen for the nominal alarm setpoint calculation.- ' Offgas is cl.osen because it represents the most - lsignificant . , contaminate of gasecus activity- in the plant. -The- following calculation will be used -for the initial Alara Setpoint. It will be recalculated if a' significant release is encountered. In that : case l
-the actual distribution of noble gases will' be, used' in the' calculation. -The listed activity . concentrations C1, - correspond to -
~
offgas concentration expected with' the plant design limit for fuel; failure. ISOTOPE ACTIVITY PLUME PLUME NAME CONCENTRATION FACTOR FACTOR' ! uCi/a1 arem-sec area /yr. yr uC1. al/sec A B C - D*'( B
- C )
(C1) (VI) (Ci*Vi) KR83 8.74E-2 - - - -- -- KR85 4.90E-4 3.28E-5 1.61E-8 KR85M 1.56E-1 3.21E-3 5.01E- 3 , KR87 5.2 3E-1 9.98E-3 5.22E-3 I KR88 5.32E-1 2.21E-2 1.18E- 2 . KR89 1.63 1.92E-2 3.13E-2 KR90 ----- 1.51E-2 - - - - - - XE131M 3.82E-4 6.55E 2.50E-8 XE133 2.06E-1 5.93E-4 1.22E-4 XE133M 7.3 5E-3 3.44E-4 2.53E-6 XE135- 5.88E-1 6.12E-3 3.60E-3 XE135M 5.91E-1 6.12E-3 3.62E-3 XE137 2.11 2.88E-3 6.08E-3 XE138 1.93 1.33E-2 2.57E-2 AR41 - -- 1.61E-2 --- TOTALS 8.36 9.28E-2 The alara setpoint equation ist Alarm Setpoint = 0.8*R*Ii(C1)/Ii(Ci*V1). ' Where the Alara Setpoint is in uCi/sec, R 'is taken as 250erea/yr, 3 T(C1) is 8.36 uC1/mi and. I(Ci*V1) is 9.28E-2 area /yr per al/sec. These values yield an alarm setpoint of 1.80E4 uCi/sec. May 1986
3.1.3.2 Vent Ef fluent Noble Gas Detector Alara Setpoint - This: detector is made ofigermanium., It is sensitive to only gassa-
~
radiation. 'However, because ~ it is. a computer, based - multichannel analysis . systen ' it is able- toD Laccurately quantif y the Eactivity ' released in teras of i C1 n of specific nuclides.. Only pure alpha . and-beta.ealtter's are not detectabla, of which there are no conson noble. i gases. A distribution .of Noble ' Gases corresponding to thatLempected - with the design limit . forH fue1% f ailure : offgas ~ is chosea f or = the l nominal alara : setpoint J calculation. . Offgasi is - chosen becausei it: 4 represents the most ~ significant contesinate of gaseous activity in -
~ the plant.- The : following calculation will; be.'used for the . initial Alara Setpoint. It will be' rec'alculated if 'a l 61gnificant release is q encountered. .In that case' the actual distribution of noble gases ;
will' be used in the calculation. ' ISOTOPE ACTIVITY IMMERSION IMMERSIO N. NAME CONCENTRATION FACTOR FACTOR uCi/mi area-m3 are e-a3 ' .: l yr uC1 yr 31 . ( A B C D=(B*C)' =l J ; KR83 8.74E-2. 7.56E-2 6.61E-3 KR85 4.90E-4 1.61E-1 7.90E- 3 . l KR85M 1.56E-1 1.17E-3 1.82E2. KR87 5. 23E 5.92E3 - 3.10E 3 KR88 5.32E-1 1.47E4 7.82E3 KR89 1.63 1.66E4 2.71E 4 KR90 ---- 1.56E4 ------- XE131M 3.82E-4 9.15El '3.50E- 2 XE133 2.0 6E- 1 2.94E2. 6. 0 6El ' .; XE133M 7.35E-3 2.31E2 1.84-XE135 5.88E-1 1.8123 1.06E3 s XE135M 5.91E-1 3.12E3 1.84E 3
- XE137 2.11 1.42E3 3.00E3 XE138 1.93 1.83E3 1.70E4 ,
AR41 - - - - - 8.84E3 ----- 3 TOTALS 8.36 6.12E4 The Vent Effluent Noble Gas Monitor Alara Setpoint equation ist Alara Setpoint = 0.FR* Ii(C1)/ [(X/Q)v*Ii(Ci*Ki)) . Where the ' Alara Setpoint is in nCi/sec,'R is 250erea/yr. Ii(C1) is. 8.36 uC1/m1, (X/Q) is 2.0E-6 sec/m3 . and Zi(Ci*Ki) is ' 6.12E4 ares /yr per al/m3. This will yield an alars setpoint of 1.41E4 oC1/sec. 4 1 May 1986 , L, __ . _ _ _ _ . - __ _ _ _ . .-~ .- - ~ ~ ~ -- --- - - - - - - - - - - - - - - - - - - - - - - -
^ ^
4 3.1.3.3 offgas Noble Cas D*tector Alarm Setpoint The Radiation Detector is a sodlum lodido crystal. It is a reletillation device and has a thin mylar window so that it is sensitive to both gamena and beta radiation. Detector response
!!(Cl*CF1) will be evaluated from isotopic analysis of offgas analyzed on a multichannel analyser, traceable to N85, prior to consnercial operation. A distribution of offges corresponding to that espected with the design limit for fuel feilure is used to establish setpoint initially, assualns the nominal response lasted on Table 3-1. The monitor nominal response values will be confirmed during initial calibration using a Transfer Standard source traceable to the primary calibratior performed by the vendor. However, a revision . to the ODCN will contain an . updated distribution : end total detector response based on actual plant esperlences. The initial calculation is presented below.
ISOTOPE ACTIVITY DETECTOR DETRCTOR' NAME CONCENTRATION RESPONSE CPN pC1/ml cpe/pC1/ml cpm A 8 C D (Cl) (CFl) (Cl*CF1) KR83 8.74E-2 ------ +------ KR85 4.90E-4 4.30E3 2.11 KR85N 1.56E-1 4.8053 7.50E3 KR87 5.23E-1 8.00E3 4.18E3 KR88 5.32E-1 7.60E3 4.04E3 KR89 1.63 ------ ------- KR90 ------ ------ ------- KE131N 3.82E-4 ------ ------- 7133 2.06E-1 1.75E3 3.60E2 Is133M 7 . 3 5 E ** ------ ------- XE133 5.88E 1 5.10E3 3.00E3 XE135N $.915 1 .---- ------ XE137 2.11 4.10E3 1.71E4 XE138 1.93 7.10E3 1.37E4 AR41 ------- ------ ------- TOTALS 8.36 4.99E4 The Offgas Noble Cas Monitor Alarm Setpoint equation 1s: Alarm Setpoint . 0.8*350,000*2.1E-38!1(CinCF1)/(f*I!(Cl)) + Skg. Where the Alarm Setpoint is in eps, !!(Cl*CF1) is 4.99E4 cm , f is 25CFM and !!(Cl) is 8.36 pC1/cc. This will yleid an alarm setpoint of 1.40E5 eps above background. Particulates and Iodines are not included in this calaulation because this is a noble gas ! monitor. To provide an alare in the event of f ailure of the offgai- ystem flow Instrumeetation, the low flow alarm setpoint will be set at or above ; 10 sefm. (well below normal system flow) and che high flow alarm TCN-setpoint will be set at or below 120 scfs, which is well above espected steady-state flow rates with a tight condenser.
-36 September 1987;
4 3.2 Geseous Effluents Dose Rate Calculation This section covers TS Section 4.11.2.1.1 and 4.11.2.1.2 concerning the calculation of dose rate from gaseous effluents for compliance with TS Section 3.11.2.1. TS Section 3.11.2.1 The dose rate from radioactive materials released in gaseoss effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the followings
- a. For noble gases: less than or equal to 500 meen/yr to the whole body and less than or equal to 3000 ares /yr to the skin, and
- b. For iodine-131, iodine-13 3, for tritius, and for all radionuclides in particulate form with half-lives greater than 8 days less than or equal to 1500 area /yr to any organ
- 3. 2.1 Whole Body Dose Rate Due to Noble Gases This calculation covers TS Section 3.11.2.1.a (for whole body) and 4.11.2.1.1. The dose f rom the plume shine of elevated releases is -
taken into account with the factor V1. The dose from Vent releases takes into account the exposure from immersion in the semi-infinite cloud and the dispersion fros the point of release to the : receptor, which is at the East site bcundary. The release rate is averaged over the period of concern. The factors are discussed in greater detail later. Whole body dose rate due to noble gases ares /yr = ti (Vi*Qie + Ki (X/Q)v*Qiv] Where V1 Is the constant accounting for the ganan radiation from the alevated finite plume of the Stack releases for each identified noble gas nuclide, 1. Listed on Table 3-2, - area /yr per uC1/sec l 1 Qis Is the release rate of each noble gas nuclide, i, f rom the Stack release averaged. over the time period of concern, uct/se c K1 Is the constant accounting for the whole body dose rate from innersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed on Table 3-3, area /yr per uCi/m3 May 1986 l )' ~..
(X/Q)v Is- the highest calculated annual average relativ e concentration at or beyond the site boundry for the Vent. Final hvironmental Statement, NUREG 1085, Table D- 2, i 2.0E-6 se e/m3 I l Qiv Is the tne ,se rate of each. noble gas nuclide, i, f rom the Vent release averaged over the time period of concern, l 41/see ! Example Calculation i As sume an analysis of the Stack and Vent Efflut.sts indicate that ! 1.81E4 and 1.26E4 uCi/see of Xe-133 are being released from each point respectively. From Table 3-2, Vi is ' $.93E-4 ares /yr per l uC1/sec . From Table 3-3 Ki is 2.94E2 area /yr per uC1/m3. (X/Q)v ! is 2.0E-6 sec/m3. These values yield a whole body dose rate of 10.7 l and 7.41 area /yt from the Stack and Vent respectively for a total of i 18.1 area /yr. This value is added to the whole. body dose rates l obtained fros' the Nine Mile Point-Unit 1 and James A. Fitzpatrick j plants to obtain the site dose rate to the whole body from noble gas i releases. The whole body dose rate due to noble gases is specified ,
]
by TS Section 3.11.2.1.a. 1 l 3.2.2 Skin Dose Rate Due to Noble Gases j i l This calculation covers TS Section 3.11.2.1.a (for skin),and 1 4.11.2.1.1. For Stack releases this calculation takes into account - the exposure f rom beta radiation of a semi infinite cloud by use of j the factor Li. Additionally the dispersion of the released activity ! from the stack to the receptor is taken into account by use of the factor (X/Q). Ganas radiation exposure frca the overhemd plume is taken into account by the f actor 1.181. ] For vent releases the calculations also take into account th e exposure from the beta and samma radiation of the semi infinate cloud by use of the factors Li and 1.1Mi respectively. Dispersion is taken { into account by use of the factor (X/Q). The release rate is ; averaged over the period of concern. The factors are discussed in ' greater detail later. l l Skin dose rate due to noble gases : I area /yr = ti ( (Li*(X/Q)s + 1.1*Bi)*Qis + (Li + 1.1*Mi)*(X/Q)v*Qiv) ) Where: Li Is the constant to take into account the skin dose due l to each noble gas nuclide, i, from immersion in the : ! seal-infinite cloud, stes/yr per uC1/m3 l ( i M1 Is the constant acc6unting for the air gamma dose rate ! from immersion ir. the semi-infinite cloud for each , identified noble gas nuclide, 1. Listed on Table 3-3, 1 ara d/yr per uCi/m3 1.1 is a unit conversio n { constant, area / rad ; 1 I May 198 6 1
. . . . . - . - . - , . - - - . - - - - . ~ - - . - . . - . , - . - _ _ _ _ _ _ - - _ _ _ _ - _ _ -
i l , 81 Is. the constant accounting for the air gamma dose rate ' from. esposure to the overhead plume of elevate d releases of each identified noble gas nuclide, 1. Listed on Table 3-2, arad/yr per uCi/sec. (X/Q)v Is the highest calculated annual average- relative coricentration at or beyond the site boundary for the l Vent. Final Environmental Statement, NUREG' ;G85, i l Table D-2, 2.0E-6 sec/m3 (X/Q)s is the highest calculated annual - average relative concentration at or beyond the site boundary for the Stack. Final Environputal Statement, NUREG 1085, Table D-2, 4.5E-8 se c/m i Qiv Is the release rate of each noble gas nuclide, i, from! l the Vent release averaged over the time period . o f. j concern, uC1/se c l
\
Qis Is the rWease rate of each noble gas nuclide, i, f ron l the Staca release averaged over the time period of . , concern, uC1/se e ! i Example Calculationi t j As sume an analysis of the Stack and Vent Effluents indicate that j 1.81E4 and 1.26E4 uCi of Xe-133 are released from each point. From Ta ble 3-2, Bi is 6.12E-4 arad/yr per uC1/sec. From Table 3-3, Li and Mi are 3.06E2 and 3.53E2 aren.arad/yr per uCi/m3 respectively. ! (X/Q) for the Stack and Vent is 4. 5 E-8 and 2.0E-6 sec/m3 l respectively. Wese values yield a skin dose rate of 12.6 and 17.5 ) ares /yr for the Stack and Vent respectively for a total rate of 30.1 area /yr. Bis value is added to the skin dose rates obtained- from Nine Mile Point-Unit 1 and the James A. Fitzpatrick plants to obtain 1 the site dose rate to the skin from noble gas releases. The skin i dose rata limit due to noble gases is specified by ' TS Section ) 3.11. 2.1. a . ' 3.2.3 Organ Dose Rate Due to I-131,1-133 Tritium, and Particulates. with Half-lives greater than 8 days. l Wis calculat%n covers TS Section 3.11.2.1.b and 4.11.2.1.2. The -) factor Pi takes into account the dose rate received from the ground plane, inhalstion and food (cow milk) pathways. We and Wy take into account tha atmospheric dispersion from the release point to the 1 i location of the most conservative receptor for each of the respective i pathways. The release rate is averaged over the period of concern. The factors are discussed in greater detail later. 1 1 l I May 1986
.-.____-_m_____ _ _ _ - _ _ _ _ _ _ _ _ _ ____ _ _ _____._ _ - , _ _ _ _ _ _ _ _ . . ,..- . - . , _.. ,_. _.. .., ._._ ._ . _.. .~... _m .. . . . _ . ,
_ . . . _ . _ _ _ . _ _ _ __ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . ~ _ . ~ . _ . . . _. . l Organ dose rates . due to iodine-131, iodine-133, tritium and all i radionuclides in particulate form with half-lives greater than 8 days ares /yr = rp [ ti Pip [WsQis + WyQiv] ] , Where l 4
- Pip Is the factor that takes - into account the dose to an '
individual organ from nuclide i P'*h"'7 '- inhalation pathwpy, ares /yr ' per throul.h vCi/a For ground and l food pathways, a%rea/yr per aci/sec . l ti Is the summation over all nuclides, i Ip Is the summation over all pathways Ws , Wy Are the dispersion parameters for stack and vent regease resqectively for each pathway as approriate se c/a or , 1/m'. See Table 3-2 2. Qis, Qiv Are the release rates for nuclide i, from the stack and ; vent respectively 41/sec. Example Calculation Assume an analysis of the Stack and Vent Effluents indicate that i 1.84E-1 and 1.26E-1 .C1/sec of I-131 are released from each. point re spectively. From Table 3-4 thru 3-6 and 3-22 the f ollowing ' table
- can be made t ORGAN Pi GROUND Pi INHALATION Pi FOOD '
or 32-are a/yr utes/yr a2-1are a/y r
? ACTOR uC1/sec uCi/m3 uCi/sec T BODY 2.46E7 1.96E4 1.43 E9 SKIN 2.98E7 - - ----
BONE -- 3.79E4 2.77E9 LIVER - - - - 4.4 AE4 3.26E9 THYROID - - - - - 1.48E7 1.07E12 KIDNEY - - - - - 5.18E4 3.81E9 LUNG - - -- --- 0I-LLI ---- 1.06E3l 1.16E8 Ws 1.34E-9 8.48E-9 3.64E-10 WV 2.90E-9 1.42E-7 4.73E-10 WsQe&WyQv 6.12E-10 1.95E-8 1.27E-10 NOTE: The Dispersion Parameters given in Table 3-22 will be revised based on the results of environmental surveys and meteorological data. From these values the following table of dose rates (area /yr) can be calculated May 1986 '
i l ORGAN GRL g INHAIATION FOOD TOTAL T BODY 1.51E-2 3.8 2E-4 1.82E-1 1.97E-1 SKIN 1.82E-2 - - 1.82E- 2 BONE ----- 7.39E-4 3.52E-1 3.53E-1 LIVER --- 8.66E-4 4.14E-1 4.15E -1 THYROID - - - - - 2.89E-1 1.36E+2 1.36E+2 KIDNEY - - - - 1.01E-3 4.84E 4.85E-1 Lupo ._-- .- -- .. _ _. GI-LLI - - - - 2.07E-5 1.47E-2 1.47E- 2 In this case the maximum dose rate to an organ is 136 ares /yr to the thyroid from I-131. This calculation would be repeated for all nuclides and age groups then summed for each- age group to obtain the dose rates to all organs. The dose rate limit to the maximum exposed organ is specified by TS Section 3.11.2.1.b. 3.3 Geseous Effluent Dose Calculation Methodology TS Section 3.11.2.2 - The air dose f ros noble gases released in gaseous effluents, . f rom each unit , to areas at or beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the f ollowing.
- a. During any calendar quarter less than ot equal to 5 arad for gaana radiation and less than or equal to 10 arad for beta radiation, and
- b. During any calendar years Less than or equal to 10 arad fer ganea radiation and less than or equal to 20 arad for beta radiation.
May 1986 1 1 l l , ., , , , , . - - . . -
TS S:ctien 3.11.2.3 ) We dose to a MEMBER OF THE PUBLIC f rom iodine-131, iodine-133, tritius, and all radioactive material in particulate form with half-lives greater than 8 days in gaseous effluents released, from , each unit , to areas at or beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be lialted to the followings
- a. During any calendar quarter: lass than or equal to 7.5 area t o any organ and,
- b. During any calendar years Less than or equal to 15 stem to any
- organ.
- TS Section 3.11.2.5 :
i The VENT 11ATION EXHAUST TREATMENT SYSTEM shall be OPERABLE and appropriate portions of.this system shall be used to reduce releases of radioactivity when the projected doses in 31 days f rom iodine and particulate releases , from each unit , to areas at or beyond the SITE BOUNDARY (see Figure 5.1.5-1) would exceed 0.3 area to any organ of a * , MEMBER OF THE PUBLIC.
- r 3.3.1 Gamma Air Dose Due to Noble Gases This calculation covers TS Section 3.11.2.2 and 4.11.2.2.
- t Gamma air dose due to noble gases released' is calculated monthly.
The factor Mi takes into account the dose from innersion in- the semi-infinite cloud of the vent release. The factor X/Q takes into account the dispersion of vent releases to the most conservative l location. The factor Bi takes into account the dose from exposure to i the plume of the stack releases. The release activity is totaled over the period of concern. The factors are discussed in greater detail later. Gamma air dose due to noble gases ! arad = Ii [Mi(X/Q)y Qiy + Bi Qge) Where the constants have all been Note that since Q is expressed as uCi/sec , previously defined. the constant 3.17E-8 see -l , given in NUREG-0133, section 5.3.1 may be omitted, provided that the annual dose calculated is divided by 4 to yield -quarter dose, or 12 to yield monthly dose, as applicable. Example Calculation
, As sume an analysis of the Stack and Vent Effluents indicate that 1.42E11 and 9.91E10 uci of Xa-133 are released from each poin t respectively over the last quarter. This correlates to 1.81E4 and
- 1.26E4 uCi/sec respectively. Pros Table 3-2, si is 6.12E-4 arad/yr per uC1/sec. From Table 3-3 Mi is 3.53E2 arad/yr per uC1/m3.
(X/Q)v is 2.0E-6 se c/m3. Rose values yield a gassa air dose rate of 11.1 and 8.9 arad/yr from the Stack and Vent respectively for a total of 20.0 arad/yr or 5.0' arad for the quarter. The samma air dose limit due to noble gases is specified by TS Section 3.11.2.2. May 1986 .__ ___ _ . _ _ . ~ _ . _ ___ - . _ _ _ _ ._ __ _ ,- _ _ . ~ . _ . _ _
_ _ _ _ g 1'l
'do 1
l 3.3.2 leta Air Dose Due to Noble Gases This esiculation covers TS Section 3.11.2.2 and 4.11.2.2. '
- Beta air dose due to noble gases released is calculated monthly. The l factor Ni takes into account the dose from immersion in the cloud of l all the releases. - The f actor X/Q takes into account the dispersion !
of releases to the most conservative location. The factors are l discussed in greater detail later. I 1 ; Beta air dose due to noble gases arad = IiWi [(X/Q), Qiy + (X/Q), Qi s) i Where the constants have all been previously defined. .; I I l Example Calculation , l J As sume an analysis of the Stack and Vent Effluents indicate that i 1.42E11 and 9.91E10 901 of Xe-133 are released from each poin t l respectively over the last month. This correlates to 1.81E4 and ! j 1.26E4 ici/sec respectively. From Table 3-3, Ni is 1.0$E3 mrad /yr , per "Ci/m3. (X/Q) for the Stack and Vent is 4.5E-6 and 2.0E-6 ! , sec/m3 respectively. These values yield a beta air dose of 0.9 and
- 26.5 arad/yr for the Stack and Vent respectively for a total of 27.4 !
I arad/yr or 6.8 arad over the last quarter. The beta air dose limit I due to noble gases is specified by TS Section 3.11.2.2. 3.3.3 Organ Dose Due to 1-131, 1-133 Tritius and Particulates with half-lives greater than 8 days. l l This calculation covers TS Section 3.11. 2.3, 3.11. 2.5, 4.11.2.3, an d , 4.11.2.5.1. Organ dose due to I-131. I-133 Tritiva and Particulates with half-lives greater than 8 days released is calculated monthly. . The f actor Ri takes into account the dose - received f rom the ground ) plane, inhalation, food (cow milk, cow seat and vegetation) pathways. We and Wy take into account the atmospheric dispersion ; from the release point to the location of the most conservative i receptor for each of the respective pathways. The release is totaled l over the period of concern. The factors are discussed in greater , detail later. Organ dose due to iodine-131, iodine-133, tritius radionuclides in l particulate form with half-lives greater than 8 !ays f area = 3.1/k-8 Ip [ Il Rip [We Qis + WV Qiv) ) - Wheret j i 3.17E-8 Is the inverse of the raaber of seconds in a year i Rip- Is the factor that takes into account the dose to an individual, organ f rom nuclide 1 through pathway p. , t May 1986 ) l
, , , , , , _ , - - , - - - , . - - - . , .-.,__w- w. . , . . . . .
ti Is the summation over all nuclides 1. l Ip Is the summation over all pathways p. . i We , WV Are the diraarsion parameters for the stack and veng l' respectfvely for each pathway as appropriate se c/a or 1/n . See Table 3-2 2. l Qis, Qiv Are th9 amount of activity of nuclide i released f rom the stack or veut respectively over the pgriod of ; c once rn, uC1. If activity released is given in- l terms of re rate, uCi/sec, then the constant ' 3.17E-8 see"{ ease may be omitted, provided that the ; annual dose calculated is divided by 4 to yield ! s q uarter dose. -or 12. to yield monthly dose, as e applicable. ! Example Calculation
)
As sume an analysis of the Stack and Vent Effluents indicate that i 1.45E6 and 9.9E5 aCi of I-131 are released from each poin t . respectively over the last quarter. This correlates 1.84E-1 and 1.26E-1 uC1/sec respectively. Calculate the. dose to a childs-organs. From Tables 3-8,11,13,16 and 19 the following table can be made: ORGAN Ri-GROUND Ri-1NHAIATION Ri-MILK Ri-MEAT Ri-VEGETATION ~ I or m2-ere s/yr ares /yr a2-erea/yr FACTOR nCi/sec uC1/m3 pC1/sec l
~
TBODY 1.72E7 2.73E4 3.72E5 4.73E6 8.16E7 l SKIN 2.09E7 - -~~- - - --- - l BONE 4.81E4 6.51E8 8.26E6 1.43E8 , LIVER - - - - 4.81E4 6.55E8 8.32E6 1.44E8 THYROID ------ 1.62E7 2.17 E11 2.75E9 4.75 E10 KIDNEY --- 7.88E4 1.08E9 1.37E7 2.36E8 LUNG - - - - - - - - -- - - - - - - - GI4LI ---- 2.84E3 5.83E7 7.40E5 1.28E7 Wo 1.34E-9 8.48E-9 3.64E-10 1.15E-9 9.s 1E-10 $ Wv 2. 90E-9 1.42E-7 4.73E-10 1.86E-9 1.5IE-9 WsQs+WyQv 6.12E-10 1.95E-8 1.29E-10 4.46E-10 3.6u-10 From these values the following table of annual dose (area) can be calculated: ORGAN GROUND INHALATION MILK -MEAT VEGE. . TOTAL ' T T55Y 1.ost-2 s.32E-4 T.TUE-2 G E-3 TY5E-2 Y tfR-2 SKIN 1.28E-2 --- - - --- - - 1.28E-2 l BONE -- - 9.38E-4 8.4 0E-2 3.69E-3 5.18E-2 1.40E-1 LIVER - --- 9.38E-4 8.45E-2 3.71E-3 5.21E-2 1.41E -1 THYROID -- - 3.16E-1 28.0 1.23 17.2 46.7 KIDNEY --- 1.54E-3 1.39E-1 6.11E-3 8.54E-2 2.32E-1 LUNG - - - - - - - - - - - - -- - -- - -- CI-LLI - - --- 5.54E-5 7.52E-3 3.30E-4 4.63E-3 1.25E-2 May 1986 ,
- - , - _ _ - . _ _ _ ~ _ . _ _ - - _ _ . - _ - _ _ . _ - - - ,
. -. . - - - _ - . ~ - . -. -- - - -. . -
la this c:se the assimu3 quart :rly deso to the child org2n is.46.7/4
= 11.7 ares to the thyroid f rom I-131. The calculation would be !
- repeated for all nuclides and age groups and summed to - find the maximum dose to any organ. The dose limit to the nazimum exposed organ is specified by TS Section 3.11.2.3 and 3.11.2.5.
3.4 caseous Effluent Dose Factor Definition and Derivation f . . 3.4.1 Bi and VI- Plume Shine Factor For Gamme and Beta Doses (Table 3-2) ; i Bi (arad/y r per uCi/sec) is calculated by modeling the. effluent l from the Stack as a line source with an elevation above ground equal i to the stack height (131a) From " Introduction to Nucitar Engineering" by 14aarsh, page 410, the flux o at a point a distance of - x from an infinite line esitting S , gammas /see per en ist j o = S/4x. S is proportional to release rate Q - (uci/sec) and inversly to wind . speed U (ca/sec): S = Q/U. De distance of an individual on the ground from the elevated plume is approximately equal to the height of the stack h (meters). .The I ganas radiation from the plume is attenuated by the sir. This is ' proportional to the exponential of the- negative product of the stack . height h (a) and the air attenuation coefficient Uo,1/as ' l exp (-Uo*h). Ris is a conservative assumption because only the portion of the - plume directly overhead is at a distance of h. The bulk 1s . auch further away. ' Additionally, there is a dose buildup factor which, f rom RG 1.109 ; Appendix F-11,12, is equal tot 1+[(Uo-Ua)*Uo*h}/Ua ,- . where Ua (1/a) is the air energy absorption coefficient . ( l May 1986
l l The dose D at a point is proportional to the flux o, energy E (Nev) i of the radiation, air energy absorption coefficient Ua (a-1) and unit conversion constant K i D = K*o*E*Ua. I Substitution in the above formula for flux from an infinite line source yields: l D = K*$*E*Ua/(4*x). Substitution for S yields: ! ! -D = K*Q*E*Ua/[4*z*U). Substitution for x of Stack height h yields: D = K*Q*E*Ua/[4*h*U). Factoring in the air attenuation and corresponding dose buildu p f actors yields. . D = K*Q*E*(Ua+(Uo-Ua)*Uo*h]exp(-Uo*h)/[4*h*U) . Bi is the ganaa air dose received on the ground for a given release rat 6 Q. Thus 1 B = D/Q = K*E*(Us+(Uo-Us)*Uo*h)*exp(-Uo*h)/[4*h*Uj. 1 l Where K = 1.447E4 arad-dis-3 3 /Mev-uCi-yr, U is 5.71 m/sec and the other symbols are as discussed above. To calculate Vi (ares /yr per uC1/sec), the factor to account for the Total Body dose rate for a given release rate Q (uC1/sec) a conversion ratio of 1.1 aren/arad is assumed between tissue and air doses. If the Total Body tissue density Td (ga/cc) is assumed to be Sga/cc (like a rock) and Ut (es2/gn) is the energy absorption for tissue then * ' l V = 1.1*8*exp(-Td*Ut). Example Calculation l Us, Ue and Ut all vary with the energy - of the radiation. Figure 3.5-6 and Table 3.5 -1 (b suscle) of the " CRC Handbook of Radiation Measurement and Protection" list values for the variables. For a 0.25 Mev gamma: Uo = 0.0145 m-1 Ua = 0.0036 a-1 Ut = 0.0306 cm2/ga . May 1986 l l g ew--. ,,m 3 'my.w .s-
,--qm. , . ,--o yy n .,vna ~,,.-,,qe p- w weo.,--
- -. I{
I I i These values will yield a factor of 4.38E-3 and 4.14E-3 ared, ares /yr l per uCi/see respectiTy for B and V. Similarily for the primary i energies of Xe135 the following table is obtainable: ; i CNERGY YIELD B V ! MEV __ arad/yr/uci/sec area /yr/uci/sec { 0.25 0.9 4.38E-3 4.14E-3 0.6 0.03 [ 9.38E-3 8.77E-3 i 0.7 0.01 1.06E-2 9.97E- 3 TOTALS FACTORING IN THE YEILDS: 4.31E-3 4. 0 7E-3 These values correspond to those listed on Table 3-2. It should be l noted that or.ly a limited number of nuclides are listed on Table 3-2. These are the a-ist consono noble gas nuclides encountered in ! e f fluents. If a nuclide is detected for which a factor is not ; listed, then it will be calculated and included in a revision to the ! ODCN. . ! 1 3.4.2 Semi-Infinite Cloud Innersion Dose Factors (Table 3-3) , Ki, Li, Mi and Ni are the factors which take into account the dose f ros innersion in the semi-infinite cloud of gaseous releases. These are taken from RG 1.109, Table B-1, and multiplied by 1E6 to convert f ros units of ares,arad/yr per pct /m3 to area,arad/yr per uCi/m3. L 3.4.3 Dose Rate Factor f o r I-131, 1-133, Tritium and Particulates with Half-lives greater than 8 days. Table 3-4. Ground Plane Pi (a2-erea/yr per uCi/sec) takes into account several factors among these are the dose rate to the total body from exposure to radiation deposited on the ground. (From NUREG 0133, section 5.2.1.2) INSERT SYMBOLS Where K' = a constant of unit coversion,106 pCi/ uci . K" = a constant of unit conversion, 8760 hr/ year. ( 1 1 = the decay constant for the ith radionculide, see -1 t = the exposure period 3.15 x 107 see (1 year) .
=
DFG1 the ground plane dgse conversion factor the the ith radionuclide (area /hr per pCi/s ) . The deposition rate onto the ground plane results in a ground plane concentration that is assuand to persist over a year with radiological decay the only operating removal mechanism for each radionuclide. The ground plane dose conversion factors for the ith radionuclide, DFG i , are presented gn Table E-6 of Regulatory Guide 1.109, in units of area /hr per pCi/a May 1986
. - . . . . . - - . ~ - . - . ,. . - . --. - - . - .. .. . - . - . - - - .
+
Resolution of the units yieldst Pi (Ground) = 8.76 a 10 9 DFGi (1-e'2 i t)/21 Example Calculation For the I-131 total body dose rate factor . for exposure from the l ground ! 21 = 9.98E-7 sec-1 . l DFCi = 2.30E-9 area /hr per Ci/s2 These values will yie.1d a Pi factor of 2.46E7 m2mres/yr per uCi/sec { a s listed on Table 3-4. It should be noted that only a limited number of nuelldes are listed on Table > 3-4. nose are the most ! common nuclides encountered in effluents. If a nuclide is detected i for which a factor is not listed, then it will be calculated and included in a revision to the ODCM. ,) i Pi (a2-ares /yr per uCi/sec) also takes into account the dose rate t o ! the skin f rom exposure to the ground. I Example Calculation For the 1-131 skin dose rate f actor for exposure from the groundt
$1 = 9.98E-7 sec-1 DFGi = 3.40E-9 ares /hr per pC1/s2 These values will yield a Pi factor of 2.98E7 m2sres/yr per uC1/sec as listed on Table 3-4. It should be noted that only a limited I
nubser of nuclides are listed on Table 3-4. Rose are the most conson nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then ith will be calculated and included in a revision to the ODCM. Table 3-5. Inhalation Pi (area /yr per uCi/m3) also takes into account the dose rate to various organs from inhalation exposure. (From NUREG 0133, section
- 5. 2.1.1 )
J Pi = K'(BR) DFAi (ares /yr per uCi/tt3) l Where K' = a constant of enit conversion,106 poi /101. l BR = the breathing rate of the infant age group, in a3/yr. DFAt= the organ inhalation dose factor for the infant age group for the ith radionuclide, in area /pci. The total body is considered as an organ in the selection of DFA i. l May 1986
l the infant group. The infant's breathing The rate is age group taken considered as .1400 m ig/yr from Table E-5 of Regulatory Guide , 1.109. The inhalation dose factors for the infant, DFAt are presented in Table E-10 of Regulatory Guide 1.10 9. in units of ares /pci. Resolution of the units yeilds: Pi (inhalation) = 1.4 x 109 DFA1 . ] 1 Example Calculationt l For the I-131 thyroid dose rate f actor for exposure from inhalation: DFAi = 1.06E-2 area per pCi This value will yield a Pi f actor of 1.48E7 area /yr per uCi/m3 as listed on Table 3-5. It should be noted that only a limited number ; of nuclides are listed on Table 3-5. These are the most conson * ' nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then it will be calculated and included
- in a revision to the ODCM. .
Table 3-6, Food (Cow Milk) Pi (m2-eres/yr per uCi/sec) also takes into account the dose rate to various organs from the ingestion of cow milk. (From NUREG 013 3, i s ection 5.2.1.3) INSERT SYM80LS HERE [ Where K' = a constant of unit conversion,106 pCi/uC1. 4 = the cow's consumption rate, in kg/ day (wat weight), t U,p = the infant's milk consumption rate, in liters /.vr. Yp = the agricultural productivity by unit area, in kg/m2
- F = the stable element transfer coefficients, in days / liter. i r = fraction of' deposited activity retained on cow's feed grass.
DFLi= the maximum organ ingestion dose factor for the ith radionuclide, in area /pci. 1 1 = the decay constant for the ith radionuclide, in sec ~1 y= the decay constant for removal of activity on leaf and j plant surfaces by weathering, 5.73 x 10~7 see- -1 (corresponding to a 14 day half-time). tg = the transport time from pasture to cow, te silk, to infant . in sec. . May 198 6
1 i A f raction of the airborne deposition is captured by the ground plane I vegetation cover. He captured material is removed from the . vegetation (grass) by both radiological decay and weathering l processe s. The values of Or, Use, and Y, are provided in Regulatory Guide ' 1 1.109 Tables E-3, E-5, and 1-15, as 50 kg/ day, 330 liters / day and- i { 0.7 kg/m2 , respectively. The value tf is provided in Regulatory j j Guide 1.109 Table E-15, as 2 days (1.73 x 105 seconds). ne f raction, r, has a value of 1.0 for radiciodines and 0.2 for ! particulates, as presented in Regulatory Guide 1.109, Table E-15. '
, i Ta ble E-1 of Regulatory Guide 1.109 provides the stable element ;
transfer coefficients, F ., and Table E-14 provides the ingestion dose factors DFLi , for the infant's organs. Resolution of the units yields a '!
*, (fooel . 4.4 l W on, [e*1 % 1f) (et amen /yr per uti/sec) for all radionuclides, except tritium.
The concentration of tritium in silk is based on i ts airborne . concentration rather than the deposition rate. ; j = ' P, K'K'*FM,0rt, [0.7HO.5/N)] (aren/pr per pr../ y Where K'" = a constant of unit conversion,103 ga/kg. H = absolute hualdity of the atmosphere, in ga/m3 / 1 0.7 5 = the f raction of total feed that is water. 0.5 = the ration of the specific activity of the feed grass water to atmospheric water. i From Table E-1 and E-14 of Regulatory Guide 1.109, the- values of F and DFLi for tritium are 1.0 x 10-2 day / liter and 3.08 x = 10*I v ares 8 grams per/seter pCi,3,respectively. the resolution Assuming of unitsan average absolute humidity of yields: Pi (food) = 2.4 x 103 area /yr per uCi/m3 for tritius, only < Example Calculation: ' For I-131 thyroid does rate factor for exposure from cow milk. I ingestion: May 1986 h
1 J-r = 1.0 unitiese for Iodines l Fa = 6E-3 days / liter DFL 139E-2 aren/pci , 21 =i =9.982-7 se c-1 !
$w = 5.73E-7 sec-1 I
, tf = 1.73E+5 see i These values will yield a Pi f actor of 1.07E12 ares /yr per uCi/sec as ) listed on Table 3-6. It should be noted that only a limited number ! of nuclides are listed on Table 3-6. These are the most common ! nuclides encountered in affluents. If a nuclide is - detected for l which a factor is not listed, then it will be calculated and included l in a revision to the ODCM. 3.4.4 Dose Factor - f or -. I-131, I-133, Trittua and Particulates with I half-lives greater than 8 days. I TABLES 3.7 to 3.10, Ri VALUES _- INHALATION R1 (area /yr per uCi/m3) takes into accout.t several factors, among I these are the dose rate to- various organs from inhalation exposure. l (From NUKEG 0133, Section 5.3.1.1). I Ri = K'(BR), (DFA1 ), (area /yr per uCi/a3) Wheres l K' = a constant of unit conversion,106 pC1/ uci . i (BR). - thg/yr. a breathing rate of the receptor of age group (a),in (DFA1 )a = the organ inhalation dose factor for the receptor of age group (a) for the ith radionuclide, in ares /pci. The total body is considered as an organ in the l selection of (DFA1)aa The bresthing rates (BR), for the various age groups are tablula'ted below, as given in Table E-5 of the Regulatory Guide 1.109. Aae Groue (a) Breathina Rate (a 3/yr) i Infant 1400 child 3700 feen 8000 Adult 8000 Inhalation dose factors (DFA 1 )a for the various age groups are given in Tables E-7 throught E-10 of Regulatory Guide 1.10 9. Example Calculation For the I-131 infant thyroid dose f actor for exposure from inhalation: DFA1 = 1.06E-2 aram per pCi j May 1986 j 1 e,_ _____._.,__m__._._.w-.r--
--v--w----w--------e~-=,vw e---ef
These values will yield a Ri f actor of 1.48E7 area /yr per uci/m3 as 10 6ed on Table 3-7. It should be noted that only a limited number ot nuclides are listed on Table 3-7 thru 3-10. These are the most conson nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then it will be calculated ud included in a revision to the ODCM. TABLE 3-11, Ri VALUES - GROUND PLANE Ri (a2-area /yr per uCi/sec) also takes into account the dose f rom exposure to radiation depcsited on the ground. ' (From NULEG 0133, S ection 5.3.1.2) .
= K'K"(SF)DFG [(1-e lt i )/q j (,2.aren/yr per uCi/sec)
Rt i Where K' = a constant of unit conversion,106 pCi/uci. K" = a constant of unit conversion, 8760 hr/ year. Sia the decay constant for the ith radionuclide, sec-1 t = the esposure time, 4.73 x 108 see (15 years) .
=
DFGi the ground plane gose conversion f actor for the ith radionuclide (area /hr per pCi/a ) . SF = the shielding f actor (dimensionless). A shielding factor of 0.7 is suggested in Table E-15 ; of Regulatory Guide 1.109. A tabulation of DFGi walues is presented in Table E-6 of Regulatory Guide 1.109. Example Calculation: For the 1-131 total body dose f ar.cor for exposure to the ground: , 4 = 9.98E-7 sec-1 DFGi = 2.80E-9 area /hr per pCi/m2 These values wil). yield a Ri factor of 1.72E7 ' a2 srea/yr per uCi/sec a s I listed on Table 3-11. It should be noted that 'only a limited number of ' nuclides are listed on Table 3-11. These are the most common nuclides ' encountered in effluents. If a nuclide is detected for which a-factor is not listed, then it will be calculated and included in a revision of the ODCM. Ri (m2 wrea/yr per uCi/sec) also takes into account the dose to the skin f rom exposure to the ground. ,' i C n May 1986
'l u
1 Example Calculation , l For the I-131 skin dose f actor for exposure to the ground
]
21 a 9.98E-7 sec-1 I DFGi = 3.40E-9 area /hr per pCi/m2. These valu== will yield a Ri factor of 2.09E7 m2 erea/yr per uCi/se c as listed on Table 3-11. It should be noted that only a limited , number of nuclides are listed on Table 3-11. These are the most ' ! common nuclides encountered in effluents. If a - nuclide is detected for which a factor is not listed, then it will be calculated an : included in a revision to the CDQl. 1 TELES 3-12 to 3-15. Ri VALUES - COW MILK Ri (a2-ares /yr per uC1/sec) also takes into account the dose rate to , various organs from the ingestion of milk for all age groups. _ (From ! NUREG 0133, Section 5.31.3).
..] i
., ,si . w , .v,
. s. w 3 . o .<.g * .
I (e er w yr pse att/ses! Where2 K' = a constant of unit conversion,106 pCi/ uci . t , Qy - the cow's consumption rate, in kg/ day (wat weight). j l Uap = the receptor's milk consumption rate, in liters /yr. , i Yp= the agriculturgi productivity by unit area of pasture feed - grase , in k g/m' , l Ys= the agrpeultural productivity by unit area of stored feed, in kg/m . - t F, = the stable element transfer coefficients, in days / liter. , r = fraction of deposited activity retained on cow's feed grass. k (DFLi )a = the organ ingestion . dose factor for the ith radionuclide for the receptor in age group (a), in l are a/pci. 1 1 = the decay constant for the ith radionuclide, in see ~1 3 May 1986
-,----.---r -m .,,, , . . . - - . . . _ . _ _ ,- - . , , ...__.y, . _ - - _-.,e.. - - _ . _ _ w..,-- __-,-.w.-.--.-o--- -._. -
1 < l l l I i i i kv = the decay constant for removal of activity on leaf and plant surfaces by weathering, 5.73 a 10-7 sec *1 ' l (corresponding to a 14 day half-time). tg = the transport time from pasture to cow, to allk, to { receptor, in sec. tha the transport time from pasture, to harvest, to cow, to t allk, to receptor, in sec. , fp= fraction of the year that the cow is on pasture i (dimensionless). i
?, = fraction of the cow feed that . is ' pasture grass while the cow is on pasture (dimensionless).
SP$CIAL NOTEt The above equation is applicable in the case l that the milk animal is a goat.
- i
! Milk cattle are considered to . be fed from two potential sources. : l pasture grass and ctored feeds. Following the developtent in i j Regulatory Guide 1.109, the value of f, will be considered unity. TCN-1 + l f pwill be considered to be 0.5 for a May to October grating season. Tabulated below are the appropriate parameter values and their reference to Regulatory Guide 1.109._ In case that the slik animal is , 1 a goat, rather than a cow, refer to Regulatory Guide 1.109 for the appropriate parameter values. t Parameter yh M , r (dimensionless) 1.0 for radiolodine B 0.2 for particulates E-15 F,(days / liter) Each stable element E-1 U,p (liters /yr) - Infant 330 E-$
- Child 330
= E-5
- Toen 400 E-5 -
- Adult 310 E-5 '
(DFLg)s (area /pci) Each radionuclide E-M to E-14 i Y (kg/m2) 0.7 E-15 Y (kg/m2) f 2.0 k 15 ! tg (seconds) 1.73 x 105 (2 days) E-15 l th (seconds) 7.78 x 106 (90 days) E-15 I Qp (kg/ day) 50 E ; The concentration of trittua in re11k is based on the airborne concentration rather than the deposition. Therefore, the Rg is I based on (x/Q): Rg = K'E"FaQF U spDFLg(0.75(0.5/N)) (area /yr por WC1/m3 ) i May 1987 i
, . _ - , . - . . - - . _ . , _ . . _ _ _ . . . _ . . _ ~ - . . _ , _ ~ , . . . _ . - . , _ . _ . _ _ . . , - . . - . _ . , _ . - -
Where l J l K" = a constant of unit conversion,103 ga/kg . . 1 H = absolute humidity of the ataasphere, in ge/m3 . 0.75 = the f raction of total feed that is water. ! 0.5 = the ratio of the specific activity of the feed grass water to l atmospheric water. and other parameters and vglues are given above. The value of H is l' considered as 8 grams / meter 3
, in lieu of site specific information.
Example Calculation . For I-131 inf ant thyroid dose f actor f rom milk ingestiont j l r = 1.0 unitiess for Iodines .) Fa = 6 E-3 days / liter for ccvs and 6E-2 for goats *
)
DFLi = 1.39E-2 ares /pci
\1 = 9.9 8E-7 s ec -1 tw = 5.7 3E-7 see -1 ';
tf = 1.73E+5 sec. . i These values will yield a f actor of 5.26E11 and 6.31E11 area /yr per ! uCi/sec respectively for cow and goat milk. However, the actual dose i to the infant thyroid is also dependant on the highest relative l deposition at respective cow and goat locations. 'At the Nine Mile I Point Nuclear Station these deposition coefficients are 4.73E-10 and l 1.33E-10 a-2 respectively for cows and goats. Because the goa t deposition is relatively so much smaller than the slightly larger Ri factor, cow milk is the limiting silk. If the' location of the cow and goat milk receptors changes so that this is no longer true then ! the Ri factor will be revised accordingly. Table 3-12 lis t the ; infant thyroid dose factor from I-131 as 5 26E11 ares /yr per i uC1/sec. It should be noted that only a limited number of nuclides ; are listed on Table 3-12 thru 3-15. These are the most common nuclides encountered in affluents. If a nuclide is detected for 1 which a f actor is not listed, then it will be calculated and included in a revision to the ODCM. TABLES 3 3-18 Ri VALUES - COW MEAT Rg (m2-ares /yr per uCi/sec) also takes into account the dose rate 4 1 to various organs from the ingestion of cowmeat for all age groupe except infant. (From NUREG 0133 Section 5.3.1.4) w ... m ,,_ ,,s 9 .v y ~ .~, ! (=I*ren/r w vet /ses) May 1986 i n - - ~ .,em-.-, ,,---a --~,,vre- e - - ~ - , - - , - - + ~ ~ , * - - ' " "
i
?
I Whore t f Ff = the stable element transfer coefficients, in days /kg. ! Uap= the receptor's seat consumption rate for age (a), in kg/yr. f i tg = the transport time f rom pasture to receptor, in sec. l th = the transport time f rom crop field to receptor, in sec. ! Ta bulated below are the appropriate parameter values and their ! reference to Regulatory Guide 1.109. t Parameter Value Table (RGl.109) r (dimensionless) 1.0 for radiciodine E-15 0.2 for particulates E-15 ; Ft (days /kg Each stable element F.-1
- Uap (kg/yr - Infant 0 E-5
- Child 41 E-5
- Teen 65 E-5
- Adult 110 E-5 (DFL ): (area /pci) i Each radionuclide E-11 to E-14 Y p (kg/m ) 0.7 E-15 .
Y, (kg/m2 ) 2.0 E-15 tf (seconds) 1.73 x 10 6 6 E-15 (20 days) th (seconds) 7.78 x 10 (90 days) E-15 , QF (kg/ day) 50 E-3 ; 1 The concentration of tritium in seat is based on the airborne i concentration rather than the deposition. Therefore, the Ri is based on (z/Q] ,
't
- K'K'*FrQpU ,( % ), (0. N o.5/N)] (em a/yr per WC1/m3) l where all terms are defined above in this manual.
1 Example Calculationt For 1-131 child thyroid dose f actor from cow seat ingestion. Ff = 2.9E-3 days l r = 1.0 unitiess f or Iodines DFLi = 5.72E-3 are n/pci. These values will yield a Ri f actor of 2.75E9 m2mrea/yr per uCi/sec ) as listed on Table 3-16. It should be noted that only a limited number of nuclides are listed on Table 3-16 thru 3-18. These are the most .oanon nuclides encountered in efflu4nts. If a nuclide is l detected for which a f actor is not listed, then it will be calcu1Ated in a revision to the ODCM. l May 1986 l
, l l l ! TABLES 3-19 to 3-21. Ri VALUES - VEGETATION - l R1 (m2 erea/yr per . uci/sec)' also takes into account the dose to : i various organs from the ingestion of vegetation for all age groups ; except infrut. (From NUREG 0133. Section 5.3.1.5) . ! l The integrated concentratica in vegetation consumed by man follows ! the expression developed in the derivation of the milk factor. Man , , is considered to consume two types of vegetation (fresh and stored) ! that differ only in the time period between harvest and consumption, i therefore e Q < d i
"[ T,p g* a ,) ,
(M ttI "a'L' g 'Vf',,'y 8 g i (m2 eren/yr per pCf/sec) I w83ro1 K'
- a constant of unit conversten. 106pct / pct.
VI
- the consuotten rete of fresh leafy vegetation by the receptor in age .
group (a). In kg/yr. I ] U
- thekg/yr.
censumptten rate of stared vegetation by the recepter in age group (a).' in fg
- the fractten of the annual intate of fresh leafy vegetation groun81ecally.
f, a the fraction of the amuel intake of stored vegetetten grown locally. tg a the averste time between harvest of leafy vegetation and its censuustion.
. in secones.
t a h the average time toteen harvest of stored vegetatten and its consumetion. in secones. a Y, the vegetation erest density. In kg/ma , , and all other factors are defined in this manual. Tabulated below are. the appropriate parameter values and their reference to Regulatory Guide 1.109. l
!.timitf. IAINR, likit -
r(dimenstenless) 1.0 for redteiedines t1 l 0.2 for partievletes t1 (0FLg ), (ares /pCl) tech radienuclide t 11 to E 14 UI(kg/yr)-Infant 0 E5
- -- ChtId 26 t4 l- - Teen et tl
~
j
- Adult 64 El '
d(kg/yr)-Infant 0 E5
- Child $20 E 5-Teen 430 El Adult Sto El fg(dimenstoeless) sitespecific(default *1.0) ,
f, (dtnensionless)- site specific (defevlt *'0.76) (see RGuttspage 28) l tg(secones) 8.6I10'(1 day) t 15 t, (secones) s.te : 1e 8(66 days) t 15 t Y, (kg/e ) . 2.0 t 1s
. 7 h_ 5 Map 1986 _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _
s l l l l i The contentration of tritius in vegetation is based on the airborne concentration rather than the deposition. Therefore, the Ri is based on [z/Q): l Ri fdo) e t'K'- 3 ujft + U f, (ort,h (o.rs(e.s/m 3 (.,,,fy, ,,, ,g,f,3), , I i where all terms have been defined above and in this manual. : Example Calculation I For I-131 child thyroid dose f actor to the from vegetation ingestion: r = 1.0 unitiess for Iodines DFLi = 5.72E-3 area.pci. , l These values will ' yield a Ri factors - of 4.75E10 m2-ures/yr pe r i nci/sec as Itsted on Table 3-19. It should be noted that only a limited number of nuclides are listed on Table 3-19 thru 3-21. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not . listed, then it will be calculated and included in a revision to the ODCH. 3.4.5 X/Q and WV - Dispersion Parameters for Dose Rate. Table 3-22 The dispersion parameters for the whole body and skin dose rate l calculation correspond to the highest annual average dispersiou ! parameters at or beyond the unrestricted area boundary. This is a t 1 the East Site boundary. These values were obtained f rom the Nine i , Mile Point Unit 2 Final Environmental Statement, NUREG 1085 Table D-2 ! for the Vent and stack. These were calculated using the sethodology l of Regulatory Guide 1.111, Rev. 1. The Stack was mode, led as an elevated release point because its height is more than 2.5 times than any adjacent building. The Vent was modeled as a ground level l release because even though it is higher than any adjacent building f it is not more than 2.5 times the height. The NRC Final Environmental Statement values for the Site Boundary i X/Q and D/Q terms were selected for use in calculating Effluent l Monitor Alara Points and compliance with Site Boundary Dose Rate ' I specifications because they are conservative when compared with the corresponding NMPC Environmental Report values. In addition, the Stack "interdt tent release" X/Q was selected in lieu of the
" continuous" value, since it is slightly larger, and also would allow )
not making a distinction between long tern and short term releases. The dispersion parameters for the organ dose calculations were obtained from the Environmental Report, Figures 78-4 (Stack) and 78-8 (Vent) by locating vs. lues corresponding to currently existing (1985) pathways. It should be noted that the most conservative pathways do not all exist at the same location. It is cones rvative to - assume that a single individual would actually be at each of the receptor locations. May 1986 _ . _ _ _-_ - . _ J
-- - . . - . - - - . - - - - - - - . . . ~ . _ - - . - - . - - . - - - . .
1l I l I 3.4.6 Wy and Wa - Dispersion Parameters for Dose, Table 3-22 ; The dispersion parameters for dose calculations were obtainct chiefly 1 from the Nine Mile Point Unit 2. Favhonmental Report Appendiz 7B, as ' noted in Section 3.4.5. These were calculated using the methodology of Regulatory Guide 1.111 and NUREG 0324. The Stack was modeled as ' an elevated release point because its height is more than 2.5 times ! than any adjacent building The Vent was modeled as.a combined I elevated / ground level release because even though it is higher than i any adjacent building it is not more than 2.5 times the height. Average seterology over the appropriate time period was used. , Dispersion parameters not available from the ER were obtained fron ; C.T. Main Data report dated November, 1985, or as described in ' Section 3.4.5, the FES. 3.5 I-133 Estimation t The Stack and Vent Effluent Monitor at Nine Mile Point-Unit 2 are on ! line isotopic monitors. They are designed to automatically collact
- I t iodine samples on charcoal cartridges and isotopically analyze them, ,
i with a sensitivity which exceeds the LLD requirement on TS Table . l 4.11-2 of 1E-12 oci/c e. During those time periods in which the l I-133 analysis cannot meet the LLD requirement, the I-133
- concentration will be estimated as 4 times the I-131 concentration, or by ratio applied to the I-131 concentration. The ratio will be determined at least quarterly by analysis of short duration samples.
3.6 Use of Concurrent Meteorological Data vs. Historical Data It is the inter.t of NNPC to use dispersion parameters' based on historical meteorological data to set alara points and to determine or predict dose and dose rates in the environment due to gescou s e f fluents. When the methodology' becomes available, it is the intent to use meteorological conditions concurrent with the time of release to- determine gaseous pathway doses. Alara points and dose predictions or estimates will still be based on historical data. The ODCM will be revised at that time. 3.7 Gaseous Radwaste Treatment Systes Operation Te chnical Specification 3.11.2.4 requires the Gaseous Radweste Treatment System to be in operation whenever the main condenser- air ejector system is in operation. Since the system was designed without a bypass, station design results in compliance with the s pecification. The components of the system which must operate to treat offgas are the Preheater, Recombiner, Condenser, Drye r , Charcoal Adsorbers, HEPA Filter, and Vacuus Pump. See Figures 3-1, 3-2, and 3-3, Offgas System. May 1986
l 1! l 3.8 Ventilation Exhaust Treatment System Operation Technical Specification 3.11.2.5 requires the . Ventilation Exhsust Treatment System to be OPERABLE when projected doses in 31 days due to iodine and particulate releases would exceed 0.3 area to any organ i of a oesber of the public. The appropriate components, which affect i iodine or particulate release, to be OPERABLE aret
]
- 1) HEPA Filter - Radwaste Decon Area 1
- 2) HEPA Filter - Radweste Equipment Ares !
- 3) HEPA Filter - Radwaste General Area j Whenever one of these filters is not OPERABLE, iodine and particulate '
dose projections will be made for = the remainder of the current calendar month, and for each month (at the time. of ' calculating 1 cumulative monthly dose contributions) that the filter remains I inoperable, in accordance with 4.11.2.5.1. Predicted relenos rate wi.11 be used, with the methodology of Section 3.3.3. See Figure 3-5, Gaseous Radiation Monitoring.
- l I
1 May 1986 , 1 l
- - - . - - - - . - - - - - - . - - - - - - . -- - -l
l l TABLE 3-1 t ! 0FFGAS PRETREATMENT* DETECT 0R RESPONS E l NUCLIDE NET CPM / uCi/c c l Kr 85 4.30E+3 l Vr 85a 4.80E+ 3 Kr 87 8.00E+3 i Kr 88 7.60E+ 3 Xe 133 1.75E+3 ' Xe 133m .- l Xe 135 5.10E+3 Xe 135m -- , Xe 137 '8.10E+3 ; Xe 138 7.10E+ 3
- Values f rom SWEC purchase specification NMP2-P281F May 1986
t. l ! TABLE 3-2 ' PLUME SHINE PARANETERS
- f h t.IDE B4 (arad/y t aci/sec) V4 (area /yr 6 Sci /sec)
Kr 83a 3.5.1E-5 3.28E-5 s Kr 85 3.3 9E-3 3.21E-3 ! Kr 85m 1.04E-2 9.98E-3 Kr 87 2.34E-2 2.21E-2 Kr 88 2.01E-2 1.92E- 2 Kr 89 1.59E-2 1.51E-2 Xe 131a 6.90E-5 6.55E-5 Xe 133 6.12E-4 5.9 3E-4 Xe 133a 3.62E-4 3.44F.-4
- Xe 135 4.31E-3 4.09E-3 Xe 135m 6.55E-3 6.12E-3 i
Xe 137 3.0 7E-3 2.88E-3 Xe 138 1.38E-2 1.33E-2 Ar 41 1.69E-2 1.61L-2
*B g and Vg are calculated for critical site boundary location: 1.6km in the easterly direction.
May 1986 1
_ ~ _ _ _ __ _ _ r i TABLE 3-3 DOSE FACTORS
- Nuclide K, ( y-Body )*
- j,4 (6-Skin)** M4 _( y- Ai r ) * *
- N4 (B- Air)* ** ,
Kr 83a 7.56E-02 ---~ - 1.93 E1 2.88E2 Kr 85m ' 1.17E3 1.46E3 1.23E3 1.97E3 Kr 85 1.61El 1.34E3 1.72 E1 1.95E3 Kr 87 5.92E3 9.73E3 6.17E3 1.03E4 ; Kr 88 1.47E4 2.37E3 1.52E4 2.93E3 l Kr 89 1.66E4 .1.01E4 1.73E4 1.06E4 : Kr90 1.56E4 7.29E3 1.63E4 7.83E3 Xe 131a 9.15El 4.76E2 1.56E2 1.11E3 Xe 133m 2.51E2 9.94E2 3.27E2 1.48E3 f l Xe 133 2.94E2 3.06E2 3.53E2 1.05R3 Xe 135m 3.12E3 7.11E2 3.36E3 7.39E2 , Xe 135 1.81E3 1 86E3 1.92E3 2.46E3 Xe 137 1.42E3 1.22E4 1.51E3 1.27E4 Xe 138 8.83E3 4.13E3 9.21E3 4.75E3 l Ar 41 8.84E3 2.69E3 9.30E3 3.28E3
*Froe, Table B-1. Regulatory Guide 1.109 Rev.1
** ares /yr per uC1/n .
***arad/yr per nC1/n .
May 1986
- - - - - , ,- w. - -
y- _.,, , . . , _ ,.,,...._._,.4~ _ . . . , _ . - . . . + , . m,,.-
e-TABLE 3 Pi VALUE*, -. GROUND PLANEee a2
~
- ares /yr
-Ci/see NUCLIDE TOTAL BODY. SKIN H3 ---- ----
C 14' ---- ---- Cr 51 6.666b 7.85E6 : Mn 54 1.10E9 1.29E9-Fe 59 3.88E8 4.56E8 ' Co 58 5.27E8 6.18E8 .
-l La 60 4.40E9 5.17E9 Zn 65 6.87E8' 7.90E8 4
3r 39 3.06E4 3.56E4 i Sr 90 ---- ---- j Zr 95 3.44E8 ' 3.99E8 !
- Nb 95 3.50E8 4.12E8 Mo 99 5.71E6 6.61E6 .
I 131 2.46E7 2.98E7 j1 1 133 3.50E6 4.26E6 I Cs 134 2.81E9 3.26E9 Cs 137 1.15E9 ' 1.34 E9 2 Ba 140 2.93E7 3.35E7
- La 140 2.10E8 2.38E8 '
Ce 141- 1.95E7 2.20E7 Ce 144 5.85E7 6a77E7
- Daughter Decay Product. Activity level and effective half life assumed to ;
t - equal parent nuclide. -i
*
- Calculated in accordance . with NUREG 0133, Section 5.2.1.2.
May-1980 - l
f TABLE' 3- 5 1 PgVALUES'- INHALATION *
- mrea/yr uti/m NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-LL1;.
H3 -- 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 : C 14 2.65E4 5.31E3 5.31E3- 5.31E3 5.31E3; 5.31E3 5.31E3 Cr 51 -- -- 8.95 El- 5.75 El 1.32E1 1.28E4 3.57E 2 - ,
-1 Mn 54 --
2.53E4 4.98E3 -- 4.98E3 1.00E6 :7.0 6E3 - , Fe 59 1.36E4- 2.35E4 9.48E3 -- -- 1.02E6 2.48E4 i Co>58 -- 1.22E3- 1.82E3 -- -- 7.77E5- 1.11E4 Co 60 -- 8.02E3 1.18E4 -- -- 4.51E6 3.19 E 4' l Zn 65 1.93E4 6.26E4 3.11E4 -- 3.25E4 6.4 7E5 5.14E4 .; Sr 89 3.98E5 -- 1.14E4 -- -- 2.03E6 6.40E4 I Sr 90 4.09E7 -- 2.59E6 -. -- 1.12E7- 1.31E5 Zr 95 1.15E5 2.79E4 2.03E4 -- 3.11E4 - 1.75E6 2.17E4 l
- Nb 9 5 1.57E4 6.43E3 3.78E3 --
4.72E3 4.79E5 1.27E4. ' Mo 99 -- 1.65E2 3.23E1 -- 2.65E2 - 1.35E5' 4.87E4 1 131 3.7 9E4 - 4.44E4 1.96E4 1.48E7 5.18E4 -- 1.06E3 I 133 1.32E4 1.92E4 5.60E3 3.56E6 2.24E4 -- 2.16E'3 i Cs 134 3.96E5 7.03E5 7.45E4 -- 1.90E5 7.97E4- 1.33E3 Cs 137 5.49E5 6.12E5 4.55E4 -- 1.72E5 7.13E4 1.33E3 Ba 140 5.60E4 5.60E1 2.90E3 -- 1.34E1 1.60E6 3.84E4
*La 140 5.05E2 2.00E2 5.15 El -- --
1.68E5 8.48E4 Ce 141 2.77E4 1.67E4 1.99E3 -- 5.25E3 5.17E5 2.16E4 Ce 144 3.19E6 1.21E6 1.76E5 -- 5.38E5 9.84E6 1.48E5
- Daughter Decay Product. Activity airci and effective half life assumed to equal parent nuclide.
** Calculated in accordance with NUREG 0133, Section 5.2.1.1.
May 1986 i
. _ _ _ _ ~ _ _ _
l i 1 4 TABLE 3-6 Pg VALUES - FOOD (Cow Milk)**
- J m
2
- ares /yr 4 .uCi/se c ')
l l NUCLIDE BONE LIVER T. BODY THYRDID KIDNEY LUNG -GI-LLI
.i *H 3 -- 2.40E3 2.40E3 2.40';3 2.40E3 2.40E3 2.40E3
- C 14 3.23E6 6.89ES 6.89E5 6.89ES 6.89E5 6.89ES 6.89E5' Cr 51 -- -- 1.64E5 -1.07E5 2.34E4 - 2.08E5 4.78E6 Mn 54 -- 3.97E7 8.99E6 -- 8.80E6 --
1.46E7 Fe 59 2.28E8 3.99E8 1.57E8 -- -- 1.18E8 1.91E8 , Co 58 -- 2.47E7 6.16E7 -- -- -- 6.15E7. Co 60 -- 8.98E7 2.12E8 -- -- -- 2.14 E8 - Zn 65 5.65E9 1.94E10 8.94E9 -- 9.40E9 -- 1.64E10 Sr 89 1.28E10 -- 3.67E8' -- -- -- 2.63E8 Sr 90 1.24E11 -- 3.15E10 -- -- --
'1.55E9 Zr 95 6.93E3 1.69E3 1.20E3- -- -1.82E3 -- 8.41E5 **Nb 95 7.07E5 2.91ES- 1.68E5 --
2.09ES -- 2.46E8 Mo 99 -- 2.12E8 4.13E7 -- 3.17E8 -- 6.98E7 I 131 2.77E9 3.26E9 1.43E9 1.07E12 3.81E9 --
'1.16E4-I 133 3.69E7 5.37E7 1.57E7 9.77E9 6.31E7 -- 9.09E0 Ca 134 3.71E10 6.92E10 6.99E9 --
1 78E10 7.31F9 -1.88E8 Cs 137 5.24E10 6.13 E10 4.35E9 -- 1.65 E10 6.67E9 1.92E8 Ba 140 2.4 5E8 2.45E5 1.26E7 -- 5.83E4 1.51E5 6.03E7
** La 140 3.79E2 1.49E2 3.84 E1 -- -- --
1.75E6 l Ce 141 4.41E4 2.69E4 3.17E3 -- 8.30E3 -- 1.39E7-
- Ce 144 2.37E6 9.69ES 1.33E5 -- 3.92E5 -- 1.36E8
l l
*mrea/yr per uCi/m3. ** Daughter Decay Product. Activity level and effective half life assumed to I equal parent nuclide.
1
*** Calculated in accordance with NUREG 0133 Section 5.2.1.3.
May 1d86 l
M 1 i TABLE 3-7 < RgVALUES - INHALATION - INFANT *
- q are a/y r ;
3 uC1/m .l 1 1 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI_ j H3 -- 6.47E2 6,47E2 6.47E2 6.47E2 6.47E2 . 6.47E 2 . C 14 2.65E4 5.31E3 5.31E3 5.31E3 5.31E3 5,31E3 5.31E3 1 Cr 51 -- - - - 8.95 El 5.75El 1.32E1 1.28E4 1.57E 2 Mn 54 -- 2.5 3E4 ' 4.9"t3 -- 4.98E3 1.00E6 1.06E3 EFe 59 1.36E4. 2.35E4 9.48E3 -- -- 1.02E6 2.48E4 Co 58 -- 1.22E3 1.82E3 -- -- 7.77E5 1.11E4 Co 60 -- 8.02E3 1.~18E4 -- -- 4.51E6 3.19E 4 - Zn 65 1.93E4 6.26E4 3.11E4 -- 3.25E4 6.47E5 5.14E4 , Sr 89 3.98E5 -- 1.14E4 -- -' 2.03E6 6.40E4 Sr 90 4.09E7 -- 2.59E6 -- -- 1.12E7 1.31E5 Zr 95 1.15ES 2.79E4 2.03E4 -- 3.11E4 1.75E6 2.17E4
*db 95- 1.57E4 6.4 3E3 3.78E3 --
.4.72E3 4.79ES 1.27E4 Mo 99 --
1.65E2 3.23 E1 -- 2.65E2 1.35ES 4.87E4 I-131 3.79E4 4.4 4E4 1.96E4 1.4SE7 5.18E4 -- 1.06E3 I 133 1.32E4 1.92E4 5.60E3 3.56E6 2.24E4 -- 2.16E 3 Cs 134 3.96ES 7.03E5 7.45E4 -- 1.90E5 7.97E4 1.33E3 Cs 137 5.49E5 6.12E5 4.55E4 -- 1.72E5 7.13E4 1.33E3 3a 140 5.60E4 5.60E1 2.90E3 -- 1.34E1 1.60E6 3.84E4
- La 140 5.05E2 2.00E2 5.15 El -- --
1.68E5 8.48E4 l Ce 141 2.77E4 1.67E4 1.99E3 -- 5.25E3 5.17ES 2.16E4 Ce 144 3.19E6 1.21E6 1 76E5 -- 5.38E5 9.84E6 1.48E5
- Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
**This and f ollowing R Tables g Calculated in accordance with NUREG ' 0133, S ection 5.3.1, excpet C 14 values in accordance with Regulatory Guide .l.109 Equation C 8.
l 1 May 1986
HTABLE 3-8 ; R VALUES 1
-INHAIATION - CHILD ares /y r uCi/m NUCLIDE BONE LIVER T. BODY THTROID KIDNEY LUNG; GI-LLI- ,
H3 -- 1.12E3 1.12E3 1.12E3 1.12E3 :1.12E3 1.12E3' C 14 3.59E4 6. 7 3E3. 6.73E3 6.73E3 6.73E3 '6.73E3- 6.73E3 Cr 51 -- -- 1.54E2 8.55 El 2.43E1 11.70E4 1.08E3-Mn 54 -- 4.29E4 9.51E3 -- 1.00E4 1.58E6' 2.29E4i . Fe 59 2.07E4 3.34E4 1.67E4 -- -- 1.27E6 7.07E4 Co 58 -- 1.77E3 3.16E3 -- --- 1.11E6
- 3.44E4 Co 60 --
1.31E4 2.26E4 -- -- 7.07E6 - 9.62E4 Zn 65 4.26E4 1.13E5 7.03E4 -- 7.14E4 '9.95ES- 1.63E4. Sr 89 5.99E5 -- 1.72E4 -- -- 2.16E6 11.67ES-Sr 90 1.01E8 -- 6.44E6 -- -- 1.48E7 3.43E5 1 Zr 95 1.90E5 4.18E4 3.70E4 -- 5.96E4 2.23E6 6.11E4 - l
- Nb 95 2.35E4 9.18E3 6.55E3 ' --
8.62E3i 6.14E5' 3.70E4 Mo 99 -- 1.72E2 4.26 El -- 3.92E2 1.35ES- 1.27ES I 131 4.81E4 4.81E4 2.73E4 1.62E7 7.88E4 -- 2.84E3 1 133 1.6624 2.03E4 7.70E3 3.85E6 3.38E4 -- 5.48E3 Cs 134 6.51E5 1.01E6 2.25ES -- 3.30E5- 1.21ES 3.85E3 Cs 137 9.07ES 8.25E5 1.28E5 -- 2.82E5 1.04E5 3.62E3 1 Ba 140 7.4 0E4 6.48E1 4.33E3 -- 2.11El 1.7.4E6 1.02E5
*La 140 6.44E2 2.25E2 7.55El -- --
1.83E5 2.26ES Ce 141 3.92E4 1.95E4 2.90E3 -- 8.55E3 5.44E5 5.66E4 Ce 144 6.77E6 2.12E6 3.61E5 -- 1.17E6' 1.20E7 3.89ES
- Daughter Decay Product. Activity level and effective half life assuced to equal parent nuclide.
May 1986
TABLE 3-9 RgVALUES - INHALATION -' TEEN - are a/y r 3 uci/m i NUCLIDE BONE LIVER T. BODY THYROID- KIDNEY LUNG GI-LLI= j H3 -- 1.27E3 1.27E3 1.27E3 '1.27E3 1.27E3 1.27E3 i C 14 2.60E4 4.87E3 4.87E3 4.87E3' 4.87E3 4.87E3 4.87E3 Ce 51 -- -- 1.35E2 7.50E1 3.07El ~ 2.10E4 3.00E3 1 Mn J4 -- 5.11E4 8.40E3 -- 1.27E4. 1.98E6 6.68E4 j Fe 59 1.59E4 3.70E4 1.43E4 --- -- 1.53E6 1.78E5 C Co 58 -- 2.07E3 2.78E3 -- -- 1.34E6 9.52E4 Co 60 -- 1.51E4 1.98E4 -- ' -- 8.72E6 - 2'.59 E5 l i t Zn 65 3.86E4 1.34E5 6.24E4 -- 8.64E4 1.24E6 4.66E4 i 1 Sr 89 4.34E5 -- 1.25E4 -- -- 2.42E6 3.71ES Sr 90 1.08E8 -- 6.68E6 -- -- 1.65E7 7.65E5-
~
Zr 95 1.46E5 4.58E4 3.15E4 -- 6.74E4 2.69E6 1.49E5 l
- Nb 9 5 '1.86E4 1.03E4 5.66E3 --
1.00E4 7.51ES 9. 6 8E4 -' : Mo 99 -- 1.69E2 3.22E1 -- 4.11E2 - 1.54E5' 2.69E5 I 131 3.54E4 4.91E4 2.6 4E4 1.46E7 8.40E4 -- 6.4 9E3 I133 1.22E4 2.95E4 6 22E3 2.92E6 3.59E4 -- 1.03E4 ; Cs 134 5.02E5 1.13E6 5.49ES -- 3.75ES 1.46ES 9.76E3- l Cs 137 6.70E5 8.48:5 3.11ES -- 3.04E5 1.21E5 8.48E3 : j Ba 140 5.47E4 6.70E1 3.52E3 -- 2.28E1 2.03E6 2.29ES ;
- La 140 4.79E2 2.36E2 6.26 El -- --
2.14E5 4.87ES Ce 141 2.84E4 1.90E4 2.17E3 -- 8.88E3 6.14E5 1.26E5 Ce 144 4.89E6 2.02E6 2.62E5 -- 1.21E6 1.34E7 8.64E5
- Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. l May 1986 (N .
TABLE 3-10 ; RgVALUES - INHALATION.- ADULT ares /y r 0 uC1/a i NUCLIDI E ik'E LIVER T. BODY THYROID KIDNEY LUNG GI-LLI__ H3 -- 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 I C 14 1.82E4 3.41E3 3.41E3 3.41E3 3.41E3- 3.41E3 3.41E3 Cr 51 -- -- 1.00E2- 5.95El' 2.28E1 1.44E4 3.32E3 Mn 54 -- 3.96E4 6.30E3 -- 9.84E3 1.40E6 7.74E4. Fe 59 1.18E4 2.78E4 1.06E4 -- -- 1.02E6 1.80E5 l l L Co 58 -- 1.58E3 2.07E3 -- -- 9.28ES: 1.06E5 , 1 Co 60 -- 1.15E4 1.48E4 -- -- 5.97E6 - 2.85ES Zn 65 3.2 4E4 1.03E5 4.66E4 -- 6.90E4 8.64E5 5.34E4. Sr 89 3.04E5 -- 8.72E3 - - - -- 1.40E6- 3.50E5 l Sr 90 9.9 2E7 -- 6.10E6 -- -- 9.60E6 7.22E5 Zr 95 1.07ES 3.44E4 2.33E4 -- 5.42E4 1.77E6 1.50E5 i L
- Nb 9 5 1.41E4 7.82E3 4.21E3 --
- 7. 7 4E3 - 5.05E5' 1.04E5 Ho 99 --
1.21E2 2.30 E1 -- 2.91E2 9.12E4 2.48E5 I 131 2.52E4 3.53E4 2.05E4 1.19E7 6.13E4 --
-6.28E3 I133 8.64E3 1.48E4 4.52E3 2.15E6 2.58E4 - --
8.88E3 Cs 134 3.7 3E5 8.48E5 7.28E5 -- 2.87E5 9.76E4- '1.04E4 Cs 137 4.78E5 6.21ES 4.28E5 -- 2.22E5 7.52E4 8.40E3 i Ba 140 3.90E4 4.90E1 2.57E3 -- 1.67El 1.27E6 2.18E5 l
- Ia 14 0 3.44E2 1.74E2 4.58 E1 -- --
'1.36ES 4.58E5 Ce 141 1.99E4 1.35E4- 1.53E3 --
6.26E3 3.62E5 1.20E5 Ce 144 3.43E6 1.43E6 1.84E5 -- 8.48E5 7.78E6 8.16E5
- Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
-70 Mat 1986 . _ _ . _ _ _ _ _ __
7 TABLE 3-11 RgVALUES - GROUND PLANE - ALL AGE GROUPS 2 m - area /yr i uCi/se c i NUCLIDE TOTAL BODY -SKIN '[ H~3 L C 14 - Cr 51 - 4.65E6 5.50E6 ; Mn 54 1.40E9 1.64E9 . Fe 59 2.73E8 3.20 E8 -' i i Co 58 3.80E8- 4.45E8 Co 60 2.15E10 2.53 E10 -; Zn 65 7.4 6E8 8.57E8 Sr 89 2.16E4 2.51E4 ~ Sr 90 -- -- Zr 95 2.45E8 2.85E8 1
- Nb 95 2.50E8 : 2.94E8 Mo 99 3.99E6 4.63E6 d
l I 131 1.72E7 2.09E7
?
I 133 2.45E6 2.98E6 Cs 134 6.83E9 7.97E9 Ca 137 1.03E10 1.20 E10 Ba 140 2.05E7 2.35E7 .
*I4 140 1.47E8 1.66E8 L Ce 141 1.37E7 1.54E7 Ce 144 6.96E7 8.07 E7
- Daughter Decay Product. Activity level and effective half life assumed )
to equal parent nuclide. l May 1986 -
i q TABLE 3-12 Rg' VALUES - COW MILK - INFAN T a -erea/yr i uC1/se c-NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG' GI-LLI
*H 3 --
2.38E3 2.38E3 2.38E3 2.38E3- 2.38E3 2.38E3
*C 14 3.23E6 6.89E5 6.89E5- 6.89E5 6.89E5 6.89E5 6.89E5 Cr 51 -- --
8.35E4 5.45E4 1.19E4 1.06E5 2.43E6~ Mn 54 -- 2.51E7 5.68E6 -- 5.56E6 >-- 9.21E6 Fe 59 1.22E8 2.13E8 8.38E7 -- -- 6.29E7 = 1 '.02E8 '
. 1 Co 58 --
1.39E7 3.46E7 -- -- -- 3.46E7 - ! i' Co 60 -- 5.90E7 1.39E8 -- -- -- 1.40E8
)
Zn 65 3.53E9 1.21E10 5.58E9 -- 5.87E9 -- 1.02E10 i
^
Sr 89 6.93E9 -- 1.99E8 -- -- -- 1.42E8 Sr 90 8.19E10 -- 2.09E10 -- -- --- 1.02E9' Er 95 3.85E3 9.39E2 6.66E2 -- 1.01E3 --- 4.68E5 li
*
- Nb 9 5 3.93E5 1.62E5 9.35E4 --
1.16E5 -- 1.37E8 i Mo 99 -- 1.04E8 2.03E7 -- 1.55E8 -- 3.43E7 i I 131 1.36E9 1.60E9 7.04E8 5.26E11 1.87E9 -- 5.72E7 I 133 1.81E7 2.64E7 7.72E6 4.79E9 ' 3.10E7 -- 4.46E6 Cs 134 2.41E10 4.49E10 4.54E9 -- 1.16E10 4.74ES 1.22E8-Cs 137 3.47E10 4.06 E10 2.88E9 -- 1.09E10 4.41E9 1.27E8 ; Ba 1401.21E8 1.21ES 6.22E6 -- 2.87E4 7.42E4 .2.97E7
** La 1401.86E2 7.35El 1.89 El -- -- --
8.63E5 Ce 1412.28E4 1.39E4 1.64E3 -- 4.28E3 -- 7.18E6 J Ce 144-1.49E6 6.10E5 8.34E4 -- 2.46ES -- 8.54E7
- mrem /yr per uC1/m3 .
** Daughter Decay Product. Activity level and effective ualf life assumed to equal parent nuclide.
May 1986
?
i TABLE ' 3 -13
.k g VALUES - COW MILK - CHILD l.
m -erea/yr t .uci/se c NUCLIDE BONE LIVER ~T. BODY THYROID KIDNEY LUNG ~ GI-LLI
*H 3 --
1.57E3 1.57E3 1.57E3 1.57E3 1.57E3~ 1.57E3
*C 14 1.65E6 3.29ES 3.29ES 3.29E5 J.29E5 3.29E5: 3.29E5J Cr 51 -- -- 5.27E4 2.93E4 .7.99E3- 5.34E4 ' 2.80E6 '
Mn 54 -- 1.35E7 3.59E6 -- 3.78E6 ---
- 1.13E7 Fe 59 6.52E7 1.06E8 -5.26E7 -- -- 3.06E7 ' '1.10E8-Co 58 -- 6.94E6 2.13E7 -- -- --
4.0$E7 ' , Co 60 -- 2.89E7 8.52E7 -- --- -- '1.60E8
'1.23E9
~
Zn 65 2. 63E9 7.00E9 4.35E9 -- 4.41E9 ~- Sr 89 3.64E9 -- 1.04E8 -- -- - - - 1.41E8 , Sr 90 7.53E10 -- 1.91E10 -- -- -- 1.01E9 3 Zr 95 2.17E3 4.77E2 4.25E2 --
-6.83E2 -- 4.98E5 *
- Nb 9 5 2.10E5 8.19E4 5.85E4 --
7.70E4 -- 1.52E8' Mo 99 -- 4.07E7 1.01E7 -- 8.69E7 -- 3.'37 E7 I 131 6.51E8 6.55E8 3.'Er8 2.17E11' 1.08E9 -- 5.83E7 I133 8.58E6 1.06E7 4.01E6 1.97E9 1.77E7 -- 4.27E6 , Cs 1341.50E10 2.4 5E10 5.18E9 -- 7.61E9' 2.73E9~ 1.32E8 l l Cs 137 2.17E10 2.08 E10 3.07E9 -- 6.78E9 2.44E9 1.30E8 l Ba 140 5.87E7 5.14E4 3.43E6 -- 1.67E4 3.07E4 '2.97E7 - l
*
- Ia 140 8.92 E1 3.12 E1 1.05 El -- -- -- 8.69ES -
Ce 141 1.15E4 5. 73E3 8.51E7 -- 2.51E3 -- 7.15E6 Ce 1441.04E6 3.26E5 5.55E4 -- 1.80ES. --
' 8.49E7
- mrem /yr per uCi/m3 ,
** Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
l L i-l
- TABLE 3-14 -
Rg-VALUES _ COW MILK - TEEN ; a grea/yr + 2 uC1/sec NUCLIDE BONE LIVER T. BODY . THYROID -KIDNEY' LUNG GI-LLI
*H 3 --
9.94E2. 9.94E2 9.94E2. 9.94E2 9.94E2 9.94E2.
- C 14 6.70E5 1.34E5 1.34E5- 1.34E5 1.34E5 l.35E5~ 1.34E5 Cr $1 -- --
.2.58E4 1.44E4 5.o6E3- 3.69E4 4.34E6 Mn 54 --
9.01E6 1.79E6 -- 2.69E6 -- 1.85E7 Fe 59 2.81E7 . 6.57E7 2.54E7 -- -- 2.07E7 1.55E8 Co 58 -- 4.55E6 1.05E7 -- -- -- 6.2 7E7 '- Co 60 -- 1.86E7. 4.19E7 -- -- --< 2.42E8 Zn 65 1.34E9 4.65E9 2.17E9 --
- 2. 9 7E9 .- --
1.97E9- ' Sr 89 1.47E9 -- 4.21E7 -- --- -- 1.75 E8 Sr 90 4.4 5E10 -- 1.10E10 -- -- -- 1.25E9 . 4 Zr 95 9.34E2- 2.95E2 2.03E2 -- 4.33E2 -- 6.80E5
**Nb 95 9.3 2E4 5.17E4 2.85E4 --
5.01E4' --
- 2. 21E8.'
Mo 99 -- 2.24E7 4.27E6 -- 5.12E7 -- 4.01E7 I 131 2.68E8 3.76E8 2.02E8 1.10E11 6.4 7E8 -- 7.44E7 I 133 3.53E6 5.99E6 1.83E6 8.36E8 1.05E7 -- 4.53E6 - 1 Cs 134 6.4 9E9 1.53E10 7.08E9 -- 4.85E9 1.85E9 1.90E8. Cs 137 9.02E9 1.20E10 4.18E9 -- 4.08E9 1.59E9 1.71E8 Ba 140 2.4 3E7 2.98E4 1.57E6 -- 1.01E4 2.00E4 3.75E7.
**I4140 3.73E1 1.83E1 4.87E0 -- -- --
1.05E6 Ce 141 4.67E3 3.12E3 3.58E2 -- 1.47E3 -- 8.91E6 Ce 144 4.22E5 1.74E5 2.27E4 -- 1.04E5 -- 1.06E8 i
*mres/yr per oC1/m3 , j
** Daughter Decay Product. Activity level and effective half life assumed to ;
equal parent nuclide. j May 1986 1
4 TABLE 3 1 RgVALUES - COW MILK - ADULT a -erea/yr t uCi/se c . . NUCLIDE BONE LIVER T. BODY . THYROID: KIDNEY LUNG- GI-LLI
*H 3 -' 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2
- C 14 3.63E5 7.26E4 7.26E4 ' 7. 2 6E4 - 7.26E4 7.26E4 7.26E4 Cr 51 -- --
1.48E4 8.85E3 ' 3.26E3 '1.96E4 3.72E6 1.03E6 1.61E6 1.66E7 Mn 54 --: 5.41E6 -- -- Fe 59 1.61E7 3.79E7 1.45E7- -- -- 1.06E7 1.26E8 Co 58 -- 2.70E6 6.05E6 -- -- -- 5.47E7 -. Co 60 -- 1.10E7 2.42E7 -- -- -- 2.06E8 Zn 65 8.71E8 2.77E9 1.25E9 -- 1.85E9 -- 1.75E9 Sr 89 7.99E8 -- 2.29E7 -- -- -- 1.28E8 Sr 90 3.15E10 -- 7.74E9 -- -- -- 9.11E8 Zr 95 5.34E2 1.71E2 1.16E2 -- 2.69E2 -- 5.43E5
* *Nb 9 5 5.46E4 3.04E4 1.63E4- --
3.00E4 --- 1.84E8 Mo 99 -- 1.24E7 2.36E6 -- 2.81E7 -- 2.87E7 I 131 1.48E8 2.12E8 1.21E8 6.94E10 3.63E8 -- 5.58E7-1 1 1 133 1.93E6 3.36E6 1.02E6 4.94E8 5.86E6 -- 3.02E6 Cs 134 3.74E9 8.89E9 7.27E9 -- 2.88E9 9.55E8 l'.56E8 Cs 137 4.97E9 6.80E9 4.46E9 -- 2.31E9 7.68E8 1.32E8 Ba 140 1.35E7 1.69E4 8.83E5 -- 5.75E3 9.69E3 2.77E7-
**Ia 140 2.07El 1.05El 2.76E0 -- -- --
7.67E5 Ce 141 2.54E3 1.72E3 1.95E2 -- 7.99E2 -- 6.58E6 Ce 144 2.29ES 9.58E4 1.23E4 -- 5.68E4 -- 7.74E7
- mrem /yr per uC1/m3
** Laughter Decay Product. Activity level and effective half life assumed to equal p_ rent nuclide.
May 1986 1
TABLE 3-16 Rg- VALUES - COW MEAT - CHILD E SteB/yr I uCi/Se C : NFCLIDE BONE LIVER T. BODY ~ THYROID: KIDNEY LUNG GI-LLI ,
- H 3. --
2.34E2 ' 2.34E2 '2.34E2- 2.34E2 2.34E2 2.34E 2 l
- C 14 5.29E5 1.06E5 1.06ES '1.06E5 1.06E5 1.06E5l 1.06E5' C r 5. -- --
4.55E3 2.52E3 6.90E2 4.61E3 2.41E5 Mn 54 -- 5.15E6 1.37E6 -- 1.44E6 -- 4.32E6 i I Fe 59 2.04E8 3.30E8 1.65E8 -- -- 9.58E7 3.44 E8 - l Co 58
~
9.41E6 2.88E7- -- -- -- 5.49E7 . Co 60 -- 4.64E7 1.37E8 -- -- -- 2.57E8 Zn 65 2.38E8 6.35E8 . 3.9 5E8 --- 4.00E8' -- 1.12E8 i Sr 89 2.65E8 -- 7.57E6 -- -- -- 1.03E7 j 9 Sr 90 7.01E9 -- 1.78E9 -- -- -- 9.44E7 ' Zr 95 1.51E6 3.32E5 2.95E5 -- 4.75ES -- 3.46E8 j
* *Nb 9 5 2.41E6 9.38E5 6.71E5 --
8.82E5 -- 1.74E9 Mo 99 -- 5.42E4 1.34E4 -- l'.16ES . -- 4.48E4 . I 131 8.27E6 8.3 2E6 4.73E6 2.75E9 1.37E7 -- 7.40E5 - I 133 2.87E-1 3.55E-1 1.34E-1 6.60E-1 5.92E-1 -- 1.43E -l ' l Ca 134 6.09E8 1.00E9 2.11E8 -- 3.10E8 l'.11E8 5.39E6 Cs 137 8.99E8 8.60E8 1.27E8 -- 2.80E8 1.01E8 5.39E6 Ba 140 2.20E7 1.93E4 1.28E6 -- 6.27E3' 1.15E4 1.11E7 !
**1a 1401 67E2 5.84E1 1.97El- -- -- --
1.63E6 Ce 141 1.17E4 5.82E3 8.64E2 -- 2.55E3 -- 7.26E6 i Ce 144 1.48E6 4.65E5 7.91E4 -- 2.57E5 -- 1.21E8 l l i
- mrem /yr per uC1/m3 .
** Daughter Decay Product. Activity level and effective half life assumed to ,
equal parent nuclide. May 1986 i
l t, v
' TABLE 3 RgVALUES . COW MEAT - TEEN m2 -ares /y 4 'uCi/se c NUCLIDE BONE LIVER T.. BODY- THYROID. KIDNEY LUNG GI-LLI
=l
*H 3 -- 1.94E2- 1.94E2 1.94E2 _1.94E2 '1.94E2 II.94E2
*C 14 2.81E5 5.62E4 5.62E4; 5.6 2E4 - 5.62E4 5.62E4" -5.62E4 i
Cr 51 -- -- 2.93E3 1.62E3 '6.39E2 4.16E3 : 4.90E5 I 9.24E6 Mn 54 -- 4.50E6 8.93E5 --- 1.34E6 --
]
Fe 59 1.15E8 2.69E8 1.04E8 -- --
' 8.47E7 6.36 E8 ' j Co 58 --
8.05E6 1.86E7 -- -- -- 1.11E8 Co 60 -- 3.90E7 8.80E7 -- -- -- 5.09E8' 1 Zn 65 1.59E8 5.52E8 2.57E8 -- 3.53E8~ -- 2.34E8 Sr 89 1.40E8 -- 4.01E6 -- -- -- 1.67E7 l Sr 90 5.4 2E9 -- 1.34E9 -- -- -- 1.52E8 { ! Zr 95 8.50E5 2.68E5 1.84E5 -- 3.94E5 -- 6.19E8
**Nb 95 1.40E6 7.74E5 4.26ES --
7.51E5 -- 3.31E9' Mo 99 -- 3.90E4 7.43E3- -- 8.92E4 -- '8E4 I 131 4.4 6E6 6.24E6 3.35E6 1.82E9 1.07E7' -- 1.23E6 1 1133 1.55E-1 2.62E-1 8.00E-2 3.66 El -4.60E -1 -- 1.99E i Ca 134 3.46E8 8.13E8 3.77E8 -- 2.58E8 9.87E7 1.01E7 i Cs 137 4.88E8 6.49E8 2.26E8 -- 2.21E8 8.58E7 9.24E6 Ba 140 1.19E7 1.46E4 7.68E5 -- 4.95E3 -9.81E3 1.84E7
*
- La 140 9.12 E1 4.48 E1 1.19 El -- -- --
2.57E6 Ce 141 6.19E3 4.14E3 4.75E2 -- 1.95E3 .-- 1.18E7-Ce 144 7.87E5 3.26ES 4.23E4 -- 1.94E5 -- 1.98E8
- mrem /yr per uC1/m3 .
** Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
_. _ . .. _ ._~ _ _ . . - . TABLE 3-18 . RgVALUES - COW MEAT - ADULT- , a grea/yr i L ci/sec u NUCLIDE BONE LIVER T. BODY THYROID' KIDNEY LUNG- GI-LLI i
*H 3 -- 3.25E2' ' 3.25E2 3.25E2 3.25E'2- 3.25E2 3.25E 2 - .
- C 14 3.33E5 6.66E4 6.66E4 6.66E4 6.66E4 6.66E4- 6.66E4 Cr 51 -- -- 3.65E3 2.18E3 8.03E2 4.84E3 - 9.17E5 Mn 54 --
5.90E6 1.13E6 -- 1.76E6 --- 1.81E7 Fe 59 1.44E8 3.39E8 1.30E8 -- -- 9.46E7' 1.13E9' Co 58 -- 1.04E7 L34E7 -- -- -- 2.12E8L ..
.l Co 60 --
5.03E7 bd Dd . -- -- -- 9.45E8 , Zn 65 4.81E8-
~
2.2 6E8 - 7.19E6 3.25E8 -- -- 4.53E8 Sr 89 1.66E8 --- 4.76E6. -- -- -- 2.66E7 i Sr 90 8.38E9 -- 2.06E9- -- -- -- 2.4 2E8 - ! Zr 95 1.06E6 3.40E5 2.30E5 -- 5.34E5 -- 1.08E9 !
* *Nb 9 5 1.79E6 9.94E5 5.35E5 --
9.83E3 -- 6.04E9-Mo 99 -- 4.71E4 8.97'43 -- 1.07ES '-- 1.09E5 ! I 131 5.3 7E6 7.67E6 4.40E6 2.52E9 1.32E7 --
- 2. 0 2E6 . !
I 133 1.85E-1 3.22E-1 9.81E-2 4.73E1 5.61E-1 -- 2.89 E -l' - l Cs 134 4.3 5E8 1.03E9 8.45E8 -- 3.35E8 '1.11E8 1.81E7 ; Cs 137 5.88E8 8.04E8 5.26E8 -- 2.73E8 - 9.07E7 1.56E7 - Ba 140 1.44E7 1.81E4 9.44E5 -- 6.15E3 1.04E4 2.97E7
*
- I4 140 1.11E2 5.59El 1.48E1 -- -- ---
4.10E6 Ce 141 7.38E3 4.99E3 5.66E2 -- 2.32E3 -- 1.91E7' l Ce 144 9.33E5 3.90E5 5.01E4 -- 2.31E5 -- 3.16E8 1' 1 I
*mres/yr per nC1/m3 , i I
** Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. l l May 1986 -
3 b
-i TABLE 3-19 R'y VALUES - VEGETATION - CHILD. ,
-i a2-eres/yr + uC1/se'c ;
NUCLIDE BONE LIVER -T. BODY: THYROID. KIDNEY. LUNG. GI-LLI
*H 3 -- 4.01E3 4.01E3 4.01E3 - 4.01E3 4.01E3 4.01E3
- C 14 3.50E6 7.01E5 7.01E5 7.01E5 7.01E5 7.01E5 7.01E5' Cr 51 -- --
1.17E5 6.49E4 1.77E4 1.18E5 6.20E6 Mn 54 -- 6.6 5E8 , 1.77E8 -- 1.86E8 -- ~5.58E8 ; Fe 59 3.97E8 6.42EB 3.20E8- -- -- l'.86E8 6.'6 9 E8 . t Co 58 -- 6.4 5E7 .- 1.97E8 -- -- --- 3.76E8 , g Co 60 -- 3.78E8 : :1.12E9 -- -- -- 2.10 E9 Zn 65 8.12E8 2.16E9 1.35E9 -- 1.36E9 -- 3.80E8 -i Sr 89 3.59E10 -- 1.03E9 -- -- -- 1.39E9 l 1.67E10 l Sr 90 1.24E12 -- 3.15E11 -- -- -- 1 l Zr 95 3.86E6 8.50E5 7.56ES' -- 1.22E6 '-- 8.86E8 '
* *Nb 9 5 7.50E5 2.92E5 2.09E5- --
.2.74E5 --
5.40E8' ; 6.37E6
~
l Mo 99 -- 7.70E6 1.91E6 -- 1.65E7 -- I 131 1.43E8 1.44E8 8.16E7 4.75E10 2.36E8 -- 1.28E7 _; I 133 3.52E6 4.35E6 1.65E6 8.08E8 7.25E6 --. 1.75E6- , Ca 134 1.60E10 2.63E10 5.55E9 -- 8.15E9 2.93E9 1.42E8. Ca 137 2.39 E10 2.29 E10 3.38E9 -- 7.46E9 2.68E9 1.43E8 Ba 140 2.77E8 2.43E5 1.62E7 -- 7.90E4 1.45E5 '1.40E8 l
**14 140 3.37E4 1.18E4 3.97E3 -- -- --
3.28 E8 ' Ce 141 6. 5 6ES - 3.2 7E5 4.85E4' -- 1.43E5 -- 4.08E8 Ce 144 1.27E8 3.98E7 6.78E6 -- 2.21E7 -- 1.04 E10 1 I
*mres/yr per uC1/m3 .
** Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. .l May 1986 l
+
TABLE 3-20 1 Rj VALUES - VEGETATION - TEEN. a -rvem/yr i uCi/se c. ; l NUCLIDE BONE LIVER' T. BODY THYROID ' KIDNEY- LUNG GI-LLI-
*H 3 --
2.59E3 2.59E3 2.59E3 2.59E3 - 2.59E3 2.59E3
*C 14 1.45E6: 2.91E5 2.91E5 2.91E5 2.91E5 2.91E5 2.91ES .
Cr 51 -- -- 6.16E4 3.42E4 1.35E4 8.79E4 1.03E7 Mn 54 - - - 4.54E8 9.01E7 -- 1.36E8 -- 9.3 2E8. ; 1 i Fe 59 1.79E8 4.18E8 - 1.61E8 -- -- 1.32E8: 9.89E8 Co 58 --- 4.37E7 1.01E8 -- -- -- 6.02E8 . Co 60 -- 2.49E8 5.60E8 -- -- --
' 3.24 E9 Zn 65 4.24E8 1.47E9 6.86E8 --
9.41E8 -- 6.23E8 Sr 89 1.51E10 -- 4.33E8 -- -- -' '1.80E9 Sr 90 7.51E11 -- 1.85E11 -- -- -- 2.11E10 l Zr 95 1.72E6 5.44E5 3.74E5 --- 7.99E5 -- 1.26E9 -j
* *Nb 9 5 3.44E5 1.91ES 1.05ES --
1.85ES -- 8.16E8 Mo 99 -- 5.64E6 1.08E6 -- 1.29E7 -- 1.01E7 -i I 131 7.68E7 1.07E8 5.78E7 3.14E10 1.85E8 -- 2.13E7- ( I 133 1.93E6 3.27E6 9.98E5 4.57E8 5.74E6 --- 2.48E6 Ca 134 7.10E9 1.67E10 7.75E9 -- 5.31E9 2.03E9 2.08E8 Cs 137 1.01E10 1.35E10 4.69E9 -- 4.59E9 1.78E9 1.92E8 Ba 140 1.38E8 1.69ES 8.91E6 -- 5.74E4 1.14E5 2.13E8
*
- 14 140 1.69E4 8.32E3 2.21E3 -- -- --
4.78E8 Ce 141 2.83E5 1.89ES 2.17E4 -- 8.89E4 -- 5.40E8 Ce 144 5.27E7 2.18E7 2.83E6 -- 1.30E7 -- 1.33 E10
*mrea/yr per LCi/m3
** Daughter Decay Product. Activity level and effective half life assumed to i equal parent nuclide.
May 1986 lf
o i i TABLE 3-21 : RgVALUES - VEGETATION - ADULT - i
; a2 -eres/yr t uCi/se c '
-f 4
NUCLIDE BONE- LIVER T. BOD _Y TRYROID KIDNEY LUNG GI-LLI
*H 3 -- 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 - *C 14- - 8.97E5 1.79E5 1.79E5 1.79E5 1.79E5 1.79E5 1.79E5 i l .:
l Cr 51 -- -- 4.64E4 2.77E4 1.02E4 6.15E4- 1.17E7' [ Mn 54 -- 3.13E8- 5.97E7 -- 9.31E7 -- 9.58E8 'i s Fe 59 1.26E8 2.96E8 1.13E8- -- -- 8.27E7 1.02E9 1 co 58 -- 3 08E7 6.90E7- -- -- --- 6.24E8~ , ] Co 60 -- 1.67E8. 3.69E8 -- -- -- 3.14 E9 ? Tj Zn 65 3.17E8 1.01E9 4.56E8 -- 6.75E8- --
~ 6.36E8 j Sr 89 9.96E9 --
2.86E8 -- ' -- -- 1.60 E9 -
;"t Sr 90 6.05E11 --
1.48E11 -- -- -- 1.75E10 Zr 95 1.18E6 3.77E5 2.55ES -- 5.92E5 ' --'
- 1.20 E9 '
- * *Nb 9 5 2.41E5 1.34E5 7.20E4 --
1.32E5 -- 8.13E8 l Mo 99 -- 6.14E6 1.17E6 -- 1.39E7 -- 1.42E7 I 131 8.07E7 1.15E8 6.61E7 3.78E10 1.98E8 --- 3.05E7 I 133 2.08E6 3.61E6 1.10E6 5.31E8 6.30E6 -- 3.25E6 Cs 134 4.67E9 1.11E10 9.08E9 -- 3.59E9 1.19E9 1.94E8-l Cs 137 6.36E9 8.70E9 5.70E9 -- 2.95E9 9.81E8 1.68E8 l Ba 140 1.29E8 1.61ES 8.42E6 -- 5.49E4 9.25E4- 2.65E8
** La 1401.58E4 7.91E3 2.11E3 -- -- --
5.86E8
- Ce 141 1.97ES 1.33E5 1.51E4 --
6.19E4 -- 5.09E8 Ce 144 3.29E7 1.38E7 1.77E6 -- 8.16E6 -- 1.11E10
- mrem /yr per uCi/m3
** Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
n'{
-i TABLE 3-22' t DISPERSION PARAMETERS AT CONTROLLING LOCATIONS
- X/Q,% and W. VALUES- l DIRECTION DISTANCE (a)' 3 VENT OQ(sec/m-[ :D/0(ad)_
Site _ . Boundary *** E 1,600- 2.00 E .2.10E-9 !
-Inhalation -
and Ground E (104') 1,800 11'.42E-7 2.90E Plane j Cow Milk ESE (130') 4,300 4.11E-8 4. 73E-10 Goat Milk ** E (89') 12,500' 1.75E-8 1;33E-10 Meat Animal E (114') 2,600 1.17E-7 1.86E ! Vegetation E (96') 2,900 1.04E-7 1.50E-9'
}
STi:1 Site Boundary *** E 1,600 4.50E-8 ~ 6. 0 0E-9 ' Inhalation and Ground E (109') 1,700 8.4 8E-9 1.34E Plane Cow Milk ESE (135') 4,200 . 1.05E-8 3.'6 4E-10 ; Goat Milk ** E (94') 12,500 1.80E 1.84E-10 ; Meat Animal E (114') 2,500 1.13E-8 1.15E-9 l -; ! Vegetation E (96') 2,800 1.38E-8 9.42E-10 L NOTE: Inhalation and Ground Plane are annual average : values. Others-l are grazing season only.
*X/Q and D/Q values f rom NMP-2 ER-OLS.
** - X/Q and D/Q from C.T. Main Data Report dated November 198 5. - l
* ** X/Q and D/Q f rom NMP-2 FES, NUREG-1085, May 1985. l 1.
l May 1986
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Title:
. OFFGAS SYSTEM
, CHARCOAL ABSORBERS , m , _..mu. -_ 4<- . -
n y-- -- *- eor - - *dw-- e~ *-+=--
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Title:
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. 58 ,e se . .r. .soses.cs, GASEOUS RADIATION MONITORNG NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2.
FNAL SAFETY ANALYSIS REPORT
' AMENDMENT 23 DECEM8ER 190'
-.M M O
, PART CtAATE , IODINE , NOntE GAS
- COttECTION =
COLLECTION asEASumradENT
- STATION : STATION I.- STATION .~
l FETER [ l FITTER j .
- ww : W ,. : Sie ou= oiturou : raaico. : oo ,E : g i<g E*AARo.
as
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- < q :-
- * : : evada STe - .
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y \ 7 w/ ,7 p i P
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P
/ V P
Ny =
- m
~
FRESH AIR M FETER hP ROL TE " DETFCT -DE TECTOR tISOsuNETICp
' CET
~
' CMCK CHECK FLOW SENSORS SOURCE SOURCE l WEVES. ETC
._ SOURCE ggg llllll A44Pt FIER z7 a AoC AnsPtriER a Acc
[a Aoc I - _ . viDEOTERhdINAL __INDUSJRIAL__ E N- -- GAULTICHAfesEL CONTROLLER Af84tVZER = COasPUTi.9 ,9-==, yp tt8CA)
-,-,E.
ooAE - m . FEURE 3-6 BLOCK DIAGRAM TYPICAL GASEOUS EFFLUENT MONITORING SYSTEM NIAGARA MOHAWK POWER CORPORATON NINE MILE POINT-UNIT 2
. FNAL SAFETY ANALYSIS REPORT-
- ~ - - ~ ' - - - - - - '- " - ^ ' - - - ~ " - --- - - ~ - - ^ -
,._-.c- -- - - - - - - - - - + ~ ~ -
1 4.0. URANIUM FUEL CYCLE . The " Uranium Fuel cycle" is - defined d u 40 CFR Part 190.02. (b) as f follows: 7 "Ursuium fuel cycle means the operatioas of milling of uranium-ore, chemical conversion of uranium, isotopic enrichment of i uranium, f abrication of uranium: f uel, generation of electricity by a lightmrater-cooled nuclear. power plant- using uranium fuel, - and reprocessing of spent uranium fuel, to the extent that thess ' directly support the production of electrical . power : for; public . . use utilizing nuclear energy, but excludes " mining- operations, operations at waste disposal sites,. transportation of any ; radioactive material' in support ' of these operations and the reuse of. recovered non-uranium. special nuclear and by produc t materials frem the cycle." Section 3/4.11.4 of the Technical Specifications requires , that when the csiculated doses associated with the effluent releases exceed twice the applicable quarter or annual limits _ the licensen shal1 ~ evaluate the calendar year doses and, if required. . submit a Special _ Report to the NRC and limit subsequent releases such that the dose 4 commitment to a real individual from all uranium fuel cycle sources is limited to 25 arem to the total body or any organ (except the thyroid, which is limited to 75 area). This report is to demonstrate that radiation exposures to all' real individuals from all uranium fuel cycle sources (including all liquid 'and gaseous effluent pathways and direct radiation) are less than the . limits. in .40 CFR Part 190. If releases that result in doses exceeding the 40 CFR.190 limits have occurred , then a variance from ' the NRC to permit such ; releases will be requested and if possible, action will be taken to reduce subsequent releases. The report to the NRC shall containt
- 1) Identification of all uranium fuel . cycle facilities or operations within 5 miles of the nuclear pouer reactor units at L the site, that contribute to the annual dose of the maximum exposed member of the public.
- 2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from all existing pathways and sources ' of radioactive ; effluents and direct radiation.
The total body and organ doses resulting f rom radioactive material in
- liquid effluents from Nine Mile Point Unit 2 will be summed with the j doses resulting f rom the releases of noble gases, radioiodines, and i l.
particulates. The direct dose components will also be determined by. either calculation or actual measurement. Actual measurements will utilize environmental TLD dosimetry. Calculated measurements will utilize engineering calculations to determine a projected direct dose component. In the event calculations are used, the methodology will be detailed as required in Section 6.9.1.8 of the Technical l Specifie tions. The doses from Nine Mile Point Unit 2 w111-bs added ! to the done: tu the maximum exposed individual that are contributed from other urante fuel cycle operations within 5 miles of tne site. May 1986 l 1 l
4.0 (Cont'd) For the purpose of calculating doses.. the results_ of .the Environmental Monitoring program. may' be included to ~ provide acre - refined estimates' of doses to a real- maximum exposed individual. Estimated doses, as calculated from station . effluents,~ 'aay be replaced by doses calculated from actual environmental sample resultr.' 4.1 Evaluation of Doses From Liquid Effluents For the evaluation'of doses to real acabers of the public.from liquid 4 offluents, the - fish consumption and . shoreline - sedimentJground; dose will be considered. Since the doses from other aquatic 1 pathways see insignificant, fish consumption and shoreline sedimenti are the only two pathways that will be considered. The dose associated with fish consumption may be calculated using effluent - data and: Regulatory Guide 1.109 methodology or by calculating a dose to man - based 1 on actual fish sample analysis data. The dose associated withEshoreline - sediment- is based on tho' assumption that the shoreline would be utilized as a recreational area. This - dose may be; derived from-liquid effluent data and Regulatory Guide 1.109.aethodology or from actual shorelina sediment sample analysis data. Equations used to evaluate fish and shoreline sediment samples - are based on Regulatory Guide l'.109 methodology. - Because of the sample medina type and the half-lives of the radionuclidesi historically observed, the decay corrected' portions of the equations are deleted. This does not reduce the conservatism of ~ the calculated doses but increases the staplicity from an evaluation point of view.- The dose from fish sample andia is calculated as: [ (1) Ewb " I t [C ig x u x 1000 x Di wb I fI , Where: Rwb = The total dose to the whole body of an adult in area per year cit = The concentration of radionuclide- i in fish samples in pci/ seam u = The consumption rate of fish for an adult (21 kg Per year) 1000 = Grams per kilogram i Diwb= The dose factor for radionuclide i for the whole body.of an adult 'R.G. 1.109. Table E-11) f = The fractional portion of the year over which the dose is applicable. (2) R1 = It [Ci g x u a 1000 x Dit x f}
-90 February 1988
m , l I l
-\
Where: .;
., s R1
= The total: dose to the: liver, of 'an1 adult (masimum exposed -
l
' organ) la arom per year Cgf = The concentration of radionuclide i in fish' samplos : in-pci/ gram: .!
y = The consumption rate of fish for an adult (21 kg per year) , 1000 = Grams per kilogram The dose factor -for radionuclide i for ths . liver of ; an I s Dg3 .= adult (R.C. Table E-11) f = The fr:ci.ional portion of the year over which the dose-is applicable. The dose fron shoreline sediment sample media is-calculated as: 3 Rwb "'It 501s a p z 40,000 x 0.3 x Diwb 3 fl and Rsk
- Il (Cg, 2'y a 40,000 x 0.3 z' Disk 3 fl j Where:
Rwb = The total dose to the whole body of a- teenager or adult 4. (maximum exposed age group) in arem per year The total dose to the skin of a teenager or adult (maximum 4 Rsk = exposed age group) in mrom per year Cg, . The concentration of radionuclide.1 in shoreline sediment in pci/ gram p = The usage factor. This is assumed as-67 hours per year by 4 a teenaber or adult 40,000= The product of the assumed density of . shoreline sediment (40 kilogram per. square meter to a depth- of- 2.5 cm) times a the number of grams per kilogram N 0.3 = The shore width factor for a lake Diwb= The dose factor for radionuclide i for the total body-(R.C. 1.109, Table E-6) Disk = The dose factor for radionuclide i for the skin (R.C. 1.109, Table E-6) f = The fractional portion of the year c.7r which the dose is applicable EqIK: Because of the nature of the receptor location and the extensive fishing in the area,. the 4 critical individual may be a teenager or an adult. 2 -91 February 1988
63 Evaluation of Doses From Gaseous Eff* ent s For the evaluation of doses to real maskr. of the public f rom gaseous effluents, the tathways contained in section 3.0 of the ODCM will be considered end include ground deposition, inhaeation, cows milk, goats milk, seat. and food products (vegetation). However, any updated field data may be utilized that concarns locations of real individuals, real time seteorological data, location of critical receptors, etc. Data f rom the most recent census and sample location surveys should be utilized. Doses may also be calculated from actual environmental sample media, as available. Environmental sasple media data such as TLD, air sample, milk sample and vegetable (food crop) sample data any be utilized in lieu of effitent calculational data. Doses to members of the public from the pathwsya contained in ODCM section 3.0 as a result of gastous effluents will.be calculated using -- the dose factors of Regulatory Guide 1.109 cr the methodology of the - - - ODOle as applicable. Doses calculated from environmental sample media will utilize the methodologies found in Reg:ulatory Guide 1.109. 4.3 Evaluation of Doses From Direct Radiation Section 3.11.4.a of the Technical Specifications requires that the = dose contribution as a result of direct radiation be considered when ' i-evaluating whethe r the dose limitations of 40 CFR 190 have been = exceeded. Direct radiation doses as a resitit of the reactor, turbine and radwaste buildings and cutside radioactive storage tanks (as applicable) any be evaluated by engineering calculations or by evaluating environmental TLD results at critical receptor locations , site boundary or other speciti interest locations. For the evaluation of direct radiation doses utilizing environmental TLDs , the critical receptor in question, such sa the critical res10ence, etc., will be compared to the control locations. The comparison involves the differenca in environmental T1D results between the receptor location and the average contrci location retult. May 1986
4.4 Dosos to Nembers of the Public Withj A the site Soundary. ! l l Section 6.9.1.8 of the N!ae Nile Point Unit 2 Technical Specifications requires that the Samiannual . Radioactive Effluent i Release Report include an assessment of the radiation dossa from ;
- radioactive 11guld and gaseous effluents to member's of the public due 4 J to their activities inside the site boundary as defined by Figure I $ 1.3 of the specifications. A member of the public, as defined by the Tect.nical Specifications, would be represented by an individual who visits the sites' Energy Information Center for the purpose of
)
] observing the educational displays or for plenicing and assoelated j getivltles. Fishing is a major recreational activity la the area and on the site i as a result of the salmon!d and trout pcpylations in Lake Ontarlo. Fishermen have been observed fishing at the shoreline near the Energy
- Information Center from April through December in all weather j conditions. Thus, fishing is the major activity performed by members
- of the public within the dite boundary. Based on the nature of the fisherun and undocumented observations, it is conservatively assumed that the maalmum esposed individual spends an average of 8 hours per week fishing from the shoreline at a location between the Energy ;
7 'ormation Center and the Unit 1 facility. This estimate is i l considered conservative but not necessarily' escessive and accounts for occasions whero individuals may fish more on weekends or on a few days in March of the year.
^
The pathways considered for the evaluation include the Inhalation pathway with the resultant lung dose, the gropod dose pathway with the resultant whole body and skin oose and the direct radiation dose. pathway with the associated total body dose. The direct radiation dose pathway, in actuality, includes several pathways, 'the s e includes the direct radiation gnassa dose to an individual from an overhead plume, a gamma submersion plume dose, possible direct radi dion dose from the facility and a ground plano dose (deposition). Because the location 1s in close proximity to the site, any beta plume submersion dose is felt to be insignificant. , Other pathways, such as the ingestion pathway, are not applicable. In addition, pathways associated with water related recreational activitleu, other t h a.. fishing, are not applicable here. These include swinsni ng , boating and wading which are prohibited at the ' facility.
-93 February 1988
1 l 4 j 1 W i 4.4 (Coat'd) . The inhalation pathway is evaluated by Identifyleg the applicable ! radionuelldes (radiolodine, tritium and particulates) in the effluent i l for the appropriate time period. The radionuclide concentrations are i then multiplied by the appropriate I/Q value, inhalation dose f actor, l air intake rate, and the fractional portion of the year _ la question. , Thus, the inhalation pathway is evaluated u=ing the following ; equation adapted from Regulatory Culde 1.109. ;
.N0ZI The following equation is adapted from equations C-3 and '
C-4 of Regulatory Culde 1.109. Since many of the f actors are in units of pC1/m 3, m /sec., 3 etc., and since the radionuclide decay espressions have been deleted because of j the short distance to the receptor location, the equation , presented here is not identical to the Regulatory Culde , equations. .
- R =I g (CgF I/Q DFA gjaRt e) ,
where l R = the maalau dose for the period in question to the lung (j) for all radionuclides (1) for the adult age group (a) in mrom per time period. 4 Cge The average concentra tior.. In the stack or vent release of . radionuclide i lh PC1/m3 for the period in question. ! F = Unit 2 average stack or vent flowrote in m 3/sse. X/Q = The plume dispersion parameter for a location approximately ! 0.50 miles west of NNP-2 (The plume dispersion . parameters are 9.6E-07 (stack) and 2.82-06 (vent) s'd were obtained ' l frcm the C.T. Main five year average annual I/Q tables. A I/Q value based on real time meteorology may also be utilized for the period in_ question, if desired.' The vent I/Q (5round level) is ten times the listed 0.50 mile I/Q because the vent is approximately 0.3 miles from the receptor location. The stack (elevated) I/Q is conservative when based on 0.50 alles because of the close proximity of the stack and the receptor location. DFAgj. a the inhalation dose factor for radlonuclide 1 the lung j, and adult age Group a in mrem per pCl found on Table E-7 of Regulatory cuide 1.109. Ra = annual air intake for individuals in a6e group a in M3 per year (this value is 8,000 m 3 per year and was obtained from Table E-5 of Re6ulatory Culde 1.109). t = fractional portion of the year for which radionuclide I was detected and for which a dose is to be calculated (in years).
-94 February 1988
l I t I I 4.4 (Cont'd) i The ground dose pathway (deposition) will be evaluated by obtaining i et least one soll or shore 11ae sediment sample in the area where j fishing occurs. The dose will then be calculated using the sample results, the time period in question, and the methowlogy based on , Regulatory cuide 1.109 as presented in section 4.1. The resultant l dose may be adjusted for a background dose by subtracting the 4 applicable off-site control soil or shore 11ae sediment sample redlonuclide activities. In the event it is noted that fishing is not performed from the shore 11oe, but is instead performed in the i water (1.e., the use of waders), then the ground dose pathway . (deposition) will not be evaluated. l The direct radiation gansna dose pathway includes any gamma doses from an overherd plume, submersion in the plume, possible radiation from j
- the feellity and ground plane dose (deposition). This general pathway will be evaluated by average environmental TLD readings. At least two environmental TLDs will be utilized at one location in the [
approximate area where fishing occurs. The TLDs will be placed in the fleid on appro Imately the beginning of each calendar quarter and re. moved on approximately the end of each calendar. quarter (guarter 2 3, and 4). The average TLD readings will be adjusted by the av6 rage control TLD readings. This is accomplished by subtracting the average quarterly control TLD value from the average . fishing location TLD value. The , applicable quarterly control TLD values will be utilized after adjusting for the appropriate time period (as applicable). In the event of loss or theft of the TLDs. results from a TLD or TLDs in the area may be utilized.
-95 February 1988 ,
l l 5.0 ENVIRONMENTAL MONITORING PROGRAM t 5.1 Samplieg Stations The current sampling locations are specified in Table 5-1 and Figures ! 5.1-1, 5.1-2. The meteorological tower location is shown on Figure ! 5.1-1. The location is shown as TLD location #17. The gnvironmental l Monitoring Program is a joint effort between the Niagara Mohawk Power [ Corporation and the New York Power Authority, the owners and ! operaters of the N!ae Nile Point Units- 1 and 2 and the James { A.FitsPatrick Nuclear Power Plants, respectively. Sampling locations are chosen on the basis of historical average dispersion or deposition parameters from both units. The environmental sampling location coordinates shown on Table 5-1 are based on the NNP-2 ; reactor conterline, i i The average dispersion and deposition parameters for the three units have been calculated for a 5_ year period, 1978 through 1982. The ' calculated dispersion or deposition parameters will be compared to ! I the results of the annual land use census. If it is determined that 1 a milk sampling location esists at a location that yields a significantly h!gher (e.g. 50%) calculated D/Q rate, the new milk sampling locatloa will be added to the n.. 'toring program within 30 l days. If a new location is added, the olo location that yields-the t lowest calculated D/Q may be dropped from the program af ter October 31 of that year. 5.2 Interlsboratory Comparison Program Analyses shall be performed on samples containing known quantitles of radioactive materials that are supplied as part of a coussis sion approved or sponsored Interlsboratory Comparison Program, such as the EPA Crosscheck Program. Participation shall be only for those media, e.g., sir, allk, water, etc., that are included in the N!ne Mlle ' Polnt Environmental Monitoring Program and for which cross check, samples are available. An attempt will be made to obtain a QC sample I 4 to program sample ratio of 5% or better. The Quality Control sample ' results shall '., e reported in the Annual Radlolos! cal Environmental Operating Report so that the Coassis sion staff may evaluate the results. Specific sample media for which EPA Cross Check Program samples are available include the followingt
- gross beta in air particulate filters '
gamma emitters in air particulate filters
- I-131 in milk l
- gamuna emitters in allk
- gasuna emitters in food product a gamuna emitters in water e tritium in water
- I-131 in water i
-96 February 1988
t 5.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosiasters used for environmental measurements required by the Technical Specifications are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use. In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabilities may be determined by the vendor who supplies the TLDs. Required detection capabilities are as follows. 5.3.1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting f rom an exposure rate of- 10 uR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less than 7.5%. A total of at least 5 TLDs shall be evaluated. 5.3.2 Reproducibility shall be determined by giving TLDs repeated exposures equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The average of the relative standard deviations of the responses shall be less than 3.0%. A total of at least 4 TLDs shall be evaluated. 5.3.3 Dependence of exposure interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be - constant. This test shall be conducted under approximate average winter temperatures, and approximate average summer temperatures. For these tests , the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85 At least 6 TLDs shall be evaluated. 5.3.4 Energy dependence shall be evaluated by the response of TLDs to photons for several energies between approximately 30 kev and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25% for photons with energies greater than 80 kev and shall not be enhanced by more than a f actor of two f or photons with energies less than 80 kev. A total of at least 8 TLDs shall be evaluated. 5.3.5 The directional dependence of the TLD response shall be determined by 4 comparing the response of the TLD exposed in the routine orientation with respect to the calibration source with the response obtained for different orientations. To accomplish this, the TLD shall be rotated through at least two perpendicular planes. The response averaged over all directions shall not differ from the response obtained in the standard calibration position by more than 101. A total of at least 4 TLDs shall be evaluated. February 1988 ff
4 503.6 Light dependecce shall be determined by placing TLDs in the field for a period equal to the field cycle under the four conditions.found in ANSI N545, section 4.3.6. The results obtained for the unwrapped TLDs shall not differ from those obtained for the TLLe wrapped in aluminus foil by more than 101. A total of at least 4 TLDs shall be : evaluated for each of the four conditions. 5.3.7 Moisture dependence shall be determined by placing TLDs (that is, the phosphors packaged for field use) for a period equal to the field cycle in an area where the exposure rate is known to be constant. The TLDs shall be exposed under two conditions: (1) packaged in a thin, sealed plastic bag, and (2) packaged in a thin, sealed plastic , bag with sufficient water to yield observable moisture throughout the field cycle. The TLD or phosphor, as appropriate, shall' be dried 4 before readout. The response of the TLD esposed in the plastic bag i containing water shall not differ from that esposed in the regular plastic bag by more than .10%. A total of at least 4 TLDs shall be evaluated for each condition. ,
'I; 5.3.8 Self irradiation shall be determined by placing TLDs for a period -
equal te the field cycle in an area where the exposure rate is less than 10 uR/hr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3) . The average exposure inferred from the responses of : the TLDs shall not differ f rom the known exposure by more than an-exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least 3 TLDs shall be [ evaluated. i i February 1988
t t , f I i Nine Mlle Point Nuclear Station i , Radlological Environmental Monitoring Program l Sampling Locations ; Table 5.1 i t i Type of-
- Map }
S-le Location colleeklos Elke (Env. Pronram No.) Location { I t, Ridlolodine and 1 Nine Nile Point Road 1.8 al $ 88* E~ Particulates (air) north (R-1) Radiolodine and 2 Co. Rt. 29 & Lake Road 1.1 al 9 104* Est Particulates (alt) (R-2) i R:dlolodine and 3 Co. Rt. 29 (R-3) 1.5 mi 0 132* SE i Particulates (alr) ' Radiolodine and 4 Village of Lycoming, NY 1.4 mi S 143' SE $ Pcrticulates (alr) (R-4) t 1 R:diolodine and 5 Nontarlo Point Road 16.4 al 9 42' NE Particulates (air) (R-5) Direct Radiation (TLD) l 6 North Shoreline Area (75) 0.1 mi 0 5' N i i Direct Radiation (TLD) 7 North Shore 11ae Area (76) 0.1 mi 9 25' NNE )' Direct Radiation (TLD) 8 North Shoreline Area (77) 0.2 al 9 45' NE i Direct Radiation (TLD) 9 North Shoreline Area (23) 0.8 al 9 70* ENE f l Direct Radiation (TLD) 10 JAF East Boundary (78) 1.0 ml 9 90* E 'i Direct Radiation (TLD) 11 Rt. 29 (79) 1.1 al 9 115* ESE ! Direct Radiation (TLD) 12 Rt. 29 (80) 1.4 al 4 133' SE ; Direct Radiation (TLD) 13 Minor Road (81)- 1.6 al 9 159' SSE Direct Radiation (TLD) 14 N!ner Road (82) 1.6 al 9 181* S Direct Radiation (TLD) 15 Lakeview Road (83) 1.2 al 9 200* SSW ! Direct Radiation (TLD) 16 Lakeview Road (84) 1.1 al 4 225' SW Direct Radiation (TLD) 17 Site Noteorolo61 cal Tower (7) 0.7 al 0 250' WSW ! Direct Radiation (TLD) 18 Energy Information Center (18) 0.4 al A 765* W
- Map - See Figures 5.1-1 and 5.1-2 I
-99 February 1988
^
l l 1 1 N!as Mlle Point Nuclear Statloa j Radiological Environmental Monitoring Program l Sampling Locations l Table 5.1 1 (continued) Type of
- Nap S---le Location Collection Site 'Env. Pramram No. ) Location Direct P.adiation (TLD) 19 North Shore 11ae (45) 0.2 al 9 294* WNW I Direct Radiation (TLD) 20 North Shorellae (86) 0.1 al 9 315' NW Direct Radiation (TLD) 21 North Shore 11ae (87) 0.1 al 9 341' NNW
. I Direct Radiation (TLD) 22 Hickory Crove Road (88)- 4.5 al 9 97* E Direct Radiation (TLD) 23 Leavitt Road (89) 4.1 al 9 111* ESE Direct Radiation (TLD) 24 Rt. 104 (90) 4.2 al 9 135* SE Direct Radiation (TLD) 25 Rt. 51A (91) 4.4 ml 9 156' SSE I
Direct Radiation (TLD) 26 Nalden Lane Road (92) 4.4 al S 183* S j Direct Radiation (TLD) 27 Co. Rt. 53 (93) 4.4 al 9 205* SSW f Direct Radiation (TLD) 28 Co. Rt. 1 (94) 4.7 al S 223' SW Direct Radiation (TLD) 29 Lake shoreline (95)~ 4.1 al 6 237* WSW Direct Radiation (TLD) 30 Phoenis, NY Control (49) 19.3 al 9 170* S Direct Radiation (TLD) 31 S.W. Oswego, Control (14) 12.6 ml 9 226' SW l l Direct Radiation (TLD) 32 Scriba, NY (96) 3.6 al 9 199* SSW Direct Radiation (TLD) 33 Alcan Aluminum, Rt. 1A (58) 3.1 ml 9 220' SW Direct Radiation (TLD) 34 Lycoming, NY (97) 1.8 mi S 143* SE 1 Direct Radiation (TLD) 35 New Haven, NY (56) 5.3 al 9 123' ESE l Direct Radiation (TLD) 36 W. Boundary, Bible Camp (15) 0.9 al 9 237' WSW Direct Radiation (TLD) 37 Lake Road (98) 1.2 al 9 101* E Surf;se Water 38 OSS Inlet Canal (NA)- '.6 al S 235* SW Surface Water 39 JAFNPP Inlet Canal (NA) 0.5 al 9 70* ENE l l '
- Nap - See Figures 5.1-1 and 5.1-2
! (NA) - Not applicable
-100 February 1988
_______a
l l Nine Mile Point Nuclear Station 4 Radiological Environmental Monitoring Program i 8ampling Locations Table 5.1 (Continued) l Env. Type of " Map Program ! s---le Location Collection site 'No. Location l i shoreline sediment 40 Sunset Bay shoreline (NA) 1.5 mi 9 80* E Fish 41 NMP Site Discharge Area (NA) 0.3 al 9 315' WW i and/or Fish 42 NNP Site Discharge Area (NA) 0.6 mi 9 $5' WE Fish 43 Oswego Harbor Area (NA) 6.2 mi 9 235' sw - l Milk 44 Milk Location #50 9.3 mi 9 93* E Milk 45 Milk 1.ocation #7 ~5.5 mi 9 107' Est J I Milk 46 Milk Location #16 5.9 mi 9 190' S Milk 47 Milk Location #65 17.0 mi 9 220* SW Food Pcoduct 48 Produce Location #6** 1.9 al 9 141' SE ' (Sergenstock) - Food Product 49 Produce Location #1** 1.8 mi 9 96' E (J. Parkhurst) l Food Product 50 Produce Location #2** 1.9 mi 9 101' E (Vitullo) (NA) ; i Food Product 51 Produce Location #5** 1.5 mi 9 114' ESE (C. S. Parkhurst) Food Product 52 Produce Location #388 1.6 mi 9 84* ESF. (C. Warowski) Food Product 53 Produce Location #4** 2.0 mi 9 110* ESE (S. Morris) (NA) Food Product (CE) 54 Produce Location #7** 15.0 mi 9 223* SW (Mc Millon) (NA)
-101 February 1988 '
. _ - -- . _ - . - - . ._ - -..- _. ... . . _.~ - .. _ .- .
i Nine Nile Point Nuclear station Radiological Environmental Monitoring Progras , sampling Locations Table 5.1 (Continued) i tav. Type of 7.:p Program 4 8---le Locattua Collection Site No .' Location Food Product (CR) 55 Produce Location #8am 12.6 mi 9 225' SW ' (Denman) (NA) Food Product $6 Produce Location #988 1.6 al 9 171' S
. (O'Connor) (NA) !
Food Product $7 Produce Location #10** 2.2 mi S 123' E8E (C. Lawton) (NA) , Food Product 58 Produce Location #11** 2.0 mi S 112' ESE (C.R. Parkhurst) (NA)- Food Product 59 Produce Location #12** 1.9 mi S 103' Est (Johnson) (NA) i e
- Map - See Figures 5.1-1 and 5.1-2 l
** Food Product samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q. 4 (N/A) = not applicable CR = Control Result (location) l
-102 February 1988
_ ~ . . . _ _..._ _ _ _, , _ _ _
6.0 DISCUSSION OF TECHNICAL SPECIFICATION REFERENCES Section 3.12.1 of the Technical S pecifications. Table 3.12-1 (Radiological Environmental Monitoring Program) references several I f octnotes to discussions in the ODCN. The following ODCN discussions are an attempt, on the part of the Commission and the licensee, to f urther clarify several of the requirements of Table 3.12-1. 6.1 Table 3.12-1. Footnota s Representative composite sample aliquots are obtained f rom sampling equipment that will obtain sample aliquots over short intervals. An example of a short interval is once per hour. Intervals of less than one hour are also acceptable. In addition, in order to be representative, the aliquot volume must be consistent over the required composite period. Sub-intervals may be designed for sample collection as long as each sub-interval's contribution to the final composite volume is proportional to the duration of the sub-interval. For example, a monthly composite may consist of equal contributions from four weekly sub-intervals, plus a contribution 3/7 - of that volume f rom a fif th weekly sub-interval, to be representative of the monthly composite period. 6.2 Table 3.12-1, Footnote h Ground water in the vicinity of the site is not currently a drinking water pathway. The hydraulic gradient and rechange properties in the vicinity of the site currently cause ground water to flow in a northerly direction to lake Ontario. The results of such hydraulic gradient and recharge property studies are documented in the NMP-2 FSAR. Thus, any ground water utilized for drinking water or irrigation purposes is not affected by the site and therefore sampling of ground water is not currently required. s In the event of significant seismic activity, however, the hydraulic gradient and recharge properties in the vicinity of the site may change. In this case it is possible that ground water utilized f or drinking water or irrigation purposes may have a potential to become contaminated. Thus, in the event of a significant seismic occurrence, samples from one or two sources will be obtained as noted in Table 3.12-1, Section 3.b of the Technical Specifications until hydraulic investigations conclude that the previous hydraulic gradient and recharge property studies are unchanged. Investigations that conclude that the hydraulic gradient and recharge propertie s have changed and that there is a potential for contamination of ground water used for drinking water and/or irrigation will result in continuing any applicable ground water sampling.
-103- February 1988 t
6.3 Table 3.12-1, footnota i Currently, there are no drinking water sources (from Lake Ontario) that can be significantly affected by the site under normal operating , conditions. The closest drinking water source is near the City of Oswego. This source is located in an "up-current" direction for the ' majority of the time based on local Lake Ontario currants. In i addition, the source is significantly affected by the " plume" from-the Oswego River which enters Lake Ontario at a point between the site and the source. The source is located approximately eight ailes to the west of the site. Other drinking water sources within 50 miles of the site range from 20 to 50 miles. . These sources are beyond any sig;nificant influence of the site. In the event a drinking water source (ovher than the source near the -l City of Oswego) is established within 10 miles of the site (current ' l ailes in contrast to air miles), then the new source will- be I evaluated for any significant dose effects . based on dilution l criteria. Sources found to be significantly affected by the site l will be added to the Radiological Environmental Monitoring Program j required by Table 3.12-1, section 3.C of the Technical Specifications. i l 6.4 Table 3.12-1, footnote 1
]
Considering the shoreline topography and land development within 10 miles of the site, and the dilution factors beyond; 10 miles,. only , major irrigation projects where food products are irrigated with Lake 4 ontario water need be considered for specification 4.C of Table 3.12-1. Major irrigation projects are defined as agricultural projects where food products for human consumption are grown and - irrigation water from Lake Ontario is used frequently. Major irrigation projects are I not considered to be small private gardens located on the lake shore at suoner residences or year-round residences where occasional use of lake water during times of draught has been observed. Major projects ' include pumps and piping systems, either permanent or temporary, that supply lake water to agricultural projects on a frequent basis. ' In-frequent use of lake water is not considered to have a significant effect on food products. Therefore, such a situation does not constitute a major irrigation project. Currently, no major irrigation projects exist within 10 miles of the site (May 1986), s
-104- February 1988
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tsand eenA3 TECHNICAL SPECIFICATIONS FIGURE 5.1.3-1 SITE BOUNDARIES NINE MILE POINT UNIT 2
l l ,i
',0TES TO FIGt1tt 5.1.3-1 i
(a) NMP1 Stack (help *.t is 350') (b) hMP2 Stack (height is 430 j (c) JAFNPP Stack (height is 31 ) ' (d) NMP1 Radioactive Liquid Discharge (Lake Ontario, botton) ! l (e) NMP2 Radioactive Liquid Discharge (Lake Ontario, botton) (f) JAFNPP Radioactive Liquid Discharge (Lake Ontario, botton) (g) Site Boundary ; (h) Lake Ontario Shoreline (i) Meteorological Tower (j) Training Center . (k) Energy Information Center Additional Informations
- NMP2 Reactor Building Vent is located 187 feet above tround level JAFNPP Reactor and Turbine Building Vents are located 173 feet above ground level l I
JAFNPP Radwaste Building Vent is 112 feet above ground level
- The F.norgy Information Center and adjoining picnic area are UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC Lake Road, a private road, is an UNRESTRICTED AREA within the SITE BOUNDARY accessible to MEMBERS OF THE PUBLIC i
NINE MILE POINT UNIT 2
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