VERMONT YANKEE NUCLEAR POWER STATION
OFF-SITE DOSE CALCULATION MANUAL
REVISION 32
This document contains Vermont Yankee proprietary information. This information may not be transmitted, in whole or in part, to any other organization without permission of Vermont Yankee.
Effective Date: 06/24/08 _
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Off-Site Dose Calculation Manual Rev. 32 Page i of x Vermont Yankee Nuclear Power Station REVISION SUMMARY
DATE REVISION DESCRIPTION
06124/08 32 Section 6.11 Factors revised in accordance with AREV A Calc. #32-9068362-000 recommendations REFERENCE Section - Added ARE VA Calc. Reference
Off-Site Dose Calculation Manual Rev. 32 Page ii of x Vermont Yankee Nuclear Power Station ABSTRACT
The VYNPS ODCM (Vermont Yankee Nuclear Power Station Off-Site Dose Calculation Manual) contains the effluent and environmental control limits, and approved methods to estimate the maximum individual doses and radionuclide concentrations occurring at or beyond the boundaries of the plant due to normal plant operation. The effluent dose models are based on the U.S. NRC Regulatory Guide 1.109. Revision 1.
With initial approval by the U.S. Nuclear Regulatory Commission and the VYNPS Plant Management and approval of subsequent revisions by the Plant Management (as per the Technical Specifications) the methods contained in the ODCM are suitable to demonstrate compliance with effluent controls.
Off-Site Dose Calculation Manual Rev. 32 Page iii of x Vermont Yankee Nuclear Power Station TABLE OF CONTENTS REVISION SUMMARY ii ABSTRACT iii TABLE OF CONTENTS iv LIST OF TABLES viii LIST OF FIGURES x
Section-Page
1.0 INTRODUCTION 1-1
1.1 Summary of Methods, Dose Factors, Limits, Constants, and Radiological Effluent Control Cross References 1-2
2.0 DEFINITIONS 2-1
3/4.0 EFFLUENT AND ENVIRONMENTAL CONTROLS 3/4-1
3/4.1 Instrumentation 3/4-1 3/4.1.1 Radioactive Liquid Effluent Instrumentation 314-1 3/4.1.2 Radioactive Gaseous Effluent Instrumentation 3/4-6 3/4.2 Radioactive Liquid Effluents 3/4-11 3/4.2.1 Liquid Effluents: Concentration 3/4-11 3/4.2.2 Dose - Liquids 3/4-15 3/4.2.3 Liquid Radwaste Treatment 3/4-16 3/4.3 Radioactive Gaseous Effluents 3/4-17 3/4.3.1 Gaseous Effluents: Dose Rate 3/4-17 3/4.3.2 Gaseous Effluents: Dose from Noble Gases 3/4-20 3/4.3.3 Gaseous Effluents: Dose from 1-131,1-133, Tritium, and Radionuclides in Particulate Form 3/4-21 3/4.3.4 Gaseous Radwaste Treatment 3/4-22 3/4.3.5 Ventilation Exhaust Treatment 3/4-23 3/4.3.6 Primary Containment 3/4-24 3/4.3.7 Steam Jet Air Ejector (SJAE) 3/4-25 3/4.4 Total Dose 3/4-26 3/4.4.1 Total Dose 3/4-26 3/4.5 Radiological Environmental Monitoring 3/4-27 3/4.5.1 Radiological Environmental Monitoring Program 3/4-27 3/4.5.2 Land Use Census 3/4-36 3/4.5.3 Intercomparison Program 3/4-37 3/4.6 Effluent and Environmental Control Bases 3/4-38
Off-Site Dose Calculation Manual Rev. 32 Page iv of x Vermont Yankee Nuclear Power Station Section-Page
5.0 METHODS TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS 5-1
5.1 Method to Determine F I ENG and C I NG •••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••••• 5-1 5.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway 5-3 5.2.1 Sample Tanks Pathways 5-3 5.2.2 Service Water Pathway 5-3 5.2.3 Circulating Water Pathway 5-4
6.0 OFF SITE DOSE CALCULATION METHODS 6-1 6.1 Introductory Concepts 6-1 6.2 Method to Calculate the Total Body Dose from Liquid Releases 6-3 6.3 Method to Calculate Maximum Organ Dose from Liquid Releases 6-9 6.4 Method to Calculate the Total Body Dose Rate from Noble Gases 6-11 6.5 Method to Calculate the Skin Dose Rate from Noble Gases 6-16 6.6 Method to Calculate the Critical Organ Dose Rate from Iodines, Tritium and Particulates with Tl/2 Greater Than 8 Days 6-22 6.7 Method to Calculate the Gamma Air Dose from Noble Gases 6-26 6.8 Method to Calculate the Beta Air Dose from Noble Gases 6-29 6.9 Method to Calculate the Critical Organ Dose from Iodines Tritium and Particulates 6-32 6.10 Receptor Points and Annual Average Atmospheric Dispersion Factors for Important Exposure Pathways 6-39 6.11 Method to Calculate Direct Dose From Plant Operation 6-45 6.12 Cumulative Doses 6-55
7.0 ENVIRONMENTAL MONITORING PROGRAM 7-1
8.0 SETPOINT DETERMINATIONS 8-1
8.1 Liquid Effluent Instrumentation Setpoints 8-1 8.2 Gaseous Effluent Instrumentation Setpoints 8-10
9.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS, AND RADWASTE TREATMENT SYSTEMS 9-1
9.1 In-Plant Radioactive Liquid Effluent Pathways 9-1 9.2 In-Plant Radioactive Gaseous Effluent Pathways 9-4
10.0 REPORTING REQUIREMENTS 10-1
R. REFERENCES R-I
Off-Site Dose Calculation Manual Rev. 32 Page v of x Vermont Yankee Nuclear Power Station Section-Page
APPENDIX A: Method I Example Calculations Deleted
APPENDIX B: Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont yankee B-1
APPENDIX C: Response to NRCIEG&G Evaluation of ODCM Update Through Revision 4 *On File
APPENDIX D: Assessment of Surveillance Criteria for Gas Releases from Waste Oil Incineration D-1
APPENDIX E: NRC Safety Evaluation for Disposal of Slightly Contaminated Soil On-Site at VY (Below the Chern Lab Floor) - TAC No. M82152 *On File
APPENDIX F: Approval Pursuant to lOCFR20.2002 for On-Site Disposal of Cooling Tower Silt F-1
APPENDIX G: Maximum Permissible Concentrations (MPCs) in Air and Water Above Natural Background Taken from lOCFR20.1 to 20.602, Appendix B G-1
APPENDIX H: H-1
1) "Request to Amend Previous Approvals Granted Under 10CFR20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil," dated June 23,1999, BVY 99-80 H-2
2) "Supplement to Request to Amend Previous Approvals Granted Under lOCFR20.302(a) to Allow for Disposal of Contaminated Soil," dated January 4, 2000, BVY 00-02. H-19
3) "Vermont Yankee Nuclear Power Station, Request to Amend Previous Approvals Granted Under lOCFR20.302(a) to Allow for Disposal of Contaminated Soil (TAC No. MA5950)", dated June 15, 2000, NVY 00-58 H-37
Off-Site Dose Calculation Manual Rev. 32 Page vi of x Vermont Yankee Nuclear Power Station Section-Page
APPENDIX I: 1-1
1) "Request to Amend Previous Approval Granted Under lOCFR20.2002 for Disposal of Contaminated Soil", dated September 11, 2000 BVY 00-71 1-2
"Vermont Yankee Nuclear Power Station - Safety Evaluation for an Amendment to an Approved 10CFR20.2002 Application (TAC No. MA9972)", dated June 26, 2001, NVY 01-66 1-5
APPENDIX J: J-l
2) "Request to Amend Previous Approval Granted Pursuant to lOCFR20.200 1 for Increase of the Annual Volume Limit and One-time Spreading of Current Inventory," dated October 4, 2004, BVY 04-110 J-2
3) "Safety Evaluation of Request to Amend Previous Approvals Granted Pursuant to lOCFR20.2002 - Vermont Yankee Nuclear Power Station (TAC No. MC5104)" dated July 19,2005, NVY 05-090 J-51
* To access this document, go to the Electronic Document Management System. Search using ODCM.
Off-Site Dose Calculation Manual Rev. 32 Page vii of x Vermont Yankee Nuclear Power Station LIST OF TABLES
Number Title Section-Page 1.1.1 Summary of Radiological Effluent Controls and Implementing Equations 1-3 1.1.2 Summary of Methods to Calculate Unrestricted Area Liquid 1-6 Concentrations 1.1.3 Summary of Methods to Calculate Off-Site Doses from Liquid 1-7 Concentrations 1.1.4 Summary of Methods to Calculate Dose Rates 1-8 1.1.5 Summary of Methods to Calculate Doses to Air from Noble Gases 1-9 1.1.6 Summary of Methods to Calculate Dose to an Individual from Tritium, 1-10 Iodine, and Particulates in Gas Releases and Direct Radiation 1.1.7 Summary of Methods for Setpoint Determinations 1-12 1.1.8 Effluent and Environmental Controls Cross-Reference 1-13 1.1.9 (Table Deleted) 1.1.10 Dose Factors Specific for Vermont Yankee for Noble Gas Releases 1-16 1.1.lOA Combined Skin Dose Factors Specific for Vermont Yankee Ground Level 1-17 Noble Gas Releases 1.1.11 Dose Factors Specific for Vermont Yankee for Liquid Releases 1-18 1.1.12 Dose and Dose Rate Factors Specific for Vermont Yankee for Iodines, 1-19 Tritium, and Particulate Releases 2.1.1 Definitions 2-2 2.1.2 Summary of Variables 2-4 3.1.1 Liquid Effluent Monitoring Instrumentation 3/4-2 3.1.2 Gaseous Effluent Monitoring Instrumentation 3/4-7 4.1.1 Liquid Effluent Monitoring Instrumentation Surveillance Requirements 3/4-4 4.1.2 Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 3/4-9 4.2.1 Radioactive Liquid Waste Sampling and Analysis Program 3/4-12 4.3.1 Radioactive Gaseous Waste Sampling and Analysis Program 3/4-18 3.5.1 Radiological Environmental Monitoring Program 3/4-28 3.5.2 Reporting Levels for Radioactivity Concentrations in Environmental 3/4-33 Samples 4.5.1 Direction Capabilities for Environmental Sample Analysis 3/4-34 6.2.1 Environmental Parameters for Liquid Effluents at Vermont Yankee 6-7 6.2.2 Usage Factors for Various Liquid Pathways at Vermont Yankee 6-8
Off-Site Dose Calculation Manual Rev. 32 Page viii of x Vermont Yankee Nuclear Power Station Number Title Section-Page 6.9.1 Environmental Parameters for Gaseous Effluents at Vermont Yankee 6-36 6.9.2 Usage Factors for Various Gaseous Pathways at Vermont Yankee 6-38 6.10.1 Atmospheric Dispersion Factors 6-42 6.10.2 Site Boundary Distances 6-43 6.10.3 Recirculation Correction Factors 6-44 7.1 Radiological Environmental Monitoring Stations 7-4 8.2.1 Relative Fractions of Core Inventory Noble Gases After Shutdown 8-21
Off-Site Dose Calculation Manual Rev. 32 Page ix of x Vermont Yankee Nuclear Power Station LIST OF FIGURES
Number
7-1 Environmental Sampling Locations in Close Proximity to Plant 7-7
7-2 Environmental Sampling Locations Within 5 km of Plant. 7-8
7-3 Environmental Sampling Locations Greater Than 5 km from Plant 7-9
7-4 TLD Locations in Close Proximity to Plant 7-10
7-5 TLD Locations Within 5 km of Plant 7-11
7-6 TLD Locations Greater than 5 km from Plant 7-12
9-1 Liquid Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee 9-9
9-2 Gaseous Effluent Streams, Radiation Monitors, and Radwaste Treatment System at Vermont Yankee 9-10
Off-Site Dose Calculation Manual Rev. 32 Page x of x Vermont Yankee Nuclear Power Station .,- INTRODUCTION
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7)0. Jt om Rtxo "Syn vo Ryxr Jk' ~ -,0= an gpn 7,..,0 btno V &
7)0/ bu stxo "ao tx Rtxo & JbPa ~ -,-.7 a Pa gp P 7,..,0 7)00 bu stxo "JD f& JbPJ ~ -,-.3 abPJ gp P 7,..,0
Vpp)bt o Jy o Ikvm vk tyx Skx kv bom tyx .,- ao , 0. Ykro .- yp .= eo wyx hkxuoo T mvok Yy o b k tyx Y=ANE ,','1 ~Ctsynszih"
Evzeynts Tikiwisgi Pzrfiw Ceyiltw Evzeynts Vigynts 1&./ Tixns Nnsiw Ywesxkiw 1',,'. DYwes; )'))-0 T Ywes + Y Ywes ~_sxmniphih"
1&.0 Msyiwrthzpew Jeu Dtxi DJeu ; -'//E&- bJeu =TN kJeu 1',,'.
1&.1 MVIVM Mshn nhzep Cexo DtxiDig&-11r; .',:E&0 c L 1',,'/ ey bixy Vnyi Atzshew & -11 riyiwx
1&.3 MVIVM Mshn nhzep Cexo DtxiDig&.))r ; -')0E&0 c ! 1',,'/ ey bixy Vnyi Atzshew & .)) riyiwx
1&.7 Dtxi ey bixy Vnyi Deg&Ytyep; ~Dig&-11rc P~-11r 1',,'/ Atzshew kwtr epp gexox ts ~ ~Dig&.))r c P~.))r ymi MVIVM Seh
Rkk&Vnyi Dtxi Cepgzpeynts Oeszep Vigynts ,') Ti ' ., Seli MM tk ,: aiwrtsy desoii Pzgpiew St iw Vyeynts eJLVO 1.1.=
d xxl zr Yp tzo rz dp zuy Np p xuyl uzy
O l uzy cprp pynp _ xmp Ml psz O l uzy dpn uzy
A-1 Vu uo Orrw py D Vu uo clo l p Nu ntl sp A.1.1.1 Yzyu z &1=/7:0' cE &n iI ~~uy dh ~ n,
Rl pz Orrw py D bwly d lnv &cc-10A-1J, cc-10A-1L' lyo JaR arrsl d px &713=,713A' _zmwp Rl Jn u u Yzyu z A-C ez lw Lzo A.3.1.1 c~~ &n x'I ;;;ds~k1- PNPLn A-10 dvuy A.3.1.1 c vuy&n xiI 7000d !~k1k s P NP"n
A-31 dTJO _zmwp Rl Jn u u A.3.3.1 cdTJO&xcSt ' I 1 ;O) ~!, 1 Yzyu z s + &1=/1:0J, 1=/1:0L'
arr-du p Nz p Mlwn wl uzy Yly lw dpn uzy 1.0 cp . 71 blsp 13 zr 1C fp xzy glyvpp _ nwpl bz p d l uzy cACRI .).);
Immt lv gvk Iv pzwvulv gt Dwv zwt Dzw 'almlzlvil
Vzpnpvgt cliovpigt al p lk VEDS Dwv zwt cwxpi bxlipmpig pwv bli pwv Dwv zwt bli pwv NTbcadSITcAcNVT agkpwgi p l Rpy pk Immt lv 0,1)=)A 0,1).). Nv z ulv g pwv Immt lv pv z ulv g pwv tp cghtl 0)=). cghtl 0).). Nv z ulv z lpttgvil zly pzlulv cghtl 1)=). cghtl 1).). agkpwgi p l Lg lw Immt lv 0,1)=)C 0,1).)/ Nv z ulv g pwv Immt lv pv z ulv g pwv tp cghtl 0)=)/ cghtl 0).)/ Nv z ulv g pwv zly pzlulv cghtl 1)=)/ cghtl 1).)/ aAENVADcNeI RN_dNE )IJJRdITcb Dwvilv zg pwv 0,1);)A 0,1)/). Rpy pk g l guxtpvn ~ gvgt p cghtl 1);). cghtl 1)/). xzwnzgu Ew l ' Rpy pk 0,1);)C 0,1)/)/ Rpy pk agk g l czlg ulv 0,1);)D 0,1)/)0 aAENVADcNeI LAbIVdb IJJRdITcb Ew l ag l 0,1);)I 0,1)0). Lg lw g l guxtpvn ~ gvgt p cghtl 1);)/ cghtl 1)0). xzwnzgu Ew l mzwu Twhtl Lg l 0,1);)J 0,1)0)/ Ew l mzwu .'.0.&.'.00& czp p u gvk 0,1);)L 0,1)0)0 agkpwv itpkl pv Ygz pi tg l Jwzu Lg lw agk g l czlg ulv 0,1);)M 0,1)0)1 elv ptg pwv I og czlg ulv 0,1);). 0,1)0)3 Yzpugz Dwv gpvulv 0,1);)R 0,1)0)7 b lgzv Ol Apz Irli wz /.4-8-:~ 0,1)0):" cVcAR EVbI cw gt Ew l 0,1);)S 0,1)1).
~ bxlipmpig pwv 0,1);)P zlugpv pv xtgv cliovpigt bxlipmpig pwv gvk p k xtpig lk pv gEDS Dwv zwt 0,1)0):) Vmm'bp l Ew l Dgti tg pwv Sgv gt bli pwv .)- al ) 0. Ygnl .0 wm .= elzuwv fgvsll T itlgz Yw lz b g pwv aCDOJ .).); ~Evu ou ki"
Syomoufs akhnuohfs Yk ozki SIEP Evu yvs avwoh _wkholohf ovu _kh ovu Evu yvs _kh ovu YCINSOSMNECO JRcNYSRPJRaCO PSRNaSYNRM Yfiovsvmohfs Ju oyvutku fs Pvuo vyoum 0,1)=)E 0,1)3). Tyvmyft Ooz oum vl ykx oyki tvuo vyoum hyo kyof afgsk 0)=)0 afgsk 0)3). Ykwvy oum sk ksz lvy yfiovfh o o ou afgsk 0)=)1 afgsk 0)3)/ zftwskz Ik kh vy hfwfgoso lvy ku oyvutku fs afgsk 1)=)0 afgsk 1)3). fufs zoz Ofui bzk Ekuz z 0,1)=)I 0,1)3)/ Nu kyhvtwfyozvu Tyvmyft 0,1)=)J 0,1)3)0
JLLObJRa ESRaYSO DC_J_ DfzkzA 0);~ 0)= 0,1)7 bRNVbJ YJTSYaNRM YJVbNYJPJRa_ Cuu fs Yfiovfh o k Jlls ku Ykskfzk 7):)E)s .-). Ykwvy Ju oyvutku fs Yfiovsvmohfs Pvuo vyoum 7):)E)0 .-)/ _wkhofs Ykwvy z 7):)E)/ .-)0 Pfpvy Enfumkz v Yfiovfh o k Oox oi& 7).1 .-)1 Mfzkv z& fui _vsoi dfz k aykf tku _ z ktz
Sll'_o k Ivzk Efsh sf ovu Pfu fs _kh ovu .)- Yk ) 0. Tfmk .1 vl .= ckytvu efurkk R hskfy Tv ky _ f ovu J01;: ,','/
~7R_RfRP"
CSS&IVfR 7ceR 3N_Og_NfVcb =NbgN_ IROfVcb ,') ERh' -, DNTR ,. cS ,/ LRdacbf MNbYRR AgO_RNd DciRd IfNfVcb cDESL r,r,.Y Juxl Mgiyuwx bvliomoi muw dlwsuty fgtpll muw Vuhrl Ngx alrlgxlx
agkoutzi.okl Ngssg cuygr Euk Elyg bpot Iushotlk bpot Elyg Dow Ngssg Dow Juxl Mgiyuw Juxl Mgiyuw Juxl Mgiyuw Juxl Mgiyuw Juxl Mgiyuw "bygip alrlgxl& JME4 JMb 4 +- 4! JMm JM= 0 s0 0 "swls) s & " wtwls) & "swls~ xli& 0 vIo) w "swgC) s & " swgC) s & vIo) w P,y IO) w vIr) w vIr) w . Dw)1. :,:1L)-0' /,3;L)-0 .,/-L)-/ 0,/:L)-0 ;,0-L)-0 Rw):0s . d! al~,f ----- r,3.L)Yb /,::L)-1 r,;0L)Yb £ ~f . .,.7L)-0 .,13L)-0 J!J l~,J .,;7L)-0 .,/0L)-0 £ ~f r,3.L)Yb . r,01L)-0 /,.bL)-0 r,;bL)-0 /!d;l~« Rw):7 . !i;l~,J ;,70L)-0 /,-3L)-/ r,-0L)-/ 3,.7L)-0 Rw):: .,17L)-/ . /,07L)-0 .,31L)-/ /,;0L)-0 y! ;l~,; Rw):; . r,33L)-/ r,YOL)-/ 0,-bL)-/ r,-3L)-/ r,70L)-/ Rw);- . y! al~,; 7,/;L)-0 ;! ;l~,; 7,:0L)-0 .,30L)-/ el).0.s . i!/ l~« 1,73L)-1 :,:7L)-1 r,r.L)-0 y! al~,M el).00s . ;! /l~,M ;,;1L)-1 /!f l~,J .,1:L)-0 0,/7L)-1 el).00 . /,;1L)-1 0,-3L)-1 7,:.L)-1 .,-bL)-0 J! Jl~,M el).0bs . 0,./L)-0 7, rrL)-1 0,;0L)-0 7,0;L)-1 0,03L)-0 el).0b .,:.L)-0 . .,:3L)-0 M! al~,J /,13L)-0 .,;/L)-0 el).07 . .,1/L)-0 .,//L)-/ /,-7L)-/ .,/7L)-/ r,brL)-0 el).0: . :,:0L)-0 1, .0L)-0 .,10L)-/ M!d l~,J ;,/.L)-0
~ :,:1L)-0 A :,:1 e .-) 0
Ymm)boyl Juxl Igrizrgyout Tgtzgr bliyout .,- al , 0. _gnl .3 um.; dlwsuty fgtpll Vzirlgw _u lw bygyout Y=ANE -)-)mR=
Cpndkogf Vlko Dptg Iceupst Vrgekhke hps _gsnpou bcolgg Jspvof Ngwgm Ppdmg Jct Tgmgctgt
Tcfkpovemkfg & ~nsgn' tge" .4! C) ~ -+ ys
=s'0- -)70E'R- Ms':/O -)1:E',0 Ms':1O 3)/0E',. Ms':1 0);-E',. Ms':7 0),3E'Rm Ms':: .) mmE'RL Ms':; 1)mRE'RL Ms';, /);;E'Rm ag'-/-O -):7E',. ag'-//O /);,E',. ag'-// -)0-E',. ag'-/1O 1)/0E',. ag'-/1 :)/3E',. ag'-/7 0)1:E'Rm ag'-/: .).3E'Rm
Rhh'Vkug Dptg Ccmevmcukpo Ocovcm Vgeukpo -), Tgw)/m Scig -7 ph -; _gsnpou bcolgg Pvemgcs Spxgs Vucukpo T=ALE 1.1.11
Dotf Ibduost Spfdigid gos Vfsmonu _bnkff gos Lirvie Rflfbtft
Rbeionvdlief Toubl Aoez Mbyimvm Oshbn Dotf Ibduos Dotf Ibduos ~ msfm" DILiuclCi DILimolCi~ msfm"
J'/ .),3E',0 .),3E',0 Nb'.0 /)/:E',. /)/:E',. Cs'1- /)-,E',0 3);3E',. Mn'10 .),:E'Ol /),,E&OO :)S/E',3 1).;E',/ 0)g:E',. .)10:',- .)0;E'Ol -):0E&OO 1);7E',. 0)/0E'Ol Co'3, .) -/E'Ol -).:E&OO an'31 :),3E&OO -)30E&O- Ss':; .)11E'Ol :);-E&OO Ss';, 0)./E&Ol -)37E&,. as';1 0).-E',0 -)/3E'Ol Mo';; 0)7;E',/ 0)1-E',. Td';;m 1),0E',3 .)//E',0 =h'l l Om 3);,E',/ 7),.E'Ol Sc'-.0 :)00E',/ .)..E'Ol Sc'-.1 7)1.E',/ -)-1E'Ol -'-/- .)17E',. -)07E&,- -'-/. /)lOE',3 -).;E',0 -'-// /)/-E',/ -)3/E&OO -'-/1 /) -3E',0 1);,E',. Ct'-/0 -).:E&,. -)3,E&,. Ct'-/7 7)1:E&Ol -).-E&,. Ab'-0, 0),:E',/ ;)7.E',. Cf'-0- .)/-E',1 0)lOE',. Y'-:7 -)-:E',. :);,E&,,
Ogg'Siuf Dotf Cbldvlbuion Mbnvbl Sfduion -), Rfw) /- Pbhf -: og -; Vfsmonu _bnkff Nvdlfbs Poxfs Subuion bDERL 1.1.12
Jvyk fui Jvyk _fzk Mfhzvxy awkholoh lvx ckxtvuz efurkk lvx sviouky, bxozo t, fui Yfxzoh sfzk _kskfyky
azfhr _kskfyk Nxv ui Rk ks _kskfyk~ _fiov- Ixozohfs Vxmfu Ixozohfs Vxmfu Ixozohfs Vxmfu Ixozohfs Vxmfu u hsoik Jvyk Mfhzvx Jvyk _fzk Mfhzvx Jvyk Mfhzvx Jvyk _fzk Mfhzvx &txkt' , &txkt-ykh ' & txkt' , & txkt-ykh ' JMN yohv --f- JMN yohf I" JMN mohv JMN mohv I" x -P~ 0 --f- .+!- 0
O-3 7.=CL-/7 0.:0L-/1 0.sVL-/1 3.7=L-Vs I-07 1.C0L-Vs C.0AL)VV ;.;AL)VV 1. ssL)/1 Ix-:0 A.0:L-/3 1.A1L-Vs 7.7:L-/1 0.:3L)VV Su-:7 C.37L-Vs 3.;CL)Vs 7.C=L)VV 0.C:L)/1 Mk-:: 7.11L-Vs 0.33L)Vs 1.1/L)VV ;.C7L)Vs Mk-:C C.33L-Vs 3./CL)Vs 7.C1L)VV 0.;3L)/1 Iv-:= 1.=3L-Vs 0./3L)Vs 0.71L)VV :.3/L)Vs Iv-:A 7.A3L-Vs 0.=3L)Vs 1.:;L)VV C.01L)Vs Iv-;/ 0./1L)Vs 7.:7L)/1 :.31L)Vs 1.31L)/3 1u-;: 7.C;L)VV 0.;/L)/1 1.:3L)Vs A.1/L)/1 ak-=: 3.0CL)VV 0./3L)/1 0.;3L)Vs :.1=L)/1 au-003 0.7=L)VV 7.=;L)Vs =.:;L)VV 1.7;L)/1 ax-AC 0.:1L)Vs 7.=CL)/1 =.=CL)Vs 1.7;L)/3 ax-C/ :.=:L)/1 0.A0L)/7 3./3L)/3 C.:;L)/7 1x-C: C.13L-Vs 3./7L)Vs 7.A7L)VV 0.:CL)/1 ag-017 0.;CL)VV :.;:L)Vs A.C:L)VV 1.CCL)/1 ag-01: 0.;=L)VV ;.:3L)Vs A.:CL)VV 3.37L)/1 0-030 0./3L)/1 3.1:L)/3 :.3AL)/1 0.=/L)/7 0-033 0.0VL)VV 3.7=L)Vs A.=:L)VV 1.=;L)/1 Iy-037 1.01L)Vs =./;L)/1 0./CL)/1 3.;3L)/3 Iy-03= 1.0=L)Vs =.70L)/1 0.00L)/1 3.=AL)/3 Ef-07/ 0.:/L-Vs 7.A;L)VV 1.1;L)VV =.0CL)Vs Ik-070 1.1;L-Vs =.11L)VV 0.1=L)VV 7./7L)Vs Ik-077 :. 07L)VV 0.;1L)/1 1.;CL)Vs A.:1L)/1
~ bnk xkskfyk wvouz xklkxkuhk oy znk Tvxzn dfxknv yk. bnkyk ivyk fui ivyk xfzk lfhzvxy fxk hvuykx fzo k lvx wvzkuzofs xkskfyk fwwsohfzovuy fyyvhofzki ozn mxv ui sk ks klls kuzy lxvt vznkx tfpvx lfhosozoky &o.k., b xgouk E osioum, _kfhzvx E osioum, DVN, fui IDE'.
Vll-aozk Jvyk Ifsh sfzovu Sfu fs akhzovu 0./ _k . 30 Yfmk 0C vl 0C ckxtvuz efurkk T hskfx Yv kx azfzovu -') CDINONUNROT
CRSd dNLeSa_ VSded MNOS_SeSa_d ~CIJVN -','," I_M MadN NbfIeSa_ gIcSIJVN _IYNd ~CIJVN -','-" hRSLR IcN feSVSiNM S_ eRN-!.+~ DE
:OO&ASeN 0adN /IVLfVIeSa_ 3I_fIV ANLeSa_ -') =Ng' .) ;IPN 1 aO1: DNcYa_e EI_TNN 7fLVNIc ;ahNc AeIeSa_ Y=ANE /).). Dfgkpkwkrpv
.) Jbvfrxv Tbezbvwf Yufbwofpw V vwfo ' Yif =xhofpwfe Rgg'Jbv V vwfo ~=RJ" kv wif hbvfrxv ubezbvwf wufbwofpw v vwfo zikdi ibv cffp efvkhpfe bpe kpvwbnnfe wr ufexdf ubekrbdwkyf hbvfrxv fggnxfpwv c drnnfdwkph sukobu drrnbpw v vwfo rgg'hbvfv guro wif sukobu v vwfo bpe surykekph gru efnb ru irnexs gru wif sxusrvf rg ufexdkph wif wrwbn ubekrbdwkykw sukru wr ufnfbvf wr wif fpykurpofpw)
/) Lrw Vwbpec ' Lrw vwbpec ofbpv rsfubwkrp zkwi wif ufbdwru dukwkdbn bpe wif obkp vwfbo nkpf kvrnbwkrp ybnyfv dnrvfe)
0) Moofekbwf ' Moofekbwf ofbpv wibw wif uftxkufe bdwkrp zknn cf kpkwkbwfe bv vrrp bv subdwkdbcnf drpvkefukph wif vbgf rsfubwkrp rg wif xpkw bpe wif kosruwbpdf rg wif uftxkufe bdwkrp)
1) Mpvwuxofpw Cbnkcubwkrp ' =p kpvwuxofpw dbnkcubwkrp ofbpv wif belxvwofpw rg bp kpvwuxofpw vkhpbn rxwsxw vr wibw kw druufvsrpev& zkwikp bddfswbcnf ubphf bpe bddxubd & wr b mprzp ybnxf~v" rg wif sbubofwfu zikdi wif kpvwuxofpw orpkwruv) Cbnkcubwkrp vibnn fpdrosbvv wif fpwkuf kpvwuxofpw kpdnxekph bdwxbwkrp& bnbuo& ru wuks) Tfvsrpvf wkof bv vsfdkgkfe kv prw sbuw rg wif urxwkpf kpvwuxofpw dbnkcubwkrp cxw zknn cf difdmfe rpdf sfu rsfubwkph d dnf)
3) Mpvwuxofpw Cifdm ' =p kpvwuxofpw difdm kv txbnkwbwkyf efwfuokpbwkrp rg bddfswbcnf rsfubcknkw c rcvfuybwkrp rg kpvwuxofpw cfibykru exukph rsfubwkrp) Yikv efwfuokpbwkrp vibnn kpdnxef& zifuf srvvkcnf& drosbukvrp rg wif kpvwuxofpw zkwi rwifu kpefsfpefpw kpvwuxofpwv ofbvxukph wif vbof ybukbcnf)
7) Mpvwuxofpw Ixpdwkrpbn Yfvw ' =p kpvwuxofpw gxpdwkrpbn wfvw vibnn cf;
b) =pbnrh dibppfnv ' wif kplfdwkrp rg b vkhpbn kpwr wif dibppfn bv dnrvf wr wif vfpvru bv subdwkdbcnf wr yfukg rsfubcknkw kpdnxekph bnbuo bpe,ru wuks gxpdwkrpv)
c) Akvwbcnf dibppfnv ' wif kplfdwkrp rg b vkhpbn kpwr wif vfpvru wr yfukg wif rsfubcknkw kpdnxekph bnbuo bpe,ru wuks gxpdwkrpv)
:) Rgg'Vkwf Drvf Cbndxnbwkrp Obpxbn ~RDCO" ' = obpxbn drpwbkpkph wif dxuufpw ofwirernrh bpe sbubofwfuv xvfe kp wif dbndxnbwkrp rg rgg'vkwf ervfv exf wr ubekrbdwkyf hbvfrxv bpe nktxke fggnxfpwv& kp wif dbndxnbwkrp rg hbvfrxv bpe nktxke fggnxfpw orpkwrukph bnbuo,wuks vfwsrkpwv& bpe kp wif drpexdwkrp rg wif fpykurpofpwbn ubekrnrhkdbn orpkwrukph surhubo) Yif RDCO vibnn bnvr drpwbkp ~." wif Tbekrbdwkyf Eggnxfpw Crpwurnv ~kpdnxekph wif Tbekrnrhkdbn Epykurpofpwbn Orpkwrukph" Surhubo uftxkufe c Yfdipkdbn Vsfdkgkdbwkrp 7): )D& bpe ~/" efvdukswkrpv rg wif kpgruobwkrp wibw virxne cf kpdnxefe kp wif bppxbn Tbekrbdwkyf Eggnxfpw Tfnfbvf Tfsruw bpe =ppxbn Tbekrnrhkdbn Epykurpofpwbn Rsfubwkph Tfsruw uftxkufe c Yfdipkdbn Vsfdkgkdbwkrpv 7)7)D bpe 7)7)E& ufvsfdwkyfn ) Rgg'Vkwf Drvf Cbndxnbwkrp Obpxbn Vfdwkrp /)- Tfy) 0- Sbhf / rg .- _fuorpw abpmff Pxdnfbu Srzfu Vwbwkrp aCDPJ 0,/,/ "Eutznt ih&
;, Yik irntl T zeli ) Yik irntl u zeli ny zmi vixnuh uk znsi fiz iit zmi ym zhu t uk zmi tnz vxnux zu e xik irntl eth zmi yzexz v uk zmi vretz y fyiw itz zu zmez xik irntl, Lux zmi v xvuyi uk hiynltezntl kxiw itg uk ziyzntl eth y x inrretgi' e xik irntl u zeli ymerr siet e xil rexr ygmih rih xik irntl u zeliA mu i ix' mixi y gm u zeliy ugg x nzmnt ; sutzmy uk zmi gusvriznut uk zmi vxi nu y xik irntl u zeli' zmi xiw nxih y x inrretgi ziyzntl tiih tuz fi vixkuxsih tznr zmi ti z xil rexr ygmih rih u zeli,
=, _nzi Du thex ) ami ynzi fu thex ny ymu t nt Vretz Ixe ntl 7=0.):037,
/., _u xgi Emigp ) ami w ernzezn i eyyiyysitz uk gmettir xiyvutyi mit zmi gmettir yityux ny i vuyih zu e xehnuegzn i yu xgi,
//, _ x inrretgi Lxiw itg ) btriyy uzmix nyi yzezih nt zmiyi yvignkngeznuty' vixnuhng y x inrretgi ziyzy' gmigpy' gernfxeznuty' eth i esnteznuty ymerr fi vixkuxsih nzmnt zmi yvignknih y x inrretgi ntzix ery, amiyi, ntzix ery se fi eho yzih vr y 07~, ami uvixezntl g gri ntzix er ny gutynhixih zu fi /; sutzmy eth zmi zurixetgi yzezih efu i ny evvrngefri,
/0, _ x inrretgi Otzix er) ami y x inrretgi ntzix er ny zmi gerithex znsi fiz iit y x inrretgi ziyzy' gmigpy' gernfxeznuty' eth i esnteznuty zu fi vixkuxsih vut et ntyzx sitz ux gusvutitz mit nz ny xiw nxih zu fi uvixefri, amiyi ziyzy triyy uzmix nyi yzezih nt zmiyi yvignkngeznuty se fi en ih mit zmi ntyzx sitz' gusvutitz' ux y yzis ny tuz xiw nxih zu fi uvixefri' f z zmiyi ziyzy ymerr fi vixkuxsih ut zmi ntyzx sitz' gusvutitz' ux y yzis vxnux zu fintl xiw nxih zu fi uvixefri,
/1, citznreznut J me yz axiezsitz _ yzis ) ami Yeh eyzi D nrhntl eth CTM D nrhntl itznreznut NJVC knrzixy exi itznreznut i me yz zxiezsitz y yzisy mngm me i fiit hiynltih eth ntyzerrih zu xih gi xehnuegzn i sezixner nt vexzng rezi kuxs nt leyiu y ikkr itz f veyyntl itznreznut enx zmxu lm NJV C knrzixy kux zmi v xvuyi uk xisu ntl xehnuegzn i vexzng reziy kxus zmi leyiu y i me yz yzxies vxnux zu xirieyi zu zmi it nxutsitz, Jtlntiixih yekiz kiez xi ezsuyvmixng griet v y yzisy' y gm ey zmi _zethf Mey axiezsitz "_DMa& _ yzis' exi tuz gutynhixih zu fi itznreznut i me yz zxiezsitz y yzis gusvutitzy,
/3, citz-V xlntl ) citz-v xlntl ny zmi gutzxurrih vxugiyy uk hnygmexlntl enx ux ley kxus zmi vxnsex gutzentsitz zu gutzxur zisvixez xi' vxiyy xi' m snhnz ' gutgitzxeznut ux uzmix uvixezntl guthnznuty,
Tkk)_nzi Iuyi Eerg reznut Ret er _igznut 0,. Yi , 1. Veli 1 uk /. cixsutz detpii S griex Vu ix _zeznut R;=ID 0,/,0
PullYry nd TYrgY_kcs
TYrgY_kc Ccdgmgtgnm Smgts
RntYk eYllY Yatgvgty anmtYgmcb gm Y rcsgm kgmcr gm Ag stnrYec bgrcatky gm kgmc wgtf Y eYo _ctwccm YbhYacmt stnrYec lnbukcs,
+! AnmacmtrYtgnm Yt ongmt nd bgsafYrec tn Ym A- : -)kAg-lk umrcstrgatcb YrcY nd bgssnkvcb Ymb cmtrYgmcb mn_kc eYs ~g~ gm kgpugb oYtfwYys drnl Ykk stYtgnm snuracs,
: RntYk Yatgvgty nd Ykk bgssnkvcb Ymb cmtrYgmcb mn_kc eYscs gm kgpugb oYtfwYys drnl Ykk stYtgnm snuracs,
AnmacmtrYtgnm nd rYbgnmuakgbc ~g~ Yt tfc ongmt nd kgpugb bgsafYrec tn Ym umrcstrgatcb YrcY,
~ AnmacmtrYtgnm nd rYbgnmuakgbc ~g,~
~ AnmacmtrYtgnm' cxakusgvc nd mn_kc eYscs' nd rYbgnmuakgbc ~g~ drnl tYmi7 ~o~ Yt ongmt nd bgsafYrec tn Ym umrcstrgatcb YrcY, : AnmacmtrYtgnm nd rYbgnmuakgbc ~g~ gm lgxturc Yt tfc lnmgtnr,
~ =ctY bnsc tn Ygr drnl stYai rckcYsc, lrYb
~ =ctY bnsc tn Ygr drnl ernumb kcvck rckcYsc, lrYb
EYllY bnsc tn Ygr drnl stYai rckcYsc, lrYb
: EYllY bnsc tn Ygr drnl ernumb kcvck rckcYsc, lrYb
Cans Cnsc tn argtgaYk nreYm drnl stYai rckcYsc, lrcl
Cane : Cnsc tn tfc argtgaYk nreYm drnl ernumb kcvck lrcl rckcYsc,
Cgrcat bnsc "Rur_gmc =ugkbgme&, lrcl
Mdd)Pgtc Cnsc AYkaukYtgnm JYmuYk Pcatgnm 0,. Ocv, 1. NYec 3 nd /. Tcrlnmt VYmicc LuakcYr Nnwcr PtYtgnm S7:JA .)-). ~;pouioved"
Vasiabme =efioiuipo Toiut
=ec'.11n 3 ERCRE Eodiwidvam ;atl =pte! Riue :pvodasz '.11 neuest) nsen
=ec'/,,n 3 ERCRE Eodiwidvam ;atl =pte! Riue :pvodasz '/,, neuest) nsen
=ac'Spuam 3 =pte au Yetu Riue :pvodasz fspn amm catlt po uhe ERCRE Oad) nsen
Pd 3 =pte saue au / feeu fspn vopbtusvcued tide pf nsen tupsage npdvme faciog tiue bpvodasz) hs
=dA 3 =isecu dpte au tiue bpvodasz res vopbtusvcued nsen tupsage npdvme ~thpsu eod") zs'npdvme
=dR 3 =isecu dpte au tiue bpvodasz res vopbtusvcued nsen tupsage npdvme ~mpog tide") zs'npdvme
=uniue 3 Danna dpte up ais& cpssecued fps fioiue cmpvd) nsad
=Dar 3 Eouesnpdvmas gar dpte rspkecued up uhe nayinvn nsen tiue bpvodasz mpcauipo fspn setio xatue opu disecumz zs thiemded bz =7 Y npdvmet)
=np 3 =pte up uhe nayinvn psgao) nsen =R 3 =pte up tlio fspn beua aod ganna) nsen
Pt 3 =pte saue au - neues fspn tpvsce io thiemded eod pf nsen Mpsuh Yasehpvte) hs
=t 3 7oovam dpte au tiue bpvodasz fspn fiyed tpvscet io nsen thiemded eod pf Mpsuh Yasehpvte) zs
PRI= 3 Layinvn dpte saue au / feeu pwes upr pf =7Y io a nsen tupsage npdvme) hs
PRIP 3 Layinvn dpte saue au / feeu pwes upr pf each setio nsen mioes io a tupsage npdvme) hs
=RI= Rlzthioe dpte au uhe tiue bpvodasz fspn =7 Y io nsen tupsage npdvmet ~vopbtusvcued upr tvsfacet") zs' npdvme Nff'Riue =pte ;amcvmauipo Laovam Recuipo.), Pew)/, Oage 0 pf +~ Vesnpou _aolee Mvcmeas Opxes Ruauipo T;=LD 1.0.1 'Apovkowfe)
Ybtkbcmf Cfgkokvkpo Vokvu
CSJR Sl uikof epuf bv vif ukvf cpwoebt gtpn tfuko mkoftu ntfn ko uvptbhf npewmfu 'wopcuvtwdvfe vpr uwtgbdfu). t -mkoft
Cpuf vp vif vpvbm cpe . ntfn
~4! . Nm7 : Tif ektfdv epuf dpoxftukpo gbdvpt gpt N-07 udbvvft ntfn gtpn vif vwtckof ibmm vp Lpdbvkpo 'L). M_ i f
- Cpuf tbvf bv dpovbdv gtpn vif wouikfmefe vpr tbe uwtgbdf pg tfuko mkoft. it
- Cpuf bv vif ukvf cpwoebt gtpn wouikfmefe ntfn npx~Iofov')gt~ukomkoftcfvyffo vtbougft dbul boe uvptbhf npewmf.
Rw - Cpuf tbvf bv +nfvft gtpn upwtdf ko wouikfmefe foe ntfn pg Nptvi _btfipwuf. it
Cw : Tif boowbm epuf bv vif ukvf cpwoebt gtpn gkzfe ntfn upwtdfu ko vif wouikfmefe foe pg Nptvi _btfipwuf. it
CE : Ckmwvkpo gbdvpt. tbvkp
CEnko - Mkoknwn bmmpybcmf ekmwvkpo gbdvpt. tbvkp CE&d : Apnrpukvf ulko epuf gbdvpt. ntfn - ufd rAk- t
- Tpvbm cpe hbnnb epuf gbdvpt gpt owdmkef "k." ntfn-n3 rAk- t
- Apnrpukvf vpvbm cpe epuf gbdvpt. ntfn-n3 rAk- t
CELvc : Skvf-urfdkgkd, vpvbm cpe epuf gbdvpt gpt b mkswke ntfn tfmfbuf pg owdmkef "k." Ak
CELnp : Skvf-urfdkgkd, nbzknwn pthbo epuf gbdvpt gpt b ntfn mkswke tfmfbuf pg owdmkef "k." Ak
Ogg-Skvf Cpuf Abmdwmbvkpo Mbowbm Sfdvkpo 1./ Rfx. 3/ Pbhf 7 pg 0/ Yftnpov abolff Nwdmfbt Ppyft Svbvkpo S;=JD 1.0.1 'Anmshmted)
Vaphabke Cefhmhshnm Tmhsr
CEIrhcn Rhse-roechfhc, cphshcak npgam dnre facsnp fnp a rsacilpel garentr pekeare nf mtckhde "h." Ah
CEI&rhcn Rhse-roechfhc, cphshcak npgam dnre pase facsnp fnp alpel-rec rsaci garentr pekeare nf mtckhde "h." ~Ah- wp
CEIghcn Rhse-roechfhc, cphshcak npgam dnre facsnp fnp a gpntmdlpel keuek garentr pekeare nf mtckhde "h." Ah
CEI&ghcn Rhse-roechfhc, cphshcak npgam dnre pase facsnp fnp alpel-rec gpntmd keuek garentr pekeare nf mtckhde "h." ~Ah- wp
CERh =esa rihm dnre facsnp fnp mtckhde "h." lpel: l& oAh- wp
CE&hr : Anlbhmed rihm dnre facsnp fnp mtckhde "h" fpnl a lpel-rec rsaci pekeare. ~Ah- wp
CE&hg Anlbhmed rihm dnre facsnp fnp mtckhde "h" fpnl a lpel-rec gpntmd keuek pekeare. ~Ah- wp
~!, Ialla ahp dnre facsnp fnp mtckhde "h." lpad:-pm & oAh- wp
CEh : =esa ahp dnre facsnp fnp mtckhde "h." lpad: l & oAh- wp
Pcnr Aphshcak npgam dnre pase dte sn hndhmer amd lpel oapshctkaser pekeared fpnl rsaci. wp
Pcng Aphshcak npgam dnre pase dte sn hndhmer amd lpel oapshctkaser pekeared fpnl gpntmd. wp
Prihmr : Rihm dnre pase dte sn rsaci pekeare nf mnbke garer. lpel wp
Prihmg : Rihm dnre pase dte sn gpntmd pekeare nf mnbke garer. lpel wp
Nff-Rhse Cnre Aakctkashnm Lamtak Recshnm 1./ Peu. 3/ Oage 7 nf _/ Veplnms Yamiee Mtckeap Onvep Rsashnm T;=LD 0,/,0 "Aonuinved&
Yasiable Cefmiuion Vniut
Rubt : Toual bodz dote saue dve uo noble gatet fsom tuack msem seleate, zs
Rubg : Toual bodz dote saue dve uo noble gatet fsom msem gsovnd lewel seleate, zs
-~. : Cepotiuion facuos fos dsz depotiuion of elemenual sadioiodinet and ouhes pasuicvlauet,
D : Isott elecusic ovupvu owes uhe pesiod of inuesetu,
~ Esacuion of a zeas uhau a tuosage modvle it in vte fsacuion xiuh an vnobtusvcued tide osienued uoxasd xetu tiue bovndasz,
fIap Esacuion of a zeas uhau uhe inuesmodvlas gap it nou fsacuion thielded,
fSJ : Esacuion of a zeas uhau a tuosage modvle it in vte fsacuion xiuh an vnobtusvcued uop tvsface,
: Elox saue ovu of ditchasge canal, gpm
: Elox saue patu lirvid sadxatue moniuos, gpm
! : Elox saue patugateovtsadxatue moniuos, cc tec
DNI Svm of uhe fsacuiont of combined efflvenu fsacuion E+ : concenusauiont in lirvid pauhxazt "eyclvding noble gatet&,
: ;nnval awesage efflvenu concenusauion limiu fos -)lAi sadionvclide ~i~ "lOAER0., /../)0.,03./ ' cc ;ppendiy =' Table 0' Aolvmn 0&
: Releate fos sadionvclide ~i~ fsom uhe poinu of cvsiet inuesetu,
: Releate saue fos sadionvclide ~i~ au uhe poinu of inuesetu, tec
Off)Siue Cote Aalcvlauion Manval Secuion 0,. Rew, 1. Page 7 of /. Yesmonu _ankee Nvcleas Poxes Suauion S:;JC 0-/-0 &=onuhnvdc'
V_sh_ald Adehnhuhon Tnhut
~ Sgd noald f_t s_chonvblhcd "h" sdld_td s_ud _u ugd ~=h pl_nu tu_bk- tdb
7 Sgd noald f_t s_chonvblhcd "h" sdld_td s_ud esom ~=h fsovnc ldwdl- tdb
7 Sgd noald f_t s_chonvblhcd "h" sdld_td s_ud _u ugd ~=h tud_m idu _hs didbuos- tdb
7 Sgd noald f_t s_chonvblhcd "h" sdld_td s_ud _u ugd ~=h dyg_vtu oe ugd :cw_nbdc Nee,D_t Rztudm tdb
7 Sgd hochnd) ushuhvm) _nc p_suhbvl_ud s_chonvblhcd~=h "h" sdld_td s_ud esom ugd pl_nu tu_bk- tdb
7 Sgd hochnd) ushuhvm) _nc p_suhbvl_ud s_chonvblhcd~=h "h" sdld_td s_ud esom fsovnc ldwdl- tdb
7 Sgd sdld_td oe noald f_t s_chonvblhcd "h" esom ugd bvshdt pl_nu tu_bk-
7 Sgd sdld_td oe noald f_t s_chonvblhcd "h" esom bvshdt fsovnc ldwdl-
Sgd sdld_td oe hochnd) ushuhvm) _nc p_suhbvl_ud bvshdt s_chonvblhcd "h" esom ugd pl_nu tu_bk-
Sgd sdld_td oe hochnd) ushuhvm) _nc p_suhbvl_ud bvshdt s_chonvblhcd "h" esom fsovnc ldwdl-
! Jhrvhc monhuos sdtpontd eos ugd lhmhuhnf bpt +tpu bonbdnus_uhon _u ugd pohnu oe chtbg_sfd-
tkhn + tpu Pdtpontd oe ugd noald f_t monhuos _u ugd lhmhuhnf bpm tkhn cotd s_ud-
ua Rtpu Pdtpontd oe ugd noald f_t monhuos uo lhmhuhnf uou_lbpm aocz cotd s_ud-
Rghdlchnf e_buos- P_uho
Adudbuos bovnuhnf deehbhdnbz esom ugd motu sdbdnubpm mPEgs f_t monhuos b_lhas_uhon- ,,os,, ~=hIbb ~=hIbb Nee,Rhud Aotd =_lbvl_uhon L_nv_l Rdbuhon 0-. Pdw- 1. O_fd 3 oe /. Vdsmonu Y_nkdd Mvbld_s Ooxds Ru_uhon T:;JC 1-0-1 &=vu ou ki'
Yfyofgsk Aklouo ovu Vuo z
z) Ak kh vy hv u oum kllohokuh lvy uvgsk mfz "o-" hwt tRIny ,,vy,, s- =o.hh Es=o.hh
+! Ak kh vy hv u oum kllohokuh lyvt nk tvz ykhku hwz sox oi tvuo vy hfsogyf ovu- Es=o.ts
+.- Ak kh vy hv u oum kllohokuh lvy yfiovu hsoik "-o" hwz Es=o.ts
TTyfu Totk nf fu uznoksiki ykzou souky oz k wvzki ou nv yz nk z vyfmk wfi fykf-
_Dfw Eu kytvi sk mfw oi n gk kku fipfhku A: _ ouhnkz z vyfmk tvi skz nohn znoksi ykzou souky z vyfmk tvi skz lyvt nk kz zo k gv uify -
a.Pz :uu fs vy svum, kyt f kyfmk uikwsk ki zkh ~ f tvzwnkyoh iozwkyzovu lfh vy lvy z fhr ykskfzk- 3 t
a.Pm :uu fs vy svum, kyt f kyfmk uikwsk ki zkh f tvzwnkyoh iozwkyzovu lfh vy lvy myv ui sk ks 3 t ykskfzk-
ca.Pe7 Cllkh o k fuu fs vy svum, kyt f kyfmk mfttf zkh f tvzwnkyoh iozwkyzovu lfh vy- ~ t 3
ca.Pe~ Cllkh o k fuu fs vy svum, kyt f kyfmk mfttf zkh f tvzwnkyoh iozwkyzovu lfh vy lvy f myv ui sk ks ~ t3 ykskfzk-
Nll,So k Avzk =fsh sf ovu Lfu fs Skh ovu 1-/ Rk - 3/ Ofmk d/ vl d/ Ykytvu bfurkk M hskfy Ov ky S f ovu 0,1)- EIIM_EOY =OD EOaLTPONEOY=M CPOYTPMV
Ylmw whfxmsr mrfoyghw xlh hiioyhrx drg hrzmvsrphrxdo fsrxvsow xldx hvh svmkmrdoo tdvx si xlh ahvpsrx cdrnhh Yhflrmfdo Vthfmimfdxmsrw) Ylhwh fsrxvsow hvh vhosfdxhg mrxs xlh PDCN mxlsyx dr wyewxdrxmdo fldrkhw& mr dffsvgdrfh mxl OTC Jhrhvmf Mhxxhv 7:'-.) Yh x drg xdeohw hvh vhisvpdxxhg xs xlh wx oh si xlh PDCN) Ylh zdvmsyw fsrxvsow hvh vhrypehvhg ivsp xlh svmkmrdo rypehvmrk wflhph si xlh Yhflrmfdo Vthfmimfdxmsrw) = fvsww'vhihvhrfh si xlh sog Yhflrmfdo Vthfmimfdxmsrw whfxmsr xs xlh rh PDCN whfxmsr mw tvhwhrxhg mr Ydeoh .).)7)
0,1). LOVYT_NEOY =YLPO
0,1).). Tdgmsdfxmzh Mmuymg Eiioyhrx Lrwxvyphrxdxmsr
CPOYTPMV
0).). Ylh vdgmsdfxmzh omuymg hiioyhrx psrmxsvmrk mrwxvyphrxdxmsr fldrrho wldoo eh sthvdeoh mr dffsvgdrfh mxl Csrxvso Ydeoh 0).). mxl xlhmv dodvp whxtsmrxw whx xs hrwyvh xldx xlh ompmxw si Csrxvso 0)/). dvh rsx h fhhghg)
=RRMLC=ALMLYc;
Dyvmrk thvmsgw si vhohdwh xlvsykl psrmxsvhg tdxl d w dw omwxhg sr Ydeoh 0).).)
=CYLPO;
d) bmxl d vdgmsdfxmzh omuymg hiioyhrx psrmxsvmrk mrwxvyphrxdxmsr fldrrho dodvp,xvmt whxtsmrx ohww fsrwhvzdxmzh xldr d zdoyh lmfl moo hrwyvh xldx xlh ompmxw si Csrxvso 0)/). dvh phx& mxlsyx ghod wywthrg xlh vhohdwh si vdgmsdfxmzh omuymg hiioyhrxw psrmxsvhg e xlh diihfxhg fldrrho sv fldrkh xlh whxtsmrx ws xldx mx mw dffhtxdeo fsrwhvzdxmzh sv ghfodvh xlh fldrrho mrsthvdeoh)
e) bmxl srh sv psvh vdgmsdfxmzh omuymg hiioyhrx psrmxsvmrk mrwxvyphrxdxmsr fldrrhow mrsthvdeoh& xdnh xlh dfxmsr wls r mr Ydeoh 0).).)
V_TaELMM=OCE TES_LTENEOYV
1).).)d Edfl vdgmsdfxmzh omuymg hiioyhrx psrmxsvmrk mrwxvyphrxdxmsr fldrrho wldoo eh xhwxhg drg fdomevdxhg dw mrgmfdxhg mr Ydeoh 1).).)
1).).)e Ylh whxtsmrxw isv psrmxsvmrk mrwxvyphrxdxmsr wldoo eh ghxhvpmrhg mr dffsvgdrfh mxl xlh PDCN ~Vhfxmsr 7).")
Pii'Vmxh Dswh Cdofyodxmsr Ndrydo Vhfxmsr 0,1 Thz) 0- Rdkh L si 13 ahvpsrx cdrnhh Oyfohdv Rs hv Vxdxmsr YACNI 1,/,/
Nltxlf Ihhnxgpw Orplwrulpi Mpvwuxogpwcwlrp
Olploxo Dkcppgnv Rsgucdng Prwgv
/, Lurvv Tcflrcewlylw Orplwruv prw Surylflpi Axwrocwle Yguolpcwlrp rh Tgngcvg
c, Nltxlf Tcfzcvwg Elvekcuig Orplwru /& /'3 ~TO)/;)17." d, Vguyleg acwgu Elvekcuig Orplwru / 0'3 ~TO )/;)17/" 0, Jnrz Tcwg Ogcvxugogpw Egylegv c, Nltxlf Tcfzcvwg Elvekcuig Jnrz Tcwg /& 1'3 Orplwru ~JMY )0.)3=7-330"
& Exulpi ugngcvgv ylc wklv scwkzc
Rhh)Vlwg Ervg Dcnexncwlrp Ocpxcn Vgewlrp 1-3 Tgy, 1. Scig 0 rh 3: _guorpw bcpmgg Pxengcu Srzgu Vwcwlrp R:;IC 0).). LMR:REML
LMRC . ' Tgsf sfc ltk_cp md afYllcir mncpY_ic icrr sfYl pcotgpcb _x sfc kglgktk afYllcir mncpY_ic pcotgpckcls& cdditcls pcicYrcr kYx amlsgltc npmugbcb sfYs npgmp sm glgsgYsgle Y pcicYrc7
Y) :s icYrs svm glbcnclbcls rYknicr Ypc YlYixycb gl YaampbYlac vgsf =mlspmi 1)/).& Ylb
_) :s icYrs svm scaflgaYiix otYigdgcb kck_cpr md sfc DYagigsx PsYdd glbcnclbclsix ucpgdx sfc pcicYrc pYsc aYiatiYsgmlr Ylb bgrafYpec iglc uYiugle)
Msfcpvgrc& rtrnclb pcicYrc md pYbgmYasguc cdditclsr ugY sfgr nYsfvYx)
LMRC / ' Tgsf sfc ltk_cp md afYllcir mncpY_ic icrr sfYl pcotgpcb _x sfc kglgktk afYllcir mncpY_ic pcotgpckcls& cdditcls pcicYrcr ugY sfgr nYsfvYx kYx amlsgltc npmugbcb sfYs& Ys icYrs mlac ncp /1 fmtpr& epY_ rYknicr Ypc amiicascb Ylb YlYixycb dmp epmrr pYbgmYasgugsx ~_csY mp eYkkY" Ys Y imvcp igkgs md bcscasgml md Ys icYrs .-'~kgapmatpgc,ki)
LMRC 0 ' Tgsf sfc ltk_cp md afYllcir mncpY_ic icrr sfYl pcotgpcb _x sfc kglgktk afYllcir mncpY_ic pcotgpckcls& cdditcls pcicYrcr ugY sfgr nYsfvYx kYx amlsgltc npmugbcb sfc dimv pYsc gr crsgkYscb Ys icYrs mlac ncp 1 fmtpr btpgle YastYi pcicYrcr) Ntkn ncpdmpkYlac atpucr kYx _c trcb sm crsgkYsc dimv)
LMRC 1 ' Tgsf sfc ltk_cp md afYllcir mncpY_ic icrr sfYl pcotgpcb _x sfc kglgktk afYllcir mncpY_ic pcotgpckcls& cwcps pcYrmlY_ic cddmpsr sm pcstpl sfc glrsptkcls~r" sm mncpY_ic rsYstr npgmp sm sfc lcws pcicYrc)
Mdd'Pgsc Amrc =YiatiYsgml JYltYi Pcasgml 0,1 Ocu) 0- NYec 0 md 13 Scpkmls VYlhcc LtaicYp Nmvcp PsYsgml T;=LD 3,/,/
Radioacuiwe Lirvid Dfflvenu Moniuosing Jntusvmenuauion Svsweillance Rervisemenut
Jntusvmenu Jntusvmen Sovsce Jntusvmenu Evncuional Jntusvmenu u Aheck Aheck Aalibsauion Tetu /, Isott Radioacuiwiuy Moniuost nou Psowiding ;vuomauic Tesminauion of Releate
a, Lirvid Radxatue Once each Psios uo each Once each Once each Citchasge Moniuos ~1" day& seleate' bvu /: monuht ~/" rvasues ~0" no mose uhan once each monuh
b, Seswice Yaues Once each Once each Once each Once each Citchasge Moniuos ~1" day monuh /: monuht ~/" rvasues ~0" 0, Elox Raue Meatvsemenu Cewicet
a, Lirvid Radxatue Once each Nou Nou Once each Citchasge Elox Raue day& ;pplicable ;pplicable rvasues& Moniuos
~ Cvsing seleatet wia uhit pauhxay,
Off)Siue Cote Aalcvlauion Manval Secuion 1-3 Rew, 1. Page 3 of 37 Vesmonu _ankee Nvcleas Poxes Suauion S;=JD 0'-'- MNS;SINM
~-" Sfc Imstrulcmt AYkg_rYtgnm dnr rYbgnYatgvgty lcYsurclcmt gmstrulcmtYtgnm sfYkk gmakubc tfc usc nd Y imnwm ~trYacY_kc tn MYtgnmYk Imstgtutc dnr RtYmbYrbs Ymb Scafmnkney" kgpugb rYbgnYatgvc snurac onsgtgnmcb gm Y rcornbuag_kc ecnlctry wgtf rcsocat tn tfc scmsnr' Sfcsc stYmbYrbs sfYkk ocrlgt aYkg_rYtgme tfc systcl nvcr gts mnrlYk nocrYtgme rYmec nd cmcrey Ymb rYtc'
~." Sfc Imstrulcmt EumatgnmYk Scst sfYkk Yksn bclnmstrYtc tfc Anmtrnk Pnnl YkYrl YmmumagYtgnm naaurs gd Ymy nd tfc dnkknwgme anmbgtgnms cxgsts:
~Y" Imstrulcmt gmbgaYtc lcYsurcb kcvcks Y_nvc tfc YkYrl sctongmt'
~_" Agraugt dYgkurc'
~a" Imstrulcmt gmbgaYtcs Y bnwmsaYkc dYgkurc'
~b" Imstrulcmt anmtrnks mnt sct gm nocrYtc lnbc'
~/" Sfc YkYrl sctongmts nd tfcsc afYmmcks sfYkk _c bctcrlgmcb Ymb Ybhustcb gm YaanrbYmac wgtf tfc lctfnbnkney Ymb oYrYlctcrs gm tfc Ndd&Rgtc Cnsc AYkaukYtgnm LYmuYk ~scc Rcatgnm 7'-"'
Ndd&Rgtc Cnsc AYkaukYtgnm LYmuYk Rcatgnm /)0 Pcv' /, OYec 1 nd 03 Tcrlnmt VYmicc MuakcYr Onwcr RtYtgnm 0,1). JOVYT_NEOY=YJPO
0,1).)/ Tdgmsdfwmyh Idvhsxv Eiioxhrw Jrvwuxphrwdwmsr
CPOYTPMV
0).)/ Ylh kdvhsxv tusfhvv drg hiioxhrw psrmwsumrk mrvwuxphrwdwmsr fldrrhov vldoo eh sthudeoh mr dffsugdrfh zmwl Csrwuso Ydeoh 0).)/ zmwl wlhmu dodup,wumt vhwtsmrwv vhw ws hrvxuh wldw wlh ompmwv si Csrwusov 0)0).)d& drg Yhflrmfdo Vthfmimfdwmsrv 0):).). drg 0):)L). ~Csrwuso 0)0)7" duh rsw h fhhghg)
=RRMJC=AJMJYc;
=v vlszr mr Ydeoh 0).)/)
=CYJPO;
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RSYL 1 ' alxk xkg rypdgv sh ekcrrgow stgvcdog ogww xkcr vguylvgf d xkg plrlpyp ekcrrgow stgvcdog vguylvgpgrx& ghhoygrx vgogcwgw zlc xkg chhgexgf tcxk c pc esrxlryg tvszlfgf wcptogw cvg esrxlrysywo esoogexgf lxk cy lolcv wcptolri guyltpgrx)
RSYL 3 ' alxk xkg rypdgv sh ekcrrgow stgvcdog ogww xkcr vguylvgf d xkg plrlpyp ekcrrgow stgvcdog vguylvgpgrx& g gvx vgcwsrcdog ghhsvxw xs vgxyvr xkg lrwxvypgrx~w" xs stgvcdog wxcxyw lxklr 0- fc w)
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RSYL = ' alxk xkg rypdgv sh ekcrrgow stgvcdog ogww xkcr vguylvgf d xkg plrlpyp ekcrrgow stgvcdog vguylvgpgrx& icwgw hvsp xkg VNDL pc dg vgogcwgf xs xkg grzlvsrpgrx hsv yt xs :/ ksyvw tvszlfgfA
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YaVbIMNNAPDI VITaMVIOIP_Y
0)/).)e _mi htxi weyi hzi yt stfpi lexix ns lexitzx ikkpzisyx xmepp fi hiyiwrnsih yt fi nymns ymi pnrnyx tk Dtsywtp /)/). ns eggtwhesgi nym ymi riymthx ns ymi REDO)
0)/).)f _mi htxi weyi hzi yt Mthnsi'./.& Mthnsi'.//& ywnynzr esh wehntszgpnhix ns uewyngzpeyi ktwr nym mepk'pn ix lwieyiw ymes : he x ns lexitzx ikkpzisyx xmepp fi hiyiwrnsih yt fi nymns ymi pnrnyx tk Dtsywtp /)/). ns eggtwhesgi nym ymi riymthx ns ymi REDO f tfyensnsl wiuwixisyeyn i xerupix esh uiwktwrnsl esep xix ns eggtwhesgi nym ymi xerupnsl esh esep xnx uwtlwer xuignknih ns _efpi 0)/).)
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Vehnuegzn i Jeyiu y beyzi Yesvrntl =th =ter yny Txulxes
Ou ix Onsnz Pntns s _ vi uk uk Dizigznut Jeyiu y Virieyi Yesvrntl =ter yny =gzn nz ~OOD" _ vi Ixiw itg Ixiw itg =ter yny ~ Cnrsr" ~:! =) Yzies Miz Stgi vix iip Stgi vix iip ci'./:& . rS= =nx Eoigzux Jxef Yesvri ci'./1& ci'.//& Nx'::& Nx':7& Nx':1P A) Cutzentsitz Txnux zu iegm Txnux zu iegm Txntgnver . .-'8 ~_! T xli xirieyi, xirieyi, Eegm Jesse Esnzzixy ~=-_! Eegm T xli Jxef T xli Yesvri kux eth .'./. Texzng reziy C) Pent Tretz Stgi vix sutzm Stgi vix sutzm Txntgnver . rS= ~I! Yzegp ~I! Jxef Yesvri Jesse Esnzzixy Ch" L'/ . .-'3 .4 Cutznt u y ~I! Stgi vix iip .'./. ~\! . .-' ~f" Cmexguer Yesvri .. Cutznt u y ~I! Stgi vix iip Txntgnver . .-' ~f" Texzng rezi Jesse Yesvri Esnzzixy ~=+_! eth .'./. .. Cutznt u y ~I! Stgi vix sutzm Jxuyy =rvme . .-' Cusvuynzi Texzng rezi Yesvri
.. Cutznt u y ~I! Stgi vix w exzix Yx':;& Yx';- . .-' Cusvuynzi Texzng rezi Yesvri Cutznt u y Rufri Jey Rufri Jeyiy . .-'1 Putnzux Jxuyy Aize ux Jesse
Skk'Ynzi Duyi Cerg reznut Pet er Yigznut /,0 Vi ) /- Teli .: uk 03 aixsutz detpii R griex Tu ix Yzeznut cIJSN 7-3-0 VYcIcPYV
i- bnn oxx wx n i- xo cikun 7-1-0-
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l- bivyuswp iwm iwiu n riuu iu x kn yn ox vnm oxuux swp r mx w) i y) x i rn viu yx n lriwpn n lnnmswp 1:" xo i nm rn viu yx n yn rx wun D &0' iwiu s rx ri rn mx n nz s iunw 0,030 lxwlnw i sxw sw rn y svi lxxuiw ri wx swl ni nm vx n riw i oil x xo 3 iwm rn n u iw lxwlnw i sxw s i uni 0 0/,0 .uLs.vuE iwm &1' rn wxkun pi vxws x rx ri noou nw il s s ri wx swl ni nm vx n riw i oil x xo 3-
m- crn y swlsyiu pivvi nvs n ox rslr rn SSM ynlsosli sxw suu iyyu i n n lu s nu rn oxuux swp imsxw lusmn D R ,A=) R ,AA) fn,033) fn,033v) fn,03: iwm fn,03A ox pi nx nvs sxw ) iwm Tw,:7) On,:C) Lx,:A) Lx,;/) hw,;:) Tx,CC) L ,037) L ,03=) Ln,070 iwm Ln,u77 ox yi sl ui n nvs sxw - crs us mxn wx vniw ri xwu rn n w lusmn i n x kn mn nl nm iwm nyx nm- Y rn ynit rslr i n vni ikun iwm smnw sosikun) xpn rn s r rn ikx n w lusmn ) riuu iu x kn smnw sosnm iwm nyx nm- V lusmn rslr i n knux SSM ox rn iwiu n rx um wx kn nyx nm i knswp y n nw i rn SSM un nu ox ri w lusmn) k i ~wx mn nl nm-~ ernw w iu ls l v iwln n u sw SSM rsprn riw nz s nm) rn ni xw riuu kn mxl vnw nm sw rn aimsxil s n Noou nw anuni n anyx -
n- crn i sx xo rn ivyun oux i n x rn ivyunm niv oux i n riuu kn twx w ox rn svn yn sxm lx n nm k nilr mx n x mx n i n liul ui sxw vimn sw illx miwln s r Lxw xu 3-3-0) 3-3-1) iwm 3-3-3-
o- crn pi nx i n ivyuswp iwm iwiu s y xp iv mxn wx n yusls u nz s n ivyuswp iwm iwiu s i i ynlsosnm SSM x mn n vswn rn 0,033 nuni n- N svi n xoP,033 nuni n riuu kn mn n vswnm k lx w swp rn nntu lri lxiu ivyun ox 0,033 &i nuu i 0,030' iwm i vn i lxw iw nuni n i n ox rn nuni n yn sxm- p- Sx n Ssvs xo Mn nl sxw &SSM' iyyusn xwu x yi sl ui n ox v imsxw lusmn smnw sosnm sw cikun Vx i sxw m- ikx n-
Yoo,bs n Mx n Liul ui sxw Tiw iu bnl sxw3.7 an - 3/ _ipn 0C xo 7; dn vxw giwtnn V luni _x n b i sxw 0,1)0 S;CJO;AVJ_D I;TDOYT DEELYDNVT
0,1)0)/ Crvg ' Nrdng Icvgv
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0)0)/ Vkg clu frvg fxg wr prdng icvgv ugngcvgf lp icvgrxv ghhnxgpwv huro wkg vlwg wr cugcv cw cpf dg rpf wkg vlwg drxpfcu vkcnn dg nlolwgf wr wkg hrnnrzlpi:
c) Cxulpi cp ecngpfcu txcuwgu:
ngvv wkcp ru gtxcn wr 3 oucf hru icooc ucflcwlrp& cpf ngvv wkcp ru gtxcn wr .- oucf hru dgwc ucflcwlrp& cpf
d) Cxulpi cp ecngpfcu gcu:
ngvv wkcp ru gtxcn wr .- oucf hru icooc ucflcwlrp& cpf ngvv wkcp ru gtxcn wr /- oucf hru dgwc ucflcwlrp)
;PPLJA;=JLJVb:
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TYS_DJLL;NAD SDRYJSDMDNVT
1)0)/ Axoxncwlyg frvg erpwuldxwlrpv hru wkg wrwcn wlog sgulrf vkcnn dg fgwguolpgf lp ceerufcpeg zlwk wkg ogwkrfv lp wkg OCAM cw ngcvw rpeg gygu orpwk)
Ohh'Tlwg Crvg Acnexncwlrp Mcpxcn Tgewlrp 0,1 Sgy) 0- Pcig /- rh 17 _guorpw bcpmgg Nxengcu Przgu Twcwlrp 0,1)0 VAEMRAD_MbI LAYIRaY IJJNaIP_Y
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DRP_VRNY
0)0)0 _nk iuyk zu f sksgkx ul znk v groh lxus Muiotk'.0.& Muiotk'.00& zxozo s& fti xfiout hroiky ot vfxzoh rfzk luxs ozn nfrl'ro ky mxkfzkx znft ; if y ot mfyku y kllr ktzy xkrkfyki lxus znk yozk zu fxkfy fz fti gk uti znk yozk gu tifx ynfrr gk rosozki zu znk lurru otm=
f) E xotm ft hfrktifx w fxzkx=
rkyy znft ux kw fr zu :)3 sxks zu ft uxmft& fti
g) E xotm ft hfrktifx kfx=
rkyy znft ux kw fr zu .3sxks zu ft uxmft)
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YaVbIMNNAPDI VITaMVIOIP_Y
1)0)0 D s rfzo k iuyk hutzxog zouty lux znk zuzfr zosk vkxoui ynfrr gk ikzkxsotki ot fhhuxifthk ozn znk skznuiy ot znk REDO fz rkfyz uthk k kx sutzn)
Rll'Yozk Euyk Dfrh rfzout Oft fr Ykhzout 0,1 Vk ) 0- Sfmk /. ul 17 bkxsutz dftpkk P hrkfx Su kx Yzfzout 0,1)0 V=DLR=C_LbE J=YERaY EIINaEP_Y
0,1)0)1 Jexiuzx Veh exyi _wieysity
CRP_VRNY
0)0)1 _mi =zlsityih Rkk'Jex Y xyis ~=RJ" xmerr fi zxih nt nyx hixnltih suhi uk uviweynut yu wihzgi tufri lexix nt lexiuzx exyi vwnuw yu yminw hnxgmewli miti iw ymi sent guthitxiw xyies oiy enw ioigyuw ~YM=E" nx nt uviweynut)
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cnym lexiuzx weh exyi kwus ymi sent guthitxiw enw ioigyuw x xyis fintl hnxgmewlih nymuzy ywieysity kuw suwi ymet 7 he x& vwivewi eth xzfsny yu ymi Cussnxxnut nymnt 0- he x& e Yvigner Vivuwy ymey ntgrzhix ymi ntkuwseynut hiyenrih nt RDCO Yigynut .-)0)/)
YaVbELNN=PCE VETaLVEOEP_Y
1)0)1)e _mi wiehntlx uk ymi wiri ety ntxywzsity xmerr fi gmigpih i iw ./ muzwx mit ymi sent guthitxiw YM=E nx nt zxi yu itxzwi ymey ymi =RJ nx kztgynutntl)
1)0)1)f _mi lexiuzx ikkrzity ywieysity x xyis xgmiseyng nx xmu t nt RDCO Inlzwi :)/)
Rkk'Ynyi Duxi Cergzreynut Oetzer Yigynut 0,1 Vi ) 0- Seli // uk 13 biwsuty detpii Pzgriew Su iw Yyeynut 0,1)0 T;CLP;AYLaD I;VDP_V DEEM_DOYV
0,1)0)3 ahsxmpdxmts D ldywx Yvhdxrhsx
APOYTPMV
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V_TaDLMM;OAD TDS_LTDNDOYV
1)0)3 Vhh Atsxvtp 1)0)/ itv wyvzhmppdsfh vhpdxhg xt ;PI dsg Tdg dwxh =ympgmsk zhsxmpdxmts impxhv w wxhr tuhvdxmts)
Pii'Vmxh Ctwh Adpfypdxmts Ndsydp Vhfxmts 0,1 Thz) 0- Rdkh /0 ti 17 ahvrtsx cdsohh Oyfphdv Rt hv Vxdxmts 0,1)0 S;CJO;AVJ_D I;TDOYT DEELYDNVT
0,1)0)3 Pulocu Arpwclpogpw
AONVSOLT
0)0)3 akgp wkg sulocu erpwclpogpw lv wr dg _gpwgfnPxuigf& lw vkcnn dg _gpwgfnPxuigf wkurxik wkg Twcpfd Icv Vugcwogpw T vwgo zkgpgygu wkg cludrupg ucflrcewlylw ngygnv lp erpwclpogpw rh Jrflpg'.0.& Jrflpg'.00 ru ucflrpxenlfgv lp scuwlexncwg hruo zlwk kcnh'nlygv iugcwgu wkcp 7 fc v g eggf wkg ngygnv vsgelhlgf lp ;ssgpfl = wr nOAES/-).--. ' /-)/1-/& Vcdng .& Arnxop 0)
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c) alwk wkg ugtxlugogpwv rh Arpwurn 0)0)3 prw vcwlvhlgf& loogflcwgn vxvsgpf cnn _gpwlpi,Pxuilpi rh wkg erpwclpogpw)
d) Cxulpi pruocn ughxgnlpi cpf oclpwgpcpeg rxwcigv zkgp sulocu erpwclpogpw lv pr nrpigu ugtxlugf& wkgp Arpwurn 0)0)0 vkcnn vxsguvgfg Arpwurn 0)0)3)
TYS_DJLL;NAD SDRYJSDMDNVT
1)0)3 Vkg sulocu erpwclpogpw vkcnn dg vcosngf sulru wr ygpwlpi,sxuilpi sgu Vcdng 1)0).& cpf lh wkg ugvxnwv lpflecwg ucflrcewlylw ngygnv lp g egvv rh wkg nlolwv rh Arpwurn 0)0)3& wkg erpwclpogpw vkcnn dg cnlipgf hru ygpwlpi,sxuilpi wkurxik wkg Twcpfd Icv Vugcwogpw T vwgo) Nr vcosnlpi vkcnn dg ugtxlugf lh wkg ygpwlpi,sxuilpi lv wkurxik wkg Twcpfd Icv Vugcwogpw ~T=IV" T vwgo)
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1,3,/ ank iuyk ux iuyk hussozsktz zu f sksgkx ul znk v grohC~ ot fxkfy fz fti gk uti znk _ozk Eu tifx lxus frr yzfzout yu xhky oy rosozki zu rkyy znft ux kw fr zu 07 sxks zu znk zuzfr gui ux ft uxmft "k hkvz znk zn xuoi' nohn oy rosozki zu rkyy znft ux kw fr zu ;7 sxks& u kx f hfrktifx kfx,
DTTONIDENONaeA
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3,3,/,f I s rfzo k iuyk hutzxog zouty lxus row oi fti mfyku y kllr ktzy ynfrr gk ikzkxsotki ot fhhuxifthk ozn Iutzxury 3,0,0' 3,1,0' fti 3,1,1,
3,3,r,g I s rfzo k iuyk hutzxog zouty lxus ioxkhz xfiofzout lxus vrftz yu xhky ynfrr gk ikzkxsotki ot fhhuxifthk ozn znk skznuiy ot znk SJIP, anoy xkw oxksktz oy fvvrohfgrk utr tikx hutiozouty ykz luxzn ot Iutzxur 1,3,/ Dhzout _zfzksktz,
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0,1)3). Epylurpogpwcn Mrplwrulpi Puriuco
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TYS_EJLL=NCE SERYJSEMENVT
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Sadipmpgicam Eoxitponeovam Npoivptiog Rtpgtan
Ezrpuwte Ravhya Ownbet pf Tanrme Tanrmiog aod V re aod Itesweoc aod,pt Tanrme Mpcavipou~-! Cpmmecvipo pf =oam uiu Itesweoc
m) =LSAPSOE
a) Sadipipdioe aod Tanrmeu ftpn Cpoviowpwu Sadipipdioe caoiuvet; Ratvicwmaveu 3 mpcavipou; pretavipo pf =oam e each uanrme fpt uanrmet yivh L'.0m) . uanrme ftpn wr uanrme cpmmecvipoRatvicwmave uanrmet; xamme & yivhio 1 nimeuyeelm pt npte Jtpuu beva tadipacvixiv pf Tive mmpwodat ) ftesweovm au aoam uiu po each uanrme ~nakpt yiod ditecvipo" teswited b dwuv fpmmpyiog fimvet chaoge) ~c" ) ) mpadiog) Cpnrpuive ~b . uanrme ftpn dpyo mpcavipo" fpt ganna xamme & yivhio 1 nimeu ) ~d" + LTPvPRLC av eauv poce pf Tive Apwodat ) ret swatvet) ~nakpt yiod ditecvipo"
. uanrme each ftpn vhe xicioiv pf vyp oeatb cpnnwoivieu& yivhio .- nimeu pf Tive Apwodat )
. uanrme ftpn a cpovtpm mpcavipo)
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xlh edodrfh si xlh wxdxmsrw xs eh todfhg mr wthfmdo mrxhvhwx dvhdw drg fsrxvso wxdxmsr dvhdw,
Sii)_mxh Iswh Edofyodxmsr Pdrydo _hfxmsr 1-3 Yhz, 1. Tdkh 0; si 3: bhvpsrx cdrnhh Ryfohdv Ts hv _xdxmsr T:;JC /)1). "=onu~d&
Radiological Cnwisonmenual Loniuosing Osogsam
Cypotvse Oauhxaz Mvmbes of Sample Sampling and Tzpe and Dservencz ~-! and,os Sample Jocauiont =ollecuion Dservencz of :nalztit /) Y:TCR;NRMC
~4! a) Svsface . tample vptuseam) Lonuhlz gsab tample) Eamma itouopic ~.! analztit of each tample) Tsiuivm analztit of compotiue tample au leatu once pes rvasues) . tample =ompotiue tample doxntuseam) collecued owes a soesiod of one monuh
E)) ~.! b) Esovnd . tample fsom xiuhin Pvasueslz) amma ISNuNOi= 7 +-8 dituance) and usiuivm analztet of each tample) . tample fsom a Pvasueslz) conusol locauion)
c) Sedimenu fsom . tample fsom Semiannvallz) Eamma itouopic ~.! Shoseline doxntuseam asea xiuh analztit of each eyituing os pouenuial tample) secseauional walve) . tample fsom nosuh Semiannvallz) tuosm dsain ovufall)
Nff'Siue Aote =alcvlauion Lanval Secuion /,0 Rew) /- Oage /- of 03 Vesmonu _ankee Mvcleas Ooxes Suauion V;=LD /)1). "Aonu~d&
Sadiological Dnwisonmenual Moniuosing Psogsam
Dypotvse Pauhxaz Nvmbes of Tample Tampling and Vzpe and Eservencz ~.! and,os Tample Locauiont Aollecuion Eservencz of ;nalztit 0) JNIDTVJON R a) Milk Tamplet fsom milking Temimonuhlz if milking Iamma itouopic animalt in / locauiont animalt ase idenuified and .'./. analztit of xiuhin 1 km dituance on patuvse: au leatu once each tample) hawing uhe highetu dote pes monuh au ouhes pouenuial) Jf uhese ase uimet) lett uhan / psimasz locauiont awailable uhen +os mose tecondasz tample fsom milking animalt in each of / aseat beuxeen 1 uo 7 km dituance xhese dotet ase calcvlaued uo be gseaues uhan . msem pes zeas) . tample fsom milking animalt in a conusol locauion) b) Eith . tample of uxo Temiannvallz) Iamma itouopic R secseauionallz imposuanu analztit on edible tpeciet in wiciniuz of posuiont) planu ditchasge asea) . tample "psefesablz of tame tpeciet& in aseat nou inflvenced bz planu ditchasge) c) Yegeuauion . gsatt tample au each Rvasueslz xhen Iamma itouopic R ais tampling tuauion) awailable) analztit of each tample) . tilage tample au each ;u uime of haswetu) Iamma itouopic R milk tampling tuauion analztit of each "at awailable&) tample) V;=LD /)1). NOV;VJON Off'Tiue Cote Aalcvlauion Manval Tecuion /,0 Sew) /- Page -+ of 03 Yesmonu _ankee Nvcleas Poxes Tuauion a Tpecific pasameuest of dituance and disecuion tecuos fsom uhe cenuesline of uhe seacuos and addiuional detcsipuiont xhese pesuinenu' thall be psowided fos each and ewesz tample locauion in Vable 1,7,/ in a uable and figvse"t& in uhe PDCN "Tecuion ;&, Dewiauiont ase pesmiuued fsom uhe servised tampling tchedvle if tpeciment ase vnobuainable dve uo ha asdovt condiuiont' teatonal vnawailabiliuz' malfvncuion of avuomauic tampling ervipmenu and ouhes legiuimaue seatont, Lf tpeciment ase vnobuainable dve uo tampling ervipmenu malfvncuion' ewesz seatonable effosu thall be made uo compleue cossecuiwe acuion psios uo uhe end of uhe neyu tampling pesiod, All dewiauiont fsom uhe tampling tchedvle thall be docvmenued in uhe Annval Sadiological Enwisonmenual Ppesauing Seposu pvstvanu uo PDCN Tecuion /.,0, Lu it secogni ed uhau' au uimet' iu maz nou be pottible os psacuicable uo conuinve uo obuain tamplet of uhe media of choice au uhe motu detised locauion os uime, Ln uhete intuancet' tviuable aluesnauiwe media and locauiont maz be choten fos uhe pasuicvlas pauhxaz in rvetuion and appsopsiaue tvbtuiuvuiont made xiuhin 1. dazt in uhe sadiological enwisonmenual moniuosing psogsam, Ln liev of a Micentee Ewenu Seposu and pvstvanu uo PDCN Tecuion /.,/' idenuifz uhe cavte of uhe vnawailabiliuz of tamplet fos uhau pauhxaz and idenuifz uhe nex locauion"t& fos obuaining seplacemenu tamplet in uhe neyu Sadioacuiwe Efflvenu Seleate Seposu and alto inclvde in uhe seposu a sewited figvse"t& and uable fos uhe PDCN seflecuing uhe nex locauion"t&,
b Pne os mose intusvmenut' tvch at a psettvsi ed ion chambes' fos meatvsing and secosding dote saue conuinvovtlz maz be vted in place of' os in addiuion uo' inuegsauing dotimeuest, Ios uhe pvspotet of uhit uable' a Vhesmolvminetcenu Dotimeues "VMD& it contidesed uo be one photphos= uxo os mose photphost in a packeu ase contidesed at uxo os mose dotimeuest, Iilm badget thall nou be vted at dotimeuest fos meatvsing disecu sadiauion, Vhe 3. tuauiont it nou an abtolvue nvmbes, Vhe fservencz of analztit os seadovu fos VMD tztuemt xill depend vpon uhe chasacuesituict of uhe tpecific tztuem vted and thovld be telecued uo obuain opuimvm dote infosmauion xiuh minimal fading,
c Aisbosne pasuicvlaue tample filuest thall be analz ed fos gsott beua sadioacuiwiuz 03 hovst os mose afues tampling uo allox fos sadon and uhosivm davghues decaz, Lf gsott beua acuiwiuz in ais pasuicvlaue tamplet it gseaues uhan uen uimet uhe zeaslz mean of conusol tamplet' gamma itouopic analztit thall be pesfosmed on uhe indiwidval tamplet,
d Jamma itouopic analztit meant uhe idenuificauion and rvanuificauion of gamma)emiuuing sadionvclidet uhau maz be auusibvuable uo uhe efflvenut fsom uhe faciliuz,
e Vhe ~vptuseam tample~ thall be uaken au a dituance bezond tignificanu inflvence of uhe ditchasge, Vhe ~doxntuseam~ tample thall be uaken in an asea bezond bvu neas uhe miying one,
f Compotiue tample alirvout thall be collecued au uime inueswalt uhau ase wesz thosu selauiwe uo uhe compotiuing pesiod in osdes uo attvse obuaining a sepsetenuauiwe tample,
g Each meueosological tecuos thall hawe an etuablithed ~innes~ and an ~ovues~ moniuosing locauion bated on eate ofsecowesz "i,e,' setponte uime& and zeas)sovnd accettibiliuz,
h Deleued,
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cop Ewv zwt xzw pkl zlg wvghtl g zgvil og zltlg l mzwu iwv gpvulv x znpvn- lv pvn wxlzg pwv ptt hl mpt lzlk ozw no ol b gvkh Mg czlg ulv b lu "bDMc& w og ol gvv gt kw l tpup wm /.ELa _gz 0. mwz Tluhlz wm ol _ htpi pv gzlg g gvk hl wvk ol bp l Dw vkgz ptt vw hl l illklk, col kw l whrli p l wm Ewv zwt 1,1,1 zl zpi x znl- lv pvn wxlzg pwv olv ol b gvkh Mg czlg ulv b lu p vw pv l gvk np l zlg wvghtl g zgvil og gtt zltlg l mzwu ol xtgv ptt hl slx ~g tw g p zlg wvght giopl ghtl,~ col xlipmpig pwv zly pzl ol l wm bDMc wvt olv Owkpvl)/1/' Owkpvl)/11 wz zgkpwv i/pkl pv xgz pi tg l mwzu p o ogtm)tp l nzlg lz ogv = kg pv iwv gpvulv l illk ol tl lt pv cghtl /' Ewt uv 1' w Cxxlvkp D wm tYELa 0.,/../)0.,03./ pvil ol mpt lz lu p vw iwv pklzlk lmmli p l pv zlk ipvn vwhtl ng zgkpwgi p p mzwu ng zlgu ,
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Chapter 5 contains the basis for plant procedures that the plant operator requires to meet ODCM Control 3.2.1 which limits the total fraction of combined effluent concentration in liquid pathways, excIuding noble gases, denoted here as, FF~at the point of discharge at any time (see Figure 9-1). Ff"" is limited to less than or equal to ten, Le.,
The total concentration of all dissolved and entrained noble gases at the point of discharge from all station sources, denoted CING,is limited to 2E-04 pCi/ml, Le.,
cy 12E - 04 pCi /ml.
Evaluation of ~yand crGis required concurrent with the sampling and analysis program in Control Table 4.2.1.
5.1 Method to Determine FfNGand CyG
Determine the total fraction of combined effluent concentrations at the point of discharge in liquid pathways (excluding noble gases), denoted FYIand determine the total concentration at the point of discharge of all dissolved and entrained noble gases in liquid pathways from all station sources, denoted CING,as follows:
and:
I Revision29 Date 1/11/02 5-1 cf”“= cFG 12E-04 i
where:
pENG= Total sum of the fractions of each radionuclide concentration in liquid I effluents (excluding noble gases) at the point of discharge to an unrestricted area, divided by each rahonuclide’s ECL value.
cpi = Concentration at point of discharge to an unrestricted area of radionuclide “i”, except for dssolved and entrained noble gases, from any tank or other significant source, p, from which a discharge may be made (including the floor drain sample tank, the waste sample tanks, the detergent waste tank and any other significant source from which a discharge can be made) (pCi/ml). This concentration can be calculated from: Cpi = CTKIx FTK/~DIL+ FTK) where: CTKIequals the concentration of radionuclide i in the tank to be discharged (pCi/ml); FDLis equal to the dilution flow provided by the liquid radioactive waste ddution pumps (20,000 gpm); FTK equals the liquid waste discharge pump flow rate which regulates the rate at which liquid horn a waste collection tank is discharged (gpm).
ECL, = Annual average effluent concentration limits of radlonuclide Y’, except I for &solved and entrained noble gases, from 1OCFR20.1001-20.2402, Appendix B, Table 2, Column 2 (pCi/ml).
CY = Total concentration at point of discharge to an unrestricted area of all dissolved and entrained noble gases in liquid pathways from all station sources (pCi/mlj.
cFG = Concentration at point of discharge to an unrestricted area of dissolved and entrained noble gas “i” in liquid pathways from all station sources (pCi/ml).
I Revision 2 Date 1/11/02 5-2 5.2 Method to Determine Radionuclide Concentration for Each Liquid Effluent Pathway
5.2.1 Sample Tanks Pathways
Cpi is determined for each radionuclide above LLD from the activity in a representative grab sample of any of the sample tanks and the predicted flow at the point of discharge to an unrestricted area.
Most periodic batch releases are made from the two 10,000-gallon capacity waste sample tanks. These tanks serve to hold all the high purity liquid wastes after they have been filtered through the waste collector and processed by ion exchange in the fuel pool and waste demineralizers. Other periodic batch releases may also come from the detergent waste tank or the floor drain sample tank.
The tanks are sampled from the radwaste sample sink and the contents analyzed for water quality and radioactivity. If the sample meets all the high purity requirements, the contents of the tank may be re-used in the nuclear system. If the sample does not meet all the high purity requirements, the contents are recycled through the radwaste system or discharged.
Prior to discharge each sample tank is analyzed for tntium, dissolved noble gases and dissolved and suspended gamma emitters.
5.2.2 Service Water Pathway
The service water pathway shown on Figure 9-1, flows from the intake structure through the heat exchangers and the discharge structure. Under normal operating conditions, the water in this line is not radioactive. For this reason, the service water line is not sampled routinely but it is continuously monitored with the service water discharge monitor (No. 17/351).
The alarm setpoint on the service water discharge monitor is set at a level which is three times the background of the instrument. The service water is sampled if the monitor is out of service or if the alann sounds.
I Revision 29 Date 1/11/02 5-3 ..
Under expected or anticipated operating conditions, the concentration at any time of radionuclides at the point of discharge from the service water effluent pathway to an unrestricted I area will not exceed ten times the effluent concentration values of lOCFR20.1001-20.2402, Appenchx B, Table 2, Column 2,
5.2.3 Circulating Water Pathway
The circulating water pathway shown on Figure 9-1, flows from the intake structure through the condenser and the discharge structure. Under normal operating conditions, the water in this line is not radioactive. For this reason, the circulating water line is not sampled routinely but it is monitored continuously by the discharge process monitor {No. 17/359) located in the discharge structure.
The alarm setpoint on the discharge process monitor is set at a level which is three times the background of the instrument. The circulating water is sampled if the monitor is out of service or if the aIarm sounds.
Under normal operating conditions, the average concentration of radionuclides at the point of discharge from the circulating water pathway to an unrestricted area will not exceed the I annual effluent concentration limits in 1OCFR20.1001-20.2402, Appendlx B, Table 2, Column 2.
I Revision29 Date 1/11/02 5-4 6.0 OFF-SITE DOSE CALCULATION METHODS
Chapter 6 provides the basis for plant procedures required to meet the IOCFR50, Appendix I, ALARA dose objectives, and the 40CFRI90 total dose limits to members of the public in unrestricted areas, as stated in the Radiological Effluent Controls (implementing the requirements of Technical Specification 6.7.D). A simple, conservative method (called Method I) is listed in Tables 1.1.2 to 1.1.7 for each of the Control requirements. Each of the Method I equations is presented, along with their bases in Sections 6.2 through 6.9 and Section 6.11. In addition, reSference is provided to more sophisticated but still conservative methods (called Method II) for use when more accurate results are needed. This chapter provides the methods, data, and reference material with which the operator can calculate the needed doses and dose rates. Setpoint methods for effluent monitor alarms are described in Chapter 8.
Demonstration of compliance with the dose limits of 40CFRI90 is considered to be a demonstration of compliance with the 0.1 rem limit of lOCFR20.1301(a)(I) for members of the public in unrestricted areas (Reference 56 FR23374, 3rd column).
6.1 Introductory Concepts
The Radiological Effluent Controls Program (Technical Specifications 6.7.D) either limit dose or dose rate. The term "Dose" for ingested or inhaled radioactivity means the dose commitment, measured in mrem, which results from the exposure to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped. The time frame over which the dose commitment is evaluated is 50 years. The phrases "annual Dose" or "Dose in one year" then refers to the fifty-year dose commitment from one year's worth of releases. "Dose in a quarter" similarly means a fifty-year dose commitment from one quarter's releases. The term "Dose," with respect to external exposures, such as to noble gas clouds, refer only to the doses received during the actual time period of exposure to the radioactivity released from the plant. Once the source of the radioactivity is removed, there is no longer any additional accumulation to the dose commitment.
Gaseous effluents from the plant are also controlled such that the maximum "dose rates" at the site boundary at any time are limited to 500 mremlyear. This instantaneous dose rate limit allows for operational flexibility when off normal occurrences may temporarily increase gaseous effluent release rates from the plant, while still providing controls to ensure that licensees meet the dose objectives of Appendix I to IOCFR50.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page I of 55 Vermont Yankee Nuclear Power Station It should also be noted that a dose rate due to noble gases that exceeds for a short time period (less than one hour in duration) the equivalent 500 mrem/year dose rate limit stated in Control 3.3. La, does not necessarily, by itself, constitute a Licensee Event Report (LER) under lOCFR Part 50.73 unless it is determined that the air concentration of radioactive effluents in unrestricted areas has also exceeded 20 times applicable concentration limits specified in Appendix B to 20.1001 - 20.2402, Table 2, Column 1 (four-hour notification per lOCFR50.72, and 30-day LER per lOCFR50.73).
The quantities D and R are introduced to provide calculable quantities, related to off-site dose, or dose rate which demonstrates compliance with the effluent controls.
The dose D is the quantity calculated by the Chapter 6 dose equations. The D calculated by "Method I" equations is not necessarily the actual dose received by a real individual but usually provides an upper bound for a given release because of the conservative margin built into the dose factors and the selection and definition of critical receptors. The radioisotope specific dose factors in each "Method I" dose equation represent the greatest dose to any organ of any age group accounting for existing or potential pathways of exposure. The critical receptor assumed by "Method I" equations is typically a hypothetical individual whose behavior - in terms of location and intake - results in a dose which is expected to be higher than any real individual. The Method I equations employ five-year historical average atmospheric dispersion factors to define receptors of maximum impact. Method II allows for a more exact dose calculation for real individuals, if necessary, by considering only existing pathways of exposure, or actual concurrent meteorology with the recorded release. Maximum receptor doses determined using quarterly meteorology may be greater than doses calculated with Method I due to short time period variability of meteorological conditions from the long-term average. Quarterly average dispersion values for maximum receptors have been observed to differ from five-year average values by as much as 54%.
Ris the quantity calculated in the Chapter 6 dose rate equations. It is calculated using the plant's effluent monitoring system reading and an annual average or long-term atmospheric dispersion factor. Dispersion factors based on actual concurrent meteorology during effluent releases can also be used via Method II, if necessary, to demonstrate compliance with off-site dose rate limits.
Each of the methods to calculate dose or dose rate are presented in separate sections of Chapter 6, and are summarized in Tables 1.1.1 to 1.1.7. Each method has two levels of complexity and are called Method I and Method II. Method I is the simplest; generally a linear equation. Method II is a more detailed analysis which allows for use of site-specific factors and variable parameters to be selected to best fit the actual release conditions, within the bounds of the guidance provided.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 2 of 55 Vermont Yankee Nuclear Power Station The plant has both elevated and ground level gaseous release points: the main vent stack (elevated release), and the North Warehouse waste oil burner (ground level release). Therefore, total dose calculations for skin, whole body, and the critical organ from gaseous releases will be the sum of the elevated and ground level doses. Appendix D provides an assessment of the surveillance needs for waste oil to ensure that off-site doses from its incineration is maintained within the ALAR A limits of the effluent Controls.
6.2 Method to Calculate the Total Body Dose from Liquid Releases
Effluent Control 3.2.2 limits the total body dose commitment to a Member of the Public from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year. Control 3.2.3 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 mrem in any month. Control 3.4.1 limits the total body dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year. Dose evaluation is required at least once per month. If the liquid radwaste treatment system is not being used, dose evaluation is required before each release.
Use Method I first to calculate the maximum total body dose from a liquid release to the Connecticut River as it is simpler to execute and more conservative than Method II.
Use Method II if a more accurate calculation of total body dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied.
If the radwaste system is not operating, the total body dose must be estimated prior to a release (Control 3.2.3). To evaluate the total body dose, use Equation 6.1 to estimate the dose from the planned release and add this to the total body dose accumulated from prior releases during the month.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 3 of 55 Vermont Yankee Nuclear Power Station 6.2.1 Method I
The increment in total body dose from a liquid release is:
(6-1) Dtb = L: Qi DFL itb i
(mrem) (Ci) (mr~im )
where:
DFLitb = Site-specific total body dose factor (mremlCi) for a liquid release. See Table 1.1.11.
Qi = Total activity (Ci) released for radionuclide "i." (For strontiums and Fe 55, use the most recent measurement available.)
Equation 6-1 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1. Normal operations (not emergency event),
2. Liquid releases were to the Connecticut River, and
3. Any continuous or batch release over any time period.
6.2.2 Basis for Method I
This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II.
Method I may be used to show that the effluent Controls which limit off-site total body dose from liquids (3.2.2 and 3.2.3) have been met for releases over the appropriate periods. Control 3.2.2 is based on the ALARA design objectives in lOCFR50, Appendix I Subsection II A. Control 3.2.3 is an "appropriate fraction," determined by the NRC, of that design objective (hereafter called the Objective). Control 3.4.1 is based on Environmental Standards for Uranium Fuel Cycle in 40CFR190 (hereafter called the Standard) which applies to direct radiation as well as liquid and gaseous effluents.
Exceeding the Objective or the Standard does not immediately limit plant operation but requires a report to the NRC within 30 days. In addition, a waiver may be required.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 4 of 55 Vermont Yankee Nuclear Power Station Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated" (lOCFR50, Appendix I). The definition, below, of a single "critical receptor" (a hypothetical individual whose behavior results in an unrealistically high dose) provides part of the conservative margin to the calculation of total body dose in Method I. Method II allows that actual individuals, with real behaviors, be taken into account for any given release. In fact, Method I was based on a Method II analysis for the critical receptor with maximum exposure conditions instead of any real individual. That analysis was called the "base case;" it was then reduced to form Method I.
The steps performed in the Method I derivation follow. First, in the base case, the dose impact to the critical receptor (in the form of dose factors DFLitb, mremlCi) for a 1 curie release of each radioisotope in liquid effluents was derived. The base case analysis uses the methods, data and assumptions in Regulatory Guide 1.109 (Equations A-2, A-3, A-7, A-13 and A-16, Reference A). The liquid pathways identified as contributing to an individual's dose are the consumption of fish from the Connecticut River, the ingestion of vegetables and leafy vegetation which were irrigated by river water, the consumption of milk and meat from cows and beef cattle who had river water available for drinking as well as having feed grown on irrigated land, and the direct exposure from the ground plane associated with activity deposited by the water pathway. A plant discharge flow rate of 44.6 fe/sec was used with a mixing ratio of 0.0356 which corresponds to a minimum regulated river flow of 1250 cfs at the Vernon Dam just below the plant discharge outfall. * Tables 6.2.1 and 6.2.2 outline human consumption and environmental parameters used in the analysis. The resulting, site-specific, total body dose factors appear in Table 1.1.11.
For any liquid release, during any period, the increment in annual average total body dose from radionuclide "i" is:
(6-2)
L1Dtb = Qi DFLtb
(mrem) (Ci) ( ~~m ) where:
DFLitb = Site-specific total body dose factor (mremlCi) for a liquid release. See Table 1.1.11.
Qi = Total activity (Ci) released from radionuclide "i."
* An Mp equal to 1.0 for the fish pathway is assumed between the discharge structure and the dam. Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 5 of 55 Vermont Yankee Nuclear Power Station Method I is conservative because it is based on dose factors DFLitb which were chosen from the base case to be the highest of the four age groups for each radionuclide, as well as assuming minimum river dilution flow.
6.2.3 Method II
If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable, such as the use of actual river flow at the time of actual discharge as opposed to the minimum river flow of 1,260 cfs that is assumed in the Method I dose factors (except for the fish pathway). The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 6 of 55 Vermont Yankee Nuclear Power Station TABLE 6.2.1
Environmental Parameters for Liquid Effluents at Vermont Yankee (Derived from Reference A)
FOOD GROWN WITH CONT AMINATED WATER POTABLE AQUATIC SHORELINE LEAFY COW VARIABLE WATER FOOD ACTIVITY VEGETABLES VEG. MEAT MILK MP Mixing Ratio 1.0 - 0.0356 0.0356 0.0356 0.0356 0.0356 TP Transit Time (HRS) - 24.0 0.000 0.0000 0.0000 480.0 48.0 YV Agricultural 2 (KGIM ) 2.0 2.0 2.0 2.0 Productivity P Soil Surface 2 (KGIM ) 240.0 240.0 240.0 240.0 Density IRR Irrigation Rate 2 (UM /HR) 0.152 0.152 0.152 0.152 TE Crop Exposure (HRS) 1440.0 1440.0 1440.0 1440.0 Time TH Holdup Time (HRS) 1440.0 24.0 2160.0 2160.0 QAW Water Uptake (UD) 50.0 60.0 Rate for Animal QF Feed Uptake Rate (KGID) 50.0 50.0 for Animal PI Fraction of Year Crops Irrigated 0.5 0.5 0.5 0.5 Location of Critical Connecticut River Below Vernon Dam Receptor
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 7 of 55 Vermont Yankee Nuclear Power Station TABLE 6.2.2
Usage Factors for Various Liquid Pathways at Vermont Yankee (From Reference A, Table E-5. Zero Where No Pathway Exists)
LEAFY POTABLE AGE VEG. VEG. MILK MEAT FISH INVERT. WATER SHORELINE (KGIYR) (KGIYR) (LlTERIYR) (KGIYR) (KGIYR) (KGIYR) (LITERlYR) (HRIYR) Adult 520.00 64.00 310.00 110.00 21.00 0.00 0.00 12.00 Teen 630.00 42.00 400.00 65.00 16.00 0.00 0.00 67.00 Child 520.00 26.00 330.00 41.00 6.90 0.00 0.00 14.00 Infant 0.00 0.00 330.00 0.00 0.00 0.00 0.00 0.00
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 8 of 55 Vermont Yankee Nuclear Power Station 6.3 Method to Calculate Maximum Organ Dose from Liquid Releases
Effluent Control 3.2.2 limits the maximum organ dose commitment to a Member of the Public from radioactive material in liquid effluents to 5 mrem per quarter and 10 mrem per year. Control 3.2.3 requires liquid radwaste treatment when the maximum organ dose estimate exceeds 0.2 mrem in any month. Control 3.4.1 limits the maximum organ dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year. Dose evaluation is required at least once per month if releases have occurred. If the liquid radwaste treatment system is not being used, dose evaluation is required before each release.
Use Method I first to calculate the maximum organ dose from a liquid release to the Connecticut River as it is simpler to execute and more conservative than Method II.
Use Method II if a more accurate calculation of organ dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied.
If the radwaste system is not operating, the maximum organ dose must be estimated prior to a release (Control 3.2.3). To evaluate the maximum organ dose, use Equation 6-3 to estimate the dose from the planned release and add this to the maximum organ dose accumulated from prior releases during the month.
6.3.1 Method I
The increment in maximum organ dose from a liquid release is:
(6-3) Dmo = L Qi DFL imo i
(mrem) (Ci) (~~ ) where:
DFLimo Site-specific maximum organ dose factor (mremlCi) for a liquid release. See Table 1.1.11.
Qi = Total activity (Ci) released for radionuclide "i." (For strontiums and Fe-55, use the most recent measurement available.)
Equation 6-3 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1. Normal operations (not emergency event), 2. Liquid releases were to the Connecticut River, and 3. Any continuous or batch release over any time period. Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 9 of 55 Vermont Yankee Nuclear Power Station 6.3.2 Basis for Method I
This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. The methods to calculate maximum organ dose parallel the total body dose methods (see Section 6.2.2). Only the differences are presented here.
For each radionuc1ide, a dose factor (rnrernJCi) was determined for each of seven organs and four age groups. The largest of these was chosen to be the maximum organ dose factor (DFLimo) for that radionuc1ide.
For any liquid release, during any period, the increment in annual average dose from radionuc1ide "i" to the maximum organ is:
(6-4)
~ Dmo = Qi DFL imo
(rnrem) (Ci) (rnr~~ ) where:
DFLimo = Site-specific maximum organ dose factor (mrernJCi) for a liquid release. See Table 1.1.11.
Qi = Total activity (Ci) released for radionuc1ide "i".
Because of the assumptions about receptors, environment, and radionuc1ides; and because of the low Objective and Standard, the lack of immediate restriction on plant operation, and the adherence to lOCFR20 concentrations (which limit public health consequences) a failure of Method I (i.e., the exposure of a real individual being underestimated) is improbable and the consequences of a failure are minimal.
6.3.3 Method II
If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 10 of 55 Vermont Yankee Nuclear Power Station 6.4 Method to Calculate the Total Body Dose Rate From Noble Gases
Effluent Control 3.3.1 limits the instantaneous dose rate at any time to the total body from all release sources of noble gases at any location at or beyond the site boundary equal to or less than 500 rnrernlyear.
Use Method I first to calculate the Total Body Dose Rate from the peak release rate via both elevated and ground level release points. The dose rate limit of Control 3.3.1.a is the total contribution from both ground and elevated releases occurring during the period of interest.
Use Method II if Method I predicts a dose rate greater than the Control limit (i.e., use of actual meteorology over the period of interest) to determine if, in fact, Control 3.3.1 had actually been exceeded during a short time interval.
Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant stack noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site dose rate limit of Control 3.3.1, or a value below it, taking into account the potential contribution of releases from all ground level sources.
Determinations of dose rates for compliance with Control (3.3.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Control 3.3.1 is unsuccessful, or as required by the notations to Control Table 3.1.2 when the stack noble gas monitor is inoperable.
6.4.1 Method I
The Total Body Dose Rate due to noble gases can be determined by multiplying the individual radionuclide release rates by their respective dose factors, summing all the products together, and then multiplying this total by a conversion constant (0.75), as seen in the following Equation 6-5:
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page II of 55 Vermont Yankee Nuclear Power Station DFB. Rtbs 0.75 1 (6-5)
3 mremJ (PC~ - se~J (~Ci) (mrem - m J ( yr ~Cl - m sec pCi - yr where:
Q ~T = In the case of noble gases, the release rate from the plant stack (J..lCi/sec)for each radionuclide, "i", identified. The release rate at the plant stack is based on measured radionuclide concentrations and distributions in periodic grab samples taken at the stack. As an alternative method, the radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) can be used during plant operations, along with the Stack Gas Monitor effluent count rate, to estimate stack radionuclide releases. The release rate at the stack when using SJAE samples can be stated as follows:
.SJAE . STQ " I Q - ---:::.!...I -::-:-:-::- M F i-I Q~JAE Sg I (6-28) uCi = (cpm) (11 Cilcc ) (cc) sec cpm sec
M Plant Stack Gas Monitor I or II count rate (cpm).
Sg = Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpm/(J...lCilcc»).
F = Stack flow rate (cc/sec) .
•SJAE Q i = The last measured release rate at the steam jet air ejector of noble gas i (J...lCilsec).
DFBi = Total body gamma dose factor (see Table 1.1.10).
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 12 of 55 Vermont Yankee Nuclear Power Station For ground level noble gas releases, the total body dose rate is calculated as follows:
7.4 DFB. R tbg 1 (6-39)
where:
Q~L = Ground level release rate (J,-ICi/sec)of noble gas.
The total body dose rate for the site is equal to Rtbs + Rtbg'
During periods (beyond the first five days) when the plant is shutdown and no radioactivity release rates can be measured at the SJAE, Xe-133 may be used in place of the last SJAE measured mix as the referenced radionuclide to determine off-site dose rate and monitor setpoints. In this case, the ratio of each Q~JAE to the sum of all
Q~JAE in Equation 6-28 above is assumed to reduce to a value of 1, and the total body gamma dose factor DFBj for Xe-133 (2.94 E-04 mrem-m'VpCi-yr) is used in Equation 6- 5. Alternately, a relative radionuclide "i" mix fraction (fj) may be taken from Table 8.2.1 as a function of time after shutdown, and substituted in place of the ratio of QSJAE to the sum of all QSJAE in Equation (6-28) above to determine the relative fraction I I of each noble gas potentially available for release to the total. Just prior to plant startup before a SJAE sample can be taken and analyzed, the monitor alarm setpoints should be based on Xe-138 as representing the most prevalent high dose factor noble gas expected to be present shortly after the plant returns to power. Monitor alarm setpoints which have been determined to be conservative under any plant conditions may be utilized at any time in lieu of the above assumptions.
Equations 6-5 and 6-39 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1. Normal operations (not emergency event), and
2. Noble gas releases via either elevated or ground level vents to the atmosphere.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 13 of 55 Vermont Yankee Nuclear Power Station 6.4.2 Basis for Method I
Method Imay be used to show that the Control limit for total body dose rate from noble gases released to the atmosphere has been met for the peak noble gas release rate.
Method Ifor stack releases was derived from Regulatory Guide 1.109 as follows:
Rtbs = 1E + 06 SF [X/Q] : L
3 (6-6) mrem J = ( pC~ J (# ) sec J( !-lCi ] (mre~ - m J ( yr !-lCI ( m3 sec pCI - yr where:
= Shielding factor = 1.0 for dose rate determination.
= Maximum annual average gamma atmospheric dispersion factor for stack (elevated) releases; = 7.51E-07 (sec/rrr').
QST = Release rate from the plant stack of noble gas "i" (!-lCi/sec). i 3 Gamma total body dose factor, (mre~- m ]. See Table 1.1.10. pC 1- yr
Equation 6-6 reduces to:
Rtbs = 0.75 (6-5)
For ground level releases, the ground level maximum long-term average gamma atmospheric dispersion factor = 7.37E-06 sec/rrr', thus leading to:
Rtbg = I E+ 06 * 7.37 E- 06 I Q~L DFBi i or (6-39) . GL Rtbg = 7.4 I Qi DFBi
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 14 of 55 Vermont Yankee Nuclear Power Station The selection of critical receptor, outlined in Section 6.10, is inherent in Method I, as are the maximum expected off-site annual or long-term average atmospheric dispersion factors. Due to the holdup and decay of gases allowed in the AOG, off-gas concentrations at the plant stack during routine plant operations are usually too low for determination of the radionuc1ide mix at the plant stack. It is then conservatively assumed that most of the noble gas activity at the plant stack is the result of in-plant steam leaks which are removed to the plant stack by building ventilation air flow, and that this air flow has an isotopic distribution consistent with that routinely measured at the SJAE.
The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.
In the case of noble gas dose rates, Method II cannot provide much extra realism
because Rtbs and Rtbg are already based on several factors which make use of current plant parameters. However, should it be needed, the dose rate analysis for critical receptor can be performed making use of current meteorology during the time interval of recorded peak release rate in place of the default atmospheric dispersion factor used in Method I.
6.4.3 Method II
If Method I cannot be applied, or if the Method I dose exceeds the limit, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 15 of 55 Vermont Yankee Nuclear Power Station 6.5 Method to Calculate the Skin Dose Rate from Noble Gases
Effluent Control 3.3.1 limits the instantaneous dose rate at any time to the skin from all release sources of noble gases at any location at or beyond the site boundary to 3,000 rnremlyear.
Use Method I first to calculate the Skin Dose Rate from both elevated and ground level release points to the atmosphere. The dose rate limit of Control 3.3.1.a is the total contribution from both ground and elevated releases occurring during the period of interest. Method I applies at all release rates.
Use Method II if Method I predicts a dose rate greater than the Control limits (i.e., use of actual meteorology over the period of interest) to determine if, in fact, Control 3.3.1 had actually been exceeded during a short time interval.
Compliance with the dose rate limits for noble gases are continuously demonstrated when effluent release rates are below the plant stack noble gas activity monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site Control dose rate limit, or a value below it, taking into account the potential contribution releases from all ground level sources.
Determinations of dose rate for compliance with Control (3.3.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Control 3.3.1 is unsuccessful, or as required by the notations to Control Table 3.1.2 when the stack noble gas monitor is inoperable.
6.5.1 Method I
The skin dose rate due to noble gases is determined by multiplying the individual radionuclide release rates by their respective dose factors, and summing all the products together as seen in the following Equation 6-7:
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 16 of 55 Vermont Yankee Nuclear Power Station R skins (6-7) ~) (.mrem~ sec) ( sec J1 CI- yr where:
QST = In the case of noble gases, the noble gas release rate from the plant stack I (f..tCiisec)for each radionuclide, "i," identified. The release rate at the plant stack is based on measured radionuclide concentrations and distributions in periodic grab samples taken at the stack. As an alternative method, the radionuclide distribution in the off-gas at the Steam Jet Air Ejector (SJAE) can be used during plant operations, along with the Stack Gas Monitor effluent count rate, to estimate stack radionuclide releases. The release rate at the stack when using SJAE samples can be stated as follows:
= Q~JAE M 1 F - I:Q~JAE Sg I (6-28)
f..tCi = (cpm) (f..tCi/cc) (cc) sec cpm sec M = Plant stack gas monitor I or IIcount rate (cpm).
Appropriate or conservative plant stack monitor detector counting efficiency for the given nuclide mix (cpml(f..tCilcc)).
F = Stack flow rate (cc/sec).
The last measured release rate at the steam jet air ejector of noble gas i (f..tCi/sec).
DFis = combined skin dose factor (see Table 1.1.10) for stack release.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 17 of 55 Vermont Yankee Nuclear Power Station For ground level releases, the skin dose rate from noble gases is calculated by Equation 6-38:
R sking = L Q ~L DF' ig (6-38) i where:
Q~L = The noble gas release rate from ground level (J.lCi/sec) for each radionuclide "i" identified.
DFi~ = Combined skin dose factor for a ground level release [see Table 1.1.lOA].
The skin dose rate for the site is equal to Rskins + Rsking,
During periods (beyond the first five days) when the plant is shutdown and no radioactivity release rates can be measured at the SJAE, Xe-133 may be used in place of the last SJAE measured mix as the referenced radionuclide to determine off-site dose rate and monitor setpoints. In this case, the ratio each of Q~JAE to the sum of all Q~JAE in Equation 6-28 above is assumed to reduce to a value of 1, and the combined skin dose factor DF'is for Xe-I33 (7.8IE-04 rurem-sec/uf.i-year) is used in Equation 6-7.
Alternately, a relative radionuclide "i" mix fraction (fi) may be taken from Table 8.2.1 as a function of time after shutdown, and substituted in place of the ratio of each Q~JAE
to the sum of all Q~JAE in Equation 6-28 above to determine the relative fraction of each noble gas potentially available for release to the total. Just prior to plant startup before a SJAE sample can be taken and analyzed, the monitor alarm setpoints should be based on Xe-I38 as representing the most prevalent high dose factor noble gas expected to be present shortly after the plant returns to power. Monitor alarm setpoints which have been determined to be conservative under any plant conditions may be utilized at any time in lieu of the above assumptions.
Equations 6-7 and 6-38 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1. Normal operations (not emergency event), and
2. Noble gas releases via both elevated and ground level vents to the atmosphere.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 18 of 55 Vermont Yankee Nuclear Power Station 6.5.2 Basis For Method I
The methods to calculate skin dose rate parallel the total body dose rate methods in Section 6.4.3. Only the differences are presented here.
Method Imay be used to show that the Control limit for skin dose rate from noble gases released to the atmosphere (Control 3.3.1) has been met for the peak noble gas release rate.
Method I was derived from Regulatory Guide 1.109 as follows:
= 1.11 (6-8)
3 mrem) (#) (mrad) P~i- yr) Ci) (se~) [mre~- m ) ( mrad yr ( Ci= sec ( yr m pCI- yr
where:
1.11 = Average ratio of tissue to air absorption coefficients will convert mrad in air to mrem in tissue.
D~ir = 3.17E +04 L Qi [X/Q]s DF! (6-9) i
mrad J (P~i - yr) (Ci J (sec) (mra~ - m 3 ) ( yr CI - sec yr m 3 pCI - yr
- Dr now D hnite - air [X / Q H/ [X / oi, (6-10) ( m;;dJ (m;;dJ (:n (::J
and 31.54 . ST Qi = Qi (6-11)
( Ci :- sec J (!lei J (~:J /lCI - yr sec
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 19 of 55 Vermont Yankee Nuclear Power Station so Rskins = 1.11 SF IE + 06 [XlQ]Z I Q~T (6-12)
mrem J( mrem J (#) (PC~ J( se~ J ( !-lCiJ (mra~ - m 3 J ( yr mrad jiCl m - sec pCl - yr
DFS.I
3 !-lCi (mre~ - m ( P~i) J sec PCl- yr
substituting
[XlQ] I = 7.S1E-07 sec/m3 XlQs = 1.S9E-06 sec/m3 SF = Shielding factor = 1.0 for dose rate determinations
gives
Rskins = 0.83 DIf + DF~ (6-13)
3 3 mremJ(pCi. - sec - mremJ (!-lCi) [mra~ - m J(pC~ - secJ(!-lCi)(mre~ - m J ( yr !-lCl- m' - mrad sec pCl- yr !-lCl- m' sec pCl- yr
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 20 of 55 Vermont Yankee Nuclear Power Station = L Q~T [0.83 DF7 + 1.59 DFS d (6-14) I define DF'is = 0.83 DF r + 1.59 DFS i (6-15) then
DI1s Rskins = L (6-7)
~Ci) (mre~ - sec) ( sec ~Cl- yr
For determining combined skin doses for ground level releases, a [XlQ)gy = 7.37E-06 sec/m3 and an undepleted XlQg = 3.65E-05 sec/m3 have been substituted into Equation 6-12 to give:
R sking = I Q~L (8.18 DF iY + 36.5 DF i) \
then DF(g = 8.18 DFjY + 36.5 DFS i (6-37)
and'Rsking = L.. " QG\, L DF\,'g (6-38) i where:
. GL Qi = The noble gas release rate from ground level release points (f..lCi/sec)for each radionuclide "i" identified.
DFig = Combined skin dose factor for a ground level release [see Table 1.1.lOA).
The selection of critical receptor, outlined in Section 6.10 is inherent in Method I, as it determined the maximum expected off-site atmospheric dispersion factors based on past long-term site-specific meteorology.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 21 of 55 Vermont Yankee Nuclear Power Station The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.
6.5.3 Method II
If Method I cannot be applied, or if the Method I dose exceeds the limit, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.
6.6 Method to Calculate the Critical Organ Dose Rate from lodines, Tritium and Particulates with T 112 Greater Than 8 Days
Effluent ControI3.3.1.b limits the dose rate to any organ, denoted Reo, from all release sources of 1-131, 1-133, H-3, and radionuclides in particulate form with half lives greater than 8 days to 1500 mremlyear to any organ. The peak release rate averaging time in the case of iodines and particulates is commensurate with the time the iodine and particulate samplers are in service between changeouts (typically a week).
Use Method I first to calculate the critical organ dose rate from both elevated and ground level release points to the atmosphere. The dose rate limit of Control 3.3.1.b is the total contribution from both ground and elevated releases occurring during the period of interest. Method I applies at all release rates.
Use Method II if Method I predicts a dose rate greater than the Control limits (i.e., use of actual meteorology over the period of interest) to determine if, in fact, Control 3.3.l.b had actually been exceeded during the sampling period.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 22 of 55 Vermont Yankee Nuclear Power Station 6.6.1 Method I
The critical organ dose rate from stack releases can be determined by multiplying the individual radionuclide release rates by their respective dose factors and summing all their products together, as seen in the following Equation 6-16:
Rcos = L DFG'sico
(6-16) /.lCi) ( mre~ - sec) ( sec /.lCI- yr where: = Stack activity release rate determination of radio nuclide "i" (lodine-131, Iodine-133, particulates with half-lives greater than 8 days, and tritium), in f.lCi/sec. For i = Sr89, Sr90 or tritium, use the best estimates (such as most recent measurements).
DFG'sico = SIte. speer ific ICcntica cri I organd ose rate f actor (mrem-sec) . uci- yr for a ground level gaseous release. See Table 1.1.12.
For ground releases (North Warehouse waste oil burner) the critical organ dose rate from Iodine, Tritium, and Particulates with T 112greater than 8 days is calculated as follows:
" . GLP (6-40) Rcog = L... Qi DFG'gico where:
= Ground activity release rate determination of radionuclide "i" (Iodine- 131, lodine-133, particulates with half-lives greater than 8 days, and tritium), in f.lCiisec. For i= Sr89, Sr90, Fe-55, or tritium, use the best estimates (such as most recent measurements). For waste oil, the release rate is the total activity by radionuclide divided by the estimated bum time. (See Appendix D for surveillance criteria on waste oil burning.
DFG'gico = S·ite speer ific ICcntica cri I organ d ose rate f actor (mrem-sec) . f.lC1-yr for a ground level gaseous release. See Table 1.1.12.
The critical organ dose rate for the site is equal to Reos + Reog . Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 23 of 55 Vermont Yankee Nuclear Power Station Equations 6-16 and 6-40 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1. Normal operations (not emergency event), and
2. Tritium, iodine, and particulate releases via either elevated or ground level vents to the atmosphere.
6.6.2 Basis for Method I
The methods to calculate critical organ dose rate parallel the total body dose rate methods in Section 6.4.3. Only the differences are presented here.
Method I may be used to show that the Control limit for organ dose rate from iodines, tritium and radionuc1ides in particulate form with half lives greater than 8 days (hereafter called Iodines and Particulates or "I+P") released to the atmosphere (ControI3.3.1.b) has been met for the peak 1+ P release rate.
The equation for ~os and ~og is derived by modifying Equation 6-25 from Section 6.9 as follows:
(6-17) (mrem) (Ci) (~~m)
applying the conversion factor, 31.54 (Ci-sec/trCi-yr) and converting Q to Q in IlCiisec as it applies to the plant stack yields:
= 31.54 • STP DFGsieo Reos Qi
(6-18) Ci~ sec IlCi) (mr~m) J ( ( IlCl- yr sec ci
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 24 of 55 Vermont Yankee Nuclear Power Station Equation 6.8 is written in the form:
R = 31.54 DFGsico
(6-19) Ci ~sec /.lCi] J ( (mr~m] ( /.lCl- yr sec Ci
DFG~ico and DFG~ico (North Warehouse waste oil burner vent releases) incorporates the conversion constant of 31.54 and has assumed that the shielding factor (SF) applied to the direct exposure pathway from radionuclides deposited on the ground plane is equal to 1.0 in place of the SF value of 0.7 assumed in the determination of DFG sicoand DFG gicofor the integrated doses over time.
The selection of critical receptor (based on the combination of exposure pathways which include direct dose from the ground plane, inhalation and ingestion of vegetables, meat, and milk) which is outlined in Section 6.10 is inherent in Method I, as are the maximum expected off-site atmospheric dispersion factors based on past long- term site-specific meteorology.
The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.
Should Method II be needed, the analysis for critical receptor critical pathway(s) and atmospheric dispersion factors may be performed with actual meteorologic and latest land use census data to identify the location of those pathways which are most impacted by these type of releases.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 25 of 55 Vermont Yankee Nuclear Power Station 6.6.3 Method II
If Method I cannot be applied, or if the Method I dose exceeds the limit, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site- specific models, data or assumptions are more applicable. The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.
6.7 Method to Calculate the Gamma Air Dose from Noble Gases
Effluent Control 3.3.2 limits the gamma dose to air from all release sources of noble gases at any location at or beyond the site boundary to 5 rnrad in any quarter and 10 mrad in any year. Dose evaluation is required at least once per month.
Use Method I first to calculate the gamma air dose for elevated and ground level vent releases during the period. The total gamma air dose limit of Control 3.3.2 is the total contribution from both ground and elevated releases occurring during the period of interest.
Use Method II if a more accurate calculation is needed.
6.7.1 Method I
The gamma air dose from plant stack releases is:
0.024 Df'Y = 1
(6-21) (mrad) 3 PCi- yr) (Ci) (mra~- m J ( Ci-m3 pCI- yr where:
Q?T = total noble gas activity (Curies) released to the atmosphere via the plant stack of each radionuclide "i'' during the period of interest.
DF? = gamma dose factor to air for radionuclide "i." See Table 1.1.10.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 26 of 55 Vermont Yankee Nuclear Power Station For ground level noble gas releases, the gamma air dose is calculated as follows:
Y, D arrg = 0.23 L: Q ~L DF! (6-41) i where:
Q9L = Total noble gas activity (curies) released to the atmosphere via ground level vents of each radionuclide, "i", during the period of interest.
The gamma air dose for the site is equal to D rirs + D ai~g •
Equations 6-21 and 6-41 can be applied under the following conditions (otherwise justify Method lor consider Method II):
1. Normal operations (not emergency event), and
2. Noble gas releases via either elevated or ground level vents to the atmosphere.
6.7.2 Basis for Method I
Method Imay be used to show that the Control limit for off-site gamma air dose from gaseous effluents (3.3.2) has been met for releases over appropriate periods. This Control is based on the Objective in lOCFR50, Appendix I, Subsection B.l, which limits the estimated annual gamma air dose at unrestricted area locations.
Exceeding the Objective does not immediately limit plant operation but requires a report to the NRC.
For any noble gas release, in any period, the dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109 with the added assumption that
Dhnite = DY [XlQ] Y /[XlQ] :
(6-22)
(mrad) (P~i- yr) (sec/m3)(Ci) (mrad- ~3 J CI- sec yr- pCI
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 27 of 55 Vermont Yankee Nuclear Power Station where:
[XJQn = maximum long term average gamma atmospheric dispersion factor for a stack release.
= 7.51E-07 (sec/rrr')
Q~T = number of curies of noble gas ''i" released from the plant stack which leads to:
0.024 = DFY1
(6-21) (mrad) PCi - yr) (Ci) [mra~-m3J ( Ci-m3 pCI- yr
For the ground level release:
D rirg = 3.17E + 04 [XlQ]~ L QpL DF! (6-42) i
where:
(XlQ)~ = Maximum long-term average gamma atmospheric dispersion factor for a ground level release
= 7.37E-06 sec/m ' leading to:
DIirg = 0.23 I QpL DF? (6-41) I
The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 28 of 55 Vermont Yankee Nuclear Power Station The main difference between Method I and Method II is that Method II would allow the use of actual meteorology to determine [X/Q]Y rather than use the maximum long-term average value obtained for the years 2002 to 2006.
6.7.3 Method II
If the Method Idose determination indicates that the Control limit may be exceeded, or if a more exact calculation is required, then Method IImay be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable.
6.8 Method to Calculate the Beta Air Dose from Noble Gases
Effluent Control 3.3.2 limits the beta dose to air from all release sources of noble gases at any location at or beyond the site boundary to 10 mrad in any quarter and 20 mrad in any year. Dose evaluation is required at least once per month.
Use Method Ifirst to calculate the beta air dose for elevated and ground level vent releases during the period. The total beta air dose limit of Control 3.3.2 is the total contribution from both ground and elevated releases occurring during the period of interest.
Use Method IIif a more accurate calculation is needed or if Method Icannot be applied.
6.8.1 Method I
The beta air dose from plant vent stack releases is:
ST DF~ D~irs = 0.050 L QII (6-23) i
3 PCi- yr) (Ci) (mra~- m ) (mrad) ( Ci- m3 pCI- yr where:
DFr = beta dose factor to air for radionuc1ide "i". See Table 1.1.10.
Q~T = total noble gas activity (curies) released to the atmosphere via the plant stack of each radionuc1ide "i" during the period of interest.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 29 of 55 Vermont Yankee Nuclear Power Station For ground level noble gas releases, the beta air dose is calculated as follows:
D~' = 1.16 L QG(, L DF~( a(fg (6-43) where:
Qar = Total noble gas activity (curies) released to the atmosphere via the ground level vents of each radionuclide "i" during the period of interest.
The beta air dose for the site is equal to D~irs + D~irg .
Equations 6-23 and 6-43 can be applied under the following conditions (otherwise justify Method I or consider Method II):
1. Normal operations (not emergency event), and
2. Noble gas releases via either elevated or ground level vents to the atmosphere.
6.8.2 Basis for Method I
This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. The methods to calculate beta air dose parallel the gamma air dose methods in Section 6.7.3. Onl y the differences are presented here.
Method I may be used to show that the Control limit for off-site beta air dose from gaseous effluents (3.3.2) has been met for releases over appropriate periods. This Control is based on the Objective in 10CFR50, Appendix I, Subsection B.l, which limits the estimated annual beta air dose at unrestricted area locations.
Exceeding the Objective does not immediately limit plant operation but requires a report to the NRC within 30 days.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 30 of 55 Vermont Yankee Nuclear Power Station For any noble gas release, in any period, the dose is taken from Equations B-4 and B-S of Regulatory Guide 1.109:
= 3.17E+04 XlQ s.c..," QSjT
(6-24) (mrad) 3 P~i - (Ci) (mra~ - m ] ( yrJ (sec] cr sec m3 pCl- yr substituting
XlQs = Maximum long term average undepleted atmospheric dispersion factor for a stack release.
= 1.S9E-06 sec/m3
We have (6-23) 0.050 DP.airs = DFf
(mrad) 3 (Ci) (mra~ - m ) pCt- yr
For the ground level release:
= 3.17E +04 (XlQ)g L Q?L DFf (6-44) j where:
(XlQ)g = Maximum long-term average undepleted atmospheric dispersion factor for a ground level release.
= 3.6SE-OS sec/m3 leading to:
D~jrg = 1.16 L: Dfi3 (6-43) i
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 31 of 55 Vermont Yankee Nuclear Power Station The calculation of ground level release dispersion parameters are based on the location of the North Warehouse with respect to the site boundary that would experience the highest exposure. The North Warehouse contains a waste oil burner that can be used for the incineration of low level contaminated waste oil, and is designated as a ground level release point to the atmosphere. Due to differences in building cross sectional areas and resulting building wake effects, the North Warehouse atmospheric dispersion factors are conservative in comparison to those associated with the main plant facilities, such as the Turbine Building. As a consequence, any potential or unexpected ground level release from the Turbine Building or adjoining structures can utilize the above ground release dose assessment equations.
6.8.3 Method II
If Method I cannot be applied, or if the Method I dose determination indicates that the Control limit may be exceeded, or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site- specific models, data or assumptions are more applicable.
6.9 Method to Calculate the Critical Organ Dose from Iodines, Tritium and Particulates
Effluent Control 3.3.3 limits the critical organ dose to a Member of the Public from all release sources of 1-131, 1-133, Tritium, and particulates with half-lives greater than 8 days (hereafter called "I+P") in gaseous effluents to 7.5 mrem per quarter and 15 mrem per year.
Use Method I first to calculate the critical organ dose from both elevated and ground level vent releases. The total critical organ dose limit of Control 3.3.3 is the total contribution from both ground level and elevated releases occurring during the period of interest
Use Method II if a more accurate calculation of critical organ dose is needed (i.e., Method I indicates the dose is greater than the limit).
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 32 of 55 Vermont Yankee Nuclear Power Station 6.9.1 Method I
o., = I Q~TP DFGsico (6-25) i
(mrem) (Ci) (mr~m)
Q 1TP = Total activity (Ci) released from the stack to the atmosphere of radionuclide "i" during the period of interest. For strontiums and tritium, use the most recent measurement.
DFGsico = Site-specific critical organ dose factor for a stack gaseous release of radionuclide "i" (mremlCi). For each radionuclide it is the age group and organ with the largest dose factor. See Table 1.1.12.
The critical organ dose is calculated for ground level releases as follows:
Deog = I Q~LP DFGgico
(6-44) (mrem) (Ci) (~~m)
Q 9LP = Total activity (Ci) released from ground level vents to the atmosphere of radionuclide "i" during the period of interest. For tritium, strontiums, and Fe-55 use the most recent measure.
DFGgico = Site-specific critical organ dose factor for a ground level release of nuclide "i" (mremlCi). For each radionuclide it is the age group and organ with the largest dose factor. See Table 1.1.12.
The critical organ dose for the site is equal to Deos+ Deog.
Equations 6-25 and 6-44 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1. Normal operations (not emergency event),
2. I+P releases via the plant stack, Turbine Building, and waste oil burner (see Appendix D for surveillance criteria on waste oil burning), to the atmosphere, and
3. Any continuous or batch release over any time period.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 33 of 55 Vermont Yankee Nuclear Power Station 6.9.2 Basis for Method I
This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II.
Method I may be used to show that the Control limit for off-site organ dose from gases (3.3.3) has been met for releases over the appropriate periods.
Method I was developed such that "the actual exposure of an individual ... is unlikely to be substantially underestimated" (10CFR50, Appendix I). The use below of a single "critical receptor" provides part of the conservative margin to the calculation of critical organ dose in Method 1. Method II allows that actual individuals, with real behaviors, be taken into account for any given release. In fact, Method I was based on a Method II analysis of the critical receptor for the annual average conditions. For purposes of complying with the Control 3.3.3, maximum annual average atmospheric dispersion factors are appropriate for batch and continuous releases. That analysis was called the "base case"; it was then reduced to form Method 1. The base case, the method of reduction, and the assumptions and data used are presented below.
The steps performed in the Method I derivation follow. First, in the base case, the dose impact to the critical receptor in the form of dose factors (mremlCi) of 1 curie release of each I+P radionuclide to gaseous effluents was derived. Then Method I was determined using simplifying and further conservative assumptions. The base case analysis uses the methods, data and assumptions in Regulatory Guide 1.109 (Equations C-2, C-4 and C-13 in Reference A). Tables 6.9.1 and 6.9.2 outline human consumption and environmental parameters used in the analysis. It is conservatively assumed that the critical receptor lives at the "maximum off-site atmospheric dispersion factor location" as defined in Section 6.10. However, he is exposed, conservatively, to all pathways (see Section 6.10). The resulting site-specific dose factors are for the maximum organ and the age group with the highest dose factor for that organ. These critical organ, critical age dose factors are given in Table 1.1.12.
For any gas release, during any period, the increment in annual average dose from radionuclide "i" is:
(6-26)
where DFGico is the critical dose factor for radionuclide "i" and Qi is the activity of radionuclide "i" released in curies.
Method I is more conservative than Method II in the region of the effluent dose Control limit because it is based on the following reduction of the base case. The dose factors DFGico used in Method I were chosen from the base case to be the highest of the set for that radionuclide. In effect each radionuclide is conservatively represented by its own critical age group and critical organ. Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 34 of 55 Vermont Yankee Nuclear Power Station 6.9.3 METHOD II
If Method I cannot be applied, or if the Method I dose exceeds the Control limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented above, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 35 of 55 Vermont Yankee Nuclear Power Station TABLE 6.9.1
Environmental Parameters for Gaseous Effluents at Vermont Yankee (Derived from Reference A)*
Variable Vegetables Cow Milk Goat Milk Meat Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural (Kg/m2) 2 2 0.70 2 0.70 2 0.70 2 Productivity
P Soil Surface (Kg/m2) 240 240 240 240 240 240 240 240 Density
T Transport Time to User(5) (Hrs) 48 48 48 48 480 480 TB Soil Exposure Time(l) (Hrs) 131400 l31400 l31400 l31400 131400 l31400 131400 l31400 TE Crop Exposure Time to Plume (Hrs) 1440 1440 720 1440 720 1440 720 1440 TH Holdup After Harvest (Hrs) 1440 24 0 2160 0 2160 0 2160 QF Animals Daily Feed (Kg/Day) 50 50 6 6 50 50 FP Fraction of Year on Pasture'" 0.50 0.50 0.50 FS Fraction Pasture When on Pasture" 1 1 1 FG Fraction of Stored Veg. Grown in Garden 0.76 FL Fraction of Leafy Veg. Grown in Garden PI Fraction Elemental Iodine = 0.5 A Absolute Humidity = 5.6(gm/rrr')'?'
*Regulatory Guide 1.109, Revision 1.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 36 of 55 Vermont Yankee Nuclear Power Station TABLE 6.9.1 (Continued)
Notes:
(1) For Method II dose/dose rate analyses of identified radioactivity releases of less than one year, the soil exposure time for that release may be set at 8760 hours (1 year) for all pathways.
(2) For Method II dose/dose rate analyses performed for releases occurring during the first or fourth calendar quarters, the fraction of time animals are assumed to be on pasture is zero (nongrowing season). For the second and third calendar quarters, the fraction of time on pasture (FP) will be set at 1.0. FP may also be adjusted for specific farm locations if this information is so identified and reported as part of the land use census.
(3) For Method II analyses, the fraction of pasture feed while on pasture may be set to less than 1.0 for specific farm locations if this information is so identified and reported as part of the land use census.
(4) For all Method II analyses, an absolute humidity value equal to 5.6 (gm/nr') shall be used to reflect conditions in the Northeast (Reference: Health Physics Journal, Vol. 39 (August), 1980; Page 318-320, Pergammon Press).
(5) Variable T is a combination of variables TF and TS in Regulatory Guide 1.109, Revision 1.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 37 of 55 Vermont Yankee Nuclear Power Station TABLE 6.9.2
Usage Factors for Various Gaseous Pathways at Vermont Yankee (from Regulatory Guide 1.109, Table E-5)
Leafy Age Vegetables Vegetables Milk Meat Inhalation Group (kg/yr) (kg/yr) (lIyr) (kg/yr) (rrr'zyr)
Adult 520.00 64.00 310.00 110.00 8000.00 Teen 630.00 42.00 400.00 65.00 8000.00 Child 520.00 26.00 330.00 4l.00 3700.00 Infant 0.00 0.00 330.00 0.00 1400.00
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 38 of 55 Vermont Yankee Nuclear Power Station 6.10 Receptor Point and Long- Tenn Average Atmospheric Dispersion Factors for Important Exposure Pathways
The gaseous effluent dose methods have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dose than anyone else. The following exposure pathways to gaseous effluents listed in Regulatory Guide 1.109 (Reference A) have been considered for radioiodines, tritium, and particulates with half lives greater than 8 days:
1. Direct exposure to contaminated ground; 2. Inhalation of air; 3 Ingestion of vegetables; 4. Ingestion of cow's milk; and 5. Ingestion of meat.
Beta and gamma air doses have also been considered for noble gases in plant effluents along with whole body and skin dose rate calculations.
Section 6.10.1 details the selection of important off-site locations and receptors. Section 6.10.2 describes the atmospheric model used to convert meteorological data into atmospheric dispersion factors. Section 6.10.3 presents the maximum atmospheric dispersion factors calculated at each of the off-site receptor locations.
6.10.1 Receptor Locations
Distances to the site boundary from the two evaluated gaseous release pathways (the Stack and North Warehouse) are provided in Table 6.10.2. Four important off-site receptor locations are considered in the dose and dose rate equations for gaseous radioactive effluents from these two release pathways. They are:
1. The point of maximum gamma exposure (maximum gamma XlQ) from an overhead noble gas cloud for determining skin and whole body dose rates and gamma air doses;
2. The point of maximum ground level air concentration (maximum undepleted XlQ) of noble gases for determining skin and beta air dose rates and doses;
3. The point of maximum ground level air concentration (maximum depleted XlQ) of radioiodines and other particulates for determining critical organ dose from inhalation; and
4. The point of maximum deposition (maximum D/Q) of radioiodines and other particulates for determining critical organ dose from ingestion.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 39 of 55 Vermont Yankee Nuclear Power Station The Stack release pathway was evaluated as an elevated release assuming a constant nominal Stack flow rate of 146,160 cfm. The point of maximum gamma exposure from Stack releases (SSE sector, 700 meters) was determined by finding the maximum five-year average gamma X/Q at any off-site location. The location of the maximum ground level air concentration and deposition of radioiodines and other particulates (WNW sector, 2150 meters) was determined by finding the maximum five- year average depleted X/Q and D/Q at any off-site location. For the purposes of determining the Method I dose factors for radioiodines, tritium, and particulates, a milk animal was assumed to exist at the location of highest calculated ground level air concentration and deposition of radioiodines and other particulates as noted above. This location then conservatively bounds the deposition of radionuclides at all real milk animal locations.
The North Warehouse release pathway was evaluated as a ground level release using the same meteorological period-of-record as the stack. The highest long-term atmospheric dispersion factors at the site boundary were determined (see Table 6.10.1) and doses and dose rates to the critical off-site receptor were calculated assuming the highest site boundary atmospheric dispersion factors all occurred at the same location.
6.10.2 Vermont Yankee Atmospheric Dispersion Model
The long-term average atmospheric dispersion factors are computed for routine releases using AEOLUS-2 Computer Code (Reference B). AEOLUS-2 is based, in part, on the constant mean wind direction model discussed in Regulatory Guide 1.111 (Reference C). Since AEOLUS-2 is a straight-line steady-state model, site-specific recirculation correction factors were developed for each release pathway to adjust the AEOLUS-2 results to account for temporal variations of atmospheric transport and diffusion conditions. The applicable recirculation correction factors are listed in Table 6.10.3.
AEOLUS-2 produces the following average atmospheric dispersion factors for each location:
1. Undepleted X/Q dispersion factors for evaluating ground level concentrations of noble gases;
2. Depleted X/Q dispersion factors for evaluating ground level concentrations of radioiodines and other particulates;
3. Gamma X/Q dispersion factors for evaluating gamma dose rates from a sector averaged finite cloud (undepleted source); and
4. D/Q deposition factors for evaluating dry deposition of elemental radioiodines and other particulates.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 40 of 55 Vermont Yankee Nuclear Power Station The North Warehouse depleted XlQ and D/Q factors were derived using the plume depletion and deposition curves provided in Regulatory Guide 1.111. However, because the Regulatory Guide 1.111 depletion and deposition curves are limited to an effective release height of 100 meters or less and the Vermont Yankee Stack effective release height (stack height plus plume rise) can exceed 100 meters, the Stack depleted XlQ and D/Q factors were derived using the deposition velocity concept presented in "Meteorology and Atomic Energy - 1968" (Reference E, Section 5-3.2) , assuming a constant deposition velocity of 1 ern/sec.
Gamma dose rate is calculated throughout this aDeM using the finite cloud model presented in "Meteorology and Atomic Energy - 1968 (Reference E, Section 7 5.2.5). That model is implemented through the definition of an effective gamma atmospheric dispersion factor, [XlQY) (Reference B, Section 4), and the replacement of XlQ in infinite cloud dose equations by the [XlQY).
6.10.3 Long- Term Average Atmospheric Dispersion Factors for Receptors
Actual measured meteorological data for the five-year period, 2002 through 2006, were analyzed to determine all the values and locations of the maximum off-site long-term average atmospheric dispersion factors. Each dose and dose rate calculation incorporates the maximum applicable off-site long-term average atmospheric dispersion factor. The values used and their locations are summarized in Table 6.10.1. Table 6.10.1 also indicates which atmospheric dispersion factors are used to calculate the various doses or dose rates of interest.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 41 of 55 Vermont Yankee Nuclear Power Station TABLE 6.10.1
Atmospheric Dispersion Factors
Release Dispersion Dose to Individual Dose to Air Pathway Factor Total Body Skin Critical Organ Gamma Beta
XJQ Depleted - - 1.25E-06 - - (sec/m ') (2150m WNW)
XJQ Undepleted - 1.59E-06 - - 1.59E-06 (sec/nr') (26500m SW) (2650m SW) Stack D/Q - - 1.25E-08 - - (11m2) (2150m WNW) XJQY 7.51E-07 7.51E-07 - 7.51E-07 - (sec/m ') (700m SSE) (700m SSE) (700m SSE)
XJQ Depleted - - 3.44E-05 - - (sec/m ') (451mENE)
XJQ Undepleted - 3.65E-05 - - 3.65E-05 North (sec/rrr') (451m ENE) (451m ENE) Warehouse D/Q - - 6.40E-08 - - 2 (1/m ) (357m S)
XJQY 7.37E-06 7.37E-06 - 7.37E-06 - (sec/m ') (451mENE) (451m ENE) (451mENE)
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 42 of 55 Vermont Yankee Nuclear Power Station TABLE 6.10.2
Site Boundary Distances
Downwind Stack North Warehouse Sector Releases Releases
N 400m 459m NNE 350m 417 m NE 350m 417 m ENE 400m 451 m E 500m 570m ESE 700m 561 m SE 750m 612 m SSE 850m 663 m S 385 m 357m SSW 300m 238m SW 250m 213 m WSW 250m 213 m W 300m 221 m WNW 400m 281 m NW 550m 697 m NNW 550m 680m
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 43 of 55 Vermont Yankee Nuclear Power Station TABLE 6.10.3
Recirculation Correction Factors
A. Stack Releases
Sector 0.5 Mi 1.5 Mi 2.5 Mi 3.5 Mi 4.5 Mi 7.5 Mi
N 1.4 1.4 1.2 1.1 1.0 1.0 NNE 1.8 1.8 1.4 1.2 1.0 1.0 NE 1.8 1.8 1.3 1.1 1.0 1.0 ENE 2.1 2.1 1.4 1.2 1.0 1.0 E 1.7 1.7 1.2 1.0 1.0 1.0 ESE 1.5 1.5 1.3 1.1 1.0 1.0 SE 1.8 1.8 1.3 1.2 1.1 1.0 SSE 1.4 1.4 1.2 1.2 1.2 1.2 S 1.3 1.3 1.1 1.1 1.2 1.2 SSW 1.8 1.8 1.5 1.4 1.4 1.2 SW 2.1 2.1 1.7 1.6 1.4 1.1 WSW 2.4 2.4 1.9 1.6 1.5 1.1 W 1.8 1.8 1.5 1.4 1.3 1.0 WNW 1.8 1.8 1.7 1.5 1.4 1.3 NW 1.5 1.5 1.3 1.3 1.3 1.1 NNW 1.5 1.5 1.2 1.2 1.1 1.1
B. North Warehouse Release
Sector 0.5 Mi 1.5 Mi 2.5 Mi 3.5 Mi 4.5 Mi 7.5 Mi
N 1.1 1.1 1.1 1.1 1.1 1.0 NNE 1.2 1.2 1.2 1.1 1.1 1.0 NE 1.1 1.2 1.1 1.1 1.0 1.0 ENE 1.2 1.3 1.4 1.4 1.4 1.3 E 1.1 1.3 1.4 1.4 1.4 1.2 ESE 1.1 1.1 1.2 1.1 1.1 1.0 SE 1.0 1.1 1.1 1.1 1.1 1.1 SSE 1.2 1.2 1.2 1.2 1.2 1.2 S 1.0 1.0 1.0 1.0 1.0 1.0 SSW 1.0 1.1 1.0 1.0 1.0 1.0 SW 1.2 1.3 1.2 1.0 1.0 1.0 WSW 1.1 1.1 1.0 1.0 1.0 1.0 W 1.2 1.2 1.1 1.0 1.0 1.0 WNW 1.2 1.4 1.3 1.2 1.2 1.0 NW 1.1 1.1 1.0 1.0 1.0 1.0 NNW 1.1 1.2 1.2 1.2 1.2 1.1
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 44 of 55 Vermont Yankee Nuclear Power Station 6.11 Method to Calculate Direct Dose From Plant Operation
Effluent Control 3.4.1 (40CFRI90) restricts the dose to the whole body or any organ to any member of the public from all station sources (including direct radiation from fixed sources on-site) to 25 rnrem in a calendar year (except the thyroid, which is limited to 75 rnrem).
6.11.1 Turbine Building
The maximum contribution of direct dose to the whole body or to any organ due to N-16 decay from the turbine is:
, = K N-16 K,-: Ks; DMSLRM (6-27a) where:
Dose contribution (rnrem) from NI6 decay to the site boundary critical receptor (West sector),
K'N-16 = 1.42E-05 [unitless correlation factor between the site boundary exposure (mR) and the time integral of the average MSLRM readings (mR)],
Ktissue = 0.60 [conversion factor, from radiation exposure in air to tissue dose (rnrernlrnR)] ,
KCa1ib = 508.8 (rnRlhr) / [Average MSLRM reading (rnRlhr) for a calibration source of 500 rnRIhr during the exposure interval of interest]
[Note: MSLRM calibrations would normally fall within the time interval of interest. In such cases, one approach would be to calculate the time integrals of the MSLRM readings separately for the periods before and after calibration, and then adjust each result by the corresponding calibration correction factor. A second (less complicated, though more conservative) approach would be to use the highest Kcalib factor during the interval for both the pre- and post-calibration periods.]
DMSLRM = Time integral of the MSLRM average reading (mR) during the interval LlT.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 45 of 55 Vermont Yankee Nuclear Power Station The last variable is defined as:
11 m
DMSLRM I[I (R),;)/ mJ;!J.ti (6-27b) i=1 )=1 where
R·J,1 = MSLRM reading (rnRlhr) by radiation monitor j at sequential time step ti,assumed to be constant during the time subinterval ~ti (as defined below),
m = Number of functioning MSLRMs during the subinterval ~ti, = Time subinterval (hr), within ~T, defined to span the interval between t, and the next monitor reading, i.e.:
= (6-27c) n = Number of subintervals ~ti within ~T, such that
~T = (6-27d)
During plant shutdown conditions, zero MSLRM readings should be assigned since there is no NI6 production in the core. As a conservative alternative, actual readings (reflecting relatively-low background radiation) may be utilized. Also, archived monitor readings which get accidentally lost should be assigned estimated values; a single monitor reading may be provided for this, for any single time interval spanning the entire period with continuous loss in the archived data base.
The dose computed using the above equations may be conservatively applied to the nearest residence. However, it may be corrected for occupancy time (by multiplying the dose by the fraction of time typically spent by the resident at the location during the period of interest), if documented. Dose reduction by the shielding provided by the residential structure may also be considered.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 46 of 55 Vermont Yankee Nuclear Power Station 6.11.2 North Warehouse
Radioactive materials and low level waste can be stored in the North Warehouse. The maximum annual dose contributions to off-site receptors (west site boundary line) from sources in the shielded (east) end and the unshielded (west) end of the North Warehouse are:
Ds = 0.25 x R s for the shielded end (6-28) mrem/yr) ("::mJ ( mrem/hr and
D-u - 0.53 x R u for the unshielded end (6-29) mrem/yr) ("::mJ ( mrem/hr where:
Ds = The annual dose contribution at the maximum site boundary location from fixed sources of radiation stored in the shielded east end of the North
Warehouse ( "::m J.
Du = The annual dose contribution at the maximum site boundary location from fixed sources of radiation stored in the unshielded west end of the North Warehouse (nrr;m}
R s = Dose rate measured at 1 meter from the source in the shielded end of the m north warehouse ( u:: ).
Ru = Dose rate measured at 1 meter from the source in the unshielded end of the m north warehouse ( u:: ).
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 47 of 55 Vermont Yankee Nuclear Power Station 0.25 = Dose rate to dose conversion factor which relates mrem/yr at the west site boundary per mrern/hr measured at 1 meter from the source in the shielded end of the warehouse assuming it is full to capacity for one year (:::;}
0.53 = Dose rate to dose conversion factor which relates mrem/yr at the west site boundary per mrern/hr measured at 1 meter from the source in the unshielded end of the warehouse assuming it is full to capacity for one year (:::;}
6.11.3 Low Level Waste Storage Pad
Interim storage of packaged Dry Active Waste (DA W) and spent ion exchange and filter media is permitted in modular concrete storage overpacks on the LLW storage pad facility adjacent to the north warehouse. The arrangement of the storage modules is such that DA W is placed in modules which shield higher activity ion exchange media from the west site boundary. The dose at the maximum site boundary receptor from both direct radiation and skyshine scatter can be calculated as follows:
(a) Direct Dose (line of sight)
DdE 0.28 x Rd X fd
(6-30) mrem mrem/yrJ (#) J ( ( yr-rnodule mremlhr
or
Dds 0.39 x Rd fd
(6-31) mrem mrem/yrJ J ( ( yr-rnodule mremfhr
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 48 of 55 Vermont Yankee Nuclear Power Station where:
DdE = The annual direct dose contribution at the maximum site boundary from a single rectangular storage module which has an unobstructed short end surface (not shielded by other modules) orientated toward
the west site boundary ( mrem ) . yr- module
DdS = The annual direct dose contribution at the maximum site boundary from a single rectangular storage module which has an unobstructed long side surface (not shielded by other modules) orientated toward
the west site boundary ( mrem ) . yr- module
Rei = Maximum dose rate measured at 3 feet from the side of the storage module whose unobstructed face (i.e., a side or end surface which is not shielded by other waste modules) is toward the west site boundary.
fd = The fraction of a year that a storage module is in use on the storage pad.
0.28 = Dose rate to dose conversion factor which relates mremlyr at the west site boundary per mrem/hr measured at 3 feet from the narrow end of the rectangular storage module when that face is orientated toward the west boundary.
0.39 = Dose rate to dose conversion factor which relates mremlyr at the west site boundary per mrem/hr measured at 3 feet from the long side of the rectangular storage module when that face is orientated toward the west boundary.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 49 of 55 Vermont Yankee Nuclear Power Station (b) Scatter From Skyshine
(6-32) DSKR = 0.016 X RSKR X fsk
mrem ) mremlyr) (#) ( ( yr-Iiner mremlhr
and
(6-33) DSKD = 0.015 X RSKD X fSK
rnrem ) rnremlyr) (#) ( ( yr- module rnrernJhr
where:
RSKR = The annual skyshine scatter contribution to the dose at the maximum site boundary from a single spent ion exchange media liner in a storage module whose top surface is not obstructed due to
stacking of modules ( rnr~m ). yr-Iiner
RSKD = The annual skyshine scatter contribution to the dose at the maximum site boundary from a rectangular storage module containing DA W whose top surface is not obstructed due to stacking of modules ( rnrem J. yr- module
RSKR = For Resins, the maximum dose rate measured at 3 feet over the top of each liner in a storage module (rnrernJhr).
RSKD = For DAW, the maximum dose rate measured at 3 feet over the top surface of a storage module with DAW (rnrernJhr).
fSK = The fraction of a year that a storage module is in use on the storage pad.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 50 of 55 Vermont Yankee Nuclear Power Station 0.016 = Dose rate to dose conversion factor for the scatter dose from each resin liner source in storage which relates rnrem/yr at the west site boundary per mremlhour at 3 feet from the top of the module.
0.015 = Dose rate to dose conversion factor for the scatter dose from DAW boxes in storage which relates rnrem/yr at the west site boundary per rnremlhr at 3 feet from the top of the module.
(c) Dose From Resin Liners During Transfer
During the movement of resin liners from transfer casks to the storage modules, the liners will be unshielded in the storage pad area for a short period of time. The maximum dose contribution at the site boundary during the unshielded movement of resin liners can be calculated from:
Dtrans = 0.0025 x R tran x T trans (6-34) (rnrem) rnremlhr) ( rad) (hr) ( radlhr hr where:
Dtrans The dose contribution to maximum site boundary resulting from the unshielded movement of resin liners between a transfer cask and a storage module (rnrem).
~:rans = Dose rate measured at contact (2") from the unshielded top surface of the resin liner in RJhr.
Ttran = The time (in hours) that an unshielded resin liner is exposed in the storage pad area.
0.0025 = The dose rate to dose conversion factor for an unshielded resin liner which relates rnremlhour at the west site boundary per radlhr at contact (2") from the unshielded surface of the liner.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 51 of 55 Vermont Yankee Nuclear Power Station (d) Intermodular Gap Dose
In addition to the above methods for determining doses at the west site boundary from the LLW storage pad, another dose assessment model has been included to address the possible condition of spaces or gaps existing between the placement of the DAW storage modules situated along the west facing side of the pad. This could result in a radiation streaming condition existing if ion exchange resin liners were placed in storage directly behind the gap. The direct dose equations (6-30 and 6-31) consider that the storage modules situated on the outside of the pad area provide a uniform shield to storage modules placed behind them. The intermodular gap dose equation (6-35) accounts for any physical spacing between the outside storage modules which have not been covered by additional external shielding.
(6-35) Doap = 2.44 E- 2 X WOap x ARL X fOap
(in) (Ci) (# ) mrem ) (mrem) ( yr yr- in- Ci
where:
DGap = The annual dose contribution at the maximum site boundary (west) from radiation streaming through the intermodular gap between DAW storage modules used to shield resin modules from direct radiation (mremlyr).
WGap = The intermodular gap width (inches) between adjacent DAW storage modules facing the west site boundary.
ARL = The total gamma activity contained in a condensate resin liner stored directly in line with the intermodular gap adjacent DAW modules ( Ci).
fGap = The fraction of a year that the intermodular gap is not shielded.
2.44E-2 = The activity to site boundary dose conversion factor for a one-inch wide intermodular gap ( ~em. J. yr-rn- ci
The site boundary dose from waste materials placed into storage on the Low Level Waste Storage Pad Facility is determined by combining the dose contribution due to direct radiation (line of sight) from Part (a) above with the skyshine scatter dose from Part (b), resin liner transfer dose from Part (c), and any intermodular gap dose from Part (d). Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 52 of 55 Vermont Yankee Nuclear Power Station 6.11.4 Independent Spent Fuel Storage Installation Dose Contribution
The Independent Spent Fuel Storage Installation (ISFSI) has been constructed to provide a secure location for the long term storage of spent nuclear fuel bundles. This installation is located just north of the Radwaste Building and just east of the North Warehouse within the Vermont Yankee Protected Area. The vendor, Holtec International, has prepared a study report to satisfy the requirements of 10CFR72.104 and this document is included as Reference H. During the first fuel storage evolution, five casks will be located on the installation pad. Over subsequent years, up to 35 casks will be stored on the pad.
The shielding analysis of the Independent Spent Fuel Storage Installation is provided in Reference H. The report analyzes the dose generated from a single cask as well as the dose generated from 5 casks for both the first row (266 meters to the west site boundary) and the last row (300 meters from the west site boundary) on the ISFSI pad. The dose from each cask to the west site boundary (DR-53 Location) is computed using the following equation:
(a) Direct Dose From Each Cask at 266 Meters Distance to the West Site Boundary
Dec-266m = 3.19E-5 x H (6-36)
[mrem/cask] [mrem/hour ] [hours per period]
where:
Dec-266m = Direct Dose calculated at the West Site Boundary where the cask is located at the west end of the pad approximately 266 meters from the DR-53 Location
3.19E-5 = Dose rate (hourly) from each cask calculated to the DR-53 Location
H = Number of hours in the period of interest (168 hours per week, 720 or 744 hours per month and 2208 hours per standard quarter)
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 53 of 55 Vermont Yankee Nuclear Power Station (b) Direct Dose From Each Cask at 300 Meters Distance to the West Site Boundary
Dec-300m = 2.05E-5 x H (6.37)
[mremlcask] [mrem/hour] [hours per period]
where:
Dec-300m = Direct Dose calculated at the West Site Boundary where the cask is located at the west end of the pad approximately 300 meters from the DR-53 Location
2.05E-5 = Dose rate (hourly) from each cask calculated to the DR-53 Location H = Number of hours in the period of interest (168 hours per week, 720 or 744 hours per month and 2208 hours per standard quarter)
(c) Dose From All Casks on the ISFSI Pad
The total dose at the west site boundary (DR-53 Location) for all casks located on the ISFSI Pad during a period of interest is calculated as follows:
Dac-Total = (Dec-266m x N(266m)) + (Dee-300m X N(300m») (6-38)
[mremlall casks] [mremlperiod] [casks per period] [mremlperiod] [casks per period]
where:
Dac-Total = Dose (mrem) at the West Site Boundary (DR-53 Location) from all Casks located on the ISFSI pad during the period of interest
Dec-266m = Direct Dose calculated at the West Site Boundary from cask(s) located at the west end of the pad approximately 266 meters from the DR-53 Location
N(266m) = Number of casks located at the west end of the pad approximately 266 meters from the DR-53 Location during the period of interest
Dec-300m = Direct Dose calculated at the West Site Boundary from casks located at the west end of the pad approximately 300 meters from the DR-53 Location
N(300m) = Number of casks located at the east end of the pad approximately 300 meters from the DR-53 Location during the period of interest Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 54 of 55 Vermont Yankee Nuclear Power Station 6.11.4 Total Direct Dose Summary
The dose contributions from the N-16 source in the Turbine Building, fixed sources in the North Warehouse, and fixed sources on the Low Level Waste Storage Pad Facility, shall be combined to obtain the estimate of total off-site dose to any member of the public from all fixed sources of radiation located on-site.
6.11.5 Other Fixed Sources
In addition to the fixed sources noted above (Turbine Building, North Warehouse, and LLW Storage Pad), other identified temporary or fixed sources that are created due to plant operations will be included in the total direct summary of 6.11.4 if the projected annual dose contribution would add any notable addition to the reported total (i.e., ;:::0.1mremlyr).
In 1995, turbine rotors and casings were replaced in the Turbine Hall with the old rotors and casings placed in storage sheds located on site west of the switch yard along the railroad spur. Radiation surveys (December 1995) of low level contamination (principally Co-60) on the components led to a projected maximum west site boundary dose of 0.2 mremlyr. This contribution will be added to the maximum site boundary total dose until the contribution is less than 0.1 mremlyr, or the components are removed from storage location.
6.12 Cumulative Doses
Cumulative Doses for a calendar quarter and a calendar year must be maintained to demonstrate a compliance with Controls 3.2.2, 3.3.2, and 3.3.3 (lOCFR50, Appendix I dose objectives). In addition, if the requirements of the Action Statement of Control 3.4.1 dictate, cumulative doses over a calendar year must be determined (demonstration of compliance with total dose, including direct radiation per requirements of 40CFR 190). To ensure the limits are not exceeded, a running total must be kept for each release.
Demonstration of compliance with the dose limits of 40CFR190 is considered as demonstrating compliance with the 0.1 rem limit of 10CFR20.1301(a)(1) for members of the public in unrestricted areas.
Off-Site Dose Calculation Manual Section 6 Rev. 32 Page 55 of 55 Vermont Yankee Nuclear Power Station :-. IPaMVRPOIP_ AN ORPM_RVMPL SVRLVAO
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Figure 9-1 shows the normal (design) radioactive liquid effluent streams, radiation monitors, and the appropriate Liquid Radwaste Treatment System. Figure 9-2 shows the normal (design) gaseous effluent systems, radiation monitors, and the appropriate Gaseous Radwaste Treatment System.
9.1 In-Plant Radioactive Liauid Effluent Pathways I The Liquid Radwaste System collects, processes, stores, and disposes of all radioactive liquid wastes. Except for the cleanup phase separator equipment, the condensate backwash receiving tank and pump and waste sample tanks, floor drain sample tank and waste surge tank, the entire Radwaste System is located in the Radwaste Building. The Radwaste System is controlled from a panel in the Radwaste Building Control Room.
The Liquid Radwaste System consists of the following components:
1. Floor and equipment drain system for handling potentially radioactive wastes.
2. Tanks, piping, pumps, process equipment, instrumentation and auxiliaries necessary to collect, process, store, and dispose of potentially radioactive wastes.
The liquid radwastes are classified, collected, and treated as either high purity, low purity, chemical or detergent wastes. "High" purity and "low" purity mean that the wastes have low conductivity and high conductivity, respectively. The purity designation is not a measure of the amount of radioactivity in the wastes.
High purity liquid wastes are collected in the 25,000-gallon waste collector tank. They originate from the following sources:
1. Drywell equipment drains. 2. Reactor Building equipment drains. 3. Radwaste Building equipment drains. 4. Turbine Building equipment drains. 5. Decanted liquids from cleanup phase separators. 6, Decanted liquids from condensate phase separators, 7. Resin rinse.
Revision 27 Date 10/09/00 9- 1 Low purity liquid wastes are collected in the 25,000-gallon floor drain collector tank. They originate from the following sources:
1. Drywell floor drains. 2. Reactor Building floor drains. 3. Radwaste Building floor drains. 4. Turbine Building foor drains. 5. Other floor drains in RCA (e.g., AOG and Service Building, stack, etc.).
Chemical wastes are collected in the 4,000-gallon chemical waste tank and then pumped to the floor drain collector tank. Chemical wastes arise from the chemical laboratory sinks, the laboratory drains and sample sinks. Radioactive decontamination solutions are classified as detergent waste and collected in the 1,000-gallon detergent waste tank.
Once the wastes are collected in their respective waste tanks, they are processed in the most efficient manner and discharged or reused in the nuclear system. From the waste collector tank, the high purity wastes are processed in one of three alternative filter demineralizers and then, if needed, in one "polishing" demineralizer. After processing, the liquid is pumped to a waste sample tank for testing and then recycled for additional processing, transferred to the condensate storage tank for reuse in the nuclear system or discharged.
The low purity liquid wastes are normally processed through the floor drain filter demineralizer and collected in the floor drain sample tank for discharge or they are combined with high purity wastes and processed as high purity wastes.
Chemical wastes are neutralized and combined with low purity wastes for processing as low purity wastes.
Although there is only one discharge pathway from the Radwaste System to the river, there are three locations within the Radwaste System from which releases can be made. They are: the detergent waste tank (detergent wastes), the floor drain sample tank (chemical and low purity wastes), and waste sample tank (high purity wastes). The contents of any of these tanks can be released directly to the river.
The liquid wastes collected in the tanks are handled on a batch basis. The tanks are sampIed from the radwaste sample sink and the contents analyzed for radioactivity and water purity. A release is allowed once it is determined that the activity in the liquid wastes will not exceed Control release limits. I
Revision 27 Date 10/09/00 9-2 A discharge from any of the tanks is accomplished by first starting the sample pumps, opening the necessary valves, and positioning the flow controller. The release rate in the discharge line is set between 0 and 50 gpm. The dilution pumps which supply 20,000 gpm of dilution water are then started. An interlock does not allow discharge to the river when dilution water is unavailable.
The effluent monitor (No. 171350) in the discharge line provides an additional check during the release. The alarm or trip setpoint on the monitor is set according to the effluent Control limits and an analysis of the contents of the tank. The monitor warns the operator if the activity of the liquid waste approaches regulatory limits. In response to a warning signal from the monitor, the operator may reduce the flow rate or stop the discharge.
Revision 27 Date 10/09/00 9-3 9.2 In-Plant Radioactive Gaseous Effluent Pathwavs
The gaseous radwaste system includes subsystems that dispose of gases from the main condenser air ejectors, the startup vacuum pump, the gland seal condenser, the standby gas treatment system and station ventilation exhausts.
The processed gases are routed to the plant stack for dilution and elevated release to the atmosphere.
The plant stack provides an elevated release point for the release of waste gases. Stack drainage is routed to the liquid radwaste collection system through loop seals.
The air ejector Advanced Off-Gas Subsystem (AOG) reduces the ejector radioactive gaseous release rates to the atmosphere. The AOG System consists of a hydrogen dilution and recombiner subsystem, a dual moisture removalldryer subsystem, a single charcoal absorber subsystem, and dual vacuum pumps. Equipment is located in shielded compartments to minimize the exposure of maintenance personnel.
Radioactive releases from the air ejector off-gas system consist of fission product noble gases, activation product gases, halogens, and particulate daughter products from the noble gases. The particulates and halogens are effectively removed by the charcoal beds and high efficiency particulate filters in the AOG System. The activation product gases that are generated in significant quantities have very short half-lives and will decay to low levels in the holdup pipe, as well as in the absorber beds. The noble gases, therefore, are expected to provide the only significant contribution to off-site dose. The charcoal off-gas system is designed to provide holdup of 24 hours for krypton and 16.6 days for xenon at a condenser air inleakage rate of 30 scfm.
Steam dilution, process control, and instrumentation systems are designed to prevent an explosive mixture of hydrogen from propagating beyond the air ejector stages. An explosive mixture of hydrogen should never exist in the recombiner subsystem, "30-minute" delay pipe, condenser/dryer, or charcoal absorber beds. To prevent a hydrogen explosion in the recombinerlpreheater and upstream lines during shutdown, the residual off-gas steam mixture containing hydrogen is purged with steam or air. Starting procedures insure sufficient steam is introduced upstream of the preheater to dilute any hydrogen entering the AOG System as the air
Revision 27 Date l0/09/00 9-4 ejector line is prepared for operation. To prevent operating unsafely, instrumentation is used to detect an explosive mixture.
Hydrogen control is accomplished by providing redundant hydrogen analyzers on the outlet from the Recornbiner System. These analyzers initiate recombiner system shutdown and switchover if the hydrogen concentration at the system outlet exceeds 2% by volume. During an automatic shutdown, two main air process valves close to isolate the recombiner system. Additionally, the recombiner bed temperatures and recombiner outlet temperature provide information about recombiner performance to insure that inflammable hydrogen mixtures do not go beyond the recombiner.
Should a number of unlikely events occur, it would be hypothetically possible for a hydrogen explosion to occur in the off-gas system. Such an explosion within the recombiner system could propagate into the large "30-minute" delay pipe, through the condenserldryer subsystem, and into the charcoal absorber tanks. However, the recombinerladsorber subsystems, piping, and vessels are designed to withstand hydrogen detonation pressures of 500 psi at a minimum so that no loss of integrity would result. Furthermore, the seven tanks of charcoal would significantly attenuate a detonation shock wave and prevent darnage to the downstream equipment.
During normal operation, the dryerladsorber subsystem may be bypassed if it becomes unavailable provided the releases are within effluent Control limits. With the dryerladsorber I subsystem bypassed, the air ejector off-gas exhausts through the recombinerlcondenser subsystems, and the 30-minute delay pipe.
The off-gas mixture combines with steam at the air ejector stage to prevent an inflammable hydrogen mixture of 4%by volume from entering the downstream hydrogen recombiners. Approximately 6,400 lbhr of steam introduced at the second stage air ejector reduces the concentration of hydrogen to less than 3% by volume.
The recombiner subsystem consists of a single path leading from the hydrogen dilution steam jet ejectors to two parallel flow paths for hydrogen recombination. Each recombination subsystem is capable of operating independently of the other and each is capable of handling the condenser off-gas at a startup design flow of 1,600 lblhr air and the normal off-gas design flow rate of 370 lblhr. The major components of each recombiner flow path are a preheater, a h ydrogen-oxygen recombiner, and a desuperheating condenser.
Revision 27 Date 10l09l00 9-5 The preheater assures that the vapor entering the hydrogen-oxygen recombiner is heated to approximately 300OF. At this temperature, the water vapor in the stream becomes superheated steam, thereby, protecting the recombiner catalyst.
During passages through the recombiner, the recombination of H2 and 0, in an exothermic reaction increases the stream temperature to approximately 52OOF. This recombination results in a maximum effluent H2 concentration of 0.1% by volume.
The desuperheating condenser is designed to remove the heat of recombination and condense the steam from the remaining off-gas. The condensers discharge the off-gas through moisture separators into the initial portion of an underground 24-inch diameter delay pipe which allows for 40% of the total system holdup volume. The pipe slopes away from the off-gas particulate (HEPA) filters in both directions for drainage purposes. Loop seals prevent gas escaping through drainage connections. Shorter lived radionuclides undergo a substantial decrease in activity in this section of the system. The preheatersh-ecombinersoperate at pressures slightly above atmospheric; the condenser and the subsystems that follow operate at subatmospheric pressures.
Particulate (mPA) filters with flame suppressant prefilters are located at the exit side of the delay pipe ahead of the moisture removal subsystem to remove radioactive particulates generated in the delay pipe.
In the moisture removalidryer subsystem, the moisture of the gas is reduced to incfease the effectiveness of the charcoal absorber beds downstream. The subsystem consists of two parallel cooling condensers and gas dryer units. Each condenser is cooled by a mechanical glycollwater refrigeration system that cools the off-gas to -40OF as it removes bulk moisture. The dryer is designed to remove the remaining moisture by a molecular sieve desiccant to a dew point of less than -40°F (1% RH). One of the dryers absorbs moisture from the off-gas; the other desorbs moisture by circulating heated air through the bed in closed cycle.
The mixed refrigerantldryer concept improves the reliability of the system. If the refrigerant system fails, the two dryer beds operate in parallel to remove the moisture and maintain the off-gas near the design dew point (-4OOF). If the dryer fails, the -40°F dew point air leaving the mechanical system can enter the guard bed for over 6 hours without affecting the performance of the charcoal beds downstream.
Revision 27 Date 10/09/00 9-6 The charcoal absorber subsystem consists of seven tanks of charcoal preceded by a smaller charcoal guard bed upstream. The guard bed protects the seven main tanks from excessive radioactivity levels or moisture in the event of a malfunction upstream in the moisture removal subsystem. The guard bed also removes compounds which might hinder noble gas delay, The seven tanks hold a minimum of approximately 90,OOO pounds of charcoal.
The first two main tanks can be bypassed and used for storing a "batch of high activity" gas for static decay. The remaining five are all in series with no bypassing features so that the off-gas to the stack must be delayed.
Redundant particulate (HEPA) after-filters are used to remove charcoal fines prior to the vacuum pumps.
A water-sealed vacuum pump boosts the gas stream pressure to slightly over-atmospheric pressure before it is vented through the stack. To assure maintaining constant operating pressures in the system, a modulating bypass valve will recirculate process gas around the pump as required. During periods of high flow rates, both pumps can be operated in parallel.
Discharge of the vacuum pump then passes through the remaining 60% of the delay pipe prior to being vented through the station stack.
The gland seal off-gas subsystem collects gases from the gland seal condenser and the mechanical vacuum pump and passes them through a charcoal filter (if required) and then through holdup piping prior to release to the stack. The gases from the gland seal condenser system are discharged to the atmosphere via the ventilation stack after passing through the filter for iodine removal (if required) and then through the same 1-314 minute holdup piping that is used for the startup vacuum pump system. One automatic valve on the discharge side of each steam packing exhauster closes upon the receipt of high level radiation signal from the main steam line radiation monitoring subsystem to prevent the release of excessive radioactive material to the atmosphere. The exhausters are shut down at the same time the valves close. In addition, the mechanical vacuum pump is automatically isolated and stopped by a main steam line high radiation signal. The filter assembly is located in the air ejector room.
The release of significant quantities of gaseous and particulate radioactive material is prevented by the combination of the design of the air ejector AOG system and automatic isolation of the system from the stack. Gas flow from the main condenser stops when the air
Revision 27 Date 10/09/00 9-7 ejectors are automatically isolated from the main condenser by either a high radiation signal in the main steam line or by high temperature and/or pressure signals from the AOG System. The gland seal off-gas system is automatically isolated and stopped by a main steam line high radiation signal. In addition, monitoring the stack release provides a backup warning of abnormal conditions.
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~." _bthnm~s" t_idm tn rdstnrd tgd hmnodr_akd dpuholdmt tn nodr_akd st_tus: _mc
~/" sull_ry cdsbrhothnm ne _bthnm~s" t_idm tn ordvdmt _ rdburrdmbd) Nee'Rhtd Ansd =_kbuk_thnm L_mu_k Rdbthnm!~ Pdv) .3 O_fd 0 ne1 Tdrlnmt Y_midd Mubkd_r Onwdr Rt_thnm 4//;.8 Jateovt Efflvenut ~Conusolt 0)0)/& 0)0)0&:.:/; and 0)0)3"
Yiuh uhe calcvlaued ais dote fsom sadioacuiwe noble gatet in gateovt efflvenut eyceeding anz of uhe limiut of Conusol 0)0)/& psepase and tvbmiu uo uhe Committion xiuhin 0- dazt a tpecial seposu xhich idenuifiet uhe cavte~t" fos eyceeding uhe limiu~t" and uhe cossecuiwe acuion~t" uaken uo attvse uhau tvbtervenu seleatet xill be in compliance xiuh uhe limiut of Conusol 0)0)/) Yiuh uhe calcvlaued dote fsom uhe seleate of Lodine'.0.& Lodine'.00& usiuivm& and,os sadionvclidet in pasuicvlaue fosm eyceeding anz of uhe limiut of Conusol 0)0)0& psepase and tvbmiu uo uhe Committion xiuhin 0- dazt a tpecial seposu xhich idenuifiet uhe cavte~t" fos eyceeding uhe limiu~t" and uhe cossecuiwe acuion~t" uaken uo attvse uhau tvbtervenu seleatet xill be in compliance xiuh uhe limiut of Conusol 0)0)0)
Yiuh gateovt sadxatue being ditchasged xiuhovu psocetting uhsovgh appsopsiaue useaumenu tztuemt at defined in Conusol:.:/; fos mose uhan tewen ~7" contecvuiwe dazt& os in eycett of uhe limiut of Conusol 0)0)3& psepase and tvbmiu uo uhe Committion xiuhin 0- dazt a tpecial seposu xhich inclvdet uhe folloxing infosmauion=
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~/" acuion~t" uaken uo setuose uhe inopesable ervipmenu uo opesable tuauvtA and
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4//;.: Toual Dote ~Conusol:/;.4~
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Off'Siue Dote Calcvlauion Manval Secuion .- Rew) /: Page 3 of7 Vesmonu _ankee Nvcleas Poxes Suauion /.,3,3 Vehnurulnger Lt nxutsitzer Putnzuxntl ~Iutzxur 1,7,/"
cnzm zmi ri ir uk xehnuegzn nz ey zmi xiy rz uk vretz ikkr itzy nt et it nxutsitzer yesvrntl sihne ez uti ux suxi uk zmi rugeznuty yvignknih nt Iutzxur _efri 1,7,/ i giihntl zmi xivuxzntl ri iry uk Iutzxur _efri 1,7,0' vxivexi eth y fsnz zu zmi Iussnyynut nzmnt 1. he y kxus zmi xiginvz uk zmi Oefuxezux Eter yiy e yvigner xivuxz mngm ntgr hiy et i er eznut uk et xirieyi guthnznuty' it nxutsitzer kegzuxy ux uzmix kegzuxy mngm ge yih zmi rnsnzy uk Iutzxur _efri 1,7,0 zu fi i giihih, _mny xivuxz ny tuz xiw nxih nk zmi siey xih ri ir uk xehnuegzn nz ey tuz zmi xiy rz uk vretz ikkr itzy' mu i ix' nt y gm et i itz' zmi guthnznut ymerr fi xivuxzih eth hiygxnfih nt zmi Ett er Vehnurulnger Lt nxutsitzer Svixezntl Vivuxz,
/.,3,7 Oeth ayi Iity y ~Iutzxur 1,7,0"
cnzm e reth yi gity y tuz fintl guth gzih ey xiw nxih f Iutzxur 1,7,0' vxivexi eth y fsnz zu zmi Iussnyynut nzmnt 1. he y e yvigner xivuxz mngm nhitznkniy zmi xieyuty m zmi y x i ey tuz guth gzih' eth mez yzivy exi fintl zepit zu guxxigz zmi ynz eznut,
/.,7 Peoux Imetliy zu Vehnuegzn i Onw nh' Neyiu y' eth Yurnh ceyzi _xiezsitz Y yzisy ~~
Ongityii)ntnznezih seoux gmetliy zu zmi xehnuegzn i eyzi y yzisy ~rnw nh' leyiu y' eth yurnh"C
E, Ymerr fi xivuxzih zu zmi gussnynut nt zmi Vehnuegzn i Lkkr itz Virieyi Vivuxz kux zmi vixnuh nt mngm zmi i er eznut ey xi ni ih f zmi Stynzi Yekiz Vi ni Iussnzzii ~SYVI", _mi hnyg yynut uk iegm gmetli ymerr gutzentC
/, E y ssex uk zmi i er eznut zmez rih zu zmi hizixsnteznut zmez zmi gmetli gu rh fi sehi nt egguxhetgi nzm rSIMV Texz 7.,7AD
0, Y kkngnitz hizenrih ntkuxseznut zu y vvuxz zmi xieyut kux zmi gmetli nzmu z fitiknz uk ehhnznuter ux y vvrisitzer ntkuxseznutD
1, E hizenrih hiygxnvznut uk zmi iw nvsitz' gusvutitzy' eth vxugiyyiy nt ur ih eth zmi ntzixkegiy nzm uzmix vretz y yzisyD
3, Et i er eznut uk zmi gmetli' mngm ymu y zmi vxihngzih xirieyiy uk xehnuegzn i sezixnery nt rnw nh eth leyiu y ikkr itzy eth-ux w etznz uk yurnh eyzi zmez hnkkix kxus zmuyi vxi nu yr vxihngzih nt zmi rngityi evvrngeznut eth esithsitzy zmixizuD
&&Ongityii se gmuuyi zu y fsnz zmi ntkuxseznut gerrih kux nt zmny xivuxzntl xiw nxisitz ey vexz uk zmi ett er MYEV vhezi, Skk)Ynzi Juyi Ierg reznut Pet er Yigznut /. Vi , 0= Teli : uk; bixsutz detpii R griex Tu ix Yzeznut /) :f VoPdnPmbgf gY maV SaPf_V& pabSa lagpl maV VrhVSmVT ePrbene VrhglnkVl mg eVeRVk~l" gY maV hnRdbS Pm maV lbmV RgnfTPks PfT mg maV _VfVkPd hghndPmbgf maPm TbYYVk Ykge maglV hkVobgnlds VlmbePmVT bf maV dbSVflV PhhdbSPmbgf PfT PeVfTeVfml maVkVmg7
0) : SgehPkblgf gY maV hkVTbSmVT kVdVPlVl gY kPTbgPSmboV ePmVkbPdl& bf dbinbT PfT _PlVgnl VYYdnVfml PfT bf lgdbT pPlmV& mg maV PSmnPd kVdVPlVl Ygk maV hVkbgT hkbgk mg paVf maV SaPf_Vl PkV mg RV ePTV7
1) :f VlmbePmV gY maV VrhglnkV mg hdPfm ghVkPmbf_ hVklgffVd Pl P kVlndm gY maV SaPf_V7 PfT
3) AgSneVfmPmbgf gY maV YPSm maPm maV SaPf_V pPl kVobVpVT PfT YgnfT PSSVhmPRdV Rs IML=)
;) MaPdd RVSgeV VYYVSmboV nhgf kVobVp PfT PSSVhmPfSV Rs IML= PfT PhhkgoPd Rs maV CVfVkPd DPfP_Vk)
IYY'MbmV AglV =PdSndPmbgf DPfnPd MVSmbgf -, LVo) .3 JP_V 1 gY 1 NVkegfm OPfcVV EnSdVPk JgpVk MmPmbgf REFERENCES
A. Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I," U.S. Nuclear Regulatory Commission, Revision 1, October 1977.
B. Hamawi, J. N., "AEOLUS-2 - A Computer Code for the Determination of Continuous and Intermittent-Release Atmospheric Dispersion and Deposition of Nuclear Power Plant Effluents in Open-terrain Sites, Coastal Sites, and Deep-River Valleys for Assessment of Ensuing doses and Finite-Cloud Gamma Radiation Exposures," Entech Engineering, Inc., P100R13A, March 1988 (Mod 5, Revised by Yankee Atomic Electric Company, March 1992).
C. Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," U.S. Nuclear Regulatory Commission, Rev. 1, July 1977.
D. National Bureau of Standards, "Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure," Handbook 69, June 5, 1959.
E. Slade, D. H., "Meteorology and Atomic Energy - 1968, USAEC, July 1968.
F. Lowder, W. M., P. D. Raft, and G. dePlanque Burke, "Determination of N-16 Gamma Radiation Fields at BWR Nuclear Power Stations," Health and Safety Laboratory, Energy Research and Development Administration, Report No. 305, May 1976.
G. Letter from Charles L. Miller of the United States Nuclear Regulatory Commission to John F. Schmidt of the Nuclear Energy Institute, dated December 26, 1995.
H. "Dose vs Distance at the Vermont Yankee ISFSI" Holtec Report No: HI-2073701, Holtec Project No: 90348, Approved on 12/10/07.
1. Certificate of Compliance, Holtec International, Final Safety Analysis Report, Amendment 2, 06/07/05.
J. AREVA Calculation #32-9068362-000 "Vermont Yankee Site Boundary Direct Dose From N-16 Methodology", April 4, 2008, Mark Strum and John Hamawi.
Off-Site Dose Calculation Manual Section R Rev. 29 Page I of I Vermont Yankee Nuclear Power Station APPENDIX B
Approval of Criteria for Disposal of S1 ightly Contaminated
Septic Waste On-Site at Vermont Yankee
d Revision -9- Date 3/2/90
E-1 UNITED STATES NUCLEAR REGULATORY COMMLSSION WASHINGTON. D, C 20555 August 30, 1989
Docket No. 50-271
Mr. L. A. Tremblay Licensing Engi neeri ng Vermont Yankee Nuclear Power Corporation Engi neeri ng Offi ce 580 Main Street Bolton, Massachusetts 01740-1398
Dear Mr. Tremblay:
SUBJECT: APPROVAL UNDER 10 CFR 20.302(a) OF PROCEDURES FOR DISPOSAL OF SLIGHTLY CONTAMINATED SEPTIC WASTE ON SITE AT VERMONT YANKEE fTAC NO. 73776)
REFERENCE: (a) June 28, 1989 letter from R. Ld. Capstick to US NRC Document Control Desk, including Attachment I and Attachment 11. (b) Final Environmental Statement related to the operation of Vermont Yankee Nuclear Power Station, dated July 1972.
In reference (a) Vermont Yankee Nuclear Power Corporation (Vermont Yankee, or the licensee) submitted an application for disposal of licensed material on site. This disposal was not previously considered by the staff in the Vermont Yankee Final Environmental Statement (FES), reference fb). This extensive application, prepared in accordance with 10 CFR 20.302(a), contains a detailed description of the licensed material, thoroughly analyzes and evaluates the information pertinent to the effects on the environment of the proposed disposal of the licensed material, and commits the licensee to follow specific procedures to minimize the risk of unexpected or hazardous exposures. In the FES, the NRC staff considered the potential effects on the environment of licensed material from operation of the plant and, in the assessment of the total radiological impact of the Vermont Yankee Station concluded that: "...operation of the Station will contribute only an extremely small increment to the radiation dose that area residents receive from natural background. Since fluctuations of the background dose may be expected to exceed the increment contributed by the plant, the dose will be immeasurable in itself and will constitute no meaningful risk to be balanced against the benefits of the plant."
Revision -9- Date 3/2/90
B-2 Mr. L. A. Tremblay -2- August 30, 1989
4 Since the disposal proposed by the licensee involves licensed material containing less than 0.1 percent of the radioactive materials, primarily cobalt-60 and cesium-137, already considered acceptable in the FES, and involves exposure pathways much less significant than those considered in the FES, we consider the site-specific application (Reference (a)) for Vermont Yankee Nuclear Power Station to have insignificant radiological impact. We accept the commitments and evaluations of the licensee, documented in reference (a), as further assurance that the proposed disposal procedures will have a negligible effect on the environment and the general population in comparison to normal background radiation.
LA1 In conclusion, we find the licensee’s procedures with commitments as 855 documented in reference (a) to be acceptable, provided that reference (a) is permanently incorporated into the licensee’s Offsite Dose Calculation Manual (ODCM) as an Appendix, and future modifications of reference (a) be re~orted to NRC in accordance with licensee commitments reqardinq ODCM chanqes.
Pursuant to 10 CFR 51.22(~)(9), no environmental assessment is required. This completes our -review under TAC No. 73776.
Sincerely ,
Morton 8. Fairtile, Project Manager Project Directorate 1-3 Division of Reactor Projects 1/11 Office of Nuclear Reactor Regulation
cc: See next page
I Revision 9 Date 3/2/90
8-3 Mr. L. A. Tremblay -3-
d cc : Mr. J. Gary Weigand G. Dean Weyman President & Chief Executive Officer Chairman, Board of Selectman Vermont Yankee Nuclear Power Corp. Post Office Box 116 R.D. 5, Box 169 Vernon, Vermont 05354 Ferry Road Brattleboro. Vermont 05301 Mr. Raymond N. McCandless Vermont Division of Occupational Mr. John DeVi ncenti s, Vice President and Radiological Health Yankee Atomic Electric Company Admi nist rat i on Bui 1 ding 580 Main Street Montpel ier, Vermont 05602 Bolton. Massachusetts 01740-1398 Honorable John J. Easton New England Coalition on Nuclear Attorney General Pol 1 ution State of Vermont Hill and Dale Farm 109 State Street R.D. 2, Box 223 Montpel ier, Vermont 05602 Putney, Vermont 05346 Conner & Wetterhahn, P.C. Vermont Publ ic Interest Research Suite 1050 Group, Inc. 1747 Pennsylvania Avenue, N.W. 43 State Street Washington, D.C. 20006 Montpelier, Vermont 05602 Diane Curran, Esq. Regional Admi nistrator, Regi on I Harmon. Curran & Tousley U.S. Nuclear Regulatory Commission 2001 S Street. N.W., Suite 430 - 475 Allendale Road Washington, D.C. 20009 King of Prussi a, Pennsyl vani a 19406 David J. Mullett, Esq. R. K. Gad, I11 Special Assistant Attorney General Ropes & Gray Vermont Department of Publ ic Service 225 Franklin Street 120 State Street Boston, Massachusetts 02110 Montpel ier, Vermont 05602
Mr. W. P. Murphy, Vice President, Jay Gutierrez and Manager of Operations Regional Counsel Vermont Yankee Nuclear Power Corporation U.S. Nuclear Regulatory Commission R.D. 5, Box 169 475 All.endale Road Ferry Road King of Prussi a, Pennsyl vani a 19406 Brattleboro, Vermont 05301 G. Dana Eisbee, Esq. Mr. George Sterzinger, Commissioner Office of the Attorney General Vermont Department of Publ ic Service Envi ronmental Protection Bureau 120 State Street, 3rd Floor State House Annex Montpelier, Vermont 05602 25 Capitol Street Concord, New Hampshire 03301-6397 Publ ic Service Board State of Vermont Atomic Safety and Licensing Board 120 State Street U. S. Nuclear Regulatory Commi ssi on Montpelier, Vermont 05602 Washington, D.C. 20555
.. , Revision 9 Date 3/2/90
8-4 Mr. L. A. Tremblay -4-
cc : Mr. Gustave A. Linenberger, Jr. Admini strative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Resident Inspector Vermont Yankee Nuclear Power Station U. S. Nuclear Regul atory Commi ssi on P.O. Box 176 Vernon, Vermont 05354
John Traficonte, Esq. Chief Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, Massachusetts 02108
Geoffrey M. Huntington, Esquire Office of the Attorney General Envi ronmental Protection Bureau State House Annex 25 Capitol Street - Concord, New Hampshire 03301-6397 Charles Bechhoefer, Esq. Administrative Judge Atomic Safety and Licensing Board U. S. Nucl ear Regul atory Commi ssi on Washington, D.C. 20555
Or. James H. Carpenter Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Adjudicatory File (2) Atomic Safety and Licensing Board Panel Docket U.S. Nuclear Regulatory Commission Washington, D.C. 20555
/ Revision -9- Date 3/2/90 ‘VERMONT YANKEE NUCLEAR POWER CURF’OELATION
Ferry Road. Brattlebom. VT 05301-7002 KCL. IO ENGINEERING OFFICE $80 LWNSTREET June 28, 1989 BOLTON. MA 01 ?# BVY 89-59 T706711
United States Nuclear Regulatory Commi ssion Washington, DC 20555
Attention: Document Control Desk
Reference: License No. DPR-28 (Docket No. 50-271).
Subject: Request to Routinely Dispose of Slightly Contaminated Septic Waste in Accordance with 10 CFR 20.302(a)
Dear Sir:
In accordance with the criteria of the Code of Federal Regulations, Title 10, Section 20.302(a) (10CFR20.302(a)), enclosed please find the subject application for the disposal of very low level radioactive waste materials. Vermont Yankee Nuclear Power Corporation (Vermont Yankee) hereby requests NRC approval of the proposed procedures for the disposal of slightly contaminated septic waste generated at the Vermont Yankee Nuclear Power Plant in Vernon, Vermont.
This appl icati on specifically requests approval to dispose of septic tank waste, contaminated at minimal levels, which have been or might be generated through the end of station operations at the Vermont Yankee Nuclear Power Plant. The proposed method of disposal is for the on-site land spreading in designated areas in compliance with State of Vermont health code requirements for septic waste. Disposal of this waste in the manner proposed, rather than at a 10 CFR Part 61 licensed facility would save Vermont Yankee not only substantial cost, but also valuable disposal site space which would then be avai 1 ab1 e for wastes of higher radioactivity 1 eve1 s. Disposal as radioactive waste would require treatment of the biological aspects of the septage and solidification to a stable waste form, thereby increasing the volume substanti a1 ly.
A radiological assessment and proposed operational controls, based upon the continued on-site disposal of septic waste as presently contained in the plant’s septic tanks, are detailed in Attachments 1 and 2. Based upon this analysis, Vermont Yankee requests approval to dispose of septic tank waste on-site by land spreading in such a manner that the radioactivity concentration limit in any batch of septage to be spread does not exceed one-tenth of the MPC values listed in 10 CFR 20, Appendix B. Table 11; and the combined radi ologi cal impact for a1 1 disposal operations shall be 1 imited to a total body or organ dose of a maximally exposed member of the public of less than one mremlyear (less than 5 mrem/year to an inadvertent intruder).
Revision -9- Date 3/2/90
B-6 United States Nuclear Regulatory Commi ssi on June 28, 1989 c Page 2
Due to our expected need to utilize the proposed methodology of land application of septic waste on-site during the spring of 1990, we request your review and approval of this proposed disposal method by the end of the first quarter of 1990.
We trust that the information contained in the submittal is sufficient; however, should you have any questions or require further information concerning this matter, please contact this office
Very truly yours,
VERMONT YANKEE NUCLEAR POWER CORPORATION
Robert W. Capstick, Jr. ti censi ng Engineer
/ MSSlemd
Enclosures
cc: USNRC - Region I USNRC - Resident Inspector, VTNPS
Revision -9- Date 3/2/90
8-7 ATTACHMENT 1
VERMONT YANKEE NUCLEAR POWER PLANT
APPLICATION FOR APPROVAL TO ROUTINELY DISPOSE OF
SEPTIC WASTE WITH M~NIMALLEVELS OF RADIOACTIVITY
- Revision -9- Date 3/2/90
B-8 ATTACHMENT 1
VERMONT YANKEE NUCLEAR POWER PLANT
Application for Approval to Routinely Dispose of Septic Waste With Minimal Levels of Radioactivity
1.O INTRODUCTION
Vermont Yankee Nuclear Power Corporation (Vermont Yankee) requests approval, pursuant to 10CFR20.302(a). of a method proposed herein for the routine disposal of slightly contaminated septic tank waste. Vermont Yankee proposes to dispose of this waste by spreading it on designated areas within the plant’s site boundary fence. This application addresses specific information requested in 10CFR20.302(a).
2 .O HASTE STREAM DESCRIPTION
The waste involved in this application consists of residual solids and water associated with the sewage collection system at Vermont Yankee. The - plant’s sewage systems are of the septic tank and disposal field type. The two systems servicing the majority of the plant’s sanitary waste are identified as (1) main septic system and (2) the south sewage disposal system.
The main septic system (design flow capacity 4,950 gallonsfday) consists of a wastewater lift station, septic tank, and dual alternating disposal fields located on the north side of the plant. This system services the main complex of buildings central to the plant and processes approximately 3,500 gallons of wastewater per day. The septic tank, shown in Figure (l), will typically contain 9,250 gallons of septage.
The south sewage disposal system is a newly-installed (January 1989) pressurized mound system, which is used in lieu of the construction office building (COB) holding tank that had previously serviced the lavatory facilities on the south end of the plant. The new system is composed of a septic tank (5,700 gallon capacity, see Figure 2). pumping station, and pressurized mound disposal field. When dosing the field. a force main pressurizes the disposal field’s piping system with the septic tank effluent, which distributes throughout the field. The south sewage disposal system has - Revision 9 Date 3/2/90 the design flow capacity to process 4,607 gallons of wastewater per day. The system is typically loaded at approximately 2,500 gallons per day during
4 normal plant operations. Figure (3) indicates diagrammatically the flow of both potable and wastewater throughout Vermont Yankee.
Both the main septic system and the south sewage disposal system’s septic tanks collect waste from the plant’s lavatories, showers, kitchens, and janitorial facilities outside the Radiological Control Area (RCA). No radioactivity is intentionally discharged to either of the septic systems. However, plant investi gations into the source of 1 ow 1 eve1 s of contami nati on found in septic waste have identified that very small quantities of radi‘oacti ve materi a1 s, which are bel ow detection 1 imits for radioactivity releases from the RCA, are carried out of the control area on individuals and accumulate in the septic waste collection tanks by way of floor wash water, showers, and hand washing. As a means of minimizing the transport of radioactive materials into the septic collection tanks, the primary source of the radioactivity ( .e., floor wash water) is now poured through a fi ter bag to remove suspended solids and dirt before the water is released into a janitorial sink.
The majority of the radioactivity found in waste sludge has been d associated with the main septic tank. Grab samples of sludge from the bottom of the COB and main septic tank were analyzed by gamma spectroscopy with the following results of plant - re1ated radi onucl ides :
Activity Concentration IsotoDe fl Siqma (DCi/kq Met)
COB Sludge CS-137 10.3 rt 1.8 (June 8, 1988) Co - 60 45.4 k 3.1
Main Tank Sludge Mn - 54 39.3 f 4.3 (June 8, 1988) Co-60 853.0 f 12.0 Zn - 65 52.7 f 8.2 CS - 134 13.0 j, 2.2 CS - 137 120.7 f 5.2
Revision -9- Date 3/2/90
B-10 The principle radionuclide is Cobalt-60, which accounts for 79% of the plant related activity in the septage samples. In comparison to in-plant 4 smear samples taken for 10CFR61 waste characterizations, the septage sample from the main tank correlates very close with the distribution of radionuclides identified in-plant as shown below:
Re1 ative I sotoDi c Dist ri buti ons
IsotoPe In-P1 ant Smears Main lank Sludge
Mn - 54 3.6% 3.6% co-60 81.5 79.1 Zn - 65 3.3 4.9 CS - 134 0.4 1.2 CS - 137 10.3 11.2
Additional analyses of the main tank septage showed that the liquid portion of the collected sample did not contain any plant-related activation or fission products, and that essentially all of the activity i.n the waste was associated with the solid sludge fraction. The average density of the collected sludge was found to be approximately equal to that of water, with a wet to dry ratio of 25.4 to 1.
"4 Both the liquid and solid fractions of the main tank septic waste were also analyzed for strontium with no detectable activity found. The liquid portion of the waste sample was also analyzed for tritium with no activity above the minimum detectable levels found. Appendix A to Attachment 2 contains the laboratory analysis reports of the samples taken from the COB and main septic tanks.
Prior to identification of the plant-related radioactivity in septage waste, the COB holding tank was being pumped on the average of twice per week, with the sludge and waste liquid transported off-site primarily to the Brattfeboro, Vermont, sewage treatment facility. Waste from the main septic tank was being pumped and transported off-site for disposal on the average of twice per year.
Revision -9- Date 3/2/90 B-11 With the replacement of the COB holding tank by the new south sewage disposal system, and the requested implementation of on-site land disposal of 4 accumulated septic waste, the frequency of collection tank pump-outs with land application of the waste is expected to be once per year. With the past pump- out frequency of the main tank being every six months, the accumulation of sludge at the bottom of the tank was well below its design capacity. During the 1988 sample collections, it was estimated that the sludge thickness was less than 1 foot of its 6-foot depth. However, for conservatism in the radiological evaluations, it is assumed that the sludge layer in the main septic tank and south disposal tank occupies 30% of their combined design volume, and that the frequency of pump-outs is semiannual as opposed to the expected annual cycle. Also, as noted above from laboratory analyses of the sludge layer taken from the bottom of the main tank, the average density of the tank contents is approximately equal to that of water, with a wet-to-dry ratio of 25.4 to 1. Hence, the weight of solids (W,,,) being disposed of is estimated, for purposes of this bounding dose assessment, to be approximately:
Wso, = 14.950 [gall x 3,785.4 Eccfgall x Ekg/ccl x 0.30 [solids fraction] x (1125.41 Cdrylwet ratio] - 700 Ekgl per pump-out of both tanks
or, 1,400 kg of dry solids per year.
3.0 DISPOSAL METHOD
Approval of this application will allow Vermont Yankee to dispose of septage by utilization of a technique of land spreading or surface injection in a manner consistent with all applicable state of Vermont health regulations regarding disposal of septic waste. Details of the chemical and biological controls necessary to satisfy state health code requirements are provided in Reference 5.
The septage will be spread or surface injected on land areas owned by Vermont Yankee and situated within the plant’s site boundary. Transportation of the septage waste to the disposal areas will involve pumping from one of the septic waste collection tanks (i.e., main septic tank, COB holding tank,
Revision -9- Date 3/2/90 B-12 new replacement COB septic tank, or from any other on-site septic waste collection point) into an enclosed truck-mounted tank. The enclosed tank -.c truck is used to prevent spillage while in transit to the disposal areas. The septage will be transported to one of the two disposal sites designated for land application for septage from Vermont ankee, and applied at a fixed rate based on either limitations imposed by the state of Vermont for heavy metals or organic content of the waste, or on the radioactivity content such that projected maximum individual doses wi 11 no exceed established dose objectives.
3.1 Seutic Haste Disposal Procedure
Gamma isotopic analysis of septic waste shall be made prior to each disposal by obtaining a representative sample from each tank prior to pump- out. At least two septic waste samples will be collected from each tank to be pumped by taking a volumetric column of sludge and waste water which allows for analysis of the solid’s distribution and content from top to bottom of each tank. The weight percent of solid content of the collected waste will be determined and applied to the gamma isotopic analysis in order to estimate the total radioactivity content of each tank to be pumped and spread on designated - disposal fields. These gamma isotopic analyses of the representative samples will be performed at the environmental Technical Specification lower limit of detection (LLD) requirements for liquids (see Technical Specification Table 4.9.31 in order to document the estimation of radiological impact from septage disposal .
The radionuclide concentrations and total radioactivity identified in the septage will be compared to the concentration and total curie limits established herein prior to disposal. The methodology and limits associated with determining compl iance with the disposal dose and activity criteria are described in Attachment 2. If the concentration and total activity limits are met, compliance with the dose assessment criteria will have been demonstrated since the radiological analysis (Section 4.5 and Attachment 2) was based on evaluating the exposure to a maximally exposed individual and inadvertent intruder after the accumul ation of twenty years of periodic semi annual
d Revision -9- Date 3/2/90 8-13 spreading of the septic waste on a single (2 acre) plot within one of the designated disposal areas. If the activity limit per disposal area is --cl projected to be exceeded, the appropriate exposure pathways as described in Section 4.5 will be evaluated prior to each additional application, or a separate plot within the designated disposal area will be utilized.
Annually, for years in which disposal occurs, the potential dose impact from disposal operations conducted during the year, including the impact from previous years, will be performed and results reported in the plant’s Semi annual Radioactive Effluent Re1 ease Report which is f 1’ 1 ed after January 1 A1 1 exposures wi11 be assessed uti1 izi ng the methodot ogy described in Attachment 2.
The established dose criteria requires that all applications of septage within the approved designated disposal areas shall be limited to ensure the dose to a maximally-exposed individual be maintained less than 1 mremlyear to the whole body and any organ, and the dose to the inadvertent intruder be maintained less than 5 mrem/year. The total activity based on the measured radionuclide distribution for any single disposal plot is not expected to exceed the following: Maximum Accumul ated Radi oacti vity A1 1 owed Per Acre Isotope -1 ~.limwCi 3
Mn - 54 1.4 Co-60 120.0 Zn - 65 1.4 CS - 134 0.7 CS-137 46.5
If any of the above radionuclides are projected to exceed the indicated activity val ues , then dose cal cul ations wi 11 be performed prior to spreading, in accordance with the methods detailed in Section 4.2.2 of Attachment 2, to make the determination that the dose limit criteria will not be exceeded.
Revision -9- Date 3/2/90 8-14 The concentration of radionuclides in any tank of septic waste to be disposed of will also be limited to a combined Maximum Permissible
1_ Concentration of Water (MPC) (as listed in 10CFR. Part 20, Appendix B, Table 11, Column 21 ratio of less than or equal to 0.1.
For radi ol ogi cal control , each appl icati on of septage wi 11 be appl ied on the designated land area by approved plant procedure which adheres to the following assumptions which were used in developing the dose impact:
0 During surface spreading or injection, the septage. and any precipitation falling onto or flowing onto the disposal field, shall not overflow the perimeter of the designated area.
0 Septage shall not be surface spread or injected into the top 6-inch soil layer within 300 feet from any drinking water well SUPPl Y .
0 Septage shall not be surface spread closer than 300 feet from the nearest dwelling or public building (or within 100 feet f injected into the top 6-inch surface layer).
0 Septage shall not be surface spread closer than 50 feet or within 25 feet if injected into the top 6-inch surface layer) from any roads or site boundary adjacent to land areas.
0 Septage shall not be surface spread within 100 feet (or within 50 feet if injected into the top 6-inch surface layer) of any surface water (rivers, streams, drainage ditches).
0 Low areas of the approved fields, subject to seasonally high groundwater levels, are excluded from the septage application
In addition to the radiological controls to limit the total accumulation of radioactive materials released by septic waste spreading, state of Vermont health code requirements will be followed to ensure the protecti on of the pub1 i c and envi ronment f rom chemi cal and bi ol ogi cal hazards. The application rate and acreage will be determined prior to each
J Revision -9- Date 3/2/90 B-15 disposal operation. This will vary with the chemical composition of the septage, the percent solids, and the radioactive concentrations. --.-
3.2 Administrative Procedures
Complete records of each disposal will be maintained. These records will include the concentration of radionuclides in the septage, the total volume of septic waste disposed, the total activity in each batch as well as total accumulated on the disposal plot at time of spreading, the plot on which the septage was applied, and the results of any dose calculations required.
The annual disposal of septage on each of the approved plot areas will be limited to within the established dose, activity, and concentration criteria noted above, in addition to limitations dictated by chemical and biol ogi cal condi tions. Dose guide1 ines, and concentration and acti vity limits, will be maintained within the appropriate values as detailed in Attachment 2.
Any farmer using land which has been used for the disposal of septic waste will be notified of any applicable restrictions placed on the site due to the land spreading or injection of waste. d
4.0 EVALUATION OF ENVIRONMENTAL IMPACT
4.1 Site Characteristics
4.1.1 Site Toposracihv
The proposed disposal sites consist of two fields located on the Vermont Yankee Nuclear Power Plant site, which is located on the west bank of the Connecticut River in southwestern Vermont at 1 atitude 42 degrees, 47 minutes north and longitude 72 degrees 31 minutes west. Both fields are on plant property within the site boundary and surrounded by a chain link fence.
i Revision -9- Date 3/2/90 B-16 Site A contains an approximate eight-acre parcel of usable land centered - approximately 2,200 feet northwest of the Reactor Building. Site B contains about two acres and is centered approximately 1,700 feet south of the Reactor Building. The usable acreage of both the north and south disposal fields is restricted to those areas which have no slopes greater than five percent to limit surface runoff. A radiological assessment based on the 1988 measured radioactivity concentrations in sludge has determined that a single two-acre plot would be sufficient for the routine disposal of septage for twenty years without exceeding the dose criteria to maximum exposed individual or inadvertent intruder. As a result, the eight-acre field to the northwest could be divided into four disposal plots, with the two-acre site at the south end of the plant site, providing a fifth plot. A portion of the United States Geological Survey topographic map (Brattl eboro quadrangle), showing the plant site. is presented in the Final Safety Analysis Report (FSAR) as Figure 2.5-1. A plan map showing the plant site and the disposal sites is given on Figure 4.
The sites are located along a glacial terrace on the west side of the Connecticut River. This terrace extends about 3,000 feet west rising gently and then more abruptly to a higher terrace and then to dissected uplands. Distance to the east from the disposal sites to the river is at least 100 feet - if septage is disposed of by surface spreading within the designated areas, or 50 feet if septage is injected directly into the soil.
Relief of the proposed disposal sites is low, with elevation ranging between 250 feet and 265 feet (msl). Mean water surface elevation of the adjacent river is about 220 feet.
The topographic character of the s te and surrounding area is compat ble with this use. The spreading of septage at these locations will have no effect on the topography of the area.
Revision -9- Date 3/2/90 B-17 4.1.2 Site Geo1oq.y
1 Profiles of site exploratory borings are shown in the FSAR in Figures 2.5-8 through 2.5-11. Current site characteristics as determined from a recent detailed site investigation can be found in Reference 5.
Composition of surfacial materials is compatible with the proposed use of the site for septic waste disposal.
4.2 Area Characteristics
4.2.1 Meteorol oqy
The site area experiences a continental-type climate with some modification due to the marine climate which prevails at the Atlantic seacoast to the east. Annual precipitation averages 43 inches and is fairly evenly distributed in each month of the year.
Potential impacts on septic waste disposal include occasional harsh weather: ice storms, severe thunderstorms, heavy rains due to hurricanes, the possibility of a tornado, and annual snowfall of from 30 to 118 inches per - year. In addition, frozen ground can occur for up to 4 months of the year.
Septage spreading will be managed by written procedure such that material which is spread or a mix of that material with precipitation will not overflow the perimeter of the disposal site.
Additional information on meteorology of the site can be found in Section 2.3 of the Final Safety Analysis Report.
4.2.2 Hvdrol 0q.y
Hydrology of the site and local area is tied closely to flow in the adjacent Connecticut River. River flow is controlled by a series of hydroelectric and flood-control dams including the Vernon Dam which is about 3,500 feet downstream of the site.
I Revision 9 Date 3/2/90 B- 18 All local streams drain to the Connecticut River and the site is in the direct path of natural groundwater flow from the local watershed easterly
I’ toward the river . Site groundwater 1 eve1 is infl uenced by both precipitation and changes in the level of ponding of the Connecticut River behind the Vernon Dam due to natural flow or dam operation.
Flood flows on the Connecticut are control ed by numerous dams inc udi ng five upstream of the site. Elevation of the 100 year flood is about 228 ft (msl); and, thus, well below the elevation of the proposed site which ranges from about 250 to 265 feet (mslf. The 100-year flood level is based on information presented in References (1) and (2).
Septage disposal by means of land spreading on the proposed site will have no adverse impact on area hydrology.
Further information about site hydrology is in Section 2.4 of the FSAR.
4.3 Water Usase
4.3.1 Surface Water
The adjacent Connecticut River is used for hydroelectric power, for cooling water for the Vermont Yankee plant, as well as for a variety of recreational purposes such as fishing and boating. The Connecticut River is not used as a potable water supply within 50 miles downstream of the plant.
Locally, water from natural springs are used for domestic and farm purposes. FSAR Table 2.4.5 and Figure 2.4-2 show springs used within a 1-mile radius of the site. FSAR Table 2.4.4 and Figure 2.4-1 show water supplies with surface water sources which are within a ten-mile radius of the site.
There will be no impact on surface water usage or quality as a result of septage disposal due to the required separation distances between surface waters and the disposal plots.
- Revision -9- Date 3/2/90 B-19 4.3.2 Groundwater - Based on a review of groundwater measurements in various site borings presented in the FSAR and References 3 and 5, an upper estimate of groundwater levels at the plant is about 240 feet. Considering the proximity of the Connecticut River and Vernon Pond, with a mean water surface elevation of 220 feet, this estimate for the groundwater level appears to be reasonable. Given the topography of the proposed disposal sites, it is highly unlikely that the groundwater level will be within 3 feet of the disposal area surface elevation. Prior to each application of septic waste to a disposal plot, the groundwater level in nearby test wells will be determined and no application will be allowed if the groundwater level in the vicinity of the disposal plot is found to be less than 3 feet.
Groundwater provides potable water for public wells as shown in FSAR Table 2.4.5 and Figure 2.4-1. Groundwater flow in the vicinity of the proposed disposal sites is towards the Connecticut River. There are no drinking-water wells located between the site and the river. Therefore, it is highly unlikely that any drinking water wells could be affected by septage disposal. FSAR Figure 2.4-2 and Table 2.4-5 present information on private wells near the plant.
2
The Vermont Yankee on-site wells provide water for plant use. This supply is routinely monitored for radioactive contamination.
To quantify the impact of septage disposal on the Connecticut River, a conservative groundwater/radionucl ide travel time analysis was performed. For an assumed average travel distance of 200 feet from the disposal site to the river. a groundwater travel time of 408 days was estimated from Oarcy's Law. This estimate is based on a permeability for the glacial till of 10 gpdift', a hydraulic gradient of 0.11 ftlft, and a soil porosity of 0.3. This analysis conservat'ively assumed that the septage placed on the ground was immediately available to the groundwater. In practice, a minimum of 3 feet separation between groundwater and the surface will be required at time of application of the septic waste.
I Revision -9- Date 3/2/90 9-20 Due to ionic adsorption of the radionuclides on solid particles in the groundwater flow regime, most radionuclides travel at only a small fraction of d the groundwater velocity. For the radionuclides present in the sludge, retardation coefficients were developed from NUREWCR-3130 (Reference 4). Retardation coefficients for Co-60, Cs-137, and Cs-134 were directly obtained from NUREGICR-3130. The coefficients for Zn-65 and Mn-54 were conservatively estimated using NUREGICR-3130 as a guide. The radionuclides, their half- lives, retardation coefficients, and their travel time to the river are summarized in Table 1.
TABLE 1
Radionuclide Travel Times
Retardation Travel Time Radi onucl ide Hal f ti fe Coefficient to River
Co-60 5.3 years 860 961 years CS-137 30.2 years 173 193 years CS - 134 2.1 years 173 193 years Zn - 65 244 days 3 1,224 days Mn - 54 312 days 3 1,224 days
The radiological impact on the river for the radionuclides reaching the river under this conservative analysis is discussed in Attachment 2. Water usage of the Connecticut River downstream from the disposal area is limited to drinking water for dairy cows, irrigation of vegetable crops, and irrigation of cow and cattle fodder.
Based on the assessments noted above, it is concluded that groundwater sources will not be adversely impacted as a result of septage disposal on the proposed site.
4.4 Land Use
Both the eight-acre and two-acre sites proposed for the disposa areas are currently part of the Vermont Yankee Nuc ear Power Plant Site ins de the plant’s site boundary which is enclosed by a chain link fence. It is
Revision -9- Date 3/2/90 E-21 undeveloped except for transmission line structures which traverse a portion of the northern disposal area. Development potential is under the control of Vermont Yankee. At present, the eight-acre site on the north end of the plant property is used by a local farmer for the growing of feed hay for use with his dairy herd. No curtailment of this activity as a result of the low levels of radioactivity in septage will be necessary.
Utilization of the proposed sites for septic waste disposal will result in no impact on adjacent land or properties because of the separation of the disposal plots from off-site properties, the general movement of groundwater toward the river and away from adjacent land areas, and the very low levels of radioactive materials contained in the waste. Administrative controls on spreading and the monitoring of disposal area conditions will provide added assurance that this proposed practice will not impact adjacent properties.
4.5 Radiolosical ImDact
In addition to state of Vermont limits imposed on septage spreading, based on nutrient and heavy metal content, the amount of septage applied on each of the proposed disposal plots will also be procedurally controlled to insure doses are maintained within the stated limits. These limits are based "_ on NRC Nuclear Reactor Regulation (NRR) staff proposed guidance (described in AIFINESP-037, August 1986). The proposed dose criteria require that the maximally exposed member of the general public receive a dose less than 1 mrem/year to the whole body or any organ due to the disposal material, and less than 5 mrem/year to an inadvertent intruder.
To assess the doses received by the maximally-exposed individual and the inadvertent intruder, six potential pathways have been identified. These incl ude:
(a) Standing on contaminated ground,
(b) Inhalation of resuspended radioactivity,
_, Revision -9- Date 3/2/90 8-22 (c) Ingestion of leafy vegetables,
._- (d) Ingestion of stored vegetables,
(f) Ingestion of meat, and
(9) Ingestion of milk.
The liquid pathway was also evaluated and determined to be insignificant. Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways with difference between the two related to the occupancy time. The basic assumptions used in the radiological analyses include:
(a) Exposure to the ground contamination and to resuspended radioactivity is for a period of 104 hours per year during Vermont Yankee active control of the disposal sites, and continuous thereafter. The 104-hour interval being representative of a farmer’s time on a plot of land (4 hours per week for 6 months).
I (b) The septic tanks are emptied every two to three years. (The 4. I assumed practice is to pump septic tanks once per year. The I actual practice may be to pump septic tanks every two to three I years. 1
(c) The tank radioactivity remains constant at the currently determined level. To account for the uncertainty associated with the counting statistics, the measured activity concentrations listed in Section 2 were increased by 3 sigmas. That is, the activity concentrations employed in dose assessment and the total radioactivity content per pump-out (at 700 kg of solids per batch) are as follows:
-I’ [ Revision 23 Date 05/12/99
B-23 Upper-Bound Activity Upper-Bound Activity I sotope Concentration EpCilks drvl Content [Ci /tankful )
fln - 54 1,348 9.436E-07 Co - 60 23,060 1.614E-05 Zn - 65 1,620 1.134E-06 CS-134 322 2.254E-07 CS - 137 4,100 2.870E-06
(d) The radiation source corresponds to the accumulation of radioactive material on a single plot (two-acre) within the proposed disposal sites over a period of 20 years (40 applications at 6-month intervals). (In actuality, the proposed sites will accommodate more than one disposal plot, and, in practice, more than one plot will most probably be used with an application frequency of once per year.)
(e) For the analysis of the radiological impact during Vermont Yankee active control of the disposal sites, a1 1 dispersed radioactive material remains on the surface and forms a source of unshielded radiation. (In practice, the septic waste will be either surface spread or directly injected within the top 6 inches of the disposal plot, in which case, the radioactive material will be mixed with the soil. This, in effect, would reduce the ground plane source of exposure by a factor of about four due to self-shielding.)
(f) No radioactive material is dispersed directly on crops for human or animal consumption, crop contamination being on y through root uptake.
(g) The deposition on crops of resuspended radioact vity is insignificantly smal 1 .
Revision -9- Date 3/2/90 8-24 (h) Pathway data and usage factors used in the analysis are the same as those used in the plant’s ODCM assessment of the off-site radiological impact from routine releases,with the exception that the fraction of stored vegetables grown on the disposal plots was conservatively increased from 0.76 to 1.0 (at present no vegetable crops for direct human consumption are grown on any of the proposed disposal plots).
(i It is conservatively assumed that Vermont Yankee relinquishes control of the disposal sites after the fortieth pump-out (i.e., the above source term applies also for the inadvertent intruder).
(j) For the analysis of the impact after Vermont Yankee control of the sites is relinquished, the radioactive material is plowed under and forms a uniform mix with the top six inches of soil: but, nonetheless, undergoes resuspension at the same rate as surface contamination.
From radio1ogi cal impact assessments associated with the disposa of septage on different plot sizes (Attachment 21, it was determined that a single two-acre plot within the disposal sites would accommodate the 1 mrem/year prescribed dose to the critical organ of the maximally exp sed individual for a period of up to 20 years, as well as the 5 mrem/year 1 prescribed dose to the inadvertent intruder after control is assumed to be relinquished. The calculated potential radiation exposures following the spreading of 40 combined (main septic system and south disposal system) tankfuls (at six-month intervals) on a single two-acre plot are as follows:
Control of Disoosal Sites Radiation Exposure Individual/Orqan
Controlled by VYNPS 0.1 mremfyr Chi 1 d/Whol e Body (Maximum Exposed Individual 1 0.2 mrem/yr Maximum ChildlCiver
Uncontroll ed 1.3 mremlyr Adul t/Whol e Body (Inadvertent Intruder) 3.9 mrem/yr Maximum Teenager/Lung
- Revision -9- Date 3/2/90
8-25 The individual pathway contributions to the total dose at the end of the 20-year accumulation of waste deposited on a single two-acre plot are as - listed below:
Pathway-Dependent Critical Orqan Doses
Maxi mal 1y Exposed Inadvertent Intruder Indi vidual /Organ Cri ticaf Individual /Organ (Chi ld/ Li verf (TeenagerlLung) Pathway (mremlyear) (mremiuear)
Ground Irradiation 0.0576 1.16 Inhalation 0.00122 2.74 Stored Vegetables 0.0913 0.00601 Leafy Vegetable 0.00467 0.00040 Milk Ingestion 0.0421 Of00229 Meat Ingestion 0.00249 0.00012
TOTAL 0.1994 3.909
In addition, an isoto'pic breakdown of the critical organ dose results listed above is shown in the following table:
Isotopic Breakdown of Maximum Radiation Exposures
Radioactivity Exposure Description Isotope [uCi /2 Acres] [ m r em/y r 3
During Vermont Yankee Mn - 54 2.831 0.000436 control of the Co-60 235.3 0.0559 disposal sites. Zn - 65 2.801 0.0230 Maxi mal 1y Exposed CS - 134 1.457 0.00231 Individual/Organ: CS-137 92.59 0.118 Chi 1 d/Li ver TOTAL 0.199
After Vermont Yankee Mn - 54 2.831 0.0144 control of sites is Co- 60 235.3 3.76 ref inqui shed. Zn - 65 2.801 0.00983 Inadvertent Intruder CS - 134 1.457 0.000505 Critical Individual / CS - 137 92.59 0.1247 Organ: Teenager/Lung TOTAL 3.91
Revision -9- Date 3/2/90
B-26 Of interest are also derived dose conversion factors which provide a means of ensuring septage disposal operations within the prescribed
I radiological guidelines. The critical -organ (worst-case) all-pathway values per acre are as follows:
A1 1 -Pathwau Critical -0rqan Dose Conversion Factors Durinq Vermont Yankee Control of Disposal Sites
Exposure
I sotope Individual /Orqan ~ Imrem/vr-uCi /acre1
Mn - 54 Adul t/GE-LLI 3.74E-4 GO-60 Teenager/ Lung 7.14E-4 Zn-65 Chi 1 d/Li ver 1.64E-2 CS-134 Child/Liver 3.18E-3 CS-137 Chi 1 d/Bone 2.66E-3
The calculational methodology and details of the radiological assessment and proposed operational controls on total activity and concentration of waste to be disposed are presented in Attachment 2.
.- 5.0 RADIATION PROTECTION
The disposal operation will follow the applicable Vermont Yankee procedures to maintain doses as low as reasonably achievable and within the specified dose and re1 ease concentration criteria.
c Revision -9- Date 3/2/90 B-27 REFERENCES
.d 1. Flood Insurance Study, Vernon, Vermont, Windham County, FEMA, Community No. 500137. July 25, 1980.
2. Flood Insurance Study, Town of Hinsdale, New Hampshire, Cheshire County, FEMA, Community No. 330022, October 15. 1980.
3. Vermont Yankee Well Development Evaluation by Wagner, Heindel, and Noyes, Inc. July 10, 1986.
4. NUREGfCR-3130, Influence of Leach Rate and Other Parameters on Groundwater Migration, by Dames & Moore, February 1983.
5. Vermont Yankee Nuclear Power Corporation On-Site Septage Disposal Plan, by Wagner, Heindel, and Noyes, Incorporated, June 1989.
Revision 9 DAte 3/2/90 8-28 FIGURE 1 MAIN SEPTIC TANK
Revision 9 Date 3/2/90 6-29 B FIGURE 2 Y SOUTH DISPOSAL SYSTEM SEPTIC TANK
Revision 9 Date 3/2/90 B-30 FIGURE 3 POTABLE AND WASTE WATER FLOW
I I I I I I 1 I 1 1 I 1 I I I I 1 I
Revision -9- Date 3/2/90 8-31 FIGURE 4 SEPTIC WASTE DISPOSAL AREAS
Revision -9- Date 3/2/90 B-32 (This attachment to Appendix B is incorporated into the ODCM by reference due to size. A complete copy is on file with Vermont Yankee Document Control as 4 part of Correspondence Letter BVY 89-59.)
ATTACHMENT 2
VERMONT YANKEE NUCLEAR POWER PLANT
RADIOLOGI CAL ASSESSMENT OF
ON-SITE DISPOSAL OF SEPTIC WASTE
and
PROPOSED PROCEDURAL CONTROLS TO ENSURE
COMPLIANCE WITH RAD10 LOGI CAL LIMITS
Revision -9- Date 3/2/90
B-33 APPENDIX D
ASSESSMENT OF SURVEILLANCE CRITERIA FOR GAS RELEASES FROM WASTE OIL INCINERATION
-J Revision 16 Date 10/28/93 D-1 APPENDIX D
ASSESSMENT OF SURVEILLANCE CRITERIA FOR GAS RELEASES FROM WASTE OIL INCINERATION
INTRODUCTION:
The Nuclear Regulatory Commission amended its regulations (10CFR20) in a Federal Register Notice (Vol. 57, No. 235; page 57649 / Monday, December 7, 1992) that permitted the on-site incineration of contaminated waste oil generated at licensed nuclear power plants without the need to , Incineration of waste oil is to be carried out under existing effluent limits, recordkeeping and reporting requirements. Specifically, licensees must comply with the effluent release limitations of lOCFR Part 20, and Part 50; Appendix I. This d, includes the site gaseous pathway dose and dose rate limits contained in the plant's Technical Specifications (Section 3.8). The dose contribution to members of the public resulting from the on-site incineration of contaminated waste oil must not cause the total dose or dose rate from all effluent sources to exceed the dose or concentration limits imposed by lOCFR20, 10CFR50; Appendix I, and the Radiological Effluent Technical Specifications (RETS). It is expected that the actual contribution to public exposures caused by waste oil burning will be a small fraction of the site's effluent limits, as well as a small portion of the total releases from the site. SOURCE DESCRIPTION Contaminated waste oil suitable for on-site incineration can be burned in the Waste Oil Burner located in the North Warehouse. The burner has its own exhaust stack situated on the roof of the warehouse. However, due to the short height of the exhaust stack above the roof line, this release point is considered to be a ground level point source for modeling discharges to the environment. In addition, the building wake effects from the North Warehouse are assumed to be independent of the larger Turbine HalllReactor Building complex due to its distance from these main structures. Consequently, the re1atively smal 1 size of the North Warehouse 1 eads to -..... Revision _16 Date 10/28/93 0-2 meteorological dispersion factors that are conservative with respect to the dispersion factors for the main plant structures. c The waste oil burner is rated to process oil at a 2 gal ./hour from a 500 gallon day tank. The offgas flow rate for the burner is rated as 199 cfm. This provides an air to oil dilution during the incineration of 44,800. WASTE OIL SAMPLINGI’SURVEILLANCE REQUIREMENTS The oil burner stack is not equipped with continuous air monitoring or sampling capability for the direct determination of radiolog‘ical effluent releases during the incineration process. As a consequence, sampling and analysis of the waste oil prior to its incineration is necessary to project the dose and dose rate consequences of burning contaminated oil. Calculations of projected dose from the incineration of total quantity of oil to be added to the Waste Oil Burn Day Tank for each series of burns will be performed in accordance with the methods in the ODCM and compared to the accumulated site total dose for that period before initiation of incineration. Dose rate determinations will be determined by averaging the projected dose for the quantity of radioactivity determined to be present in the oil over the expected duration of the burn necessary to incinerate the total volume to be added to the Day Tank. Inherent in this determination is the assumption that all radioactivity found to be present in each batch of oil will be released to the .- atmosphere during the incineration. No retention of activity in the combustion chamber is assumed in calculating the offsite radiological impact. Normal sampling and analysis methods for gaseous release streams cannot be applied directly to liquids (waste oil 1. Therefore, the sampling and analysis requirements for liquids as identified in Technical Specification Table 4.8.1 shall be used to determine the level of contamination in waste oil. The stated Lower Limits of Detection (LLD) given on Table 4.8.1 provide assurance that undetectable levels of contamination up to the LLD values will not result in a significant dose impact to the maximum offsite receptor. If waste oil was burned continuously for an entire calendar quarter, and the radionuclides listed in the ODCM Dose Conversion Factor Table 1.1-12 were assumed to be present in the oil at the LLD values specified in Technical Specification Table 4.8.1, the resultant maximum organ dose would amount to only 0.28% of the ALARA quarterly limit of 7.5 mrem. The principle limitation in the incineration of waste oil is that the site release limits contained in RETS, and implemented by the ODCM methodology, shall not be exceeded. The use of the liquid LLDs on waste oil sample analyses provide sufficient sensitivity to ensure that site dose limits will not be exceeded as a consequence of burning slightly contaminated oil. Revision 16 Date 10/28/93 D-3 APPENDIX F APPROVAL PURSUANT TO 10CFR20.2002 FOR ONSITE DISPOSAL OF COOLING TOWER SILT d Revision 21 Date 08/14/97 F-1 UNWED STATES NUCLEAR REGULATORY COMMISSION WASISWINOTON. D.C. po565OM)l June 18, 1997 Nn 97-85 Mr. Donald A. Reid Vice President, Operations Vermont Yankee Nuclear Power Corporation Ferry Road Brattleboro, VT 05301 SUBJECT: REVISED SAFETY EVALUATION - APPROVAL PURSUANT TO 10 CFR 20.2002 FOR ONSITE DISPOSAL OF COOLING TOWER SILT - VERMONT YANKEE NUCLEAR POWER STATION (PAC NO. M96371) Dear Mr. Reid: By letter dated August 30, 1995, Vermont Yankee Nuclear Power Corporation (VYNPC) requested approval, pursuant to 10 CFR 20.2002, for the onsite disposal of slightly contaminated silt material removed from Vermont Yankee Nuclear Power Station’s (Vermont Yankee’s) cooling towers. In a safety evaluation (SEI dated March 4, 1996, the NRC staff approved the proposed silt disposal. However, because of discrepancies VYNPC identified between the safety evaluation and VYNPC’s letter of August 30, 1995, VYNPC postponed implementation of the silt disposal until resolution of the discrepancies. By letter dated August 2, 1996, VYNPC informed the NRC staff of the discrepancies and requested that the SE be revised accordingly. Recognizing the discrepancies, the NRC staff has prepared the enclosed SE to resolve the discrepancies and to replace the SE of March 4, 1996. LA I The NRC staff’s approval of VYNPC’s silt disposal request is granted provided 1130 the enclosed replacement SE is permanently incorporated into Vermont Yankee’s Offsite Dose Calculation Manual as an appendix. Any modification to the proposed action that may be considered in the future must have prior NRC staff approval Pursuant to the provisions of 10 CFR Part 51, the Commission has published in the Federal Reqister an Environmental Assessment and Finding of No Significant Impact (61 FR 6662). Revision 21 Date 08/14/97 F-2 D. Reid -2- If you have any further questions regarding this matter, please contact Mr. Kahtan Jabbour at (301) 415-1496. 4 Sincerely , Patrick D. Milano, Acting Director Project Directorate 1-3 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket No. 50-271 Enclosure: Safety Eva1 uati on cc w/encl: See next page _. Revision 21 Date 08/14/97 F-3 Vermont Yankee Nuclear Power Vermont Yankee Nuclear Power Station Corporati on cc : e' Mr. Peter LaPorte, Director Regional Administrator, Region I ATTN: James Muckerheide U. S. Nuclear Regulatory Commission Massachusetts Emergency Management 475 Allendale Road Agency King of Prussia, PA 19406 400 Worcester Rd. P.O. Box 1496 R. K. Gad, I11 Framingham, MA 01701-0317 Ropes & Gray One International Piace Mr. Raymond N. McCandless Boston, MA 02110-2624 Vermont Divi si on of Occupati onal and Radi ologi cal Heal th Mr. Richard P. Sedano, Commissioner Administration Building Vermont Department of Public Service Montpel ier, VT 05602 120 State Street, 3rd Floor Montpel ier, VT 05602 Mr. J. J. Duffy Licensi ng Engi neer Pub1 ic Service Board Vermont Yankee Nuclear Power State of Vermont Corpo'rati on 120 State Street 580 Main Street Montpel ier, VT 05602 Bolton, MA 01740-1398 Chai rman, Board of Sel ectman Mr. Robert J. Wanczyk Town of Vernon Di rector of Safety and Regulatory P.O. Box 116 Affairs w Vernon, VT 05354-0116 Vermont Yankee Nuclear Power Corp. Ferry Road Mr. Richard E. McCullough Brattleboro. VT 05301 Operating Experience Coordinator Vermont Yankee Nuclear Power Station Mr. Ross B. Barkhurst, President P.O. Box 157 Vermont Yankee Nuclear Power Governor Hunt Road Corporation Vernon, VT 05354 Ferry Road Brattleboro, VT 05301 G. Dana Bisbee, Esq. Deputy Attorney General Mr. Gregory A. Maret. Plant Manager 33 Capitol Street Vermont Yankee Nuclear Power Station Concord, NH 03301-6937 P.O. Box 157 Governor Hunt Road Resident Inspector Vernon, VT 05354 Vermont Yankee Nuclear Power Station U.S. Regulatory Commission P.O. Box 176 Vernon, VT 05354 Chief, Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108 4 Revision 21 Date 08/14/97 F-4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASlikNOTON, 0.C. zOSdSQOol SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO ONSITE DISPOSAL OF SLIGHTLY CONTAMINATED COOLING TOWER SILT VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 1.O INTRODUCTION By letter dated August 30, 1995, Vermont Yankee Nuclear Power Corporation (VYNPC) requested approval for the onsite disposal of slightly contaminated si1 t materi a1 removed from Vermont Yankee Nuclear Power Station’s (Vermont Yankee’s) cooling towers. In a safety evaluation (SEI dated March 4, 1996, the NRC staff approved the proposed silt disposal. However, because of discrepancies between the SE and VYNPC’s letter of August 30, 1995, VYNPC postponed implementation of the silt disposal until resolution of the discrepancies. By letter dated August 2, 1996, VYNPC informed the NRC staff of the discrepancies and requested that the SE be revised accordingly. Recognizing the discrepancies, the NRC staff has prepared this SE to resolve the discrepancies and to replace the SE of March 4, 1996. 2.0 BACKGROUND VYNPC has previously obtained NRC staff approval of the onsite disposal of very-low-level radioactive material similar to the proposed silt disposal. By letter dated June 28, 1989, VYNPC proposed the onsite disposal of slightly contaminated septic waste material by land application at Vermont Yankee. By letter dated August 30, 1989, the NRC staff approved this request pursuant to 10 CFR 20.302 (now 10 CFR 20.2002). The NRC staff considered this site-specific application for Vermont Yankee to have insignificant radi ologi cal impact because the proposed septic waste materi a1 di sposal invol ved 1 icensed materi a1 containing 1 ess than 0.1 percent of the radioactive material, primarily cobalt-60 and cesium-137, already considered acceptable in the Final Environmental Statement (FES) of July 1972, and involved exposure pathways much less significant than those in the FES. In addition, the proposed septic waste materi a1 di sposal satisfied the following appl icab1 e boundary conditions for the disposal of licensed material: a. The whole body dose to the hypothetical maximally exposed individual must be less than 1.0 mremiyear. Revision 21 Date 08/14/97 F-5 -2- b. Doses to the whole body and any organ of an inadvertent intruder -.-- from the probable pathways of exposure are less than 5 mremlyear c. The disposal must be at the same site. Following the NRC staff’s approval on August 30, 1989, VYNPC implemented the disposal of the contaminated septic waste material as proposed. By letter dated August 30, 1995, VYNPC requested that the previous authorization for the onsite disposal of very-low-level radioactive material be amended to permit the onsite disposal of slightly contaminated silt material, within the boundary conditions of the-previously approved septic waste material disposal . 3.0 EVALUATION In its fetter of August 30, 1995, VYNPC stated that the proposed si t disposal method is the same as the previously approved septic waste disposal method, and utilizes land spreading in the same onsite areas approved for s ptic waste disposal. The volume of silt proposed for onsite disposal consists of 14,000 cubic feet 1396 cubic meters) accumulated through August 1335 plus approximately 4,000 cubic feet 1113 cubic meters) to be removed from the cool ing towers during each 18-month operating cycl e. The acti vity contained in the currently accumulated silt, based on samples taken by VYNPC in June 1995, is 0.193 millicuries, principally from 0.034 millicuries of cobalt-60 and 0.159 millicuries of cesium-137. The activity contained in the 1_ additional silt to be removed from the cooling towers each 18-month operating cycle is anticipated to be 0.059 millicuries, principally from 0.012 millicuries of cobalt-60 and 0.047 millicuries of cesium-137. VYNPG’s radiological assessment enclosed with its August 30, 1395, letter demonstrates that the combined radiological impact for all onsite disposal operations, the proposed disposal of silt and the previously approved disposal of septic waste material, will continue to meet the applicable boundary conditions (given above) for the disposal of licensed material. Therefore, the proposed onsite disposal of slightly contaminated silt is acceptable. As discussed in VYNPC’s letter of August 2, 1966, if the onsite disposal of cooling tower silt or septic waste material would result in exceeding the applicable boundary conditions (given above). then VYNPC must obtain prior LA I NRC staff approval of the disposal. In addition, VYNPC made the following 1193 commi tments: 1130 a. VYNPC will report in the Annual Radiological Effluent Release Report a list of the radionuclides present and the total radioactivity associated with the onsite disposal activities at Vermont Yankee. “-0 Revision 21 Date 08/14/97 F-6 c -3- VYNPC 4 AI b. will maintain records of radionuclide concentrations and 1130 total activity associated with onsi te disposal activities at Vermont Yankee in accordance with 10 CFR 50.75(g). 4.0 CONCLUSION The NRC staff finds that the radiological conditions at the Vermont Yankee site (see attachment) that would result from the onsite disposal of slightly contaminated silt material, as proposed by VYNPC pursuant to 10 CFR 20.2002, and the previously approved onsi te disposal of slightly contaminated septic waste material, are within the applicable boundary conditions (given above) for the disposal of licensed material. Therefore, the proposed onsite disposal of slightly contaminated silt removed from Vermont Yankee’s cooling towers is acceptable. LA I VYNPC is required to permanently incorporate this SE into the Vermont Yankee 1130 Offsite Dose Calculation Manual as an Appendix to document the the radioactive material onsite disposal activities approved for Vermont Yankee, and VYNPC’s re1ated commitments regarding reporting and record keeping. Any additional modification of VYNPC’s disposal activities which go beyond those proposed in the August 30, 1995, submittal, and are not addressed above must have prior NRC staff approval. In addition, any onsite disposal of cooling tower silt or septic waste material that would result in exceeding the applicable boundary conditions (given above), must also have prior NRC staff approval. I Principal Contributors: J. Minns D. Dorman 6. Harbuck Date: June 18, 1997 Attachment: Vermont Yankee Site Area Map Revision 21 Date 08/14/97 F-7 FIGURE 4 SEPTIC WASTE DISPOSAL AREAS Revision 21 Date 8/14/97 F-8 VERMONTYANKEE RESPONSIBILITY NUCLEARPOWER CORPORATION n am7 TO ENGINEERINGOFF(CE 58oM(rfff STREET BOLTON. MA 01740 (soel 719-671 I August 30, 1995 BVY 95-97 United States Nuclear Regulatory Commission Washington, DC 20555 ATTN: Document Control Desk References: (1) License No. DPR-28 (Docket No. 50-271) (2) Letter from R. W. Capstick. Vermont Yankee, to USNRC, "Request to Routinely Dispose of Slightly Contaminated waste in Accordance with 10CFR20.302(a)", BVY 89-59, June 18, 1989. (3) Letter from M. B. Fairtile, USNRC, to L. A. Tremblay, Vermont Yankee, "Approval Under 10 CFR 20.3021a) of Procedures for Disposal of Slightly Contaminated Septic Waste on Site at Vermont Yankee (TAC No. 73776)". dated August 30, 1989. Subject: Request to Amend Previous Approval Granted Under 10 CFR 20.302(a) for Disposal of Contaminated Septic Waste In accordance with the criteria of the Code of Federal Regulations, Title 10, Section 20.2002 (previously cited 10CFR20.302(a)), enclosed please find the subject application to amend the previously granted approval (Reference 3) to dispose of slightly contaminated septic waste on site at Vermont Yankee by expanding the allowable waste stream to include slightly contaminated Cooling Tower silt material. This application specifically requests approval to dispose of Cooling Tower silt deposits, contaminated at minimal levels, which have been or might be generated through the end of station operations at the Vermont Yankee Nuclear Power Plant. The proposed silt disposal method is the same as the septic waste disposal method requested in Reference 2 and approved in Reference 3. The disposal method utilizes on site land spreading in the same designated areas used for septic waste. Disposal of this waste in the manner proposed, rather than holding it for future disposal at a lOCFR Part 61 licensed facility when access to one becomes available, will save substantial casts and valuable disposal site space for waste of higher radioactivity levels. A radiological assessment and proposed operational controls based on continued on site disposal of accumulated river silt removed from the basins of the plant's mechanical draft cooling towers is contained in Enclosure A. The assessment demonstrates that the dose impact expected from the disposal of silt removed from the cooling towers during normal maintenance will not exceed the dose limits already imposed for septic waste disposal. The combined radiological impact for all on site disposal operations shall be limited to a total body or organ dose of a maximally exposed member of the public of less than one rnrerniyear during the period of active Vermont Yankee control of the site, or less than five mremlyear to an inadvertent intruder after termination of active site control. Enclosure B contains a copy of the original assessment and disposal procedures for Revision 21 Date 08/14/97 F-9 VERMONT YANKEE NUCLEAR POWER CORPORATION septic waste (References 2 and 31 for your use and reference in evaluating the proposed amendment. -r Upon receipt of your approval, Enclosure A will be incorporated into the Vermont Yankee ODCM. We trust that the information contained in the submittal is sufficient, however, should you have any questions or require further information concerning this matter, please contact this office. Sincerely. VERMONT YANKEE NUCLEAR POWER CORPORATION James J. Duffy Licensing Engineer Enclosures A & B c: USNRC Region I Administrator (Letter and Enclosure A) USNRC Resident Inspector - VYNPS (tetter and Enclosure A) USNRC Project Manager - VYNPS (Letter and Enclosure A) Revision 21 Date 08/14/97 F-10 ENCLOSURE A VERMONT YANKEE NUCLEAR POWER PLANT ASSESSMENT OF ROUTINE DISPOSAL OF COOLING TOWER SILT IN AREAS ON SITE PREVIOUSLY DESIGNATED FOR SEPTIC WASTE DISPOSAL - Revision 21 Date 08/14/97 F-11 Table of Contents Page 1.0 ~NTRO~UC~IO~...... 3 2.0 WASTE STREAM DESCRIPTION ...... 4 3.0 DISPOSALMETHOD ...... 9 3 . I Si1 t Disposal Procedure Requi rements ...... 9 3.2 Administrative Procedure Requirements ...... 12 4.0 EVALUATION OF ENVIRONMENTAL IMPACTS ...... 13 4.1 Site Characteristics ...... 13 4.2 Radiological Impact ...... 13 5.0 RADIATION PROTECTION ...... 23 6.0 CONCLUSIONS ...... 23 7.0 REFERENCES ...... 24 . 2 c Revision 21 Date 08/14/97 F-12 VERMONT YANKEE NUCLEAR POWER PLANT Assessment of Routine DiSDOSaf of Coolinq Tower Silt in Areas On-Si te Previous1Y Designated for Septic Waste Disposal 1.O INTRODUCTION: Vermont Yankee Nuclear Power Corporation (Vermont Yankee) requested from the NRC in 1989 permission to routinely dispose of slightly contaminated septic waste in designated on-site areas in accordance with 10CFR20.302(a). The NRC responded to this request on August 30, 1989 by granting approval of the proposed procedures for on-site disposal of septic waste concluding that the commitments, as documented in our request, were acceptable provided that our request and analysis be permanently incorporated into the plant’s Offsite Dose Calculation Manual (ODCM). Revision 9 to the ODCM (Appendix Bf incorporated the assessment and approval of the methods utilized for on-site disposal of slightly contaminated sewage sludge. In addition to the previously identified solids content of septic waste as a source of envi ronmental , 1 ow 1 eve1 radi oacti ve contaminated materi a1 , cool ing tower silt deposits resulting from the settling of solids from river water passing through the mechanical draft cooling tower system have been identified - to also contain low levels of plant-specific radionuclides. Periodic removal of the silt from the cooling tower basins is a necessary maintenance practice to insure operability of the cooling system. However, due to the presence of by-product materi a1 s in the si1 t, proper disposal requi rements must be appl ied to insure that the potential radiological impact is within acceptable limits. This assessment of silt disposal expands the original septic waste disposal assessment to include earthen type materials (cooling tower silt deposits) whi 1 e maintaining the original radi ologi cal assessment model ing and dose 1 imit criteria that have been approved for septic waste spreading on site. This assessment demonstrates that cooling tower silt can be disposed of in the same manner and under the same dose limit criteria as previously approved for septic waste in Appendix B to the Vermont Yankee ODCM. Implementation of the following commitments as an amendment to the original 10 CFR 3 .d Revision 21 Date 08/14/97 F-13 Part 20.302(a) approval for septic waste shall be incorporated into the Vermont Yankee ODCM upon approval by the PJRC. 4 2.0 WASTE STREAM DESCRIPTION: The waste involved in this assessment is residual solids (silt) collected in the basins of the plant’s mechanical draft cooling tower system. The silt consists of organic and inorganic sediments and earthen type materials that have settled from the cooling water flow taken from the Connecticut River as it passes through the towers. As a result of de-sludging the tower basins in 1993, an estimated 14,000 cubic feet of silt was accumulated on site. Clean-out operations will also occur periodically to ensure continued system operability. Sample analysis performed to the plant’s environmental lower 1 imits of detection requirements, as contained in Technical Specification Table 4.9.3., has identified Cobalt-60 and Cesium-137 in low concentrations as being present in silt collected in 1993. The cooling towers are located at the southern end of the plant facility complex but are not directly connected to any system in the plant that contains radioactivity. The postulated mechanism of how plant-related radionuclides have been introduced into the cooling system silt assumes that past routine effluents discharged from nearby plant gaseous release points 1 were entrained in the large mechanically-induced air flow that is pulled through the towers as a heat exchange medium. The cooling water flow provides a scrubbing action as it is breaks up into fine water droplets due to the splash pans of the towers. This scrubbing action washes any airborne particulates out of the. air. Over long periods of operation, any radioactivity removed from the air flow could buildup to measurable levels in silt that settles out in the basins at the bottom of the towers. Table 1 lists the analyses of twenty-one samples collected from the silt pile removed from the cool i ng tower basins. Radi oacti vi ty measurements, averaged over all the samples, indicate that the silt material can be characterized as containing approximately 50 pCi/kg (dry wt.) of Cobalt-60 and 198 pCifkg (dry wt.) of Cesium-137. Eight of the samples indicated no positive Cobalt-60 above a minimum detectable level achieved for the analysis. 4 Revision 21 Date 08/14/97 F-14 Table 1 Cool inq Tower Si1 t Radi oacti vity ( 1993 sampl es*) Sample # co - 60 CS - 137 (pCi/kg dry) (pCi/kg dry) G12759 53 144 G12758 72 172 G12757 (14 201 G12756 (17 245 G12755 73 206 G12754 (16 165 G12753 (27 240 G12752 79 181 G12751 (29 180 G12750 35 107 G12749 59 171 G12748 <19 205 G12747 <38 209 G12746 <7 218 G12745 50 241 G12744 40 220 G12743 68 2 64 612742 45 195 -.- 612724 71 115 G12723 104 264 G13940 126 218 Average: 50 198 Max. 126 264 Min. (7 107 Standard deviation: 30 42 * Average wet to dry sample weight ratio equal to 1.6 Dry weight silt density equal to 1.3 gm/cc. 5 - Revision 21 Date 08/14/97 F-15 For the purpose of estimating the total activity in the silt pile, the less than values in Table 1 are included as positives in the calculation of the - average radioactivity concentration. Cobalt-60, due to its relatively short half life, is typically associated with plant operations when measured in the near environment. However, Cesium-137 when measured in the environment may have a background component that is not related to power plant operations. Past weapons testing fallout has imposed a man made background level of Cesium-137 in New England soils and sediments that can vary over several hundred pCi/kg. The plant’s Environmental Monitoring Program has shown that Connecticut River sediment in the vicinity of Vermont Yankee averages about 123 pCi/kg (dry wt.) of Cesium-137 (Table 2) with no plant related detectable level of Cobalt-60. The value of 123 pCi/kg may represent an estimate of background level of Cesium-137 in sediment that would be subject to entrainment in cooling water flow that enters the plant. In comparing the measured levels of Cesium-137 on Table 1 with the past river sediment level, the average concentration in the cooling tower silt is higher than that of the river sediment data but does fall within the observed range of recorded sediment Cs-137. The river sediment Cesium-137 concentration averages about 62% of the concentration value detected in the tower silt. For purposes of this assessment of plant-related dose impact from the on-site disposal of si1t material, it is conservatively assumed that all detectable 1 Cs-137 in cooling tower silt is directly related to plant operations. No background component is subtracted from the measured values for this case study since only a single sampling location (down stream) is included in the Environmental Monitoring Program which may not fully describe the true background levels in the region. The total radioactivity for the current 14,000 cubic feet of silt collected on site can be estimated by multiplying this volume by its “as is” density of 2.1 gm/cc (i.e. 1.3 gm/cc dry weight density x 1.6 wetldry weight ratio) and then conservatively assume that the measured average dry weight radioactivity concentrations for Cobalt and Cesium would be the same as in the collected silt. Multiplying the average Cobalt-60 and Cesium-137 concentration in silt by the mass of the collected material produces estimates (Table 3) of total radioactivity that was removed from the cooling tower basin in September 1993. 6 - Revision 21 Date 08/14/97 F-16 Table 2 Cesium-137 in Connecticut River Sediments* CS- 137 Date fpCilkg dry) 61 a5 60 137 176 178 230 a4 62 (174 179 115 62 Average: 132 Max. 230 Min. 60 Standard deviation: 56 * Sampl es coll ected as part of the Vermont Yankee Radio1 ogical Envi ronmental Monitoring Program (REMPI for river sediment sample location SE-11. 7 Revision 21 Date 08/14/97 F-17 Table 3 Estimated Total Radioactive Material for 1993 Tower Clean-Out Volume of Mass Average Total Act. Decayed Act. Silt Concentration (as of 11/93) (as of 6/95) (ft3) fkg) ( pCi / kg 1 (uCi 1 (uCi 1 CO- 60 14.000 8.32E+5 50 42 34 CS - 137 14,000 8.32 E+5 198 165 159 In addition to 14,000 cubic feet of silt already accumulated, it is anticipated that periodic maintenance work in cleaning out the cooling tower basins will generate approximately 4,000 cubic feet of new silt material over each successive 18 month operating cycle. Assuming the same level of plant-related radioactivity concentration that was originally observed, the addi tional amounts of radi oact vity that will require on site disposal foll owing each refuel ing cycl e can also be estimated. Table 4 lists an estimate of the total radioact vity that might be present at each 18 month cl ean - out cycl e. - Table 4 Estimated Total Radioactive Material for Each 18 Month Maintenance Cycle Volume of Mass Average Total Silt Concentration Activity (ft3) (kg) (pCi /kg) (uCi 1 ...... ~.,.,...... ~,.”..__.___.._.._~____._.,.. * CO-60 4,000 2.38E+5 50 12 CS-137 4.000 2.38E+5 198 57 8 - Revision 21 Date 08/14/97 F-18 3.0 DISPOSAL METHOD: ci The method of silt disposal shall utilize a technique of land spreading in a manner consistent with the current commitments for the on-site disposal of septic waste as approved by the NRC and implemented as Appendix B of the Vermont Yankee ODCM (Reference I). The same land areas designated and approved for septic waste disposal shall be used for the placement of silt removed from the cool ing tower basins. Determination of the radi ol ogi cal dose impact shall also be made based on the same models and pathway assumptions as indicated in Appendix B of the ODCM. 3.1 Si1 t Disposal Procedure Reaui rements : Gamma isotopic analysis of silt samples shall be made prior to each disposal by obtaining representative composite samples in sufficient numbers to characterize the material removed from the cooling tower basins. Each gamma isotopic analysis shall be required to achieve the environmental lower limits of detection as indicated for sediment on Table 4.9.3 of the Vermont Yankee Technical Specifications. The estimation of total radioactivity to be disposed of shall be made based on the average of all composite sample analyses. The estimation of total I radioactivity and projected dose impact shall be made prior to placing the collected silt on the designated disposal plots. The dose impact from each disposal operation shall be included with all past septic waste and silt spreading operations to ensure that the appropriate dose limits are not exceeded on any waste disposal area for the combination of all past operations. The established dose criteria requires that all applications of earthen type materials within the approved designated disposal areas shall be limited to ensure that dose to a maximally exposed individual (during the Vermont Yankee control period) be maintained less than 1 mremlyear to the whole body and any organ, and the dose to an inadvertent intruder following termination of site control be maintained less than 5 mrem/year to the whole body and any organ. 9 Revision 21 Date 08/14/97 F-19 The limits on concentrations of radionuclides as addressed in Appendix B to the ODCM for septic waste (i.e., each tank of septic sludge to be disposed are d limited to a combined MPC ratio of less than 0.1) were included to ensure proper control was in place to address the situation of small quantities of re1ati vel y high Concentration mater1a1 . This 1 imitation does not di rectly apply to silt deposits since the silt is handled as dewatered sediments as opposed to liquid slurries of septic waste. For dry, earthen type material such as silt, a specific radionuclide concentration limit shall be applied in place of the septic waste liquid MPC ratio. No soil associated with a sample analysis that identifies a plant-related radionuclide in excess of the concentration limits of Table 5 will be permitted regardless of the total pathway dose assessment determined for the quantity material under consideration. For the case where more than one radionuclide is detected, the sum of the ratio rule will be applied. The measured concentration of each radionuclide divided by its limiting concentration value shall be added with the sum of all fractions equal to or less than 1. This limiting condition will prevent small volumes of relatively high specific radioactivity from being spread on the disposal plots, and therefore reduce the potential for creating unexpected hot spots of concentrated material. Table 5 lists, by radionuclide, soil concentration values that would generate an annual external effective dose equivalent of 25 mremlyear if it were assumed that an individual continuously stood on an infinite plane of soil contaminated to a depth of 15 cm. The assumptions of an infinite plane and continuous occupancy are conservative for situations where the amount of contaminated soil identified would not provide for a 15 cm soil depth over an extended surface area and where disposal site access is limited. Twenty-five mremlyear was selected as a reference value based on the fact that it was a suitable fraction of the NRC annual dose limit (100 mrem/year per 10 CFR Part 20.1301) applied to members of the public from all station sources. The 25 mrem/year also equals the EPA dose limit from 40 CFR Part 190 which would apply to real members of the public offsite and allow for credit to be taken in accounting for actual usage patterns such as occupancy time. The external dose factors provided on Table 5 were derived from Table E-2 of NUREWCR-5512 (Reference 3). 10 - Revision 21_ Date 08/14/97 F-20 Table 5 Dry Soil Maximum Concentration Values Soi 1 Concentration pCi / kg Radi onucl ide (equal to 25 mremfyr) Cr-51 1.51E+05 Mn-54 5.50E+03 Fe-59 3.83E+03 CO-58 4.70E+03 Co-60 1.82E+03 Zn - 65 7.85€+03 Zr-95 6.18E-f-03 Ag-llOm 1.66E-t-03 Sb-124 2.51 €+03 CS - 134 2.95€+03 CS-137 8.13E+03 Ce-141 7.85E+04 Ce-144 8.75E+04 Assumptions include infinite planar distribution. uniform depth distribution to 15 cm, soil density at 1.625E+06 gm/m3 and external direct dose pathway only with a 100% occupancy factor. 11 I Revision 2 Date 08/14/97 F-21 3.2 Administrative Procedure Requirements: .4 Dry silt material shall be dispersed using typical agricultural dry bulk surface spreading practices only in approved disposal areas on site. Complete records of each disposal will be maintained. These records will include the concentration of radionuclides detected in the silt, an estimate of the total volume of silt disposed of, the total radioactivity in each disposal operation as well as the total accumulated on each disposal plot at the time of the spreading, the plot on which the silt was applied, and the results of any dose calculations or maximum allowable accumulated activity determinations required to demonst ate that the dose limits imposed on these land spreading operations have not been exceeded. The determination of the total radioactivity and dose calcu ations shall also include all past septic waste and silt disposal operations that placed low level radioactive material on the designated disposal plots. The periodic disposal of silt on each of the approved land spreading areas will be limited to within the same established dose and radioactivity criteria that have been approved for septic waste disposal. Concentration limits that are applied to the disposal of earthen type .-- materials (dry soil) shall restrict the placement of small volumes of material that have relatively high concentrations of radioactivity such that direct exposure could not exceed a small proportion (25%) of the annual dose limits to members of the public that is contained in 10 CFR Part 20.1301, Any farmer leasing land used for the disposal of silt deposits will be notified of the applicable restrictions placed on the site due to the land spreading of low level contaminated material. These restrictions are the same as detailed for septic waste spreading as given in Reference 1. 12 Revision 21 Date 08/14/97 F-22 4.0 EVALUATION OF ENVIRONMENTAL IMPACTS: 'd 4.1 Site Characteri stics The designated disposal sites consist of two fields located on the Vermont Yankee Nuclear Power Plant site. Both fields are on the plant property within the site boundary security fence. Site A contains an approximate ten-acre parcel of land centered approximately 2,000 feet northwest of the Reactor Building. Site B consist of approximately two acres and is centered approximately 1,500 feet south of the Reactor Building. These are the same land parcels approved by the NRC for the land disposal of septic waste and are described in detail in Reference 1 along with the boundary restrictions for the placement of contaminated materi a1 . Radi ol ogi cal assessments of septic waste disposal have determi ned that a single two-acre plot would be sufficient for the routine disposal of that waste stream over a 20-year period without exceeding the dose criteria to a maximum exposed individual or inadvertent intruder. As a result, the ten-acre field to the northwest can be divided into five disposal plots, with the two-acre site at the south end of the plant site providing a sixth plot. It is therefore concluded that there is sufficient space within the a? ready approved disposal plots to accommodate additional materi a1 f rom the cool ing - tower basins along with the septic waste without the likelihood of exceeding the approved dose limit criteria. Since the residual organic and inorganic solids associated with river sediment (silt) are similar to the sand and residual organic material remaining after decomposition of septic waste that is removed from the plant's septic tanks, the conclusions of no significant environmental (non-radiological) impact associated with the disposal of septic waste are not changed by the addition of another earthen type material, namely silt. 4.2 Radioloqical ImDact: The amount of cooling tower silt, in combinat on with any septic waste di sposal s, wi11 be procedural 1 y control 1ed to insure doses are maintained within the prior approved limits (Reference 1 . These limits are based on 13 Revision 21. Date 08/14/97 F- 23 past NRC proposed guidance (described in AIF/NESP-O37), August 1986). The dose criteria require that the maximally exposed member of the general public d receive a dose less than 1 mrem/year to the whole body or any organ due to the disposed material and less than 5 mrem/year to the whole body or any organ of an inadvertent intruder. To assess the doses received by the maximally exposed individual and inadvertent intruder resulting from silt spreading, the same pathway modeling, assumptions and dose calculation methods as approved for septic disposal are used. These dose models implement the methodologies and dose conversion factors as provided in Regulatory Guide 1.109 (Reference 2). Six potential pathways have been identified and include: (a) Standing on contaminated ground, (b) Inhalation of resuspended radioactivity, (cf Ingestion of leafy vegetables, (d) Ingestion of stored vegetables. (f) Ingestion of meat, and (g) Ingestion of cow’s milk Based on the septic waste evaluations, the liquid pathway was determined to be insignificant. / Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways with the difference between them related to occupancy time. The basic assumptions used in the radiological analyses include: (a) Exposure to ground contamination and resuspended radioactivity is for a period of 104 hours per year during the Vermont Yankee active control of the disposal sites and continuous thereafter. The 104-hour interval is representative of a farmer’s time spent on a plot of land (4 hours per week for 6 months). (b) For the purpose of projecting and illustrating the magnitude of dose impacts over the remaining life of the plant, it is assumed that the current concentration levels of activity detected in sift remain 14 Revision 21 Date 08/14/97 F-24 constant. Table 1 indicates the measured radioactivity levels for Cobalt-60 and Cesium-137 first noted in silt material. (c) The maximum radiation source corresponds to the accumulation of radioactive material on a single plot (two acre> within the approved disposal sites over a period of 13 operating cycles. This extends over the next 18 years until after the operating license expires in 2012. The initial application (referenced to June 1995) consists of 14,000 cubic feet of silt collected in 1993 along with the first periodic clean out of the tower basins that adds an additional 4,000 cubic feet. All subsequent applications of 4,000 cubic feet occur at 18-month interval s. (d) For the analysis of the radiolog cal impact during the Vermont Yankee active control of the disposal sites, no plowing is assumed to take place and all dispersed radioact ve material remains on the surface forming a source of unshielded direct radiation. (e) No radioactive material is dispersed directly on crops for human or animal consumption. Crop contamination is only through root uptake. (f) The deposition on crops of suspended radioactivity is insignificantly small. (g) Pathway data and usage factors used in the analysis are the same as those used in the plant’s OOCM assessment of off-site radiological impact from routine releases with the exception that the fraction of stored vegetables grown on the contaminated land was conservatively increased from 0.76 to 1.0 (at present no vegetable crops for human consumption are grown on any of the approved disposal plots). (h) It is conservatively assumed that Vermont Yankee relinquishes control of the disposal sites after the operating license expires in 2012 (i.e., the source term accumulated on a single, 2-acre disposal plot applies also for the inadvertent intruder). 15 - Revision 21 Date 08/14/97 F-25 (i) For the analysis of the impact after Vermont Yankee control of the site is relinquished, the radioactive material is plowed under and forms a uniform mix with the top six inches of soil: but nonetheless, undergoes resuspension in air at the same rate as the unplowed surface contarninati on. For di rect ground pl ane exposure the sel f shi el ding due to the six-inch plow layer reduces the surface dose rate by about a factor of four. The dose models and methods used to generate deposition values and accumulated activity over the operating life of the plant are documented in Attachment 2 to Reference 1. Based on the measured concentrations and silt volumes noted in Section 2.0 above, the total radioactivity that remains on the disposal plots after the operating license expires is estimated on Table 6. Table 6 Projected Radi oacti vi tv Bui 1 duD Due to Si 1t SDreadi nq Nucl i de Contribution Accumulation from Total Remaining from initial 13 cycles at in year 2013 14,000 ft3 (uCi) 4,000 ft3/ea. (uCif (uCi 1 Cobal t - 60 3.2 61.9 65.1 Cesi urn- 137 104.9 500.5 605.4 The calculated potenti a1 radiation exposure fo7 1 owing the spreading of a1 7 silt material anticipated to be generated through the remainder of the operating license on a single, two-acre plot is provided on Table 7. 16 v Revision 21 Date 08/14/97 F-26 Table 7 Dose Impact Due to Continued Spreadinq to End of License Disposal Site Access Radi ation Exposure Individual /Organ Control 1 ed by VYNPS 0.228 mrem/yr adul tlwhol e body (rnax. exposed individual 1 0.820 mrem/yr max. childibone Uncontrolled by VYNPS 1.46 mremlyr adul tfwhofe body (inadvertent intruder after 2.41 mremlyr max. childlbone license termination) The individual pathway contr butions to the total dose due to cont nued silt spreading are shown on Table 8. Table 8 Pathway-Dependent Critical Orsan Doses Maximally exposed Inadvertent Intruder Individuaf/Organ Critical Individual /Organ (ChildIBone) (Chi 1 d/Bone Pathway (mrem/year) (mremfyear 1 Ground Irradiation 0.0474 0.957 Inhalation 0.00814 0.685 Stored Vegetables 0.528 0.528 Leafy Vegetables 0.0265 0.0265 Milk Ingestion 0.201 0.201 Meat Ingestion 0.00833 0.00833 Total : 0.82 2.41 17 - Revision 21 Date 08/14/97 F-27 In addition, the isotopic breakdown of the critical organ doses listed above (Table for the two detected radionuclides is seen to be: - 8) Table 9 Isotopic Breakdown of Maximum Radiation Exoosures Radioactivity Dose Percent of Description I sotope f uCi 12 acres 1 (mrem/yr 1 total During CS - 137 605.4 0.805 98.2 control of Co-60 65.1 Of0144 1.8 disposal 0.82 sites Max. organ : chi 1 d/bone Termination CS-137 605.4 2.12 88.0 of disposal Co-60 65.1 -0.29 12.0 site Max. 2.41 organ : chi 1 d/bone For comparison to the total dose calculated assuming the continued disposal of silt removed from the tower basins through the end of the operating license, the dose from just the original 14.000 cubic feet collected is shown on Table 10. 18 Revision 21 Date 08/14/97 F- 28 Table 10 c Dose Impact Due to Sinqle 114,000 ft3) Silt SDreadinq Disposal Site Access Radiation Exposure Individual /Organ Controlled by VYNPS (max. 0.064 mrem/yr adul tlwhol e body exposed indi vidual in 1995 1 0.219 mrem/yr max.' chi 1 d/ bone Uncontroll ed by VYNPS 0.224 mremlyr adul tlwhol e body (inadvertent intruder after 0.393 mrem/yr max. childlbone license termination) Tab e 10 shows that the applicat on of the silt material initially collected (14 000 cubic feet) accounts for about 27 percent of the maximum individual organ dose during the control period as compared to the scenario of continued periodic silt spreading over the balance of the operating license. This illustrates that dose impacts from the material currently collected are well below the acceptance criteria of limiting the dose from any two-acre plot to no more than 1 mrem/year during the control period and 5 mrem/year after termination of the license, and is expected to remain below the acceptance - criteri a throughout the plant 1 ife. If unexpected bui 1 dup of radi oacti vity in future silt clean-out operations were to occur, the option for use of alternate disposal plots remains available to ensure that the impact from any single, two-acre plot stays within the acceptance criteria. Also of interest are derived dose conversion factors which provide a means of ensuring that septic and si1 t disposal operations remain within the prescribed radi ol ogi cal gui del ines noted above. The critical organ (worst case) and whole body dose factors for all pathways on a per acre bases are 19 Revision 21 Date 08/14/97 F-29 given on Table 11 for periods during Vermont Yankee control of the disposal site and on Table 12 for post control periods associated with the inadvertent .4 intruder scenario. The dose conversion factors have been expanded to include other potential radionuclide beyond the original five that were addressed in Reference 1. This provides a means to assess other nuclides if future disposal operations identi fy addi tional radi onucl ides not previ ously observed. The devel opment of these additional nucl ide dose conversion factors uti 1 i ze the same modeling and pathway assumptions as used to derive the factors for the -original five radionuclides identified in septic waste. The models f-or these site and pathway-specific dose factors are those in Regulatory Guide 1.109 (Reference 2) and are described in detail in Attachment If to Reference 1. 20 + Revision 21 Date 08/14/97 F -30 Table 11 4 A1 1 -Pathway Critical Orqan / Who1 e Body Dose Conversion Factors Durinq Vermont Yankee Control of Disposal Sites Critical Organ Whole Body Dose Dose Factor Factor Nucl ide Individual /Organ fmremlyr per uCi /acre) Cr-51 Teen/Lung 1.14E-05 5.76E-06 Mn - 54 Adul t/GI - LLI 3.75E-04 1.93E-04 Fe-55 Chi 1 dfBone 6.45E-06 1.06E-06 Fe-59 Teen/Lung 4.61 E- 04 2.13E-04 CO-58 Teen/ Lung 3.27 E - 04 2.OlE-04 Co-60 Teen/Lung 7.17E-04 5.31E-04 Zn - 65 Childiliver 1A4E-02 1.03E-02 Zr-95 Teen/ Lung 4.47E-04 1.34E - 04 AgllOm Teen/GI-LLI 1.32E-02 5.24E-04 Sb-124 Teen/ Lung 8.34E- 04 3.54E-04 CS - 134 ChitdlLiver 3.18E-03 1.28E-03 CS-137 Chi ld/Bone 2.66E-03 7.02E-04 I Ce-141 Teen/ Lung 1.54E-04 1.5OE-05 Ce-144 Teen/Lung 6. OOE- 04 2.44E-05 21 - Revision 21 Date 08/14/97 F-31 Table 12 All -Pathway Critical Orqan / Whole Bodv Rose Conversion Factors Post Vermont Yankee Control of RisRosal Sites (Inadvertent Intruder) Critical Organ Whole Body Rose Rose Factor Factor Nucl i de Indi vi dual /Organ (mrernlyr per uCi/acre) Cr-51 TeenlLung 5.89E-04 1.19E-04 Mn - 54 Teen/ Lung 1.02E-02 3.12E-03 Fe - 55 TeenILung 3.50E-04 2.27E -05 Fe-59 Teen/ Lung 2.55E-02 4.43E-03 CO-58 TeenILung 1.59E-02 3.72E-03 Co-60 Teen/ Lung 3.19E-02 9.09E- 03 Zn-65 ChildILiver 1.89E-02 1 AE-02 Zr-95 TeenILung 2.93E-02 2.99E-03 AgllOm TeenILung 3.59E-02 9.53E-03 Sb-124 Teen/ Lung 4.73E-02 7.04E-03 CS - 134 Chi 1 d/Li ver 1.21E-02 9.36E-03 - CS - 137 Chi ldf8one 6.98E-03 3.85E-03 Ce-141 TeenILung 1.21E-02 3.44E -04 Ce-144 Teen/ Lung 5.00E-02 1.52E-03 22 - Revision 21 Date 08/14/97 F-32 5.0 RADIATION PROTECTION: i The disposal operation of silt material from the cooling tower basins will follow the applicable Vermont Yankee procedures to maintain doses as low as reasonably achievable and within the specific dose criteria as previously approved for septic waste disposal (Reference 1). 6.0 CONCLUSIONS: Silt collected from the cooling tower basins is an earthen type material that is similar in characteristics to septic waste residual solids with respect to the radiological pathway behavior and modeling and can be disposed of through on-site land spreading on the same disposal plots as previously evaluated and approved for septic waste disposal. The radiological assessment of low level contaminated silt shows that the projected dose from the on-site periodic spreading of this material will have no significant dose impact to members of the public and can be maintained below the approved dose limitations already in place for septic waste. 23 - Revision 21 Date 08/14/97 F-33 7.0 REFERENCES: d (1) Vermont Yankee ODCM, Appendix 8; "Approval of Criteria for Disposal of Slightly Contaminated Septic Water On-Site at Vermont Yankee". (Included NRC approval letter dated August 30, 1989, VY request for approval dated June 28. 1989 with Attachments I and 11). (2) USNRC Regulatory Guide 1.109, Rev. 1; "Calculation of Annual Doses to Man from Routine Re1 eases of Reactor Effluents for the Purpose of Eva1 uating Compliance with 10 CFR Part 50, Appendix I", dated October 1977. (3) NUREGfCR-5512, Vole 1, "Residual Radioactive Contamination From Decommissioning" , Final Report, dated October 1992. file: vsilt.mss 24 k Revision 21 Date 08114197 F-34 APPENDIX G MAXIMUM PERMISSIBLE CONCENTRATIONS (MPCs) IN AIR AND MATER ABOVE NATURAL BACKGROUND TAKEN FROM 10CFR20.1 TO 20.602. APPENDIX B 4 Revision 21 Date 08/14/97 G-1 APPENDIX G With the implementation of the revised Part 20 to Title 10 of the Code of Federal Regulations (10CFR20.1001-20.2401), the Maximum Permissible Concentrations (MPCs) that were part of the old 10CFR20 were replaced by a new Appendix B to lOCFR20 for the limits that apply to effluents released to unrestricted areas. However, MPC values were also used and accepted as licensing conditions for the control of radioactive materials in situations other than those directly covered by the requirements of the regulations. One example is the on-site disposal of septic waste which used the MPC values as one criteria of acceptability for land spreading. Appendix B to the ODCM references 10CFR20.1-20.601 Appendix B MPCs as concentration criteria for this disposal option. With the final publication of the revised 10CFR20.1001-20.2401, the original MPC tables of the old 10CFR20 are no longer in print. As such, this appendix to the ODCM is added to provide a reference source for the MPC values contained in the original Appendix B to 10CFR20.1-20.601 for those conditions that still refer to these requirements. Revision 21 Date 08/14/97 G-2 APPENDIX G MAXIMUM PERMISSIBLE CONCENTRATIONS (MPCs) (FROM 10CFR20.1 TO 20.602. APPENDIX B) APPENDIX B TO 1820.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at e n d of Appendix Bl Table I I Table I1 ~ Col. 1 Col. 2 "":r ~ Water Air ~ Water Element (atomic number) I sotopel (pCi /ml 1 (pCi::>-: /ml ) I (pCi /ml 1 (pCi/ml) Actinium (89) ...... Ac 227 S 2x10-12 6~10-~ 8~10-l~ 2x10-6 I 3x1o-I1 9~10'~~3~10'~ Ac 228 S I Americium (95) ..... Am 241 Am 242m Am 244 S I Antimony Sb 122 S I 5~10-~ 3~10-~ 2x104 1x10'~ 3~10~~3~10~~ 9x10-10 1~10-~ Argon (18) ...... A3 7 Sub2 6~10-~ ...... 1~10-~ ...... A4 1 Sub 2x10-6 ...... 4~10'~ ...... Arsenfc (33) ...... As 73 S 2x10-6 1x10-z 7~10-~ 5~10'~ I 4~10-~ 1x10-2 1x10-8 5~10'~ As 74 S 3~10-~ I 1~10-~ As 76 S I As 77 S ~ 5x10-: 1 2x10-i 1 2~10"~ 8x10a5 I 4x10- 2x10- 1~10-~1 8~10-~ Revision 21 Date 08/14/97 G-3 APPENDIX B TO 8520.1 * 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appenc ix 81 I Tab e1 Tab e I1 Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope' (pCi /ml 1 (pCi /ml (pCi /ml 1 (pCi /ml 1 Astatine (85) ...... 5~10-~ 2x10-10 2x10-6 2~10-~ 1x10-9 7~10-~ Barium (56) ...... Ba 131 1x10-6 4~10-~ I i I 4~10-~ 1x10-8 Ba 140 Berkelium (97) . . . . . Bk 249 2x10-2 2x10-2 Bk 250 6x10e3 1x10-6 6~10-~ Beryllium (4) ...... Be 7 5x10-' 5x10'' Bismuth (83) ...... Bi 206 Bi 207 6x10*' 5xlO-l' Bi 210 6x10-' 2 x 10-I0 I I 6x10-' 2x10-10 Bi 212 lxlo-z 3x10-' 1x10-2 7x10'' Bromine (35) ...... 4x10" 6~10~' Cadmium (48) ...... Cd 109 2x10-9 3x10-' Cd 115m Cd 115 Calcium (20) ...... Ca 45 S 3~10-~ 9x10-6 I 1~10-~ 2~10-~ Ca 47 S ~~10'~ 6x10e9 I 2x10'~ 6x10'' Revision 21 Date 08114197 G-4 APPENDIX B TO JIZO.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix B1 Elemen t ( atomi c number 1 Californium (98) . . . . Carbon (6) ...... Cerium (58) . . . , . . . Cesium (55) . . . , . , . Revision 21 Date 08/14/97 G-5 APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND Table I Table I1 Col. 1 Col. 2 Col. 1 Col. 2 Air Water Air Water Element (atomic number) Isotope1 (pCi /ml f (pCi /ml 1 (pCi /ml 1 (pCi /ml 1 Chlorine (17) ...... C1 36 S 4~10~~2~10-~ 1x10'8 8~10-~ I 2x10-8 2~10-~ ~xIo-~O 6~10-~ C1 38 S 3~10'~ 1x10-2 9x10'8 4x10m4 I 2x10-6 1x10-2 7~10-~ 4~10-~ Chromium (24) ...... Cr 51 S 1~10-~ 5x10-' 4~10~' 2x10'~ I 2x10+ 5x10-' 8~10-~ 2~10-~ Cobalt (27) ...... Co 57 S 3~10-~ 2x10'2 1~10-~ 5~10'~ I 2~10-' 1x10-2 6~10'~ 4~10'~ Co 58m S 2~10-~ 8x10-' 6~10-~ 3~10-~ I 9x10-6 6x10-' 3x10'' 2x10'~ Revision 21 Date 08/14/97 G-6 APPENDIX B TO BBZO.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See f ootnotl-e s at e nId of Appendix B1 Table I Table I1 Col. 1 Col. 2 Water Water 1 Element (atomic number) Is 0 tope Dysprorium (66) . . . . . S I Dy 166 S II Einsteinium (99) . . . . I Es 253 Es 254m ~ S I Es 254 I Es 255 S I ~ S I I Er 171 S I ~___ Europium (63) . . , . . . Eu 152 S 4x1OW7 I 2~lO-~I 1x10"; I 6x10A5 I fT12-9.2 hrs) I 3x10-' 2~10-~ 1x10- 6x10-' Eu 152 S 1x10-8 2~10-~ 4x10-10 8x10'' I (T/2-13 yrs) I 2x10-8 ZX10-3 6x10-" 8~10-~ Eu 154 4~10'~ 6~10-~ 1~10-~' 2x10" 7x10'' 6~10-~ 2x10"' ZxlO-' f Eu 155 9x10-8 6~10~~3x10-' 2~10-~ 7x10-' 6~10'~ 3x10-' 2~10-~ Fermium (100) ...... Fm 254 I Fm 255 Fm 256 S I I Fluorlne (9) ...... F 18 S 5~10~~2x10-2 2~10-~ 8~10-~ I 3~10'~ 1x10'2 9x10-8 5~10-~ Revision 21 Date 08/14/97 G-7 APPENDIX B TO 5520.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix 81 Table I Table I1 Col. 1 CO?. 2 Col. 1 Col. 2 Air Water Air Water Elemen t ( atomi c number ) Isotope1 (pCi /ml 1 (pCi /ml 1 (pCi /nl 1 CpCi /ml Gadolinium (64) . . . . . Gd 153 S 2~10-~ 6~10*~ 8x10'' 2~10-~ I 9x10-8 6~10-~ 3x10" 2~10-~ Gd 159 S 5~10~~2~10-~ 2x104 8x I 4~10-~ 2x10'3 1x10-8 8~10'~ Gallium (31) ...... Ga 72 S 2x10'~ 1~10-~ 8x10-' 4~10-~ I 2~10-~ 1~10-~ 6x10-' 4~10'~ Germanium (32) . . . . . Ge 71 S 1~10-~ 5~10-~ 4~10-~ 2~10-~ I 6~10-~ 5~10-~ 2~10-~ 2~10-~ Gold (79) ...... Au 196 S 1x10-6 5~10-~ 4~10-~ 2~10-~ 1 6~10-~ 4~10-~ 2x10-8 1x10'~ Au 198 S 3~10~~2~10-~ 1x10-8 5~10'~ I ZX~O-~ 1~10-~ SxlO-' 5~10-~ Au 199 S lXlO+j 5~10'~ 4~10-~ 2~10-~ I 8~10-~ 4~10-~ 3~10-~ 2~10-~ Hafnium (72) , . . . . . Hf 181 S 4~10"~ ZX~O-~ 1~10-~ 7~10'~ I 7~10-~ 2~10-~ 3x10-' 7~10'~ Holmium (67) ...... Ho 166 S 2~10-~ 9~10-~ 7x10-' 3~10-~ I 2~10-~ 9~10~~6~10~' 3~10-~ . ____ Hydrogen (1) ...... H3 S 5~10-~ 1x10-1 2~10-~ 3x I 5~10-~ 1x10'1 2~10-' 3~10-~ Sub ZX~O-~ ...... 4~10'~ ...... Indium (49) ...... In 113111 S 8~10-~ 4~10-~ 3~10-~ 1~10-~ I 7~10-~ 4x10-' 2x10'~ 1~10-~ In 114111 S 1~10-~ 5~10'~ 4x10-' 2x10'5 I 2x10-8 5~10-~ 7xlO-l' 2~10-~ In 115m S 2x10-6 1x10-2 8~10-~ 4x10T4 I 2x204 1x10-2 6~10-~ 4~10-~ In 115 S 2~20-~ 3~10-~ 9x10*' 9~10-~ I 3~20-~ 3~10-~ 1~10-~ 9x10-5 Iodine (53) ...... I 125 S 5x30-' 4~10-~ 8~10-l~ 2~10-~ I 2~50-~ 6~10-~ 6x10-' 2~10-~ I 126 S 8~20~' 5~10'~ 9xlO-l' 3~10-~ I 3~80~~3~10"~ 1x10-8 9~10'~ I 129 S Zx30-' 1x10'~ ZXIO-; ~XIO-~ I 7x20-' 6~10-~ 2x 10 - 2x10-4 Revision 21 Date 08/14/97 6-8 1 Table I Table I1 Col. 2 Col. 1 Col. 2 ~ "":;I-Water ~ Air Water Element (atomic number) Isotope1, (pCi /ml 1 I (pCi /ml 1 (pCi/ml) (pC1 /ml 1 Iodine (53) ...... I 131 9x70-' 6~10~' 1x10-'10 (Continued) 3x90-' 2~10-~ 1x10'8 I 132 S 2~10-~ 2~10-~ 3x10'' I 9x10-' 5x10m3 3~10'~ I 133 s 3~10-~ 2~10'~ ~x~O-~O I 2x1~~1x10-3 7~10-~ I 134 S 5~10-~ 4~10-~ 6x10-' I 3~10-~ 2x10-2 1x10-7 I 135 S 1x10'~ 7~10-~ 1~10-~ I 4x10" 2~10-~ 1x10-8 Iridium (77) . , . . . . Ir 390 S 1x10-6 6~10-~ 4~10-~ I 4~10'~ 5~10'~ 1x10-8 I I Ir 192 S 1~10-~ 1x10'~ 4~10~' I 3~10"~ 1~10-~ ~XIO-~~ Ir 194 Iron (26) ...... Fe 59 1 ; I 1x10-7 1 2x10-3 I 5x10'9 5x10-8 2~10-~ 2~10-~ Krypton (36) . . . . , . Kr 85m f Sub 1 6~10'~I ...... I lxlOm7 ...... . . .. * ...... Lanthanum (57) . . . . . 2~10-~ 2~10-~ ~ Lead (82) ...... Pb 212 s 2x104 6~10-~ 6x10-" I 2x10-8 5~10-~ 7xlO-'O Revision 21 Date 08/14/97 G-9 APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND -. [See footnotes at end of Appendix 83 1 e1 Tab 1e II Col. 2 Col. 1 Col. 2 Water Air Water El emen t ( a tomi c number 1 lsotopt (pCi /ml 1 (pCi /ml 1 (pCi /ml 1 (pCi/ml) Lutetium (71) ...... Lu 177 Manganese (25) . . . . . 1x10-8 1~10-~ 3~10-~ I 2x10-8 S I Hg 197 S lxlO+j 4x10-' I I 3~10~~ 9x10-8 Hg 203 S I 7x10F8 I 1x10.~ Molybdenum (42) . . . . . Mo 99 Neodymium (60) . . . . Nd 144 3x10-" I 2x10'~ 1x 10-11 Nd 147 1x10'8 8x10-' Nd 149 8x loa3 8~10-~ ~~ ~~ Neptunium (93) . . . . . Np 237 1x 10- l3 1 4x10-" INp 239 S axio-7 I 7x10-' I ~~ Nickel (28) ...... Ni 59 6~10-~ 2x10'8 I 6x10" 3x10-' Ni 63 8~10-~ 2~10-~ 2x10-2 1x10-8 Ni 65 3~10-~ F 2x10'8 Revision 21 Date 08/14/97 G-10 APPENDIX B TO 5520.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix B1 I Table I I Table 11 Col. 1 Col. 2 Col. 1 Col. 2 Air 1 Water I Air 1 Water El emen t ( a tomi c number Isotopef (pCi /ml 1 (pCi /ml 1 (pCi /ml 1 tpCi /in1 1 Nlobium (Columbium) (41) Nb 93m S 1~10-~ 1x10-2 4x10-' 4~10-~ I 2x10'~ 1x10-2 SX~O-~ 4~10-~ Nb 95 S 5~10~~3~10-~ 2x10-8 1~10-~ I 1~10-~ 3~10-~ 3~10-~ 1~10-~ Nb 97 S 6~10-~ 3x10" 2~10-~ 9x10-' I 5x10m6 3x10-* zX10+ 9x10e4 Osmium (76) ...... Os 185 2x10'8 Os 191m 2x10'~ 7x10-' 6~10-~f ;;3;:; I s 1 9~10-~I 7x10" I ~xIO-~ os 191 1x10-6 5~10-~ I s I 4x10e7 I 5~10-~ Os 193 I ; 1 mo-; I 2x10-3 1x10-8 6~10-~ 3x10- I 5x10-5 Palladium (46) . , . . . Pd 103 1x10-6 1x10-2 5x10-' 3~10'~ I 1 7~10-~I 8~10-~I 3x10-' I 3~10-~ Pd 109 6~10-~ 3~10"~ 2x10-8 9~10~~ I I 4~10-~I ZX~O-~I 1~10-~I 7~10-~ Phosphorus (15) . . . . . Platinum (78) . , . . . Pt 193m S I Pt 193 Pt 197m Pt 197 Revision 21 Date 08/14/97 G-11 CONCENTF I I Table I Table I1 Col. 2 Col. 1 Col. 2 1 "":;, I Water 1 Air 1 Water El ement ( atomi c number 1 isotope1 (pci /m1) (pCi /ml ) (pCi Imll (pCi /ml 1 Plutonium (94) . . . . . Pu 238 S 2x10-12 1x10'~ 7~io-l~ 5~10'~ I 3~10-l~ 8~10-~ 1x 10- 3~10-~ Pu 239 S 2x 10- 1x10'~ 6~10"~ 5~10-~ I 4~10'~~8~10-~ 1x10- 3~10-~ Pu 240 2x 10- Pu 241 Pu 242 S 2x10-12 1~10-~ 6~10-l~ 5~10-~ I 4~10''~ 9~10~~1x 10 - 3~10'~ Pu 243 S 2x10-6 1x10-2 6x10-* 3~10-~ I 2x104 1x10-2 8~10~~3~10'~ Pu 244 S 2x10-12 1x10'~ 6~10~'~4~10-~ I 3~10-l~ 3~10-~ 1~10-'~ 1x10'~ Polonium (84) ...... Po 210 S ~x~o-~O 2~10+ 2~10-~l 7~10-~ I 2x10-lo 8~10'~ 7~10-l~ 3~10'~ ~~ Potassium (19) . . . . . K4 2 S 2x10-6 9~10-~ 7x10T8 3~10-~ Praseodymium (591 . . . . Pr 142 2x10-7 Pr 143 1x10-8 5~10-~ I 6x10-' I 5x10e5 Promethium (61) . . . . . Pm 147 6~10-~ 6x10b3 2x10'~ I ? 1 1~10'~I 6~10-~I 3x10-' Pm 149 3x10e7 1~10-~ 1x10-8 4~10-~ I i I ZX~O-~I l~lO-~I 8x10-' 1 4~10-~ Protoactinium (91) . . . Pa 230 2x10-:' 1 7~10'~ 6~10"~1 2~10'~ 1 1 8x10- 7x10-' 1 3xlO-l' 2~10-~ Pa 231 lx10 - l2 3~10-~ 4~10-l~I 9~10'~ I I 1~10"~I 8~10'~1 4~10-l~ 2x10'~ Pa 233 S 6~10-~ 4~10'~ 2x10'8 1x10-4 I 2~10-~ 3~10-~ 6x10'' 1~10-~ Revision 21 Date 08/14/97 G-12 APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AN0 WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix BI I Table I I Table I1 Element ( atomi c number 1 Radium (88) ...... Ra 228 7~10'~~8~10~~ ZX~O-~~ 1 3x1OV8 1 T 1 4~10"~1 7x10'' I 1~10-I~ 3~10'~ Radon (86) , . . . . . , Rn 220 1 S I 3~10-~I ...... I lxlO-' I ...... ~~ ~ ~~ Rn 22Z3 1 I 3~10'~I ...... 1 3x10-' I ...... Rhenium (75) ...... Re 183 S 3~10-~ 2x10-2 9x10-8 6~10~~ I 2~10-~ 8~10-~ 5x10-' 3~10-~ Re 186 S 6~10-~ 3~10-~ 2x10-8 9~10'~ I 2~10-' 1~10-~ 8x10-' 5~10'~ Re 187 9x10-6 7~10~~3~10-~ 3~10'~ I I 5~10-~I 4~10-~I ZX~O-~I 2~10'~ Re 188 4~10-~ 2~10-~ 1xlO" 6~10-~ I s I ZXlO-' 1 9~1O-~1 6~10~' I 3; Rhodium (45) ...... , Rh 103m 8~10-~I 4x10-'1 I 3x1OV6 1x10'~ I s I 6~10*~ 3xlO-l 2x10-6 I 1x10-2 Rh 105 S 8~10'~ 4~10'~ 3x10'' 1~10-~ I 5x10*' 3~10-~ 2x10'8 1x10'~ Rubidium (37) ...... 1 Rb 86 S 3~10-~ 2~10-~ 1x10-8 7~10-~ I 7~10-~ 7~10'~ 2~10-~ ZX~O-~ Rb 87 1 ; I 5~10-~I 3~10-~ 1~10-~ 7~10'~ I 2x10-4 Ruthenium (44) . . . . . ZX~O-~I 1~10-~I 8x10m8 4x10.' I : I 2x10'6 1x10-2 6x10-' I 3~10-~ Ru 105 1 ; 1 7x10-i I 3~10'~ 2x10-8 1~10-~ 5x10- 3~10~~I 2x10-8 I 1x10-4 Ru 106 8~10-~ 4~10'~ 3x10-' 1x10'~ I I 6x10" I 3~10-~I 2x10-l' I 1x10-5 Revision 21 Date 08/14/97 6-13 APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix Bl El emen t ( atomi c number ) Isotc Samarium (62) ...... Sm 147 S -I Sm 151 S I Sm 153 S -I Scandium (21) . . . . /. . Sc 46 S -I sc 47 S I Sc 48 S -I Selenium (34) . . . . . Se 75 S -I Silicon (14) ...... Si 31 S -I Silver (47) ...... Ag 105 S I Ag llOm S I Ag 111 S 1 Sodium (11) ...... Na 22 S 2~10-~ 1~10-~ 6x10" 4~10-~ -I 9x10'9 9~10-~ 3~10'~' 3~10-~ Na 24 S 1x10-6 6~10-~ 4~10-~ 2~10-~ -I 1~10-~ 8~10'~ 5x10" 3~10-~ Strontium (38) . . . . . Sr 85m S 4~10-~ 2x10'1 1x10'6 7~10-~ I 3~10-~ 2x10'1 1x10-6 7x10e3 Sr 85 S ZxlO-; I 3~10'~ 8x10-' 1~10-~ -I 1x10- 5~10'~I 4x10" I 2x10-4 Sr 89 S 3~10-~I 3x10-i I -I 4~10-~ 8x10- Sr 90 S I Revision 21 Date 08/14/97 G-14 APPENDIX B TO 8820.1 * 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix B1 Table I Table I1 Col. 1 Col. 2 Col. 1 Cot. 2 Air Water Air Water Element (atomic number) Isotope1 (pCi /ml 1 CpCi Iml 1 (pCI /ml 1 (pCi /ml 1 Strontium (38) , . . . , Sr 91 S 4~10'~ 2~10-~ 2x10-8 7x10e5 (Continued) I 3~10-~ 1~10-~ 9~10-~ 5~10'~ Sr 92 S 4x10a7 ~~10'~2x10-8 7~10'~ I 3x10'' 2~10-~ 1x10-8 6~10'~ Sulfur (16) ...... s 35 S 3~10'~ 2~10-~ 9x10-' 6~10~~ I 3~10'~ 8~10-~ 9x10-' 3~10-~ Tantalum (73) ...... Ta 182 S 4x10a8 1x10'~ 1x10;90 4x10e5 I 2x10-8 1~10-~ 7x10- 4~10~~ Technetium (43) . . . . . Tc 96m S 8~10-~ 4x10-' 3~10~' 1x10-2 I 3~10-~ 3xlO-l 1x10'6 1x10'2 Tc 96 S 6x10e7 3~10-~ 2x10-8 1~10'~ I 2~10-~ 1~10-~ 8x10-' 5~10"~ Tc 97m S 2x104 1x10'2 8~10'~ 4x10-' I 2x1~-7 5~10'~ 5x10-' 2~10-~ Tc 97 S 1~10-~ 5x10-' 4x10" 2~10-~ I 3~10-~ 2x10-2 1x10'8 8x10-4 Tc 99m S 4~10'~ 2x10-1 1x10'6 6~10-~ I 1x10'~ 8x10e2 5~10-~ 3~10"~ Tc 99 S 2x10-6 1x10'2 7~10'~ 3x10a4 I 6~10~~5~10-~ 2~10-' ~~10'~ Tellurium (52) . . . . . Te 125111 S 4~10-~ 5~10-~ 1x10-8 2~10-~ I 1~10-~ 3~10-~ 4x10-' 1~10-~ Te 127m S 1~10-~ 2~10-~ 5x10-' 6~10-~ I 4~10-~ 2~10-~ 1~10-~ 5~10'~ Te 127 S 2XlO+j 8~10'~ 6~10'~ 3~10~'" I 9~10-~ 5~10'~ 3~10-~ 2x10'~ Te 129m S 8~10-~ 1~10-~ 3x10-' 3~10-~ I 3~10'~ 6~10'~ 1x10'~ 2~10-~ Te 129 Revision 21 Date 08/14/97 G-15 APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at c. d of Appendix 81 ITable I Table I1 Col. 1 Col. 2 Col. 1 Col. 2 Air Water Ai r Water (pCi /mi ) (pCi /ml 1 (pCi /ml 1 (pCi /m1 ) 1x10'~ 1~10-~ 3x10-' 4~10-~ 3~10-~ 1~10-~ 1~10-~ 4~10-~ 3~10'~~5~10'~ 1x10-11 2~10-~ 2x 10 - '0 5~10-~ 6~lO-~~2~10-~ Th 228 9x 10"'2 2~10-~ 3~10-l~ 7x10-' 6~10"~ 4x10'~ 2~10-l~ 1~10-~ .. Th 230 2x10-" 5~10-~ 8~10"~ 2x10-6 1~10-l~ 9~10~~3~10-l~ 3~10-~ Th 231 1x10-6 7~10-~ 5x10+ 2~10-~ 1x10-6 7~10~~4~10-~ 2~10-~ Th 232 3x10-'' 5x10m5 1~10-~* 2x10-6 I If 3~10-l~ 1~10-~ lx10 -I2 4~10-~ Th 234 s I Thulium (69) ...... Tm 170 S I I Tm 171 I: Tin (50) ...... + 1 Sn 113 ~ ; ~~ Sn 125 Revision 21 Date 08/14/97 G-16 APPENDIX B TO 5tiZO.l - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of ADpendix 61 Element (atomic number) Tungsten (Wolfram) (74) . Uranium (92) ...... Vanadium (23) ...... Xenon (54) ...... Ytterbium (70) ..... Yttrium (39) ...... Revision 21 Date 08/14/97 6-17 APPENDIX B TO §§20.1 - 20.602 CONCENTRATIONS IN AIR AND WATER ABOVE NATURAL BACKGROUND [See footnotes at end of Appendix B] Table I1 Col. 1 Col. 2 Col. 1 Col. 2 Ai r Water Ai r Water Element (atomi c number 1 I sotopel (pCi /ml 1 (NCi /ml 1 (rrCi /ml 1 (uCi/ml) '~ 2~10-~ 2x10-5 Y 93 S I Zinc (30) ...... Zn 65 S I Zn 69m S 1x10-8 I 1x10'8 Zn 69 S I Zirconium (40) . . . . . Zr 93 S 2x10-2 4~10-~ f 2x10-2 1x10-8 1x10'~ 9x10-8 Any single radi onucl ide 1x10-'5 ...... 3x10'' ...... not listed above with decay mode other than alpha emission or spontaneous fission and with radioactive half-life less than 2 hours. Any single radionuclide 9~10-~ 1x10-10 3~10'~ not listed above with decay mode other than alpha emission or spontaneous ffssion and with radioactive half -1 ife greater than 2 hours. Any single redionuclide 4~10-~ 2~10-l~ 3~10-~ not 1 i sted above which decays by alpha emission or spontaneous fission. I Revision 21 Date 08/14/97 G-18 'Soluble (SI; Insoluble (I). '"Sub" means that values given are for submersion in a semispherical infinite cloud of -" airborne material. 3These radon concentrations are appropriate for protection from radon-222 combined with its short-lived daughters. Alternatively, the value in Table I may be replaced by one-third (113) "working level ." (A "working level" is defined as any combination of short-lived radon-222 daughters, polonium-218, lead-214, bismuth-214 and polonium-214. in one liter of air, without regard to the degree of equilibrium, that will result in the ultimate emission of 1.3 x lo5 MeV of alpha particle energy.) The Table I1 value may be replaced by one-thirtieth (1130) of a "working level." The limit on radon-222 concentrations in restricted areas may be based on an annual average. 4For soluble mixtures of U-238, U-234 and U-235 in air chemical toxicity may be the limiting factor. If the percent by weight-enrichment of U-235 is less than 5. the concentration value for a 40-hour work week, Table I. is 0.2 milligrams uranium per cubic meter of air average. For any enrichment, the product of the average concentration and time of exposure during a 40-hour work week shall not exceed 8 x SA pCi-hrlml, where SA is the specific activity of the uranium inhaled. The concentration value for Table I1 is 0.007 milligrams uranium per cubic meter of air. The specific activity for natural uranium is 6.77 x curies per gram U. The specific activity for other mixtures of U-238, U-235 and U-234, if not known, shall be: SA = 3.6 x curiesigram U U-depleted SA = (0.4 + 0.38 E + 0.0034 E2) E 2 0.72 where E is the percentage by weight of U-235. expressed as percent. NOTE: In any case where there is a mixture in air or water of more than one radionuclide, the limiting values for purposes of this Appendix should be determined as follows: 1. If the identity and concentration of each radionuclide in the mixture are known, the limiting values should be derived as follows: Determine, for each radionuclide in the mixture, the ratio between the quantity present in the mixture and the limit otherwise established in Appendix 5 for the specific radionuclide when not in a mixture. The sum of such ratios for a71 the radionuclides in the mixture may not exceed "1" (i.e., "unity"). EXAMPLE: If radionuclides A. 5, and C are present in concentrations C,, C,, and C,. and if the applicable MPC's are MPC,, and MPC,, and MPC, respectively, then the concentrations shall be limited so that the following relationship exists: (C,/MPC,) + (C,IMPC,) + fCc/MPCc) 1 2. If either the identity or the concentration of any radionuclide in the mixture is not known, the limiting values for purposes of Ap endix B shall be: a. For purposes of Table I, Col. 1 - 6x10- E b. For purposes of Table I, Col. 2 - 4~10-~ c. For purposes of Table 11, Col . 1 2~lO-'~ d. For purposes of Table 11, Col . 2 - 3~10~~ 3. If any of the conditions specified below are met, the corresponding values specified below may be used in lieu of those specified in paragraph 2 above. a. If the identity of each radionuclide in the mixture is known but the concentration of one or more of the radionuclides in the mixture is not known. the concentration limit for the mixture is the limit specified in Appendix "6" for the radionuclide in the mixture having the lowest concentration limit: or b. If the identity of each radionuclide in the mixture is not known, but it is known that certain radionuclides specified in Appendix "5" are not present in the mixture, the concentration limit for the mixture is the lowest concentration limit specified in Appendix "B for any radionuclide which is not known to be absent from the mixture: or Revision 21 Date 08/14/97 G- 19 Table I Table I1 Col. 1 Col. 2 Col. 1 Col. 2 Air Water Ai r Water c. Element (atomic number) and isotope (PT1pCi/m1) (pci /ml) (pCi /ml (WCi /ml If it is known that Sr 90, I 125, I 126, I 129, I 131 (I133, Table I1 only), Pb 210, Po 210, At 211, Ra 23, Ra 224, Ra 226, Ac 227, Ra 228, Th 230, Pa 231, Th 232, Th-nat, Cm 248, Cf 254, and Fm 256 are not present ...... ~xIO-~ ...... 3~10-~ If it is known that Sr 90, I 125, I 126, I 129 (I 131, I 133, Table I1 only), Pb 210, Po 210, Ra 223, Ra 226, Ra 228, Pa 231, Th-nat, Cm 248, Cf 254, and Fm 256 are not present ...... 6~10.~ ...... 2x10-6 If it is known that Sr 90, I 129 (I 125. I 126, I 131, fable I1 only), Pb 210, Ra 226, Ra 228. Cm 248, and Cf 254 are not present ...... zX10+ ...... 6x10.' If it is known that (I 129, Table I1 only). Ra 226. and Ra 228 are not present ...... 3~10'~ ...... 1x10-7 If it is known that alpha-emitters and Sr 90, I 129, Pb 210. Ac 227. Ra 228, Pa 230. Pu 241, and Bk 249 are not present ...... 3x10.' ...... 1x10-10 ...... - - If it is known that alpha-emitters and Pb 210. Ac 227, Ra 228, and Pu 241 are not present ...... 3x10-" ...... lxlo-'' ...... If it is known that alpha-emitters and Ac 227 are not present ...... 3x10-" ...... 1x10-12 ...... ~~ 1 I If it is known that Ac 227. Th 230, Pa 231, Pu 238, Pu 239, Pu 240, Pu 242, Pu 244. Cm 248. Cf 249 and Cf 251 are not present 3~10-l~ ...... iX10-l3 ...... 4. If a mixture of radionuclides consists of uranium and its daughters in ore dust prior to chemical separation of the uranium from the ore, the values specified below may be used for uranium and its daughters through radium-266, instead of those from paragraphs 1, 2, or 3 above. a. For purposes of Table I, Col . 1 - lxlO-lo pCi/ml gross alpha activity; or 5~10.'~pCi/ml natural uranium or 75 micrograms per cubic meter of air natural uranium. b. For purposes of Table 11, Col. 1 - 3x10"l2 pCi/ml gross alpha activity: 2~10-l~pCi/ml natural uranium; or 3 micrograms per cubic meter of air natural uranium. 5. For purposes of this note, a radionuclide may be considered as not present in a mixture if (a) the ratio of the concentration of that radionuclide in the mixture (CAI to the concentration limit for that radionuclide specified in Table I1 of Appendix "8" (MPC,) does not exceed 1/10, (i.e. C,/MPC, A 1/10} and (b) the sum of such ratios for all the radionuclides considered as not present in the mixture does not exceed 1/4 i.e. (C,/MPC, + C,/MPC, ... + - 1/41. T Revision 21 Date 08/14/97 G-20 Appendix H 1. “Request to Amend Previous Approvals Granted Under 10CFR20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil”, dated June 23rd, 1999, BVY 99-80 2. “Supplement to Request to Amend Previous Approvals Granted Under 10CFN0.30(a) to AlIow for Disposal ofContaminated Soil”, dated January 4*, BVY 00-02 3. “Vermont Yankee Nuclear Power Station, Request to Amend Previous Approvals Granted under 10CFR20.302 a to AIlow for Disposal of Contaminated Soil (TAC No. MA5950), dated June 15L) ,2000, NVY 00-58 I I Revision29 Date 1/11/02 I H- 1 VERMONTYANKEE NUCLEARPOWER CORPOFMTKIN I85 Old Ferry Road, Brattleboro, VJ' 05301 -7002 (802) 257-5271 June 23, 1999 BW99-80 United States Nuclear Regulatory Commission A1TN: Document Control Desk Washington, DC 20555 References: (a) Letter fiom €LW. Capstick, Vermont Yankee, to USMRC, "Request to Routinely Dispose of SlightIy Contaminated Waste in Accordance with iOCFR20.302(a)", BVY 89-59, June 28, 1989 (b) Letter &om M.B. Fairtile, USNRC, to L. A. Trembley, Vermont Yankee, "Approval Under 10CFR20.302(a) of Procedures for Disposal of Slightly Contaminated Septic Waste on Site at Vermont Yankee (TAC No. 73776)", NVY 89-189, dated August 30, 1989 (c) Letter from J.J. DufFy, Vermont Yankee, to USNRC, "Request to Amend Previous Approval Granted Under IOCFR20.302(a) for Disposal of Contaminated Septic Waste", BVY 95-97, dated August 30, I995 (d). Letter from S.A. Varga, USNRC, to D. A. Reid, Vermont Yankee, "ApprovaI Pursuant to lOCFR20.2002 for Onsite Disposal of Cooling Tower Silt - Vermont Yankee Nuclear Power Station (TAC No. M93420)", Nw 94-39, dated March 4, 1996 (e) Letter from P.D. Milano, USNRC, to D. A. Reid, Vermont Yankee, "Revised Safety Evaluation - Approval Pursuant to lOCFR20.2002 for Onsite Disposd of Cooling Tower Silt - Vermont Yankee Nuclear Power Station (TAC No. M9637I)", NVY 97-85, dated June 18,1997 Subject: Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271) Request to Amend Previous Amrovals Granted Under 10 CFR 20.302fa) for Disoosai of Contaminated SeDtic Waste and cool in^ Tower Silt to Allow for Disnosal of Contaminated Soil In accordance with lOCFR 20.2002 (previously 10CFR20.302(a)), Vermont Yankee submits this application to amend the previously granted approvals to dispose of slightly contaminated septic waste and cooling tower silt on-site, This application expands the allowable waste stream to include slightly contaminated soil generated as a residual by-product of on-site construction activities. This application specificatly requests approval to dispose of soil contaminated at minimal levels, which has been or might be generated through end of station operations at the Vermont Yankee Nuclear Power Plant. The proposed soil disposd method is the same as the septic .waste and coohg tower silt disposal methods requested in References (a) and (c), and approved in References (b) and (e). The disposal method utilizes on-site land spreading in the same designated areas used for septic waste and cooling tower silt. Disposal of this waste in the manner proposed, rather than holding it for future disposal at a lOCFR Part 61 licensed facility will save substrtntid costs and reserve valuable disposal site space for waste of higher radioactivity levels. Revision 29 Date 1/11/02 1 H- 2 VERMONTYANKEE NUCLEAR POWER CORPORATION I BVY 99-80 i Page 2 of2 A radiological assessment and proposed operational controls for the inclusion of additional earthen material (soil) for on-site disposal with septic waste and cooling tower silt is provided in Attachment A. The assessment demonstrates that the dose impact expected from the existing accumulation of approximately 25.5 cubic meters of so& in total with dI past waste spreading operations, will not approach the dose limits already imposed for septic and cooling tower silt disposal. In addition to the existing accumulated soil, VY also requests that any future low leveI contaminated soil that might be generated as a by-product of plant construction and maintenance activities be dowed to be disposed of in the same manner providcd the approved acceptance dose criteria are met. AN soil dyses will be to environmental lower limits of detection. The results of dl disposal operations will also be reported in the Annual Radioactive Efnuent Release Report. The combined radiological impact for all on-site disposd opedons wiIl continue to be limited to a total body or organ dose of a maximaIly exposed member ofthe public of less than one mredyear during the period of active'fremont Yankee control of the site, or less thaa five mredyear to an inadvertent intruder after termination of active site control. The Vermont Yankee Off-Site Dose Calculation Manual (ODCM) contains ampy of the original assessment and NRC approval €or septic waste disposal (Refereucac a and b) and the previous amendment for cooIing tower silt Peferences c and e). Upon receipt of your approval, the information contained in Attachment 1 as well as the basis for approval will be incorporated into .the ODCM. We trust that the information contained in the submittd is sufficient. However, should you have any questions or requir:: Merinformation concerning this matter, piease contact Mr. Jim DeVincentis at 802-258-4236. Sincerely, mmom YANKEE:NUCLEAR POWER CORYOIWTION Licensing M-er Attachment cc: USNRC Region I Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS VT Department of Public Service Revision 29 Date 1/11/02 I H- 3 - SUMMARY OF VERMONT YANKEE COMMZTMENTS BVYNO.: 99-80 The following table identifies commitments made in this document by Vermont Yankee. Any other actions discussed in the submittal represent intended or pIanned actions by Vermont Yankee. They are described to the NRC for the NRC’s information and are not regulatory commitments. Please notify the Licensing Manager of any questions regarding this document or any associated commitments. COMMITMENT I COMMITTED DATE I OR “OUTAGE” h I VYAPF 0058.04 (Sample) AP 0058 Original Page 1 of 1 H- 4 llRevision 29 Date 1/11/02 Docket No. 50-27 1 BVY 99-80 Attachment 1 Vermont Yankee Nuclear Power Station Assessment of On-site Disposal of Contaminated Soil by Land Spreading 1 Revision 29 Date 111 1/02 H- 5 - BW99-80 /Attachment 1 I Page 1 TABLE OF CONTENTS Page LIST OF TABLES ...... 2 I .0 INTRODUCTION ...... 3 1.1 Background ...... 3 1.2 Objective ...... 3 2.0 WASTE DESC.ON ...... 4 ' 3.0 SOIL DISPOSAL AND ADMINISTRATIVE PROCEDURE REQUrREMENTS...... 5 4.0 EVAL,UATION OF ENVIRONMENT& IMPACTS ...... 6 4.1 Site Characteristics ...... 6 4.2 RadioIogical Impact ...... 6 5.0 RADIOLOGICAL PROTECTION...... 8 6.0 CONCLUSIONS...... 9 7.0 REFERENCES ...... 9 Revision 29 Date 1/11/02 I I I H- 6 L_ BVY 99-80 I Attachment 1 I Page 2 LIST OF TABLES Table 1: Radioanalytical Results of Composite Soil Samples...... lo Table 2: Estimated TotaI Radioactivity in Soil Volume...... 10 Table 3: Total Activity on South Field Mer Last Spreading Event...... 11 Table 4: Total Projected Radioactivity Remaining on South Filed at License ... Exp~~ona...... :...... ,;....*...... I1 Table 5: All-Pathway Critical Orgadwhole Body Dose Conversion Factors During Vermont Yankee Control of Disposal Site ...... 11 Table 6: MI-Pathway Critical Organ/WhoIe Body Dose Conversion Factors Post . Vermont Yankee Control of DisposaI Sites (Inadvertent Intruder) ...... 12 Table 7: Dose Contribution fiom CO-60 and Cs-137 in Soil Volume &a Land Spreading ...... 12 Table 8: Present and Future Dose Impact Due to the SoiI Spreading for Two Cases...... :...... 13 IIRevision 29 Date 1/11/02 I - BVY 99-80 I Attachment 1 1 Page 3 1.0 INTRODUCTION 1.1 Background In 1989, Vermont Yankee Nuclear Power Corporation requested approval from the NRC to routinely dispose of slightly contaminated septic waste in designated on-site areas in accordance with lOCFR20.302(a). Approval fiom the NRC was granted on August 30, 1989 and the information was permanently incorporated into the Offsite Dose Calculation Manual (ODCM)as Appendix B. In 1995, Vermont Yankee Nuclear Power Corporation requested that the previous authorization for on-site disposal of very low-level radioactive material in septic waste be amended to permit the on-site disposal of slightly con-ted cooling tower silt material. Approval fiom the NRC was granted on June 18,1997 and the information was permanently incorporated into the ODCM as Appendix IF. In 1994, approximately 25.5 rn3 of excess soil was generated during on-site construction activities. SampIing of the soil revealed low levels of radioactivity that were similar in radionuclides and activity to the septic waste and cooling tower silts previously encountered. An evaluation determined that the soil could be managed in similar fashion as the septic waste and cooIing tower silts; however, prior approval fioom the NRC would be required under 10 CFR 20.2002 (formerly 20.302(a)). 1.2 Objective The objective of this assessment is to present the data and radiological evaluation to demonstrate that the proposed disposition of the soil will meet the existing boundary conditions as approved by the NRC for septic waste and cooling tower silt. The boundary conditions established for disposal of the septic waste and cooling tower silts on the designated plots are: The dose to the whole body or any organ of a hypothetical maximally exposed individual must be less than 1.0 medyr. Doses to the whole body and any organ dose of an inadvertent intruder fiom the probable pathways of exposure are less than 5 mredyr. Disposal operations must be at one of the approved on-site locations. Revision 29 Date 1/11/02 I I I H- 8 BVY 99-80 I Attachment 1/ Page 4 2.0 WASTE DESCRIPTION The soil that is the subject of this evaluation was derived from excavations resulting fiom activities associated with a new security fence along the laut’s Protected Area boundary, The volume of soil generated was approximately 25.5 mP , and is typical of fill material containing light to dark brown poorly sorted soils with some small stones, and includes small incidental pieces of asphalt. The soil was removed from its original location by shovel, backhoe and fiont-end loader, and placed into dump trucks for transport to the location between the cooling towers where it was deposited on the ground surface and covered to prevent erosion. This location was selected because it was away fiom areas routinely occupied by pIant staff, and could easily be controlled. The most probable source of the low levels of radioactive contamination is the presence of below detectable removable contamination redistributed by foot traffic from inside the plant to walkways and parking areas. Subsequent surface ruxioff carries the contamination to nearby exposed soil near the Protected Area boundary where it had accumulated over time to low level detectable concentrations. In April 1995, a total of 20 composite soil samples were collected to characterize the volume. Composites were obtained by collecting a grab sample fiom one side, the top and the opposite side at equal distances along the length of the pile, then combining the thee into one sample. Soil samples were sent to the Yankee Atomic Environmental Laboratory for analysis and counted to environmental lower limits of detection required of environmental media. ResuIts of the andyses are presented in Table 1. AualyticaI results are provided for when the samples were collected and decay corrected to the present. The results identified both Cs-137 and Co-60 in most of the composite sampIes, which verified that plant-related radioactivity was present in the soil. For the purpose of estimating the total activity in the soil pile, the actual analytical result was used for those samples that were less than the MDC to calculate the average radioactivity concentration. The mass of soiI (dry) was estimated by multiplying the totaI in-situ volume (25.5 m3)by its wet density, 1.47E+03 kg/m3, and then dividing by the wetdry ratio of I .12, thus yielding a mass of 3.35E+04 kg (dry). The mass of the soil was then multiplied by the average Co-60 and Cs-I37 concentrations measured in the soil to obtain the total activity of each radionuclide in the 25.5 m3. Table 2 presents the estimated total radioactivity in the 25.5 m3 of soil at the time of sample collection and analysis, and decay corrected to the present. H- 9 Revisloll20 Date 1/11/02 ~ & BVY 99-80 1 Attachment 1 J Page 5 3.0 SOIL DISPOSAI, AND ADMINISTRATTVE PROCEDURE REQuxxIlEbD3NTS The method of soil disposal will use the technique of land spreading in a manner consistent with the current commitments for the on-site disposal of septic waste and cooliig tower silts as approved by the NRC, The accumulation of radioactivity on the disposal plot for this soil spreading operation will be treated as if cooling tower silt or septic waste was being disposed of since the characteristics of all these residual solids are similar (earthen-type matter), The south field (approximately 1.9 acres in size) designated and approved for septic waste and cooling tower silts disposal has been used for all past disposal operations, and will be used for the placement of this soil. Determination of the radiological dose impact has been made based on the same models and pathway assumptions used in the previous submittals. Dry soil material will be dispersed using typical agricultural dry bulk surface spreading practices in approved disposd areas on-site. Incidental pieces of asphalt and stones that were picked up with the soil from area where paving ran along the fence hewill be screened out before the soil is spread and disposed of as radioactive material at an off-site licensed facility if detectable radioactivity is found. Records of the disposal that will be maintained include the following: (a) The radionuclide concentrations detected in the soil (measured to environmental lower limits of detection). (b) The total volume of soil disposed of. (c) The total radioactivity in the disposal operation as well as the total accumulated on each disposal plot at the time of spreading. (d) The plot on which the soil was applied. (e) Dose calculations or maximum allowable accumulated activity determinations required to demonsbate that the dose limits imposed on the land spreading operations have riot been exceeded. To ensure that the addition of the soil containing the radioactivity will not exceed the boundary conditions, the total radioactivity and dose calcdation will include all past disposals of septic waste and coohg tower silt containing low-level radioactive material on the designated disposal plots. In addition, concentration limits applied to the disposal of earthen type materials (dry soil) restrict the placement of small volumes of materials that have relatively high radioactivity concentrations. Any farmer leasing land used for the disposal of soil will be notified of the applicable restrictions placed on the site due to the spreading of low level contaminated material. These restrictions are the same as detailed for the previously approved septic waste spreading application. Revision 29 Date 1/11/02 H-10 L_ BW99-80 I Attachment I I Page 6 4.0 EVALUATION OF ENVIRONMENTAL IMPACTS 4.1 Site Characteristics The designated disposal site is located on the Vermont Yankee Nuclear Power Plant site and is within the site boundary security fence. The south field consists of approximately 1.9 acres and is centered approximately 1500 feet south of the Reactor Building. This parcel of land has been previously approved by the NRC for the land disposal of septic waste and cooling tower silt. 4.2 Radiological Impact The amount of radioactivity added to the south field soil is procedurally controlled to emure that doses are maintained witbin the prior approved limits of the boundary conditions. To assess the dose received (after spreading the soil) by the maximally exposed individual during the period of plant controls over the property, and to an inadvertent intruder after plants controls of access ends, the same pathway modeling, assumptions ind dose calculation methods as approved for septic and cooling tower silt disposal were used. These dose modeIs implement the methods and dose conversion factors as provided in Regulatory Guide 1.1 09. The following six potential pathways were identified and included in the andysis: (a) Standing on contaminated &ound. (b) Inhalation of resuspended radioactivity. (c) Ingestion of leafy vegetables. (d) Ingestion of stored vegetables. (e) Ingestion of meat. (0 Ingestion of cow’s milk. Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways; the difference between them is due to the occupancy time: The basic assumptions used in the radiological analyses include: (a) Exposure to ground contamination and re-suspended radioactivity is for a period of 104 hours per year during the Vermont Yankee active control of the disposal sites and continuous thereafter. The 104-hour interval is representative of a farmer’s hespent on a plot of land (4 hours per week for 6 months). Revision 29 Date 1/11/02 I B-11 - BVY 99-80 / Attachment 1 /Page 7 (b) For the purpose of projecting and’illusmtingthe magnitude of dose impacts over the remaining life of the plant, it is assumed that the concentration levels of activity as of April 1,1999 remain constant. Table 1 indicates the measured radioactivity leveIs for CO-60 and Cs-137first noted in the soil, and decay corrected to April 1,1999. (b) For the analysis of the radiological impact during the Vermont Yankee active control of the disposal sites until 20 13, no plowing is assumed to take place and all dispersed radioactive material remains on the surface forming a source of unshielded direct radiation. (c) The crop exposure time was changed fiom 21 60 hours to 0 hours to reflect the condition that no radioactive material is dispersed directly on crops for human or animal consumption. Crop contamination is only through root uptake. (d) The deposition on crops of re-suspended radioactivity is insignificant. (e) The pathway data and usage factors used in the analysis are the same as those used in the Vermont Yankee’s ODCM assessment of off-site radiological impact fiom routine releases, with the following exceptions. The fraction of stored vegetables grown on the contaminated land was conservatively increased fiom 0.76 to 1.0 (at present no vegetable crops for human consumption are grown on any of the approved disposal plots). Also, the soil exposwe time to account €or buildup was changed from the standard 15 years to 1 year. (f) It is conservatively .assumed that Vermont Yankee rdiquishes control of the disposal sites after the operating license expires in 2012 (i.e., the source term accumulated on a singIe disposal plot applies also for the inadvertent intruder). (g) For the analysis of the impact after Vermont Yankee control of the site is relinquished, the radioactive material is plowed under and forms a uniform mix with the top six inches of the soil; but nonetheless, undergoes re- suspension in the air at the same rate as the unpIowed surface contamination. However, for direct ground plane exposure the self- shielding due to the six-inch plow layer reduces the surface dose rate by about a factor of four. I Revision29 Date 1/11/02 B-12 A BVY 99-80 / Attachment 1 f Page 8 As shown in the previous submittals, in which the concentrations of Co-60 and Cs-137 in septic waste exceed those identified in the soil, the liquid pathway was found to be an insignificant contributor to the dose. Therefore, the liquid pathway is not considered in this analysis. The dose models and methods used to generate deposition values and accumulated activity over the operating life of the plant are documented the ODCM. Table 3 presents the radioactivity that currently exists on the south field after the Iast spreading event which occurred on September 28,1998 (total elapsed time fiorn September 28,1998 to April 1 , 1999 is 184 days). In addition, the total activity on the south field is presented assuming the addition of the 25.5 m3 of soil subject of this evaluation. The total activiw that would be present ofi south field at license termination (Le., total elapsed time of 14 years post April 1,1999, or 2013), assuming no future additions of material containing radioactivity after disposal of the proposed soil volume was also evaluated and is presented in Table 4. In order to demonstrate compliance with the boundary conditions, the critical organ and whole body dose from all pathways to a maximally exposed individual during Vermont Yankee control, and to the inadvertent intruder were calculated. The dose calculations were performed using the dose conversion factors presented in TabIe 5 and 6 beIow which were obtained fiom the ODCM. The contribution to dose fiom Co-60 and Cs-137 to the whole body and organ at the present and at license expiration is presented in Table 7. The present and future total dose impact fiom the south field with and without spreading of the soil is presented in TabIe 8. These resdts demonstrate that disposal of the approximately 25.5 m3of accumulated soil will be well within the accepted dose limit criteria of 1 mremfyr to any organ or whole body during the control period, and 5 mredyr to an inadvertent intruder after control of the site is assumed to be relinquished. This analysis shows that significant dase margin still exists on the approved disposal plots to accommodate potential future spreading operations. 5.0 RADIOLOGICAL PROTECTION The disposal operation of soil piles will follow the applicable Vermont Yankee procedures to maintain doses as low as reasonably achievable and within the specific dose criteria as previously approved for septic waste and coohg tower silt disposal. I Revision 29 Date 1/11/02 H-13 - BVY 99-80 / Attachment 1 I Page 9 6.0 CONCLUSIONS Soil generated from on-site construction activities reflects an earthen type material similar in characteristics to septic waste residual soIids and cooling tower silt with respect to the radiological pathway behavior and modeling. Based on the similarity in characteristics between the proposed soil volume and waste streams that have already been approved for disposal, and the evaluation of the incrementaI dose impact, it is concluded that disposal of the approximately 25.5 m3 of existing soil through on-site land spreading wiIl meet the boundary conditions specified in the ODCM. That is, with respect to the addition of the approximately 25.5 m3soil pile to the existing radioactivity already spread on the south field: 1. Total doses to the whole body and critical organ to the hmothetically maximally exposed individual were estimated & 3.00E-02 mredyr and 1.04E-01,respectively, which are Iess than the prescribed 1.0 mredyr. 2. Total doses to the whole body and critical organ of an inadvertent intruder fiom the probable pathways of exposure were estimated as 1.13E-01 mredyr and 2.21E-01 mredyr, respectively, which are less than 5 mredyr. 3. The disposal is assumed to take place on the south field that is the same site approved for disposal of septic waste and cooling tower silts. If the soil were spread on an approved plot which had not yet been used for disposal, the dose impact from the approximately 25.5 m3 of soil alone would at present be 4.17E-03 mredyr whole body and a maximum organ dose of 1.46E-02 mredyr. In addition, for the inadvertent intruder, the whole body dose would be 1.60E-02 mrerdyr, and a maximum organ dose of 3.1 IE-02 mredyr. Each of these doses also meet the boundary conditions specified in the ODCM. 7.0 REFERENCES (1) Vermont Yankee ODCM,Rev 23, Appendix B, “Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee”. (2) Vermont Yankee ODCM, Rev 23, Appendix F, “Approval Pursuant to 1OCFR20.2002 for On-Site Disposal of Cooling Tower Silt”. (3) USNRC Regulatory Guide I .109, Rev 1, “Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance With 1OCFR Part 40, Appendix I”, dated October 1997. Revision 29 Date 1/11/02 I1 H-14 L BVY 99-80 / Attachment 1 / Page 10 Table 1 Radioanalytical Results of Composite Soil Samples Muximum Vdue 810 739 143 84 Minimum Value I07 98 5 3 Standard Deviation 191 174 37 22 * Average wet to dry sample weight ratio equal to 1.12. Average wet density equal to 1.47 gdcc. ** The apparent response of the gamma isotopic analysis which was lcss than Minimum Detectable Concentration. - Table 2 Estimated Total Radioactivity in Soil Volume I I Volume I t Average Concentration I Total Activity of Soil Mass 6Ci/kg-dry) (m Radionuclide (m3) (kg-dry) - April, 1995 April, 1999 ApriI,l995 April, 1999 CS-137 25.5 3.35E-i-04 328 299 11.0 10.0 CO-60 25.5 3.35Ei-04 85 50 2.8 1,.7 Revision29 Date 1/11/02 ~ H-15 7 - BVY 99-80 / Attachment 1 / Page 11 Total activity after Total activity decay Total activity after Iast spreading event corrected to April 1, proposed soil Radionuclide (@/acre) ,1999(pCi/acre) disposal OrWacre) Mn-54 0,17 0.1 1 0.1 1 Co-60 5.93 5.5s 6.44 Zn-65 0.074 0.044 0,044 CS-137 32.27 3 1.90 37.16 Table 4 Total Projected Radioactivity on South Field Remaining at License Expiration Total Activity as of License Expiration Radionuclide (pCVacre) - Mn-54 8.9E-07 Co-60 0.89 211-65 7.6E-09 CS-137 26.33 Table 5 Critical Organ Whole Body Dose Factor Dose Factor Radionuclide IndividuaVOrgan (mtemlyr per pCi/acre) (mrernlyr per pCi/acre) Mrl-54 AdddGI-LLI 3.75E-04 I .93E-04 CO-60 TeenLung 7.17E-04 5,3 1E-04 Zn-65 Child/Liva 1 A4E-02 1 .a3~-02 I1Revision 29 Date 1/11/02 H-16 P BW99-80 / Attachment 1 /Page 12 CriticaI Organ Whole Body Dose Factor Dose Factor Radionuclide IndividuaYOrgan (mremlyr per @Xiacre) (rnnmlyr per pcdacre) Mn-54 Teen/Lung 1.02E-02 3.12E-03 CO-60 Teen/Lung 3.19E-02 9.09E-03 Zn-65 ChildCLiver 1.WE-02 1.258-02 CS-137 ChildiBone 6.98E-03 3.8 5E-03 Table 7 Dose Contribution from Co-60 and (3-137in 25.5 m3 Soil Volume after Land Spreading Present Dose Impact' Future Dose Impacv (Maximally exposed (Inadvertent Intruder) Case individual) Dose Individual/ Dose Individual/ (mrem/yr) Organ (mredyr) Organ Cobalt-60 4.75E-04 Whole body 1.29E-03 Whole body 6.42E-04 Max. Organ 4.51E-03 Max. Organ Cesium-137 3.69E-03 Whole body 1.47~-02 Whole body 1.40E-02 Max. Organ 2.66E-02 Max. Organ Based on inventory of Co-60 of 0.895 pCi/acre and Cs-137 of 5.26 pCllacre in Apnl 1, 1999. Based on inventory of CO-60of 0.14 1 pCi/acre and Cs-137 of 3.82 pcilacre in April 1, 2013. Revision 29 Date 1/11/02 1 I I H-17 BVY 99-80 / Attachment 1f Page 13 Table S, Present and Future Dose'Jrnpact Due to the Soil Spreading for Two Cases Present Dose Impact (Maximally exposed Case individual) . Dose Individual/ Dose Individual/ (mremEyr) Organ (mredyr) Organ Case One South Field as it currently 2.58E-02 Whole body 9.7OE-02 Whole body exists 8.9%-02 MS. brgm 1.89E-01 Max. Organ Case Two South Field if disposal of 3.00E-02 Whole body 1.13E-01 Whole body soil volume is approved 1.04E-02' Max. Organ 2.2 1E-0 1 ChiWBone Increase in dose impact 4.17E-03 Whole body 1.60E-02 Whole body fiom disposal of soil 1.46E-02 Max. Organ 3.1 IE-02 Max. Organ Revision 29 Date 1/11/02 ll I ' k-18 VERMONTYANKEE . NUCLEARPOWER CORPOFWTION I 185 Old Ferry Road, Brattleboro, VT 05301-7002 / (802)257-5271 January 4,2000 BW00-02 United States Nuclear Regulatory Commission ATM: Document Control Desk Washington, DC 20555 References: (a) Letter, VYNPC to USNRC, "Request to Amend Previous Approvals Granted Under 10 CFR 29.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil," BVY 99-80, dated June 23,2999 Subject: Vermont Yankee Nuclear Power Station ' License No. DPR-28 Docket No. 50-271) Supplement to Request to Amend Previous Approvals Granted Under 10 CFFt 203021a) to Allow for Disposal of Contaminated Soil Reference (a) provided Vermont Yankee's application to amend the previously granted approvals to dispose of slightly contaminated septic waste and cooling tower silt on-site to include slightly cbntaminated soil generated as a residua1 by-product of on-site construction activities. The request was to allow the disposal of approximately 25.5 cubic meters of waste that has been accumuIated to date and to allow for disposal of future waste from construction related activities. Based on discussions with USNRC staff, additional information reIated to the estimated volume and dose consequences of the proposed future material was needed to complete your review. . Attachment 1 has been revised accordingly to include the .infomation requested. Attachment 1 supercedes the evaluation previously submitted. We trust that the information will allow you to compIete your review of our submittal. However, should you have any questions or require further information concerning this matter, please contact Mr. Jim DeVincentis at 802-258-4236. ? Sincerely, Vermont Yankee Nuclear Power Corporation Gautam Sen 1Licensing Manager Attachment cc: USNRC Region I Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS VT Deparbnent of Public Service . H-19 IIRevision 29 Date 1/11/02 7 -. SUMMARY OF VERMONT YANKEE COMilMTMENTS BVY NO.: 00-02 The foUowing tabIe identifies commitments made in this document by Vermont Yankee. Any other actions discussed in the submittal represent intended or planned actions by Vermont Yankee. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager of any questions regarding this document or any associated commitments, COMMITMENT COMMLTIXD DATE 0R"OUTAGE'' . None NIA 11 Lb VYAPF 0058.04 AP 0058 OriginaI Page 1 of 1 Revision 29 Date 1/11/02 H-20 llI .. I 1. Docket No. 50-27 I . BVY00-02 Attachment 1 Vermont Yankee Nuclear Power Station Assessment of On-site Disposal of Contaminated Soil by Land Spreadkg Revision 29 Date 111 1/02 H-21 7 b BVY 00-02/ Attachment I I Page 1 of 17 -' TABLE OF CONTENTS - Page LIST OF TABLES ...... 2 I .O,. INTRODUCTION ...... , ...... , , . , . , +. .. +, ., , . , ...... 3 1.1 Background ...... 3 . .I 1.2 Objective ...... , ...... -...... 3 ' 2.0 WASTE DESCRIPTION ...... ;:...... f.... 4 3.0 SOIL DISPOSAL AND ADMINISTRATNE PROCEDURE REQUIREMENTS...... , . . . . . + .: ...... ,...... 5 4.0 EVALUATION OF ENVIRONMENTALIMPACTS ...... 6 4.1 Site Chcteristics ,. , ...... :...... :. ,. . . . .;. 6 4.2 RadioIogicd Impact ...... 7 5.0 RADIOLOGICAL PROTECTION ...... 9 6.0 CONCLUSIONS ...... 9 7.0 REFERENCES ...... *....,,...... ,.,,...11 Revision 29 Date 1/11/02 I H-22 .-. BW00-02 / Attachment 1 /Page 2 of 17 _. LIST OF TABLES Table 1: Radioanalytical Results of Composite Soil Samples...... 12 Table 2: Estimated Total Radioactivity in 25.5 m3Soil Volume...... :. 12 Table 3: Estimated Total Radioactivity in Future Soil Additions...... 13 Table 4: Record of Septic and Silt Radioactive Material Spread each Year on South Field Disposal Plot ...... 13. Table 5: Total.Projected Radioactivity Remaining on South'Filed at License' . Expiration...... is...... I4 Table 6: All-Pathway Critical Orgadwhole Body Dose Conversion Factor&...... 14 Table 7: Dose Impact from Past Septic and SiIt Spreading Activity..:...... 15 Table 8: Dose Impact ftom Past Septic and Silt Spreading 'and Single 25.5m3 Soil ..... -. .. . -- ._- .- .- .. - . Disposal...... I5 Table 9: Dose Impact fiom Present and Future Soil Disposal Along with Past Septic and Silt Disposal,, ...... 16 TabIe 10: Dose Impact from Past Disposals through 7/15/99 Plus all Annual Projected 1 Disposals of Septic, Silt and Soil...... I6 .- Table 11 : Summary of Dose Impacts Associated with Different Disposal - Scenarios ...... :...... 17 Revision 29 Date 1/11/02 H-23 *. I I BW00-02 I Attachment I I Page 3 of 17 1.0. INTRODUCTION . 1.1 Background In 1989, Vermont Yankee Nuclear Power Corporation requested approval from the NRC to routinely dispose of slightly contaminated septic waste in designated on-site areas in accordance with lOCFR20.302(a). Approval fiom the NRC was granted on August 30,1989 and the information was permanently incorporated into the Offsite Dose Cdcdation Manual (ODCM) as Appendix E. In 1995, Vermont Yankee Nuclear Power Corporation requested that the previous authorization for on-site disposal of very low-level radioactive material in septic waste be amended to permit the on-site disposal of slightly contaminated coohg tower silt material. Approvd fiom the NRC was granted ori June 18,1997 and the information was permanently incorporated into the ODCM as Appendix F. In 1994, approximately 25.5 m3of excess soil was generated during on-site construction activities. Subsequent sampling and analysis of the soil revealed low levels of radioactivity that were similar in radionuclides and activity to the septic waste and cooling tower silts pfeviously encountered. An evaluation determined that the soil could be managed in similar fashion as the septic waste and cooling tower silts; however, prior approval from the NRC would be required under 10 CFR 20.2002 (formerly 20.302(a)). -. 1.2 Objective The purpose of this assessment is to present the data and radiological evaluation to demonstrate that the proposed disposition of the soil (Le., on-site disposal via land spreading on designated fields) will meet the existing boundary dose conditions as approved by the NRC for septic waste and cooling tower silt. The boundary conditions established for -’ disposal of the septic waste and cooling tower silt on designated plots are: I. The dose to the whole body or any organ of a hypothetical maximally exposed individual must be less than 1.O mredyr. 2. Doses to the whole body and any organ of an inadvertent intruder &om the probable pathways of exposure are less than 5 rnredyr. 3. Disposal operations must be at one of the approved on-site locations. I Revision 29 Date 1/11/02 H-24 V L BVY 00-02 / Attachment 1 / Page 4 of 17 2.0 WASTE DESCRIPTION The existing accumulation of contaminated soil was derived fkom excavation activities associated with the construction of a new security fence along the plant’s Protected Area boundary. The volume of soil generated was approximately 25.5 m3,and is typical of fill material containing light to dark brown poorly sorted soils with some small stones, and includes small incidental pieces of asphalt. The soil was removed from its original location by shovel, backhoe and fiont-end loader, and placed into dump trucks for transport to a location between the cooling towers where it was deposited on the ground surface and covered to prevent erosion. This location was selected because it was away fiom areas routinely occupied by plant sW,and could easily be controlled. The most probable source of the low levels of radioactive contamination is the presence of below detectable removable contamination redistributed by foot tr&c fiom inside the plant to walkways and parking areas. Subsequent Surface runoff carries the contamination to nearby exposed soil near the Protected Area boundary where it had accumulated aver time to low level detectable cdncentrations. In April 1995, a total of 20 composite soil samples were collected to characterize the accumulated volume. Composites were obtained by taking a grab sample from opposite sides of the pile and the top at equal distances along its length. These three grab samples were then combined into one composite sample. Soil samples were sent to the Yankee Atomic Environmental Laboratory for analysis and counted to environmental Iower Emits of detection required of environmental media Results of the analyses are presented in Table 1. For estimating the total activity in the soil pile, the actual analytical resdt was used for those samples that were Iess than the MDC to calculate the average radioactivity concentration. Analytical results are provided for both the times when the samples were collected as well as decay corrected to the present (7/15/99).The resdts identified both Cs-137 and CO-60 in most of the composite samples, which verified that plant-related radioactivity, was present in the soil. The mass of accumulated soil (dry) was estimated by multiplying the total in-situ volume (25.5 m3) by it’s wet density, 1.47E-tO3 kg/m3, and then dividing by the wet:* ratio of 1.12, thus yielding a mass of 3.35E+04 kg (dry). The mass of the soil wits then multiplied by the average measured CO-BO and Cs-137 concentrations to obtain the total activity of each radionuclide in the 25.5 m3.Table 2 presents the estimated total radioactivity in the 25.5 rn3 volume at the time of sample collection and analysis, and decay corrected to the date of the most recent disposal (septic waste) spreading operation @e., July 15,1999). I Revxion 29 Date 1/1 1/02 BW00-02 Attachment 1 i Page 5 of 17 evaluating the radiological impact of potential future soil disposals, it is assumed that an additional 28.3 m3 per year of sand / soil is contaminated at the same concentration levels as originally observed (April 1995) in the currently collected 25.5~11~of soiI. It is also assumed that this materid is placed on the same approved disposd fieId used for all past septic and cooling tower silt disposal operations. Table 3 shows the estimated amount of radioactivity associated with the annual disposal of the 28.3m3of soil. It is assumed that this material is disposed of each year for the next 14 years (until the end of pIant operating license in 20 13) on the same field (South Disposal Plot) dong with the continued application of septic waste and cooling tower silt. Table 4 shows a record of the acid amount of septidsilt. materid that has been spread on. south field for the past 10 years. A review of the actual waste disposal operations show that the annuaI average radioactivity content placed on the 1.9 &re South field horn septic and silt disposds are as follows: Mu-54 0.147 uCi/year CO-60 2.58 uCi/year Zn-65 0.269 uCi/year CS-134 0.010 uCiiyear CS-137 6.21 uciyear The maximum radioactivity inventory resulting fiom the accumulated buildup of past and projected future disposal operations (i.e., septic waste, cooling tower silt;plus the existing 25:5m3 of accumulated soil along with a projected annual addition of 28.3m' of soil each year until the termination of the operating license) is shown on Table 5. 3.0 SOIL DISPOSAL AND ADMlTyISTRATIvE PROCEDURE REQUIREMENTS The method of soil disposd will use the technique of land spreading in a manner consistent ** with the current commiments for the on-site disposal of septic waste and cooling tower silts as approved by the NRC. The accumulation of radioactivity on the disposal plot for this soil spreading operation will be treated as if cooling tower silt or septic waste was being disposed of since the characteristicsnf all these residual solids are similar (earthen-type matter). .The south field (approximateIy 1.9 acres in size) has been designated and approved for septic waste and cooling tower silt disposal and has been used for all past disposal operations, and is expected to be used for the placement of soil. Determination of the radiological dose impact has been made based on the same models and pathway assumptions used in the previous submittals. Dry soiI material will be dispersed using typical agricultural dry bulk surface spreading practices in approved disposal areas on-site. Incidental pieces of asphalt and stones that are picked up with the soil will be screened out before the soil is spread and disposed of as radioactive material at an off-site licensed facility if detectable radioactivity is found. H-26 Revision 29 Date 1/11/02 'I .. . BVY 00-02I Attachment 1 / Page 6 of 17 Records of the disposal that will be maintained include the following: .- (a) The radionuclide concentrations detected in the soil (measured to environmentd lower limits of detection). (b) The total voIume of soil disposed of. (c) The total radioactivity in the disposal operation as well as the total accumulated OR each disposal plot at the time of spreading. (d) The plot on which the soil was applied. (e) Dose calculations or npxhum allowable accumulated activity determinations required to demonstrate that the dose limits imposed on the land spreading operations have not been exceeded. .. To ensure that the addition of soil containiug low levels of adioactivity will not exceed the boundary dose limits, each new spreading operation will require an estimate of totat radioactivity that includes all past disposds of septic waste, cooling tower silt and soil material on the designated disposal plots. This will be compared with the boundary dose limits or equivalent radioactivity limits on a per acre basis, In. addition, concentration limits applied to the disposal of earthen type materials (dry soil) restrict the placement of small volumes of materials that have AativeIy high radioactiviv concentrations. Any farmer leasing land used for the disposd of soil (or other approved waste) will be notified of the applicable resbictions placed on the site due to the spreading of low level contaminated material. These restrictions are the same as detailed for the previously approved septic waste spreading application. 4.0 EVALUATION OF ENVIRONMENTAL EMPACTS 4.1 Site Characteristics All designated disposal sites are located OR the Vermont Yankee NucIear Power Plant site and are within the site boundary security fence. The south field consists of approximately 1.9 acres and is centered approximateIy 1500 feet south of the Reactor Building. This parcel of land has been previously approved by the NRC for the land disposal of septic waste and cooling tower silt, and is the only portion of the approved disposal areas which has been utilized to-date for the spreading of contaminated material. For estimating the rnaxhw future radiological impact, it is assumed in the analysis that all future disposal operations Will continue to use the South field as the disposal plot. Revision 2 Date 1/11/02 H-27 I I BVY 00-02 1 Attachment 1 I Page 7 of 17 4.2 Radiological Impact _. The amount of radioactivity added to the south field is procedurally controlled to ensure that doses are maintained within the prior approved limits of the boundary conditions. To assess the dose received by the maximaIIy exposed individuaI during the period of plant controls over the property, and to an inadvertent intruder after it is assumed plat access controIs end, the same pathway modeling, assumptions and dose calculation methods as approved for septic and cooling tower silt disposd were used. These dose models implement the methods and dose conversion factors as provided in Regulatory Guide 1.109, Revision 1 (1 977). The following six potential pathways were identified and &cIuded in the analysis,: (a) Standing on contaminated ground. . (b) Inhalation of resuspended radioactivity. (c) Ingestion of Ieafl vegetables. (d) Ingestion of stored vegetables. (e) Ingestion of meat. .+ (f) Ingestion of cow’s milk. As shown in the previous application for septic waste disposd, the liquid pathway was found to be an insignificant contributor to the dose for the radionuclides idenaed fixed in the soil type matrixes associated these waste forms. Therefore, the liquid pathway is not considered in this analysis. Both the maximum individual and inadvertent intruder are assumed to be exposed to these pathways, the difference between them being due to the occupancy time. The basic assumptions used in the radiologicd analyses include: (a) Exposure to ground contamination and re-suspended radioactivity is for a period of 104 hours per year during the Vermont Yankee active control of the disposal sites and continuous thereafter. The 104-hour interval is representative of a farmer’s time spent on a plot of land (4 hours per week for 6 months). H-28 Revi\ion= Date 1/11/02 ~ BW00-02 I Attachment 1 I Page 8 of I7 occurs at the saqe time (2013) for both the Conk01 Periodand Intruder scenarios. (c) For the analysis of the radiologicd impact during the Vermont Yankee active control of the disposal sites until 2013, no plowing is assumed to take place and aI1 dispersed radioactive material remains on the surface forming a source of unshielded direct radiation. (d) The crop exposure &ne was changed from 2160 hours to 0 hours to reflect the condition that no radioactive material is dispersed directly on crops for human or animal consumption. Crop contamination is only through root uptake. I (e) The deposition on crops ofxe-suspended radioactivity is insignificant. (f) Most of the pathway data and usage factors used in the analysis are the same as those used in the Vermont Yankee's ODCM assessment of off-site radiological impact fiom routine releases. The hction of stored vegetables grown on the contaminated land was conservatively increased fiom 0.76 to 1 .O (at present no vegetable crops for human consumption are grown on any of the approved disposal plots). For each year's spreading operations, the soil exposure time to account for buildup was changed from the standard 15 years to 1 year. .. - . (g) It is conservatively assumed that Vermont Yankee relinquishes control of the disposal sites after the operating license expires in 2013 (Le., the source term accumuIs?ted on a single disposal plot applies also for the inadvertent intruder). (h) For the analysis of the impact derVermont Yankee control of the site is relinquished, the radioactive material is plowed under and forms a uniform * mix with the top six inches of the soil; but nonetheless, undergoes re- suspension in the air at the same rate as the unplowed surface contamination. However, for direct ground plane exposure the self-shielding due to the six- inch pIow layer reduces the surface dose rate by about a factor of four. . The dose models and methods used to generate deposition values and accumulated activiw over the operating life of the plant are documented in the Vermont Yankee ODCM. The total activity that would be present on south field at the end of the operating period (Le., total elapsed time of 14 years post July 15,1999, or 20 13) from the buildup of all waste streams @e., septic, cooling tower silt and soil) is presented in Table 5. Revision 29 Date 1/11/02 I H-29 BW00-02 I Attachment 1 I Page 9 of I7 In order to evaluate the dose iqpact associated with the different disposal streams, a dose assessment was performed for the following four disposal scenarios: 0) Impact fiorn past septic and silt spreading only - Table 7 (II) Impact fiom past septic and silt spreading, plus a single 25.5m3soil disposal operation for the existing accumulated soil - Table 8 (III) Impact from past septic and silt disposals along with the existing 25.5 m3of accumulated soil and postulated future annual soil disposal volumes (28.3 m3 /y).-Table 9 0 Impact from past septic and silt disposals plus annual projected disposals of septic, silt and soil. -Table 10 For each scenario, the critical organ and whole body dose Gom all pathways to a maximally exposed individual for both the Vermont Yankee control period and the inadvertent intnzder situation were calculated. The dose calculations were performed using the dose conversion factors presented in Table 5, which were obtained fiom the Vermont Yankee ODCM, Appendix F, “Approval Pursuaut to lOCFR20.2002 for On-Site Disposal of Cooling Tower Silt.” A: summary of the cdcdated dose impact associated with the four different scenarios is shown in Table 11. These results demonstrate that disposal of the 25.5 m3 of accumulated soil will be well witbin the accepted dose limit criteria of 1 mredyr to any organ or whole body during the control period, and 5 mredyr to an inadvertent intruder. In addition, if continued soil spreading is necessary, the resulting dose is expected to dso remain below the established limits even assuming the annual application of already approved disposal media (Le., septic waste and cooling tower silt). + 5.0 RADIOLOGICAL PROTECTION l. The disposal operation of the soil will follow the applicabIe Vermont Yankee procedures to maintain doses as low as reasonably achievabk and within the specific dose criteria as previously approved for septic waste and cooling tower silt disposal. 6.0 CONCLUSIONS Soil generated fiom on-site construction and maintenance activities constitutes an earthen me material similar in characteristics to septic waste residual solids and cooling tower silt with respect to the radiological pathway behavior and modeling. Based on the similarity in characteristics between the proposed soil voIume and waste streams that have already been approved for disposal and the evaluation of the incremental dose impact, it is concluded that the disposal of the existing 25.5 m3 and the projected 28.3 m3/yearof soil through on-site land spreading will meet the existing NRC approved boundary dose conditions specified in H-30 I Revision 29 Date 1/11/02 3W00-02 f Attachment 1 I Page 10 of 17 the Vermont Yankee ODCM (see Appendix B for Septic Waste Disposal). That is, with respect to the addition of the initial 25.5 m3of soil along witti the projected 28.3 m3/yearof soi1 and the projected future disposal of septic and silt waste to the existing radioactivity already spread on the south field 1. Total doses to the whoIe body and criticd organ of the hypothetically maximally exposed individual were estimated as 1.15E-01 mredyr and 4.03E-01 medyr, respectively, which are Iess than the prescribed 1.O mredyr limit during the period of active site contxol. . 2. Total doses to the whoIe body and critical organ of an inadvertent intruder fiom the . probabIe pathways of exposure were estimated as 7.57E701 mredyr and 1-17mredyr, respectively, which are Iess than 5 mdyr limit associated with an intruder scenario following assumed loss of site accesS control as the end of the operating license. 3. For purposes of projecting maximum impact, all disposals (past and We)are assumed to take place on the south disposal plot. Therefore, the disposition of the present 25.5 m3 and the projected 28.3 m3/yearof soil will chimeto meet the existing boundary conditions as approved by the NRC for septic waste and cooIing tower silt. .- - Revision 29 Date 1/11/02 #I I I -_ BW00-02 f Attachment 1 I Page 1 l’of 17 .. 7.0 REFEFCENCES (1) Vermont Yankee ODCM, Rev 23, Appendix B, “Approval of Criteria for Disposal of Slightly Contaminated Septic Waste On-Site at Vermont Yankee”. , (2) Vermont Yankee ODCM, Rev 23, Appendix F, “Approval Fbrsuant to lOCFR20.2002 . for On-Site Disposal of Cooling Tower SiIt”. (3) USNRC Regulatory Guide 1.109, Rev I, “Calculation of Amd Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evalua~gCompliance with ZOCFR Part 40, Appendix I”, dated October 1977. .. . -- ._ - . . . H-32 Revision 29 Date 1/11/02 - BVY 00-02 / Attachment 1 J Page 12 of 17 TabIe 1 RadioanaIytical Results of Composite Samples Taken from 25.5m3Soil Pile CS-137 CO-60 @Ci/kg) (Pew Smule TD Ad. 1995 July 15,1999 Auril. 1995 July 25.1999 (322716 234 212.2 49 28.0 * 022717 522 473.4 143 81.7 G22718 .- 337 305.7 37 21.1 * G22719 29 1 263.9 111 63.4 G22720 348 315.6. , 47 26.8 * . G22721 135 122.4 ; 73 41.7 G22722 107 97.0 82 46.8 G22723 222 20 1.4 140 80.0 G22724 180 163.3 92 52.6 G22725 269 244.0 118 67.4 G22726 810 734.7 . 114 65.1 G22727 378 342.8 106 60.6 G22728 810 734.7 124 . 70.8 G22729 376 341 .O 62 35.4 G22730 33 1 300.2 87 49.7 G2273 1 253 229.5 5 2.9 * G22732 150 136.0 58 33.1 G22733 247 224.0 105 60.0 G22734 326 2957 54 30.8 * G2273 5 23 5 213.1 100 57.1 Average 328 298 85 49 Maximum 810 735 143 82 Minimum 107 97 5 3 StdDev. 186 169 36 20 *The apparent response of the gamma isotopic analysis was less than the Minimum DetectabIe Conceneation. Table 2 Estimated TotaI Radioactivity in 25.Sm3Accumulated Soil ".. - _-. H-33. . I Revision 29 Date 1/11/02 .-. . -I. . -_ ? A BVY 00-02 /Attachment 1 /Page I3 of 17 Table 3 Estimated Total Radioactivity in Future Soil Additions Volume Soil Average Concentration of soiI Mass (pcfig-dry assuming) Total Activity Nuclide (m3) Ike-drvl cA~riI.1995concentrations1 lUCi/VT) Cs-137 28.3 3.72E+04 328 12.8 . Co-60 28.3 3.72E+04 85 3.16 Table 4 Record of Septic and Sit Radioactive Material Spread Each Year on the South Field . ,I Spreading Materid Mn-54 Cod0 ZU-65 0-134 c5137 Ce-141 Year Date Type (uCi/acre) (uCVacre) (uCi/acre) (ucilacre) (nCilacre) (uCi/acre) t990 10/3 1/90 Septage 0.00 3.89 0.00 0.00 026 0.00 1990 11/20/90 Septage 0.17 2.03 0.41 0.00 0.29 1.40E-08 1991 no 0.00 0.00 0.00 0.00 0.00 0.00 spreading 1992 10/19/92 septage 0,ll 1.73 0.52 0.05 0.32 0.006 1993 10/14/93 septage 0.05 1.41 0.2 I 0.00 0.30 0.00 1994 06/14/94 septage 0.08 .0.43 0.00 0.00 0.09 0.00 1995 06/29/95 septage 0,OO 0.88 0.00 0.00 0.00 0.00 1996 no 0.00 0.00 0.00 0.00 0.00 .a00 spreading 1997 06/18/97 septage 0.12 1 .oo 0.00 0.00 0.19 0.00 1998 omo/ga septage- 0.14 0.72 0.09 0.00 0.12 0.00 1998 09/28/98 Silt 0.00 0.00 0.00 0.00 30.87 0.00 1999 07/15/99 Septage-- 0.1 1 1.47 0.20 0.00 0.25 0.00 4verage Activitylyr (UCVacre): 0.08 1.36 0.14 0.01 3.27 0.0 1 Average Activity (uCi/yr) to be 0.147 2.58 0.269 0.010 6.21 0.001 not disposed of on 1.9 significant acre field H-34 Revision 29 Date 1/11/02 BW00-02 1 Attachment I I Page 14 of 17 Table 5 Total Projected Radioactivity Remaining on South FieId after'license Termination Accum. Activity Accum. Activiw Accum. Activity Accum. Activity in Silt & Septic in soil Total All Paths Total All Paths @ Year 2013 @ year 2013 @ year 2013 @ Year 2013 Nuclide a :luci') cuci) ~ (uCi/acre>* Mn-54 , 026 - - 0.26 0.14 CO-60 19.68 21.83 41.51 21.85 - 0.42 Zn-65 0.42 .. 0;22 CS-137 119.03 154.74 273.78 144.09 CS-134 0.04 - ~' ' 0.04 0-02 * The total activity is assumed to be spread on the 1.9 acre South field to generate the uCi/acre value. Table 6 AII Pathway Critical Organ/WhoIe Body Dose Conversion Factors During W Control Post W Control putruder Scenario) Critical Organ Whole Body Critical Organ Whole Body . Dose Factor Dose Factor Dose Factor Dose Factor (mremlyr per (mremlyr per Lmremfvr uer (mredyr per Radionuclide XndividuaVOrvan yciacre) pcilacre) $ifacre) pciiacre) - Mn-54 AdddGI-LLI 3.75E-04 1.938-04 1.02E-02 3.12E-03 CO-60 Teedung - - 7. I E-04 5.31E-04 3. I9E-02 9.09E-03 211-65 ChiId/L.iver I .62E-02 1.03E-02 1.89E-02 . 1.25E-02 cs-I37 Child50ne 2.66E-03 7.02E-04 6.98E-03 3.85E-03 Revision 29 Date 1/11/02 ' H-35 ... ' :', L BVY 00-02 1 Attachment 1 I Page 15 of 17 Table 7 (Scenario I) Dose Impact from Past Septic and SiIt Spreading on South Field (as of 7/15/99) Control Sccnarfa: AI1 Other Spdmgs HaISLifC To Datc .fEiEl fUCiiaCrrl I~Wd~l~mrrmh.rluc'iacre) fmdwl frnrcm/vr\ . Mn-54 086 030 3.7sE04 133844 73935 3.788-05 CoaO 527 6.86 7.17E-04 531E-04 4.m-03 3.63E-03 zp-65 .,0.67 023 - 1.62E-M 1.03EM 3.77E-03 2.40E-03 Cs-137 30.17 3 192 2.6633 .7.m-04 8.49E-02 224E-02 Total. 937E-02 . I 2.85E-02 .. I I Dose Limit: 1' I ?4of Limit 937% 2.85?6 Intruder Scenario: AI1 mer Activity on Plat Spdmgs Decayed to MaximumOrgm WholeBody Maximum WhoteBody Half-Life To Datc Year2013 Dost Factor DoritFactDr OrganDosc Dose Iyears) ~uciiacrc~ ~UCiacfcl ~mrrmlvr~d~~IdvKd~] irnrcrn/yr> {mrcdvrl Mn-54 0.86 020 23 IE-06 1.02EM 3.12EM 236E-08 722E-09 CQ-60 527 6.86 1.OXE+OO 3.19E-02 9.09E-03 3.46E-02 9.85E-03 20-65 0.67 0.23 1.1 SE-07 1.89EM 1.2SE4Q 217E-09 I.43E-09 - c5-137 30.17 31.92 231EM1 698E-03 3.uEm 1.62E-O 1 8.9 IE-OZ . _,I_.. .. % OfLimit 3.92% 1.98% Table 8 (Scenario IC) Dose Impact from Past SeptidSilt Spreading and Single 25.5m3 Soil Disposal Controt Scenario: All spreaftiigs Half-Life toDate' r(eassl Mn-54 0.86 -0 527 7.70 7.17E-04 5.3 IE-04 552E-03 4.09E-05 zn-65 0.67 0233 1.&?E42 1.03E-02 3.77E-03 2.10E-03 CS-I37 30.17 37.19 2.66E-03 7.02E-04 9.89E-M 2.6 1E-02 Totalal~~~:1 1.08E-01 I 326E-E 1 Dose Liir 1' 1 % of Liir 10.83% 3.2694 Intruder Sctnaria: All Spreadiigs Activity on Plot Maximum Orgm Wholc Body Maximum Whole Body Half-Life tom Decaycdto2013 DoscFaaor Dose Factor Organ Dose Dose cycan) luc'iaml fucilaerc) ~mmrtmlvr/uciiaete) (rnrcmh/uCidacre> $mrcm/yr) Imrcm/vtl Mu-54 0.86 0.196 23 1E-06 1.ME-OZ 3.12E-03 236E-08 7.22509 (2-60 527 7.70 1uE+oo 3.19E02 9.G9E43 3.8XE-02 1.11E42 211-65 0.67 0233 1.15E-07 1.89E-02 I25E-M 2.17E-09 1.43E-09 &I37 30.17 37.19 2.70E+01 6.98E-03 3.85E-03 1.88E-01 1.04E-01 Totalhsc 1 227E-01 I l-I5E*Ol DoscLimit: . 5 5 % of Lit 454% 2.30% Revision 29 Date 1/11/02 H-36 UNITED STATES :EAR REGULATORY COMMISS or WASHINGTON. D.C. x)555ooo1 June .15, 2000 NXq 00-58 Mr. Samuel L. Newton Vice President, Operations Vermont Yankee Nuclear Power Corporation 185 Old Ferry Road Brattleboro, VT 05301 SUBJECT: VERMONT YANKEE NUCLEAR POWER STATION, REQUEST TO AMEND *- PREVIOUS APPROVALS GRANTED UNDER IO CFR 20.302(a) TO ALLOW FOR DISPOSAL OF CONTAMINATED SOIL (TAC NO. MA5950) Dear Mr. Newton: By !etter dated June 23,1999, as suppiemented on January 4,2000, you submitted a request . to amend a previously approved apptication granted by the Nudear Regulatory Commission (NRC) pursuant to 10 CFR 20.2002 (previously 10 CFR 20.302) to allow the addition of slightly contaminated soil and soiVsand material to the list of already approved materials-(i.e.,septic waste and cooling tower silt) for on-site disposal via land spreading on designated fields. We have completed our review of your proposal and find it to be acceptable because the previously approved bounding conditions will continue to be met. Pursuant to the provisions of 10 CFR Part 51, the NRC has published an Environmental Assessment and Finding of No Significant Impact in the Federal Register on June. 15 .- ,2000 (65 FR 37583 ). Johnb4 .Zwolinski, Director Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271 Enclosure: Safety Evaluation cc w/encl: See next page Revision 29 Date 1/11/02 H-37 Vermont Yankee Nuclear Power Station cc: Regional Administrator, Region I Mr. Raymond N. McCandless U. S. Nuclear Regulatory Commission Vermont Department of Health 475 Allendale Road Division of Occupational King of Prussia, PA 19406 and Radiological Health 108 Cherry Street Mr. David R. Lewis Burlington, VT 05402 Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W. Mr. Gautam Sen .- Washington, DC 20037-1 128 Licensing Manager. Vermont Yankee Nuclear Power Mr. Richard P. Sedano, Commissioner Corporation Vermont Department of Public Service 185 Old Ferry Road 112 State Street ' P.0, Sox 7002 Montpelier, VT 05620-2601 Brattleboro, VT 05302-7002 Mr. Michael H. Dworkin, Chairman Resident Inspector Public Service Board . Vernont Yankee Nuclear Power 'Station State of Vermont U. S. Nuctear Regulatory Commission 112 State Street P.O. Box 176 Montpelier, VT 05620-2701 Vernon, VT 05354 .' Chairman, Board of Selectmen Director, Massachusetts Emergency Town of Vernon Management Agency ' P.O. Box 116 ATTN: James Muckerheide Vernon, VT 05354-01 16 . 400 Worcester Rd. Framingham, MA 01702-5399 Mr. Richard E. McCullough Operating Experience Coordinator Jonathan M. Block, Esq. Vermont Yankee Nuclear Power Station Main Street P.O. Box 157 - P. O..BOX 566 Governor Hunt Road Putney, VT 05346-0566 Vernon, VT 05354 G. Dana Bisbee, Esq. Deputy Attorney General 33 Capitol Street Concord, NH 03301 -6937 Chief, Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108 Ms. Deborah 6. Katz .. Box 83 Shelburne Falls, MA 01 370 - I . H-38 1 1 Revision29 Date 1/11/02 I . 7 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205554CQY SAFETY EVALUATION BY TH€ OFFICE OF NUCLEAR REACTOR REGULATION VERMONT YANKEE NUCLEAR POWER STATION DOCKET NO. 50-271 .. 1.O INTRODUCTION By letter dated June 23,1999, as supplemented on January 4,2000, Vermont Yankee Nuclear Power Corporation (the licensee), submitted a request to amend a previously approved application granted by the Nuclear Regu!atory Comrnjssion (NRC) pursuant to 10 CFR 20,2002 (previously 10 CFR 20.302) to allow the addition of slightly contaminated soil and soilkand material to the list of already approved materiak (Le., septic waste and cooling tower silt) for on-site disposal via land spreading on designated fields. 2. In 1989, pursuant to IO CFR 20.302 (current 10 CFR 20.2002), the licensee received approval from the NFIC to routinely dispose of contaminated septic waste in designated on-site areas. In 1997, the NRC amended the approved on-site disposal application to also include contaminated ' cooling tower silt material. In this 10 CFR 20.2002 amendment application, the licensee identified 25.5 cubic meters of soil to be disposed of on-site immediately, and approximately 28.3 cubic meters of soikand material to be disposed of on an annual basis until the expiration of the plant's operating license in 2013. The 25.5 cubic meters of contaminated soil were generated as a result of on-site construction activities. The anticipated 28.3 cubic meters of soikand material will be generated from the annual winter spreading of sand on roads and walkways at the plant site. The licensee has performed a comprehensive radiological evaluation that includes all of the anticipated materials (Le., the current 25.5 cubic meters and the 28:3 cubic meters generated annually thereafter). The licensee's evaluation that the soiVsand can be managed on-site in the ' shows same manner as the septic waste and cooling tower silt (Le-, by land spreading on designated fields). .. 2.0 €VALUATION The licensee will dispose of the soil and future soiVsand material using a land spreading technique consistent with its current commitments for on-site disposal of septic waste and cooling tower silts previously approved by the NRC. The licensee will continue to use the designated and approved areas of their property (approximately 1.9 acres in size) which currently receives the septic waste and cooling tower silts, Determination of the radiological dose impact of the new material has been made based on the same dose assessment models and pathway assumptions used in the previously approved submittals. I Revision29 Date 1/11/02 H-39 2 The licensee wit! procedurally control and maintain records of all disposals. The following information will be recorded: 1. The radionuclide concentrations detected in the materia! (measured to radiation levels consistent with the licensee's radiological environmental monitoring program). 2. The total volume of material disposed. 3. The total radioactivity in the disposal operation as well as the total radioactivity accumulated on each disposal plot at the time of spreading. 4. The plot on which the material was applied. 5. Dose calculations or maximum allowable accumulated activity determinations required to demonstrate that the dose condition values imposed (Le., imposed by this 10 CFR 20.2002 application) on the [and spreading operation have. not been exceeded. The bounding dose conditions for the on-site disposals are as follows: 1. The annual dose fo. the whole body or any organ of a hypothetical maximally exposed individual must be less than 1.0 mrem; 2. Annual doses to the whole body and any organ of an inadvertent intruder from the probable pathways of exposure must be less than 5 mrern. , 3. Disposal operations must be at one of the approved on-site locations. To ensure that the addition of new material containing low levels of radioactivity will not exceed the bounding dose conditions, for each new spreading operation the licensee will calculate an estimate of the total radioactivity applied to the designated disposal plots. These calculated estimates will include all past disposaIs of septic waste, cooling tower silt, soil and soilkand material on the designated disposal plots. This wit1 be compared with the bounding dose condition value or equivalent radioactivity value on a pei-acre basis. In addition, Concentration limits will be applied to the disposed material to restrict the placement of small volumes of material that may have relatively high radioactivity concentrations. The licensee assessed the dose that may be received by the maximally exposed individual during the period of plant control over the property, and to an inadvertent intruder after plant access control ends using the same pathway m.odeling , assumptions, and dose calculation methods that were previously approved by the NRC for the septic waste and cooling tower silt disposals. The dose models are based on the guidance in NRC Regulatory Guide 1.109, Revision 1 (1977). The licensee's dose assessment is as follows: 1 Reviwnz Date 1/11/02 H-40 . L .- .. 3 2. Total annual doses .J the whole bAy and critical organ of an inadvertent intruder fr rn the probable pathways of exposure were estimated to be 0.757 mrem and 1.17’ mrem. These values are less than the prescribed annual dose condition value of 5.0 mrem for the time period after active site control. 3. The dose calculations are based on projecting the maximum potential impact of all disposals (past and future) on the designated disposal plot of land. 3.0, CONCLUSION _- The staff finds the licensee’s proposal to dispose of the tow-level radioactive soil and soilkand material, pursuant to 10 CFR 20.2002;in the same manner, location, and within the bounding dose conditions as the materials (i.e., septic waste and cooling tower silt) previously approved by the NRC to be acceptable because the bounding conditions will continue to be met. Principal Contributor: S. Klementowicz Date: 2000 June 15, r. c c . .. Revision _29 Date 1/11/02 . H-41 llI I -. Appendix I 1. “Request to Amend Previous Approval Granted Pursuant to 10CFR20.2002 for DisposaI of Contaminated Soil”, dated September 11*, 2000, BVY 00-71 2. Vermont Yankee Nuclear Power Station - Safety Evaluation for an Amendment to an Approved 10CFR20.2002 AppIication (TAC No. MA9972)”, dated June 2dh, 2001, NVY 01-66 1-1 llRevision 29 Date 1/11/02 VERMONTYANKEE NUCLEAR POWER CORPORATION 185 OLD FERRY ROAD, PO BOX 7002, BRATTLEBORO, VT 05302-7002 (802)257-527 1 September I 1,2000 BVY 00-71 United States Nuclear Regulatory Commission A’ITN: Document Control Desk Washington, DC 20555 References: (a) Letter, VYNPC to USNRC, “Request to Amend Previous Approvals ..- Granted under 10 CFR 20.302(a) for Disposal of Contaminated Septic Waste and Cooling Tower Silt to Allow for Disposal of Contaminated Soil,” BVY 99-80, dated June 23, 1999. (b) Letter, VYNPC to USNRC, “Supplement to Request to Amend Previous ApFrovaIs Granted under 10 CFR 20.302(a) to Allow for Disposal of Contaminated Soil,” BVY 00-02, dated January 4,2000. (c) Letter, USNRC to VYNPC, crVermontYankee Nuclear Power Station, Request to Amend Previous Approvals Granted under 10 CFR 20.302(a) to AHow for Disposaf of Contaminated Soil (TAC No. MA5950),” NVY 00-58,dated June 15,2000. (d) Letter, USNRC to VYNPC, “Revised Safety Evaluation - Approval Pursuant to 10 CFR 20.2002 for Onsite Disposal of Cooling Tower Silt - Vermont Yankee NucIear Power Station (TAC No. M96371),” NVY 97-85, dated June 18,1997. Subject: Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271) Request to Amend Previous Approval Granted Pursuant to 10 CFR 20.2002 for Disposal of Contaminated Soil In accordance with 10 CFR 20.2002 (previously 10 CFR 20.302(a)), Vermont Yankee (VU) submits this application to amend the previously granted approval to dispose of slightly contaminated soil. This application expands the allowable waste stream to include slightly contaminated soil generated as a residual by-product of other types of on-site construction activities. In References (a) and (b), VY requested approval to dispose of approximately 25.5 m3 of accumulated soil that was generated due to construction .activities. In addition, it was requested that VY be aiIowed to dispose of approximately 28.3 m3of soil that is spread annually on station roads and walkways during the winter, NRC acceptance is documented in Reference (c). This application specifically requests approval to dispose of contaminated soil that is created due to other on-site construction related activities including but not limited to design change implementation and land maintenance. I Revirion2 Date 1/11/02 1-2 - VERMONTYANKEENUCLEAR POWERCORPORATION BVY 00-7I / fage 2 of 2 in addition, VY requests that NRC’s review recognize that, although VY indicated in Reference (b) that the south disposal field (approximately 1.9 acres in size) is currently expected to be used for disposal of the subject material, VY is also authorized to use the alternate north disposal field (approximately i 0 acres in size). Approval to use both the north and south fields for disposal was granted in Reference (d). VY’s radiological impact assessments have conservatively assumed all of the disposal activities occur on the smalter south field to maximize potential caIculated doses. These assessments bound the situation where a portion of the land spreading occurs an the north field. VY will continue to limit the total activity spread, from approximately 28.3 m3 of soil generated each year, to within the limits assumed in the radiological assessment previously submitted in Reference (b). A radiological assessment and proposed operational controls for the inclusion of the additional material for on-site disposal was provided in Reference (b). The assessment demonstrates that the dose impact expected from the proposed activity, in total with all past waste spreading operations, will not approach the dose limits already imposed for septic and cooling tower silt disposal. All soil analyses will be to environmental lower limits of detection. The results of all disposal operations will be reported in the Annual Radioactive Effluent Release Report. The combined radiological impact, for ail on-site disposal operations, will continue to be limited to a total body or organ dose of a maxirnaIly exposed member of the pubIic of less than . oie mremjyear during the period of active VY control of the site, or less than five rnrem/year to an inadvertent intruder after termination of active site control. Upon receipt of your approval, this request as well as the basis for approval will be incorporated into the Off-Site Dose Calculation ManuaI. We trust that the information contained in the submittal is sufficient. However, should you have any questions or require further information concerning this matter, please contact Mr. Jim. DeVincentis at 802-258-4236. Sincerely, Vermont Yankee Nuclear Power Corporation n Licensing Manager cc: USNRC Region I Administrator USNRC Resident Inspector - VYNPS USNRC Project Manager - VYNPS VT Department of Pubtic Service I Revision 29 Date 1/11/02 1-3 SUMMARY OF VERMONT YANKEE COMMITMENTS BWNO.: 00-71 The following table identifies commitments made in this document by Vermont Yankee. Any other actions discussed in the submittal represent intended or planned actions by Vermont Yankee. They are described to the NRC for the NRC's information and are not regdatory commitments. Please notify the Licensing Manager of any questions regarding this document or any associated commitments. r COMMITMENT COMMITIED DATE OR "OUTAGE" I 1.- None NIA VYAPF 0058.04 AP 0058, Revision 1 Page I of 1 1-4 I Revision 29 Date 1/11/02 June 26, 2001 Mr. Michael A. Balduzzi NVY 01-66 Vice President, Operations Vermont Yankee Nuclear Power Corporation 185 Old Ferry Road P.O.Box 7002 Brattleboro, VT 05302-7002 SUBJECT: VERMONT YANKEE NUCLEAR POWER STATION- SAFETY EVALUATION FOR AN AMENDMENT TO AN APPROVED io CFR 20.2002 APPLICATION (TAC NO. MA9972) Dear Mr. Balduzzi: The US. Nuclear Regulatory Commission (NRC) staff hqs completed its review of the Vermont Yankee Nuclear Power Corporation (VYNPC)request dated September 11, 2000, to amend an approved 10 CFR 20.2002 (previously 10 CFR 20.302) application dated June 23,1999, as supplemented on January 4,2000. The licensee requested NRC approval to allow the addition of slightly contaminated soil resulting from on-site construction-related activities, including but not limited to, design change-implementation and land maintenance, to the list of already approved materials (i.e., septic waste, cooling tower silt and soiVsand from roads and walkways) for on-site disposal. Based on our review, we find the proposed changes to be acceptable because the previousIy to approved bounding conditions will continue ’ be met. The enclosure to this letter provides our safetyevaluation of WNPC’s application. Pursuant io the provisions of 10 CFR Part 51, the NRC has published an Environmental Assessment and finding of No Significant Impact in the Federal Register on June 14,2001 (66 FR 32399). Sincerely,yk47JQ. Joh A. Zwolinski, Director Divis\ on of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-271 Enclosure: Safety Evaluation cc w/encl: See next page I Revision 29 Date 1/11/02 1-5 Vermont Yankee Nuclear Power Station cc: . Regional Administrator, Region I Ms. Deborah B. Katz . U. S. Nuclear Regulatory Commission Box 83 475 Allendale Road Shelburne Falls, MA 01370 King of Prussia, PA 19406 Mr. Raymond N. McCandless -f Mr. David R. Lewis Vermont Department of Health Shaw, Pittman, Potts & Trowbridge Division of Occupational 2300 N Street, N.W. and Radiological Health Washington, DC 20037-1 128 t 08 Cherry Street Burlington, VT 05402 Ms. Christine S. Salembier, Commissioner Vermont Department of Public Service Mr. Gautam Sen 112 State Street Licensing Manager Montpelier, VT 05620-2601 Vermont Yankee Nuclear Power Corporation Mr. Michael H. Dworkin, Chairman 185 Old Ferry Road Public Service Board P.O. Box 7002 State of Vemiont Brattleboro, VT 05302-7002 112 State Street MontpeIier, VT 05620-2701 Resident Inspector Vermont Yankee Nuclear Power Station Chairman, Board of Selectmen U. S. Nuclear.Regutatory Commission Town of Vernon P.O. €!ox 176 P.O. Box 116 Vernon, VT 05354 Vernon, vf 05354-01 16 Director, Massachusetts Emergency Mr. Richard E. McCullough Management Agency Operating Experience Coordinator ATTN: James Muckerheide Vermont Yankee Nuclear Power Station 400 Worcester Rd. P.O. Box 157 Framingham, MA 01702-5399 Governor Hunt Road Vernon, VT 05354 Jonathan M. Block, Esq. Main Street G. Dana Bisbee, Esq. P. 0. Box566 Deputy Attorney General Putney, VT 05346-0566 33 Capitol Street Concord, NH 03301 -6937 Chief, Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108 Revision 29 Date 1/11/02 . 1-6 7 UNITED STATES NUCLEAR REGULATORY COMMISSION & WASHINGTON, D.C. 20555-0001 ‘SAFETYEVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION VERMONT YANKEE NUCLEAR POWER CORPORATION VERMONT YANKEE NUCLEAR POWER STATION f DOCKET NO. 50-271 1.O INTRODUCTION By letter dated September 11,2000, Vermont Yankee Nuclear Power Corporation (VYNPC/ licensee) submitted a request to amend.s Title 10 of the Code of Federal Regulations (10 CFR) Section 20.2002 (former 10 CFR 20.302) application, dated June 23, 1999, as supplemented on January 4,2000, that was approved by the US. Nuclear Regulatory Commission (NRC). This amendment will allow the addition of slightly contaminated soil resulting from on-site construction-related activities, including but not limited to, design change implementation and land maintenance, to the list of already approved materials (Le., septic waste, cooling tower silt and soiVsand from roads and walkways) for on-site disposal via land spreading on designated disposal fields. In 1989, pursuant to 10 CFR 20.302 (current 10 CFR 20.2002), the licensee received approval - from the NRC to routinely dispose of contaminated septic waste in designated on-site areas. In 1997, the NRC amended the approved on-site disposal application to also include contaminated cooling tower silt material. In 2000, the NRC amended the approved on-site disposal application to aIso include a one-time disposal of slightly contaminated sol and an annual .- disposal of 28.3 cubic meters of slightly contaminated soiVsand material. 1 In this 10 CFR 20.2002 amendment application, the licensee requested that slightly contaminated soil resulting from on-site construction-related activities be disposed of on-site on an annual basis until the end of the plant’s operating license in 2013. The anticipated annual volume of soil generated by on-site construction, as identified by the licensee, combined with the soiVsand generated from the annual winter spreading of sand on roads and walkways at the plant site will not exceed 28.3 cubic meters. This volume is the same volume that was approved in the January 4,2000, request. The licensee performed a comprehensive radiological evaluation which included the annual disposal of 28.3 cubic meters of soil and soiVsand materials, and shows that these materials can be managed on-site in the same manner as the septic waste and cooling tower silt (i.e., land spreading on designated fields). 2.0 EVALUATION The licensee will dispose of the future soil material using a land spreading technique consistent with the current commitments for on-site disposal of septic waste, cooling tower silts and sandsoil material previously approved by the NRC. The licensee will continue to use the I- 7 I Revision 29 Date 1/11/02 - -2- designated and approved areas of their property which include approximately 1.9 acres, which currently receive the septic waste, cooling tower silts and soiVsand material, and approximately 10 acres which have not been previously used for disposal. Determination of the radiological dose impact of the new material has been made based on the same dose assessment models and pathway assumptions used in the previously approved submittals. The licensee will proceduratty control and maintain records of all disposals. The foIlowing information will be recorded: 1I The radionuclide concentrations detected in the material (measured to radiaon levels consistent with the licensee's radiological environmental monitoring program); 2. The total volume of material disposed; 3. The total radioactivity in the disposal operation as well as the total radioactivity ' accumulated on each disposal plot at the time of spreading; 4. The plot on which the material was applied; 5. Dose calculations or maximum allowable accumulated activity determinations required to demonstrate that the dose condition values imposed (Le., imposed by the approved 10 CFR 20.2002 application dated June 23,1999) on the land spreading operation have not been exceeded. ,..' .:. , .I....: The bounding dose conditions for the on-site disposals are as follows: 1. The annual dose to the whole body or any organ of a hypothetical maximally exposed individual must be less than 1.O mrern. 2. Annual doses to the whole body and any organ of an inadvertent intruder from the probable pathways of exposure must be less than 5 mrem. 3. Disposal operations must be at one of the approved on-site focations. 4. Total annual combined volume of soil and soiVsand materials must not exceed 28.3 cubic meters. To ensure that the addition of new material containing low levels of radioactivity will not exceed the bounding dose conditions, for each new spreading operation the licensee will calculate an estimate of the total radioactivity that includes all past disposals of septic waste, cooling tower silt, and soil/sand and soil material on the designated disposal plots. This will be compared with the bounding dose condition value or equivatent radioactivity value on a per acre basis. The licensee assessed the dose from soil and soiVsand material that may be received by the maximally exposed individual during the period of plant control over the property, and to an inadvertent intruder after plant access control ends using the same pathway modeling, assumptions, and dose calculation methods that were previously approved by the NRC for the .. septic waste and cooling tower silt disposals. The dose models are based on the guidance in NRC Regulatory Guide 1 .I 09, Revision I (1 977). 1-8 I Revision 29 Date 1/11/02 :I ~ 1 3- The licensee's dose assessment is as follows: 1. Total annual doses to the whole body and critical organ of the hypothetically maximally . exposed indwidual were estimated to be 0.1 15 mrem and 0.403 rnrern respectively. .Thesevalues are less than the prescribed annual dose condition value of t .O mrem for the time period of active site control. 2. Total annual doses to the whole body and critical organ of an inadvertent intruder from the probable pathways of exposure were estimated to be 0.757 mrem and 1.17 mrem. These values are less than the prescribed annual dose condition vahe of 5.6 mrem for the time period after active site contro1. 3. The dose calculations are based on projecting the mdmum potential impact, of all disposals (past and future) on the approved disposal site. 3.0 CONCLUSION The staff finds the licensee's proposal to dispose of the low-level radioactive soil material, pursuant to 10 CFR 20.2002, in the same manner, tocation;and within the bounding dose conditions as the materials (i.e., septic waste, cooling tower silt and soiVsand from roads and walkways) previously approved by the NRC to be acceptable. The licensee has committed to permanently incorporate this modification into their Offsite Dose Calculation Manual. Principal Contributor: A. 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Jzznxms S _ spsxiv _oo.cs n Ny n Mivl vi syx Vix iv aipn C yo =C en wyx hixunn Y lvni ay n c i syx JYYNTMPe - )Lwv pv lk, bw lz pt ug lzpgt/ bol gxxtpig pwv gvgt lk ol l xli lk zgkpwtwnpigt puxgi mzwu hw o ol l p pvn pv lv wz g og pul wm ghw 3;-111 .4 +).:EA u+ ~wm gii u tg lk pt - gtwvn p o gv wxlzg pvn i itl )3Duwv o , nlvlzg pwv zg l wm ghw ;111 .4 + ).33: u+~! bol T_L zl zvlk olpz gml l gt g pwv- kg lk Sgzio;/ 3EEA-nzgv pvn gxxzw gt mwz ol xzwxw lk pt kp xw gt/ apuptgz w ol l gnl g l kp xw gt- T_L giilx gvil zly pzlk og gtt kp xw gt wxlzg pwv hl iwvk i lk io og hw o ol xzwrli lk kw l )mwz hw o i zzlv gvk gtt xg kp xw gt wxlzg pwv , gvk zgkpwv iRpkl iwvilv zg pwv tpup gzl g p mplk/ bol wpt iwvilv zg pwv tpup )mwz gv guxtl, lzl hg lk wv tpup pvn l lzvgt gvv gt kw l w 7= uzlu g upvn iwv pv w wii xgvi wv gv pvmpvp l xtw g og iwvilv zg pwv vpmwvvt xzlgk w g 3= lu klx o/ Tw xlipmpi tpup wv gvv gt wt ul wm Lwwtpvn bw lz pt og iw tk hl kp xw lk wm g pvit klk pv ol gxxzw gt/ bol T_L gt w zly pzlk og gv m z olz uwkpmpig pwv w ol xzwxw lk gi pwv og l xzpwz T_L gmm gxxzw gt/ Pv 3EEE) p o g xxtlulv gt mptpvn pv7111,- clzuwv fgvsll mptlk g opzk zly l vklz31 LO_ 71/7117 p o ol T_L w gulvk ol xzl pw t gxxzw lk gxxtpig pwv mwz wv. p l tgvk kp xw gt wm tpno t iwv gupvg lk lgz o ~ ug lzpgt ) lx pi t knl gvk Lwwtpvn bw lz pt , w pvit kl gxxzw pug lt E11 IvR)7=/= u , wm gii u tg lk iwv gupvg lk wpt nlvlzg lk k zpvn iwv z i pwv gi p p pl p opv ol cf Yzw li lk Jzlg/ aguxtpvn wm ol wpt zl lgtlk tw tl lt wm zgkpwgi p p og lzl puptgz pv zgkpwv itpkl gvk gi p p tl lt w ol lx pi g l gvk Lwwtpvn bw lz pt xzl pw t lviw v lzlk/ bol zly l w ol T_L mwz op gkkp pwvgt ug lzpgt gt w pvkpig lk og gkkp pwvgt guw v wm iwv gupvg lk wpt" gvk g wipg lk p o zwgk llxpvn mwttw pvn pv lz gvkpvn wm zwgk gvk dgts g pv ol Yzw li lk Jzlg iw tk zl t pv gv l pug lk3111 .4 : xlz lgz )7D/: u : xlz lgz, og upno vllk w hl kp xw lk wm g tpno t iwv gupvg lk ug lzpgt / bol T_L zly l lk og ol pvp pgt hup gt)3EEE,wm ol wpt xzlgkpvn71/7117 gxxtpig pwv pvit kl gv gvgt p og l gt g lk xzwrli lk m zl gkkp pwv wm gv l pug lk gvv gt wt ul wm wpt hlpvn gkklk w ol kl pnvg lk kp xw gt xtw / bop pvmwzug pwv g zly pzlk pmclvvwv fgvsll pv lvklk w l ol71/7117 wpt kp xw gt gxxtpig pwv mwz gxxzw gt w kp xw l wm xw lv pgt m zl wt ul )p/l/- vw r g wvl. pul kp xw gt gxxtpig pwv, wmtw tl lt iwv gupvg lk wpt pv ol gul ugvvlz g gtzlgk gxxzw lk mwz lx pi g l gvk Lwwtpvn bw lz pt / clzuwv fgvsll zl p lk p gxxtpig pwv h gkkpvn gv gvgt p mwz g xzwrli lk gvv gt wt ul3111 wm IvR-wz ly p gtlv t 7D/: uR- wm iwv gupvg lk wpt gz pvn pv ol lgz7111 gvk iwv pv pvn wv g lgzt hg p v pt lvk wm xtgv tpilv l pv 713:/ J ol lvk wm ol xzwrli lk kp xw gt zlgu- ol gii u tg lk h ptk x wm iwv gupvg pwv mzwu gtt w zil ) lx pi g l- Lwwtpvn bw lz pt gvk wpt 0 gvk up l , wv ol kp xw gt mpltk g l gt g lk mwz hw o ol kw l puxgi w ol izp pigt zlilx wz g ol lvk wm ol iwv zwt xlzpwk gvk ol g ulk pv z klz/ bol l kw l puxgi lzl mw vk giilx ghtl olv iwuxgzlk gngpv ol wzpnpvgt wv. p l xzlgkpvn kw l giilx gvil izp lzpg wm P uzlut z )Lwv zwt Ylzpwk, gvk = uzlut z )Pv z klz ailvgzpw,/ bol 3111 .4+)7D/: u&, gvv gt nlvlzg pwv zg l wm wpt g hg lk wv xtgv gmm l pug l og gxxzw pug lt og guw v wm wpt gvk gvk p iwttli lk mzwu zwgk gvk gts g llxpvn pv pkl ol Yzw li lk Jzlg mwttw pvn lgio lgz' pv lz itlgv. x g xgz wm zw pvl ugpv lvgvil/ bop p ol wvt xl wm lgz olv ug lzpgt og og g xlipmpi gvv gt wt ul tpup g wipg lk p o p pv gkkp pwv w ol xzwrli lk kw l gvk iwvilv zg pwv tpup g wipg lk p o ol kp xw gt wm lx pi g l gvk Lwwtpvn bw lz pt / Tw wt ul l pug l mwz vpklv pmplk m zl p l l ig g pwv gvk iwv z i pwv gi p p pl g xzw pklk/ = Jxxlvkp R Vzpnpvgt Vmm.ap l Mw l Lgti tg pwv Sgv gt Ygnl D wm =C clzuwv fgvsll T itlgz Yw lz a g pwv EVVLSJNe O &Iwv pv lk' 1-/ _bRREYf TM YL_bPa_ aol l gt g pwv wm ol zgkpwtwnpigt puxgi wm gtt xg ) gii u tg lk wzgnl pv lv wz ) gvk xzwrli lk m zl g l xzlgkpvn wxlzg pwv wv g pvntl kp xw gt xtw &0-C gizl g _w o lvk wm p l' og l pvkpig lk ol mwttw pvn zl t gvk iwvit pwv D Ett xg xzlgkpvn wm lx pi g l) Iwwtpvn aw lz pt - gvk wpt. gvk up l ozw no ol lvk wm1//3 og l zl t lk pv g ug pu u wzngv kw l w g izp pigt zlilx wz &iwv zwt Vlzpwk l' og giiw v mwz wvt 03-=" wm ol 0 uzlut lgz tpup - dp o zl xli w ol Nv z klz kw l tpup wm : uzlut lgz) gpt xg xzlgkpvn ozw no ol lvk wm 1//3 giiw v mwz wvt =-1" wm ol ug pu u wzngv kw l w ol tpup pvn zlilx wz g ol lvk wm xtgv tpilv l- aol puxgi mzwu ol xzwrli lk xzlgkpvn wm ol l p pvn g l ug lzpgt pv wzgnl p l pug lk w giiw v mwz wvt :-3" wm ol 0 uzlu. lgz Iwv zwt Vlzpwk kw l tpup - wz wvt 0=-=" wm ol gul tpup olv gtt xg xzlgkpvn gzl iwuhpvlk p o ol ug lzpgt i zzlv t pv wzgnl &g wm ol lvk wm Sw luhlz 1//3'- aol ug pu u Nv z klz wzngv kw l mzwu gtt xg xzlgkpvn gvk wzlk ug lzpgt p /-7:; uzlut z) wz C-0" wm ol : uzlut z Nv z klz ilvgzpw kw l tpup - aol l zl t pvkpig l og ol l p pvn pv lv wz wm g l ug lzpgt pv wzgnl igv hl xtgilk wv ol _w o kp xw gt xtw p ow l illkpvn kw l puxgi tpup xzl pw t l ghtp olk mwz g pvntl kp xw gt mpltk- E upvn ol gul gvv gt g lzgnl nlvlzg pwv zg l wm zgkpwgi p p pv g l ug lzpgt & lx gnlt pt t wpt' og og hllv wh lz lk w lz ol tg mw z llv lgz p gkklk w gtt xg g l xzlgkpvn gvk wzlk wpt iwuup ulv ) ol xzwrli lk kw l g ol lvk wm ol i zzlv xtgv tpilv pvn xlzpwk & lgz 1/03' pltk g tpup pvn ug pu u izp pigt zlilx wz kw l &mwz lp olz ol iwv zwt xlzpwk wz pvgk lz lv pv z klz' ly p gtlv w wvt 1:-3" wm ol uw zl zpi p l gvv gt kw l tpup &g wipg lk p o ol N uzlu. lgz tpup mwz ol ug pu u wzngv k zpvn ol Iwv zwt Vlzpwk'- aop mpvkpvn kluwv zg l og ol iwv pv lk l wm ol _w o kp xw gt xtw ) l lv p o ol gkkp pwv wm 0A);:3m 3wm tpno t iwv gupvg lk wpt i zzlv t pv wzgnl) ptt vw l illk ol gxxzw lk tpup pvn kw l izp lzpg l ghtp olk pv ol TJIR- ahl kw l puxgi ul owkwtwn lk w kl lzupvl iwuxtpgvil p o ol wv, p l xzlgkpvn kw l tpup gzl vw kzp lv h ol wt ul wm g l ug lzpgt kp xw lk wm) h h ol w gt zgkpwgi p p iwv lv wm ol ug lzpgt- L p pvn tpup wv ol iwvilv zg pwv wm zgkpwgi p p pv g l ulkpg xzw pkl xzw li pwv mzwu ugtt wt ul wm ~hw ~ wz opno xlipmpi gi p p ug lzpgt mzwu hlpvn xzlgk wv ol kp xw gt mpltk- ; Exxlvkp O Tzpnpvgt Tmm,_p l Jw l Igti tg pwv Rgv gt Vgnl C wm := clzuwv fgvsll S itlgz Vw lz _ g pwv D__LVJPg R 'I z uz po) 3./ TLcOYJ YM LeDSdDcPYV ctp yp t o r p lx l uzs tp uy ln r y tp z- u p plouzs u tp lyp l po lzo l po uz tp usuzlx p py u z p p ylop tp VaI zop PYIMa1/.R/1 r p un l p lzo tunt tl mppz l xupo uz lxx m p pz lypzoypz p p r loou u zlx p r pl tpz yl p ulx mp ou po m xlzo plouzs z- u p. ctp l t l lzo o p y opx r zo uz aps xl N uop P.R/C 'aprp pznp 1) l p py x po uz p r yuzs tp lou x sunlx o p uy ln l p ypz . ctp l xunl u z r tp o p y opx mpsuz u t tp ntl ln p u l u z r tp l p yl p ulx tl l p mp mvpn tp o p p lx l u z. 3.0 fl p Itl ln p u l u z ctp p u uzs lnn y xl po ux~ lzo tl tl n pl po l mlnwx s r lnn y xl po yl p ulx p tp xl p p lx pl u uopz urupo z clmxp 0, lx zs u t tp p uyl po x yp lzo usuzl u z r yl p ulx. ctp ux yl p ulx p p uyl ux op u po r y p nl l u z ln u u up l nul po u t tp n z n u z r zp pn u rpznp lx zs tp xlz & _ pn po D pl m zol lzo tp n z n u z r zp xlz uz lxxl u z l nul po u t tp nl lmuxu p r y t o spz l p ntpyu pl ypz r tp xlz & n xlz py. Dx uznx opo l p lo pl ypz lzo po r uz p ln u z uz uop tp pn po l pl. ctp ux~ lzo yu u unlx r ruxx yl p ulx n z luzuzs xust ol w m z x po ux u t yp ylxx zp , lzo yl uznx op ylxx uznuopz lx upnp r l tlx . ctp ux l py po r y u usuzlx x nl u z m t px, mlnwt p lzo r z -pzo x lop , lzo xlnpo uz o y nw r lz tp py l lsp l pl x nl po mp ppz tp I xuzs c p tp pu l op u po z tp s zo rlnp lzo n p po p pz p u z. ctu x nl u z l pxpn po mpnl p u l l l r y l pl uzpx nn upo m xlz lrr, lzo n xo pl ux mp n z xxpo. ctp y mlmxp np r tp x xp px r lou ln u p n z lyuzl u z u o p tp p pznp r mpx op pn lmxp py lmxp n z lyuzl u z pou um po m r lrrun r y uz uop tp xlz flxw l lzo l wuzs l pl . b m p pz rlnp z rr nl up tp n z lyuzl u z zpl m p po ux zpl tp _ pn po D pl m zol tp p u lnn y xl p p uyp x -xp px op pn lmxp n znpz l u z . J z l t lzo op u u z r l un xl p ln u u pxpl po r y tp xlz & _ uyl epz b lnw l l r uzp sl p pyu u z yl lx tl p n z um po tp x xp px r op pn lmxp ln u u . M pz ulx r p ou lx x yp r lzo lzo ux, tp n pz x yp xuyu x '0///i1A.3 y 3k, YJIT D pzou O, aprp pznp P) l ml po z tp p pn po l p r lo lzo po r uz p lo lzo lxw l ln u z, m ouo z lz unu l p prxpn tp pz ulx r r p u p n z n u z ln u u up tl n xo p nl l p ux z- u p tl lx n z luz x xp px r xlz pxl po lou ln u u . ctp p pz uz pz r po ux~ lzo mp ppz tp I xuzs c p uznx op l uyl px 3 0A,;:3 r 3':1A y ) r yl p ulx r tunt zx 3.3" ;0; x'0=.7 z &) usuzl po l lo l pp uzs . Pz lnn olznp u t D pzou p E lzo M r tp YJIT, ou lx r p un l p lzo I xuzs c p ux yl p ulx u z xuyu po m lz lzz lx x yp xuyu m m lx o p uy ln pxl po tp lou ln u u n z pz r tp ux lzo tp n znpz l u z ryou ln u u n z luzpo u tuz u . 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Tkg flwtswco stgvcxlsr sh xkg wslo tlogw loo hsoos xkg cttolecdog Vgvpsrx _crngg tvsegfyvgw xs pclrxclr fswgw cw os cw vgcwsrcdo ceklgzcdog crf lxklr xkg wtgelhle fswg evlxgvlc cw tvgzlsywo cttvszgf hsv wgtxle crf Dssolri Ts gv wlox cwxg flwtswco) 0)0 Elwtswco Posx Dkcvcexgvlwxlew Aoo fgwlircxgf flwtswco wlxgw ~wl flhhgvgrx tosxw" cvg osecxgf sr xkg Irxgvi Nyeogcv Nsvxkgcwx Vgvpsrx _crngg tocrx wlxg crf cvg lxklr xkg wlxg dsyrfcv wgeyvlx hgreg) Tkg Ssyxk hlgof esrwlwxw sh cttvs lpcxgo .)= cevgw crf lw egrxgvgf cttvs lpcxgo .3-- hggx Ssyxk sh xkg Rgcexsv Cyloflri) Tklw hlgof kcw dggr xkg sro srg sh xkg NRD cttvszgf hlgofw xkcx kcw cexycoo dggr yxlol gf hsv xklw tyvtswg xs'fcxg) Jx lw crxleltcxgf xkcx hyxyvg flwtswco stgvcxlsrw loo cows yxlol g xkg Ssyxk hlgof wlreg wyhhlelgrx pcvilr lr esptcvlwsr xs xkg cttvszgf fswg olplx evlxgvlc wxloo g lwxw hsv crxleltcxgf cwxg flwtswco sh xkg g lwxlri dcenosi sh wslo rs lr wxsvcig& toyw coo g tgexgf hyxyvg flwtswcow sh wgtxle cwxg& Dssolri Ts gv wlox crf wslo , wcrf pl gw cwwyplri xkg wcpg sdwgvzgf igrgvcxlsr vcxgw ~wgg Tcdogw .0 xkvsyik .=" tgvwlwx xs xkg grf sh xkg tocrx olegrwg lr /-.0) Jr cfflxlsr xs xkg Ssyxk hlgof& xkg rsvxk grf sh xkg wlxg kcw cr cfflxlsrco xgr cevg tcvego egrxgvgf cttvs lpcxgo /&--- hggx rsvxk gwx sh xkg Rgcexsv Cyloflri) Pvlsv cwwgwwpgrxw kczg fgpsrwxvcxgf xkcx c wlriog tosx sh cdsyx / cevgw lw wyhhlelgrx xs pggx vsyxlrg sv g tgexgf flwtswco rggfw) Tkgvghsvg xkg rsvxkgvr wlxg esyof dg wydflzlfgf lrxs 3 tosxw lh cfflxlsrco ectcelx cw rggfgf) 0)1 Rcflsosileco Jptcex Mgxksfsosi Tkg cpsyrx sh vcflscexlzlx cffgf xs cr sh xkg flwtswco hlgofw lw tvsegfyvcoo esrxvsoogf xs lrwyvg xkcx fswgw cvg pclrxclrgf lxklr xkg tvlsv cttvszgf olplxw sh xkg dsyrfcv esrflxlsrw ~wgg Sgexlsr .)- cdszg") Ts cwwgww xkg fswg vgeglzgf ~chxgv xkg wtvgcflri sh xkg g lwxlrib ;&730 hx0 a3/;cosri p0 lxkb dsxk tcwx vgesvfgf flwtswco cttolecxlsrw& toyw tvsmgexgf hyxyvg cttolecxlsrw" d xkg pc lpcoo g tswgf lrflzlfyco fyvlri xkg tgvlsf sh tocrx esrxvso& crf xs cr lrcfzgvxgrx lrxvyfgv chxgv tocrx esrxvso sh ceegww grfw ~vghgvgreg gcv sh/-.0"& xkg wcpg tcxk c psfgolri& cwwyptxlsrw crf fswg ecoeyocxlsr pgxksfw cw cttvszgf hsv wgtxle cwxg& Dssolri Ts gv wlox crf tcwx wslo , wcrf flwtswcow cvg ywgf) = Attgrfl L Ovlilrco Ohh'Slxg Eswg Dcoeyocxlsr Mcryco Pcig ./ sh 3: Vgvpsrx _crngg Nyeogcv Ps gv Sxcxlsr E__MVLRg S &Jzy ty on' cso o nz o xznow tx woxoy so xo szn kyn nz o mzy o tzy pkm z k z tnon ty aor wk z O tno 0-0/A &aopo oymo 1'- cso pzwwz tyr t z oy tkw k s k o o tnoy tpton kyn tymw non ty so kykw t C &k' b kyntyr zy mzy kxtyk on r z yn) &l' Ryskwk tzy zp o oynon kntzkm t t ) &m' Ryro tzy zp wokp oro klwo ) &n' Ryro tzy zp z on oro klwo ) &o' Ryro tzy zp xok ) kyn ~.! 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Nvx znk vznkx ikzkhzki xfiovu hsoiky 'Jv-=0, du-=; fui Ru-;:), znoy gofyoum vl znk ifzf fy uvz fwwsoki youhk znk vhh xxkuhk vl wvyozo k fs ky vus xkwxkykuzki f ytfss lxfhzovu vl znk zvzfs u tgkx vl yftwsky zfrku. _nk zvzfs fhzo oz ikzkxtoufzovu lvx kfhn wosk oy znku hfsh sfzki g zfroum znk f kxfmk tkfy xki hvuhkuzxfzovu zotky znk tkfy xki yvos ikuyoz zotky znk kyzotfzki vs tk vl znk yzvxfmk wosk, hvxxkhzoum lvx ikhf zotk gkz kku znk ifzk vl znk yftwsk fufs yoy ifzk zv znk kyzotfzki ifzk vl loksi ioywvyfs '=/PsT:). _noy hfsh sfzki zvzfs xfiovu hsoik fhzo oz fs k lvx kfhn wosk oy fyy tki zv gk wsfhki vu znk youmsk 1.D fhxk Yv zn ioywvyfs wsvz ozn znk xky szoum y xlfhk hvuhkuzxfzovu ' Jo/fhxk) lvx znk wxvpkhzki ioywvyfs hfsh sfzki lvx kfhn wosk fui zvzfski lvx fss yo wosky h xxkuzs ou yzvxfmk. _nk kyzotfzki f kxfmk hvuhkuzxfzovu lvx kfhn vl znk lv x ikzkhzki xfiovu hsoiky oy ikhf hvxxkhzki lxvt znk ifzk vl znk yftwsk hvsskhzovu zv P uk 1,300:, fy znk xklkxkuhk ifzk lvx kyzotfzoum ivyk otwfhzy lxvt znk wxvwvyki ioywvyfs vl znk fhh t sfzki tfzkxofs ou yzvxfmk. _fgsk 13 y ttfxo ky znk xfiovfhzo oz fyyvhofzki ozn znk yvos/yfui ou yzvxfmk ikhf ki zv P uk 1,300:, fui zv znk kyzotfzki kui vl znk h xxkuz wsfuz sohkuyk ou 3017 lvx ikzkxtououm znk Ouzx ikx ivyk otwfhz fz znfz zotk. _fgsk 30 hvtgouky znk xfiovfhzo oz hvuzkuz vl f11yvos/yfui tfzkxofsy yzvxki gkz kku znk Jvvsoum _v kxy ozn fss xktfououm fhzo oz wxk ov ys ywxkfi vu znk Yv zn ioywvyfs wsvz, ikhf ki zv 3017 lvx yk ou ikzkxtououm znk Ouzx ikx ivyk fz znk kui vl wsfuz sohkuyk. _fgsk 37 fwwsoky znk ivyk hvu kxyovu lfhzvxy lxvt _fgsk 3 lvx tf ot t vxmfu fui nvsk gvi ivyky ozn znk wxvpkhzki fhzo oz vu znk Yv zn ioywvyfs wsvz lxvt _fgsky 13, 1C fui 30 zv loui znk Jvuzxvs Vkxovi ivyk lvx fss wfyz tfzkxofs ywxkfioumy, yzvxki tfzkxofs fiiozovuy, fui znk y t zvzfs vl wfyz ywxkfioumy fui wxvwvyki yzvxki tfzkxofs fiiozovuy. _fgsk 3: oss yzxfzky znk yftk ivyk otwfhz hvtgoufzovu vl fyzk yzxkfty fy fwwsoki zv znk yozk Ouzx ikx fz znk kui vl wsfuz sohkuyk. _nk hvtgoufzovu vl fss wfyz ywxkfioumy fui znk h xxkuz yzvxki yvos tfzkxofs xky szy ou f tf ot t vxmfu ivyk vl vus 1A.A & '0.1AA txkts x) vl znk e txkt/ x Jvuzxvs Vkxovi ivyk sotoz. _nk tf ot t Ouzx ikx vxmfu ivyk oy kyzotfzki zv gk 0.:;= txkt/ x vx D.1& vl znk ; txkt/ x Ouzx ikx yhkufxov ivyk sotoz. 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