ML20206D798
| ML20206D798 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 11/10/1998 |
| From: | Odell W Maine Yankee |
| To: | |
| Shared Package | |
| ML20206D725 | List: |
| References | |
| PROC-981110, NUDOCS 9905040244 | |
| Download: ML20206D798 (92) | |
Text
I 1
- 'O - .
l MAINE YANKEE OFF-SITE DOSE CALCULATION MANUAL O -
APPROVED:
W.H. Odell, Plant Manager APPROVAL DATE: "!'*/i 9 nU m4A - _
'0' A K 309 O R PDR A Q1
I l- ODCM PAGE CHANGE
SUMMARY
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(( _ MAINE' YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL l r l ABSTRACT l The Maine Yankee Nuclear Power Station Off-Site Dose Calculation l Manual (MY ODCM) contains the approved methods to estimate the doses and l radionuclide concentations occurring beyond the boundaries of the plant caused [ by normal plant operation. -(The site boundary is shown in Appendix D, SITE [ BOUNDARY) With initial approval by the U.S. Nuclear Regulatory Commission I [ and the MYNPS Plant Management and approval of subsequent revisions by the [ Plant Management (as per the Technical Specifications), this ODCM is suitable to show compliance where referred to by the Plant Technical Specifications. ,% Sufficient documentation of each method is provided to allow regeneration of the ' methods with few references to other material. Most of the methods are presented at two levels. The first, Method I, is a linear equation which provides an upper l-bound and the second, Method II, is an in-depth analysis which can provide more realistic estimates. i l i 2/96
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL n v TABLE OF CONTENT'S DER r [- AB S TRA CT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i
' TAB L E O F C ONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . il LI ST O F FIG U RES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v LIS T O F TAB L E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vi 1.0 INTRO D UCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. y . . . . . . . . . . I 2.0 RELEASE OF RADIOACTIVE EFFLUENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 ,
1 2.1 Release of Liquid Radioactive Effluents . . . . . . . . . . . . . . . . . . . ........... .2 2.1.1 Applicability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1.2 Objective . ........... ..... . ....... ....... ..... . ... . . 2 2.1.3 Liquid Effluents: Concentration .................... ..... ... .. 2 i 2.1.4 Liquid Effluents: Dose........................................... 3 ( 2.1.5 Liquid Radwaste Treatment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.2 Release of Gaseous Radioactive Waste . . . . . . . . . . . . . ........ . ... .. ... 5 2.2.1 Applicability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.2.2 O bj ectiv e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 2.2.3 Gaseous Effluents: Dose Rate . . . . . . . . . . . . . . ..... .. ..... .... .6 2.2.4 Gaseous Effluents: Dose From Noble Gases . . . . . . . . . . . . . . . . . . . ... 7 2.2.5 Gaseous Effluents: Dose From Iodine-131. Iodine-133. s Tritium, and Radioactive Material in Particulate Form . . . . . . . . . . . . . . 8 2.2.6 Gaseous Radwaste Treatment System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.3 Radioactive Effluent Monitoring Systems . . . . . . . . . . . . . . . . . . . . . . . . . ........ 1I 2.3.1 Applicability ... ... 4 ............................... ..... . 11 2.3.2 Obj ective . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................. ... 11 2.3.3 Radioactive liquid Effluent instrumentation . . . . . . ................. 11 2.3.4 Radioactive Gaseous Effluent Instrumentation . . . . . . . . . . . . . . . . . ... .. 12 ii 4/98 A U
o , I I 1 MAINE YANKEE ATOMIC POWER COMPANY l
. n OFF-SITE DOSE CALCULATION MANUAL i
TABLE OF CONTENTS (CONTINUED) P.agn [ 2.4 Radiological Environmental Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
~ 2.4.1 Applicability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 2.4.2 Obj ecti ve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 ;
2.4.3 Radiological Environmental Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 i 2.4.4 Land Use Census . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 2.4.5 Interlaboratory Comparison Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 2.5 Radioactive Emuent Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 2.5.1 Applicability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 2.5.2 O bjective . . . . . . . . . . . . . . . . . . . . . . . . . . . . ....... .... .......... 26 2.5.3 Liquid Effluents: Sampling and Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 2.5.4 Liquid Effluents: Instrumentation . . . . . . . . . . . . ................... 26 2.5.5 Gaseous Effluents: Sampling and Analysis . . . . . . . . . . . . . . . . . . . . . . . . 26 2.5.6 Gaseous Emuents: Instrumentation . . . . . . . . . . . . . . . . . . . . . ....... 27
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- 2. 5. 7 B as i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 l 3.0 LIQUID EFFLUENT DOSE CALCULATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.1 Liquid Effluent Dose to an Individual . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 .
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3.1.1 Dose to the Total Body . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.1.2 Dose to the Critical Organ .................................. .... 34 4.0 GASEOUS EFFLUENT DOSE CALCULATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 4.1 Gaseous Effluent Dose Rate . . . . . . . . . . . . . . . . . ........................... 36 4.1.1 Dose Rate to the Total Body From Noble Gases . . . . . . . . . . . . . . . . . . . . . . 36 4.1.2 Dose Rate to the Skin From Noble Gases . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 4.1.3 Dose Rate to the Critical Organ From Radiciodines and Particulates . . . . . . . 38 4.2 Gaseous Effluent Dose From Noble Gases . . . . ...... .......... ... . 39 4.2.1 Gamma Air Dose . . . . . . . . . . . . . . . . . . ............. ............ 39 4.2.2 Beta Air Dose . . . .... .. ...... .. ............... ........ .40 4.3 Gaseous Effluent Dose from Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 (_. 4.3.1 Dose to the Critical Organ . ................. ... ... .... . . 41
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r % MAINE YANKEE ATOMIC POWER COMPANY OFF. SITE DOSE CALCULATION MANUAL OU i TABLE OF CONTENTS (CONTINUED) P.aga ! 5.0 ENVIRONMENTAL MONITORING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 6.0 MONITO R S ETPOINTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 6.1 Liquid Emuent Monitor Setpoints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 [ 6.1.1 Intemal Setpoints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54 [ 6.1.A Extemal Setpoints ................. ............... ..... ..... 55 6.2 Gaseous Effluent Monitor Setpoints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 6.2.1 Allowable Concentrations of Radioactive Materials in Gaseous E m uents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 8 6.2.2 Monitor Response for Gaseous Emuents . . . . . . . . . . . . . . . . . . . . . . . .. 58 7.0 M ETE O RO L OG Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61 l APPENDIX A BASIS FOR THE DOSE CALCULATION METHODS . . . . . . . . . . . . . 63 'Q Q-A.1 Liquid E ftluent Doses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 A.2 Total Body Dose Rate from Noble Gases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 A.3 Skin Dose Rate From Noble Gases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 A.4 Critical Organ Dose Rate From Iodines and Particulates . . . . . . . . . . . . . . . . . . . . . . 69 : A.5 G amma Air Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71 A.6 B e ta Air Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 A.7 Dose from lodines and Particulates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 73 APPENDIX B METEOROLOG Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 78 APPENDIX C ROUTINE REPORTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 80 [ APPENDIX D SITE BOUNDARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 82 REF E REN C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 3 w
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m l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL LIST OF FIGURES - Number Tids P.agn 5.1 Environmental Radiological Samnlino Locations l- Within 1 Kilometer of Maine Yankee . . . . . . . . . . . . . . . . r. . . . . . . . . . . . . . . . . . 50 5.2 Environmental Radiological Samnling Locations [ Outside of 1 Kilometer of Maine Yankee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 [ 5.3 Direct Radiation Monitorine Locations l [ Within i Kilometer o f Maine Yankee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 [ 5.4 Direct Radiation Monitorine Locations [ Outside of 1 Kilometer of Maine Yankee . . . . . .... ....... ............. .53 6.1 Maine Yankee Liquid Radwaste System . . . . . . .................... . .... 59 6.2 Maine Yankee Gaseous Radwaste System . ....... ...... .. ... ... 60 APP.D S ite B o undary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 2 lO l l - 1 i l l l l v 4/98 r is t
p l l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL (V\ LIST OF TABLES Number Ilds P.agt 1
- 2.1 Radioactive Liquid Effluent Monitoring Instrumentation 14 2.2 Radioactive Gaseous Effluent Monitoring Instrumentation 15 2.3 Radiological Environmental Surveillance Program 20 l 2.4 Detection Capabilities for Environmental Sample Analysis Lower Limits ofDetection 22 l 2.5 Reporting Levels for Radioactivity Concentrations l in Environmental Samples 25 2.6 Radioactive Liquid Waste Sampling and Analysis Program 28 2.7 Radioactive Gaseous Waste Sampling and Analysis Program 30 2.8 Radioactive Waste Oil Incineration Sampling and Analysis Program 32 3.1 Maine Yankee Dose Factors for Liquid Releases 35 4.1 Maine Yankee Dose Factors for Noble Gas Releases 43 l
4.2 Maine Yankee Dose Factors for Iodine, Tritium. [ and Particulate Releases For Primary Vent Stack 44 4.3 Maine Yankee Dose Factors for Iodine, Tritium, and Particulate Released Via the Auxiliary Boiler 45 I [ 4.4 Maine Yankee Dose Factors for Tritium and Particulates Released Via the [ Fuel Building Exhaust 46 5.1 Radiological Environmental Monitoring Stations 48 l l ['7.1 Maximum Off-Site Long Term Average Atmcspheric Dispersion Factors 62 l A-1 Usage Factors for Various Liquid Pathways at Maine Yankee 65 A-2 Usage Factors for Various Gaseous Pathways at Maine Yankee 75 A-3 Environmental Parameters for Gaseous Effluents at Maine Yankee 76 vi 4/98 i L l
e MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL
1.0 INTRODUCTION
i l The purpose of this manual is to provide methods to ensure compliance with the dose
' requirements of Appendix I to 10 CFR Part 50 (Reference 1). Each method is based on a plant-specific application of the models presented in Regulatory Guide 1.109 (Reference 2).
Methods are included to calculate the doses to individuals from both gaseous and liquid . L releases from the plant. Under normal operations, experience has shown that the plant will be - operated at a small fraction of the dose limits. For this reason, the dose evaluations are presented , at different levels of sophistication. 'Ihe first method being the most conservative, but simplest to use; the second method requiring a full analysis following the guidance presented in Regulatory Guide 1.109 (Reference 2). The first method, Method I, is based on a critical organ, critical age group, and critical receptor location; as such, it provides a conservative estimate of the doses. If the dose limits are l met by application of the first method, no further analysis will be required. If, however, it
- indicates that the dose limits may be approached or exceeded, a more realistic estimate may be i obtained by application of the second method.
f The second method, Method II, will calculate the dose to seven organs of four age groups for s potentially critical individuals; It is based on measured releases for each nuclide, site-specific parameters, and measured meteorological parameters. . Method II is more accurate, but less conservative than Method I, and will be used to assess doses for the Estimated Dose and [ Meteorological Summary Report. 1 .
- Liquid effluent dose calculation methods are presented in Section 3. Gaseous effluent dose
_ calculation methods in Section 4. In both Sections relevant Technical Specifications are followed by the appropriate Method I dose equations. When necessary, Method II analyses may
. be performed by applying the site-specific parameters and measured meteorological parameters to the appropriate dose equations specified in Regulatory Guide 1.109 (Reference 2). The basis for each of the dose calculation methods is described in Appendix A.
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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 2.0 RFT FASE OF RADIOACTIVE EFFLUENTS 2.1 - Releace of Liould Radioactive EfYInents 2.1.1 Anolicability The requirements in this section apply at all times to the release of all liquid waste discha'r ged [ from the plant which may contain radioactive materials. j s 2.1.2 Obiective The objective is to establish conditions for the release ofliquid waste contaimng radioactive materials and to assure that all such releases are within the concentration limits specified j radioactive in 10 CFR Part 20, and also assure that the releases from the site of radioactive j j materials in liquid wastes (above background) are kept "as low as is reasonably achievable" in # accordance with 10 CFR Part 50, Appendix I.
. 1 2.1.3 Liquid ERluents Concentration
- 1. The concentration of radioactive materialin liquid effluents released from the site to unrestricted areas shall be limited to the concentrations specified in 10 CFR, Part 20, .
Appudix B, Table 2, Column 2, for radionuclides other than noble gases, and 2 x 10" J microcuries/ml total activity concentration for all dissolved or entrained noble gases. Remedial Action: With the concentration of radioactive material released from the site to unrestricted areas exceeding the above limits, without delay take action to restore the concentration to within the above limits. Basis. These requirements are provided to ensure that the concentration of radioactive .
. materials released in liquid waste effluents from the site to unrestricted areas (at the point of discharge into Back River; discharge from the submerged multiport diffuser) will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2.
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F v l MAINE YANKEE ATOMIC POWER COMPANY ( OFF-SITE DOSE CALCULATION MANUAL ! l l d This limitation provides additional assurance that the levels of radioactive materials in bodies of l - ( water outside the site will result in exposures within (1) Section II.A design objectives of l Appendix I,10 CFR Part 50, to a member of the public; and (2) the limits of 10 CFR Part 20 to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope, and its ECL in air (submersion) was converted to an equivalent concentration in water using the methods described in Internal Commission on Radiological ! Protection (ICRP), P;.blication 2 (Reference 3). 2.1.4 Liould Effluents Dmg
- 1. The dose or dc,, .ommitment to a member of the public from radioactive materials in liquid effluents released from the site to unrestricted areas shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrem to the total body, and to less than or equal to 5 mrem to any organ; and
- b. During any calendar year to less than or equal to 3 mrem to the total body and i less than or equal to 10 mrem to any organ.
Remedial Action: With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission a report within 30 days from the end of the quarter. The report shall identify the cause(s) for exceeding the limit (s) and define the corrective actions to be taken to reduce the releases and j the corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. . l Remedial Action: With the calculated dose from the release of radioactive materials in liquid effluents exceeding twice the above limits, calculations should be made including direct j radiation contributions from significant plant sources to determine whether the limits of 40 CFR 190 (Reference 4) have been exceeded. l I i i l l 6/94 n ' L] 3
r . MAINE YANKEE ATOMIC POWER COMPANY ,fm OFF-SITE DOSE CALCULATION MANUAL
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[ If such is the case, prepare and submit a report to the Commission within 30 days. The report shall define the corrective acticn to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits and include the schedule for achieving conformance with the limits. If the release condition resulting in violation of 40 CFR Part 190, has not already been [ corrected, the report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the requestis complete. l Basis: These requirements are provided to implement the guidance of Sections ILA, III.A, and IV.A of Appendix I,10 CFR Part 50. The specification provides the required operating flexibility and, at the same time, assures that the releases of radioactive material in liquid l effluents will be kept "as low as is reasonably achievable" as set forth in Section IV.A of l Appendix I. In addition, since the facility is located on a saltwater estuary, the release of ! radioactive waste in liquids will not result in radionuclide concentrations in finished drinking l water, which would be in excess of the requirements of 40 CFR Part 190. The dose calculations performed in accordance with the methods and parameters in this ODCM implement the guidance in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The remedial action requiring calculations when releases exceed two times the design . 4 objectives is included to assure that appropriate reports and requests for variance are made should effluents exceed the limits set forth in 40 CFR Part 190. 2.1.5 Liould Radwaste Treatment t
- 1. The Liquid Radwaste Treatment System shall be used in its designed modes of l I
operation to reduce the radioactive materials in the liquid waste prior to its discharge when the estimated doses due to the liquid effluent from the site, when averaged with all other liquid releases over the last 31 days, would exceed 0.06 mrem to the total body, or 0.2 mrem to any organ. j i 10/98 I t 4
MAINE YANKEE ATOMIC POWER COMPANY ) I l OFF-SITE DOSE CALCULATION MANUAL i Remedial Action: With radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission a report with the next Annual Radioactive Effluent Release Report which includes the following information:
- a. Explanation of why liquid waste was being discharged without treatment and in excess of the above limits, identification of any inoperable liquid waste equipment which prevented treatment prior to discharge, and the reason for the inoperability; l b. Actions taken to restore the inoperable equipment back to operable status; and l c. Summary description of action (s) taken to prevent a recur'rence.
1 Basis: The requirement that the appropriate portions of the Liquid Radwaste System (as indicated in this ODCM) be used when specified provides assurance that the releases of L radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." l This specification implements the requirements of 10 CFR Part 50.36a and the design objective guidance given in Section II.D of Appendix I to 10 CFR Part 50. l The specified limits governing the use of appropriate portions of the Liquid Radwaste i Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I,10 CFR Part 50, for liquid effluents. 2.2 Release of Gaseous Radioactive Waste i 2.2.1 Applicability The requirements of this section apply at all times to the releases of all gaseous waste
- [ - discharged from the plant which may contain plant-related radioactive materials.
- 2.2.2 Objective r
l The objective is to establish conditions in which gaseous waste containing radioactive ! materials may be released and to assure that all such releases are within the dose limits specified in 10 CFR Part 20 and also assure that the releases of radioactive materials in gaseous waste (above background) from the site are kept "as low as is reasonably achievable" in accordance with 10 CFR 50, Appendix 1. 10/98
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MAINE YANKEE ATOMIC POWER COMPANY ! OFF-SITE DOSE CALCULATION MANUAL
' 2.2.3 Gaseous Effluents Dose Rate -
- 1. The dose rate (when averaged over one hour) due to radioactive materials released in gaseous effluents from the site to areas at and beyond the site boundary shall be limited to the following:
- a. For noble gases to less than or equal to 500 mrem / year to the total body, and less i than or equal to 3,000 mrem / year to the skin; and
- b. ' For Iodine-131. Iodine-133, tritium, and radioactive matenals in particulate form with half-lives greater than eight days to less than or equal to 1.500 mrem / year to any organ.
1 Remedial Action: With the dose rates averaged over a period of one hour exceeding the above limits, without delay take action to decrease the release rate to comply with the limit. Basis These requirements are provided to ensure that the dose rate at any time at the site [ area boundary and beyond from gaseous effluents from all effluent release points combined [ (i.e., primary vent stack. fuel building exhaust, and auxiliary boiler discharges when burning [ ' waste oil) will be within the annual dose limits of 10 CFR Part 20 while still providing operational flexibility, compatible with considerations of health and safety, which may temporarily result in releases higher than the absolute value of the concentration values in Appendix B. Reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a member of the public in an unrestricted area to annual doses exceeding the limits specified in 10CFR 20.1001-20.2402 is provided. For members of the public who may at times be within the site boundary area, the occupancy-t
~ ime will be sufficiently low to compensate for any increase in the atmospheric diffusion
, factor above that at the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site area boundary to less than or equal to 500 mrem / year to the total body, or to less than or equal to 3,000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the milk-infant pathway to less than or equal to 1,500 mrem / year for the nearest real milk animal to the plant. L 4/98 tO x) l l 6
MAINE YANKEE ATOMIC POWER COMPANY
- OFF SITE DOSE CALCULATION MANUAL Q 2.2.4 ' Gaseous Efnuents Dese From Noble Gases
- l. - The air dose due to noble gases released in gaseous emuents from the site to areas at and beyond the site boundary shall be limited to the following: .
~
- a. .During any calendar quarter to less than or equal to 5 mrad for gamma radiation, ,
i and less than or equal to 10 mrad for beta radiation; and i
- b. During any calendar year to less than or equal to 10 mrad for gamma radiation, l and less than or equal to 20 mrad for beta radiation. ,
Remedial Action: With the calculated air dose from radioactive noble gases in gaseous [ emuents exceeding any of the above limits, prepare and submit a report to the Commission within 30 days from the end of the quarter. The report shall identify the cause(s) for exceeding limit (s) and define the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous emuents and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. Basis: These requirements are provided to implement the guidance of Sections II.B III.A, and IV.A of Appendix I,10 CFR Part 50. The limiting condition for operation implements the guides set forth in Section II.B of Appendix I. This section provides the required operating flexibility, and, at the same time, assures that the releases of radioactive material in gaseous emuents from all emuent release points combined will be kept "as low as is reasonably achievable." Sampling and analysis requirements of Section 2.5 implement the guidance in Section III.A of Appendix I, i.e., that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through the appropriate pathways is unlikely to be substantially underestimated. The appropriate dose equations are specified in the ODCM equations for determining the air doses at the site area boundary and beyond, and are based upon the historical average atmospheric conditions. 10/98-p V 7
g MAINE YANKEE ATOMIC POWER COMPANY g OFF-SITE DOSE CALCUCATION MANUAL U 2.2.5 Gaseous Effluents Dose From Iodine-131. Iodine-133. Tritium. and Radioactive l Material in Particulate Form
- 1. The dose to a member of the public from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than eight days in gaseous effluents released to areas at and beyond the site boundary shall be limited to the following:
- a. During any calendar quarter to less than or equal to 7.5 mrem to any organ; and
- b. During any calendar year to less than or equal to 15 mrem to any organ. l
- c. Less than 0.1% of the limits specified in 2.2.5.1.a and b as a result of barning contaminated oil.
Remedial Action: With the calculated dose &om the release ofIodine-131, Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than eight days in [ gaseous effluents exceeding any of the above limits, prepare and submit a report to the Commission within 30 days from the end of the quarter. The report shall identify the cause(s) for exceeding the limit (s) and define the corrective 4 , actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. Remedial Action: With the calculated dose &om the release of radioactive materials in j
~
gaseous effluents exceeding twice the limits in Section 2.2.4 or Section 2.2.5, calculations should be made including direct radiation contributions from significant plant sources to determine whether the limits of 40 CFR 190 have been exceeded. [ If such is the case, prepare and submit a report to the Commission within 30 days. The report shall define the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits and include the schedule for achieving conformance with the limits. If the release condition resulting in violation of 40 CFR Part 190, has not already been [ corrected, the report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. 10/98 ( , 8 1
p i i MAINE YANKEE ATOMIC POWER COMPANY- i o _ OFF-SITE DOSE CALCULATION MANUAL ! !O V Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. Basia: These requirements are provided to implement the guidance of Sections II.C, III.A. and IV.A of Appendix I to 10 CFR Part 50. The limiting conditions for operation are the guides set forth in Section II.C of Appendix 1. The specification provides the required . , operating flexibility and at the same time assures that the releases of radioactive materials in L -[ _ gaseous effluents from all effluent release points combined will be kept "as low as is , i reasonably achievable." The ODCM calculational methods implement the guidance m ) l Section III.A of Appendix I that conformance with the guides of Appendix I be shown by l calculational procedures based on models and data such that the actual exposure of a member j ! . of the public through appropriate pathways is unlikely to be substantially underestimated. ! These equations also provide for determining the actual doses based upon the historical ! ( L average atmospheric conditions. < 1 The release rate specifications for Iodine-131, Iodine-133, tritium, and radioactive matenal in i particulate form with half-lives greater than eight days are dependent on the existing radionuclide pathways to man in areas at and beyond the site boundary.
- The pathways which are examined in the development of these calculations are
- 1. Individual inhalation of airborne radionuclides.
- 2. Deposition of radionuclides onto green leafy vegetation with subsequent consumption by man.
l 3. Deposition onto grassy areas where milk animals and meat-producing animals graze with consumption of the milk and meat by man: and
- 4. Deposition on the ground with subsequent exposure to maa.
I-The remedial action requiring calculations if releases exceed two times the design objectives is included to assure that appropriate reports and requests for variance are made should effluents exceed the limits set forth in 40 CFR Part 190. 4/98 9 L
n MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL
. (D V '2.2.6 Gaseous Radwaste Treatment Svstem
- 1, The Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment
. System shall be used to reduce radioactive materials in gaseous waste prior to their '
discharge when the estimated gaseous effluent air doses due to gaseous effluent j L releases from the site to areas at and beyond the site boundary would exceed 0.2 mrad ; for gamma radiation and 0.4 mrad for beta radiation over 31 days. l The Ventilation Exhaust Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the estimated doses due to gaseous effluent releases from the site to areas at and beyond the site boundary would exceed 0.3 mrem to any organ over 31 days. Remedial Action: With gaseous waste being discharged without processing through l appropriate treatment systems, as defined in the ODCM and in excess of the above limits, ' prepare and submit to the Commission a report with the next Annual Radioactive Effluent ! Release Report that includes the following information:
- a. - Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reasons for the ;
inoperability; V
- b. Action (s) taken to restore any inoperable equipment to operable status; and
- c. Summary description of action (s) taken to prevent a recurrence.
B. asis: The requirement that the appropriate portions of the Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be
.[ kept "as low as is reasonably achievable." This section implements the requirements of 10
[ CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives of Appendix I to 10 CFR Part 50. The action levels governing the use of '
. appropriate portions of the Gaseous Radwaste Treatment System were specified as a suitable fraction of the guides set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.
d L 12/97
~10
~
T. L; _- . , MAINE YANKEE ATOMIC POWER COMPANY l
. OFF-SITE DOSE CALCULATION MANUAL 2.3 Ruicactive Effluent Monitoring Svstems 2.3.l' Apolicability The requirements in this section apply at all times to Radioactive Emuent Monitoring Systems which perform a surveillance, protective, or controlling function on the release of
[ radioactive effluents from the plant. 2.3.2 Obiective The objective is to assure the operability of the Radioactive Effluent Monitoring Systems to perform their design functions. 2.3.3 RMioactive Liould Effluent Instrumentation j
- 1. The radioactive liquid effluent monitoring instmmentation channels shown in Table 2.1 shall be operable with their alarm / trip setpoints set to ensure that the limits of Section 2.1.3.1 are not exceeded during periods of release of radioactive material through the pathway monitored.
The alarm / trip setpoints of these channels shall be determined in accordance with the methodology in this ODCM. Remedial Action: With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than a value which will ensure that the limits in Section 2.1.3.1 are met, without delay:
- a. Take action to suspend the release of radioactive liquid effluents monitored by the affected channel, or
- b. Declare the channel inoperable, or change the setpoint so it is acceptably conservative.
Remedial Action: With less than the minimum number of radioactive effluent monitoring instrumentation channels operable, take action shown in Table 2.1. Exert reasonable efforts
, to:
- a. Retum the instrument (s) to operable status within 30 days;and
[ b. If unsuccessful, explain in the next Annual Radioactive Effluent Release Report the reason for the delay in correcting the inoperability. L 10/98
. 11
l l I MAINE YANKEE ATOMIC POWER COMPANY L OFF-SITE DOSE CALCULATION MANUAL l ' O Buis: The radioactive liquid effluent instrumentation is p'rovided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases ofliquid effluents. The alarm / trip setpoints for these instruments are to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, ! 63, and 64 of Appendix A to 10 CFR Part 50. 2.3.4 Rndioactive Gaseous Effluent Instrumentation
- 1. The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 2.2 shall be operable with their alarm / trip setpoints set to ensure that the limits in Section 2.2.3.1 are not exceeded during release of radioactive material via this pathway.
l l The alarm / trip setpoints of these channels shall be determined in accordance with the methodology in this ODCM. l I Remedial Action: With a radioactive gaseous process effluent monitoring instrumentation l channel alarm / trip setpoint less conservative than a value which will ensure that the limits in Section 2.2.3.1 are met, without delay take action to: i a.. Suspend the release of radioactive gaseous effluents monitored by the affected channel,
- b. Or declare the channel inoperable or change the setpoint so it is acceptably l conservative.
l Remedial Action: With less than the minimum number of radioactive effluent momtoring instrumentation channels operable. take action shown in Table 2.2. Exert reasonable efforts to:
- a. Return the instrument (s) to operable status within 30 days; and l [ b. If unsuccessful, explain in the next Annual Radioactive Effluent Release Report the l
reason for the delay in correcting the inoperability. j i I ll (- 2/96 l 12 L.
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL Hasn The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for these instruments are to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The operability and use of this instrumentation is consistent with the requirements of General ,
' Design Criteria 60,63, and 64 of Appendix A to 10 CFR Part 50.
O O 1/92 13
E MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.1 , Radioactive Liauid Effluent Monitorine Instrumentation Minimum Channels Remedial Operable Action Instrument i
- 1. Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release 1
- a. Liquid Radwaste Effluent Line (Test Tanks) (1)
- 2. Gross Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release Service Water System Effluent Line (1) 2 a.
- 3. Flow Rate Measurement Devices 3 I
- a. Liquid Radwaste Emuent Line (1)
Table Notation ACTION 1 With the number of channels operable less than required by the minimum channels operable requirement, emuent releases may continue provided that prior to initiating or continuing a release:
- 1. At least two independent samples are analyzed in accordance with Section 2.5, Table 2.6.
- 2. At least two technically qualified members of the facility staffindependently verify the release rate calculations and discharge valving.
i Othenvise, suspend release of radioactive emuents via this pathway. ACTION 2 With the number of channels operable less than required by the minimum channels operable requirement, emuent releases via this pathway may continue provided that, at least once per 24 hours, grab samples are collected and analyzed [ for radioactivity at a lower Limit of detection of no more than 10 pCi/ml. ACTION 3 With the number of channels operable less than required by the minimum channels operable requirements, effluent releases via this ?athway may continue provided the flow rate is estimated at least once per eight Tours during actual i release. Pump performance curves generated in situ may be used to estimate How. 10/98 s l l 14 L
(-, I MAINE YANKEE ATOMIC POWER COMPANY l- OFF-SITE DOSE CALCULATION MANUAL l TABLE 2.2 Rndioactive Gaseous EfYluent Monitorino Instrumentation Minimum Channels Remedial Doerable Action Instrument l ,
- 1. Primary Vent Stack (1) 6
- a. Particulate Sampler Filter **
(1) 4
- b. Emuent System Flow Rate Measuring Device (1) 4
- c; . Sampler Flow Measuring Device
- 2. Fuel Building Exhaust Vent (1) 5
- a. Noble Gas Activity Monitor (1) 6
- b. Particulate Sampler Filter" (1)- 4
- c. Emuent System Flow Rate Measuring Device l
(1) 4
- d. Sampler Flow Measuring Device ,
i Table Notation ACTION 4 With the number of channels operable less than required by the minimum channels operable requirement, eMuent releases via this pathway may continue provided the flow rate is estimated at least once per eight hours. ACTION 5 - With the number of channels operable less than required by the minimum channels operable requirement, emuent releases via this pathway may continue provided grab samples are taken at least once per 24 hours and these samples are analyzed [- for radioactivity within 24 hours. ACTION 6 With the number of channels operable less than required by the minimum channels l operable requirement:
- Take immediate action to suspend activities that may increase the potential for particulate releases via this pathway until such time that the channel is j
restored or auxiliary sampling equipment is operational, mid
- Within 24 hours, commence the collection of samples with auxiliary equipment. For the Fuel Building Exhuast Vent, equipment may include an air sampler at the intake of the exhaust vent.
" Normal shutdown for filter changeout does not constitute inoperability.
10/98 15 I
( l-MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL [ 2.4 Radiological Environmental Monitorina Procram . [ A program shall be provided to monitor the radiation and radionuclides in the environs of the [ plant. The program shall provide (1) representative measurements of the radioactivity in the [ highest potential exposure pathways, and (2) verification of the accuracy of the effluent [ monitonng program and modeling of environmental exposure pathways. The program shall l [ (1) be contamed in the ODCM,(2) conform to the guidance of Appendix 1 to 10 CFR Part 50, [ and (3) include the following: [ 1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the [ environment in accordance with the methodology and parameters in the ODCM. [ 2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE [ BOUNDARY are identified and tha: modifications to the monitoring program are made [ if required by the results of this census, MR [ 3) Participation in a Inter-laboratory Comparison Program to ensure that independent [ checks on the precision and accuracy of the measurements of radioactive materials in [ environmental sample matrices are performed as part of the Quality Assurance Program [ for environmental monitoring. 2.4.1 Aeolicability This section applies at all times to radiological environmental surveillance and land use census. 2.4.2 Obiective The objective of this section is to verify that plant operations have no significant radiological effect on the environment and that continued operation will not result in radiological effects l detrimental to the environment. The program also shall verify that any measurable concentrations of radioactive materials related to plant operations are not significantly higher than expected based on effluent measurements and modeling of the environmental exposure pathways. 2.4.3 Radiolocical Environmental Monitorine
- 1. The Radiological Environmental Monitoring Program shall be conducted as specified in i Table 2.3 with Lower Limits of Detection (LLDs) as specified in Table 2.4.
- 2. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 2.3, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
2/98 16
m f-MAlNE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL L 3. With the level of radioactivity in an environmental sampling medium at a location specified in l Table 2.3 exceeding a reporting level of Table 2.5 when averaged over any calendar quarter. l [ ?repare and submit to the Commission with the next Annual Radioactive Effluent Release L Report, following receipt of the laboratory analyses, a report which includes an evaluation of i any release conditions, environmental factors, or other aspects which caused the limits of Table 2.5 to be exceeded. When more than one of the rac ionuclides in Table 2.5 are detected
. in the sampling medium, this report shall be submitted if: "-
concentration (D + concentration (2) + ...>l.0 , reporting level (1) reporting level (2) .! l
)
Exception: When radionuclides other than those in Table 2.5 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an ] individual is equal to or greater than the calendar year limits in Sections 2.1.4, 2.2.4 and 2.2.5. This report is not required if the measured level of radioactivity was not the result of alant effluents; however. in such an event, the condition shall be reported and
- describec;in the Annual Radiological Environmental Operating Report.
l
- 4. With milk samples no longer available from one or more of the sample locations required by Table 2.3, identify the new location (s) if available, for obtaining replacement samples and add to the Radiological Environmental Monitoring Program within 30 days. The specific location (s) from which samples were no longer available may then be deleted from the Monitoring Program. Identify the cause of the samples no longer being available and identify the new location (s) for obtaining available
[ replacement samples in the next Annual Radiological Environmental Operating Report. Basis. The radiological environmental monitoring required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures ofindividuals resulting from the station operation. This monitoring program thereby supplements the Radiological Effluent Momtoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurement and modeling of the environmental exposure pathways. Program changes may be initiated based on operational experience. A two-zone sample collection network has been established for environmental surveillance. Samples are collected in Zone I at locations in the vicinity of the plant where concentrations of plant effluents may be detectable.
' These samples are compared to samples which have been collected simultaneously at locations in Zone II where the concentration of plant effluents is expected to be negligible.
. The Zone Il samples provide a running background which will make it possible to distinguish significant radioactivity introduced into the environment by the operation of the plant from that introduced by weapons testing or other sources.
12/97 L 17
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL The detection capabilities required by Table 2.4 are considered optimum for routine environmental measurements in industrial laboratories. It'should be recognized that the LLD is defined as an a paQri(before the fact) limit representing the capability of a measurement system and not as an a costeriori (after the fact) limit for a particular measurement. This does not preclude the calculation of an a costeriori LLD for a particular measurement based upon the actual parameters for the sample in question. 2.4.4 Land Use Census
- 1. An annual and use census within the distance of five miles shall be conducted to identify e iocation of the r.earest milk animal, the nearest residence, and the nearest garden of 50 m2.
In lieu of a garden census, broad leaf vegetation of at least three different kinds may be sampled at or near the site boundary in two different sections.
- 2. With a land use census identifying a location (s) which yields a calculated dose l commitment (via the same exposure pathway) at least twice than at a location from 1 which samples are currently being obtained in accordance with Section 2.4.3.1. identify l the new locations in the next Annual Radiological Environmental Operating Report. l l
If permission from the owner to collect samples can be obtained and sufficient sample j volume is available. then this new location shall be added to the Radiological Environmental Monitoring Program within 30 days. The sampling location having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted at this time.
- 3. The land use census shall be conducted at least once per 12 months between the dates of June 1 and Octoiser 1. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report. -
Hasir This specification is provided to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census. The addition of new sampling locations to Section 2.4.3.1 based on the land use census is limited to those locations which yield a dose commitment at least twice the calculated dose commitment at any location currently being sampled. This eliminates the unnecessary changing of the Environmental Radiation Monitoring Program for new locations which, within the accuracy of the calculation, contribute essentially the same to the dose or dose commitment as the location already sampled. The substitution of a new sampling point for one already sampled when the calculated difference in dose is less than a factor of 2 would not be expected to result in a significant increase in the ability to detect plant effluent-related nuclides. Changes in the location of monitoring locations are not to be done lightly since frequent changes disrupt time series and may make interpretation of data more difficult. 1/92 18
F l l l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL j 2.4.5 Interlaboratorv Comearison Program , Analyses shall be performed on applicable radioactive environmental samples supplied as part of,an interlaboratory comparison program which has been approved by NRC, if such a program eX1stS. If analyses are not performed as required above, a report shall be made in the next Annual Radiological Environmental Operating Report. I Basis: Participation in an NRC-approved interlaboratory comparison program (if one exists) provides quality assurance for the environmental laboratory, similar to programs in place for i other environmental monitoring efforts, such as that for water quality. 1 I l l l l 1 1/92 l 19 I
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a 8 Y ( s l i i C 1 5 1 5' Mp 1 I NL i s y AA l PU a n _ MN A _ OA e s _ e _ CM lp n t a _ R r E N m o g) _ a i . O t et e WI S c e t rw . OT l _ a e t e/ 0 0 0 0 0 0 P LA 4 t nD vg nk 3 6 3 6 3 5 2 ef 1 2 1 2 1 1 _ CU m o i/ i I MC E. n s dC np 2 2 L l B o r i t a( O T A A i v i. m h AC d T E n 1 i s e s E E F d S r r e u e s E O fo w b e u K ND s o. e 1 s y b e a a A TE i i t G m y a YI l i r l E S - b a o l i m N: 7 1 p e Cp l l a t i I I AO C a 0 Cp lu ') 0 n M cm i o i/ 1 7 5 6 0, 5 1 t t i 0 0 0 0 3 f c e rC a f o o t e P (p e e u D e n r o l u a v l a v - b a a _ r , s s i t A t s i s i x x e e y y a a _ w w - r )l * * ' - h el 5' 8 h t i 4 0 0 5 0 3 5 0 3
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p I-MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.4 (Continued) ' i-Table Notation a. The LLD is the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability and that only a 5% ' probability exists of falsely concluding that a blank observation represents a "real" signal.
, For a particular measurement system (which may include radiochemical separation):
4.66
- Sm LLD = E
- V
- 2.22
- Y
- Exp (-l
- At) where:
LLD is the "a priori" lower limit of detection as defined above (as picocuries per unit mass or volume). 4.66 is a constant derived from the K ,,, and K% values for the 95% confidence level. S is the standard deviation of the background counting rate or of the counting rate of a
- blank sample as appropriate (as counts per minute).
i E is the counting efficiency (as counts per disintegration). V is the sample size (in units of mass or volume). 2.22 is the number of disintegration per minute per picocuries. l Y is the fractional radiochemical yield (when applicable). A is the radioactive decay constant for the particular radionuclide. At is the elapsed time between sample collection and analysis. Typical values of E, V, Y, and at can be used in the calculation. 1/92 L l 23
l 4 l l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2 A (Continued) Table Notation This equation results in an LLD in terms of picocuries. For the purposes of Section 2.5 (Tables ' [! 2.6,2.7, and 2.8), where the required LLD is set forth in microcuries, the terms 2.22 in the denominator should be replaced by 2.22E6, which is the number of disintegrations per minute per microcurie. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the ssmples (e.g., Potassium-40 in milk samples). The analyses shr.Il be performed in such a manner that the stated LLDs will be achieved under routice conditions. Occasionally, background fluctuations, unavoidably small sample sizes, the presence ofinterfering nuclides, or other uncontrollable circumstances may render these LLDs unavailable. In such cases, the cor.tributing factors will be identified and described in the Armual Radiological Environmental Operating Report.
- b. It should be recognized that the I. LD is defined as an a pnori (before the fact) limit representing the capability of a measurement sy stem and not as an a costeriori (after the fact) limit for a particular measurement. This does not preclude the calculation of an a costeriori LLD for a particular 'neasurement based upon the actual parameters for the sample in question and appropriate decay correction parar.ieters, such as decay while sampling and during analysis.
- c. Parent only.
- d. If the measured concentration minus the three standard deviation uncertainty is found to exceed the specified LLD, the sample does not have to be analyzed to meet the specified LLD.
- e. . This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the listed nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 5.9.1.5.
- f. The Ba-140 LLD and concentration can be determined by the analysis ofits short-lived daughter product. La-140, subsequent to an eight-day period following collection. The calculation shall be predicated on the normal ingrowth equations for a parent-daughter situation and the assumption that any unsupported La-140 in the sample would have decayed to an insignificant amount (at least 3.6% ofits original value). The ingrowth equations will assume that the supported La-140 activity at the time of collection is zero.
2/98 24
i MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.5 Reoortina Levels for Radioactivity Concentrations in Environmental Samoles Airborne Particulate Fish and Food Water orGas Invertebrates Milk Products Analysis [gCjd} (oCi/m3) (oCi/ka/ wet) {gCid) (pCi/l) H-3 20,000* Mn-54 1,000 30,000 ; Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95$ 400 1-131 2* O.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-1406 200 300 4 If no drinking water pathway exists, a value of 30,000 pCi/l may be used. b Parent only. If no drinking water pathway exists, a value of 20 pCi/l may be used. 1/92 25
p l l- MAINE YANKEE ATOMIC POWER COMPANY l OFF-SITE DOSE CALCULATION MANUAL 2.5 Rndioactive Effluent Monitoring' l
-2.5.1 Annlicability -
This section applies to monitoring radioactive emuents, both liquid and gaseous. 2.5.2. Obiective The objective of this section is to specify the nature and frequency of radioactive emuent l monitoring requirements.- ! s 2.5.3 Liauid Effluents: Samoling and Analysis
- 1. Liquid radioactive waste sampling and activity analysis shall be performed in accordance with Table 2.6.
- 2. The results of the radioactivity analysis shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Section 2.1.3.1.
- 3. Cumulative dose contributions from liquid emuents for the current calendar quarter and .
. the current calendar year shall be determined in accordance with the methodology and l parameters in this ODCM at least once per 31 days.
2.5.4 Liauid Effluents: Instrumentation
.- Discharge ofliquid radioactive emuents shall be continuously monitored with the alarm / trip setpoints of the monitor set in accordance with the methods outlined in the ODCM such that the requirements of Section 2.1.3 are met. l 1
2.5.5 Gaseous Effluents: Samoling and Analvsis
- 1. Gaseous radioactive waste sampling and activity analysis shall be performed in j
[: accordance with Table 2.7 or 2.8. as applicable. ! i
- 2. The cumulative doses due to gaseous emuents for the current calendar quarter and l calendar year sha:1 be determined to.be within the limits of Sections 2.2.3,2.2.4, and 2.2.5 in accordance with the methodology and parameters of the ODCM at least once per l
! 31 days. 2/98 I i ,.. ) i ! 26 i
p a MAINE YANKEE ATOMIC POWER COMPANY
- OFF-SITE DOSE CALCULATION MAuNUAL
- 3. Doses due to gaseous releases from the site to areas at or beyond the site boundary shall be compared with the limits of Section 2.2.6 in accordance with the methodology and parameters in the ODCM at least once per 31 days. If all gaseous releases for the period have been processed via a design mode of the Gaseous Radwaste Treatment System, dose l estimates for compliance with Section 2.2.6 are not required.
2.5.6 Gaseous Effluents Instrumentation [ With the exception of waste oil incineration discharges, radioactive gaseous effluents shall be continuously monitored with the alarm / trip setpoints of the monitors set in accordance with the { methods outlined in the ODCM such that the requirements of Section 2.2.3 will be met. 2.5.7 E uis The sampling analysis and instrumentation requirements set forth in this Specification provide reasonable assurance that all significant radioactive releases will be monitored and that the effluents l will not result in exceeding the requirements of 10CFR20. l l 2/98 27
c MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.6 l Radioactive Liouid Waste Samolina and Analvsis Program l Minimum Lower Limit of l Sampling Analysis Type of Activity Detection (LLD) Liould Release Tvee Freauenevh Frecuencvh Analvsis (uCi/ mil'
- A. Batch Waste PR PR Principal Gamma 5 x 103 Release Tanksd Each Batch Each Batch Emitters' I-131 1 x 104 PR M Dissolved and 1 x 10 5 One Batch /M Entrained Gases (Gamma Emitters) I PR M" H-3 1 x 10-5 Each Batch Composite Gross Alpha 1 x 10" PR Q Sr-89. Sr 90 5 x 10 8 Each Batch Composite 5 Fe-558 1 x 104 B. Plant D* W Principal Gamma 5 x 10d Continuous Grab Sample Composite5 Emittersr Releases'
[ (Turbine Building L Sump) W M H-3 1 x 10 5 Grab Sample Composite 5 Gross Alpha ! x 10 3 W Q Sr-89. Sr-90 5 x 10 8 [ Grab Sample Composite
- FE-558 1 x 104
[ 2/98 28 l
C~ - l 1
. MAINE YANKEE ATOMIC POWER COMPANY
!' OFF-SITE DOSE CALCULATION MANUAL j TABLE 2.6 (Continued) l Table Notation
- a. ' The Lower Limit of Detection (LLD) is defined in Table Notation a of Table 2.4 of Section 2.4.
l
- b. A composite sample is one in which the quantity ofliquid sampled is proportional to the quantity ofliquid waste discharged and in which the method of sampling employed results in a
- specimen which is representative of the liquids released.
- c. To be representative of the quantities and concentrations of radioactive materials in liquid I effluents, samples shall be collected during release and composited in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be 4 thoroughly mixed in order for the composite sample to be representative of the effluent release. I
~ d.- A batch release is the discharge ofliquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.
- e. A continuous release is the discharge ofliquid wastes of a non-discrete volume; e.g., from a volume of system that has an input flow during the continuous release.
- f. The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 104 This list does not mean that only these nuclides are to be considered. Other gamma peaks which are identifiable, together with the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level.
[ [ g. If, after a period of two years, the results indicate that Fe-55 is likely to contribute 1% or less of the total dose attributable to this pathway, the licensee may discontinue the analysis.
- h. Frequency notations: PR = Prior to Release D = Daily W = Weekly M = Monthly Q = Quarterly
' l 2/98 1 l i 29 l l
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i TABLE 2.7-Radioactive Gaseous Waste Samnline and Analysis Program Minimum Lower Limit of Sampling Analysis Type of Activity Detection (LLD) Gaseous Release Tyne Freauencv d Freauencvd Analysis (uCi/mlY A. Primary Vent M M Principal Gamma 1 x 10" Stack Grab Emitters' Continuous6 W Principal Gamma 1 x 10-" ] Emitters
- Particulate (I-131, Others)
Sample l Continuous6 M Gross Alpha 1 x10 " Composite Particulate . J Sample Continuous
- Q Sr-89, Sr-90 1 x 10-" l Composite Particulate i Sample
[ 4 M' Grab M Tritium 1 x 10 B. Fuel Building M M Principal Gamma 1 x 10" Exhaust Vent Grab Emitters
- i Continuous6 W Principal Gamma 1 x 10 "
Emitters
- Particulate -(I-131, Others) i Sample Continuous
- M Gross Alpha 1 x 10 "
Composite Particulate Sample Continuous6 Q Sr-89, Sr-90 1 x 10 " l Composite Particulate I Sample Continuous6 Noble Gas Noble Gases 1.0 x 10-8 Monitor Gross Beta Or Ganuna j W Grab W Tritium 1 x 10 4 8/98 i i 30
l MAINE YANKEE ATOMIC POWER COMPANY
- OFF-SITE DOSE CALCULATION MANUAL TABLE 2.7 (Continued)
Table Notation a.' The Lower Limit of Detection (LLD) is defined in Table Notation a of Table 2.4 of Section 2.4.
- b. - The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Sections 2.2.3,
' 2.2.4, and 2.2.5.
- c. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported in the Annual Radioactive Effluent Release Report. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported in the Annual Radioactive Effluent Release Report. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide but as "not detected." . When unusual circumstances result in LLDs higher than requirei, the reasons shall be documented in the Annual Radioactive Effluent Release Report.
'd. Frequency notations are the same as in Table 2.6.
[ e. Tritium grab samples shall be taken weekly whenever the refueling cavity is flooded. I i l l l l (. 6/98 i 31
i l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.8 [ Radioactive Waste OilInceneration Sampling And Analysis Program
- i Minimum .. Lower Limit of Sampi.mg
,;3 Type of Activity Release Type Detection (LLD)
"*" Frequencyd 7" ( Ci/ml)* l i
; A. Incinerated Oil Principal Gamma PR6 Grab Sample PR* Each Batch 5 x 104 Emitters
[- Table Notiation [ a. Incinerated oil may be discharged via points other than the primary vent stack (e.g., auxiliary [ boilers). [ b! The release of radioactive material shall be accounted for based on the pre-release grab sample [ data. The grab sample shall be representative of the contaminated oil in liquid form prior to [ incineration. [ c. The Lower Limit of Detection (LLD) is defined in Table Notification a. of Table 2.4 of [ Section 2.4. [ d. Frequency notations are the same as in Table 2.6. 2/98 32
MAINE YANKEE ATOMIC POWER COMPANY l OFF-SITE DOSE CALCULATION MANUAL l 1 3.0 LIOUID EFFLUENT DOSE CALCULATIONS 3.1 i inuid Effluent Dose to an individual Section 2.1.4.1 limits the dose or dose commitment to a member of the public from radioactive materials in liquid effluents released from the site to Back River: a.~ During any calendar quarter to less than or equal to 1.5 mrem to the total body, and to less than or equal to 5 mrem to any organ; and b.' During any calendar year to less than or equal to 3 mrem to the total body, and to less ! than or equal to 10 mrem to any organ. 3.1.1.a Dose to the Total Body (Method n
'Ihe total body dose, D., in mrem for a liquid release is: J D. = K E Qi DFL. i (3-1)
' i where:
Q3 is the total activity released for radionuclide i,in Ci(for strontiums use the most recent measurement available). DFL. i is the site specific Total Body Dose Factor for radionuclide i, in mrem /Ci (see Table 3.1). I K ~ is equal to 935/F,; where F, is the average (typically monthly average) dilution flow of [ the Circulating / Service Water System at the point of discharge from the multiport l [ diffuser (in ft2/sec). For waste tank discharge periods when there is no plant dilution flow [ via the Circulating / Service Water System above 3800 gpm,4F is set at a minimum flow [ of 8.5 ft)/sec. due to tidal flushing from the plant forebay through the diffuser system. l 10/98 1 33 1
T l MAINE YANKEE ATOMIC POWER COMPANY ! OFF-SITE DOSE CALCULATION MANUAL ! l 3.1.1.b Dose to the Total Body (Method m l . Method 11 consists of the models, input data and assumptions (bioaccumulation factors, shore-width factor, dose conversion factors, and transport and buildup times) in Regulatory Guide , 1.109, Rev.1 (Reference 2), except where site-specific data or assumptions have been identified in the ODCM. The general ec untions (A-3 and A-7) taken from Regulatory Guide 1.109, and used in the derivation of the simpliaed Method I approach as described in the Bases Section A.1, are also applied : to Method II assessments, except that d oses calculated to the whole body from radioactive emuents l ; are evaluated for each of the four age groups to determine the maximum whole body dose of an i age-dependent individual via all existing exposure pathways. Table A-1 lists the usage factors for [ Method II calculations. During periods when the Circulatmg/ Service Water System provide dilution [ flow of effluent relec.as from the discharge diffuser to the Back River, the mixing ratio for the [ diffuser's nearfield mixing zone is set at 0.10. For periods when no plant dilution flow is provided by [ the Circulating / Service Water System, the mixing ratio may be redued to 0.020 in Method II [ calculations to account for tidal flushing of effluent materials from the plant's forebay. [ 3.1.2.a Dose to the Critical Organ (Method D
- The critical organ dose, D., in mrem for a liquid release is:
D. = K E Qi DFL % (3-2) I where: Q is the total activity released for radionuclide i, in Ci (for strontiums use the most recent measurement available).
~s DFL % is the site specific Critical Organ Dose Factor for radionuclide i, in mrem /Ci (see Table 3.1).
K is equal to 935/F4 ; where F4 is the average (typically monthly average) dilution flow of [ Service Water System at the point of discharge from the multiport [ the Circulating diffuser (in ft /sec). For waste tank discharge periods when there is no plant d [ via the Circulating / Service Water System above 3800 gpm, F, is set at a minimum flow [ of 8.5 ft3/sec. due to tidal flushing from the plant forebay through the diffuser system. 3.1.2.b Dose to the Critical Organ (Method m Method II consists of the models, input data and assumptions (bioaccumulation factors, shore-width factor, dose conversion factors, and transport and buildup times) in Regulatory Guide 1.109, Revision 1 (Reference 2), except where site-specific data or assumptions have been identified in the ODCM. The general equations (A-3 and A-7) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I a pproach as described in the Bases Section A.1, are also applied to Method II assessments, except that doses calculated to critical organs from radioactive effluents are evaluated for each of the four age groups to determine the maximum critical organ of an - age-dependent individual via all existing exposure pathways. Table A-1 lists the usage factors for [ Method II calculations.During wriods when the Circulating / Service Water System provide dilution [ flow of effluent releases from tae discharge diffuser to the Back River, the mixing ratio for the [ diffuser's nearfield mixmg zone is set at 0.10. For periods when no plant dilution flow is provided by [ the Circulating / Service Water System, the mixing ratio may be redued to 0.020 in Method II l ' [ calculations to account for tidal flushing of effluent materials from the plant's forebay. 10/98 34
l l MAINE YANKEE ATOMIC POWER COMPANY )
- OFF-SITE DOSE CALCULATION MANUAL TABLE 3.1 4
Maine Yankee Dose Factors for Liauid Releases Total Body Critical Organ Dose Factor Dose Factor mrem /Ci mrem /Ci Nuclide DFL . DFL_ H-3 2.96E-07 2.96E-07 Na-24 2.46E-05 2.83E-05 Cr-51 1.54E-05 1.45E-03 Mn-54 4.26E-03 2.55E-02 l Mn-56 1.89E-06 4.09E-05 Fe-55 1.24E-02 7.53 E-02 l , Fe-59 8.58E-02 6.54E-01 Co-58 2.21E-03 1.35E-02 l Co-60 4.79E-02 7.80E-02 I Zn-65 2.68E-01 5.38E-01 i Sr-89 2.13E-04 ' 7.45E-03 Sr-90 3.16E-02 1.29E-01 Zr-95 5.03E-04 1.73E-02 Mo-99 2.95E-05 2.62E-04 Tc-99m 4.06E-07 1.98E-06 Sb-124 1.34E-03 9.36E-03 1-131 2.07E-04 9.86E 02 I-132 2.54E-06 3.29E-06 I-133 2.46E-05 1.13E-02 I-135 7.12E-06 4.17E-04 Cs 134 2.79E-02 3.12E-02 Cs-137 2.92E-02 3.41 E-02 Ba-140 1.54E-04 3.41E-03 Ce-141 2.81E-05 9.13E-03 W-187 6.28E-06 1.32E-03 Ag-110m 7.92E-03 6.26E-01 Sb-125 4.81E-03 6.81E-03 Other 1.51E-01 3.40E+00 l/92
'4 35
r: MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL
. 4.0 GdSEOUS EFFLUENT DOSE CALCUL ATIONS -
4.1. Gaeacus Effluent Dose Rate
- Section 2.2.3.1 limits the dose rate (when averaged over I hour) due to radioactive materials released in gaseous effluents from the site to areas at and beyond the site boundary:
- a. . for noble gases: less than or equal to 500 mrem /yr to the total body, and less than or
. equal to 3000 mrem /yr to the skin, and;
- b. for Iodine-131. Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than 8 days; less than or equal to 1500 mrem /yr to any organ.
4.1.1.a Dose Rate to the Total Body From Noble Gases (Method n [- The total body dose rate, D , in mrem /yr from noble gases released via the primary vent stack is: 5,, = 1.06E d, ors, (4-1) where: [. d, is the release rate of noble gas i released via the primary vent stack, in uCi/sec: and l
- DFB i is the Total Body Dose Rate Factor for noble gas i, in mrem-m3 /pCi-yr (see l Table 4.1).
- 1.06 is as defined in Section A.2 of Appendix A. in sec-pCi/m3 -uCi ,
I l l [ The to:al body dose rate, d,,, in mrem /yr from Kr-85 released via the Fuel Building Exhaust l l ! [. is: [ 5,, = 4 . 2 0 s - 0 5 = d,, . ,, (4-la) L [- Where: L j [ d,, , ,, is the release rate of Kr-85 released via the Fuel Building exhaust, in units of l l [- Ci/sec; and i l i [ 4.20E - 0.5 is defined in Section A.2, in units of mrem - sec / Ci - yr. [. The total dose rate from the site is the combination of dose rates from the Primary Vent Stack t
-[ and the Fuel Building ediuast.
4/98 3e
o MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 4.1.1.b Dose Rate to the Total Bodv From Noble Gases (Method in Method II consists of the model and input data (whole body dose factors) in Regulatory Guide 1.109, Revision 1 (Reference 2), except where site-specific data or assumptions have been identified in the ODCM. The general ec uation (B-8) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method : approach as described in the Bases Section A.2, is also - applied to a Method II assessment. No credit for a shielding factor (Sr) associated with residential i structures is assumed. Concurrent meteorology with the release period may be utilized for the gamma atmospheric dispersion factor identified in Appendix B for the release point from which recorded effluents have been discharged. In sectors where the site boundary is adjacent to Back River, the total body dose rate will be evaluated on the nearest opposite shoreline where the potential exists for uncontrolled occupancy. On-site areas or areas with limited and controlled occupancy will be evaluated with those occupancy factors included. The most restrictive location in any of the 16 sectors will be used in determining the dose rate. 4.1.2.a Dose Rate to the Skin From Noble Gases (Method D [ The skin dose rate, d,,,, , in mrem /yr from noble gases released via the primary vent stack is: b,,,,
- Es beFl where:
[ d, is the release rate of noble gas i released via the primary vent stack, in uCi/sec; and orl is the Combined Skin Dose Rate Factor for noble gas i, in mrem-sec/pCi-yr (see Table 4.1). The skin dose rate, d,,,, , in mrem /yr. from Kr-85 released via the Fuel Building Exhaust is: [ [ d,,,, = 1.39E - 02 = d,,,,, [ [ Where: [ [ d,, , ,, is the release rate of Kr-85 released via the Fuel Building Exhaust, in Ci/sec; and [ 1.39E - 02 is as defined in Section A.3, in units of mrem-sec/ Ci - yr. [ [ The total dose rate from the site is the combination of dose rates from primary vent stack and the fuel l [ building exhaust. 4/98 I 37 1
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 4.1.2.b Dose Rate to the Skin From Noble Gases (Method in Method II consists of the model and input data (skin dose factors) in Regulatory Guide 1.109,
. Revision 1 (Reference 2), except where site-specific data or assumptions have been identified in this ,
ODCM. The general equation (B-9) taken from Regulatory Guide 1.109, and used in the derivation of ! the simplified Method I approach as described in the Bases Section A.3, is also applied to a Method II . assessment. No credit for a shielding factor (S,) associated with residential structures is assumed. Concurrent meteorology with the release period may be utilized for the gamma atmospheric dispersion factor and undepleted atmospheric dispersion factor identified in ODCM Appendix B for the release point from which recorded effluents have been discharged. In sectors where the site boundary is adjacent to Back River, the Skin Dose Rate will be evaluated on the nearest opposite , shoreline where the potential exist for uncontrolled occupancy. On-site areas or areas with limited and controlled occupancy will be evaluated with those occupancy factors included. The most restrictive location in any of the 16 sectors will be used in determining the dose rate. i 4.1.3.a Dose Rate to the Critical Orean From Radiciodines and Particulates (Method D ! The dose rate to the critical organ, d , in mrem /yr from Iodine-131 Iodine-133, tritium, and radioactive materials in particulate form dith half-lives greater than 8 days released via the , [ primary vent stack is: d.. - tkoml.. [' where: [ d, is the release rate of radionuclide i released via the primary vent stack in Ci/sec; and j [ om[,, is the site specific Ciritical Organ Dose Rate Factor for radionuclide i, [ in mrem-sec/ Ci - yr. (See Table 4.2) , [ The dose rate to the critical organ, 5" , in mrem /yr from tritium and radioactive materials in [ particulate form with half-lives greater than eight days released to the atmosphere from the Fuel [ Building Exhaust is: d" e =E,d" orrs y, (4-3a) [ [ Where: [ [ d'" is the release rate of radionuclide i released via the fuel building exhaust, in Ci/sec; [ and ) [ orrs -'" is the site specific Critical Organ Dose Rate Factor for radionuclide i for a gaseous [ release from the fuel building exhaust, in mrem - sec/ Ci - yr (See Table 4.4) [ -[ . _ [ The total dose rate from the site is the combination of dose rates from primary vent stack and the fuel -[ building exhaust. 4/98 l 38
p I MAINE YANKEE ATOMIC POWER COMPANY
- OFF-SITE DOSE CALCULATION MANUAL l '
4.1.3.b Dose Rare to the Critical Oroan From Radiciodines and Particulates (Method In
~ Method II consists of the models, input data and assumptions in Appendix C of Regulatory Guide 1.109, Revision 1 (Reference 2), except where site-specific data or assumptions have been identified in this ODCM (see Tables A-2 and A-3). The entical organ dose rate will be determined i
i based on the location (site boundary, nearest resident, or farm) of receptor pathways as identified in the most recent annual land use census, or b pathways such as ground plane, inhalation,y mgestionconservatively of stored and 1 assumin the existence vegetables, of all possible milk, and meat at an otT-s(ite location of maximum potential dose. Concurrent meteoro ogy with ! may be utilized for determination ot atmospheric dispersion factors in accordance with Appendix B l for the release point from which recorded effluents have been discharged. The maximum critical organ dose rates will consider the four age groups independent! , and take no credit for a shielding factor Sc associated with residential structures. Site boun ocations adjacent to the river will be evalua(ted)on the nearest opposite shoreline. Mud flats expose at low tide will in factor of 0.037 for evaluation of doses at those locations. 4.2 Gaseous Effluent Dose From Noble Gases Section 2.2.4.1 limits the air dose due to noble gases released in gaseous effluents to areas at and beyond the site boundary to the following:
- a. During any calendar quarter: less than or e than or equal to 10 mrad for beta radiation; and qual to 5 mrad for gamma radiation, and less
- b. During any calendar year: less than or equal to 10 mrad for gamma radiation, and less than or equal to 20 mrad for beta radiation.
4.2.1.a Gamma Air Dose (Method n [ The gamma air dose D%, in mrad from noble gases released via the primary vent stack is: D[,, = 0.034 E Q, DF[ (4-4) i where: i [ Q4 is the total ac interest. In C,tivity i; and of noble gas i released via the primary vent stack during the period of DF7 is the Gamma Dose Factor to air for noble gas i, in mrad-m'/pCi-yr (see Table 4.1). 1 0.034 is as de,ined in Sect;on A.5 of Appendix A. in pCi-yr/Ci-m'. l [ The gamma al.- dose, 9[,, , in mrad from Kr-85 released from the Fuel Building Exhaust is:
. I of,, = 1.42s - 06 o,,.,, (4-4a) l Where:
- is the total activity of Kr-85 released via the fuel building exhaust during the period of l l (( Q o.c interest, in Ci
- and
[ 1.42E - 06 is as defined in Section A.5 of Appendix A, in units of mrad /ci l [ The total dose rate from the site is the combination of dose rates from primary vent stack and the fuel [ building exhaust. 4/98 39
- f. ]
[ - MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 4.2.1.b Gamma Air Dose (Method in Guide 1.109, Revision 1 (Reference 2), except where(site-specific data or assumptions h identified in this ODCM. The general equations (B-4 and B-5 taken from Regulatory Guide 1.109, l and used in the derivation of the simplified Method I approach)as described in the Bases Sectio I are also applied to Method II assessments. Concurrent meteorology with the release period may be for the release utilized for theeffluents which recorded gamma haveatmospheric dispersion been discharged. factors For sectors adjacent(see to tApoendix B) he Back River, the nearest point f l opposite shoreline with an assumed potential occupancy factor of 100% will be used to evaluate doses. On-site areas with limited and controlled occupancy will be evaluated with those occupancy l factors included. 4.2.2.a Beta Air Dose (Method D [ The beta air dose, D(, in mrad from noble gases released via the primary vent stack is:
- o[,, = 0.037 E o, or[ (4-5)
L l \ where: l [ Qi is the total activity of noble gas i released via the primary vent stack during the period ofinterest, in Ct; and j I of,, 2 is the Beta Dose Factor for noble gas i, in mrad-m /pCi-yr (see Table 4.1). 0.037 is as defined in Section A.6 of Appendix A,in pCi-yr/Ci-m 3. r The beta air dose, o[,, in mrad from Kr-85 released from the Fuel Building Exhaust is: o[,,=6.37s-04.o,,,,, (4-Sa) [ Where: [ Qx,.u is the total activit [ ofinterest,in Ci;y andof Kr-85 released from the fuel building exhaust during the period l [ 6.37E - 04 is as defined in Section A.6 of Appendx A,in mrad /ci. 1 [ The total dose from the site is the combination of doses from the primary vent stack and the fuel
- l. building exhaust.
4/98 l 40
c: l l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL
~
4.2.2.b Beta Air Dose (Method IIT ) ! and assumptions in Regulatory Method 11 consists Reference 2of the models,
, except input data (dose factors) data or assum where site-specific Guide identified1.109, Revision 1 (The general) equations (B-4 and B-5) taken from Reguiatory Guid in the ODCM.
and used in the derivation of tiie simplified Method I approach as described in the Bases Section A.6, are also applied to Method II assessments. Concurrent meteorology with the release period may be
- utilized for the atmospheric dis,persion factors (see Appendix B) for the release point from which -
recorded effluents have been discharged. For sectors adjacent to the Back River, the nearest opposite shoreline with an assumed potential occupancy factor of 100% will be used to evaluate doses. On-site areas or areas with limited and controlled occupancy will be evaluated with those occupancy factors included. 4.3 _ aaseous Ef Suent Dose from Iodine-131. lodine-133. Tritium. and Radioactive Material in
) articulate ;orm Sections 2.2.5.1.a and 2.2.5.1.b limit the dose to a member of the Iodine-133, tritium, and radioactive materials inparticulate livesform with greater thanhalfpublic eight from days in gaseous effluents released to areas at and beyond the site boundary to the following:
- a. during any calendar quarter: less than or equal to 7.5 mrem to any organ; and
- b. during any calendar year: less than or equal to 15 mrem to any organ.
ODCM Section 2.2.5.1.c limits the dose to a member of the public from these same radionuclides to less than 0.1 percent of the limits noted above, as a result of burning contaminated oil. 4.3.1.a Dose to the Critical Organ (Method I) - ) The dose to the critical organ. Dco, in mrem from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than eight days released via the [ pnmarv vent stack is: D,. .= 2 Q, DFG,,, (4-6) I s where: [ Qi is the total activity of radionuclide i released via the primary vent stack during the period ofinterest,in Ci; and DFG ,,3 is the site specific Critical Organ Dose Factor for radionuclide i for a gaseous release [ from the pnmary vent stack, m mrem /Ci(see Table 4.2). 4/98 [ r 41 J
p MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL ! The dose to the critical organ, D* ' in mrem from Iodine-131, Iodine-133, tritium, and
- radioactive materials in particulate form wEth half-lives greater than eight days released to the atmosphere from the auxiliary boiler due to the burning of contaminated waste oil is:
o,", = E o*". orwy, (4 7) L . where: g'* = is the total activity of radionuclide i released via the auxiliary boiler stack during the period ofinterest,in Ci; and DFWu, is the site specific Critical Organ Dose Factor for radionuclide i for gaseous release from the auxiliary boiler, in mrem /Ci(see Table 4.3). l The dose to the critical organ, D,,, in mrem from Tritium and radioactive materials in particulate form with half-lives greater than eight days released to the atmosphere from the Fuel
. Building Exhaust is:
[ D'8,9 = E Qys
- DFFBa, (4-7a)
[ where:
' Q/s s the total activity of radionuclide i released via the fuel building exhaust during that period ofinterest,in Ci; and
[ DFFBa, is the site specific Critical Organ Dose Factor for radionuclide i for a release from the fuel building exhaust, in mrem /Ci (See Table 4.4). The total dose from the site is the combination of doses from the primary vent stack, the l
. auxiliary boiler, and the fuel building exhaust.
l I 4.3.1.b Dose to Critical Oroan (Method II) l Method II consists of the models, input data and assumptions in Appendix C of Regulatory Guide 1.109, Revision 1 (Reference 2), except where site-specific data or assumptions havehen identified in this ODCM (see Tables A-2 and A-3). The entical organ dose will be determined based on the location (site boundary, nearest resident, or farm) of receptor pathways, as identified in the most recent annual land use census, or by conservatively assummg the existence of all possible pathways (such as ground plane, inhalanon, ingestion of stored and leafy vegetables, milk, and meat) at an off-site location of maximum potential dose. Concurrent meteorology with the release period may be utilized for determination of atmospheric dispersion factors in accordance with Appendix B for the release point from which recorded effluents have been discharged. The maximum critical and use a shielding factor (Sp) of 0.7 organ dosewith associated willresidential consider structures. the four age Mudgroups independently, flats exposed at low tide in areas where the Back River is i adjacent to the site boundary will include an occu cy factor of 0.037 for evaluation of doses at l those locations. Only the inhalation and ground ane exposure pathways are included in the assessment of doses on the mudflats (for 10 CF 50, Appendix I, and 40 CFR 190 considerations). f 4/98 l 42 l
I MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 4.1 . l Maine Yankee Dose Factors for Noble Gas Releases [ - For Primarv Vent Stack Total Body Combined Skin Gamma Air Beta Air i Dose Rate Factor Dose Rate Factor Dose Factor Dose Factor - . 1 (mrem-m3 /pCi-yr) (mrem-sec/uCi-yr) (mrad-m3 /pCi-yr) (mrad-m3 /oCi-yr) Nuclide DFB: DF? DFL DFJ Kr-83m 7.56E-08 2.28E-05 1.93E-05 2.88E-04 Kr-85m 1.17E-03 3.17E-03 1.23E-03 1.97E-03 Kr-85 1.61E-05 1.60E-03 1.72E-05 1.95E-03 Kr-87 5.92E-03 1.88E-02 6.17E-03 1.03E-02 Kr-88 1.47E-02 2.07E-02 1.52E-02 2.93E-03 Kr-89 1.66E-02 3.23E-02 1.73E-02 1.06E-02 Kr-90 1.56E-02 2.78E-02 1.63E-02 7.83E-03 Xe-131m 9.15E-05 7.46E-04 1.56E-04 1.llE-03
' 2.15E-04 1.56E-03 3.27E-04 1.48E-03 Xe-133m Xe-133 2.94E-04 7.78E-04 3.53E-04 1.05E-03 i Xe-135m 3.12E-03 4.80E-03 3.36E-03 7.39E-04 l Xe-135 1.81 E-03 4.46E-03 1.92E-03 2.46E-03 l Xe-137 1.42E-03 1.62E-02 1.51 E-03 1.27E-02 Xe-138 8.83E-03 1.57E-02 9.21E-03 4.75E-03 l Ar-41 8.84E-03 1.41E-02 9.30E-03 3.28E-03
[ For Soent Fuel Building Exhaust Vent Total Body Combined Skin Gamma Air Beta Air , Dose Rate Factor Dose Rate Factor Dose Factor Dose Factor 3 3 (mrem-m3 /pCi-yr) (mrem-sec/uCi-yr) (mrad-m (mrad-m /pCi-yr) Nuclide DFB: DF? D /gi-yr) DF Kr-85 1.61 E-05 1.39E-02 1.72E-05 1.95E-03 (Included in dose equation) l \ 4/98 43 l
7 l j 1 MAINE YANKEE ATOMIC POWER COMPANY. OFF-SITE DOSE CALCULATION MANUAL i TABLE 4.2 Maine Yancee Dose Factors for Iodine. Tritium. ! [ and Particu ate Releases For Primarv Vent Stack j l Critical Organ Critical Organ . Dose Factor Dose Rate Factor - (mrem /Ci) (mrem-sec/uCi-yr) ! [ Nuclide DFG_ DFG"e H-3 3.56E-04 1.12E-02 C-14 2.16E-01 6.81 E-00 l Cr-51 9.34E-03 2.95E-01 l Mn-54 1.09E+00 3.44E+01 Fe-59 1.05E+00 3.31 E+01 Co-58 5.55E-01 1.75E+01 Co-60 1.18E+0l' 3.72Et02 i Zn-65 5.55E+00 1.75E-02 Sr-89 1.76E+01 5.55E-02 Sr-90 6.68E+02 2.l l E-04 Sb-124 1.94E+00 6.12E-01 I-131 1.12E+02 3.53E-03 I-133 1.16E+00 3.66Ev01 Cs-134 2.42E+01 7.63E-02 I Cs-137 2.49E+01 7.85E-02 Ba-140 1.73E-01 5.46E-00 Ce-141 2.62E-01 8.26Ee00 q Ce-144 5.97E+00 1.88E-02 Ag-110m 1.02E+01 3.22E+02 , Sb-125 1.93E+00 6.09E-01 ( Other 4.51E+00 1.42E-02 ) i I
)
i l f 4/98 t 44
i MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 4.3 . Maine Yank ee Dom Factors for Iodine. Tritium. and 3artien ue Re eM Via the Auxi iary Boiler
- Critical Organ
- Dose Factor (mrem /Ci)
Nuclide DFW , H-3 2.19E-03 C-14 1.77E+00 Cr-51 2.I1E-02 Mn-54 2.39Ev00 Fe-59 2.35E+00 Co-58 1.24E+00 Co-60 2.59E+01 Zn-65 1.22E-01 1 Sr-89 3.86E+01 Sr-90 1.48E"03 Sb-124 4.34Ev00 I-131 2.48E"02 1-133 3.28Et00 Cs-134 5.31E-01 Cs-137 5.45E-01 Ba-140 5.96E-01 Ce-141 6.00E-01 Ce-144 1.33E+01 l10m - 2.24E+01 Ag SS- 125 4.25E+00 Other 9.93E+00
- DFWw, for use with the burning of contaminated waste oil.
l i 3/93 45 t
r i l MAINE YANKEE ATOMIC POWER COMPANY
. OFF-SITE DOSE CALCULATION MANUAL
! [ TABLE 4.4 . [ Maine Yankee Dose Factors for Tribium and Particulates 1 j' [ Released Via the Fuel Building Exhaust an Dose Factor Critical Organ Dose Rate Nuclide Critical (mremOrfCi) DFFBw, Factor (mrem - sec/ Ci - yr) )j DFFB'w, [ H-3 3.10E-03 9.78E-02 [ C-14 1.88E + 00 5.93E + 01 [ Cr-51 2.14E - 02 7.38E - 01 [ Mn-54 2.40E + 00 9.49E + 01 1 [ Fe-55 1.09E + 00 3.44E + 01 [ Fe-59 2.42E + 00 8.01E + 01 [ Co-58 1.25E + 00 4.48E + 01 [ Co-60 2.60E + 01 1.16E + 03 [ Zn-65 1.27E + 01 4.10E + 02
'[ Sr-89 3.88E + 01 1.22E + 03
[ Sr-90 1.49E + 03 4.70E + 04 [ Ag-110m 2.31E + 01 7.76E + 02 j [. Sb-124 4.38E + 00 1.46E + 02 [ Sb-125 4.28E + 00 1.67E + 02 1 [ Cs-134 5.42E + 01 1.80E + 03 [ Cs-137 5.54E + 01 1.89E + 03 j [ Ba-140 6.34E - 01 2.03E + 01 ,. [ Ce-141 6.04E - 01 1.92E + 0' [ Ce-144 1.33E + 01 4.19E + 0'[ [ Other 1.06E + 01 3.34E + 02 1 4/98 46 t l
E j l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 5.0 ENVIRONMENTAL MONITORING The Radiological Environmental Monitoring Stations are listed in Table 5.1. The locations of i these stations with respect to the Maine Yankee facility are shown on the maps in Figures 5.1 through 5.6.
\
i l I
!/92 47
r l MAINE YANKEE ATOMIC POWER COMPANY i OFF-SITE DOSE CALCULATION MANUAL TABLE 5.1 . Radiological Environmental Monitorina Stations
- Distance Direction Exposure Pathway Sam ale Location From the From the and/or Samnie and Designated Code5 Plant (knD Plant
- 1. AIRBORNE AP/CF-11 Montsweag Brook 2.7 NW
[ (RADIOIODINE & AP/CF-13 Bailey Farm (ESL) 0.7 NE NNE PARTICULATE) AP/CF-14 Mason Steam Station 4.8 AP/CF-16 Westport Firehouse 1.8 S [ AP/CF-29 Dresden Substation 20.1 N
- 2. DIRECT RADIATION
[ TL-1 Old Ferry Rd. 0.9 N TL-2 Old Ferry Rd. 0.8 NNE 1, TL-3 Bailey House (ESL) 0.7 NE l,, TL-4 Westport Island, Rt.144 1.3 ENE L TL-5 MY Information Center 0.2 ENE
; TL-6 Rt.144 and Greenleaf Rd. 1.0 E TL-7 Westport Island, Rt.144 0.9 ESE
[ TL-8 MY Screenhouse 0.2 ESE [. TL-9 Westport Island, Rt.144 0.8 SE 0.3 SSE [ TL-10 Bailey Point 4.8 NNE TL-11 Mason Station TL-12 Westport Firehouse 1.7 S l l; l 1 TL-13 FoxoirdIsland 0.3 SSW 1 TL-14 Eaton Farm 0.7 SW TL-15 Eaton Farm 0.8 WSW ' TL-16 Eaton Farm 0.7 W TL-17 Eaton Farm Rd. 0.6 WNW TL-18 Eaton Farm Rd. 0.8 NW [ TL-19 Eaton Farm Rd. 0.9 NNW , TL-20 Bradford Rd., Wiscasset 6.4 N (L TL-21 Federal St., Wiscasset 7.1 NNE TL-22 Cochran Rd., Edgecomb 8.3 NE TL-23 Middle Rd.. Edgecomb 6.4 ENE [ 7.8 E TL-24 River Rd., Edgecomb TL-25 River Rd. and Rt.27 7.7 ESE [ SE
- r. TL-26 Rt. 27 and Boothbay RR Museum 7.9 F TL-27 Barters Island 7.2 SSE
' S TL-28 Westport Island, Rt.144 & East Shore Rd. 7.9 l., TL-29 Harrison's Trailer 6.2 SSW l
'. TL-30 Leeman Farm. Woolwich 7.8 SW
[ TL-31 Barley Neck Rd., Woolwich 6.8 WSW [ TL-32 Baker Farm, Woolwich 7.3 W TL-33 Rt.127, Woolwich 7.4 WNW [ TL-34 Rt.127, Woolwich 7.9 NW TL-35 Rt.127, Dresden 9.1 NNW TL-36 Boothbay Harbor Fire Sta. 12.2 SSE ( 10.7 WSW [ TL-37 Bath Fire Station [_ TL-38 Dresden Substation 20.1 N 12/97 48
- MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 5.1 (Continued) -
Radiolonical Environmental Monitorina Stations
- 2 Distance Direction Exposure Pathway Sample Location From the From the ,
and/or Samnle and Designated Codeb Plant (km) Plant j
- 3. WATERBORNE i l
[ a. Surface WE-12 Plant Outfall' O.3 SSW (Estuary) '(Composite Sample) I [ WE-20 Kennebec River (Grab Sample) 9.5 WSW ! [ b. Groundwater WG-13 Bailey Farm (ESL) 0.7 NE [ WG-24 Morse Well 9.9 W [- c. Sediment from SE-18 Foxbird Island 0.6 S [' Shoreline SE-16 Old Outfall Area 0.6 S l
)
- 4. INGESTION I
[ a. Milk TM-15 Mitman Farm 5.5 S [ TM-18 Chewonki Foundation 1.9 WSW [ TM-25 Hanson Farm 18.3 W i [ b. Fish and FH/MU/CA/HA-ll Long Ledge Area 0.9 S i S [ Invertebrates d FH/MU/CA/HA 11.1 j i
- c. Food Crop' TV-IX Indicator (to be determined) -
l Vegetation TV-lX Indicator (to be determined) TV-2X to be determined - - Footnotes: ! a Sample locations are shown on Figures 5.1 to 5.6. [b With the exception of DIRECT RADIATION locations, Station-lX's are indicator stations and Station- l 2X's are control stations. c A dilution factor of 10 shall be applied to any radioactivity detected in a sample at this station. d The station code letters will vary with the sample media collected. The sampling of all four media types is not required during each samp!,ing penod. , e Food crop sampling is not required while milk sampling is being done. [ MY Audit 94-02 SSCA: 0004 paragraph 3. 12/97 49
t I MADJE YANKEE ATOMIC POWER COMPA.NY OFF-SITE DOSE CALCULATION MANUAL i i FIGME it I l Environmental Radiolodcal Samelinc 9 cations 1 Within i Kilometer of Maine Yarh i ,., . .j =mc- :-
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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL FIGl'RE 12 l Environmental Radioloeical Sameline Locations ' [ Outside of 1 Kilometer of Maine Yankee W
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F l NL-\INE Y.O*KEE ATOMIC POER COMP ANY OFF-SITE DOSE CALCULATION MANUAL
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W MAINE YANKEE ATOMIC POWER COMPANY > OFF-SITE DOSE CALCULATION MANUAL 6.0 MONITOR SETPOINTS [ .6.1 Liould Effluent Monitor Setooints , [ This section describes the methodology to determine alarm / trip se,tpoints ofliquid e31uent l [ monitors specified in Table 2.1, Radioactive Liquid Effluent Momtonng Instrumentation. -
~[ 6.1.1 Intemal Setpoints
[- Intemal monitor setpoints shall be established to monitor compliance with the release
' [ concentration limits specified in Section 2.1.3.1.
[ The total allowable concentration of radioactivity for all releases entering the Back River at any [ [ given time less than shall unity whenbe limitedas calculated tofollows: a total Effluent Concentration Limit Ratio, ECL Ratio, (R) e [ R= E R = [C i shall be equal to or less than 1 (6.1) , [ ECL, [~ Where: [ R = Total ECL ratio (dimensionless) [ R, = ECL ratio (dimensionless) for each individual release l [ C, = diluted activity concentration of radionuclide (i), in Ci/ml, which is the - [ concentration entering the Back River, and is equal to the undiluted [. concentration of radionuclide (i)(CJi times the flowrate through the monitored [ pathway (in gpm) (Qi) divided by the total of the dilution flow (in gpm) (D i ) [ plus the release flowrate (Qi). ' [ = (Cl
- O g
[ (Di + Qi) [ ECL, = Effluent Concentration Limit (ECL) of radionuclide (i) in Ci/mi as [ specified in Section 2.1.3.1. (Includes non-gamma emitters such as [ tritium.) 2/98 f 54 l 6
l l l 1 MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL A pathway factor (PF i) (a value s 1.0) may be applied to each monitor setpoint calculation. Application of the pathway factors shall be such that the total ECL ratio for releases via multiple pathways, should they exist, are maintained less than or equal to one: PF = E PFi shall be equal to or less than 1 Setpoints shall be calculated based on the relationship: Setpoint, = ECL * [(D + Qi)/Qi]
- PFi
- RF (6.2)
Where: Setpointi = Monitor response (CPM) for the release pathway "i" ECL = Effluent Concentration Limit (ECL) as s limiting gamma emitting radionuclide (i)pecified which potentially may in be Section present in 2.1.3.1 of the mo the release pathway ( Ci/ml). D= Minimum expected total Dilution F:ow entering the forebay, which is equal to the available Circulating Water and the Service Water flowrate (gpm). Qi = Maximum expected release flowrate through the monitored release pathway, "i" (gpm). PF,= Pathway Factor as described above (dimensionless). RF = Radiation monitor response factor (sensitivity factor)(cpm / Ci/ml). 6.1.2 Extemal Setpoints [ The Liquid Radwaste (Test Tank) Monitor. RM3801, is also equipped with an extemal alarm / trip setpoint. The intent of this setpoint is to provide assurance that the pre release analysis is representative of the release bemg made through that monitor, and to alert the operator if a problem does exist. This setpoint shall be determined for each release as follows: Calculate the expected radiation monitor response (ER), as follows:- ER = Activity
- RF Where:
ER = Expected radiation monitor response (CPM) Activity = Undiluted activity concentration of gamma emitting radionulides ( Ci/ml) RF = Radiation monitor response factor (sensitivity factor) as determined by the most recent monitor calibration (CPM / Ci/ml) Calculate the setpoint as follows: Setpoint_,i = 2
- ER If the Setpoint_,i calculates to a value less than 1000 cpm, the monitor setpoint should be set at 1000 cpm above background, provided that 1000 cpm is less than the internal setpoint.
4/98 55
f-l l ! . MAINE YANKEE ATOMIC POWER COMPANY l OFF-SITE DOSE CALCULATION MANUAL E In the event that the extemal setpoint alarms and/or trips a release, comply with ACTION 1 of the
' Table Notations for Table 2.1: Radioactive Liquid Emuent Monitoring Instrumentation.
[. If the independent verification is in agreement with the initial analysis, the external setpoint may [ be established up to the value of the mtemal setpoint, and the release may proceed. [ -but-If the independent verification is not in agreement with the initial analysis, the reason for the variation shall be determined, and appropnate corrective action shall be taken prior to recommencing the release. In this event, the extemal setpoint shall be recalculated and reestablished as described above before proceeding with the release, r 2/98 I 1 56
I I l ! MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 6.2 Gaseous Efhent Monitor Setnoints l Section 2.5.6 requires that radioactive gaseous efDuents be continuously monitored with the alarm / trip setpoints of the monitors set to ensure that the requirements of Section 2.2.3 are met. Section 2.5.6 ensures that the dose rate at any time at the site area boundary and beyond from gaseous effluents will be within the annual dose rate limits specified in section 2.2.3. These limits . provide reasonable assurance that radioactive material discharged in gaseous effluents will not result m the exposure of a member of the [ CFR Pan 20, Appendix B,Column Table1. 2,public in an unrestricted area in excess of This section of the ODCM describes the methodology that may be used to determine the setpoints of the gaseous effluent monitors. Gaseous effluent flow paths and release points, as well as the locations and identification numbers of the gaseous efDuent radiation detectors, are shown in Figure i 6.2. The methodology for determining alarm / trip setpoints is divided into two parts. The first consists of calculating an allowable concentration for the radionuclide mixture to be released. The second consists of determining monitor response to this mixture in order to establish the physical settings on the monitors. l l l l l l 4 2/98 57
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 6.2.1 Allowable Concentrations of Radioactive Materials in Gaseous EfY1uents [ The ECL fraction, Re, for each aseous effluent release point is calculated by the relationship Y Appendix B: defined by Note 1 of 10 CFR Part _0, c' (6-5) R' = (X/01 F EL ECL, , , i where: 1 1 [R 3 is the ECL-fraction for the release point j, dimensionless;
'[X/Q] is the most conservative sector site boundary or off-site long- l term average dilution factor (see Table 7.1)(8.99E 06 sec/m3 );
F is the release flow rate (in m 3/sec); Ci is the concentration of radionuclide i, in uCi/cc; [ ECL, is the effluent concentration of radionuclide i as specified in 10 CFR Part 20, Appendix B, Table 2, Column 1, in uCi/ce. [ The ECL-fractions for the various release points are then summed to yield the total ECL-fraction, R: R ER (6-6) 7 3 [ The total ECL-fraction, R, at the most conservative site boundary or off-site location must be less than or equal to one. R $ 1. (6-7)
' 6.2.2 Monitor Resnonse for Gaseous Effluents Normal radioactivity releases consist mainly of well-decaved fission gases. Therefore, monitor response calibrations are performed using fission gas typical of normal releases (mainly Xenon-133).
The total concentration of radioactive materials in gaseous effluents, in uCi/ce, at the monitor is calculated. The calibration curve or constant, in epm /(uCi/cc) is applied to determine the expected cpm for the mix of radionuclides. The setting of the monitor is established at some factor, b, greater than one but less than 1/R (see Equation 6-6). 2/98 58
7. l MAINE YANKEE ATOMIC POWER COMPANY l OFF-SITE DOSE CALCULTAION MANUAL j l i FIGURE 6.1 Maine Yankee Liauid Radwaste Svstem Tenue l l A.rs.ed Ormen Tanns s Pet.r w Pet.r
,r.
2 =, ,,:. _ n.L"*:La (MartTl Funcuon)
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RM-1601 - y iP 'P nMC. WM I l l 2/98 r 59 i L.
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL { FIGURE 6.2 Maine Yankee Gaseous Ralwaste System , 1 l l l
- Flow Indicator dL i
Stack Continuous S@ [ntaintnent Purge s
- Fues Sudeng i f.eaus:
Primary Auxiliary ven t o e ed Areas) e 1 P " Primary l Vent w l Stack l Suelding Radallon Monder
~
- RM-SFP 19 conenuou:
s.wi.r
=
n 3 Fusi SuMilne P = Particulate Prefilter H = HEPA filter 10/98 60
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 7.0 METEOROLOGY The atmospheric dilution factors in the dose calculation methods assume an individual whose behavior leads to a dose higher than expected for anyone else. Since long term (5-year) average ! meteorology is expected to be representative of the area, the location of the critical receptor can be ' predicted by scanning all the reasonable off-site locations to find the location with the most limiting dilution factors. Important off-site locations are: site boundaries and nearest residences in each of the sixteen meteorological sectors, as well as all milk farm locations within five miles of the plant. , Exposure pathways assumed to exist at site boundary locations are direct exposure from radioactive materials in the air, direct exposure from radioactive materials deposited on the ground, ) and exposure from inhalation of radioactive materials. In addition to the pathways present at site ' sp boundary radionuclides locations, exposure in home grown vegeta pathwges. resentinclude Farm locations at each residence all exposure are assumed pathways found at to melude residences plus ingestion of radionuclides in meat and milk. Meteorological data for the year 1986 through 1990 were analyzed for the values of the maximum l i average dilution factors at the important receptor locations described above. Yankee Atomic Electric ' ' Company's (YAEC) AEOLUS-2 computer code (Reference 5) calculated all atmospheric dilution factors. Appendix B briefly describes the YAEC AEOLUS-2 computer code model. Table 7.1 lists the maximum average diluuon factors for all important receptor locations for releases via the plant stack. Each dose and dose rate calculation method incorporates the maximum applicable off-site average dilution factors listed in Table 7.1. The maximum potential dose to a member of the public due to plant stack releases in any year will be conservatively estimated by the dose calculated for a full-time l resident living on a hypothetical milk farm 670 to 700 meters from the plant in the southeast sector. 3/93 61
f l
- MAINE YANKEE ATOMIC POWER COMPANY l
l OFF-SITE DOSE CALCULATION MANUAL TABLE 7.1 i l [ Maximum Off-Site Long Term Averace
; Maximum Long-Term Primary Vent Stack Aux Boiler Fuel Building Vent .' Dispersion Factor
[ Undepleted X/Q (sec/m') 1.18E - 06 (670 m SE) 9.67E - 06 (670 m SE) 1.03E - 05 (670 m SE) [ Depleted X/Q (sec/m3 ) 1.09E - 06 (670 m SE) 8.93 - 06 (670 m SE) 9.51E - 06 (670 m SE) 1 [ 2 D/Q (1/m ) 1.46E-08 (670 m SE) 3.20E-08 (670 m SE) 3.20E - 08 (670 m SE) ) [ Gamma X/Q (sec/m ) 3 1.06E - 06 (670 m SE) 2.17E - 06 (670 m SE) 2.61E - 06 (670 m SE) I 4/98 i 62 L
l MAINE YANKEE ATOMIC POWER COMPANY l OFF-SITE DOSE CALCULATION MANUAL APPENDIX A Basis for the Dose Calculation Methods l A.1 Liauid Efiluent Doses Method I is used to demonstrate compliance with Section 2.1.4 which limits the dose , commitment to a member of the public from radioactive materials in liquid effluents. Liquid pathways contributing to individual doses at the Maine Yankee Nuclear Power Station are: ingestion of fish and shellfish, and direct exposure from shoreline deposits. The potable water pathway and the irrigated foods pathway are not considered since the receiving water is not suitable for either dnnking or irri Guide 1.109 (Reference k)ation. Method I is derived from Equations A-3 a A-7 from shoreline deposits. The use of the methodology of Equations A-3 and A-7 for a 1 curie release of each radionuclide in liquid effluents yielded the dose impact to the critical organ. Table 3.1 lists the resulting site specific total body and critical organ dose conversion factors giving the number of millirem per curie released for each radionuclide. Smce the dose factors of Table 3.1 represent a variety of critical organs, Method I conservatively calculates a critical organ dose consisting of the maximum critical organ for each radionuclide of any of the four age groups, and combines them into a composite individualindependent of age. Except for the site specific values noted below, the parameter values recommended in Regulatory Guide 1.109 (Reference 2) were used to derive the liquid dose factors for Method I. Table A 1 lists the usage factors for liquid pathways utilized in the dose analysis. Liquid effluents discharge from the plant via a submerged multi port diffuser which extends approximatel T . (935 fWsec).y 1000 For the aquatic feet foods intothe pathway the tidal dilution for estuary andofhas the mixing effect a design the diffuser is set at circulatm
, a minimum of 10 to 1 in the Method I dose f' actors (Reference 6). This dilution applies to the edge of
, the initial mixing zone where the effluent has undergone prompt dilution only. For release conditions
, when no circulating,or service water p ps are runnmg, tidal flushing of the plant's forebay provides
, = a conservative dilution factor of 50 to e., mixing ratio = 1/ dilution factor, or 0.020) for near-field
. mixing effects of discharges from the di to the river. For shoreline de sits, the nearest point where tidal flats could be occupied on a recurring basis is in Bailey, Cove w 'ch borders the site on the south and west. The estimated average dilution for Bailey Cove with respect to the discharge is
- approximately 25 to I (Reference 6).
I i 10/98 63
I l i MAINE YANKEE ATOMIC POWER COMPANY i OFF-SITE DOSE CALCULATION MANUAL APPENDIX A - I I Shoreline activities in the vicinity of the site include a commercial worm digging industry along the tidal flats of Montsweag Bay. In the area of the plant (Bailey Cove), a commercial worm digger could occupy the mud flats for as long as 325 hours per year. This occupancy time is applied to both adults and teenagers in the dose calculations. For Method I, the period of time for which sediment is exposed to the contaminated water is fifteen years. This time period represents the approximate mid-point of plant operatin lifetime, and-thus allows for the calculation of a plant lifetime average concentration of radioactivi in sediment. No credit is taken for the decay of activity in transit from the discharge point to the se iment in Bailey Cove. l l 3/93 i 64 l
I l L l MAINE YANKEE ATOMIC POWER COMPANY l OFF-SITE DOSE CALCULATION MANUAL !
- {
l APPENDIXA - l TABLE A-1 l Usace Factors for Various Liould Pathways at Maine Yankee
~ (From Reference 1, Table E-5*, except as noted.
Zero where no pathway exists.) LEAFY POTABLE AGE E VEG. . MILK MEAT FISH INVERT. WATER SHORELINE (KG/YR) (KG/YR) (LITER /YR) (KG/YR) (KG/YR) (KG/YR) (LITER /YR) (HR/YR) Adult 0.00 0.00 0.00 0.00 21.00 5.00 0.00 334.00 " Teen 0.00 0.00 0.00 0.00 16.00 3.80 0.00 67.00 Child 0.00 0.00 0.00 0.00 6.90 1.70 0.00 14.00 Infant 0.00 0.00 0.00 0.00 0.00 0.00 0.00 0.00
- Regulatory Guide 1.109.
" Regional shoreline use associated with mudflats - Maine Yankee Atomic Power Station Environmental Report.
l 3/93 65 L
rz MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL l APPENDIX A A.2 Total Body Dose Rate from Noble Gases L Method I can be used to demonstrate compliance with Section 2.2.3.1.a. which limits total body ~ l dose rate from noble gases released to the atmosphere. Method I applies the methods of Equation B-8 in Regulatory Guide 1.109 (Reference 2) as
.follows:
D. = Sr 3.17E+04 [X/Q]7 E Q, DFB i
~
(A-1) l where: D. is the annual total body dose, in mrem /yr, . Sr is the attenuation factor that accounts for the dose reduction due to shielding provided by residential structures, but for all dose rate calculations is assumed to be equal to I (dimensionless); 3.17E+04 is the number of pCi per Ci divided by the number of seconds per year; J l is the effective long term average gamma dilution factor, in sec/m ; 3 j [X/Q]Y Qi is the annual release rate of radionuclide i, in Ci/yr; and 3 DFB i is the total body gamma dose factor for radionuclide i, in mrem-m /pCi-yr. [ For a release from the primary vent stack, the analysis of Maine Yankee five-year average meteorology presented in Section 7.0 yielded a maximum effective average gamma dilution factor. Y. of 1.06E-06 sec/m3 . The maximum gamma dilution factor was identified for an off-site point X/Q]d 670 meters southeast of the plant. This location is along the opposite shoreline of Back
,ocate from the plant in a sector where the site boundary is adjacent to the river. The maximum gamma dilution factor for the site boundary along the river's near shoreline has been determined to be a more restrictive 3
value (south sector at 129 meters, [X/Q Y = 3.33E-6 sec/m ). However, the definition of site boundary in
~t he Technical Specifications allows for]the use of occupancv factors in assessing, definition of unrestricted area in NUREG-0133 (Reference 7) also does not require dose evaluations over
- water. For those portions of the adjacent shoreline to the site boundary where mudflats are exposed during low tide, an occupancy factor for worm diggers (0.037) is applied to the average gamma dilution factor at those locations. As a result, the opposite shoreline atmospheric gamma dilution factor becomes limiting due to its assumed full time occu?ancy since physical constraints (areas over water) do not exist, and there is no control on occupancy avai.able. It should be noted that controlling the maximum dose rate to 500 mrem per year at a location on the opposite shoreline from the plant still ensures that the dose rate on the exposed mudflats during low tide will not exceed a value which would give rise to two mrem in one hour (10 CFR 20] even assuming continuous occupancy during the hour.
Incorporating the above int.o Equation A-1 and converting from annual release Q (Ci/yr) to maximum instantaneous release rate Q (uci-sec), and multiplying by the conversion constant 31.54 Ci-sec/uCi yr l yields the method to calculate total body dose rate from noble gases: l b = 1.06 E Q, DFB 4 (A-2) 4/98 66 b
Fr i l L i MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION, MANUAL APPENDIX A
; I The maximum long term gamma dilution factor [X/Q]Y for a Fuel Building Exhaust Vent release is 1 2.61E - 06 sec/m3
[ Correcting for the Fuel Bldg. vent [X/Q]Y yields: [ 6, = 2.61[, d, ors, for releases from the Fuel Building Exhaust
; Since the only noble gas applicable to Maine Yankee due to the , Kr-85, DFBx,.., = 1.61E-05 mrem - m /pCi - yr, Table B-1, 3
Ref. 5, the permanent equation plant shutdown is can be simplified
. to- l i
[ 6, = 4 . 2 0 E- 0 5 = d,, ,,, (4-l A) [ (for a fuel building release of Kr-85). ' l A.3 Skin Dose Rate From Noble Gases Method I is used to demonstrate compliance with Section 2.2.3.1.a, which limits skin dose rate
- from noble gases released to the atmosphere, for the peak noble gas release rate.
l Method I applies the methods of Equation 11 in Regulatory Guide 1.109 (Reference 2) as follows: (A-3) D,,,, = 1.11 S, 3.17K+04 [r/c]Y E 0, or[ + 3.17t+04 X/o E og ors, 1 i where: Dai, is the annual skin dose rate, in mrem /yr, 1.11 is the average ratio of tissue to air energy absorption coefficient: Sr is the attenuation factor that accounts for the dose reduction due to shielding provided by residential structures, but for all dose rate calculations is assumed to be equal to I (dimensionless);
'3.17E+04 is the number of pCi per Ci divided by the number of seconds per year;
[X/Q]Y is the effective long term average gamma dilution factor in sec/m3 ; Q3- is the annual release rate of radionuclide i, in Ci/yr, orYi is the amma air dose factor for a uniform semi-infinite cloud of radionuclide i, in mrad-m3 /pCi-yr, X/Q. is the long term average undepleted dilution factor in sec/m3 ; . and-4/98 67
MAINE YANKEE ATOMIC POWER COMPANY
- OFF-SITE DOSE CALCULATION MANUAL APPENDIX A . j DFS i is the beta skin dose factor for a semi-infinite cloud of radionuclide i, which includes the attenuation by the outer l
" dead" layer of the skin, in mrem-m'/pCi-yr (taken from
[ Reference 2, Table B-1). l
~[ For a release from the primary vent stack, the maximum effective five year av::: age gamma dilution factor [X/Q]7, is 1.06E-06 sec/m 2(see Table 7.1), and the maximum five year average 2
l undepleted dilution factor, X/Q, is 1.18E-06 sec/m (see Table 7.1). Incorporating these constants into Equation A-3 and converting from annual release Q (Ci/yr) to maximum instantaneous release rate 6 (uCi/sec) and multiplying by the conversion factor 31.54 Ci-sec/uCi-yr yields: Y b,,,g = 1.18 E , Dr + 1.18 E 6, DFS, a i i (A-4) Y
= E Q, (1.18 Dr + 1.18 DrS,].
A i [ A combined skin dose factor, Dr , may be defined: Dr = 1.18 Dr[ + 1.18 DFS,. Incorporating the combined skin dose factor, Dr ,into Equation A-4 yields the method to i calculan skin dose rate from noble gases: l l
< 1 -[ 6,,,, = E 6, Dry (For a primary vent stack release) i 4/98 L 68 l
L.
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX A [ The maximum long term gamma dilution factor [X/Q]7 for a Fuel Building Exhaust is 2.61E-06 Sec/m' [ and the maximum long term average undepleted dilution factor X/Q is 1.03E-05 sec/m2 . [ Correcting for the SFPI X/Q's in the skin dose rate equation, A-3, yields [ 5,,,,=2.9[,d,Dr,Y + 10.3 [, d, Drs, [ Since the only noble gas applicable to Maine Yankee is Kr-85, the equation can be simplified with 3 2 [ DFu.u v equal to 1.72E-05 (mrad-m /pCi-yr) and DFSo.n equal to 1.34E-03 (mrad-m /pCi-hr) as taken [ from Regulator Guide 1.109, Table B-1 (Ref. 2). This reduces the equation for Fuel Building exhaust [ skin dose rate in mrem /yr. from KR-85 to: [ d,,,, = d,,,,, (2. 9 (1. 72E - 05) + 10.3 (1.34E-03)) = 1. 3 9E - 02 d,, ,,, [ d,,,, = 1. 3 9E - 02
- d,, ,,, (4 - 2a)
A.4 Critical Orean Dose Rate From lodines and Particulates Method I is used to demonstrate compliance with Section 2.2.5, which limits the dose rate from Iodine-131, lodine-133, tritium, and radioactive materials in particulate form with half-lives greater than 8 days. The method to calculate the critical organ dose rate from radioactive iodines and particulates is derived from ODCM Equation 4-6 which limits the dose to the critical organ from radioactive iodines and particulates. D.,= E Qi DFG,,, (A-5) i where: D, is the dose to the critical organ from lodine-131, Iodine-133, tritium, and radioactive
. materials in particulate form with half-lives greater than 8 days, in mrem; Q, is the total activity of radionuclide i released via the plant stack during the period ofinterest, in Ci; and DFG,,, is the site specific critical organ dose factor for radionuclide i for a gaseous release, in mrem /Ci (see Table 4.2).
4/98 69
p MAINE YANKEE ATOMIC POWER COMPANY t OFF-SITE DOSE CALCULATION MANUAL-l . l APPENDIX A Applying the conversion factor,31.54 (Ci-sec/uCi-yr), to convert DFG;,, (mrem /Ci) to an organ dose rate factor ord g , (mrem-sec/uCi-yr) for use for iodines and particles and changing the shi l . factor (S,) from 0.7 to 1.0 for exposure from a contaminated ground plane yields a new critical organ dose rate factor om (see Table 4.2), and a dose rate equation in the same form as Equation A-5 above, where the activity release rate 6,is in uCi/sec. 5,, = E 0, Dm,'.. (A-6) L [ The dose rate to the critical organ, D'8co, in mrem /yr from tritium and radioactive materials in particulate [ form with half-lives greater than eight days released to the atmosphere from the Fuel Building exhaust is: [ 6" = [ 6" = orra'u , [ where:
' is the release rate of radionuclide i, released via the fuel building, in Ci/sec; and
[ [# "- is the site specific Critical Organ Dose Rate Factor for radionuclide i, for a gaseous release [ from the fuel building exhuast, in mrem-sec/ Ci-yr. (See Table 4.4) i l' i l 4/98 l 70
p MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX A . A.5 Gamma Air Dose Method I is used to demonstrate compliance with Section 2.2.4, which limits the gamma air dose [ L due to noble gases released in gaseous effluents via the primary vent stack to areas at and beyond the site ,
- - boundary. -
l t 1 Method I is derived from the methods of Equations B 4 and B-5 in Regulatory Guide 1.109 (Reference 2) which gives: of,,,,,;,,, =' 3.17t+04 (x/ol! E o, or[
- (A-7) i A \
where: D%,,,, is the gamma air dose, in mrad due to a finite cloud release; 3.17E+04 is the number of pCi per Ci divided by the number of seconds per year; [X/Q)' is the effective long-term average gamma dilution factor in sec/m' (see Appendix B for use of effective gamma atmospheric dilution factors); I Qi - is the total activity of noble gas i released via the plant stack during the period ofinterest, in Ci; and DFY i-is the gamma dose factor to air for noble gas i, in mrad-m2 /pCi-yr (taken from Reference 2). 3 Incorporating the maximum effective long-term average gamma dilution factor of 1.06E-06 sec/m (see Table 7.1) yields: DY,i, = 0.034 E Qi DFYi (4-4) i C For a Fuel Building Exhaust release the maximum effective long term average gamma dilution factor is [ 2.61E-06 sec/m .3 Incorporating this value into equation (A-7) yields: C [- DYg, = 0.083 E Qi DFYi (for a fuel building release)
-[
[ DFY for Kr-85 = 1.72E-05 mrad-m3 /pci-yr; therefore
'[-
[ DY ,S = 1.42E-06
- Qg,.n (for a fuel building Kr-85 release)
[
- [ The gamma air dose, Db in mrad from Kr-85 released from the fuel building exhaust is:
[ [ DY,i, = 1.42E-06
- Qx,.c (4-4a) 4/98 l
l L 71 l:_
MAINE YANKEE ATOMIC POWER COMPANY - OFF-SITE DOSE CALCULATION MANUAL 1 l APPENDIX A .
)
A.6 Beta Air Dose l Method I is used to demonstrate compliance with Section 2.2.4, which limits the beta air dose due !
--[ to noble gases released in gaseous effluents via the primary vent stack to areas at and beyond the site boundary.
Method I is derived from the methods of Equations B-4 and B-5 in Regulatory Guide 1.109 (Reference 2) which gives: o[,, = 3.17t+04 x/o E o, or[ (A-8) L where: o, is the beta air dose,in mrad; 3.17E+04 is the number of pCi per Ci divided by the number of seconds per year; X/Q 'is the long-term (5-year) average undepleted dilution factor, in sec/m); ) Qi is the total activity of noble gas i released via the plant stack during the period of interest, in Ci; , and is the beta dose factor to air for noble gas i, in mrad-m'/pCi-yr. orf 3 Incorporating the maximum long-term average undepleted dilution factor of 1.18E-06 sec/m (see Table 7.1) yields: o[,,=0.037Eo,or[' (4-5) i [ - For a Fuel Building Exhaust release the maximum effective long term average undepleted dilution factor j[ is 1.03E - 05'sec/m . 2Incorporating this value into equation (A-8) yields: 1 [ o[,, = 0.033 E o, or[ (for a fuel building release) L [ DF8 for Kr-85 = 1.95E-03 mrad-m'/pCi- yr; therfore a fuel building Kr-85 release the Beta dose in mrad [ can be expressed as: , [ o[,, = . 6. 37s - 04
- o,,.,, (4-Sa)
[ 4/98 72
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATIOff MANUAL APPENDIX A A.7 Dose from iodines and Particulates Method I is used to demonstrate compliance with Section 2.2.5, which limits the dose commitment to a member of the public from Iodine-131, Iodine-133, tritium, and radioactive materials in [ particulate form with half-lives greater than eight days in gaseous effluents released via the primary vent ( stack, the fuel building exhaust or auxiliary boiler stack to areas at and beyond the site boundary. For site boundaries adjacent to Back River, the off-site atmospheric dispersion parameters were determined (see Table 7.1) for locations on the opposite shore where there is a potential for exposure pathway's to exist on a continuous basis. The maximum of all off-site atmospheric dispersion parameters in any direction was selected in the determination of potential doses from iodines and particulates. The dose commitments to an individual from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than eight days released to the atmosphere via the plant stack are calculated using the methods of Equations C-2, C-4, and C-13 in Regulatory Guide 1.109 (Reference 2). Gaseous pathways assumed to contribute to individual doses at Maine Yankee are: extemal irradiation from radionuclides deposited on the ground surface, inhalation of radionuclides in air, and ingestion of atmospherically released radionuclides in food. The use of the methodology of Equations C-2, C-4, and C-13 for a one curie release of each radionuclide in gaseous effluents yielded the dose impact to the critical organ. Table 4.2 lists the resulting site specific critical organ dose factors for plant stack releases giving the number of millirem per curie released for each radionuclide. Since the dose factors of Table 4.2 represent a variety of critical organs, Method I conservatively calculates a critical organ dose consisting of a combination of critical organs of different age groups. Similarly, Table 4.3 list the site specific dose factors for releases via the [ auxiliary boiler resulting from the buming of contaminated waste oil. Table 4.4 lists the site specific [ dose factors for releases via the fuel building exhuast. Parameter values used to derive the critical organ dose factors for iodines and particulates are listed on Tables A-2 and A-3. 4/98 l 73
(; i l MAINE YANKEE ATOMIC POWER COMPANY ! OFF-SITE DOSE CALCULATION MANUAL l APPENDIX A Milk and meat animals are asstuned to be on pasture 50 percent of the time, consuming 100 l percent of their feed from pasture during that period. This assumption is conservative since most dairy operations use supplemental feeding of animals when on pasture or actually restrict animals to full time . silage feeding throughout the year. i 4 1 l l i I i l 4 l i l i l l 3/93 I 74 1 i
g MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX A TABLE A-7 Uuoe Factors for Various Gaseous Pathways at Maine Yankee . (From Reference 1, Table E-5*) AGE LEAFY GROUP VEG. VEG. MILK MEAT INHALA110N (KG/YR) (KG/YR) (1/YR) (KG/YR) (M'/YR) Adult 520.00 64.00 310.00 110.00 8,000.00 Teen 630.00 42.00 400.00 65.00 8,000.00 Child 520.00 26.00 330.00 41.00 3,700.00 Infant 0.00 0.00 330.00 0.00 1,400.00
- Regulatory Guide 1.109.
l l~ ! l l f 3/93 75 j l j
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. e s
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, a s9 y3 oo s otud a ed l df d s s e a .
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s s oea d y on os d ol n e a r p 1 1 l e a d a _ 1 1 1 1 1 b y 1 1 d on do an ht ru ( d s d o a or otd m e o - ht u o ht e eP hd e t ei MJs Meh0 l c MmF f - Mir n t l as i r6 r u . e r y F8 o7 oss0 F a1 o Fi d FP oh s e t o ) ) ) ) 2 3 4( N 1 ( ( (
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX B Meteorotony Long term (annual and five year) average dilution factors based on on-site meteorological data [ were computed for routine primary vent stack, auxiliary boiler stack, and fuel building exhaust vent, [ releases by the Yankee Atomic Electric Company's (YAEC) AEOLUS-2 (Reference 5) computer code. AEOLUS-2 is based, in part, on the straight-line airflow model as discussed in Regulatory Guide 1.111 (Reference 8). The following AEOLUS-2 features were used in the assessment of dilution factors for the Maine Yankee site:
- hourly meteorological data input (wind direction, wind speed, and vertical temperature difference)
- straight-line air flow model with Gaussian diffusion,
[- part-time ground level and part-time elevated releases (split-H model),*
- multi-energy sector-averaged finite cloud dilution factors for gamma dose calculations,
[- terrain height correction features,* [- plume r se (momentum),* depletian in transit, [- wind speed extrapolated as a function of release height.* dry deposition rates (based on Regulatory Guide 1.111). The following sector-average dilution and deposition factors were produced: non-depleted dilution factors for evaluating ground level concentrations of noble gases, tritium, carbon 14 and non-elemental iodines, depleted dilution factor for estimating ground level concentrations of elemental radiciodines and other particulates, effective gamma dilution factors for evaluating gamma dose rates from a sector-averaged finite cloud (multiple-energy undepleted source), and
- deposition factors for computing dry deposition ofelemental radiciodines and other particulates.
- [
- Primary Vent Stack Only 4/98 78
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX B Gamma dose rates are calculated throughout the ODCM using the finite cloud model presented in j
" Meteorology and Atomic Energy - 1968" (Reference 9, Section 7-5.2.5). That model is implemented l through the definition of an effective gamma atmospheric dispersion factor, [X/Q]v j (Reference 5, Section 6), and the replacement of X/Q in infinite cloud dose I l
equations by the [X/Q]Y. 1 l l l i . (. 3/93 79 l
l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX C Routine Reoorts
- 1. Annual Radiological Environrnental Ooeratino Reoort l The Annual Radiological Environmental Operating Reports covering the operation of the unit
[ during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and an analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period, and an assessment of the environmental impact of plant operation, if any. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50. The reports shall also include the results of the land use censuses required by Section l 2.4.4 of the ODCM. The Annual Radiological Environmental Operating Reports shall include summarized and tabulated l results of radiological environmental samples taken during the report period pursuant to the tables and figures in the ODCM. In the event that some results are not available for inclusion with the i report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a summary description of the radiological , environmental monitoring program including a map of all sampling locations keyed to a table giving distances and directions from the reactor; and a discussion of all analyses in which the LLD required by Table 2.4 of the ODCM was not achievable.
- 2. Annual Radioactive Effluent Release Reoort
[ The Annual Radioactive Effluent Release Report covering the activities of the unit during the [ previous year shall be submitted prior to May 1 of each year in accordance with 10CFR50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents released from the unit summarized on a quarterly basis. The report shall also include a summary of the solid waste released from the unit summarized on a semiannual basis. The material provided ahall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance l with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50. 10/98 1 80 l
MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped off-site during the report period:
- a. Container volume.
- b. Total curie quantity (specify whether determined by measurement or estimate).
- c. Principal radionuclides (specify whether determined by measurement or estimate).
- d. Source waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms). 1
- c. Type of container (e.g., LSA, Type A, Type B, Large Quantity).
- f. ' Solidification agent or absorbent (e.g., cement, asphalt, "Dow"). s The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site boundary of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Radioactive Effluent Release Reports shall include a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Section 2.4.4 of the ODCM. The Radioactive Effluent Release Report shall include changes to the ODCM in the form of a [ complete, legible copy of the entire ODCM in accordance with Technical Specification 5.6.2.
- 3. Estimated Dose and Meteorological Summary Reoort A report of the estimated maximum potential dose to the members of the public from radioactive effluent releases for the previous calendar year shall be submitted within 120 days after January 1 of each year. The assessment of the radiation doses shall be performed in accordance with the Off-Site Dose Calculation Manual (ODCM).
An annual summary report of the hourly meteorological data collected over the previous calendar year shall either be submitted within 120 days after January 1 of each year or be retained in a file that shall be provided to the NRC upon request. This annual summary may be either in the form or an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured) or in the form ofjoint frequency distributions of wind speed, wind direction, and atmospheric stability. 10/98 81
.'. 1 i
i l t i~ I MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATIDN MANUAL l APPENDN D SITE BOUNDARY s., i
- . :: . .::. ;;;; 3.e . e.:
.- .-~
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2/96 t l 82
MAINE YANKEE ATOMIC POWER COMP.OlY OFF-SITE DOSE CALCULATION MANUAL REFERENCES
- 1. Title 10, Code of Federal Regulations. The Office of the Federal Register, National Archives and .
Records Admiai*ation.
- 2. Regulatory Guide 1.109, " Calculation of Annual Doses to Man From Routine Releases of Reactor
. Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I", U.S. Nuclear Regulatory Commission, Revision 1, October 1977.
- 3. Intemational Commission on Radiological Protection (ICRP) Publication 2. Oxford: Pergammon.
- 4. Title 40, Code of Federal Regulations. The Office of the Federal Register, National Archives and Records Administration.
- 5. Hamawi, J.N., "AEOLUS Technical Description", Entech Engineering, Inc., Document No.
P100-R13-A, YAEC - Revised Sofhvare Release MOD 05, dated March 1992.
- 6. " Supplemental Information for the Purposes of Evaluation of 10 CFR 50, Appendix I", Maine Yankee Atomic Power Company, including Amendments 1 and 2, October 1976.
- 7. NUREG -0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", U.S. Nuclear Regulatory Commission.
- 8. 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, March 1976.
- 9. Slade, D. H., " Meteorology and Atomic Energy - 1968," USAEC, July 1968.
[ 10. MYC-2044, " Proposed ODCM Change for No Service Water Flow Conditions," Duke Engineering [- and Services, July 1998. 10/98 83 E}}