ML20056G750

Semiannual Radioactive Effluent Release Rept Jan-June 1993

ML20056G750
Person / Time
Site:	Maine Yankee
Issue date:	06/30/1993
From:	Hebert J Maine Yankee
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
JRH-93-177, MN-93-80, NUDOCS 9309070120
Download: ML20056G750 (110)

Text

MaineYankee pgTxhil;dKfhTciinstucttu2 i

EDISON DRIVE

• AUGUSTA. MAINE 04330 + (207) 622-4868 i

August 31, 1993 MN-93-80 JRH-93-177 UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk Washington, DC 20555

Reference:

(a)

License No. DPR-36 (Docket No. 50-309)

Subject:

Semiannual Radioactive Effluent Release Report Gentlemen:

Enclosed, in accordance with the requirements of Maine Yankee Technical Specification 5.9.1.6, is the Maine Yankee Semiannual Effluent Release Report for the period January to June 1993.

Please contact John Arnold if you have questions or comments.

Very truly yours, James R. Hebert, Manager Licensing & Engineering Support Department JHA/ jag Enclosures c:

Mr. Charles S. Marschall Mr. E. H. Trottier Mr. Patrick J. Dostie Mr. Thomas T. Martin Mr. Donald Hoxie - DHE Mr. Allan Prysunka - DEP - Oil & Hazardous Waste American Nuclear Insurers 1

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MAINE YANKEE ATOMIC POWER COMPANY SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT j

1 JANUARY - JUNE 1993

1.0 INTRODUCTION

i Tables 1 and 2 summarize the quantity of radioactive gaseous and liquid effluents, respectively, for the first and second quarters of 1993. Table 3 summarizes the solid waste shipped off-site for burial or disposal during the first half of 1993. Table 4 contains supplementary information.

i Appendices A throuah D indicate the status of reportable items per the requirements of ODCM sections 2.1.5, 2.2.6, 2.3.3, 2.3.4, 2.5 and Appendix C.

Changes to the ODCM issued during this period are summarized and included in Attachment I.

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l TABLE 1A Maine Yankee Atomic Power Station Effluent and Waste Disposal Semiannual Report First and Second Quarters. 1993 Gaseous Effluents - Summation of All Releases Unit Quarter Quarter Est. Total 1st 2nd Error. %

A. Fission and Activation Gases

1. Total release Ci 2.28 E+00 5.35 E+00 2.50 E+01

2. Averaae release rate for period UCi/sec 2.90 E-01 6.80 E-01

3. Percent of reaulatory limit 8.50 E-04 1.82 E-03 B. Iodines

1. Total Iodine-131 Ci 1.20 E-05 1.44 E-05 2.50 E+bl

2. Averaae release rate for period uti/sec 1.53 E-06 1.83 E-06

3. Percent of reaulatory limit 6.69 E-06 8.03 E-06 C. Particulates

1. Particulates with T-1/2 > 8 days Ci 1.08 E-06 9.80 E-06 2.50 E+01

2. Averaae release rate for period uCi/sec 1.37 E-07 1.25 E-06

3. Percent of reaulatory limit 6.02 E-07 5.47 E-06

4. Gross alpha radioactivity Ci 5.59 E-07 2.65 E-07 D. Tritium

1. Total release Ci 7.21 E-01 6.43 E-01 2.50 E+01

2. Averaae release rate for period uti/sec 9.17 E-02 8.18 E-02

3. Percent of reaulatory limit 8.26 E-04 7.37 E-04 ON/D = Not Detected l

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I.8BLE_lB Maine Yankee Atomic Power Station Effluent and Waste Disposal Semiannual Report Firth and Second Quarters.1993 Gaseous Effluents - Elevated Release Continuous Mode Batch Mode Nuclides Released Unit Quarter Quarter Quarter Quarter 1st 2nd 1st 2nd

1. Fission Gases Krypton-85 Ci N/D*

N/D*

6.98 E-02

_1.19 f+00 Krypton-85m Ci 6.41 E-04 3.50 E-03 N/D*

N/D*

Krvoton-87 Ci 1.02 E-03 6.16 E-03 N/D*

N/D*

Krypton-88 Ci 1.31 E-04 3.30 E-03 N/D*

N/D*

Xenon-133 Ci 1.88 E400 3.80 E+00 8.83 E-03 9.58 E-02 Xenon-135 Ci 1.80 E-01 1.53 E-01 1.56 E-05 2.67 E-05 Xenon-135m Ci 1.31 E-02 5.31 E-02 N/D*

N/D*

Xenon-138 Ci 3.79 E-03 2.11 E-02 N/D*

N/D*

Xenon-133m Ci 1.28 E-01 2.95 E-05 N/D*

1.45 E-06 Arcon-41 Ci 2.00 E-03 1.09 E-02 8.04 E-06 2.62 E-06 Xenon-131m Ci N/D*

N/D*

N/D*

9.47 E-03 Unidentified Ci N/D*

N/D*

N/D*

N/D*

Total for period Ci 2.21 E+00 4.05 E+00 7.87 E-02 1.30 E+00

2. Iodines lodine-131 Ci 1.20 E-05 1.44 E-05 1.70 E-09 3.75 E-08 Iodine-133 Ci 4.63 E-05 4.23 E-05 3.10 E-08 6.37 E-08 lodine-135 Ci N/D*

N/D*

6.05 E-09 1.52 E-08 Total for period Ci 5.83 E-05 5.67 E-05 3.88 E-08 1.16 E-07

3. Particulates Strontium-89 Ci N/D*

N/D*

N/D*

N/D*

Strontium-90 Ci N/D*

N/D*

N/D*

N/D*

Cesium-134 Ci N/D*

N/D*

N/D*

N/D*

Cesium-137 Ci 1.08 E-06 9.80 E-06 N/D*

N/D*

Barium-Lanthanum-140 Ci N/D*

N/D*

N/D*

N/D*

Zinc-65 Ci N/D*

N/D*

N/D*

N/D*

Cobalt-58 Ci N/D*

N/D*

N/D*

N/D*

Cobalt-60 Ci N/D*

N/D*

N/D*

N/D*

Others-Ci N/D*

N/D*

N/D*

N/D*

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i TABLE IC Maine Yankee Atomic Power Station Effluent and Waste Disposal Semiannual Report First and Second Duarters 1993 Gaseous Effluents - Ground level Releases There were no routine measured ground level continuous or batch mode gaseous releases during the first and second quarters of 1993.

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TABLE 2A Maine Yankee Atomic Power Station Effluent and Waste Disposal Semiannual Report j

First and Second Quarters. 1993 Liouid Effluents - Summation of All Releases j

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I Unit Quarter Quarter Est. Total Ist 2nd Error. %

A. Fission and Activation Products l

1. Total release (not including f

tritium. aases. alpha)

Ci 5.47 E-03 1.65 E-02 1.50 E+01 r

2. Average diluted concentration durina period uCi/ml 2.68 E-ll 7.96 E-11

3. Percent of applicable limit 9.49 E-04 4.24 E-03 i

B. Tritium

1. Total release Ci 1.46 E+02 9.08 E+01 1.50 E+01

2. Average diluted concentration durina period uCi/ml 7.16 E-07 4.39 E-07 i

3. Percent of acolicable limit 7.16 E-02 4.39E E-02 l

C. Dissolved and Entrained Gases

1. Total release Ci 9.08 E-03 1.82 E-02 1.50 E+01

2. Average diluted concentration durina period uCi/ml 4.55 E-11 8.80 E-Il

3. Percent of applicable limit 2.22 E-05 4.41 E-05 j

D. Gross Alpha Radioactivity

1. Total release Ci 7.41 E-07 1.75 E-06 1.50 E+01

2. Average diluted concentration durina period uCi/ml 3.63 E-15 8.45 E-15 E. Volume of waste released (prior to dilution) liters 2.84 E+07 2.67 E+07 1.00 E+01 l

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durina period liters 2.04 E+11 2.07 E+11 1.00 E+01 cN/D = Not Detected i

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TABLE 2B Maine Yankee Atomic Power Station Effluent and Waste Disposal Semiannual Report First and Second Quarters. 1993 Liouid Effluents Continuous Mode Batch Mode Nuclides Released Unit Quarter Quarter Quarter Quarter 1st 2nd 1st 2nd Strontium-89 Ci N/D*

N/D*

N/D*

N/D*

Strontium-90 Ci N/D*

N/D*

N/D*

N/D*

Cesium-134 Ci N/D*

N/D*

1.29 E-04 7.04 E-04 Cesium-137 Ci 2.00 E-06 N/D*

8.51 E-04 3.73 E-03 Iodine-131 Ci N/D*

5.74 E-05 7.22 E-04 2.70 E-03_

Cobalt-58 Ci 2.08 E-05 N/D*

3.21 E-04 4.32 E-04 Cobalt-60 Ci 3.08 E-06 6.52 E-06 1.01 E-04 2.96 E-03 Iron-59 Ci N/D*

N/D*

N/D*

N/D*

Zinc-65 Ci N/D*

N/D*

N/D*

N/D*

Manaanese-54 Ci N/D*

N/D*

N/D*

1.36 E-05 Chromium-51 Ci N/D*

N/D*

3.58 E-04 1.33 E-04 Zirconium-Niobium-95 Ci N/D*

N/D*

3.70 E-06 N/D*

Molybdenum-99 Ci N/D*

N/D*

N/D*

N/D*

Technetium-99m Ci N/D*

N/D*

N/D*

6.24 E-06 Barium-Lanthanum-140 Ci N/D*

N/D*

6.27 E-05 3.78 E-05 Cerium-141 Ci N/D*

N/D*

N/D*

1.57 E-06 Others-Iron-55 Ci N/D*

N/D*

1.93 E-03 2.28 E-03 Iodine-133 Ci N/D*

N/D*

1.04 E-04 2.36 E-04 Iodine-135 Ci N/D*

N/D*

3.12 E-05 N/D*

Antimony-124 Ci N/D*

N/D*

3.31 E-05 N/D*

Antimony-125 Ci N/D*

N/D*

1.18 E-04 9.00 E-04 Silver-110M Ci N/D*

N/D*

6.54 E-04 2.12 E-03 Cerium-144 Ci N/D*

1.22 E-05 4.58 E-06 1.63 E-04 Cadmium-109 Ci N/D*

N/D*

2.57 E-05 N/D*

Unidentified Ci N/D*

N/D*

N/D*

N/D*

Total for period (above)(1) Ci 2.59 E-05 7.61 E-05 5.45 E-03 1.64 E~02 Xenon-133 Ci 1.88 E-05 N/D*

8.56 E-03 1.76 E-02 Xenon-135 Ci 5.55 E-06 N/D*

4.45 E-04 4.64 E-04 1

Xenon-133m Ci N/D*

N/D*

4.56 E-05 1.81 E-04 l

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TABLE 3 Maine Yankee Atomic Power Station Effluent and Waste Disposal Semiannual Report First Half.1993 j

Solid Waste and Ir, adiated Fuel Shioments A. Solid Waste Shipped Off-Site for Burial or Disposal (Not Irradiated Fuel) l E

l None shipped.

l B. Irradiated Fuel Shipments (Disposition): None shipped.

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TABLE 4 Supplemental Information i

1.

Reaulatory Limits 1

Effluent Concentration l

a.

Fission and activation gases:

10CFR20; Appendix B, Table 2, Column 1 i

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Iodines:

10CFR20; Appendix B, Table 2, Column 1 c.

Particulates, (with half lives greater than 8 days) 10CFR20; Appendix B, T ale 2, Column 1 d.

Liquid effluents:

10CFR20; Appendix B, Table 2, Column 2 i

e.

Total noble gas concentration 2E-04 uCi/ml 2.

Averaae Enerov - Not Apolicable 3.

Measurements and Aporoximations of Radioactivity i

a.

Fission and Activation Cases i

Continuous Discharge - Vent stack samples are analyzed monthly. Activity levels determined are assumed constant for the surveillance interval.

The continuous vent stack monitor reading is used as a basis for increasing periodic sample frequency.

Batch Discharges - Direct measurements of the waste gas hold-up drums are taken prior to discharge. Containment vents and purges are analyzed by direct measurement of the containment atmosphere at periodic intervals during discharge.

b.

Iodines Primary vent stack iodine totals are taken from a minimum of weekly measurements i

of an in-line charcoal filter.

c.

Particulates Primary vent stack particulate totals are taken from a minimum of weekly measurements of an in-line particulate filter.

d.

Liouid Effluents Samples of secondary systems' liquid effluents are analyzed weekly for gross beta-gamma, alpha, tritium, dissolved gases, and gamma emitting isotopes.

Each batch release is analyzed for gross beta-gamma, alpha, tritium, dissolved gases, and gamma emitting isotopes prior to discharge.

Composite samples are made of secondary and primary system liquid effluents for a quarterly analysis of Strontium-90 and Strontium-89.

Primary system liquid effluents are also analyzed quarterly for Iron-55.

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TABLE 4 t

(Continued) 4.

Batch Releases a.

Liouids 1.

Number of batch releases: 68 2.

Total time period for batch releases:

134 hours0.00155 days <br />0.0372 hours <br />2.215608e-4 weeks <br />5.0987e-5 months <br />, 16 minutes I

3.

Maximum time period for a batch release:

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 38 minutes 4.

Average time period for batch releases:

I hour, 58 minutes i

5.

Minimum time period for a batch release:

50 minutes i

6.

Average stream flow during periods of release of effluents into a flowing stream: N/A j

7.

Maximum gross release concentration (uci/ml):

3.58 E-08

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Gaseous i

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Number of batch releases:

18 2.

Total time period for batch releases:

18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, 53 minutes 3.

Maximum time period for a batch release:

3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, 33 minutes 4.

Average time period for batch releases:

I hour, 3 minutes 5.

Minimum time period for a batch release:

15 minutes 6.

Maximum gross release rate (uCi/sec):

1.48 E+02 5.

Unplanned Releases f

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Liouid There were no abnormal liquid releases during the reporting period.

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Gaseous There were no abnormal gaseous releases during the reporting period.

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On-line Containment Purae i

On-line containment purge was not employed during this reporting period.

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APPENDIX A Radioactive Effluent Monitorino Instrumentation Recuirement:

Radioactive effluent monitoring instrumentation channels are required to be operable in accordance with ODCM Sections 2.3.3 and 2.3.4.

With less than the minimum number of channels operable and reasonable efforts to return

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the instrument (s) to operable status within 30 days being unsuccessful, ODCM Sections 2.3.3 and 2.3.4 requires an explanation for the delay in correcting the inoperability in the next Semiannual Effluent Release Report.

Response

On March 31, 1993, the Steam Generator #1 Blowdown Monitor Inlet Isolation Valve (BD-16) was found shut. An investigation concluded that the valve had been shut since March 3,1993, following maintenance on that valve.

The Maine Yankee Offsite Dose Calculation Manual (0DCM) requires that Steam Generator Blowdown Line effluent monitoring instrumentation be operable whenever steam generator blowdown is not being recycled.

If the instrumentation is not operable, remedial action requires grab sampling at 4

least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, with analysis for gross gamma activity at a limit 4

of detection of at least 10 uci/ml. With the valve shut, the Steam Generator #1 Blowdown Monitor (RM-2601) was isolated and, therefore, not operable from March 3 to March 31, 1993.

During that period, blowdown was discharged overboard and not recycled on five occasions:

From 0720 to 0944 on March 5, 1993 From 0540 to 1235 on March 13, 1993 From 0818 on March 15 to 1530 on March 16, 1993 From 0910 to 1230 on March 19, 1993 From 0115 on March 22 to 1020 on March 26, 1993 The weekly Steam Generator blowdown surveillance sampling and analysis 4

puformed on March 16, 1993, and March 23, 1993, satisfied the remedial action at the required sensitivity on those dates. Although the Steam Generator blowdown is routinely monitored by a daily gross gamma 4

surveillance, the lower limit of detection was only 1.44 x 10 uci/ml, and, i

therefore, the ODCM remedial action for Steam Generator blowdown effluent monitoring was not satisfied for the discharges which occurred on March 5, March 13, March 15, March 19, March 22, March 24, and March 26, 1993.

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APPENDIX B l

Liauid Radwaste Treatment System Reauirement:

With radioactive liquid waste being discharged without treatment with estimated doses in excess of the limits in ODCM Section 2.1.5, a report must be submitted to the Commission in the Semiannual Effluent Release Report for the period.

Response

The requirements of ODCM Section 2.1.5 were met during this period and, therefore, no report is required.

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APPENDIX C Gaseous Radwaste Treatment System Reauirement:

With radioactive gaseous waste being discharged without treatment with doses in excess of the limits in ODCM Section 2.2.6, a report must be submitted to the Commission in the Semiannual Effluent Release Report for the period.

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Response

The requirements of ODCM Section 2.2.6 were met during this period and, therefore, no report is required.

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l APPENDIX D l

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Lower Limit of Detection for Radioloaical Analyses l

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l Recuirement:

ODCM Section 2.5 requires that when unusual circumstances result in LLD's I

higher than required, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report.

Response

All samples were counted in such a manner as to satisfy the specified a priori lower limits of detection.

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ATTACHMENT I Chanaes to the Off-Site Calculation Manual l

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Yankee Nuclear Services Division (Bolton, Mass) (YNSD) (REG 309/92) provided a) new long-term average atmospheric /dcposition factors for both elevated and ground level release conditions, b) new dispersion factors to incorporate the stack exit velocity and site-specific recirculation factors, c) revised dose factors to incorporate the new atmospheric dispersion factors.

Meteorological data were updated using the site-specific data for the period 1986 through 1990. Dilution factors were calculated using the latest version of the AE0LUS f

Program.

One page (71) was deleted on YNSD advice because the information is not necessary for ODCM calculations and may vary depending on the results of the Annual Land Use Census.

2.

The revisions to 10;FR20 require the use of ECLs (Effluent Concentration Limits) rather than the old term "MPC".

The ODCM revision changed MPC to ECL.

3.

The Technical Support Department Manager was listed as the responsible manager since i

the REMP Program became the responsibility of Tech Support on 4/1/93.

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4.

The page summary was appropriately revised.

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r MAINE YANKEE i

OFF-SITE DOSE CALCULATION MANUAL' r

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t APPROVEDi S. E. Nichols, Manager i

Technical Support Department-

'i APPROVED:

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R. W. Blackmore, Plant Manager i

s DATE:

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ODOI PAGE CHANGE SDDIARY CHANGE NO. 2 DATE: 3/23/93 PAGE DATE PAGE DATE PAGE DATE PAGE DATE Cover 3/93 22 1/92 52 1/92 82 3/93 i

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL l

I Yankee Atomic Electric Company Nuclear Services Division 580 Main Street Bolton, MA 01740-1398 i

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l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company (" Yankee").

The use of information contained in this document by anyone other than Yankee, or the t

Organization for which the document was prepared under contract, is not authorized and, with respect to any unauthorized use, neither Yankee nor its officers, directors, agents, or employees assume any obligation, responsibility, or liability or make any warranty or representation as to the accuracy or completeness of the material contained in this document.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL I

ABSTRACT The Maine Yankee Nuclear Power Station Off-Site Dose Calculation Manual (MY ODCM) contains the approved methods to estimate the doses and radionuclide concentrations occurring beyond the boundaries of the plant caused by normal plant operation. With initial approval by the U.S. Nuclear Regulatory Commission and the MYNPS Plant Management and approval of subsequent revisions by the Pl nt Management (as per the Technical Specifications), this ODCM is suitable ts :how compliance where referred to by the Plant Technical Specifications.

Suffi _.

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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 bound and the second, Method II, is an in-depth analysis which can provide more realistic estimates.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE OF CONTENTS S

Pace DISCLAIMER OF RESPONSIBILITY.......................

i ABSTRACT................................. if TABLE OF CONTENTS iii b

LIST OF FIGURES ix LIST OF TABLES..............................

x

1.0 INTRODUCTION

1 2.0 RELEASE OF RADI0 ACTIVE EFFLUENTS.....................

2 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 2.1.4 Liquid Effluents: Dose 3

2.1.5 Liquid Radwaste Treatment 4

j 2.2 Release of Gaseous Radioactive Waste 5

2.2.1 Applicability 5

2.2.2 Objective 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, Tritium, anj Radioactive Material in Particulate Form...

8 2.2.6 Gaseous Radwaste Treatment System 10 2.3 Radioactive Effluent Monitoring Systems..............

11 1

2.3.1 Applicability 11 2.3.2 Objective 11 2.3.3 Radioactive Liquid Effluent Instrumentation 11 2.3.4 Radioactive Gaseous Effluent Instrumentation........

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MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL TABLE OF CONTENTS (CONTINUED)

Paae 1

i 2.4 Radiological Environmental Surveillance Program..........

18 l

2.4.1 Applicability 18 j

2.4.2 Objective 18 i

2.4.3 Radiological Environmental Monitoring 18 2.4.4 Land Use Census 20 2.4.5 Interlaboratory Comparison Program.............

22

-i 2.5 Radioactive Effluent Monitoring..................

30 2.5.1 Applicability 30 2.5.2 Objective 30 2.5.3 Liquid Effluents:

Sampling and Analysis..........

30 i

2.5.4 Liquid Effluents:

Instrumentation.............

30 2.5.5 Gaseous Effluents: Sampling and Analysis 30

{

2.5.6 Gaseous Effluents:

Instrumentation 31 2.5.7 Basis 31

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3.0 LIQUID EFFLUENT DOSE CALCULATIONS 36 i

3.1 Liquid Effluent Dose to an Individual...............

36 3.1.1 Dose to the Total Body 36 3.1.2 Dose to the Critical Organ 37 4

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4 MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE OF CONTENTS (CONTINUED)

Pace i

4.0 GASEOUS EFFLUENT DOSE CALCULATIONS 39 i

4.1 Gaseous Effluent Dose Rate

....................39 I

4.1.1 Dose Rate to the Total Body From Noble Gases

......39 i

4.1.2 Dose Rate to the Skin From Noble Gases

.........40 4.1.3 Dose Rate to the Critical Organ From Radioiodines and l

Particulates

......................41 4.2 Gaseous Effluent Dose From Noble Gases

..............42 i

l 4.2.1 Gamma Air Dose 42 l

4.2.2 Beta Air Dose....................... 43 l

l 4.3 Gaseous Effluent Dose from Iodine-131, Iodine-133, Tritium, l

and Radioactive Material in Particulate Form...........

44 4.3.1 Dose to the Critical Organ

...............45 i

i 5.0 ENVIRONMENTAL MONITORING........................

50 I

i 6.0 MONITOR SETPOINTS 60 i

6.1 Liquid Effluent Monitor Setpoints.................

60 l

6.1.1 Allowable Concentrations of Radioactive Materials in Liquid Effluents 61 6.1.2 Monitor Response for Liquid Effluents 62 l

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l MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL

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i TABLE OF CONTENTS (CONTINUED) l Pace i

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6.2 Gaseous Effluent Monitor Setpoints 63 6.2.1 Allowable Concentrations of Radioactive Materials in Gaseous Effluents 64 6.2.2 Monitor Response for Gaseous Effluents...........

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7.0 METEOROLOGY 68 APPENDIX A BASIS FOR THE DOSE CALCULATION METHODS A.1 Li quid Effl uent Doses.......................

71 A.2 Total Body Dose Rate from Noble Gases...............

74

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A.3 Skin Dose Rate From Noble Gases..................

76 A.4 Critical Organ Dose Rate From Iodines and Particulates

......78 A.5 Gamma Air Dose

..........................79 A.6 Beta Air Dose.......................

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A.7 Dose from Iodines and Particulates 81 APPENDIX B METEOROLOGY.........................

86

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APPENDIX C ROUTINE REPORTS.......................

88 REFERENCES...............................

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL LIST OF FIGURES i

Number Title Pace t

5.1 Environmental Radiological Sampling Locations Within 1 Kilometer of Maine Yankee..................

54 i

t 5.2 Environmental Radiological Sampling Locations Within 12 Kiloneters of Maine Yankee 55 5.3 Environmental Radiological Sampling Locations Outside 12 Kilometers From Maine Yankee............

56 5.4 Direct Radiation Monitoring Locations Within 1 Kilometer of Maine Yankee 57 5.5 Direct Radiation Monitoring Locations Within 12 Kilometers of Maine Yankee 58 t

5.6 Direct Radiation Monitoring Locations Outside 12 Kilometers From Maine Yankee............

59 i

6.1 Maine Yankee Liquid Radwaste System..............

66 l

6.2 Maine Yankee Gaseous Radwaste System.............

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1 1

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i MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i

LIST OF TABLES t

Number Title Pace 2.1 Radioactive liquid Effluent Monitoring Instrumentation 14 2.2 Radioactive Gaseous Effluent Monitoring Instrumentation......

16 i

2.3 Radiological Environmental Surveillance Program..........

23 i

2.4 Detection Capabilities for Environmental Sample Analysis Lower Limits of Detection 25 2.5 Reporting Levels for Radioactivity Concentrations in Environmental Samples 29 l

2.6 Radioactive Liquid Waste Sampling and Analysis Program

......32 i

2.7 Radioactive Gaseous Waste Sampling and Analysis Program......

34 3.1 Maine Yankee Dose Factors for Liquid Releases...........

38 4.1 Maine Yankee Dose Factors for Noble Gas Releases

.........47 l

4.2 Maine Yankee Dose Factors for Iodine, Tritium, and Particulate Releases

.....................48 4.3 Maine Yankee Dose Factors for Iodine, Tritium, and Particulates Released Via the Auxiliary Boiler

........49 5.1 Radiological Environmental Monitoring Stations 51 7.1 Maine Yankee Maximum 5-Year Average Atmospheric Dilution Factors..............................

69 A-1 Usage Factors for Various Liquid Pathways at Maine Yankee.....

73 A-2 Usage Factors for Various Gaseous Pathways at Maine Yankee.....

83

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A-3 Environmental Parameters for Gaseous Effluents at Maine Yankee...........................

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MAINE YANV,EE ATOMIC POWER COMPANY i

0FF-SITE DOSE CALCULATION MANUAL

1.0 INTRODUCTION

j 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 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.

The first method being the most conservative, but simplest to use; the second r

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 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 obtained by i

application of the second method.

The second method, Method II, will calculate the dose to seven organs of four age groups for 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 l

will be used to assess doses for the Semiannual Radioactive _ Effluent Release l

Report.

l Liquid effluent dose calculation methods are presented in Section 3.

Gaseous effluent dose calculation methods in Section 4.

In both chapters relevant l

Technical Specifications are followed by the appropriate Method I dose equations.

l 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 l

of the dose calculation methods is described in Appendix A.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i

2.0 RELEASE OF RADI0 ACTIVE EFFLUENTS l

2.1 Release of Liouid Radioactive Effluents 2.1.1 Acolicability The requirements in this sectinn apply at all times to the release of all liquid waste discharged from the plant which may contain radioactive materials.

The provisions of Technical Specifications 3.0.A.2 and 3 do not apply to the limiting conditions for operation established in th1s section.

2.1.2 Objective The objective is to establish conditions for the release of liquid waste containing radioactive materials and to assure that all such releases are within the concentration limits specified in 10 CFR Part 20, and also assure that the releases from the site of radioactive materials in liquid wastes (above background) are kept "as low as is reasonably achievable" in accordance with 10 CFR Part 50, Appendix I.

2.1.3 Liouid Effluents:

Concentration 1.

The concentration of radioactive material in liquid effluents released frcm the site to unrestricted areas shall be limited to the concentrations specified in 10 CFR, Part 20, Appendix B, Table II, Column 2, for radionuclides other than noble gases, and

[

l x 10" 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 l

unrestric'ed areas (at the point of discharge into Back River; discharge from i

the submerged multiport diffuser) will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II.

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MAINE YANKEE ATOMIC POWER COMPANY l

OFF-SITE DOSE CALCULATION MANUAL This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within (1)Section II.A design objectives of Appendix I, 10 CFR Part 50, to a member

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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 MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP), Publication 2 (Reference 3).

2.1.4 Liauid Effluents:

Dose 1.

The dose or dose commitment 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 less than or equal to 10 mrem to any organ.

Remedial Action:

With the calculated dose from the release of radioactive I

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 the' corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

t Remedial Action:

With the calculated dose from the release of radioactive f

I materials in liquid effluents exceeding twice the above limits, calculations t

should be made including direct radiation contributions from significant plant sources to determine whether the limits of 40 CFR 190 (Reference 4) have been exceeded.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL If such is the case, prepare and submit to the Commission within 30 days pursuant to Maine Yankee Technical Specification 5.9, a Special Report.

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 Special 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 request is complete.

Basis:

These requirements are provided to implement the guidance of Sections II.A, 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 effluents i

l will be kept "as low as is reasonably achievable" as set forth in I

Section IV.A of Appendix 1.

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 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 l

conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated.

i The remedial action requiring calculations when releases exceed two times the i

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.

2.1.5 Licuid Radwaste Treatment I.

The Liquid Radwaste Treatment System shall be used in its designed modes of 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.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 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 Semiannual Effluent Release Report which includes the following information:

a.

Explination 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 eason for the inoperability; b.

Actions taken to restore the inoperable equipment back to operable status; and c.

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

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 radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36(a) and the design objective guidance given in Section II.D of Appendix I to 10 CFR Part 50.

The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents.

l 2.2 Release of Gaseous Radioactive Waste 1

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.

The provisions of Technical Specifications 3.0.A.2 and 3 do not apply to the limiting conditions for operation established in this section.

2.2.2 Objective 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 racioactive 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.

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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 l

released in ;v eous 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 i

body, and less than or equal to 3,000 mrem / year to the skin; and I

b.

For Iodine-131, Iodine-133, tritium, and radioactive materials in l

particulate form with half-lives greater than eight days to less I

than or equal to 1,500 mrem / year to any organ.

j Remedial Action: With the dose rates averaged over a period of one hour l

exceeding the above limits, without delay take action to decrease the release j

rate to comply with the limit.

l Basis:

These requirements are provided to ensure that the dose rate at any time at the site area boundary and beyond from gaseous effluents will be

{

within the annual dose limits of 10 CFR Part 20.

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 average concentrations exceeding the limits specified in Appendix B, Table II of 10

[

CFR Part 20.

For members of the public who may at times be within the site boundary area, the occupancy time 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 tctal 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-inf ant pathway to less than or equal to 1,500 mrem / year for the nearest real milk animal to the plant.

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MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL I

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2.2.4 Gaseous Effluents:

Dose From Noble Gases I.

The air dose due to noble gases released in gaseous effluents from the i

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, and less than or equal to 10 mrad for beta J

radiation; and b.

During any calendar year to less than or equal to 10 mrad for gamma radiation, and less than or equal to 20 mrad for beta radiation.

Remedial Action: With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission, a report within 30 days from the end of the quarter.

f 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 i

gases in gaseous effluents and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

r Basis: These requirements are provided to implement the guidance of i

Sections II.B, III.A, and IV.A of Appendix I, 10 CFR Part 50. The limiting 4

condition for operation implements the guides set forth in Section II.B of Appendix I.

7 4

This section provides the required operating flexibility, and, at the same time, assures that the releases of radioactive material in gaseous effluents 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.

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MAINE YANKEE ATOMIL POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 2.2.5 Gaseous Effluents:

Dose From Iodine-131. Iodine-133. Tritium, and Radioactive Material in Particulate Form l

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 I

b.

During any calendar year to less than or equal to 15 mrem to any l

organ.

c.

Less than 0.1% of the limits specified in 2.2.5.1.a and b as a result of burning contaminated oil.

Remedial Action: With the calculated dose from the release of Iodine-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 to the Commission a report within 30 days from the end of the quarter.

l The report shall identify the cause(s) for exceeding the limit (s) and define the corrective 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 from the release of radioactive materials in 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.

j If such is the case, prepare and submit to the Commission within 30 days pursuant to Maine Yankee Technical Specification 5.9, a Special Report.

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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 Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190.

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MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL j

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

Basis: 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 l

conditions for operation are the guides set forth in Section II.C of Appendix I.

The spe efication provides the required operating flexibility and at the same time assures that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods implement the guidance in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by l

i calculational procedures based on models and data such that the actual

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exposure of a member of the public through appropriate pathways is unlikely 4

to be substantially underestimated.

These equations also provide for determining the actual doses based upon the historical average atmospheric conditions.

l The release rate specifications for Iodine-131, Iodine-133, tritium, and i

radioactive material in 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.

f 2.

Deposition of radionuclides onto green leafy vegetation with subsequent consumption by man.

3.

Deposition onto grassy areas where milk animals and meat-producing animals graze with consumption of the milk and meat by man; and i

1 4.

Deposition on the ground with subsequent exposure to man.

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 l

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MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL 2.2.6 Gaseous Radwaste Treatment System 1.

The Gaseous Radwaste Treatment System and the Ventilation Exhaust

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Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the estimated gaseous t

effluent air doses due to gaseous effluent 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.

The Ventilation Exhaust Treatment System shall be used to reduce i

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 r

31 days.

Remedial Action: With gaseous waste being discharged without processing i

through 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 Semiannual 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; b.

Action (s) taken to restore any inoperable equipment to operable status; and c.

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

Basis: 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 50.36(a), General Design Criterion 50 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 i

System were specified as a suitable fraction of the guides sei forth in i

Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 2.3 Radioactive Effluent Monitorina Systems 2.3.1 Acolicability The requirements in this section apply at all times to Radioactive Effleent Monitoring Systems which perform a surveillance, protective, or contro'lling function on the release of radioactive effluents from the plant.

The provisions of Technical Specifications 3.0.A.2 and 3 do not apply to the limiting conditions for operation established in this section.

2.3.2 Ob.iective The objective is to assure the operability of the Radioactive Effluent Monitoring Systems to perform their design functions.

2.3.3 Radioactive Liouid Effluent Instrumentation 1.

The radioactive liquid effluent monitoring instrumentation channels shown in Table 2.1 shall be operable with their alarm / trip setpoints set i

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 l

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.

i Exert reasonable efforts to:

'i a.

Return the instrument (s) to operable status within 30 days;and b.

If unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report the reason for the delay in correcting the inoperability.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL Basis: The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid ef fluents.

The alarm / trip setpoints for these instruments are to ensure that the alarm / trip wili occur prior to exceeding the limits of 10 CFR Part 20.

The i

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 Radioactive 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 i

setpoints set to ensure that the limits in Section 2.2.3.1 are not exceeded during release of radioactive material via this pathway.

The alarm / trip setpoints of these channels shall be determined in accordance with the methodology in this ODCM.

Remedial Action:

With a radioactive gaseous process effluent monitoring instrumentation 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:

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 conservative.

Remedial Action: With less than the minimum number of radioactive effluent monitoring instrumentation channels operable, take action shown in Table 2.2.

I Exert reasonable efforts to:

a.

Return the instrument (s) to operable status within 30 days; and b.

If unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report the reason for the delay in correcting the inoperability.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL Basis:

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in j

gaseous effluents during actual or potential releases of gaseous effluents.

l The alarm / trip setpoints for these instruments are to ensure that the i

alarm / trip will occur prior to exceeding the limits of 10 CFR Part'20.

i The operability and use of this instrumentation is consistent with the l

requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

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MAINE YANKEE ATOMIC POWER CVMPANY OFF-SITE DOSE CALCULATION MANUAL 1

TABLE 2.1 Radioactive Liauid Effluent Monitorina Instrumentation Minimum Channels Remedial i

l Instrument Operable Action L

t 1.

Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release a.

Liquid Radwaste Effluent Line (1) 1 l

(Test Tanks) l l

2.

Gross Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination of Release t

a.

Service Water System Effluent Line (1) 3 b.

Steam Generator Blowdown Line (1)*

2 3.

Flow Rate Measurement Devices a.

Liquid Radwaste Effluent Line (1) 4 1

• Not required during steam generator blowdown recycle.

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.1 (Continued) t Table Notation ACTION 1 With the number of channels operable less than required by the minimum channels operable requirement, effluent 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.

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2.

At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valving.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 2 With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided grab samples are analyzed for gross radioactivity i

4 (beta or gamma) at a limit of detection of at least 10 uCi/ml:

1.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 uCi/ gram dose equivalent Iodine-131.

2.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 uCi/ gram dose equivalent Iodine-131.

ACTION 3 With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided that, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a 4

lower limit of detection of at least 10 uti/ml.

ACTION 4 With the number of channels operable less than required by the minimum channels operable requirements, effluent releases via this pathway may continue provided the flow rate is estimated at least once per eight hours during actual release.

Pump performance curves generated in situ may be used to estimate flow.

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OFF-SITE DOSE CALCULATION MANUAL i

TABLE 2.2 Radioactive Gaseous Effluent Monitorino Instrumentation j

Minimum l

Channels Instrument Operable Action l.

Waste Gas Holdup System

• W a.

Noble Gas Activity Monitor (1) 5 b.

Effluent System Flow Rate (1) 6 2.

Condenser Air Ejector 1

a.

Noble Gas Activity Monitor *

(1) 7 3.

Plant Stack (Vent Header System) a.

Noble Gas Activity Monitor (1) 7 b.

Iodine Sampler Cartridge **

(1) 8 1

c.

Particulate Sampler Filter **

(1) 8 d.

Effluent System clow Rate (1) 6 Measuring Device e.

Sampler Flow Rate Measuring Device (1) 6 (a) Monitor provides alarm and automatic isolation functior..

During power operations (Operating Condition 7).

Normal shutdown for filter changeout does not constitute inoperability.

1/92 L:\\DS\\15.EED 16

MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL TABLE 2.2 (Continued)

Table Notation ACTION 5 With the number of channels operable less than required by the minimum channels operable requirement, the contents of the drum (s) may be released to the environment provided that prior to initiating or continuing the release:

1.

At least two independent samples of the drum's contents are analyzed, and 2.

At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

i Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 6 With the number of channels operable less than required by the minimum l

channels operable requirement, effl>

,t releases via this pathway may continue provided the flow rate is t - imated at least once per eight hours.

ACTION 7 With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may i

continue provided grab samples are taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 8 With the number of channels operable less than required by the minimum thannels operable requirement, effluent releases via this pathway may continue provided samples are collected with auxiliary equipment.

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1/92 L:\\DSUS.EED 17

i MAINE YANKEE ATOMIC POWER COMPANY I

0FF-SITE DOSE CALCULATION MANUAL 2.4 Radioloaical Environmental Surveillance Proaram 2.4.1 Applicability This section applies at all times to radiological environmental surveillance and land use census.

2.4.2 Objective 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 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.

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2.4.3 Radioloaical Environmental Monitorino 1.

The Radiological Environmental Monitoring Program shall be conducted as specified in 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.

3.

With the level of radioactivity in an environmental sampling medium at a location specified in Table 2.3 exceeding a reporting level of Table 2.5 i

when averaged over any calendar quarter, prepare and submit to the i

l Commission with the next Semiannual Effluent Release Report, following receipt of the laboratory analyses, a report which includes an l

evaluation of any release conditions, environmental factors, or other l

aspects which caused the limits of Table 2.5 to be exceeded.

When more than one of the radionuclides in Table 2.5 are detected in the sampling medium, this report shall be submitted if:

concentration (1)

+ concentration (2)

+

...>1.0 reporting level (1) reporting level (2) 1 l

1/92 L;\\DS\\15.EED 18

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l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL Exception:

When radicnuclides 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 ta 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 plant effluents; however, in such an event, the condition shall be rep-orted and described in the Annual Radiological Environmentti Operating Report.

4.

With milk samples no long9r 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 Monitoring 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 of individuals resulting from the station operation.

This monitoring program thereby supplements the Radiological Effluent Monitoring 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 wh the concentration of plant effluents is expected to be negligible.

'the Zone II 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.

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l 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 oriori (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 posteriori LLD for a particular-measurement based upon the actual parameters for the sample in question.

2.4.4 Land Use Census 1.

An annual land use census within the distance of five miles shall be conducted to identify the location of the nearest milk animal, the 2

nearest residence, and the nearest garden of 50 m,

In lieu of a garden census, broad leaf vegetation of at least three i

different kinds may be sampled at or near the site boundary in two different sections.

l 2.

With a land use census identifying a location (s) which yields a i

calculated dose commitment (via the same exposure pathway) at least twice than at a location from which samples are currently being obtained in accordance with Section 2.4.3.1, identify the new locations in the l

next Annual Radiological Environmental Operating Report.

If permission from the owner to collect samples can be obtained and i

sufficient sample volume is available, then this new location shall be added to the Radiological Environmental Monitoring Program within 30 days.

ihe sampling location having the lowest calculated dose or dose commitment (via the same exposure pathway) may b3 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 October 1.

The results of the land use census shall be included in the Annual Radiological Environmental Operating Report.

Basis: This specification is provided to ensure that changes in the use of i

areas at and beyond the site boundary are identified and that modifications I

to the monitoring program are made if required by the results of this census, j

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t MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 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.

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HAINE YANKEE ATOMIC POWER COMPANY i

0FF-SITE DOSE CALCULATION MANUAL i

2.4.5 Interlaboratory Comparison Procram 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 exists.

If analyses are not performed as required above, a report shall be made in the next Annual Radiological Environmental Operating Report.

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 other environmental monitoring efforts, such as that for water quality.

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1/92 L:\\CS\\15.EED 22 l

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.3 Radioloaical Environmental Surveillance Proaram""'

Exposure Pathway Number of Sampling and Type and Frequency and/or Samnle Sample locations Collection Frecuency of AnalysisW l.

Airborne

a. Radioiodine and 5

Continuous operation of sampler with Radioiodine canister.

Analyze at Particulates sample collection as required'by least once per week for I-131.

dust loading but at least once per week.

Particulate sampler.

Analyze for gross beta radioactivity at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following filter change.

Perform gamma isotopic analysis on composite (by location) sample at least once per quarter.

2.

Direct Radiation 38 Quarterly.

Gamma dose quarterly.

3.

Waterborne

a. Surface (Estuary) 2 Composite

• sample collected over a Gamma isotopic analysis of each

. period of one month, monthly sample.

Tritium analysis of composite sample at least once per quarter.

b. Ground **

2 At least once per quarter.

Gamma isotopic and tritium analysis of each sample.

c. Sediment from shoreline 2

At least once per six months.

Gamma isotopic analysis of each sample, o Composite sample.shall be collected by collecting an aliquot at intervals not exceeding two hours.

Control station samples'may be grab samples rather than composite.

N Groundwater samples'shall be taken when this source is tapped-for drinking or irrigation purposes in areas where hydraulic gradient or recharge properties.are suitable for contamination.

1/92 L:\\DS\\i5.EED 23

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MAHUAL TABLE 2.3 (Continued)

Radioloqical Environmental Surveillance Programox2x33 Exposure Pathway Number of Sampling and Type and Frequency I

and/or Sample Sample Locations Collection Freauency of AnalysisW 4.

Ingestion a.

Milk

• 3 At least once per month.

Gamma isotopic and I-131 analysis of j

each sample.

b.

Fish and 2

One sample in season, or semiannually Gamma isotopic analysis on edible Invertebrates if not seasonal, of each of at least portions.

two commercially or recreationally important species.

c.

Food Products, 3

Monthly when available.

Gamma isotopic and I-131.

consisting of at least three types of broad leaf vegetation.

Performed only if milk sampling is not done.

(1). Specific sample locations for all media are specified in the Off-Site Dose Calculation Manual and reported in the Annual Radiological Environmental Operating Report.

(2) See Table 2.4 for maximum values for the lower limits of detection.

(3) Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, to seasonal unavailability or to malfunction of sampling equipst.

10 ihe latter occurs, every effort shall be made to complete corrective action prior to the end of the next sampling period.

(4) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to effluents from the plant.

O Food products (4.c) may be substituted for milk samples.

1 1/92 mosus.uo 24

MAINE YANKEE ATOMIC POWER COMPAl1Y OFF-SITE DOSE CALCULATION MANUAL TABLE 2.4 Detection Capabilities for Environmental Sample Analysis (a)(b)(d)

Lower limits of Detection Airborne Particulate food Water or Gas Fish and Invertebrates Milk Sediment Products M

Analysis faci /1)

(nCi/m )

(pCi/ka/ wet) pCi/1 Inci/ka/ dry)

(pCi/ka/ wet)

Gross Beta 4

.01 11 - 3 2000*

Mn-54 15 130 Fe-59 30 260 Co-58, Co-60 15 130 Zn-65 30 260 Zr-Nb-95 15*

l-131 1**

.07 1

60 Cs-134 15

.05 130 15 150 60 Cs-137 18

.06 150 18 180 80 Ba-La-140 15'#

15'#

0 If no drinking water pathway exists, a value of 3,000 pCi/1 may be used.

O If no drinking water pathway exists, a value of 15 pCi/1 may be used.

1/92 L:\\DS\\l5.EED 25

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i

TABLE 2.4 (Continued) l 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.

I For a particular measurement system (which may include radiochemical l

separation):

l l

4.66

• S 3 LLD =

E

• V

• 2.22

• Y

• Exp (-A

• At) where:

LLD is the "a priori" lower limit of detection as defined above (as picocuries per unit mass or volume).

l 4.66 is a constant derived from the K,p, and K values for the 95%

confidence level.

S is the standard deviation of the background counting rate or of the e

counting rate of a blank sample as appropriate (as counts per minute).

l E is the counting efficiency (as counts per disintegration).

V is the sample size (in units of mau or volume).

2.22 is the number of disintegration per minute per picocuries.

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 i

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1 MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL j

o j

TABLE 2.4 (Continued)

J Table Notation This equation results in an LLD in terms of picocuries.

For the purp7ses of Section 2.5 (Tables 2.6 and 2.7), where the required LLD is set forth in microcuries, the terms 2.22 in the denominator should be replaccd 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 samples (e.g., Potassium-40 in milk samples).

The analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.

Occasionally, background fluctuations, i

unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unavailable.

In such cases, the contributing factors will be identified and described in the l

Annual Radiological Environmental Operating Report.

b.

It should be recognized that the LLD is defined as an a oriori (before the l

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 posteriori LLD for a particular measurement based upon the actual parameters for the sample in question and appropriate decay correction parameters, such as decay while sampling and during analysis.

c.

Parent only.

I 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 l

to meet the specified LLD.

j 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.

1/92 L:\\DS\\l5.EED 27 l

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MAINE YANKEE ATOMIC POWER COMPANY

]

0FF-SITE DOSE CALCULATION MANUAL TABLE 2.4 (Continued)

Table Notation i

f.

The Ba-140 LLD and concentration can be determined by the analysis of its short-lived daughter product, La-140, subsequent to an eight-day period i

I following collection.

The calculation shall be predicated on the normal ingrowth equations for a parent-daughter situation and the assumption that l

any unsupported La-140 in the sample would have decayed to an insignificant amount (at least 3.6% of its original value). The ingrowth equations will l

assume that the supported La-140 activity at the time of collection is zero.

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l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.5 Reportina Levels for Radioactivity Concentrations in Environmental Samples I

Airborne Particulate Fish and Food Water or Gas Invertebrates Milk Products Analysis (oCi /1 )

(oCi /m')

(oCi/ka/ wet)

(oCi/1)

(0C1/11 H-3 20,000*

i Mn-54 1,000 30,000 I

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 l

6 l-131 2*

0.9 3

100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140' 200 300 t

If no drinking water pathway exists, a value of 30,000 pCi/l may be used.

Parent only.

If no drinking water pathway exists, a value of 20 pCi/l may be used.

1/92 L: IDS \\15.EED 29

t MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 2.5 Radioactive Effluent Monitorino 2.5.1 Applicability This section applies to monitoring radioactive effluents, both liquid and gaseous.

2.5.2 Ob.iective The objective of this section is to specify the nature and frequency of radioactive effluent monitoring requirements.

2.5.3 Liauid Effluents: Samplina and Analysis i

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 l

concentrations at the point of release are maintained within the limits t

of Section 2.1.3.1.

3.

Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in this ODCM at least once per 31 days.

2.5.4 Licuid Effluents:

Instrumentation l

Discharge of liquid radioactive effluents shall be continuously monitored with the alarm / trip setpoints of the moritor set in accordance with the methods outlined in the ODCM such that the requirements of Section 2.1.3 are met.

2.5.5 Gaseous Effluents:

Samplino and Analvsis 1.

Gaseous radioactive waste sampling and activity analysis shall be performed in accordance with Table 2.7.

2.

The cumulative doses due to gaseous effluents for the current calendar quarter and calendar year shall 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 31 days.

1/92 L:\\DS\\15.EED 30

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MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL 3.

Doses due to gaseous releases from the site to areas at or beyond the t

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 estimates for compliance with Section 2.2.6 are not required.

2.5.6 Gaseous Effluents:

Instrumentation l

Radioactive gaseous effluents shall be continuously monitored with the alarm / trip setpoints of the monitors set in accordance with the methods outlin'd i

in the ODCM such that the requirements of Section 2.2.3 will be met.

l 2.5.7 Basis The sampling analysis and instrumentation requirements set forth in this Epecification provide reasonable assurance that all significant radioactive releases will be monitored and th;t the effluents will not result in' exceeding the requirements of 10 CFR 20.

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i MAINE YANKEE ATOMIC POWER COMPANY l

OFF-SITE DOSE CALCULATION MANUAL i

TABLE 2.6 i

I Radioactive Liouid Waste Samolino and Analysis Procram I

Minimum Lower Limit of Sampling' Analysis Type of Activity Detection (LLD) i Liouid Release Type Frecuency Frecuency, Analysis (uCi /ml )*

A. Batch Waste PR PR Principa) Gamma 5 x 10'7 Release Tanks' Each Batch Each Batch Emitters 4

1-131 1 x 10 l

PR M

Dissolved and I x 10~5 One Batch /M Entrained Gases (Gamma Emitters)

PR M"

H-3 1 x 10 i

4 Each Batch Composite l

Gross Alpha 1 x 10-7 PR Q

Sr-89 Sr-90 5 x 10*4 d

Each Batch Composite

• Fe-55 I x 10 i

Emitters) Gamma 5 x 10~7 B. Plant D*

W Principa Continuous Grab Sample

• Composite' Releases

• j W8 I-131 1 x 10 4

Grab Sample **

4 M

M Dissolved and I x 10 Grab Sample Entrained Gases W8 M

H-3 1 x 104 Grab Sample ** Composite" Gross Alpha 1 x 10#

W8 4

Q Sr-89, Sr-90 5 x 10 i

Grab Sample ** Composite

• Turbine Building Sump

• • Steam Generator Blowdown Only 1/92 L:\\csus.tED 32

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MAINE YANKEE ATOMIC POWER COMPANY i

0FF-SITE DOSE CALCULATION MANUAL i

TABLE 2.6 (Continued)

Table Notation I

a.

The Lower Limit of Detection (LLD) is defined in Table Notation a of Table 2.4 of Section 2.4.

b.

A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid 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 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 thoroughly mixed in order for the composite sample to be representative of the effluent release.

d.

A batch release is the discharge of liquid wastes of a discrete volume.

Prior to sampling for analyses, each batch shall be isolated, and then f

thoroughly mixed to assure representative sampling.

l e.

A continuous release is the discharge of liquid wastes of 1 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,Cg-137,andCe-141.

Ce-144 shall also be measured, but with an LLD of 5 x 10.

This list does not mean that only these nuclides are to r

be considered. Other gamma peaks which are identifiable, together with the l

above nuclides, shall also be analyzed and reported in the Semiannual I

Radioactive Effluent Release Report.

Nuclides which are below the LLD for i

the analyses should not be reported as being present at the LLD level.

l g.

Weekly grab samples from the steam generator blowdown only during periods when blowdown is being discharged overboard, and not when blowdown is recycled to the main condenser.

l h.

Fe-55 shall be analyzed on quarterly composite samples commencing with July 1,1986.

If, after a period of two years, the results indicate that Fe-55 is likel this pathway, y to contribute 1% or less of the total dose attributable to the licensee may discontinue the analysis.

i.

Frequency notations:

PR = Prior to Release D = Daily W = Weekly i

M = Monthly Q = Quarterly 1/92 L:\\DS\\15.EED 33 l

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 2.7 Radioactive Gaseous Waste Samolina and Analysis Procram Minimum Lower Limit of Sampling '

Analysis Type of Activity Detection (LLD)

Gaseous Release Tvoe Freauency Freauency' Analysis (uCi/ml)*

A. Waste Gas PR PR Principal Gamma 1 x 10" Storage Tank Each Tank Each Tank Emitters

• Grab Sample B. Containment PR PR Principal Gaseous 1 x 10" 6

Purge Each Purge" Each Purge Gamma Emitters

• 4 Grab Sample H-3 1 x 10 C. Plant Vent I;"

M" Principal Gamma 1 x 10" j

Stack Grab Emitters

• l Continuous' W*

I-131 1 x 10 l

42 Charcoal Sample Continuous' W*

Principal Gamma 1 x 10'"

Emitters

• Particulate (I-131, Others)

Sample Continuous' M

Gross Alpha 1 x10'"

Composite Particulate Sample Continuous' Q

Sr-89, Sr-90 1 x 10'"

Composite I

Particulate i

Sample Continuous' Noble Gas Noble Gases 3.5 x 10'5 Monitor Gross Beta or 6

Gamma i

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1/92 L:\\DS\\15.EED 34 l

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0FF-SITE DOSE CALCULATION MANUAL l

TABLE 2.7 (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.

Sampling and analyses shall also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following shutdown, startup, or a thermal power change exceeding 15% of the rated thermal power in one hour unless:

(1) analysis shows that the dose equivalent I-131 concentration in primary coolant has not increased more than a factor of 3, and the resultant concentration is at least 1 x 10" uCi/ml; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

c.

Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least seven days following each shutdown, startup, or a thermal power change l

l exceeding 15% of rated thermal power in one hour, and analysis shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing the samples. This requirement to sample at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for seven days applies only if:

}

(1) analysis shows that the dose equivalent I-131 concentration in the primary coolant has increased more than a factor of 3 and the resultant concentration is at least 1 x 10" uti/ml; and (2) the noble gas monitor shows that effluent activity has increased more than a factor of 3.

When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10.

i d.

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.

e.

The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, l

Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, i

l Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions.

l This list does not mean that only these nuclides are to be detected and l

reported n-the Semiannual Radioactive Effluent Release Report.

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported in the Semiannual 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 i

detected." When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Semiannual Radioactive Effluent Release Report.

f.

Frequency notations are the same as in Table 2.6.

1/92

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 3.0 LIOUID EFFLUENT DOSE CALCULATIONS 3.1 Liouid 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 I)

The total body dose, D, in mrem for a liquid release is:

De - K E Q,DFl (3-1) e 1

where:

Q, is the total activity released for radionuclide i, in Ci (for strontiums use the most recent measurement available).

DFL.

is the site specific Total Body Dose Factor for radionuclide i, in mrem /Ci (see Table 3.1).

K is equal to 935/F ; where F, is the average (typically monthly average) o dilution flow of the Circulating W'ater System at the point of discharge from the multiport diffuser (in ft /sec).

1/92 L:\\DS\\15.EED 36

MAINE YANKEE ATOMIC POWER COMPANY l

OFF-SITE DOSE CALCULATION MANUAL 3.1.1.b Dore to the Total Body (Method II)

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, Rev.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 approach as described in the Bases Section A.1, are also applied to Method II assessments, except that doses calculated to the whole body from radioactive effluents are evaluated for each of the four age groups to determine the maximum whole body dose of an age-dependent individual via all existing exposure pathways. Table A-1 lists the usage factors for Method II calculations.

3.1.2.a Dose to the Critical Oroan (Method I)

The critical organ dose, D, in mrem for a liquid release is:

co D, = K Z Q DFL,c, (3-2) c 1

where:

Q is the total activity released for radionuclide i, in Ci (for strontiums use the most recent measurement available).

DFl,

is the site specific Critical Organ Dose Factor for radionuclide i, in ic mrem /Ci (see Table 3.1).

X is equal to 935/Fo; where Fa is the average (typically monthly average) dilution flow of the Circulating Water System at the point of discharge from the multiport diffuser (in ft /sec).

3.1.2.b Dose to the Critical Oroan (Method II)

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 approach 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.

1/92 mcsus.Eto 37

i MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL TABLE 3.1 Maine Yankee Dose Factors for Liauid Releases Total Body Critical Organ Dose Factor Dose Factor mrem /Ci mrem /Ci Nuclide DFle DFle 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 t

Mn-56 1.89E-06 4.09E-05 Fe-55 1.24E-02 7.53E-02 Fe-59 8.58E-02 6.54E-01 l

Co-58 2.21E-03 1.35E-02 Co-60 4.79E-02 7.80E-02 Zn-65 2.68E-01 5.38E-01 Sr-89 2.13E-04 7.45E-03 Sr-90 3.16E-02 1.29E-01 Zr-95 5.03E-04 1.73E-02 i

Mo-99 2.95E-05 2.62E-04 i

Tc-99m 4.06E-07 1.98E-06 Sb-124 1.34E-03 9.36E-03 I-131 2.07E-04 9.86E-02 I-132 2.54E-06 3.29E-06 1-133 2.46E-05 1.13E-02 I-135 7.12E-06 4.17E-04 Cs-134 2.79E-02 3.12E-02 l

Cs-137 2.92E-02 3.41E-02 l

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 i

Sb-125 4.81E-03 6.81E-03 Other 1.51E-01 3.40E+00 i

l l-I/92 L:\\DS\\15.EED 38

i i

l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 4.0 GASEOUS EFFLUENT DOSE CALCULATIONS 4.1 Gaseous Effluent Dose Rate Section 2.2.3.1 limits the dose rate (when averaged over 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) 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 I) l The total body dose rate, D, in mrem /yr from noble gases released via the e

plant stack is:

l t

1.06 E ~Q DFB, (4-1)

[D

=

3 7

l where:

t Q

is the release rate of noble gas i released via the plant stack, in uCi/sec; and

DFB, is the Total Body Dose Rate Factor for noble gas i, in mrem-m'/pCi-yr (see Table 4.1).

[ 1.06 is as defined in Section A.2 of Appendix A, in sec-pCi/m'-uCi 4.1.1.b Dose Rate to the Total Body From Noble Gases (Method II)

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 equation (8-8) taken i

from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases Section A.2, is also applied to a Method II assessment.

No credit for a shielding factor (Sp) associated with residential j

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 j

3/93 L:\\D5\\15.EED 39

1 MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL

)

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 I)

The skin dose rate, Dg,n, in mrem /yr from noble gases released via the plant stack is:

D

,n = 1 Q DF/

(4-2) a i

where:

is the release rate of noble gas i released via the plant stack, in Q

uCi/sec; and DFj is the Combined Skin Dose Rate Factor for noble gas i, in mrem-sec/pCi-yr (see Table 4.1).

4.1.2.b Dose Rate to the Skin From Noble Gases (Method II)

Method II consists of the model and input data (skin dose factors) in l

Regulatory Guide 1.109, Revision 1 (Reference 2), except where site-specific data l

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 l

Method I approach as described in the Bases Section A.3, is also applied to a Method II assessment.

No credit isr a shielding factor (S ) associated with p

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 i

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 i

with those occupancy factors included. The most restrictive location in any of the 16 sectors will be used in determining the dose rate.

1/92 L M SUS.EED 40 l

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MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i

4.1.3.a Dose Rate to the Critical Oroan From Radioicdines and Particulates (Method 1)

Ine dose rate to the critical organ, D, in mrem /yr from Iodine-131, m

Iodine-133, tritium, and radioactive materials in particulate form with half-lives greater than 8 days released via the plant stack is:

D

= I Qi D F G',,

(4-3) i where:

Q, is the release rate of radionuclide i released via the plant stack, in uCi/sec; and DFG'im is the site specific Critical Organ Dose Rate Factor for radionuclide i, in mrem-sec/ Ci-yr (see Table 4.2).

4.1.3.b Dose Rate to the Critical Oraan From Radiciodines and Particulates (Method II)

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 havo been identified in this ODCM (see Tables A-2 and A-3).

The critical organ dose rate will be determined based on the location (site boundary, nearest resident, or farm) of receptor pathways as identified in the most recent i

annual land use census, or by conservatively assuming the exi 'ence of all possible pathways (such as ground plane, inhalation, 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 organ dose rates will consider the four age groups independently, and take no credit for a shielding factor (Sp) associated with residential structures.

Site boundary locations adjacent to the river will be evaluated on the nearest opposite l

shoreline. Mud flats exposed at low tide will include an occupancy factor of l

0.037 for evaluation of doses at those locations.

1/92 L:\\DS\\15.EED 41

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MAINE YANKEE ATOMIC POWER COMPANY I

0FF-SITE DOSE CALCULATION MANUAL.

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 equal to 5 mrad for gamma radiation, and less than or equal to 10 mrad for beta radiation; and b.

During any calendar year:

less than or equal to 10 mrad for gamma radiation, and less than or equal to 20 mrad for beta radiation.

4.2.1.a Gamma Air Dose (Method Il The gamma air dose, Di,, in mrad from noble gases released via the plant stack is:

Di, = 0.034 E Q, DF[

(4-4) i

[

where:

Qi is the total activity of noble gas i released via the plant stack during the period of interest, in Ci; and DF( is the Gamma Dose Factor to air for noble gas i, in mrad-m*/pCi-yr (see Table l

4.1).

l

[ 0.034 is as defined in Section A.5 of Appendix A, in pCi-yr/Ci-m'.

4.2.1.b Gamma Air Dose (Method II)

Method II consists of the models, input data (dose factors) and assumptions in Regulatory Guide 1.109, Revision 1 (Reference 2), except where site-specific data or assumptions have been identified in this ODCM. The general equations (B-4 and B-5) taken from Regulatory Guide 1.109, and used in the derivation of the simplified Method I approach as described in the Bases Section A.5 are also applied to Method II assessments. Concurrent meteorology with the release period may be utilized for the gamma atmospheric dispersion 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 poteatial occupancy factor of 100% will be used to evaluate doses.

On-site areas with limited and controlled occupancy will be evaluated with those occupancy factors included.

3/93 l

L \\DS\\15.EED 42

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 4.2.2.a Beta Air Dose (Method I)

The beta air dose, D,,,, in trad from noble gases released via the plant stack is:

D, = 0.037 E Q, DF[

(4-5) i

[

where:

Q is the total activity of noble gas i released via the plant stack during the period of interest, in Ci; and DF 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'.

4.2.2.b Beta Air Dose (Method II)

Method II consists of the models, input data (dose factors) and assumptions 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 (B-4 and B-5) taken from Regulatory Guide 1.109, and used in the derivation of the 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 te utilized for the atmospheric dispersion factors (see Appendix B) for the reletse point from which recorded effluents have been discharged.

For sectors adjacent to the Back River, the nearest opposite shoreline with an assumed potertial 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.

3/93 L:\\C5\\15.EED 43

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 4.3 Gaseous Effluent Dose from Iodine-131. Iodine-133. Tritium. and Radioactive Material in Particulate Form Sections 2.2.5.1.a and 2.2.5.1.b limit the dose to a member of the public i

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 to the following:

during any calendar quarter:

less than or equal to 7.5 mrem to any a.

I orga 1; and b.

during any calendar year:

less than or equal to 15 mrem to any organ.

l 0DCM 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.

l l

l t

l i

1/92 L:\\Ds\\15.EED 44 1

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i MAINE YANKEE ATOMIC POWER COMPANY i

0FF-SITE DOSE CALCULATION MANUAL t

4.3.1.a Dose to the Critical Oroan (Method 11 l

l The dose to the critical organ, D, in mrem from Iodine-131, Iodine-133, co tritium, and radioactive materials in particulate form with half-lives greater than eight days released via the plant stack is:

D,={Q,DFG,,,

(4-6) c l

where:

Q, is the total activity of radionuclide i released via the plant stack during the period of interest, in Ci; and l

D FG;,,

is the site specific Critical Organ Dose Factor for radionuclide i for a gaseous release from the plant stack, in mrem /Ci (see Table 4.2).

(

The dose to the critical organ, D*,,, in mrem from Iodine-131, Iodine-133, l

tritium, and cadioactive materials in particulate form with half-lives greater than eight days released to the atmosphere from the auxiliary boiler due to the burning of contaminated waste oil is:

l Dll = 5 Ql"". DFW (4-7) u i

where:

q[""

is the total activity of radionuclide i released via the auxiliary boiler stack during the period of interest, in Ci; and DFW;c, is the site specific Crii.ical Organ Dose Factor for radionuclide i for gaseous release from the auxiliary boiler, in mrem /Ci (see Table 4.3).

1/92 L:\\DS\\15.EED 45 l

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i MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 1

i 4.3.1.b Dose to Critical Oroan (Method II)

Method 11 consists of the models, input data and assumptions in Appendix C c:

Regulatory Guide 1.109, Revision 1 (Reference 2), except where site-specific data l

or assumptions have been identified in this ODCM (see Tables A-2 and A-3).

The critical 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 assuming the existence of all possible pathways (such as ground plane, inhalation, 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 organ dose will consider the four age groups independently, and use a shielding factor (Sp) of 0.7 associated with residential structures. Mud flats exposed at low tide in areas where the Back River is adjacent to the site boundary will include an occupancy factor of 0.037 for evaluation of doses at those locations.

1 Only the inhalation and ground plane exposure pathways are included in the assessment of doses on the mudflats (for 10 CFR 50, Appendix I, and 40 CFR 190 l

considerabons).

l I

e 1/92 L:\\DS\\15.EED 46 I

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i MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL TABLE 4.1 Maine Yankee Dose Factors for Noble Gas Releases Total Body Combined Skin Gamma Air Beta Air Dose Rate Factor Dose Rate Factor Dose Factor Dose Factor (mrem-m /pCi-yr)

(mrem-sec/uci-yr) (mrad-m'/pCi-yr) (mrad-m'/pCi-yr)

Nuclide DFB:

DF!

DF[

DF[

[ 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 i

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 I

[ 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 1.11E-03

[ Xe-133m 2.15E-04 1.56E-03 3.27E-04 1.48E-03

[ 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

[ Xe-135 1.81E-03 4.46E-03 1.92E-03 2.46E-03

[ Xe-137 1.42E-03 1.62E-02 1.51E-03 1.27E-02

[ Xe-138 8.83E-03 1.57E-02 9.21E-03 4.75E-03

[ Ar-41 8.84E-03 1.41E-02 9.30E-03 3.28E-03 3/93

- tnosuS.EED 47 Effective: 05-24-93 j

v.,

g-

l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 4.2 Maine Yankee Dose Factors for Iodine. Tritium, and " rticulate Releases Critical Organ Critical Organ Dose Factor Dose Rate Factor (mrem /Ci)

(mrem-sec/uci-yr)

Nuclide DFG.ce D FG;,,

1 l

[ H-3 3.56E-04 1.12E-02

[ C-14 2.16E-01 6.81E+00 i

[ Cr-51 9.34E-03 2.95E-01

[ Mn-54 1.09E+00 3.44E+01

[ Fe-59 1.05E+00 3.31E+01

[ Co-58 5.55E-01 1.75E+01

[ Co-60 1.18E+01 3.72E+02 l

[ Zn-65 5.55E+00 1.75E+02

[ Sr-89 1.76E+01 5.55E+02

[ Sr-90 6.68E+02 2.11E+04

[ Sb-124 1.94E+00 6.12E+01 l

[ I-131 1.12E+02 3.53E+03

[ I-133 1.16E+00 3.66E+01

[ Cs-134 2.42E+01 7.63E+02

[ Cs-137 2.49E+01 7.85E+02 l

[ Ba-140 1.73E-01 5.46E+00 l

[ Ce-141 2.62E-01 8.26E+00

[ 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

3/93 L:\\CSUS.EED 48 Effective:

05-24-93 L

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MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL 1

TABLE 4.3 I

Maine Yankee Dose Factors for Iodine. Tritium, and Particulate Released Via the Auxiliary Boiler

• Critical Organ l

Dose Factor (mrem /Ci)

Nuclide DFW -

e

[

H-3 2.19E-03 i

[

C-14 1.77E+00

[

Cr-51 2.IlE-02 i

Mn-54 2.39E+00

[

Fe-59 2.35E+00

[

Co-58 1.24E+00 l

[

Co-60 2.59E+01

[

Zn-65 1.22E+01

[

Sr-89 3.86E+01

[

Sr-90 1.48E+03

[

Sb-124 4.34E+00

[

I-131 2.48E+02 I

[

I-133 3.28E+00

[

Cs-134 5.31E+01 i

[

Cs-137 5.45E+01

[

Ba-140 5.96E-01 l

[

Ce-141 6.00E-01 Ce-144 1.33E+01

)

[

Ag-110m 2.24E+01 1

[

Sb-125 4.25E+00

[

Other 9.93E+00 DFW, for use with the burning of contaminated 51ste oil.

i I

3/93 Effective: 05-24-93 L:\\DS\\15.EED 49 l

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l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i

5.0 ENVIRONMENTAL MONITORING The Radiological Environmental Monitoring Stations are listed in Table 5.1.

The locations of these stations with respect to the Maine Yankee facility are shown on the maps in Figures 5.1 through 5.6.

l j

l 1/92 L:\\DSUS.EED 50

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL TABLE 5.1 Radioloaical Environmental Monitorina Stations' Distance Direction Expusure Pathway Sample Location From the From the and/or Samole and Desianated Code" Plant (km)

Plant

1. AIRBORNE AP/CF-ll Montsweag Brook 2.7 NW (RADIOI0 DINE &

AP/CF-13 Bailey Farm (ESL) 0.6 NE PARTICULATE)

AP/CF-14 Mason Steam Station 4.8 NNE AP/CF-16 Westport Firehouse 1.8 S

AP-CF-29 Dresden Substation 19.8 N

2. DIRECT RADIATION l

TL-1 Old Ferry Rd.

1.0 N

TL-2 Old Ferry Rd.

0.8 NNE TL-3 Bailey House (ESL) 0.6 NE TL-4 Westport Island, Rt.144 1.2 ENE TL-5 MY Information Center 0.2 E

TL-6 Rt.144 and Greenleaf Rd.

0.9 E

TL-7 Westport Island, Rt. 144 0.8 ESE TL-8 MY Screenhouse 0.2 SE I

TL-9 Westport Island, Rt.144 0.9 SE TL-10 Bailey Point 0.4 SSE TL-Il Mason Station 4.8 NNE TL-12 Westport Firehouse 1.8 S

TL-13 Foxbird Island 0.4 SSW TL-14 Eaton Farm 0.8 SW l

TL-15 Eaton Farm 0.8 WSW l

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.

1.0 NNW i

TL-20 Bradford Rd., Wiscasset 6.5 N

TL-21 Federal St., Wiscasset 7.2 NNE TL-22 Cochran Rd., Edgecomb 8.3 NE TL-23 Middle Rd., Edgecomb 7.0 ENE TL-24 River Rd., Edgecomb 7.8 E

TL-25 River Rd. and Rt. 27 7.5 ESE I

TL-26 Rt. 27 and 7.8 SE l

Boothbay RR Museum 1/92 i

Lt\\DS\\15.EED 51 l

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i MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL i

TABLE 5.1 (Continued)

Radiolooical Environmental Monitorino Stations' i

Distance Direction i

1 Exposure Pathway Sample Location From the From the and/or Sample and Desianated Code" Plant (km)

Plant DIRECT RADIATICd TL-27 Barters Island 7.0 SSE (CONTINUED)

TL-28 Westport Island, Rt. 144 and i

East Shore Rd.

7.7 5

TL-29 Harrison's Trailer 6.3 SSW TL-30 Leeman Farm, Woolwich 7.6 SW i

TL-31 Barley Neck Rd., Woolwich 6.7 WSW TL-32 Baker Farm, Woolwich 7.2 W

TL-33 Rt. 127, Woolwich 7.3 WNW i

TL-34 Rt. 127, Woolwich 7.9 NW TL-35 Rt. 127, Dresden 9.1 NNW TL-36 Boothbay Harbor Fire Sta.

11.4 SSE TL-37 Bath Fire Station 10.4 WSW TL-38 Dresden Substation 19.8 N

3. WATERBORNE a.

Surface WE-12 Plant Outfall' O.3 SW (Estuary)

(Composite Sample) i WE-20 Kennebec River (Grab Sample) 9.8 SW b.

Groundwater WG-13 Bailey Farm (ESL) 0.6~

NE WG-24 Morse Well 9.8 W

c.

Sediment from SE-18 Foxbird Island 0.7 S

i Shoreline SE-16 Old Outfall Area 0.4 SW I

f 1/92 L:\\DS\\15.EED 52

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l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i

TABLE 5.1 (Continued) i i

Radiolooical Environmental Monitorina Stations

• Distance Direction Exposure Pathway Sample Location From the From the b

and/or Samole and Desionated Code Plant (km)

Plant 4.

INGESTION a.

Milk TM-15 Mitman Farm 5.8 S

TM-16 Baker Farm 7.2 W

TM-18 Chewonki Foundation 1.2 WSW TM-25 Hanson Farm 16.0 W

b.

Fish and FH/MU/CA/HA-11 Long Ledge Area 1.1 S

Invertebrates' MU/CA HA/24 Sheepscot River 11.2 S

f c.

Food Crop

• TV-1X Indicator Vegetation (to be determined)

TV-lX Indicator 4

(to be determined)

TV-2X to be determined i

Footnotes:

a Sample locations are shown on Figures 5.1 to 5.6.

b Station-lX's are indicator stations and Station 2-X'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 l

sampling of all four media types is not required during each sampling period.

Food crop sampling is not required while milk sampling is being done.

i e

1/92 j

L:\\D5\\15.EED 53

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i MAINE YANKEE ATOMIC POWER COMPANY l

OFF-SITE DOSE CALCULATION MANUAL i

FIGURE 5.1 l

Environmental Radiolocical Samolino locations Within 1 Kilometer of Maine Yankee t

N

-5G-13

,,ey,g

)

R$ -

l i

1-t i

1 T

g'

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l i l t i MAINE YANKEE ATOMIC POWER COMPANY i 0FF-SITE DOSE CALCULATION MANUAL 6.0 MONITOR SETPOINTS 6.1 Liouid Effluent Monitor Setooints Section 2.5.4 requires that the discharge of liquid radioactive effluents be l continuously monitored with the alarm / trip setpoint of the monitor set to ensure I that the requirements of Section 2.1.3 are met. i Section 2.1.3.1 requires that the concentration of radioactive material in liquid effluents relaased from the site to Back River be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for [ radionuclides other than noble gases and 1E-04 microcuries/ml total activity concentration for all dissolved or entrained noble gases. This section of ODCM describes the methodology that may be used to determine the setpoints of the liquid effluent monitors. Liquid effluent flow paths and release points, as well as the locations and identification numbers of the liquid effluent radiation monitors, are shown in Figure 6.1. l l 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 j response to this mixture in order to establish the physical settings on the monitors. 4 3/93 uusu5.EED 60

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 6.1.1 Allowable Concentrations of Radioactive Materials in Liouid Effluents Section 2.1.3.1 requires that the concentration of radioactive material in liquid effluents released from the site to Back River be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for [ radionuclides other than noble gases and IE-04 microcuries/ml total activity concentration for all dissolved or entrained noble gases. To ensure compliance with Section 2.1.3.1, the following method may be used for each release of liquid effluent. Prior to each release a grab sample will be analyzed to determine the activity concentration of each radionuclide. The [ ECL-fraction, R;, for each liquid effluent release point will be calculated by the relationship defined by Note 1 of 10 CFR Part 20, Appendix B: l l C# R=E (6-1) j i ECL, [ where: [ R, is the ECL-fraction for the release point j, dimensionless. C, is the diluted activity concentration of radionuclide i, in uCi/ml and is equal to undiluted concentration of radionuclide i times fj/F, where fj is the flow rate from the release point and F is the total dilution finw in gpm. [ ECL, is the effective concentration limit of radionuclide i as specified in 10 CFR Part 20, Appendix B, Table II, Column 2. in uCi/ml. [ The ECL-fractions for the various relecse points are then summed to yield the total MPC-fraction, R: R=Z R, (6-2) i [ The total ECL-fraction, R, at the discharge to the Back River must be less than or equal to one. R=s1 (6-3) 3/93 LnosuS.EED 61

i l l i MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL l 6.1.2 Monitor Response for ticuid Effluents The response of each liquid effluent monitor is established by combining the appropriate concentration, flow rate, dilution, principal gamma emitter, geometry, and detector efficiency. The radiation monitor alarm / trip setpoint for a test tank release is set such [ that the sum of the ECL ratios of the diluted nuclides is less than or equal to 0.6 at the discharge to the Back River. The setpoint is determined in the following manner: 1. The CPM of each undiluted nuclide is determined from the response graphs. If a nuclide other than Co-58, Co-60, I-131, I-133, Cs-134, and Cs-137 or unidentified nuclides contributes more than 10% of the total activity, use the conservative I-131 response curve to determine the CPM. (Unidentified nuclide activity is determined by subtracting [ identified activity from gross gamma activity.) i j 2. Set the alarm setpoint at background plus the calculated CPM from Step 1. I i [ 3. If the discharge ECL ratio is calculated to be less than 0.6, the alarm setpoint CPM may be increased by a factor of 0.6/ actual ratio. [ 4. The ECL ratio for a test tank discharge may be increased to less than 1 if it is determined that no other radioactivity is being released to the Back River. i I 3/93 L:\\DS\\15.EED 62

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 6.2 Gaseous Effluent Monitor Setooints Section 2.5.6 requires that radioactive gaseous effluents 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 limits of 10 CFR Part 20. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive reterial discharged ir ;neanus i effluents will not result in the exposure of a member of the public in unrestricted area, to annual average concentrations exceeding the limits specified i.n 10 CFR Part 20, Appendix B, Table II, Column 1. 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 effluent radiation detectors, are shown in Figure 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. t 1/92 L:\\DS\\15.EED 63

4 I MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL 6.2.1 Allowable Concentrations of Radioactive Materials in Gaseous Effluents l [ The DAC-fraction, R,, for each gaseous effluent release point is calculated by the relationship defined by Note 1 of 10 CFR Part 20, Appendix B: ) C'. R = [X/0] F g (6-5) i j i DAC. [ where: [R is the DAC-fraction for the release point j, dimensionless; 3 [X/Q] is the most conservative sector site boundary or off-site long- [ term average dilution factor (see Table 7.1) (8.99E-06 sec/m'); t F is the release flow rate (in m'/sec); i C, is the concentration of radionuclide i, in uCi/cc; t [ DAC, is the derived air concentration of radionuclide i as specified in 10 CFR Part 20, Appendix B, Table II. Column 1, in uCi/cc. [ The DAC-fractions for the various release points are then summed to yield the [ total DAC-fraction, R: l R=ER (6-6) 3 1 [ The total DAC-fraction, R, at the most conservative site boundary or off-site location must be less than or equal to one. R s 1. (6-7) 3/93 L:\\DS\\15.EED 64 i i l 1 l 1

i i l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL 6.2.2 Monitor Response for Gaseous Effluents Normal radioactivity releases consist mainly of well-decayed fission gases. Therefore, monitor response calibrations are performed using fission gas typical of normal release (mainly Xenon-133). The total concentration of radioactive materials in gaseous effluents, in uCi/ct, at the monitor is calculated. The calibration cur'!e or constant, in i cpm /(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). f l i i t l I i l i 1/92 L:\\DS\\15.EED 65

l l r MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL l FIGURE 6.1 Maine Yankee licuid Radwaste System i Pr mary $ysten a g yarogenated Crains vent to waste cas surge Tane P Baron j P rtu ry l Waste Soren 2 storase

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i 1/92 L:\\03\\15.EED 66 I

- 1 l i 1 i ) l l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL l FIGURE 6.2 Maine Yankee Gaseous Radwaste System i I i Conminment Air Rac5a:en Monitor Alarm funeson of on line purge incia:on l AM-61c2X, RS61c2Y RC* MC"" Conminment Purge l Stack Raciaton IW2rutor A:atm & on line purge isciaten RM3SC2X. RPMSC2Y S:ack Canonuous Sampier Pnmary Aux #1ary Building (Stuaiced Areas) l i RCA Storage Area - Radiacon Moniw. Alarm Funcan ~ n RM-1801 Condenser Air Rec: ors w I w Radianon Moniw. Alarm & tsolaeon l RM-3901 Func:en i ~ ~ ~ ~ ~ "s Cas systems" * ~ ~ ~ ~ ~ ~ ~ ~ 8 Wast E:owdown Vent . t C H P i y s T _ sowineicanon ll 8 8

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i l 1 2 i 5 I MAINE YANKEE ATOMIC POWER COMPANY i 0FF-SITE DOSE CALCULATION MANUAL 7.0 METEOROLOGY 1 i 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 l 1 the reasonable off-site locations to find the location with the most limiting i j 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 i locations within five miles of the plant. 1 i Exposure pathways assumed to exist at site boundary locations are direct ( i exposure from radioactive materials in the air, direct exposure from radioactive I l materials deposited on the ground, and exposure from inhalation of radioactive I materials. In addition to the pathways present at site boundary locations, exposure pathways present at each residence are assumed to include ingestion of radionuclides in home grown vegetables. Farm locations include all exposure pathways found at residences plus ingestion of radionuclides in meat and milk. l [ Meteorological data for the year 1986 through 1990 were analyzed for the values of the maximum average dilution factors at the important receptor locations 1 [ described above. Yankee Atomic Electric Company's (YAEC) AE0LUS-2 computer code

i (Reference 5) calculated all atmospheric dilution factors. Appendix B briefly

[ describes the YAEC AEOLUS-2 computer code model. Table 7.1 lists the maximum a i average dilution factors for all important receptor locations for releases via the } plant stack. i 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 f will be conservatively estimated by the dose calculated for a full-time resident living on a hypothetical milk farm 670 to 700 meters from the plant in the southeast sector. + i i 3/93 t n SU S.EED 68

i l i t MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL j TABLE 7.1 f Maine Yankee I l Maximum 5-Year Averaae Atmospheric Dilution FactorsM Y l Direction From Distance From X/Q U1dep'leted. X/Q Depleted. (sec/m{) D/ [X/Q (m',Q) the Plant the Plant (m) (sec/m ) ( s ec /m')- j i Site Boundaries [ N 1219 6.28E-07 5.92E-07 4.87E-09 6.33E, [ NNE 2209* 3.65E-07 3.45E-07 1.77E-09 3.21E-07 l [ NE 1280* 2.17E-07 2.02E-07 1.74E-09 2.58E-07 l l [ ENE 914* 2.01E-07 1.86E-07 1.59E-09 2.85E-07 l [ E 731* 2.89E-07 2.70E-07 2.46E-09 4.20E-07 l [ ESE 670* 5.75E-07 5.33E-07 6.57E-09 5.89E-07 i [ SE 670* 1.18E-06 1.09E-05 1.46E-08 1.06E-06 l [ SSE 823* 6.80E-07 6. 6.49E-09 7.60E-07. l [ S 1310* 3.03E-07 2.bu 2.59E-09 3.58E-07 i [ SSW 2986* 1.33E 1.25E-07' 6.98E-10 _1.49E-07 l [ NNE 1368** 6.19E-07 5.90E-07 3.54E-09 5.58E-07 i [ NE 370** 8.22E-07 7.82E-07 6.87E-09 '8.16E-07 [ ENE 306"* 8.56E-07 8.19E-07 5.49E-09 7.22E-07 l [ E 209** 1.80E-06 1.74E-06 1.03E-08 1.C5E-06 l [ ESE 209** 3.67E-06 3.55E-06 3.08E-08 1.98E-06 l [ SE 274** 4.78E-06 4.59E-06 5.02E-08 2.85E-06 [ SSE 209** 5.85E-06 5.66E-06 3.82E-08 2.92E-06 [ S 129** 8.99E-06 8.79E-06 4.69E-08 3.33E-06 l l [ SSW 129** 6.36E-06 6.22E-06 3.21E-08 2.46E-06 l [ SW 914 2.22E-07 2~07E-07 2.19E-09 3.31E-07 i [ WSW 762 1.77E-07 1.67E-07 1.67E-09 3.24E-07 [ W 670 1.93E-07 1.86E 1.58E-09 3.52E-07. [ WNW 670 1.74E-07 1.65E-07 1.55E-09 3.04E-07 i [ NW 762 2.60E-07 2.47E-07 2.01E-09 4.07E-07 [ NNW 1036 3.82E-07 3.55E-07 3.33E-09 4.66E-07 l [ (1) 1986 through 1990 for release via the plant stack. l

• 0pposite shore of Back River where site boundary coincides with the water j

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• • Near shoreline where site boundary coincides.with water boundary of river.

j Occupancy factors for worm diggers apply l l l l 3/93 l L:\\DSUS.EED 69 Effective: 05-24-93

i i i i MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i TABLE 7.1 (Continued) Maine Yankee Maximum 5-Year Averace Atmospheric Dilution FactorsN i Y Direction From Distance From X/Q Undep'leted X/Q Depleted D/ (sec/m{) [X/Q (m',Q) the Plant the Plant (m) (sec/m ) _ ( sec /m') Mud Flats (Worm Diggers)*** [ E 250 1.31E-06 1.26E-06 8.23E-09 8.91E-07 [ S 450 1.02E-06 9.59E-07 9.26E-09 9.30E-07 [ SSW 700 3.96E-07 3.67E-09 3.96E-09 4.69E-07 i [ SW 400 5.77E-07 5.48E-07 4.78E-09 6.24E-07 [ WSW 300 5.07E-07 4 87E-07 3.54E-09 5.42E-07 [ W 250 4.58E-07 4.41E-07 3.73E-09 4.27E-07 [ WNW 250 5.46E-07 5.26E-07 4.33E-09 4.60E-07 [ NNW 250 2.43E-06 2.34E-06 1.81E-08 1.65E-06 Nearest Residences i [ N 1300 5.95E-07 5.61E-07 4.48E-09 5.93E-07 [ NNE 2300 3.47E-07 3.28E-07 1.67E-09 3.06E-07 [ NE 1300 2.14E-07 2.00E-07 1.71E-09 2.54E-07 [ ENE 900 1.88E-07 1.73E-07 1.55E-09 2.58E-07 [ E 900 2.41E-07 2.25E-07 1.92E-09 3.43E-07 [ ESE 1400 3.26E-07 3.04E-07 2.53E-09 3.56E-07 i [ SE 700 1.11E-06 1.03E-06 1.37E-08 1.01E-06 [ SSE 900 6.09E-07 5.60E-07 5.77E-09 6.93E-07 [ S 1700 3.50E-07 3.31E-07 1.96E-09 3.70E-07 [ SSW 3000 1.33E-07 1.24E-07 6.93E-10 1.48E-07 [ SW 1400 1.91E-07 1.81E-07 1.45E-09 2.55E-07 [ WSW 1000 1.51E-07 1.43E-07 1.29E-09 2.52E-07 l [ W 2600 9.83E-08 9.50E-08 3.73E-10 1.02E-07 [ WNW 800 1.63E-07 1.55E-07 1.33E-09 2.63E-07 [ NW 2000 1.68E-07 1.61E-07 7.19E-10 1.77E-07 [ NNW 1100 3.63E-07 3.37E-07 3.08E-09 4.38E-07 [ (1) 1986 through 1990 for release via the plant stack.

• • Dilution factors must be corrected for occupancy factors of worm diggers on the mud flats at low tide equal to 325 hours per year (0.037).

Reference:

Maine Yankee Environmental Report. 3/93 Effectivet 05-24-! U VSUS.EED 70

i MAINE YANKEE ATOMIC POWER COMPANY I 0FF-SITE 00SE CALCULATION MANUAL APPENDIX A Basis for the Dose Calculation Methods A.1 Liouid Effluent 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 drinking or irrigation. Method I is derived from Equations A-3 and A-7 of Regulatory Guide 1.109 (Reference 2). Equation A-3 calculates radiation doses from aquatic foods. Equation A-7 from shoreline deposits. l 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 i dose conversion factors giving the number of millirem per curie released for each radionuclide. Since 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 individual independent 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. I Liquid effluents discharge from the plant via a submerged multi-port diffuser ] which extends approximately 1000 feet into the tidal estuary and has a design circulating water flow of 420,000 gpm (935 ft'/sec). For the aquatic foods pathway, the dilution for the mixing effect of the diffuser is 10 to 1 (Reference 3 6). This dilution applies to the edge of the initial mixing zone where the i i effluent has undergone prompt dilution only. For shoreline deposits, the nearest l point where tidal flats could be occupied on a recurring basis is in Bailey Cove which borders the site on the south and west. The estimated average dilution for l Bailey Cove with respect to the discharge is approximately 25 to I (Reference 6). 1/92 1 L:\\D5115.EE3 71 l \\

MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL APPENDIX A Shoreline activities in the vicinity of the site include a commercial worm diggir.g 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 l contaminated water is fifteen years. This time period represents the approximate mid-point of plant operating lifetime, and thus allows for the calculation of a plant lifetime average concentration of radioactivity in sediment. No credit is taken for the decay of activity in transit from the discharge point to the sediment in Bailey Cove. i 6 i J i 3/93 L:\\DS\\15.EED 72 i s ,,e

MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL APPENDIX A TABLE A-1 I Usaae Factors for Various Licuid Pathways at Maine Yankee (From Reference I, Table E-5*, except..s noted. Zero where no pathway exists.) LEAFY P0 TABLE AGE VEG. 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** i 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 t

• Regulatory Guide 1.109.
  • Regional shoreline use associated with mudflats - Maine Yankee Atomic Power l

Station Environmental Report. l r P b k 3/93 L.\\osus.Eto 73

i MAINE YANKEE ATOMIC PrNER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX A A.2 Total Body Dose Rate from Noble Gases Method I can be used to demonstrate compliance with Section 2.2.3.1.a, which limits total body dose rate from noble gases released to the atmosphere. Method I applies the methods of Equation B-8 in Regulatory Guide 1.109 l l (Reference 2) as follows: D. - Sp 3.17E+04 [X/Q]" Z Q, DFB, (A-1) i i where: De is the annual total body dose, in mrem /yr; S is the attenuation factor that accounts for the dose reduction due to p shielding provided by residential structures, but for all dose rate calculations is assumed to be equal to 1 (dimensionless); 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'; t Q, is the annual release rate of radionuclide i, in Ci/yr; and

DFB, is the total body gamma dose factor for radionuclide i, in l

mrem-m'/pCi-yr. The analysis of Maine Yankee five-year average meteorology presented in i Section 7.0 yielded a maximum effective average gamma dilution factor, [X/Q]r, of l [ 1.06E-06 sec/m. The maximum gamma dilution factor was identified for an off-site l [ point located 670 meters southeast of the plant. This location is along the l opposite shoreline of Back River 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 value (south sector at 129 meters, [X/Q]r = 3.33E-6 sec/m ).

However, the definition of site boundary in the Technical Specifications allows for the use of occupancy factors in assessing doses, and the expanded 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 l 1ccations. As a result, the 3/93 L:\\csus.tto 74

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL opposite shoreline atmospheric gamma dilution factor becomes limiting due to its assumed full time occupancy since physical constraints (areas over water) do not exist, and there is no control on occupancy available. 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 into Equation A-1 and converting from annual release Q (Ci/yr) to maximum instantaneous release rate 6 (uCi-sec), and multiplying by the conversion constant 31.54 Ci-sec/uci-yr yields the method to calculate total body dose rate from noble gases: b, = 1.06

DDFB, (A-2) i

[ i I i \\ i l 3/93 L:\\DS\\15.EED 75 i

MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL APPENDIX A l A.3 Skin Dose Rate From Noble Gases f 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. Method I applies the methods of Equation 11 in Regulatory Guide 1.109 (Reference 2) as follows b,gn = 1.11 S 3.17E-04 [X/0]' E 0, DF[ + 3.17E+04 X/Q E 0, DFS, (A-3) f y i i where: D,en is the annual skin dose rate, in mrem /yr; 1.11 is the average ratio of tissue to air ener'gy absorption coefficient; i Sg 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]' is the effective long term average gamma dilution factor in sec/m'; Qi is the annual release rate of radionuclide i, in Ci/yr; I DF( is the gamma air dose factor for a uniform semi-infinite cloud of radionuclide i, in mrad-m'/pCi-yr; X/Q is the long term average undepleted dilution factor in sec/m'; and 1 i 3/93 i i L:\\DS\\15.EED 76 l

i l MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL APPENDIX A

DFS, is the beta skin dose factor for a semi-infinite cloud of radionuclide i, which includes the attenuation by the outer

" dead" layer of the skin, in mrem-m /pCi-yr (taken from Reference 1, Table B-1). The maximum effective five year average gamma dilution factor [X/Q]', is I 1.06E-06 sec/m' (see Table 7.1), and the maximum five year average 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 Q (uCi/sec) and multiplying by ( the conversion factor 31.54 Ci-sec/uCi-yr yields: Y [ D,g;n = 1.18 Q, M + 1.18 E Q, E, i i i (A-4) Y l [ = E Q, [1.18 DF + 1.18 DFS,]. l i A combined skin dose factor, DF(, may be defined: =1.18DF(+1.18DFS;. [ DF Incorporating the combined skin dose factor, DF j, into Equation A-4 yields the method to calculate skin dose rate from noble gases:

b. kin ={,D 1

3/93 Effective: 05-24-9: U\\DSUS.EED 77

i MAINE YANKEE ATOMIC POWER COMPANY i 0FF-SITE DOSE CALCULATION MANUAL APPENDIX A f A.4 Critical Oroan Dose Rate From Iodines and Particulates Method I is used to demonstrate compliance with Section 2.2.5, which limits the dose rate from Iodine-131, Iodine-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 0DCM Equation 4-6 which limits the dose to the critical organ from radioactive iodines and particulates. r D,, = E Q, DFG,c, '(A-5) l i j where: D,, is the dose to the critical organ from Iodine-131, Iodine-133, tritium, l 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 of interest, in Ci; and D FG,c, is the site specific critical organ dose factor for radionuclide i for a gaseous release, in mrem /Ci (see Table 4.2). Applying the conversion factor, 31.54 (Ci-sec/uCi-yr), to convert DFG,, i 4 (mrem /Ci) to an organ dose rate factor' DFG; g (mrem-sec/uci-yr) for use for iodines and particles and changing the shielding factor (Sr) from 0.7 to 1.0 for exposure from a contaminated ground plane yields a new critical organ dose rate factor DFGico (see Table 4.2), and a dose rate equation in the same form as Equation A-5 above, where the activity release rate Q,is in uCi/sec. D,, = E 0, DFGlc, (A-6) i 3/93 L:\\DS\\15.EED 78 4

i l MAINE YANKEE ATOMIC POWER COMPANY i 0FF-SITE DOSE CALCULATION MANUAL APPENDIX A l A.5 Gamma Air Dose i Method I is used to demonstrate compliance with Section 2.2.4, which limits the gamma air dose due to noble gases released in gaseous effluents via the plant 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: D[ny,,,, = 3.17E+04 [X/Q][ E Q,0F[ (A-7) 1 where: D'un,.,, 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); Q, is the total activity of noble gas i released via the plant stack during the period of interest, in Ci; and DF/ is the gamma dose factor to air for noble gas 1, in mrad-m'/pci-yr (taken from Reference 2). Incorporating (the maximum effective long-term average gamma diiution factor [ of 1.06E-06 sec/m see Table 7.1) yields: [ D',,, = 0.034 E 0, DF( (4-4) i l 3/93 L:\\05\\15.EED 79 i I

MAINE YANKEE ATOMIC POWER COMPANY l OFF-SITE DOSE CALCULATION MANUAL I l APPENDIX A A.6 Beta Air Dose Method I is used to demonstrate compliance with Section 2.2.4, which limits l the beta air dose due to noble gases released in gaseous effluents via the plant stack to areas at and beyond the site boundary. i Method I is derived from the methods of Equations B-4 and B-5 in Regulatory i Guide 1.109 (Reference 2) which gives-i D[r = 3.17E-04 X/Q E Q, DF[ (A-8) i l where: l t Dfir is the beta air dose, in mrad; l I 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'; Q, is the total activity of noble gas i released via the plant stack during i the period of interest, in Ci; and i DFf is the beta dose factor to air for noble gas i, in mrad-m /pci-yr. 3 Incorporating the maximum long-term average undepleted dilution factor of [ 1.18E-06 sec/m' (see Table 7.1) yields: Df, = 0.037 E Q, DF[ (4-5) [ i l l 3/93 L:\\DSUS.EED 80 l

i i i MAINE YANKEE ATOMIC POWER COMPANY i 0FF-SITE DOSE CALCULATION MANUAL APPENDIX A A.7 Dose from lodines 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, Iodint-133, tritium, and radioactive materials in particulate form with half-lives greater than eight days in gaseous effluents released via the plant stack 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 l potential for exposure pathway's to exist on a continuous basis. The maximum of l 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: external irradiation from radionuclides deposited on.the ground surface, l inhalation of radionuclides in air, and ingestion of atmospherically released radionuclides in food. The use of the methodology cf Equations C-2, C-4, and C-13 for a one curie l release of each radionuclide in gaseous effluents yielded the dose impact to the l 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 burning of contaminated waste oil. Parameter values used to derive the critical organ dose factors for iodines and particulates are listed on Tables A-2 and A-3. l L:\\DS\\15.EED 81 i

r J MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX A Milk and meat animals are assumed to be on pasture 50 percent of the time, consuming 100 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. 1 3 l l \\ l 3/93 l L:\\DS\\15.EED 82

1 1 4 J 1 1 i MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL 1 APPENDIX A TABLE A-2 Usage Factors for Various Gaseous Pathways at Maine Yankee l (From Reference 1, Table E-5*) AGE LEAFY l GROUP VEG. VEG. MILK MEAT INHALATION (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 t Child 520.00 26.00 330.00 41.00 3,700.00 Infant 0.00 0.00 330.00 0.00 1,400.00 a

• Regulatory Guide 1.109.

4 3/93 L:\\DS\\15.EED 83 i I

l l MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL t I APPENDIX A TABLE A-3 i Environmental Parameters for Gaseous Effluents at Maine Yankee i (Derived from Reference 1)* Veaetables Cow Milk __ Goat Milk Meat Variable Stored Leafy Pasture Stured Pasture Stored Pasture Stored YV Agricultural (kg/m') 2. 2. 0.75 2. 0.75 2. 0.75 2. Productivity 2 P Soil Surface Density (kg/m ) 240. 240. 240. 240. 240. 240. 240. 240. i T Transport Time to User (hrs) 48. 48. 48. 48. 480. 480. TB Soil Exposure Time (" (hrs) 131400. 131400. 131400. 131400. 131400. 131400. 131400. 131400. TF Crop Exposure Time (hrs) 1440. 1440. 720. 1440. 720. 1440. 720. 1440. T to Plume TH Holdup After llarvest (hrs) 1440. 24. O. 2160. O. 2160. O. 2160. QF Animals Daily Feed .(kg/ day) 50, 50. 6. 6. 50. 50. FP-FractionogYear 0.50 0.50 0.50 i on Pasture FS FractionPasturgFeed 1. 1. 1. When on Pasture FG ~ Fraction of Stored 0.76 Veg. Grown in Garden FL Fraction' of Leafy 1.0 Veg. Grown in Garden i 3/93 j t:sossis.tto 84

MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL APPENDIX A TABLE A-3 Environmental Parameters for Gaseous Effluents at Maine Yankee (Derived from Reference 1)* Venetables Cow Milk Goat Milk Meat Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored FI Fraction Elemental Iodine = 0.5 H Absolute (gm/m ) W llumidity = 5.6 o Regulatory Guide 1.109 Notes: (1) For Method II dose / dose rate analyses of identified radioactivity releases of less than one year, the soil exposure time for that release may be set at 8760 hours (1 year) for all pathways. (2) For Method II dose / dose rate analyses performed for releases occurring during the first or fourth calendar quarters, the fraction of time animals are assumed to be on pasture is zero (non-growing season). For the second and third calendar quarters, the fraction of time on pasture (FP) will be set at 1.0. FP may also be adjusted for specific farm locations if this information is so identified and reported as part of the land use census. (3) For Method II analyses, the fraction of pasture feed while on pasture may be set to less than 1.0 for specific farm locations if this information is so identified and reported as part of the land u:e census. (4) For all Method II analyses, an absolute humidity value equal to 5.6 (gm/m') shall be used to reflect conditions in the Northeast (

Reference:

llealth Physics Journal, Vol. 39 (August), 1980; Page 318-320, Pergammon Press). 3/93 t nosus.tro 85

MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL APPENDIX B r Meteoroloay i Long term (annual and five-year) average dilution factors based on on-site meteorological data were computed for routine primary vent stack releases by the [ Yankee Atomic Electric Company's (YAEC) AEOLUS-2 (Reference 5) computer code. [ AE0LUS-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 l temperature difference) straight-line air flow model with Gaussian diffusion, i part-time ground level and part-time elevated releases (split-H model), multi-energy sector-averaged finite cloud dilution factors for gamma dose i calculations, terrain height correction features, plume rise (momentum), depletion 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, i 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 of elemental radiciodines and other particulates. 3/93 L:\\DS\\15.EED 86 i

l I l i d ~ MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL i APPENDIX B Gamma dose rates are calculated throughout the ODCM using the finite 3 cloud model presented in Metcorology and Atomic Energy - 1968" (Reference 9, Section 7-5.2.5). That model is implemented through the definition of an effective gamma atmospheric dispersion factor, [X/Q]r (Reference 5, Section 6), and the replacement of X/Q in infinite cloud dose equations by the [X/Q]r. i t i 1 I i I l ~ i 3/93 ' L:\\DS\\15.EED 87

t 1 i i MAINE YANKEE ATOMIC POWER COMPANY ^ OFF-SITE DOSE CALCULATION MANUAL I APPENDIX C Routine Reports 1. Annual Radiolooical Environmental Operatina Recort l The Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 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 2.4.4 of the ODCM. l t The Annual Radiological Environmental Operating Reports shall include summarized and tabulated 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 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. Semiannual Radioactive Effluent Release Report The Semiannual Radioactive Effluent Release Report covering the operating of the unit during the previous six months of operation shall be submitted within 60 days after January 1 and July 1 of each year. 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 shail also include a summary of the solid waste released from the unit summarized on a semiannual basis. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50. 3/93 L:\\DS\\1s.EED 88

I I MAINE YANKEE ATOMIC POWER COMPANY 0FF-SITE DOSE CALCULATION MANUAL i The Radioactive Effluent Release Reports shall include the following l information for each class of solid waste (as defined by 10 CFR Part 61) l 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). e. Type of container (e.g., LSA, Type A, Type B, large Quantity). i f. Solidification agent or absorbent (e.g., cement, asphalt, "Dow"). 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. l 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. i The Radioactive Effluent Release Report shall include changes to the ODCM for i information. 1 3/93 L:\\Dsus.Etc 89

1 l \\ MAINE YANKEE ATOMIC POWER COMPANY OFF-SITE DOSE CALCULATION MANUAL i l REFERENCES 1. Title 10, Code of Federal Regulations. The Office of the Federal Register, National Archives and Records Administration. 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, j October 1977. I 3. International Commission on Radiological Protection (ICRP) Publication 2. l 0xford: Pergammon. 4. Title 40, Code of Federal Regulations. The Office of the Federal Register, l National Archives and Records Administration. [ 5. Hamawi, J.N., "AEOLUS Technical Description", Entech Engineering, Inc., [ Document No. P100-R13-A, YAEC - Revised Software Release MOD 05, dated March [ 1992. t 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. l l 3/93 Lacsu5.EED 90}}