annual radioactive effluent report 1995 iannual radioactive effluent report 1995 i northeast nuclear...

AnnualRadioactive

Effluent Report1995

I

Northeast Nuclear Energy CompanyMillstone Nuclear Power Station

Unit123

LicenseDPR-21DPR-65NPF-49

Docket50-24550-33650-423

9605030141 960426PDR ADOCK 05000245R PDR

Table of Contents

Introduction

1.0 Doses

1.1 Dose Calculations

1.2 Dose Results

2.0 Radioactivity

2.1 Airborne Effluents

2.2 Liquid Effluents

2.3 Solid Waste

3.0 REMM/ODCM Changes

4.0 Inoperable Effluent Monitors

5.0 Errata

Introduction

This report is being submitted for Northeast Nuclear Energy Company's Millstone Power Station's Units 1,2, and 3, in accordance with 10CFR50.36a and the Radiological Effluent Technical Specifications and inthe US NRC Regulatory Guide 1.21 format. A combined report is being submitted for all three unitsbecause they share some common effluent facilities.

The report provides radioactivity information for airborne and liquid effluents and solid waste. Doses andregulatory limits are provided for airborne and liquid effluents. If applicable, any changes to theREMM/ODCM, description of any effluent monitors inoperable for more than 30 days, and any correctionsto previous reports are included.

The annual capacity factor for: Unit 1 was 75.4%, Unit 2 was 35.6%, and Unit 3 was 79.0%. Listedbelow are the major shutdowns for each unit during 1995:

Period

Unit 1 06/2708/1811/04

Unit 2 01/0108/0812/14

07/1108/2212/31

08/0408/1512/20

Reason

Both Emergency AC Power Trains Declared InoperableSafety Relief Valve RepairsRefuel Outage

Refuel OutageHeater Drain Pump Recirculation Pipe RepairsCharging Header Thermal Relief Valve Repairs

Refuel OutageRCS Valve Repairs

Unit 3 04/14 06/0711/30 - 12/15

1.0 Doses

This report provides a summary of the estimated off-site radiation doses from routine releases ofradioactive materials in airborne and liquid effluents for each unit. Included are the annual populationdose commitments (person-rem) for the area within 50 miles of the site, the annual average dosecommitment (mrem) to the population, and the annual maximum dose commitment (mrem) to any realmember of the public. Also provided are the maximum gamma and beta air doses.

The doses are compared with the regulatory limits and with the annual average population dose

commitments from natural background and other sources to provide perspective.

1.1 Dose Calculations

The off-site dose to humans from radioactive airborne and liquid effluents have been calculatedusing measured radioactive effluent data, measured meteorological data, and dose computermodels developed by the US Nuclear Regulatory Commission (NRC) and the EnvironmentalProtection Agency (EPA). These doses generally tend to be conservative because of theconservative assumptions used in these models. More realistic estimates of the off-site dosecan be obtained by analysis of environmental monitoring data. A comparison of dosesestimated by each of the above methods will be presented in the Annual RadiologicalEnvironmental Monitoring Report.

1.1.1 Population and Maximum Individual Dose Commitment

Population dose commitment is defined as the total radiation dose received by thespecified population in a specified time period from an identified radiation source. Forthis report, the specified population is defined as the population within 50 miles of theMillstone nuclear site. The doses are based upon continuous exposure to the airborneand liquid effluents for a one year period and an associated dose commitment over a50-year period from initial exposure, taking into account radioactive decay and biologicalelimination of the radioactive materials contributing to the dose. The population dosecommitment (person-rem) is the sum of the individual doses.

The dose calculations are based upon these three types of input: radioactive sourceterm, site specific data, and generic factors. The radioactive source term (Curies) ischaracterized in the Effluent Activity section of this report. The site specific dataincludes: meteorological data (e.g. wind speed, direction, stability, etc.) to calculate thetransport and dispersion of airborne effluents, dilution factors for liquid effluents, thepopulation distribution and demographic profile surrounding the site by compass sector.Other site specific data include the average annual production of milk, meat, vegetation,fish, and shellfish. The generic data includes the average annual consumption rates (forinhalation of air and ingestion of fruits, vegetables, leafy vegetables, grains, milk,poultry, meat, fish, and shellfish) and occupancy factors (for air submersion and groundirradiation, shoreline activity, swimming, boating, etc.) for determining the dose to theindividual who would receive the maximum dose (maximum individual). All these inputsare used in the appropriate dose models to calculate the population and individual dosecommitments from radioactive site airborne and liquid effluents.

1.1.1.1 Airborne Effluents

Maximum individual doses and population doses due to the release of noblegases, radioiodines, and particulates were calculated using the computercode GASPAR (Ref 1), with the exception of Unit 1 noble gas doses. Themaximum individual dose due to direct exposure from the Unit 1 noble gas

plume was calculated using the computer code AIREM (Ref 2).

The GASPAR code is an NRC code which uses a semi-infinite cloud modelto implement the NRC Regulatory Guide 1.109 (Ref 3) dose models. TheAIREM code is an EPA code which uses a sector averaged Gaussiandiffusion model that includes ground and inversion lid reflections, radionuclidedecay, first daughter in-growth, ground deposition, cloud depletion, and dosecontributions from radionuclides in clouds at all azimuths. The finite cloudmodel used in AIREM is a modified version of R.E. Cooper's code EGAD (Ref4).

The values of average relative effluent concentration (X/Q) and averagerelative deposition (D/Q) used in the GASPAR code were generated using ameteorological computer code which implements the assumptions cited inNRC Regulatory Guide 1.111 (Ref 5), Section C. The annual summary ofhourly meteorological data (in 15-minute increments), which includes windspeed, direction, atmospheric stability, and joint frequency distribution, is notprovided in the report but can be retrieved from computer storage.

Unit 1 (375 ft) stack releases are always considered elevated releases; and,Pasquill stability classes are determined based upon the temperature gradientbetween the 33 ft and 447 ft meteorological tower levels. BecauseGASPAR's semi-infinite cloud model would underestimate the direct exposuredose from the plume within 2 miles of the stack, AIREM was used todetermine the maximum individual exposure from an overhead finite gammacloud.

Unit 2 (159 ft) vent releases are considered mixed mode (partially elevatedand partially ground) releases; and, Pasquill stability classes are determinedbased upon the temperature gradient between the 33 ft and 142 ftmeteorological tower levels. GASPAR was used to calculate doses for Unit2 continuous releases through the Unit 2 vent (building ventilation) and steamgenerator blowdown tank flashed gases; and, doses for Unit 2 batch releasesthrough the Unit 1 stack (containment vents and waste gas tanks) and Unit2 vent (containment purges). These doses were analyzed to determine thetotal Unit 2 airborne effluent dose.

Unit 3 (142.5 ft) vent releases are considered mixed mode (partially elevatedand partially ground) releases; and, Pasquill stability classes are determinedbased upon the temperature gradient between the 33 ft and 142 ftmeteorological tower levels. GASPAR was used to calculate doses for Unit3 continuous releases through the Unit 3 vent (building ventilation) and forUnit 3 batch releases from containment purges and drawdowns. Thesedoses were analyzed to determine the total Unit 3 airborne effluent dose.

1.1.1.2 Liquid Effluents

Maximum individual and population doses from the release of radioactiveliquid effluents were calculated using the LADTAP (Ref 6) code, which usesthe dose models and parameters cited in NRC Regulatory Guide 1.109.

1.1.2 Gamma and Beta Air Doses

Maximum gamma and beta air doses from the release of noble gases were calculatedusing the GASPAR code, with the exception of the Unit 1 gamma air dose which usedthe greater dose from either GASPAR or AIREM.

1.2 Dose Results

1.2.1 Airborne Effluents

For population doses, the GASPAR code calculates the dose to the whole body, GI-tract, bone, liver, kidney, thyroid, lung, and skin from each of the following pathways:direct exposure from the plume and from ground deposition, inhalation, vegetation,cow's milk, and meat. The values presented are a total from all pathways; however,only the whole body, skin, and maximum organ dose are presented.

For the dose to the maximum individual, the GASPAR code calculates the dose to thesame organs listed above for the following pathways: direct exposure to the plume(except for Unit 1 finite cloud doses), exposure from ground deposition, inhalation, andingestion of vegetation, meat, cow's milk, and goat's milk. The doses are calculated foradults, teenagers, children, and infants separately.

For the plume and inhalation pathways, the maximum individual dose is calculated atthe off-site location of highest decayed x/Q where a potential for dose exists.

For ground deposition, the maximum individual dose is calculated at the off-sitemaximum land location of highest x/Q and highest D/Q where a potential for doseexists.

For the vegetation pathway, the maximum individual dose is calculated at the vegetablegarden of highest D/Q. For the meat, cow's milk, and goat's milk pathways, thecalculated dose is included for the maximum individual's dose only at locations andtimes where these pathways actually exist. Doses were calculated at the cow farm andgoat farm of maximum deposition.

The AIREM code calculates the individual whole body and skin dose for each sector-segment. The maximum individual dose is obtained by taking the maximum AIREMresult at the off-site location where a potential for dose exists and multiplying by a factorof 0.7 to compensate for building shielding and occupancy.

To determine compliance with 10CFR50, Appendix I (Ref 7), the maximum individualwhole body dose only includes the external pathways (i.e. plume and ground exposure)while the maximum individual organ dose only includes the internal pathways.Population doses include all applicable pathways.

The air dose includes only the dose from noble gases in the plume. Hence, if theground shine contribution was significant, there may be cases where the maximumwhole body or skin dose is greater than the maximum gamma or beta air doserespectively. To determine the gamma air dose from the overhead finite cloud, theAIREM results were divided by 0.7 to convert mrem whole body to mrad gamma airdose.

The off-site dose commitments from airborne effluents are presented in Table 1-1.These doses are the maximum doses observed.

1.2.2 Liquid Effluents

The LADTAP code performs calculations for the following pathways: fish, shellfish,algae, drinking water, irrigated food, shoreline activity, swimming, and boating. AtMillstone, the algae, drinking water, and irrigated food pathways do not exist; and, thus,only the other pathways are included in the totals. Doses are calculated for the wholebody, skin, thyroid, GI-LLI, bone, liver, kidney, and lung. Calculations are performedseparately for adults, teenagers, and children.

The off-site dose commitments from liquid effluents are presented in Table 1-2. Thesedoses are the maximum doses observed.

1.2.3 Analysis of Results

The doses are well below the permissible levels and small in comparison to the dosefrom natural background radiation. The statistical expectation of adverse health effectsfrom the calculated radiation dose due to nuclear plant operations is essentially zero.

Refer to Table 1-3 for the annual effluent whole body airborne and liquid doses for themaximum and average individuals and a quantitative comparison between the dosesfrom nuclear plant operations and those doses from other sources such as naturallyoccurring background radiation. Also provided are legally allowed 40CFR190 (Ref 8)levels to a real member of the public. Note that the maximum individual in Tables 1-1,1-2, and 1-3 is conservatively assumed to be a real member of the public.

Prior to its refuel outage, Unit 3 was operating with a degraded fuel rod. Duringrefueling, this caused higher than normal primary side activity to be vented. This, inturn, resulted in a relatively higher second-quarter airborne thyroid dose for Millstone.

For compliance with 40CFR190, any direct dose from the site must be added to theeffluent dose to a "real member of the public." At Millstone, the only potential directdose of significance would be from Unit 1 turbine shine and radwaste storage.However, all radwaste storage during this year was well within storage criteria thatensure the public dose would be less than 1 mrem. Hence, the direct dose would beless than 1 mrem and inconsequential for 40CFR190 compliance. The Unit 1 turbineshine dose is 2.0 mrem to the maximum individual, who is assumed to be a lobstermanthat frequents the water immediately outside the Unit 1 turbine building. Adding thisdose to the 0.1 maximum whole body dose from effluents yields a total of 2.1 mrem,which is less than ten percent of the 25 mrem limit.

References

1. NUREG-0597 User Guide to GASPAR Code, KF Eckerman, FJ Congel, AK Roecklien,WJ Pasciak, Division of Site Safety and Environmental Analysis, Office of NuclearReactor Regulation, US Nuclear Regulatory Commission, Washington, DC 20555,manuscript completed January 1980, published June 1980.

2. EPA-520/1-74-004, AIREM Program Manual - A Computer Code for Calculating Doses.Population Doses. and Ground Depositions Due to Atmospheric Emissions ofRadionuclides, JA Martin Jr, CB Nelson, PA Cuny, Field Operations Division, Office ofRadiation Programs, US Environmental Protection Agency, Washington, DC 20460, May1974.

3. NRC Regulatory Guide 1.109 Calculation of Annual Doses to Man from RoutineReleases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFRPart 50. Appendix I, Revision 1, October 1977.

4. DF-1304 EGAD - A Computer Program to Compute Dose Integrals from ExternalGamma Emitters, RE Cooper, Mathematics and Computers (TID-4500, VC32),Savannah River Laboratory, Aiken, SC, September 1972.

5. NRC Regulatory Guide 1.111 Methods for Estimating Atmospheric Transport andDispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,Revision 1, July 1977.

6. NUREG/CR-1276, ORNL/NUREG/TDMC-1 User's Manual for LADTAP II - A ComputerProgram for Calculating Radiation Exposure to Man from Routine Release of NuclearReactor Liquid Effluents, DB Simpson, BL McGill, prepared by Oak Ridge NationalLaboratory, Oak Ridge, TN 37830, for Office of Administration, US Nuclear RegulatoryCommission, manuscript completed 17 March 1980.

7. 10 CFR Ener•y, Part 50 Domestic Licensing of Production and Utilization Facilities,Appendix I Numerical Guides for Design Obiectives and Limiting Conditions forOperation to Meet the Criterion "As Low As Reasonably Achievable" for RadioactiveMaterial in Light-Water-Cooled Nuclear Power Reactor Effluents, 1 January 1994Edition.

8. 40 CFR Environmental Protection Agency, Part 190 Environmental Radiation ProtectionStandard for Nuclear Power Operation, Revision 0, 13 January 1977.

Table 1-1

1995 Off-Site Dose Commitments from Airborne EffluentsMillstone Units 1, 2, 3

Unit I 1st Quarter 2nd Quarter 3rd Quarter 4th QuarterMax Air (mrad) (mrad) (mrad) (mrad)

Beta 3.09E-06 a 1.6miE 1.55E-03 @ 3.5rmiWNW 1.17E-04 0 3.5miNNE 7.53E-05 0 1.6 miESEGamma 1.32E-06 @1.6SmiE 1.10E-02 @0.3miNE 1.55E-04 @3.5miNNE 1.OOE-04 @1.6miESE

Max Individual (mrem) (mrem) (mrem) (mrem)

Whole Body 5.05E-04 @09miE 7.71E-03 00.3miNE 4.64E-04 @0.5rmiNE 3.14E-04 I.0 miESEThyroid 4.48E-06 0 1.6miE(c) 3.74E-03 0 1.8miN(i) 4.16E-03 0 1.8miN(i) 3.42E-03 @2.0miENE(i)

Skin 5.93E-04 @0.9rmiE 3.39E-03 @3.5miWNW 5.69E-04 @0.5miNE 3.92E-04 C.0 miESE

Population (person-rem) (person-rem) (person-rem) (person-rem)

Whole Body 3.06E-03 3.04E-01 1.60E-02 6.1OE-03Thyroid 3.32E-03 3.28E-01 3.04E-02 1.66E-02Skin 3.75E-03 6.82E-01 3.25E-02 1.24E-02

Avg Individual (mrem) (mrem) (mrem) (mrem)

Whole Body 1.05E-06 1.04E-04 5.48E-06 2.09E-06Thyroid 1.14E-06 1.12E-04 1.04E-05 5.68E-06Skin 1.28E-06 2.34E-04 1.11E-05 4.25E-06

Unit 2 1st Quarter 2nd Quarter 3rd Quarter 4th QuarterMax Air (mrad) (mrad) (mrad) (mrad)

Beta 8.16E-08 0 1.6rmiENE 0.00E+00 1.08E-03 @ 0.4miNE 9.55E-08 0 7.7miiSSE

Gamma 7.20E-10 @ 1.6miENE 0.00E-f-00 1.85E-03 @0.4 miNE 3.95E-08 @7.7miSSE

Max Individual (mrem) (mrem) (mrem) (mtero)

Whole Body 4.72E-10 @ 1.6miENE 4.33E-08 @0.4miNE 1.23E-03 @0.4rmiNE 2.42E-08 07.7miSSE

Thyroid 1.39E-10 0 1.6 miENE(t) 1.05E-04 @2.0rniENE(i) 4.87E-04 @0.6miiNE(c) 1.15E-05 @2.0miiENE(i)

Skin 5.66E-08 @ 1.6miENE 5.26E-08 @0.4 miNE 2.25E-03 @0.4 miNE 6.85E-08 07.7miSSEPopulation (person-rem) (person-rem) (person-rem) (person-rem)

Whole Body 5.02E-09 3.01 E-05 3.75E-03 8.35E-08Thyroid 5.02E-09 2.13E-04 3.88E-03 3.35E-05Skin 6.01 E-07 2.98E-05 6.90E-03 7.36E-08

Avg Individual (inrem) (intern) (mtero) I (tero)

Whole Body 1.72E-12 1.03E-08 1.28E-06 2.86E-11Thyroid 1.72E-12 7.29E-08 1.33E-06 1.15E-08Skin 2.06 E- 10 1.02E-08 2.36E-06 2.52E- 11

Unit 3 1st Quarter 2nd Quarter 3rd Quarter 4th QuarterMax Air (mrad) (mrad) (mrad) (mrad)

Beta 4.16E-05 @0.4miENE 1.09E-02 00.4miENE 0.OOE+00 2.96E-03 @0.4miENE

Gamma 1.40E-05 @ 04 mi ENE 3.68E-03 @ 0.4 mi ENE 0.OOE+00 9.97E-04 @ 0.4 mi ENE

Max Individual (mrem) (mrem) (mrem) (mrem)

Whole Body 3.55E-05 @0-4rmiENE 3.17E-03 00.4miENE 6.08E-06 00.4miENE 5.84E-04 00.4miENE

Thyroid 3.37E-04 @ 0.4 mi ENE (c) 1.04E+00 0 2.0 mi ENE i) 6.29E-03 @ 0.5 miNE (c) 3.90E-03 a 2.0 mi ENE (i)

Skin 5.53E-05 0 0.4 mi ENE 7.28E-03 @ 0.4 mi ENE 7.38E-06 0 0.4 mi ENE 1.64E-03 0 0.4 mi ENE


Whole Body 3.73E-05 8.56E-03 2.25E-05 1.61 E-03Thyroid 9.51 E-04 1.79E+00 1.10E-02 1.24E-02Skin 9.75E-05 1.75E-02 3.53E-06 4.93E-03

Avg Individual (mrem) (itoero) (mrem) (mrem)

Whole Body 1.28E-08 2.93E-06 7.71 E-09 5.51 E-07Thyroid 3.26E-07 6.13E-04 3.77E-06 4.25E-06Skin 3.34E-08 5.99E-06 1.21 E-09 1.69E-06

(a)=Adult, (c)=Child, (i)=Infant, (t)=Teen

Table 1-2

1995 Off-Site Dose Commitments from Liquid EffluentsMillstone Units 1, 2, 3

Unit 1 1st Quarter 2nd Quarter 3rd Quarter 4th QuarterMax Individual (mrem) (mrem) (mrem) (mrern)

Whole Body 5.06E-03 (c) 5.24E-03 (c) 1.33E-02 (c) 1.09E-02 (c)Thyroid 1.16E-05 (a) 7.14E-06 (a) 1.32E-05 (a) 2.75E-05 (a)Max Organ 1.01 E-02 (a) (H) 1.05E-02 (a) (h) 2.66E-02 (a) (1i) 2.1 8E-02 (a) (1i)


Whole Body 5.63E-03 5.26E-03 1.34E-02 1.17E-02Thyroid 1.33E-04 1.02E-04 1.74E-04 2.95E-04Max Organ 1.14E-02 (1i) 1.07E-02 (00 2.79E-02 (1i) 2.35E-02 (0)


Whole Body 1.93E-06 1.80E-06 4.59E-06 4.01 E-06Thyroid 4.55E-08 3.49E-08 5.96E-08 1.01 E-07Max Organ 3.90E-06 (w 3.66E-06 (0i) 9.55E-06 (i) 8.05E-06 (0I)

Unit 2 1st Quarter 2nd Quarter 3rd Quarter 4th QuarterMax Individual (mrem) (mrem) (mrem) (mrem)

Whole Body 2.55E-04 (a) 1.82E-03 (c) 1.49E-03 (t) 3.64E-04 (a)

Thyroid 9.75E-05 (a) 3.19E-04 (a) 4.69E-04 (a) 1.74E-04 (a)Max Organ 1.33E-03 (a) (w) 9.68E-03 (a) (gi) 9.39E-03 (a) (gi) 3.36E-03 (a) (gi)


Whole Body 5.41 E-03 2.22E-02 2.04E-02 8.01 E-03Thyroid 9.14E-04 2.99E-03 4.47E-03 3.10E-03Max Organ 2.30E-02 (w) 1.68E-01 (gi) 2.19E-01 (gi) 6.69E-02 (w)


Whole Body 1.85E-06 7.60E-06 6.99E-06 2.74E-06Thyroid 3.13E-07 1.02E-06 1.53E-06 1.06E-06Max Organ 7.88E-06 (gi) 5.75E-05 (gi) 7.50E-05 (gi) 2.29E-05 (91)

Unit 3 Ist Quarter 2nd Quarter 3rd Quarter 4th QuarterMax Individual (mrem) (mrem) (mrem) (mrem)

Whole Body 8.26E-04 (a) 4.97E-03 (a) 1.OOE-03 (a) 6.57E-03 (c)Thyroid 2.38E-04 (a) 3.14E-03 (a) 3.41 E-04 (a) 4.49E-04 (a)Max Organ 1.47E-02 (a) (gi) 3.34E-02 (a) (gi) 1.68E-02 (a) (gi) 4.OOE-02 (a) (gi)


Whole Body 9.95E-03 8.59E-02 2.43E-02 8.66E-02Thyroid 4.64E-03 5.25E-02 3.40E-03 7.OOE-03Max Organ 1.90E-01 (gi) 8.47E-01 (00) 3.33E-01 (gi) 8.72E-01 (9')


Whole Body 3.41 E-06 2.94E-05 8.32E-06 2.97E-05Thyroid 1.59E-06 1.80E-05 1.16E-06 2.40E-06Max Organ 6.51E-05 (90 2.90E-04 (gi) 1.14E-04 (gi) 2.99E-04 (gi)

(a)=Adult, (c)=Child, (i)=Infant, (t)=Teen (a)=du~t (c=Chid, i)=(~an( (t=Tee (b)=Bone; (gi)=GI-LL1, (ki)=Kidney, (Ii).Liver, (1u)=Lung, (tt,)=Thyroid

Table 1-3

1995 Off-Site Dose SummaryMillstone Units 1, 2, 3

Population Dose Committments (person-rem)

Airborne I Uquid

Whole Body I Thyroid I Skin I Whole Body I Thyroid I Max Organ

Unit 1 0.3292 0.3783 0.7307 0.0360 0.0007 0.0735

Unit 2 0.0038 0.0041 0.0069 0.0560 0.0115 0.4769

Unit 3 0.0102 1.8144 0.0225 0.2068 0.0675 2.2420

Station 0.3432 2.1968 0.7601 0.2988 0.0797 2.7924

Max Individual Doses vs Limits

Whole Body Thyroid Max Organ Skin Air

Airborne Liquid Airborne ILquid Liquid Airborne Beta Gamma

(mrem) (mrem) (mrem) I (mrem) (mrad)

Uni Lirnit 5 I 3 15 j 10 10 15 20 10

Unit 1 Actual 0.0090 0.0345 0.0113 0.0001 0.0690 0.0049 0.0017 0.0113

Unit 2 Actual 0.0012 1 0.0039 0.0006 j 0.0011 0.0238 0.0023 0.0011 0.0019

Unit 3 Actual 0.0038 0.0134 1.0505 0.0042 0.1049 0.0090 0.0139 0.0047

Station Limit **

Station Actual

25

0.0658

75

1.0677

25

0.1977

Connecticut ResidentAverage Whole Body Doses (mrem)

Cosmic . 27

Cosmogenic 1

Terrestial (Atlantic and Gulf Coastal Plain) 16

Inhaled 200

In the Body 40

Average CT Resident Whole Body Dose from Background 284

Average CT Resident (within 50 miles) Whole Body Dose from

Millstone Station Radioactive Effluents 0.0002

I10CFRSO Appendix I

40CFR190

NCRP94

2.0 Radioactivity

2.1 Airborne Effluents

2.1.1 Measurement of Radioactivity

2.1.1.1 Unit 1 Stack

2.1.1.1.1 Fission and Activation Gases

Stack monitors continuously record the effluent activity and flowrate. Weekly grab samples are taken from the stack andanalyzed for isotopic content. When the recombiner is notoperable, sampling frequency is increased. The isotopicconcentrations at the release point are multiplied by the totalstack flow to obtain the total activity released for each isotope.

Tritium collection is accomplished by the gas washing bottlemethod. The sample is counted on a liquid scintillation detector.Concentration is computed using the worst possible case, 100%humidity. Concentration is multiplied by volume to give the totalactivity released.

2.1.1.1.2 lodines and Particulates

Charcoal cartridges and particulate filters are used to collectiodines and particulates, respectively. These filters are thenanalyzed for isotopic content using a gamma spectrometer.Particulate filters are also analyzed for Sr-89, Sr-90 and grossalpha. Isotopic concentrations are multiplied by the release flowrate and sampling time to determine the total amount of activityreleased.

2.1.1.2 Unit 2 Vent

Total effluent volume from the Unit 2 vent per month is multiplied by theisotopic concentrations as measured by gamma spectrometer HPGe analysisfor gases and liquid scintillation detector for tritium to obtain the total activityreleased from the vent.

Charcoal cartridges and particulate filters are used to collect iodines andparticulates, respectively. These filters are then analyzed for isotopic contentusing a gamma spectrometer. Particulate filters are also analyzed for Sr-89,Sr-90 and gross alpha. Isotopic concentrations are multiplied by the releaseflow rate and sampling time to determine the total amount of activity released.

Tritium collection is accomplished by the gas washing bottle method. Thesample is counted on a liquid scintillation detector. Concentration iscomputed using the worst possible case, 100% humidity. Concentration ismultiplied by volume to give the total activity released.

2.1.1.3 Unit 2 Containment Purges

Grab samples are taken for gaseous, particulates, and iodine. These are

analyzed on a HPGe gamma spectrometer and liquid scintillation detector fortritium. Computed concentrations are then multiplied by the purge volume forthe total activity released.

Tritium collection is accomplished by the gas washing bottle method. Thesample is counted on a liquid scintillation detector. Concentration iscomputed using the worst possible case, 100% humidity. Concentration ismultiplied by volume purged to give the total activity released.

2.1.1.4 Unit 2 Steam Generator Blowdown Tank Vent

A decontamination factor (DF) across the steam generator blowdown tankvent has been determined for iodines by comparison of the results of gammaspectrometry, HPGe, analysis of steam generator blowdown water and grabsamples of condensed steam exiting the blowdown tank vent. This DF wasthen applied to the total iodine releases via the steam generator blowdownwater to determine the iodine releases out of the blowdown tank vent. Anadditional factor of 0.33 was utilized to account for the fraction of blowdownvolume actually flashing to steam in the blowdown tank.

2.1.1.5 Unit 3 Vent and ESF Building Vent

The Unit 3 ventilation vent collects gas streams from the auxiliary, fuel, wastedisposal, and service building exhausts, containment purge, and gaseouswaste process vent. The Unit 3 Engineered Safety Features (ESF) buildingvent collects gas streams from the ESF building ventilation system. This ventis located on the south wall and discharges 23 feet above grade. Totaleffluent volume per month is multiplied by isotopic concentrations from theanalysis of grab samples and composites to obtain the total activity released.These samples are obtained monthly for fission gas, weekly composites offilters for iodines and particulates, monthly composites of particulate filters forgross alpha and strontium.

2.1.1.6 Unit 3 Containment Drawdown and Purge

Unit 3 containment is drawn down and purged intermittently. The initialdrawdown is accomplished by using the containment vacuum steam jetejector and releases through an unmonitored vent on the roof of the auxiliarybuilding. The containment vacuum pump discharge, which maintainssubatmospheric pressure following initial drawdown, is released through theUnit 1 stack. The purge is the process of discharging air from containmentto maintain temperature, humidity, pressure, concentration, etc., where air isreplaced. Purges are filtered and normally released through the Unit 3 ventbut may use the Unit 1 stack. Purges and drawdowns are intermittent andare therefore considered batch releases. Calculated volume discharged ismultiplied by isotopic concentrations from the analysis of grab samples toobtain total activity released.

2.1.2 Estimate of Errors

Estimates of errors associated with radioactivity measurements were made using thefollowing guidelines:

Sampling/Data Collection 10% Variation in data collectionCalibration 10% Calibration to NBS standardsSample Counting 10% Maximum error for counting statisticsFlow & Level Measurements 10% Maximum error for release volumes

2.1.3 Batch Releases

Unit 2 Purge WGDT Summary

Number of Batches 4 4 8Total Time (min) 1,215 1,390 2,605Maximum Time (min) 604 590 604Average Time (min) 304 348 326Minimum Time (min) 54 145 54

Unit 3 Purge Drawdown Summary

Number of Batches 3 2 5Total Time (min) 581 123 704Maximum Time (min) 217 84 217Average Time (min) 194 62 141Minimum Time (min) 179 39 39

2.1.4 Abnormal Releases

None.

2.2 Liquid Effluents

2.2.1 Measurement of Radioactivity

2.2.1.1 Liquid Tanks

There are numerous tanks which are used to discharge liquids containingradioactivity to the environs; they are:

Unit 1 Decontamination Solution TankFloor Drain Sample Tanks (2)Waste Sample Tanks (2)

Unit 2 Clean Waste Monitor Tanks (2)Aerated Waste Monitor Tank

Unit 3 High Level Waste Test Tanks (2)Low Level Waste Tanks (2)

Prior to release, a tank is recirculated for two equivalent tank volumes, asample is drawn and analyzed on the HPGe gamma spectrometer and liquidscintillation detector for individual radionuclide composition. Isotopicconcentrations are multiplied by the volume released to obtain the totalactivity released. A proportional aliquot of each discharge is retained forcomposite analysis for Sr-89, Sr-90, Fe-55 and gross alpha.

2.2.1.2 Unit 2 and Unit 3 Steam Generator Blowdown

Steam generator blowdown water grab samples are taken and analyzed onthe HPGe gamma spectrometer and liquid scintillation detector. Total volumeof blowdown is multiplied by the isotopic concentrations to determine the totalactivity released via blowdown. A proportional aliquot of each discharge is

retained for composite analysis for Sr-89, Sr-90, Fe-55 and gross alpha.Tritium is determined through liquid scintillation counting; and, strontiums areanalyzed by radiochemical separations and appropriate counting techniques.

Unit 3 steam generator blowdown is typically recycled and therefore noradioactivity is discharged.

2.2.2 Estimate of Errors

Estimates of errors associated with radioactivity measurements were made using thefollowing guidelines:

Sampling/Data CollectionCalibrationSample CountingFlow & Level Measurements

10%10%10%10%

Variation in data collectionCalibration to NBS standardsMaximum error for counting statisticsMaximum error for release volumes

2.2.3 Batch Releases

Unit 1 Unit 2 Unit 3

Number of BatchesTotal Time (min)Maximum Time (min)Average Time (min)Minimum Time (min)Average Stream Flow

230 185 52718,395 16,859 44,588

195 476 94580 91 85

9 1 1Not Applicable - Ocean Site

2.2.4 Abnormal Releases

None.

Effluent Release Tables

Table 2.1-1 Unit 1 Airborne Effluents, Release SummaryTable 2.1-2 Unit 1 Airborne Effluents, Elevated ContinuousTable 2.1-3 Unit 1 Liquid Effluents, Release SummaryTable 2.1-4 Unit 1 Liquid Effluents, Batch

Table 2.2-1 Unit 2 Airborne Effluents, Release SummaryTable 2.2-2 Unit 2 Airborne Effluents, Mixed Continuous-Vent & SGBD Tank VentTable 2.2-3 Unit 2 Airborne Effluents, Mixed Batch-Containment PurgesTable 2.2-4 Unit 2 Airborne Effluents, Elevated Batch-WGDTTable 2.2-5 Unit 2 Liquid Effluents, Release SummaryTable 2.2-6 Unit 2 Liquid Effluents, Continuous-SGBDTable 2.2-7 Unit 2 Liquid Effluents, Batch

Table 2.3-1 Unit 3 Airborne Effluents, Release SummaryTable 2.3-2 Unit 3 Airborne Effluents, Mixed Continuous-Normal VentilationTable 2.3-3 Unit 3 Airborne Effluents, Mixed Continuous-ESF Building VentilationTable 2.3-4 Unit 3 Airborne Effluents, Mixed Batch-Containment DrawdownTable 2.3-5 Unit 3 Airborne Effluents, Mixed Batch-Containment PurgesTable 2.3-6 Unit 3 Liquid Effluents, Release SummaryTable 2.3-7 Unit 3 Liquid Effluents, Continuous-SGBDTable 2.3-8 Unit 3 Liquid Effluents, Batch-LWSTable 2.3-9 Unit 3 Liquid Effluents, Batch-CPF WN Sumps

Table 2.1-1Millstone Unit No. 1

Airborne Effluents - Release Summary

I 1 1°° 9 IUnits 11st Otr f 2nd Otr I 3rd Qtr I 4th Qtr I Total

D. Gross AlphaI1. GReleased Ci 1.04E-06 I 5.21E-07 1.44E-06 9.23E-07 3.93E-06

N/D = Not Detected


Airborne Effluents - Elevated Continuous

iNuclides 1 1 9oo9 5Released I Units 1 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total

A. Fission & Activation GasesAr-41 Ci 7.39E-01 8.16E-02 - 8.21E-01Kr-85m Ci 1.21 E+01 7.83E-01 4.57E-01 1 .33E+01Kr-87 Ci 3.51 E-02 4.22E+01 2.84E+00 1.15E+00 4.62E+01Kr-88 Ci 3.62E+01 1.59E+00 1.04E+00 3.88E+01Xe-131m Ci 5.89E-01 - 5.89E-01Xe-133 Ci 5.81E-01 1.12E+01 4.35E+00 7.41E+00 2.36E+01Xe-1 35 Ci 5.80E-02 5.29E+01 2.96E+00 3.52E+00 5.94E+01Xe-135m Ci 3.49E+01 9.92E-01 1.47E+00 3.73E+01Xe-138 Ci 1.79E-02 1.30E+02 3.55E+00 2.45E+00 1.36E+02Total Activity Ci 1.28E+00 3.20E+02 1.71 E+01 1.75E+01 3.56E+02

B. lodines1-131 Ci 6.37E-05 8.26E-04 3.82E-04 2.45E-04 1.52E-031-133 Ci 3.42E-04 1.56E-03 2.58E-03 1.60E-03 6.08E-03Total Activity Ci 4.05E-04 2.38E-03 2.96E-03 1.84E-03 7.59E-03

C. Particulates1-131 Ci 2.54E-06 - 2.06E-06 - 4.60E-06Ag-1im Ci - - 3.66E-05 9.41E-06 4.60E-05Ba-1 40 Ci 3.36E-06 1.46E-04 3.47E-05 - 1.84E-04Co-58 Ci 3.06E-06 2.54E-06 1.05E-05 4.57E-06 2.07E-05Co-60 Ci 1.66E-04 1.55E-04 9.17E-05 4.69E-05 4.60E-04Cr-51 Ci 2.25E-03 1.61 E-03 3.OOE-03 1.53E-03 8.40E-03Cs-137 Ci 1.21E-04 9.04E-05 2.20E-04 3.13E-05 4.62E-04Fe-59 Ci - 2.51E-05 2.51 E-05Mn-54 Ci 1.52E-05 2.62E-05 6.19E-05 2.19E-05 1.25E-04Sr-89 Ci 2.29E-05 9.09E-05 3.82E-05 9.02E-06 1.61 E-04Sr-90 Ci 1.22E-06 1.46E-06 8.26E-07 4.48E-07 3.96E-06Zn-65 Ci 8.91 E-05 2.07E-04 8.44E-04 3.57E-04 1.50E-03Total Activity Ci 2.68E-03 2.33E-03 4.37E-03 2.01E-03 1.14E-02

D. Gross AlphalGross Alpha Ci I 1.04E-06 1 5.21 E-07 I 1.44E-06 I 9.23E-07 I 3.93E-06

E. TritiumH-3 Ci I 2.93E-01 2.93E-01

N/D = Not Detected


Liquid Effluents - Release Summary

i 1 1 Q n 5.Units 11st tr T 2nd Qtr I 3rd Qtr I 4th Qtr I Total

D. Gross Alpha1. Total Activity Ci N/D N/D 4.53E-05 N/D 4.53E-05

Released

E. Volume1. Released Waste Liters 1.85E+06 1.82E+06 2.14E+06 1.80E+06 7.61 E+06

Volume

2. Dilution Volume Liters 1.22E+10 6.29E+09 6.65E+09 3.41 E+09 2.85E+1 0During Releases_

3. Dilution Volume Liters 2.16E+11 2.12E+11 2.06E+11 8.99E+10 7.23E+11During Period I I I II_ I

N/D = Not Detected


Liquid Effluents - Batch

Nuclides I 1 1 99 5ReleasedI Units 1 1st Qtr [ 2nd Qtr I 3rd Qtr I 4th Qtr Tot~a~l:

A. Fission & Activation ProductsCo-57 Ci 2.88E-06 - - 2.88E-06Co-60 Ci 7.96E-04 5.26E-04 1.30E-03 2.85E-03 5.48E-03Cr-51 Ci 2.49E-04 1.84E-05 2.28E-03 1.32E-04 2.68E-03Cs-1 34 Ci 4.69E-05 - 4.69E-05Cs-1 37 Ci 2.25E-03 5.34E-04 9.78E-04 2.61 E-03 6.37E-03Fe-59 Ci - 2.52E-04 2.52E-04Hf-181 Ci 1.40E-05 1.40E-051-131 Ci 2.15E-05 2.15E-05Mn-54 Ci 1.33E-04 2.49E-05 1.03E-03 5.33E-04 1.72E-03Mo-99 Ci 2.85E-05 2.96E-05 2.26E-05 8.07E-05Na-24 Ci - 1.59E-04 6.20E-06 4.42E-05 2.09E-04Sb-125 Ci 7.53E-06 7.53E-06Sn-117m Ci - 2.73E-06 2.73E-06Sr-89 Ci - 4.02E-05 4.02E-05Sr-90 Ci - 5.40E-05 2.61 E-05 8.01 E-05Zn-65 Ci 1.33E-03 1.03E-03 3.66E-03 2.60E-03 8.62E-03Zn-69m Ci 2.31 E-06 2.31 E-06Total Activity Ci 4.84E-03 2.32E-03 9.68E-03 8.80E-03 2.56E-02

B. TritiumIH-3 Ci I 2.60E+00 I 2.84E+00 I 3.89E+00 I 3.75E+00 1.31E+01

C. Dissolved & Entrained GasesIXe-i 33 Ci I5.OOE-06 I - I1.83E-05 I - I2.33E:-:05Total Activity Ci 5.OOE-06 - 1.83E-05 2.33E-05

D. Gross AlphaIGross Alpha Ci - 4.53E-05 I I 4.53E-05

N/D = Not Detected


Airborne Effluents - Release Summary

Units 1 1st Qtr I 2nd Qtr I 3rd 1 tr I 4th 9tr Total

D. Gross Alpha

I1 Totl:Activi Ci 7.75E-06 2.19E-08 9.50E-09 N/D 7.78E-o6

N/D = Not Detected


Airborne Effluents - Mixed Continuous - Vent & SGBD Tank Vent

SNuclides I 1 1 9 9 5Released Units 1 1st Qtr 2nd Qtr 3rd Qtr I 4th Qtr I Total

A. Fission & Activation GasesAr-41 Ci 6.55E-01 - 6.55E-01Xe-1 33 Ci 2.48E-01 - 2.48E-01Xe-135 Ci 7.35E-01 - 7.35E-01Total Activity Ci 1.64E+o0 - 1.64E+00

B. lodines1-131 Ci 1.62E-06 1.14E-06 3.25E-07 3.08E-061-133 Ci 1.59E-06 4.11E-06 8.66E-07 6.56E-06Total Activity Ci 3.21 E-06 5.24E-06 1.19E-06 9.64E-06

C. Particulates1-131 Ci

Total Activity Ci - N/D

D. Gross AlphaIGrossAlpha I Ci I 7.75E-06 I 2.19E-08 I 9.50E-09 I I 7.78E&06

E. TritiumIH-3 I Ci - - 6.24E-01 - 6.24E-01

N/D = Not Detected


Airborne Effluents - Mixed Batch - Containment Purges

Nuclides I I 1995Released I Units 11st Otr I2nd Otr 3rd Qtr I 4th Qtr I Total I

A. Fission & Activation Gases

I Ci- I - - - - -

ITotal Activity C - I - N/D

B. lodines1-131 Ci - -

Ci - -

Total Activity Ci - - N/D

C. Particulates1-131 Ci-

CiTotal Activity Ci - N/D

D. Gross AlphaIGross Alpha I Ci I -" N/D

E. TritiumIH-3 I Ci - 2.03E-02 -" 2.03E-02

N/D = Not Detected


Airborne Effluents - Elevated Batch - WGDT

SNuclides 1 1 9 9 5Released IUnits 1 1st Qtr I 2nd Qtr I 3rdQtr 4th Qtr Total

A. Fission & Activation GasesAr-41 Ci - - 1.29E-04 1.29E-04Kr-85 Ci 3.74E-03 1.02E-01 2.35E-03 1.08E-01Kr-85m Ci - - - 9.68E-05 9.68E-05Kr-88 Ci - - - 7.43E-05 7.43E-05Xe-131m Ci - - - 1.03E-03 1.03E-03Xe-133 Ci - - - 1.05E-01 1.05E-01Xe-133m Ci - - - 1.67E-03 1.67E-03Xe-135 Ci - - 1.01E-02 1.01E-02Total Activity Ci 3.74E-03 1.02E-01 1.20E-01 2.26E-01

B. lodines1-131 Ci - _ _"Ci "To tal Ac t ivity Ci - - N/D

C. Particulates1-131 Ci -

Ci-

:Total Activity Ci - - N/D

D. Gross AlphaIGross Alpha I Ci I " -" N/D

E. TritiumIH-3 Ci I 9.77E-06 - 5.73E-03 1.40E-05 5. 75E-03

N/D = Not Detected



Units 11stQtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total

D. Gross Alpha1. Total Activity Ci N/D 4.18E-05 N/D N/D 4.18 E:59Released

E. Volume1. Released Waste Liters 3.83E+05 4.53E+05 1.34E+06 5.91EE+05 2.77E+06

Volume

2. Dilution Volume Liters 2.86E+09 3.1OE+09 5.08E+09 4.25E+09 1.53E+10During Releases

3. Dilution Volume Liters 1.95E+11 1.57E+11 2.44E+11 2.77E+11 8.72E+11During Period _I IIII

N/D = Not Detected


Liquid Effluents - Continuous - SGBD

<< No Activity Detected >>

i N u clide s I --! , 9 , 9 ]_ o aReleased I Units 1 st~tr I 2ndQtr 3rd Otr 4th Qtr Total

A. Fission & Activation Products

ITotal ActiviyII NIDB. TritiumH-3 I " i_ N/

C. Dissolved & Entrained GasesI I I - I -

ITotalctvy I -i -I NED]

D. Gross Alpha

IGross Alpha I Ci I " - " N/D

N/D = Not Detected


Liquid Effluents - Batch

INuclids 1 1 9° ° 5Released I Units 1 st Qtr I 2nd Qtr 3rd Qtr I 4th Qtr Total I

A. Fission & Activation ProductsAg- 1 Om Ci 1.57E-04 1.45E-03 1.21 E-03 5.56E-04 3.36E-03As-76 Ci 5.71 E-06 - 5.71 E-06Co-57 Ci 2.63E-04 2.87E-04 5.87E-04 1.03E-04 1.24E-03Co-58 Ci 2.72E-02 1.22E-02 1.98E-02 7.1OE-03 6.62E-02Co-60 Ci 1.51 E-02 3.82E-02 8.41E-02 2.07E-02 1.58E-01Cr-51 Ci 5.69E-04 4.72E-04 1.39E-03 2.43E-03Cs-134 Ci 6.34E-04 9.29E-04 1- .56E-03Cs-137 Ci 2.33E-03 3.87E-03 9.28E-04 2.11E-04 7.33E-03Fe-55 Ci 6.80E-03 3.87E-02 4.12E-02 6.73E-03 9.34E-02Fe-59 Ci 7.04E-04 7.04E-04La-140 Ci 3.35E-04 7.23E-04 1.06E-03Mn-54 Ci 2.25E-03 1.93E-03 5.79E-03 8.95E-04 1.09E-02Nb-95 Ci 6.68E-04 1.95E-03 2.87E-04 2.90E-03Ru-105 Ci 1.41E-04 1.41E-04Sb-122 Ci 3.31 E-05 - 3.31 E-05Sb-124 Ci 4.53E-05 1.55E-04 2.83E-06 2.03E-04Sb-125 Ci 5.1OE-04 1.09E-02 4.89E-03 5.17E-03 2.14E-02Sn-113 Ci 2.19E-04 3.77E-04 - 5.97E-04Sn-i 17m Ci 7.30E-06 5.90E-06 1.32E-05Sr-89 Ci 3.15E-05 2.18E-05 1.98E-06 5.53E-05Sr-90 Ci 1.13E-04 9.35E-04 5.32E-05 1.43E-05 1.12E-03Sr-91 Ci - 2.52E-05 - 2.52E-05Zn-65 Ci - 1.68E-04 1.25E-04 - 2.93E-04Zr-95 Ci - 2.26E-04 5.55E-04 1.86E-04 9.67E-04Zr-97 Ci - 2.21E-05 2.17E-05 - 4.38E-05Total Activity Ci 5.60E-02 1.11E-01 1.62E-01 4.50E-02 3.74E-01

B. Tritium-3 Ci 1 1.01E+00 1.06E+00 1.27E+01 i1.42E+02 i1.57E+02 ]

C. Dissolved & Entrained GasesKr-85 Ci I1.67E-06 3.18E-03 3.18E-03Kr-88 Ci 2.25E-05 - 2.25E-05Xe-1 33 Ci 2.40E-03 2.51 E-02 2.75E-02Xe-135 Ci 1.33E-04 3.14E-04 4.47E-04Total Activity Ci 1.67E-06 2.25E-05 5.71 E-03 2.55E-02 3.12E-02

D. Gross AlphaIGrossAlpha Ci " 4.18E-05 - - 4.18E-05

N/D = Not Detected


Airborne Effluents- Release Summary

I 1 9 9 51Units 1 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total


D. Gross Alpha

1. Total Activity Ci 2.51E-07 1.18E-08 7.76E-09 9.41E-08 3.64E-07Released

E. Tritium

N/D = Not Detected


Airborne Effluents - Mixed Continuous- Normal Ventilation

INuclides 1 1 199 5Released Units 1st Qtr [ 2nd Qtr I 3rd trI 4th Qtr I Total

A. Fission & Activation GasesXe-133 Ci 4.63E-01 I 4.87E+01 1 4.41E+01 I 9.33E+01 ITotal Activity Ci 4.63E-01 4.87E+01 4.41 E+01 9.33E+01

B. lodines1-131 Ci 2.06E-04 1.65E-02 1.OOE-04 1.09E-04 1.70E-021-133 Ci 2.20E-04 2.39E-03 5.29E-04 2.95E-04 3.43E-03Total Activity Ci 4.27E-04 1.89E-02 6.29E-04 4.04E-04 2.04E-02

C. Particulates1-131 CiCs-137 Ci 3.84E-06 3.84E-06Co-58 Ci 7.27E-05 - 7.27E-05Cr-51 Ci 3.56E-05 1.20E-05 4.76E-05Mn-54 Ci 4.15E-06 1.10E-05 1.51E-05Total Activity Ci 4.15E-06 1.23E-04 1.20E-05 1.39E-04

D. Gross AlphalGross Alpha I Ci I 2.33E-07 I - 7"OOE-08 1303E.07

E. TritiumJH-3 I Ci I - 1.68E-01 I 1.68E-01

N/D = Not Detected


Airborne Effluents - Mixed Continuous - ESF Building Ventilation

Nuclides I I 1 995 IReleased Units 1 1st Qtr f 2nd Qtr 3rd Qtr 4th Qtr I Total

A. Fission & Activation GasesIXe-133 I Ci 2.87E-02 I 4.99E-02 I 7.86E-02

ITotal Activity Ci 2.87E-02 4.99E-02 I 7.86E-02

B. lodines1-131 Ci - 5.15E-04 9.20E-07 5.16E-041-133 Ci - 2.75E-04 - 2.75E-04Total Activity Ci - 7.90E-04 9.20E-07 7.91 E-04


Ce-141 Ci 6.89E-08 5.1OE-08 1.20E-07Ce-144 Ci - 2.23E-07 - 2.23E-07Co-58 Ci 1.11 E-07 9.76E-07 9.03E-08 - 1.1 8E-06Cr-51 Ci 2.17E-07 - 2.17E-07Sb-125 Ci 1.97E-07 - 1.97E-07Sn-113 Ci - 5.37E-08 - 5.37E-08Total Activity Ci 5.94E-07 1.03E-06 3.13E-07 5.1OE-08 1.99E-06

D. Gross AlphalGross Alpha I Ci 1 1.81E-08 I 1.18E-08 I 7.76E-09 1 2.41E-08 I 6.17E'08

E. TritiumIH-3 I - - I 1.69E-03 I 1.69E-03

N/D = Not Detected


Airborne Effluents- Mixed Batch - Containment Drawdown

INuclides 1 1 9° 9 5Released IUnits 1 lstQtr 2nd Qtr I 3rd Qtr I 4th Qtr I Total


I i I-- -

TIotal ctivti I N/D I

B. lodines1-131 Ci 3.15E-08 - 3.15E-08

Ci-Total Activity Ci 3.15E-08 - 3.15E-


Ci -

Total Activity Ci - - N/D

D. Gross Alpha[Gross Alpha I Ci I - N " N/D

E. TritiumIH-3 I Ci 1.54E-04 - 7.86E-02 7.88E-02

N/D = Not Detected


Airborne Effluents - Mixed Batch - Containment Purges

Nuclides I I 1 9 9 5Released Units 1 1st tr 2nd Qtr I rd 4th Qtr I Total

A. Fission & Activation GasesKr-85 Ci 6.24E-02 - 6.24E-02Kr-85m Ci 1.58E-03 - 1.58E-03Xe-131m Ci 2.86E-02 - 2.86E-02Xe-133 Ci 3.15E+00 - 5.56E-02 3.21E+00Xe- 133m Ci 5.40E-02 - I 5.40E-02Xe-135 Ci 9.71E-02 2.36E-02 1.21E-01Total Activity Ci 3.39E+00 7.92E-02 3.47E+00

B. lodines1-131 Ci - 5.38E-04 5.38E-04

CiITotal Activity Ci 5.38E-04 5.38E-04

C. Particulates1-131 Ci

CiTotal Activity Ci N/D

D. Gross AlphaIGross Alpha I Ci I " - - N/D

E. TritiumIH-3 Ci - 1.32E-02 - 2.66E-01 2.79E-01

N/D = Not Detected



Units 1,1st tr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total

D. Gross Alpha1. Total Activity Ci N/D N/D 5.94E-05 N/D 5.94E05]

Released

E. Volume1. Released Waste Liters 3.50E+06 3.33E+06 3.44E+06 3.43E+06 1.37E+07

Volume

2. Dilution Volume Liters 1.71E+10 1.85E+10 2.03E+10 1.79E+10 7.37E+10During Releases

3. Dilution Volume Liters 4.51E+11 2.81E+11 4.70E+11 4.28E+11 1.63E+12During Period I I III_ I

N/D = Not Detected


Liquid Effluents- Continuous - SGBD

<< No Activity Detected>>

iNuclides 1 1 199 5Released Units 1st Qtr I 2nd Qtr 3rd I 4th Qtr I Total


I~tlActivity I[C I -I - N/ED

B. TritiumH-3 I -" N/D

C. Dissolved & Entrained GasesI ________I___ I I - - I

ITotal Actti c I - I - NID

D. Gross Alpha

IGross Alpha ICi I " " N/D

N/D = Not Detected


Liquid Effluents - Batch - LWS

Nuclides 1st 1 9 9 5Released IUnits• 1 1st Qtr J. 2nd Qtr 3rd Qtr J. 4th Otr Total

A. Fission & Activation ProductsAg- 1 O1m Ci 5.33E-03 2.92E-03 4.88E-03 6.83E-03 2.OOE-02As-76 Ci - 8.41 E-06 8.41 E-06Ba-1 39 Ci - 2.37E-05 - 2.37E-05Ba-141 Ci - 1.31E-05 1- .31E-05Be-7 Ci - 3.62E-04 - 3.62E-04Ce-141 Ci - 9.17E-06 - 9.17E-06Ce-144 Ci 1.71 E-05 - 8.83E-06 2.59E-05Co-57 Ci 2.23E-04 4.83E-04 1.47E-04 3.01E-04 1.15E-03Co-58 Ci 7.94E-03 1.87E-01 2.90E-02 8.36E-02 3.08E-01Co-60 Ci 2.78E-02 6.50E-02 5.01 E-02 5.34E-02 1.96E-01Cr-51 Ci 3.33E-02 7.69E-04 2.05E-03 3.61 E-02Cs-1 34 Ci 7.89E-05 3.01 E-02 4.56E-03 9.62E-05 3.49E-02Cs-136 Ci 1.94E-03 1.94E-03Cs-137 Ci 4.64E-04 3.14E-02 1.16E-02 1.01E-03 4.45E-02Fe-55 Ci 7.72E-03 4.22E-02 4.40E-02 2.76E-01 3.70E-01Fe-59 Ci 4.25E-05 4.96E-03 1.68E-04 2.17E-03 7.35E-03Hf-181 Ci 2.46E-05 2.46E-051-131 Ci 5.35E-05 2.12E-02 5.OSE-05 2.13E-021-133 Ci 1.81 E-04 1.88E-05 2.OOE-04La-140 Ci 1.50E-04 3.29E-05 1.83E-04La-142 Ci 5.82E-05 6.73E-06 1.95E-05 8.44E-05Mn-54 Ci 8.50E-03 4.10E-02 2.35E-02 2.02E-02 9.32E-02Mo-99 Ci 9.07E-07 - 9.07E-07Na-24 Ci - 9.71 E-06 9.71 E-06Nb-95 Ci 1.08E-04 5.73E-03 2.15E-03 6.23E-03 1.42E-02Nb-97 Ci 2.78E-04 5.46E-04 4.16E-04 2.89E-05 1.27E-03Re-188 Ci 8.14E-06 2.05E-05 2.86E-05Ru-105 Ci 2.25E-04 2.16E-04 1.71E-05 4.58E-04Ru-106 Ci - 2.48E-05 2.48E-05Sb-122 Ci 7.11E-05 7.02E-06 7.81E-05Sb-124 Ci 3.14E-03 4.69E-05 3.18E-03Sb-1 25 Ci 6.54E-03 9.04E-02 1.59E-02 1.32E-02 1.26E-01Sn- 113 Ci 3.57E-06 1.41 E-04 3.31E-04 5.21E-04 9.96E-04Sn-117m Ci 4.02E-06 4.72E-05 - 5.12E-05Sr-87m Ci - 9.95E-07 - 9.95E-07Sr-89 Ci 2.81E-04 2.50E-04 2.85E-04 8.16E-04Sr-92 Ci 2.61 E-05 9.99E-05 1.60E-05 4.20E-05 1.84E-04Tc-101 Ci 1.30E-05 - 5.66E-06 - 1.86E-05Tc-104 Ci 8.86E-06 8.86E-06Tc-99m Ci 9.84E-07 - 9.84E-07Y-91 m Ci 1.05E-05 1.05E-05Y-92 Ci 1.17E-05 1.17E-05


Liquid Effluents - Batch - LWS

SNuclides 1 9 1995Released Units 1St Qtr I 2nd Utr 3rd Otr I 4th Qtr Total

A. Fission & Activation Products (Continued)Zn-65 Ci 1.18E-04 4.15E-04 6.33E-04 1.17E-03Zr-95 Ci 6.23E-05 2.82E-03 6.42E-04 1.71 E-03 5.24E-03Zr-97 Ci - 3.15E-05 - 3.15E-05Total Activity Ci 6.54E-02 5.66E-01 1.89E-01 4.68E-01 1.29E+00

B. Tritium-3 ci I 2.53E+02 I 1.50E+02 2.66E+01 2.66E+02 6 96E*027

C. Dissolved & Entrained GasesAr-41 Ci - 4.04E-05 4.04E-05Kr-85 Ci 1.41 E-03 - 1.41 E-03Kr-85m Ci 1.09E-06 - 1.09E-06Kr-87 Ci - 1.04E-05 2.91E-06 1.33E-05Kr-88 Ci - 1.52E-05 1.52E-05Xe-131m Ci - 3.18E-05 3.18E-05Xe-1 33 Ci 3.02E-03 2.32E-03 5.08E-06 5.35E-03Xe-135 Ci 8.37E-04 8.30E-04 1.90E-05 3.60E-05 1.72E-03Xe-135m Ci 1.59E-04 - 1.59E-04Total Activity Ci 5.43E-03 3.19E-03 4.46E-05 7.58E-05 8.74E-03

D. Gross AlphaIGrossAlpha Cj I " 5.94E-05 5.9"94E-05

N/D = Not Detected


Liquid Effluents - Batch - CPF WN Sumps

<< No Activity Detected >>

N uclides 1 1 9°°9 5Released Units 1 st Qtr I 2nd Otr 3rdItr 4th Qtr I Total


ITotal ________y - N/IDý

B. TritiumIH-3 . -I - N/D

C. Dissolved & Entrained Gases

I~tlActivity Ci I - I- N/DýD. Gross Alpha

lGross Alpha I Ci I - N/D- N

N/D = Not Detected

2.3 Solid Waste

Solid waste shipment radioactivity summaries for each unit are given in the following tables:

Table 2.1-5 Unit 1 Solid Waste and Irradiated Component ShipmentsTable 2.2-8 Unit 2 Solid Waste and Irradiated Component ShipmentsTable 2.3-10 Unit 3 Solid Waste and Irradiated Component Shipments

The principal radionuclides in these tables were from shipping manifests.

Solidification Agent(s):No solidification on site for 1995

Containers routinely used for radioactive waste shipment include:55-gal Steel Drum DOT 17-H container 7.5 ft3

Steel Boxes 45 ft3

87 ft3

91 ft3

95 ft3

122 ft3

Steel Container 202.1 ft3

Steel "Sea Van" 1280 ft3

Polyethylene High Integrity Containers 120.3 ft3

132.4 ft3

173.4 ft3

202.1 ft3

Table 2.1-5

Solid Waste and Irradiated Component ShipmentsMillstone Unit 1

January 1, 1995 through December 31, 1995

1. Type of Waste

a. Spent resins, filter sludges, evaporator bottoms, etc.

Disposition Units Annual Totals [Est. Total{ AI Error. %

From Millstone Nuclear Power Station To Chem-Nuclear Services, Inc. - m3 4.79 E+01Bamwell, SC Ci 4.20 E+02 25%

From Millstone Nuclear Power Station To Scientific Ecology Group - Oak { m 3 4.82 E+00Ridge, TN[Ci 1.59 E+01 25%

b. Dry compressible waste, contaminated equipment, etc.

Dposition Units Annual Totals EstoTotalFrom Millstone Nuclear Power Station To Chem-Nuclear Services, Inc. - m3 1.47 E+01___EoBamwell, SC Ci 1.61 E+00 1 25% 7

From Millstone Nuclear Power Station To FW Hake & Associates - m3 3.63 E+01Memphis, TN Ci 2.21 E-02 2517r

From Millstone Nuclear Power Station To Scientific Ecology Group -Oak m 3 1.76 E+02Ridge, TN Ci 8.34 E-01 25%

From Scientific Ecology Group - Oak Ridge, TN To Chem-Nuclear Service m 3 1.31 E+01Inc. - Barnwell, SC Ci 6.28 E-01 25%

d. Other (contaminated used oil for incineration)

IDisposition Units Annual Totals Est. TotalI I I IError, %

From Millstone Nuclear Power Station To Scientific Ecology Group - Oak Lm 3 6.37 E-01Ridge, TN Ci 7.77 E-06 25%

Type of WasteMillstone Station, Unit I

Page 1.1

2. Estimate of major nuclide composition (by type of waste)


From Millstone Nuclear Power Station to Chem-Nuclear Services, Inc. - Barnwell, SC for Burial

% of TotalRadionuclide I (Estimate) I Curies

Ag-Il Om

Am-241

C- 14

Ce- 141

Cm-242

Cm-244

Co-58

Co-60

Cr-51

Cs- 134

Cs- 137

Fe-55

Fe-59

H-3

1-131

La- 140

Mn-54

Ni-63

Pu-238

Pu-239

Pu-241

Pu-242

Sr-89

Sr-90

Tc-99

Zn-65

0.22%

<0.01%

0.03%

<0.01%

<0.0 1%

<0.01%

0.03%

4.29%

0.19%

<0.01%

9.71%

9.83%

0.05%

<0.0 1%

<0.0 1%<0.01%

2.23%

0.28%

<0.01%

<0.0 1%

<0.01%

<0.01%

<0.0 1%

0.01%

<0.01%

73.11%

9.21 E-01

1.26 E-03

1.39 E-01

3.91 E-03

7.60 E-05

8.96 E-04

1.09 E-01

1.80 E+01

7.99 E-01

1.18 E-02

4.08 E+01

4.13 E+01

2.30 E-01

3.03 E-02

1.84 E-05

7.82 E-21

9.35 E+00

1.18 E+00

8.13 E-04

5.67 E-04

2.89 E-02

1.46 E-06

2.70 E-04

5.22 E-02

1.41 E-04

3.07 E+02

Total: 4.20 E+02 Ci

Estimates of Major Nuclide Composition (By Type of Waste)

Millstone Station, Unit IPage 2.1


From Millstone Nuclear Power Station to Scientific Ecology Group - Oak Ridge, TN for Super-Compaction,Incineration, etc.

% of TotalRadionuclide (Estimate) Curies

Am-241 <0.01% 8.36 E-04

C-14 <0.01% 1.37 E-04

Cm-242 <0.01% 6.90 E-06

Cm-244 <0.01% 8.32 E-04

Co-60 13.85% 2.20 E+00

Cs-137 0.04% 7.11 E-03

Fe-55 74.94% 1.19 E+01

H-3 7.18% 1.14 E+00

Mn-54 2.15% 3.42 E-01

Ni-63 1.18% 1.88 E-01

Pu-238 <0.01% 5.23 E-04

Pu-239 <0.01% 2.36 E-04

Pu-241 0.11% 1.75 E-02

Pu-242 <0.01% 1.87 E-06

Sr-90 <0.01% 8.90 E-04

Zn-65 0.52% 8.20 E-02

Total: 1.59 E+01 Ci

Estinwmes of Major Nuclide Composition (By Type of Waste)






C- 14 <0.01% 1.29 E-04

Co-60 8.80% 1.42 E-01

Cr-51 0.20% 3.23 E-03

Cs- 134 0.05% 8.23 E-04

Cs- 137 1.08% 1.73 E-02

Fe-55 79.12% 1.27 E+00

H-3 0.06% 9.82 E-04

Mn-54 4.12% 6.63 E-02

Ni-63 1.16% 1.87 E-02

Zn-65 5.40% 8.68 E-02

Total: 1.61 E+00 Ci




2. Estimate

of major nuclide

composition

(by type of waste)

From Millstone Nuclear Power Station to FW Hake & Associates - Memphis, TN for Decontamination


Co-60 8.64% 1.91 E-03

Cr-51 0.20% 4.50 E-05

Cs- 137 0.75% 1.66 E-04

Fe-55 79.57% 1.76 E-02

H-3 0.48% 1.07 E-04

Mn-54 4.19% 9.26 E-04

Ni-63 0.67% 1.48 E-04

Zn-65 5.49% 1.21 E-03

Total: 2.21 E-02 Ci

Estimates of Major Nuclide Composition (B'y Type of Waste)

Millstone Station. Unit IPage 2.4


2. Estimate of major

nuclide composition

(by type of waste)



Co-60 8.73% 7.28 E-02

Cr-51 0.20% 1.70 E-03

Cs-137 0.76% 6.36 E-03

Fe-55 79.88% 6.66 E-01

H-3 0.04% 3.27 E-04

Mn-54 4.21% 3.51 E-02

Ni-63 0.67% 5.57 E-03

Zn-65 5.51% 4.60 E-02

Total: 8.34 E-01 Ci




From Scientific Ecology Group - Oak Ridge, TN to Chem-Nuclear Services, Inc. - Barnwell, SC for Burial


Ag- II0m <0.01% 4.55 E-06

Be-7 0.51% 3.17 E-03

C-14 0.06% 3.53 E-04

Co-58 2.51% 1.58 E-02

Co-60 9.34% 5.87 E-02

Cr-51 0.16% 1.03 E-03

Cs- 134 0.54% 3.38 E-03

Cs-137 3.00% 1.88 E-02

Fe-55 71.26% 4.48 E-01

H-3 0.21% 1.29 E-03

Mn-54 3.80% 2.38 E-02

Ni-63 1.52% 9.55 E-03

Sb- 125 0.07% 4.10 E-04

Zn-65 7.03% 4.42 E-02

Total: 6.28 E-01 Ci







Co-60 0.06% 4.83 E-09

Cs- 134 0.11% 8.23 E-09

Cs- 137 0.70% 5.44 E-08

Fe-55 4.26% 3.31 E-07

H-3 94.33% 7.33 E-06

Ni-63 0.06% 4.38 E-09

Ra-226 0.18% 1.40 E-08

Sr-89 0.17% 1.31 E-08

Th-234 0.12% 9.31 E-09

Zn-65 0.02% 1.84 E-09

Total: 7.77 E-06 Ci


Millstone Station, Unit /Page 2.7

Solid Waste and Irradiated Component Shipments

Millstone Unit 1


3. Solid Waste Disposition

Number of Shipments

II

Mode of Transportation

Truck (Sole Use Vehicle)



+IL-F

Destination

Chem-Nuclear Services, Inc. - Barnwell, SC

FW Hake & Associates - Memphis, TN

Scientific Ecology Group - Oak Ridge, TN

II

3]1

Solid Waste Disposition


Table 2.2-8

Solid Waste and Irradiated Component ShipmentsMillstone Unit 2


1. Type of Waste


Disposition Units Annual Totals Est. TotalI I Error. %From Millstone Nuclear Power Station To Chem-Nuclear Services, Inc. - m 3 9.93 E+00Bamwell, SC Ci 3.69 E+02 25%

From Millstone Nuclear Power Station To Scientific Ecology Group - Oak m 3 4.83 E+00Ridge, TN Ci 4.63 E+00 25%


Disposition Units Annual Totals Est. TotalI JError%From Millstone Nuclear Power Station To Chem-Nuclear Services, Inc. - m3 7.87 E+00Barnwell, SC Ci 1.20 E-O1 25%

From Millstone Nuclear Power Station To Scientific Ecology Group -Oak m 3 2.63 E+02 1Ridge, TN Ci 5.86 E-01 ] 25%

From Scientific Ecology Group - Oak Ridge, TN To Chem-Nuclear Service m 3 1.73 E+01 TInc. - Barnwell, SC Ci 5.62 E-0l 1 25%

c. Irradiated components, control rods, etc.

Disposition Units Annual Totals Est. Total1__Error, %

From Millstone Nuclear Power Station To Scientific Ecology Group - Oak m3 3.68 E-02Ridge, TN Ci 7.04 E-02 25%


Disposition Units Annual Totals Est. TotalError, %I

From Millstone Nuclear Power Station To Scientific Ecology Group - Oak m3 4.25 E-01Ridge, TN Ci 7.43 E-06 25% ]

Type of WasteMillstone Station, Unit 2

Page 1.)





Ag-110m

Am-241

Ba-140

C- 14

Cm-242

Cm-244

Co-57

Co-58

Co-60

Cr-51

Cs- 134

Cs- 137

Fe-55

Fe-59

H-3

1-129

Mn-54

Nb-95

Ni-63

Pu-238

Pu-239

Pu-241

Pu-242

Ru- 103

Sb- 124

Sb- 125

Sr-89

Sr-90

Tc-99

Zr-95

<0.01%

<0.01%

<0.01%

0.52%

<0.01%

<0.01%

0.34%

3.55%

18.20%

<0.01%

4.64%

8.88%

19.54%

<0.01%

<0.01%

<0.01%

2.79%

<0.01%

4 1.26%

<0.01%

<0.01%

0.01%

<0.01%

<0.01%

<0.01%

0.20%

<0.01%

0.05%

<0.01%

<0.01%

2.23 E-04

5.71 E-04

2.77 E-14

1.90 E+00

6.60 E-04

1.53 E-03

1.24 E+00

1.31 E+01

6.72 E+01

7.84 E-06

1.71 E+01

3.27 E+01

7.21 E+01

3.98 E-05

1.44 E-02

1.77 E-03

1.03 E+01

2.63 E-05

1.52 E+02

9.67 E-04

5.00 E-04

5.47 E-02

9.14 E-07

4.52 E-06

7.13 E-08

7.29 E-01

8.31 E-03

1.94 E-01

1.22 E-02

3.74 E-03

Total: 3.69 E+02 Ci


Millstone Station, Unit 2Page 2.1


2. Estimate

of major nuclide

composition

(by type of waste)


% of Total

Radionuclide (Estimate) Curies

Ag-110m <0.01% 2.10 E-05

Am-241 0.01% 5.10 E-04

C-14 3.45% 1.60 E-01

Cm-242 <0.01% 3.56 E-06

Cm-244 <0.01% 2.37 E-04

Co-57 <0.01% 7.00 E-06

Co-58 0.02% 9.14 E-04

Co-60 22.31% 1.03 E+00

Cr-51 <0.01% 1.68 E-04

Cs- 134 0.40% 1.86 E-02

Cs- 137 4.38% 2.03 E-01

Fe-55 8.78% 4.06 E-01

Fe-59 <0.01% 2.70 E-05

H-3 <0.01% 2.16 E-05

Mn-54 <0.01% 3.80 E-05

Nb-95 <0.01% 2.30 E-05

Ni-63 60.30% 2.79 E+00

Pu-238 <0.01% 1.52 E-04

Pu-239 <0.01% 2.48 E-04

Pu-241 0.29% 1.36 E-02

Pu-242 <0.01% 2.00 E-09

Sb- 125 <0.01% 1.30 E-05

Sr-90 0.02% 1.11 E-03

Zr-95 <0.01% 1.00 E-05

Total: 4.63 E+00 Ci







C-14 0.72% 8.59 E-04

Co-58 1.37% 1.64 E-03

Co-60 18.88% 2.26 E-02

Cs-134 5.03% 6.03 E-03

Cs-137 27.48% 3.29 E-02

Fe-55 34.84% 4.18 E-02

H-3 0.82% 9.88 E-04

Mn-54 0.55% 6.55 E-04

Ni-63 10.05% 1.20 E-02

Zn-65 0.27% 3.20 E-04

Total: 1.20 E-01 Ci


Millstone Station. Unit 2Page 2.3




C-14 1.16% 6.81 E-03

Co-58 2.87% 1.68 E-02

Co-60 19.24% 1.13 E-01

Cs-134 6.68% 3.91 E-02

Cs- 137 32.70% 1.91 E-01

Fe-55 24.94% 1.46 E-01

H-3 0.04% 2.30 E-04

Mn-54 0.59% 3.48 E-03

Ni-63 11.78% 6.90 E-02

Tc-99 <0.01% 3.00 E- 1I

Total: 5.86 E-01 Ci

Estimates of Major Nuclide Composition (Byv Type of Waste)





Ag-110m

Am-241

Be-7

C-14

Cm-242

Cm-244

Co-57

Co-58

Co-60

Cr-51

Cs- 134

Cs- 137

Fe-55

Fe-59

H-3

Mn-54

Nb-95

Ni-63

Pu-238

Pu-239

Pu-241

Sb- 125

Sr-89

Sr-90

Zn-65

Zr-95

<0.01%

<0.0 1%

0.04%

0.85%

<0.01%

<0.0 1%

<0.0 1%

8.78%

15.45%

0.01%

4.93%

23.46%

35.69%

<0.01%

0.12%

1.20%

<0.0 1%

8.67%

<0.01%

<0.01%

<0.01%

0.22%

<0.01%

<0.01%

0.55%

<0.01%

2.42 E-05

1.12 E-07

2.12 E-04

4.77 E-03

4.20 E-07

2.24 E-07

8.96 E-06

4.93 E-02

8.68 E-02

7.62 E-05

2.77 E-02

1.32 E-01

2.00 E-01

1.89 E-06

6.56 E-04

6.76 E-03

8.61 E-06

4.87 E-02

2.24 E-07

1.54 E-07

2.07 E-05

1.24 E-03

6.30 E-07

2.73 E-06

3.10 E-03

7.00 E-07

Total: 5.62 E-01 Ci



2. Estimate of maior nuclide composition (by type of waste)

c. Irradiated components, control rods, etc.



Co-58 <0.01% 1.00 E-06

Co-60 0.05% 3.52 E-05

Fe-55 95.87% 6.75 E-02

Mn-54 3.05% 2.15 E-03

Nb-94 <0.01% 1.99 E-10

Ni-59 <0.01% 6.06 E-06

Ni-63 1.02% 7.18 E-04

Tc-99 <0.01% 5.67 E-09

7.04 E-02 CiTotal:



'2. Estimate of major nuclide composition (by type of waste)




Co-60 0.07% 4.83 E-09

Cs- 134 0.11% 8.23 E-09Cs- 137 0.69% 5.10 E-08

H-3 98.65% 7.33 E-06

Ra-226 0.19% 1.40 E-08

Sr-89 0.18% 1.31 E-08

ITh-234 0.12% 8.97 E-09

Total: 7.43 E-06 Ci




Millstone Unit 2



Number of Shipments Mode of Transportation Destination

2 Truck (Sole Use Vehicle) Chem-Nuclear Services, Inc. - Barnwell, SC

6 Truck (Sole Use Vehicle) Scientific Ecology Group - Oak Ridge, TN



Table 2.3-10


Millstone Unit 3


1. Type of Waste


Disposition Units Annual Totals Est. Total

I I _Error, %From Millstone Nuclear Power Station To Chem-Nuclear Services, Inc. - m3 6.85 E+00Bamwell, SC Ci 4.24 E+01 25%

From Millstone Nuclear Power Station To Scientific Ecology Group - Oak m 3 1.07 E+01 TRidge, TN Ci 5.10 E+00 25%


Disposition Units Annual Totals Est. Total

I I_ Error, %From Millstone Nuclear Power Station To Chem-Nuclear Services, Inc. - m 3 5.76 E-01Barnwell, SC Ci 1.67 E-03 25%

From Millstone Nuclear Power Station To Scientific Ecology Group - Oak m3 2.83 E+02Ridge, TN Ci 6.49 E-01 25%

From Scientific Ecology Group - Oak Ridge, TN To Chem-Nuclear Service m 3 1.38 E+01Inc. - Barnwell, SC Ci 1.04 E+00 25%


Disposition Units Annual Totals Est. TotalError, %

From Millstone Nuclear Power Station To Scientific Ecology Group - Oak m3 2.76 E+00Ridge, TN Ci 7.69 E-06 25%

Type of WasteMillstone Station, Unit 3

Page 1.1





Ag-II Om

Am-241

C- 14

Ce- 144

Cm-242

Cm-244

Co-57

Co-58

Co-60

Cr-51

Cs-134

Cs- 137

Fe-55

Fe-59

H-3

Hf-I181

Mn-54

Nb-95

Ni-63

Pu-238

Pu-239

Pu-241

Sb- 124

Sb- 125

Sr-89

Sr-90

Tc-99

Zn-65

Zr-95

0.04%<0.01%

0.05%

0.13%

<0.01%

<0.01%

0.21%

2.60%

20.14%

<0.01%

0.67%

2.40%

34.26%

<0.01%

0.13%

<0.01%

4.88%

<0.01%

31.87%

<0.01%

<0.01%

0.05%

<0.01%

2.43%

<0.01%

0.02%

<0.01%

0.13%

<0.01%

1.56 E-02

1.08 E-04

2.08 E-02

5.33 E-02

8.86 E-05

3.87 E-04

8.92 E-02

1.10 E+00

8.53 E+00

2.84 E-10

2.82 E-01

1.02 E+00

1.45 E+01

2.30 E-07

5.30 E-02

4.93 E-08

2.07 E+00

3.90 E-08

1.35 E+01

3.69 E-04

1.37 E-04

2.29 E-02

3.33 E-07

1.03 E+00

1.89 E-04

6.68 E-03

2.53 E-05

5.51 E-02

1.11 E-04

Total: 4.24 E+01 Ci






Ag-1 10m <0.01% 3.21 E-04

Am-241 <0.01% 5.26 E-05

C-14 0.12% 5.99 E-03

Cm-242 <0.01% 1.98 E-05

Cm-244 <0.01% 6.99 E-05

Co-57 <0.01% 1.21 E-04

Co-58 0.44% 2.25 E-02

Co-60 19.37% 9.88 E-01

Cs-134 1.14% 5.82 E-02

Cs-137 6.61% 3.37 E-01

Fe-55 51.16% 2.61 E+00

H-3 0.13% 6.48 E-03

Mn-54 1.77% 9.05 E-02

Nb-95 0.05% 2.56 E-03

Ni-63 17.68% 9.02 E-01

Pu-238 <0.01 % 8.52 E-05

Pu-239 <0.01% 4.20 E-05

Pu-241 0.10% 4.92 E-03

Pu-242 <0.01% 1.33 E-08

Sb- 125 1.37% 6.97 E-02

Sr-89 <0.01% 9.00 E-06

Sr-90 0.05% 2.35 E-03

Zn-65 <0.01% 4.14 E-04

Total: 5.10 E+00 Ci






% of Total


C- 14 0.40% 6.74 E-06

Co-60 17.53% 2.92 E-04

Cs- 134 3.31% 5.53 E-05

Cs-137 22.22% 3.71 E-04

Fe-55 48.29% 8.05 E-04

H-3 1.33% 2.23 E-05

Mn-54 0.16% 2.67 E-06

Ni-63 6.48% 1.08 E-04

Zn-65 0.26% 4.39 E-06

Total: 1.67 E-03 Ci


Millstone Station. Unit 3Page 2.3



% of Total


Co-58 41.78% 2.71 E-01

Co-60 5.85% 3.80 E-02

Cs- 134 3.20% 2.08 E-02

Cs-137 9.50% 6.17 E-02

Fe-55 29.61% 1.92 E-01

H-3 0.09% 5.75 E-04

Mn-54 3.15% 2.05 E-02

Ni-63 5.75% 3.73 E-02

Sb- 125 1.07% 6.96 E-03

Total: 6.49 E-01 Ci






Ag-110Inm

Am-241

Be-7

C-14

Cm-242

Cm-244

Co-57

Co-58

Co-60

Cr-51

Cs- 134

Cs- 137

Fe-55

Fe-59

H-3

Mn-54

Nb-95

Ni-63

Pu-238

Pu-239

Pu-241

Sb- 125

Sr-89

Sr-90

Zn-65

Zr-95

0.03%

<0.01%

<0.01%

0.22%

<0.0 1%

<0.01%

0.01%

20.41%

10.53%.

0.02%

2.29%

8.10%

48.08%

<0.01%

0.07%

2.45%

0.01%

6.68%

<0.0 1%

<0.0 1%

0.03%

0.73%

<0.01%

<0.01%

0.31%

<0.01%

3.18 E-04

1.49 E-06

9.39 E-05

2.31 E-03

5.58 E-06

2.98 E-06

1.19 E-04

2.13 E-01

1.10 E-01

2.47 E-04

2.39 E-02

8.46 E-02

5.02 E-01

2.51 E-05

7.72 E-04

2.55 E-02

1.14 E-04

6.98 E-02

2.98 E-06

2.05 E-06

2.75 E-04

7.65 E-03

8.37 E-06

3.63 E-05

3.27 E-03

9.30 E-06

Total: 1.04 E+00 Ci







Co-60 0.06% 4.83 E-09

Cs-134 0.14% 1.08 E-08

Cs- 137 2.23% 1.72 E-07

Fe-55 0.40% 3. 10 E-08

H-3 95.36% 7.33 E-06

Ni-63 0.08% 6.02 E-09

Ra-226 1.09% 8.36 E-08

Sr-89 0.17% 1.31 E-08

Ta- 182 0.02% 1.83 E-09

Th-234 0.44% 3.41 E-08

Total: 7.69 E-06 Ci




Millstone Unit 3



Number of Shipments Mode of Transportation Destination

2 Truck (Sole Use Vehicle) Chem-Nuclear Services, Inc. - Barnwell, SC

7 Truck (Sole Use Vehicle) Scientific Ecology Group - Oak Ridge, TN



3.0 REMMIODCM Changes

In 1995, several changes were made to the REMM/ODCM; however, they are not bound with thisreport, but are included as a separate enclosure to the report submittal.

4.0 Inoperable Effluent Monitors

During the period January 1 through December 31, 1995, the following effluent monitor was inoperablefor more than 30 consecutive days:

4.1 Unit 2 Condensate Polishing Facility (CPF) Discharge Radiation Monitor (RM245)

The CPF Discharge Radiation Monitor was out of service from October 31, 1995 to December1, 1995 because the recorder failed. That particular monitor was no longer in production, so asubstitute device has temporarily replaced it until an engineered solution can be implemented.

5.0 Errata

In the 1993 Annual Radioactive Effluent Report, there is an error in Table 2.1-2 Millstone Unit No.1 Airborne Effluents - Elevated Continuous (enclosed), i.e. the annual Total for Xe-133 is missingand is not included in the Total Activity. However, Table 2.1-1 Millstone Unit No. 1 AirborneEffluents - Release Summary does not need to be corrected.


Airborne Effluents- Elevated Continuous

3rddtrQt4thQtr I TotalReleased Units 1 1st Qtr I 2nd I 3rd I 4th tr

# Sum formula missing. Total Activity only changedA. Fission & Activation GasesAr-41 Ci 1.16E+00 8.86E-02 2.38E-02 1.27E+00Kr-85 Ci 1.29E+01 1.29E+01Kr-85m Ci 7.28E+00 1.18E-01 5.29E-01 7.92E+00Kr-87 Ci 3.82E+01 6.37E-01 3.18E+00 4.20E+01Kr-88 Ci 2.50E+01 4.62E-01 2.04E+00 2.75E+01Xe-131 m Ci 2.22E+00 2.22E+00Xe-133 Ci 1.40E+00 2.12E-01 1.61E+00Xe-1 35 Ci 4.56E+01 8.75E-01 3.32E+00 4.98E+01Xe-1 35m Ci 3.48E+01 5.33E-01 2.85E+00 3.82E+01Xe-138 Ci 1.34E+02 2.29E+00 9.13E+00 1.45E+02Total Activity Ci 3.OOE+02 5.22E+00 2.38E-02 2.33E+01 3.29E+02

B. lodines1-131 Ci 1.16E-03 1.21 E-04 3.40E-05 7.59E-05 1.39E-03

I-133 Ci 1.18E-02 9.650E-04 1.56E-04 5.49E-04 1.35E-02Total Activity Ci 1.30E-02 1.09E-03 1.90E-04 6.25E-04 1.49E-02

C. Particulates1-131 Ci 1.24E-06 1.24E-06Ba-1 40 Ci 3.21 E-04 2.44E-05 7.84E-05 4.24E-04Be-7 Ci 8.31 E-06 8.31 E-06Ce-141 Ci 9.49E-07 9.49E-07Co-58 Ci 2.62E-06 7.48E-05 5.76E-06 2.01 E-05 1.03E-04Co-60 Ci 3.90E-05 3.94E-04 3.34E-05 7.20E-05 5.39E-04Cr-51 Ci 1.03E-04 4.OOE-04 6.79E-05 3.04E-04 8.75E-04Cs-137 Ci 1.26E-05 2.49E-05 1.78E-05 3.39E-05 8.92E-05Fe-59 Ci 5.97E-05 8.56E-06 6.83E-05Mn-54 Ci 1.04E-06 3.11E-04 1.1 OE-05 2.56E-05 3.48E-04Sr-89 Ci 1.26E-04 1.98E-05 4.77E-06 6.32E-05 2.14E-04Sr-90 Ci 3.68E-07 2.53E-07 6.21 E-07Zn-65 Ci 8.23E-05 4.99E-04 4.96E-05 3.88E-04 1.02E-03Total Activity Ci 6.88E-04 1.81 E-03 1.99E-04 9.94E-04 3.69E-03

D. Gross AlphaIGross Alpha I Ci I 1.14E-06 I 1.85E-06 I 8.37E-07 I 9.93E-07 I 4"82E'06

E. TritiumIH-3 I Ci I 1.06E+01 I 2.83E+00 4.45E+00 7.66E+00I 2.55E+01I

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Radiological Environmental Reviewfor REMM change 95-7

MP Radiological Effluent Monitoring Manual(Section I, REMODCM)

The proposed change eliminates liquid waste treatment systems that are no longer needed to bein service. They include the unit 2 boric acid evaporator and unit 3 regenerative evaporator,demineralizer and demineralizer filter. Table C-3, Radioactive Liquid Waste Sampling andAnalysis Program, was updated to reflect the elimination of this equipment. In addition,changes in Section E represent an update of current farm locations for goat and cow milk. Thejustification for these changes is in the attached Radiological Environmental Reviews. Thesechanges do not constitute an Unreviewed Radiological Environmental Impact.

The proposed changes will maintain the level of radioactive effluent control required by10CFR20,106, 40CFR190, 10CFR50.36a, and Appendix I of 10CFR5O and will not adverselyimpact the accuracy or reliability of effluent, dose or setpoint calculations.

Prepared by: Date_ /

Reviewed by: -/.6 6'2A&UY Date L•4LL"Ray Crandall-Rad Eng Sup

Radiological Environmental Review for theMillstone CPF Liquid Waste Systems

MP2 PDCR #MP3 PDCR #

REMM CR # 95-7

The MP2/MP3 Condensate Polishing Facility (CPF) indudes liquid waste treatment equipmentfor the treatment of regenerant evaporator waste, if necessary. Some of this equipment iscommon to both units 2 & 3. None of this treatment equipment has been needed or used, and isnever expected to be needed or used. The proposed PDCRs and REMODCM change requestwill officially remove this equipment from the plant design. This Radiological EnvironmentalReview provides the justification for such removal by analyzing the affect on liquid effluentsdesign bases.

The justification is based on the analysis of data from actual experience. During the first quarterof 1983, MP2 experienced the worst credible combination of primary coolant activity andprimary to secondary leak rate allowed by Technical Specifications. The fact that the CPF wastetreatment equipment was not necessary at that time is the justification that it would never berequired.

A review of data from Millstone Unit 2 operating conditions in early 1983 was performed todetermine if the Condensate Polishing Facility (CPF) liquid waste system would be neededunder abnormal conditions. During this period, the primary system DEQ Iodine concentrationwas dose to the current tech specs and primary/secondary leakrates 3 times greater thancurrently allowed. Therefore, the conditions that existed then are worse than current limitingconditions for operation allow for both Units 2 & 3.

The CPF Waste Neutralization Tank effluent discharge data was evaluated for this time perioddue to the abnormally high fuel failure rate and high primary to secondary steam generator leakrate. The maximum number of curies discharged from a waste neutralization sump during thisperiod was 2.89 E-03 Ci.

Using the maximum whole body dose conversion factor for unit 2 or 3 effluents,DwB/Ci = 1.7E-02 mrem/Ci, and assuming I discharge /day for a 31 day period with thismaximum worst case quarterly sump discharge:

DwB- 1 2.89 E-03 Ci] [ 31] [1.7E-02] =f 0.0015 mrem

REMODCM limit = 0.06 mrem*

Using the maximum organ dose conversion factor for unit 2 or 3 effluents,

DoRGAN/Ci = 1.4E-01 mrem/Ci:

DORGAN - 0.0125 mrem

REMODCM limit - 0.2 mrem* 'offsite dose limits requiring liquidwaste treatment systems to be operable.

I

These doses are well below their respective REMODCM limits for effluent doses to the offsitepublic without the use of the CPF liquid waste treatment systems. Based on the calculateddoses during abnormal conditions at or approaching current Technical Specification LimitingConditions for Operation, the CPF liquid waste systems should not be required to bemaintained and operable.

Based on the fact that the CPF radwaste treatment equipment has never been used, there will beno change in actual releases as the deletion of the equipment from design documents simplymakes it consistent with actual practice.

The potential estimated dose, under worst case operating conditions, without treatment isinsignificant compared to the Technical Specification Limits and compared to the expected dosefrom primary side liquid radwaste releases. Since there would be essentially no increase in totalliquid effluent releases under design basis conditions, the proposed change does not constitutean Unreviewed Radiological Environmental Impact.

A determination has been made that the change will maintain the level of radioactive effluentcontrol required by 10CFR20.106, 40CFR190, 10CFR50.36a, and Appendix I of 10CFR50 and thatthis change will not adveresely impact the accuracy or reliability of effluent, dose or setpointcalculations.

Prepared by: g Date // DateLen'Lnr rRdn

Reviewed by: __ __ _ Date 4 9 "Ray Crandall-Rad Eng Sup

2

REM'vP.: MA CA _lGE REQUEST # 9S- 7

P3CR #

MP2 BORIC ACID EVAPORATOR ELIMINATIONRADIOLOGICAL ENVIRONMENTAL REVIEW

References

1. "Demonstration of Compliance with 1 OCFR50 Appendix I," D. C. Switzer toDirector, NRR, November 15, 1976

2. Memo, Bob Griffin to Jerry Kinsman, November 8, 1995, CHEM-95-525

3. E-Mail, Jerry Kinsman to Ray Crandall, November 22, 1995

Reference (1) provided the justification for MP2s ability to meet the effluent ALARAguidelines of 1OCFR50, Appendix I. In order to meet the liquid effluent dose guidelines,credit was taken for the removal of radioactivity by the boric acid evaporator. As such,a requirement to use the boric acid evaporator to process waste if effluent doseguidelines are approached is included in the REMODCM.

The proposed changes to the REMODCM and plant design would eliminate the boricacid evaporator and replace it with the connections necessary to ensure a timelyinstallation of a portable demineralizer if high activity conditions exist. This reviewevaluates the effluent dose consequences of the proposed changes.

Currently, the boric acid evaporator is not required to operate in order to maintain liquideffluent doses ALARA and meet the waste treatment requirements of the REMODCM.This is because of the excellent fuel conditions which exist. Primary coolantradioactivity concentrations are typically 100-1000 times less than coolant activity limits.Since there is always the possibility we may experience an operating cycle with fueldegradation, the radwaste systems must be designed to meet 10CFR50 Appendix Iwhile operating at Technical Specification coolant activity limits.

Reference (1) assumed the inputs and removal mechanisms as shown in the attachedFigure 1 for clean liquid waste. We will be replacing the boric acid evaporator asshown in Figure 1 with a portable demineralizer. This evaluation looks at the netconsequences of this change on liquid effluent curie releases and resulting doseconsequences.

First we must consider the time period of no replacement for the evaporator as thedemineralizer is not in place and would have to be ordered at the time high inputactivity was recognized. Reference (3) confirms that a disposable demineralizer couldbe installed within two weeks. Since the design basis input into clean waste isapproximately 3,000 gallons per day and since the coolant waste receiver tanks have acombined capacity of 120,000 gallons, there is sufficient capacity to allow for twoweeks of accumulation prior to processing. Conversations with MP2 operationspersonnel confirm the fact that there should be no problem delaying the processing of

the coolant waste receiver tank for two weeks. Therefore, there would be no timeperiod without an assumed portable demineralizer should higher than expected activityexist. Radioactivity level criteria will be added to chemistry procedures to ensureportable demineralizers are ordered in a timely manner if necessary.

Next, we must address the difference in removal efficiency between the evaporator andthe portable demineralizer. Demineralizer efficiency (DF) is dependent on inletconcentration and on the remaining ion exchange capacity. We will only need theseportable demineralizers when the inlet activity is high. Additionally, the demineralizerwill come with new resins. Therefore, high DFs can be assumed. Reference (2) notesthat current exerience with new mixed bed resins yields the following DFs: OverallDF > 1.5 x 10 , iodine DF > 1 x 10 3 and CS DF >2. Comparing this to the DFsassumed for the evaporator as shown in Figure 1, we see that the demineralizer isbetter for iodines and comparable for all other nuclides except Cs (also Rb, but Rb hasnot been observed in MP2 clean waste). Therefore, the only expected increase inactivity calculated to be released from clean waste would be an increase in Cs nuclidesby a factor of 500.

What is the overall increase in Cs activity and what are the increased doseconsequences of the increased Cs compared to the doses calculated in Reference (1)and our ability to meet 10CFR50 Appendix I Fortunately, Cs nuclides are one of theprimary nuclides in the assumed secondary side activity in Reference (1). Thesecondary side activity is released via blowdown, untreated. This results in Cs releasesfrom the secondary side significantly greater than from clean waste. Additionally, theaerated waste system only has an assumed total DF of 2 for Cs. This also results insignificantly more Cs from aerated waste than clean waste. The amount of Cscalculated to be released in Reference (1) from blowdown and aerated waste is 1500times more than Cs from clean waste. Therefore, a 500 factor increase is Cs fromclean waste would only result in an overall increase of 30 percent in total Cs released.

The effect on calculated dose consequences is even less significant. The calculatedwhole body dose from all liquid effluents is only 0.03 mrem/yr. Cs contributes 5 percentof this dose. Hence, the Increased dose due to an increase of 30 percent in Csreleases is only 0.0005 mrem/yr, or essentially 0 compared to the 10CFR50 Appendix Idesign guideline of 3 mrem/yr. The calculated maximum organ dose from all liquideffluents is 0.82 mrem/year. Cs contributes only 0.1 percent of this dose. Therefore, a30 percent increase in Cs releases would only result in an additional 0.0003 mrem/yr,or essentially 0 compared to the 10CFR50 Appendix I design guideline of 10 mremnyr.

Conclusion

Based on the above, it can be concluded that the replacement of the boric acidevaporator with a portable demineralizer that can be installed within 2 weeks results inno increase in calculated doses and does not degrade the ability to meet I0CFR50Appendix I. As such, the change does not constitute an Unreviewed RadiologicalEnvironmental Impact.

Prepared By

A

FIGURE 1ASSUMED CLEAN LIQUID WASTE INr; .;''%. iOVAL

342 GPD*

EQUIPMENT 16 GPD*DRAIN TANK 0.1 PCA-

DF = 10 FOR ALLLETDOWN NUCLIDES

= 2 FOR Cs, Rb= 100 FOR OTHERS

2670 GPDBUT

0.05 PCA, Cs & Rb

DF = 100 FOR I= 1000 FOR OTHERS

DF = 10 FOR ALL

ENVIRONMENT

* GPD = GALLONS PER DAYPCA = PRIMARY COOLANT ACTIVITY CORRESPONDING TO TECH SPEC LIMIT

RADIOLOGICAL ENVIRONMENTAL REVIEW

REMODCMCY AND MP MILK SAMPLING CHANGES

Positive indications of plant related strontium have never been observed in milk. There is norequirement in the NRC Standard RETS to perform strontium analyses on milk samples.Therefore, as a cost savings measure the requirement in Section E of the CY and MP REMM toanalyze milk samples monthly for Sr-89 and Sr-90 is being changed to quarterly analyses.Quarterly analyses will be performed to continue following the general downward trend of positiveSr-90 resulting from past nuclear weapon's testing. The State of Connecticut also analyzes milksamples for Sr-90. We feel it is important to have an independent check of their results. Quarterlyanalyses are sufficient for these two functions.

Two milk sample locations, one goat farm and one cow farm, used as part of the MillstoneRadiological Environmental Monitoring Program (REMP) went out of business. The goat farmlocated 14 miles NNW was used as a control goat milk location and the cow farm located 9.5 milesWNW was used as the sole indicator cow milk lo~ation.

Based on the 1995 census of milking cows in a 15 mile'radius around Millstone, no replacementfarm with milking cows exists within the 10 mile indicator location radius. Therefore, thesampling of cow milk around Millstone is being discontinued until such time that milking cowsagain exist within 10 miles of Millstone. The annual requirement to conduct a cow and goatcensus will continue. In the event that a new farm opens or a present existing non-milking farmresumes operation the sampling of cow milk will resume. Goat milk will continue to be sampledfrom four locations. Goat milk is a more sensitive indicator of the effects of station effluents.

In another cost savings measure, the control goat farm samples taken as part of the CY REMP willbe used in the MP REMP. Control samples are required to be taken in an area similar to the sitearea yet distant enough from the station to not be influenced by station discharges. The farmlocated 23 miles N of MP and 17 miles NE of CY is sufficiently distant from both stations to beconsidered an adequate control location for both. The data obtained from the analyses of thesesamples will be used in both programs. The NRC Region 1 inspector for REMP concurs with thischange.

None of the changes represent an unreviewed radiological environmental impact.

Prepared by:

Reviewed by: ____ ____--

Radiological Environmental Reviewfor ODCM change 95-8

MP Offsite Dose Calculation Manual(Section II, REMODCM)

The proposed changes resulted from audits performed during the calendar year 1995. Thechange to the dose per curie conversion factor in Section C provides more accuracy to initialmethod 1 effluent dose calculations performed by station personnel. The changes to the goatand cow milk locations in Appendix G represent an update of current farm locations. Detailedjustification for these changes is included in the attached Radiological Environmental Reviews.These changes do not constitute an Unreviewed Radiological Environmental Impact.

The proposed changes will maintain the level of radioactive effluent control required by10CFR20.106, 40CFR190, 10CFR50.36a, and Appendix I of 10CFR50 and will not adverselyimpact the accuracy or reliability of effluent, dose or setpoint calculations.

Prepared by: Ang

Reviewed by: I6Ray Crandall-Rad Eng Sup

Date______r

Date

Radiological Environmental Review for theMP Liquid Dose Calculations

Section C of the Millstone ODCM, Liquid Dose Calculations, was audited during the year 1995.The maximum individual whole body dose per curie factors in Appendix A, Derivation ofFactors for Section C1 - Liquid Doses, were reviewed and updated through 1994 data. It wasfound that the worst case whole body dose conversion factors, derived from actual data,increased to 2.5 due to the increase in the percent of zinc-65 contained in liquid effluents fromunit 1. The change will reflect, more accurately, the mix of radionuclides discharged and thesubsequent conservative estimate of the quarterly whole body dose associated with liquideffluents.

Currently method 1 dose calculation estimates performed by station chemistry are conservativedue to other factors such as dilution flow. This increase in the dose conversion factor moreaccurately reflects the current mixture of liquid effluents and will ensure that we will maintaincontrol of liquid effluents below regulatory limits.

Prepared by: kL~n11~ry-WRed Eng

Reviewed by: _ _____Ray Crandall-Rad Eng Sup

Date

Date


REMODCMCY AND WP MILK SAMPLING CHANGES

Positive indications of plant related strontium have never been observed in milk. There is norequirement in the NRC Standard RETS to perform strontium analyses on milk samples.Therefore, as a cost savings measure the requirement in Section E of the CY and MP REMM toanalyze milk samples monthly for Sr-89 and Sr-90 is being changed to quarterly analyses.Quarterly analyses will be performed to continue following the general downward trend of positiveSr-90 resulting from past nuclear weapon's testing. The State of Connecticut also analyzes milksamples for Sr-90. We feel it is important to have an independent check of their results. Quarterlyanalyses are sufficient for these two functions.

Two milk sample locations, one goat farm and one cow farm, used as part of the MillstoneRadiological Environmental Monitoring Program (REMP) went out of business. The goat farmlocated 14 miles NNW was used as a control goat milk location and the cow farm located 9.5 milesWNW was used as the sole indicator cow milk lodation'.

Based on the 1995 census of milkling cows in a 15 mile'radius around Millstone, no replacementfarm with milking cows exists within the 10 mile indicator location radius. Therefore, thesampling of cow milk around Millstone is being discontinued until such time that milking cowsagain exist within 10 miles of Millstone. The annual requirement to conduct a cow and goatcensus will continue. In the event that a new farm opens or a present existing non-milking farmresumes operation the sampling of cow milk will resume. Goat milk will continue to be sampledfrom four locations. Goat milk is a more sensitive indicator of the effects of station effluents.

In another cost savings measure, the control goat farm samples taken as part of the CY REMP willbe used in the MP REMP. Control samples are required to be taken in an area similar to the sitearea yet distant enough from the station to not be influenced by station discharges. The farmlocated 23 miles N of MP and 17 miles NE of CY is sufficiently distant from both stations to beconsidered an adequate control location for both. The data obtained from the analyses of thesesamples will be used in both programs. The NRC Region 1 inspector for REMP concurs with thischange.

None of the changes represent an unreviewed radiological environmental impact.

Prepared by:

Reviewed by: _______________

NortheastUtilities I

RADIOLOGICAL EFFLUENT MONITORING&

OFFSITE DOSE CALCULATION MANUAL

REMODCM

NORTHEAST NUCLEAR ENERGY COMILLSTONE NUCLEAR POWER STATION

Waterford, CT

MILLSTONE UNIT 1, DOCKET NO. 50-245MILLSTONE UNIT 2, DOCKET NO. 50-336MILLSTONE UNIT 3, DOCKET NO. 50-423

RADIOLOGICAL EFFLUENT

MONITORING MANUAL

FOR THEMILLSTONE NUCLEAR POWER STATION

UNIT NOs. 1, 2, & 3

DOCKET NOs. 50-245, 50-336, 50-423

December 1995Revision 8

D483XW.091

12/31/95Revision 8

MILLSTONE STATION

RADIOLOGICAL EFFLUENT MONITORING MANUAL

TABLE OF CONTENTS

SECTION PAGE NO.

A. INTRODUCTION

B. RESPONSIBILITIES

C. LIQUID EFFLUENTS

A-1

B-1

REV. NO.

3

3

DATE

12/1/94

12/31/94

C.1 LIQUID EFFLUENTS SAMPLINGAND ANALYSIS PROGRAM

C.2 LIQUID RADIOACTIVE WASTE

TREATMENT

D. GASEOUS EFFLUENTS

D.1 GASEOUS EFFLUENTSSAMPLING AND ANALYSISPROGRAM

D.2 GASEOUS RADIOACTIVEWASTE TREATMENT

C-1C-2C-3C-4C-5C-6C-7C-8C-9C-10

C-11

2232223323

3

7/1/907/1/9012/31/947/1/907/1/907/1/9012/31/9412/15/957/1/9012/31/94

12/15/95

I

I

D-1D-2D-3D-4D-5D-6D-7D-8

D-9

22323223

2

7/1/907/1/9012/31/947/1/9012/31/947/1/907/1/9012/31/94

7/1/90

D483XW.091

12/31/95Revision 8

SECTION PAGE NO. REV. NO. DATE

E. RADIOLOGICAL ENVIRONMENTAL MONITORING

E.1 SAMPLING AND ANALYSIS E-1 5 8/31/94E-2 5 8/31/94E-3 6 12/31/95E-4 6 12/31/94E-5 5 8/31/94E-6 5 8/31/94

E.2 LAND USE CENSUS E-7 5 8/31/94E.3 INTERLABORATORY

COMPARISON PROGRAM E-8 5 8/31/94

F. REPORT CONTENT

F.1 ANNUAL RADIOLOGICALENVIRONMENTAL OPERATINGREPORT F-1 3 3/15/95

F.2 ANNUAL RADIOACTIVEEFFLUENT REPORT F-2 3 8/31/94

D483XW 091

12/15/95Revision 3

MILLSTONE 3

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM

Lower Limitof Detection

Sampling Analysis Type of Activity (LLD)*Liquid Release Type Frequency Frequency Analysis (•Ci/ml)

A. Batch Releaseb

1. CondensatePolishing Prior to Each Prior to Each Principal Gamma 5 x 10-Facility - Waste Batch Batch EmitterscNeutralizationSumpe

1-131, Mo-99 I x 10bCe-141, Ce-144 5 x 10b

2. Waste TestTanks Dissolved and I x 10.

Entrained Gassesd

3. Low Level Prior to Each MonthlyWaste Drain Batch Compositef'-Tank H-3d I x 10.1

Gross alpha" I x 10"4. Boron Test Prior to Each Quarterly

Tanks Batch Compositef, Sr-89d, St-90d 5x10 8

Fe-550 1 x 10"

B. ContinuousRelease

1. SteamGenerator Daily Grab Weekly Principal Gamma 5 x 10'Blowdownh Sample' Compositeg Emittersc

1-131, Mo-99 1 x 10"0Ce-141, CE-144 5 x 10-0

2. Service Water Monthly Grab Monthly Dissolved and 1 x 10"1Effluent Sample Entrained Gasses)

Weekly Grab MonthlySample Composite' H-31 I x 10.1

Gross alpha) 1 x 10"3. Turbine Building Weekly Grab Quarterly

Sumps' Sample Compositeg Sr-891, Sr-906 5 x 10.8_Fe-55' I x 10"b

C-8D483XW.091

12/15/95Revision 3

C.2 Liquid Radioactive Waste Treatment

All applicable liquid radioactive waste treatment systems will be operated when theprojected dose due to liquid effluents averaged over 31 days exceeds 0.06 mrem tothe total body or 0.2 mrem to any organ.

The term "all applicable liquid radioactive waste treatment" is defined as thatequipment applicable to a waste stream responsible for greater than ten percent (10%)of the total projected dose. The liquid radioactive waste treatment system equipmentis specified below for each unit.

Millstone Unit No. I

1. Filtration and ion exchange system.

2. Waste demineralizer A or B.

Millstone Unit No. 2

1. Degasifier, clean liquid primary demineralizer, portable disposable demineralizer,clean liquid secondary demineralizer and filter.

2. The aerated waste demineralizer and filter.

Millstone Unit No. 3

1. Degasifier, ion exchanger, boron evaporator, boron demineralizer and wastedemineralizer filter.

2. High level waste demineralizer or waste evaporator, waste demineralizer andwaste demineralizer filter.

If actual 31-day doses exceed 0.06 mrem to the total body or 0.2 mrem to any organand the doses from the untreated pathway exceed 10% of these limits, prepare andsubmit to the Commission a report that includes the following information:

1. Explanation of why liquid radwaste was being discharged without treatment,identification of any inoperable equipment or subsystems, and the reason for theinoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLE status, and

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

If the above treatment systems are not routinely operating, doses due to liquideffluents from the untreated pathway to UNRESTRICTED AREAS shall be projected atleast once per 31 days in accordance with the methodology and parameters in theODCM. If any of these dose projections exceed 10 percent (10%) of the above limits,then best efforts shall be made to return the inoperable equipment to service.

1 1D483XW.091 Lo,° I I

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TABLE E-1

MILLSTONE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

Exposure Pathway Number of Sampling and Type and Frequency ofand/or Sample Locations Collection Frequency Analysis

la. Gamma Dose - 17 Monthly Gamma Dose - MonthlyEnvironmentalTLD

lb. Gamma Dose - 22 Quarterly"'I N/A-a)Accident TLD

2. Airborne 8 Continuous sampler - Gross Beta - WeeklyParticulate weekly filter change Gamma Spectrum - Quarterly

on composite (by location), andon individual sample if grossbeta is greater than 10 timesthe mean of the weekly controlstation's gross beta results

3. Airborne Iodine 8 Continuous sampler - 1-131 - Weeklyweekly canister change

4. Vegetation 5 One sample near Gamma Isotopic on eachmiddle and one near sampleend of growing season

5. Milk 4 Monthly Gamma Isotopic and 1-131 oneach sample - MonthlySr-89 and Sr-90 - Quarterly

5a. Pasture Grass 4 Sample as necessary to Gamma Isotopic and 1-131substitute forunavailable milk

6. Sea Water 2 Quarterly - Composite Quarterly - Gamma Isotopic,of 6 weekly grab and Tritium on each compositesamples

7. Bottom Sediment 5 Semiannual Gamma Isotopic on eachsample

8. Fin Fish-Flounder 2 Quarterly Gamma Isotopic on eachand one other sampletype of edible finfish

9. Mussels 4 Quarterly Gamma Isotopic on eachsample

10. Oysters 4 Quarterly Gamma Isotopic on eachsample

11. Clams 2 Quarterly Gamma Isotopic on eachsample

12. Lobsters 2 Quarterly Gamma Isotopic on eachsample

(a) Accident monitoring TLDs to be dedosed at least quarterly.

E-3D483X.091

SECION 1

OFFSITE DOSE

CALCULATION MANUAL


UNIT NOs. 1, 2, & 3

DOCKET NOs. 50-245, 50-336, 50-423

December 1995Revision 8D483XW 091

12/15/95Revision 8

MILLSTONE STATION


TABLE OF CONTENTS

SECTION PAGE NO.

A. INTRODUCTION

B. RESPONSIBILITIES

A-1

B-1

REV. NO.

3

3

DATE

12/31/94

12/31/94

C. LIQUID DOSE CALCULATIONS

C.1 QUARTERLY - TOTAL BODY DOSEa. Method I - Any Unitb. Method 2 - Any Unit

C.2 QUARTERLY - MAXIMUM ORGAN Da. Method 1 - Any Unitb. Method 2 - Any Unit

C.3 ANNUAL - TOTAL BODY DOSEC.4 ANNUAL - MAXIMUM ORGAN

DOSEC.5 MONTHLY DOSE PROJECTIONS

C-1C-1OSEC-1C-2C-2

C-2

44

444

4

444

4

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12/15/9512/15/9512/15/95

12/15/95

12/15/9512/15/9512/15/95

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a. Unit 1 C-3b. Unit 2 and Unit 3 C-4

C-5C.6 QUARTERLY DOSE CALCULATIONS

FOR ANNUAL RADIOACTIVEEFFLUENT REPORT C-5

D. GASEOUS DOSE CALCULATIONS

D.1 10CFR20 LIMITS ("INSTANTANEOUS")a. Noble Gases - All Units D-1b. lodines, Particulates and D-1

Other - All Units D-2D-3

D.2 10CFR50 APPENDIX I - NOBLE GAS LIMITSa. Quarterly Air Dose - Method 1

All Units D-4

3343

2

33

5/1/935/1/933/15/953/15/95

7/1/90

12/31/9412/31/94

b. Quarterly Air Dose - Method 2All Units

c. Annual Air Dose - All UnitsD-5D-5

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MILLSTONE STATION


TABLE OF CONTENTS (Cont'd)


D. GASEOUS DOSE CALCULATIONS (Cont'd)

D.3 10CFR50 APPENDIX I - IODINE ANDPARTICULATE DOSESa. Quarterly Doses - Unit I D-6 3 5/1/93

D-7 3 5/1/93b. Quarterly Doses - Unit 2 & D-8 4 12/31/94

Unit 3 D-9 3 5/1/93c. Annual Doses - All Units D-10 3 12/31/94

D.4 GASEOUS EFFLUENT MONTHLYDOSE PROJECTIONSa. Unit 1 D-11 2 7/1/90

D-12 2 7/1/90b. Unit 2 D-12 2 7/1190c. Unit 3 D-13 3 12/31/94

D.5 QUARTERLY DOSE CALCULATIONSFOR ANNUAL RADIOACTIVEEFFLUENT REPORT D-13 3 12/31/94

D.6 COMPLIANCE WITH 40CFR190 D-14 3 12/31/94

E. LIQUID MONITOR SETPOINT CALCULATIONS,

E.1 UNIT 1 LIQUID RADWASTEEFFLUENT LINE E-1 3 12/31/94

E.2 UNIT 1 SERVICE WATEREFFLUENT LINE E-2 3 12/31/94

E.3 UNIT 2 CLEAN LIQUID E-2 3 12/31/94RADWASTE EFFLUENT LINE E-3 3 12/31/94

E-4 3 12/31/94E.4 UNIT 2 AERATED LIQUID

RADWASTE EFFLUENT LINEAND CPF - WASTE NEUT.SUMP EFFLUENT LINE E-4 3 12/31/94

E.5 UNIT 2 STEAM GENERATORBLOWDOWN E-4 3 12/31/94

E-5 3 12/31/94E.6 UNIT 2 CONDENSER AIR

EJECTOR E-5 3 12/31/94E.7 UNIT 2 REACTOR BUILDING

CLOSED COOLING WATER E-5 3 12/31/94E-6 3 12/31/94

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MILLSTONE STATION




E.8 UNIT 3 LIQUID WASTE E-6 3 12/31/94EFFLUENT LINE E-7 3 12/31/94

E-8 3 12/31/94E.9 UNIT 3 REGENERANT

EVAPORATOR EFFLUENT LINE E-8 3 12/31/94E.10 UNIT 3 WASTE NEUTRALIZA-

TION SUMP EFFLUENT LINE E-8 3 12/31/94Ell UNIT 3 STEAM GENERATOR

BLOWDOWN E-8 3 12/31/94E-9 3 12/31/94

E.12 UNIT 3 TURBINE BUILDINGFLOOR DRAINS EFFLUENT LINE E-9 3 12/31/94

F. GASEOUS - MONITOR SETPOINTCALCULATIONS F-1 2 7/1/90

F.1 UNIT 1 HYDROGEN MONITOR F-1 2 7/1/90F.2 UNIT 1 STEAM JET AIR EJECTOR

OFF GAS MONITOR F-1 2 7/1/90F.3 UNIT 1 MAIN STACK NOBLE

GAS MONITOR F-1 2 7/1/90F.4 UNIT 1 MAIN STACK SAMPLER

FLOW RATE MONITOR F-2 2 7/1/90F.5 UNIT 2 VENT NOBLE GAS

MONITOR F-2 2 7/1/90F.6 UNIT 2 WASTE GAS DECAY

TANK MONITOR F-2 2 7/1/90F.7 UNIT 3 VENT NOBLE GAS

MONITOR F-3 2 7/1/90F.8 UNIT 3 ENGINEERING

SAFEGUARDS BUILDINGMONITOR F-3 2 7/1/90

WWWaVW.I T of C3

. 12/15/95Revision 8

MILLSTONE STATION



FIGURE NAME

FIGURE NUMBER

G-1

G-2

G-3

REV. NO.

Inner Air Particulate andVegetation Monitoring Stations

Outer Terrestrial MonitoringStations

Aquatic Sampling Stations

2

2

2

DATE

7/1/90

7/1/90

7/1/90

D4&,XW.091 T of C4

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MILLSTONE STATION



APPENDICES

REV. NO. DATE

APPENDIX A

APPENDIX B

APPENDIX C

APPENDIX D

APPENDIX E

APPENDIX F

APPENDIX G

Derivation of Factors for SectionC.1 - Liquid Doses


Liquid Dose Calculations - LADTAP

Derivation of Factors for Section DGaseous Doses

Gaseous Dose Calculations - GASPAR

Gaseous Dose Calculations - AIREM

Environmental Monitoring ProgramSampling Locations

3

2

2

4

2

2

7

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7/1/90

7/1/90

3/15/95

7/1/90

7/1/90

12/31/95

I

I

048MXW.0A1 T of C5

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C.1 Quarterly - Whole Body Dose

a. Method I - Any Unit

Step I Determine CF which is total gross curies of fission andactivation products, excluding tritium and dissolved noblegases, released during the calendar quarter.

Step 2 Determine CH which is total curies of tritium releasedduring the calendar quarter.

Step 3 Determine Dow which is the quarterly dose to the wholebody in mrem.

For Unit 1:

Dow = 2.5 CF + 5.6 x 10.7 CHt

For Units 2 and 3:

Dow = 2 x 10-2 CF + 5.6 x 10,7 CH

Step 4 If Dow is greater than 0.5 mrem, go to Method 2.

Note: See Appendix A for derivation of these factors. ForUnit 1, the dose contribution from tritium can be neglectedsince it has never contributed to more than 2% of thewhole body doses.

b. Method 2 - Any Unit

If the calculated dose using Method 1 is greater than 0.5 mrem, usethe NRC computer code LADTAP II to calculate the liquid doses. Theuse of this code and the input parameters are given in RadiologicalAssessment Branch Procedure, Liquid Dose Calculations -LADTAP II.

C.2 Quarterly - Maximum Oroan Dose


Step I Determine CF which is total gross curies of fission andactivation products, excluding tritium and dissolved noblegases, released during the calendar quarter - same asStep C. I.a.

Step 2 Determine Dao which equals the quarterly dose to themaximum organ in mrem.

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For Unit 1:

Doo = 2.1 CF

For Units 2 and 3:

Doo = 0.2 CF

(See Appendix B for derivation of these factors)

Step 3 If Doo is greater than 2 mrem, go to Method 2.


If the calculated dose using Method 1 is greater than 2 mrem, use theNRC computer code LADTAP to calculate the liquid doses. The useof this code and the input parameters are given in RadiologicalAssessment Branch Procedure, Liquid Dose Calculations -LADTAP II.

C.3 Annual - Whole Body Dose - Any Unit

Determine Dyw which equals dose to the whole body for the calendar yearas follows:

Dyw = 2: Dow, where the sum is over the first quarter through the presentquarter whole body doses.

The following should be used as Dow:

(1) If the detailed quarterly dose calculations required per Section C.6 forthe Annual Radioactive Effluent Report are completed for anycalendar quarter, use that result.

(2) If the detailed calculations are not complete for a particular quarter,use the results as determined in Section C. 1.

(3) If Dyvw is greater than 3 mrem and any Dow determined as inSection C. I was not calculated using Method 2 of that section,recalculate Dow using Method 2 if this could reduce Dyw to less than3 mrem.

C.4 Annual - Maximum Organ Dose - Any Unit

Determine Dyo which is the dose to the maximum organ for the calendaryear as follows:

Iyo = Z Doo, where the sum is over the first quarter through the present

quarter maximum organ doses.

The following guidelines should be used:

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(1) 11 the. jetailed quarterly dose calculations required per Section C.6 forthe Annual Radioactive Effluent Report are completed for anycalendar quarter, use that result.

(2) If the detailed calculations are not complete for a particular quarter,use the results as determined in Section C.2.

(3) If different organs are the maximum for different quarters, they maybe summed together and Dyo can be recorded as a less than value aslong as the value is less than 10 mrem.

(4) If Dyo is greater than 10 mrem and any value used in its determinationwas calculated as in Section C.2, but not with Method 2, recalculatethat value using Method 2 if this could reduce Dyo to less than 10mrem.

C.5 Monthly Dose Projections

a. Whole Body and Maximum Organ - Unit 1

Step 1 Determine D'mw which is the whole body dose from thelast typical* previously completed month as calculated perthe methods in Section C. 1.

Step 2 Determine D'MO which is the maximum organ dose fromthe last typical* previously completed month as calculatedper the methods in Section C. 1.

Step 3 Estimate R1 which is tbe ratio of the total estimatedvolume of liquid batches to be released in the presentmonth to the volume released in the past month.

Step 4 Estimate R 2 which is the ratio of estimated primary coolantactivity for the present month to that for the past month.

Step 5 Determine F which is the factor to be applied to theestimated ratio of final curies released if there areexpected differences in treatment of liquid waste for thepresent month as opposed to the past month (e.g.,bypass of filters or demineralizers). NUREG-0016 or pastexperience should be used to determine the effect ofeach form of treatment which will vary. F = 1 if there areno expected differences.

Step 6 Determine DEmw which is the estimated monthly wholebody dose as follows:

DEmw = D'mw * R1 * R2 * F

C-3D483XW.001

12/15/95Revision 4

Step 7 Determine DEMo which is the estimated monthly maximumorgan dose as follows:

D EM = D'MO * R1 * R2 * F

The last typical month should be one without significantoperational differences from the projected month.

For example, if the plant was down for refueling the entiremonth of February and startup is scheduled for March 3,use the last month of operation as the base month toestimate March's dose.

Or, if there were no releases during September, do notuse September as the base month for October if it isestimated that there will be releases in October.

If the last typical month's doses were calculated usingLADTAP II (or similar methodology), also multiply theLADTAP doses by R5 where R5 = total dilution flow fromLADTAP run divided by estimated total dilution flow.

b. Whole Body and Maximum Organ - Unit 2 and Unit 3

Step I Determine D'MW which is the whole body dose from thelast typical* previously completed month as calculated perthe methods in Section C. 1.

Step 2 Determine D'MO which is the maximum organ dose fromthe last typical* previoysly completed month as calculatedper the methods in Section C.2.

S - See footnote in Section C.5.a.

Step 3 Estimate R, which is the ratio of the total estimatedvolume of liquid batches to be released in the presentmonth to the volume released in the past month.

Step 4 Estimate R 2 which is the ratio of estimated volume ofsteam generator blowdown to be released in presentmonth to the volume released in the past month.

Step 5 Determine F, which is the fraction of curies released lastmonth coming from steam generator blowdown.

i.e. F1 = curies from blowdowncuries from blowdown + curies from batch tanks

Step 6 Estimate R3 which is the ratio of estimated secondarycoolant activity for the present month to that for the pastmonth.

C-4D03M.091

12/15/95Revision 4

Step 7 Estimate R4 which is the ratio of estimated primary coolantactivity for the present month to that for the past month.

Step 8 Determine F2 which is the factor to be applied to theestimated ratio of final curies released if there areexpected differences in treatment of liquid waste for thepresent month as opposed to the past month (e.g.,bypass of filters or demineralizers). NUREG-0017 or pastexperience should be used to determine the effect ofeach form of treatment which will vary. F2 = 1 if there areno expected differences.

Step 9 Determine DEMw which equals estimated monthly total

body dose as follows:

DEMw = D'MW [(1 - F1) R1 R4 F2 + F1 R2 R3]

Step 10 Determine DEMO which equals estimated monthlymaximum organ dose as follows:

DEWM = D'MO [(1 - F1) R1 R4 F2 + F1 R2 R3]

C.6 Quarterly Dose Calculations for Annual Radioactive Effluent Report

Detailed quarterly dose calculations required for the Annual RadioactiveEffluent Report shall be done using the NRC computer code LADTAP I1.The use of this code, and the input parameters are given in RadiologicalAssessment Branch Procedure Liquid Dose Calculations - LADTAP I1.

C-5D483XW.O01

12/15/95Revision 3

APPENDIX A

DERIVATION OF FACTORS FOR SECTION C.1. - LIQUID DOSES

1. .ection C.l.a - Step 3

CF= Curies of fission and activation products released duringcalendar quarter.

Down = Calculated whole body dose to the maximum individual (mrem)due to fission and activation products. Dose calculated usingcomputer code LADTAP.

CH = Curies of tritium released during calendar quarter.

DQW(H) " Calculated whole body dose to the maximum individual (mrem)due to tritium releases. Dose calculatedLADTAP.

UNIT 1 - LIQUID - WHOLE BODY DOSES

using computer code

Dow(F)/CFCF Down (mrem/Ci)

Dow(H)/CHCH Dpow(H) (mrem/Ci)Year Quarter

1980

1981

1982

1983

1984

0.0130.0140.0110.686

0.3140.0420.0290.009

0.0080.0300.5770.538

0.7770.0070.0070.016

0.0170.0160.0030.002

6.2 (-5)1.6 (-4)1.2 (-4)1.2 (-2)

5.8 (-3)7.6 (-4)3.5(-4)1.2 (-4)

1.2 (-4)1.8(-4)7.4 (-3)6.1 (-3)

3.9 (-3)7.3 (-5)1.0 (-4)2.0 (-4)

1.6 (-4)2.3(-4)3.1 (-5)2.0 (-5)

4.8 (-3)1.1 (-2)1.1 (-2)1.8 (-2)

1.9 (-2)1.8 (-2)1.2 (-2)1.3 (-2)

1.5 (-2)6.0 (-3)1.3 (-2)

1.1 (-2)

5.0 (-3)1.0 (-2)1.4 (-2)1.3 (-2)

9.4 (-3)1.4 (-2)1.0 (-2)1.0 (-2)

2.404.966.45

13.50

1.420.880.310.006

0.120.123.882.08

3.04 (-7)1.54 (-6)1.67 (-6)

1,27 (-7)3.10 (-7)2.59 (-7)

1.611.873.641.26

0.775.561.061.19

3.96 (-7)1.16 (-6)

1.5 (-6)1.5 (-7)2.4 (-7)

2.12 (-7)3.19 (-7)

2.7 (-7)1.4 (-7)2.0 (-7)

-1-0483XW.001

12/15/95Revision 3


Dow(Fp/CFCF DownF (mrem/Ci)

Dow(H/CHCHi D~owfHo (mrem/Ci)Year Quarter

1985

1986

1987

1988

1989

1234

1234

1234

1234

1234

1234

1234

1234

1234

0.0380.0250.3540.049

0.0190.5110.2390.004

0.0120.1420.4130.577

0.5980.2800.1450.059

0.0870.3440.3530.110

0.0420.0270.0120.060

0.0610.5980.3400.354

0.2240.0710.1100.058

0.0570.0350.0260.011

4.5(-4)3.3 (-4)3.1 (-3)6.7(-4)

2.2 (-4)5.2 (-3)2.2 (-3)4.0(-5)

1.1 (-4)1.5 (-3)6.6 (-3)5.9 (-3)

6.2 (-3)1.3 (-2)7.5(-3)1.1 (-2)

1.1 (-2)

3.3 (-1)2.8(-1)3.1 (-2)

1.51 E-021.80E-024.77E-037.17E-03

4.40E-035.92E-016.66E-013.27E-01

1.58E-012.26E-022.87E-022.56E-02

2.38E-023.61 E-024.64E-021.20E-02

1.2 (-2)1.3 (-2)8.8 (-3)1.4 (-2)

1.2 (-2)1.0 (-2)9.2 (-3)1.0 (-2)

9.2 (-3)1.1 (-2)1.6 (-2)1.0 (-2)

1.0 (-2)4.6 (-2)5.2 (-2)1.9 (-1)

1.3 (-1)9.6 (-1)

7.9 (-1)2.8 (-1)

2.612.435.267.56

1.731.781.010.81

1.659.932.663.57

2.888.42

11.814.7

11.411.5

3.8619.0

4.9 (-7)4.6 (-7)2.4 (-7)2.4 (-7)

4.0 (-7)2.4 (-6)6.6 (-7)9.5(-7)

8.3 (-7)2.0 (-6)2.6 (-6)3.6 (-6)

3.0 (-6)3.8 (-6)8.8 (-7)4.6 (-6)

2.8 (-7)2.6 (-7)2.4 (-7)3.0 (-7)

2.4 (-7)2.4 (-7)2.5(-7)2.7 (-7)

2.9 (-7)2.4 (-7)2.2 (-7)2.4 (-7)

2.6 (-7)3.3 (-7)2.3(-7)2.4 (-7)

1990

1991

1992

1993

3.57E-016.74E-014.08E-011.19E-01

7.18E-029.90E-011.96E+009.24E-01

7.05E-013.17E-012.61E-014.43E-01

4.IBE-011.03E+001.78E+001.09E+00

-2-D4U83XW.CG

* 12/15/95Revision 3


Dow(F)/CFDow(F) (mrem/Ci)

Dow(H)/CHDDOW(H) (mrem/Ci)Year Quarter C• CH

Year Quarter CF

1994 1234

0.0060.0360.0060.012

1.26E-026.18E-021.50E-023.02E-02

2.1 OE+001.72E+002.50E+002.50E+00

-3-D483XW.091

12/15/95Revision 3

UNIT 2 - UQUID - WHOLE BODY DOSES

Year Quarter CF

1

1980 2341

1981 2341

1982 2341

1983 2341

1984 2341

1985 2341

1986 2341

1987 24

11988 2

341

1989 234

0.6350.2851.170.7230.4350.3430.2653.141.659.941.141.141.480.6852.423.221.490.860.410.801.172.290.830.300.400.361.552.181.581.080.111.30

4.562.840.2411.257.013.070.1390.448

DQW(F) /CFDQW(F) (mrem/Ci)

4.0 (-3) 6.3 (-3)1.7 (-3) 6.0 (-3)7.9 (-3) 6.8 (-3)1.2 (-2) 1.7 (-2)6.8 (-3) 1.6 (-2)5.8 (-3) 1.7 (-2)1.6.(-3) 6.0 (-3)1.0 (-2) 3.2 (.3)1.0(-2) 6.1(-3)8.4(-3) 8.5(-4)8.1 (-3) 7.1(-3)1.3 (-2) 1.10(-)1.1 (-2) 7.4 (-3)

7.2 (-3) 1.1 (-2)3.6(-2) 1.5(-2)4.5 (-2) 1.4 (-2)9.4 (-3) 6.3 (-3)1.4 (-2) 1.6 (-2)3.9 (-3) 9.5 (-3)4.7 (-3) 5.9 (-3)8.4 (-3) 7.2 (-3)3.0 (-2) 1.3 (-2)8.5 (-3) 1.0 (-2)4.9 (-3) 1.6 (-2)4.2 (-3) 1.1 (-2)4.0 (-3) 1.1 (-2)

1.1 (-2) 5.0(-3)4.1 (-3) 2.6 (-3)5.7 (-3) 5.3 (-3)7.4 (-4) 6.7 (-3)9.8 (-3) 7.5 (-3)

1.9 (-2) 4.2 (-3)1.2 (-2) 4.2 (-3)1.4 (-3) 5.8 (-3)1.1 (-2) 8.8 (-3)1.9 (-2) 2.7(-3)1.9 (-2) 6.2(-3)8.6(-4) 6.2 (-3)3.1 (-3) 6.9 (-3)

DQW(H) /CHCH DQW(H) (mremlCi)

97.757.048.864.855.3

149.087:279.9

7.488.3

113.0.82.6

70.736.7

6.56.8

47.477.1

13613734.9

5.725.1

100.050.9

134.081.812.955.753.7

10868.6

20.596.442.399.988.521.476.7

179.

1.09 (-5)

2.28 (-5)

1.91 (-7)

3.52 (-7)

5.41 (-5) 3.63P(-)1.77 (-5) 2.03 (-7)

4.91 (-5) 5.56 (-7)

2.2 (-5)2.6 (-5)2.5 (-s)

1.5 (-6)3.3 (-6)2.9 (-5)1.5 (-5)3.5 (-5)2.1 (-5)3.8 (-6)1.3 (-5)1.3 (-5)2.7 (-5)1.8 (-5)

5.9 (-6)2.3 (-5)9.1 (-6)2.4 (-5)2.4 (-5)7.0 (-6)1.7 (-5)4.4 (-5)

2.9 (-7)1.9 (-7)1.8 (-7)

2.6 (-7)1.3 (-7)2.9 (-7)2.9 (-7)2.6 (-7)2.6(-7)2.9 (-7)2.3 (-7)2.4 (-7)2.5(-7)2.6 (-7)

2-9 (-7)2.4 (-7)2.2 (-7)2.4 (-7)2.7 (-7)3.3 (-7)2.2 (-7)2.5 (-7)

* For data prior to 1986, tritium doses not listed when doses from tritium were less than 1/2%of the whole body doses.

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DOW/CFCF Down (mrem/Ci)

DOw(H)/CHDDown-n (mrem/Ci)Year Quarter CM

Year QuarterCH

1986

1987

1988

1989

1234

1234

1234

1234

0.060.881.960.11

1.991.760.431.22

1.460.720.300.66

1.281.132.471.05

9.1 (-6)2.5(-4)1.3 (-3)2.7(-4)

3.3 (-3)3.4 (-3)1.4 (-3)1.4 (-2)

1.5 (-2)3.3(-3)1.5 (-3)3.8 (-3)

6.2 (-3)6.9 (-3)2.8(-2)1.3 (-2)

1.5 (-4)2.8 (-4)6.6 (-4)2.5(-3)

1.7 (-3)1.9 (-3)3.3 (-3)1.1 (-2)

1.0 (-2)4.6 (-3)5.0 (-3)5.8 (-3)

4.8 (-3)6.1 (-3)1.1 (-2)1.2 (-2)

4.099.6

286169

243171

98.377.4

52.0124

95.1276

18719472.5

244

1.1 (-6)2.6 (-5)6.8 (-5)5.0 (-5)

5.9 (-5)4.2 (-5)2.4 (-5)2.1 (-5)

1.5 (-5)2.9 (-5)2.1 (-5)6.7 (-5)

4.9 (-5)6.3 (-5)1.6 (-5)5.9 (-5)

2.8 (-7)2.6 (-7)2.5 (-7)3.0 (-7)

2.4 (-7)2.5 (-7)2.4 (-7)2.7(-7)

2.9 (-7)2.3 (-7)2.2 (-7)2.4 (-7)

2.6 (-7)3.2 (-7)2.2 (-7)2.4 (-7)

DoseProiected Releases* Ci/vr (mrem/year) Dose/Ci

Total Fission and Activation(excl. H-3) 0.18 8.8(-4) 4.9(-3)

H-3 730 1.6(-4) 2.2(-7)

*From Unit 3 EROLS Table 5.2-4

Maximum Value of Dawff9CF =

Maximum Value of DoW(H/CH =

Unit 1 = 2.5 x 100 mrem/CiUnit 2 = 1.7 x 10.2 mrem/CiUnit 3 = 1.2 x 10*2 mrem/Ci (From 1986-1989 data)

Unit I = 4.3 x 10.? mrem/CiUnit 2 = 5.6 x 10.7 mrem/CiUnit 3 = 3.2 x 10.7 mrem/Ci (From 1986-1989 data)

I

I

The maximum value of Dow(W/CF is much higher for Unit I than for Units 2 and 3. Thisdifference in dose/curie values was caused by the change in the isotopic mixtureresulting from operation of the zinc injection system at Unit 1; therefore, a separatevalue for Unit I is necessary. The values for Units 2 and 3 are within a factor of 2 andcan be conservatively applied to both units.

D483XW.091 -5-

12/15/95Revision 3

Since the maximum value Daw(H/CH is not much different for Units 1, 2 and 3, the same

factor can be used for all three units (for simplicity).

* Dose factor values for Unit I have increased significantly because of the zinc injection

process.

Thus, for Unit 1: DQW(F9/CF =

Daw(H/Ct =

DOW(F9CF =

Daowq/Cm =

2.5 x 100 mrem/Ci

5.6 x 10"7 mrem/Ci

2.0 x 10.2 mrem/Ci

5.6 x 10"7 mrem/Ci

for Units 2 and 3:

*Not.e: Although operation of Unit 3 increases the dilution flow, the near fielddilution factor is reduced from 5 to 3. Therefore, the net effect is to reducethe doses by only a factor of 0.86. As can be seen in the 1986-1989 data,this has a negligible effect on the doses.

-6-D83XW.0MI

12/31/95Revision 7

ENVIRONMENTAL MONITORING PROGRAM

Samplina Locations

The following lists the environmental sampling locations and the types of samplesobtained at each location. Sampling locations are also shown on Figures G-1, G-2,and G-3.

LocationNumber

Direction & Distance FromName Release Point*** Sample Types

1-1" On-site - Old Millstone Road2-1 On-site - Weather Shack3-1 On-site - Bird Sanctuary4-1 On-site - Albacore Drive5-1 MP3 Discharge6-1 Quarry Discharge7-1 Environmental Lab Dock8-1 Environmental Lab9-1 Bay Point Beach

10-1 Pleasure Beach11-1 New London Country Club12-C Fisher's Island, NY13-C Mystic, CT14-C Ledyard, CT15-C Norwich, CT16-C Old Lyme, CT17-1 Site Boundary18-1 Pleasure Beach21-1 Goat Location #122-1 Goat Location #223-1 Goat Location #324-C Goat Location #425-1 Fruits & Vegetables26-C Fruits & Vegetables27-1 Niantic28-1 Two Tree Island29-1 West Jordan Cove30-C Golden Spur31-1 Niantic Shoals

32-1 Vicinity of Discharge

33-1 Seaside Point34-1 Thames River Yacht Club35-1 Niantic Bay36-1 Black Point37-C Giant's Neck38-1 Waterford Shellfish Bed #1

0.6 Mi, NNW0.3 Mi, S

0.3 Mi, NE1.0 Mi, N

0.1 Mi, SSE0.3 Mi, SSE0.3 Mi, SE0.3 Mi, SE0.4 Mi, W1.2 Mi, E

1.6 Mi, ENE8.7 Mi, ESE11.5 Mi, ENE12.0 Mi, NE14.0 Mi, N8.8 Mi, W0.5 Mi, NE1.2 Mi, E2.0 Mi, N

5.2 Mi, NNE2.0 Mi, ENE22.0 Mi, N

Within 10 MilesBeyond 10 Miles

1.7 Mi, WNW0.8 Mi, SSE0.4 Mi, NNE4.7 Mi, NNW1.8 Mi, NW

1.5 Mi, NNW

1.8 Mi, ESE4.0 Mi, ENE

0.3 Mi, WNW3.0 Mi, WSW3.5 Mi, WSW1.0 Mi, NW

TLD, Air Particulate, Iodine, VegetationTLD, Air Particulate, IodineTLD, Air Particulate, IodineTLD, Air Particulate, IodineTLDTLDTLDTLDTLDTLD, Air Particulate, IodineTLD, Air Particulate, IodineTLDTLDTLDTLD, Air Particulate, IodineTLDVegetationVegetationMilkMilkMilkMilkVegetationVegetationTLD, Air Particulate, IodineMusselsClams

Bottom Sediment, OystersMusselsBottom Sediment, Oysters, Lobster,Fish, SeawaterBottom SedimentBottom SedimentLobster, FishOystersBottom Sediment, Oysters, SeawaterClams

II

*1 = Indicator '*C = Control***The release points are the MP1 stack for terrestrial locations and the end of the quarry for aquatic locations.

-1 -D4&3XW.091

12/31/94Revision 3

TABLE Q-3 (Cont'd.)

TABLE NOTATIONS

a. The lower limit of detection (LLD) is defined in Table Notations, Item a, ofTables C-1, C-2, or C-3.

b. For gaseous samples, Mhe LLD will be 1 x 10-4 iCi/cc and for particulate samples,the LLD will be 1 x 10" 1 i.Ci/cc. The principal gamma emitters for which theseLLDs apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-1 35, and XE-138 for gaseous emission and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions.The list does not mean that only these nuclides are to be detected and reported.Other peaks which are measurable and identifiable, together with the abovenuclides, shall also be identified and reported. Nuclides which are below the LLDfor the analyses should not be reported as being present at the LLD level for thatnuclide. When unusual circumstances result in a pdriori LLDs higher than required,the reasons shall be documented in the Annual Radioactive Effluent Report.

c. Sampling and analysis shall also be performed within 24 hours following anunexplained increase, as indicated by the Unit 3 vent noble gas monitor, ofgreater than 50%, after factoring out increases due to changes in THERMALPOWER levels, containment purges, or other explainable increases.

d. The ratio of the sample flow rate to the sampled stream flow rate shall be known.

e. Analyses for 1-133 will not be performed on each charcoal sample. Instead, atleast once per month, the ratio of 1-133 to 1-131 will be determined from acharcoal sample changed after 24 hours of sampling. This ratio, along with theroutine 1-131 activity determination will be used to determine the release rate of1-133.

f. Samples shall be changed at least once per seven days and analyses shall becompleted within 48 hours after changing. Special sampling and analysis ofiodine and particulate filters shall also be performed whenever reactor coolant1-131 samples (which are taken two to six hours following a THERMAL POWERchange exceeding 15% of RATED THERMAL POWER in one hour perTable 4.4-4 of the Safety Technical Specifications) show an increase of greaterthan a factor of 5. These filters shall be changed following such a five-foldincrease in coolant activity and every 24 hours thereafter until the reactor coolant1-131 levels are less than a factor of 5 greater than the original coolant levels oruntil seven days have passed, whichever is shorter. Sample analyses shall becompleted within 48 hours of changing. The LLDs may be increased by a factorof 10 for these samples.

g. Grab samples for tritium shall be taken weekly from the ventilation vent wheneverthe refueling cavity is flooded and there is fuel in the cavity.

h. Subsequent to medical emergencies, for initial determination of isotopic content ofthe containment air, a Health Physics sample may be used in place of the normalchemistry sample.

D483xw9 ogD-8

REMM CHANGE 95-1


The proposed change clarifies a requirement related to the contents of theannual Radiological Environmental Monitoring Report. The current requirementis based on the generic RETS wording and states, "if harmful effects aredetected by the monitoring, the report shall provide an analysis of the problemand a planned course of action to alleviate the problem.'

During a 1994 NRB audit, the auditor noted that what constitutes "harmfuleffects' is a nebulous term subject to many interpretations. Therefore, to providea clear threshold for the need for the additional analysis, we are specifying thespecific dose levels of 1 OCFR50 Appendix I. These levels are well below levelsat which harmful effects would occur, but are levels at which we are required totake corrective actions to ensure we do not exceed them again. We have notexceeded such levels since the 1970's.

The proposed change adds clarity and is more conservative and hence does notconstitute an unreviewed radiological environmental impact. A determinationhas been made that the change will maintain the level of radioactive effluentcontrol required by 10CFR20.106, 40CFR190, 10CFR50.36a, and Appendix I toIOCFR50 and not adversely impact the accuracy or reliability of effluent, dose,or setpoint calculations.

PREPARED BY 1 6d / -R. A. Crandall, Supervi orRadiological Engineering

RES95002

ODCM CHANGE 95-2


The proposed change will eliminate the consideration of the milk and vegetationpathways in the method for calculating the instantaneous release rate limit foriodines, particulates, and tritium. Only the inhalation pathway need beconsidered for the instantaneous release rate limit. This is consistent with thewording in the LCO of the Technical Specification and with the requirements ofthe standard RETS. A corresponding change to the MPI, MP2, and MP3 RETSbases section is also being submitted for clarity and consistency. The need toonly consider inhalation is also consistent with the instantaneous release ratelimit that existed in the Environmental Tech Specs (ETS) up to the switch toRETS in 1986. The ETS were based on the concentration limits of 10CFR20Appendix B Table II. These concentration limits are based on the inhalationpathway only.

The dose to the public from all pathways will continue to be maintained lessthan 10CFR50 Appendix I limits which are much more restrictive than the10CFR20 limits. The RETS and ODCM requirements for integrated dose over aquarter or year must include the milk, vegetation, and other pathways. Hence,the proposed change does not allow an increase in integrated dose, but doesallow additional flexibility in the instantaneous release rate criteria.

A determination has been made that the change will maintain the level ofradioactive effluent control required by 10CFR20.106, 40CFRI90,10CFR50.36a, and Appendix I to IOCFR50 and not adversely impact theaccuracy or reliability of effluent, dose, or setpoint calculations.

PREPARED BY ___________//qs

R. A. Crandall, Superviso'rRadiological Engineering

RES9001

Docket Nos.:

License Nos.:


MONITORING AND

OFFSITE DOSE

CALCULA TION MANUAL

MILLSTONE UNIT NOs 1, 2, & 3

Northeast Nuclear Energy Company

50-24550-33650-423

DPR-21DPR-65NPF-44

March 1995Revision 8D483XW.091


MONITORING MANUAL


UNIT NOs. 1, 2, & 3

DOCKET NOs. 50-245, 50-336, 50-423

March 1995Revision 7D483XW.091 I

3/15/95Revision 7

MILLSTONE STATION


TABLE OF CONTENTS

SECTION PAGE NO.

A. INTRODUCTION

B. RESPONSIBILITIES

C. LIQUID EFFLUENTS

A-1

B-1

REV. NO.

3

3

DATE

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C.A LIQUID EFFLUENTS SAMPLINGAND ANALYSIS PROGRAM


TREATMENT




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E. RADIOLOGICAL ENVIRONMENTAL MONITORING

E.1 SAMPLING AND ANALYSIS E-1E-2E-3E-4E-5E-6

E.2 LAND USE CENSUS E-7E.3 INTERLABORATORY

COMPARISON PROGRAM E-8

F. REPORT CONTENT

F.1 ANNUAL RADIOLOGICALENVIRONMENTAL OPERATINGREPORT F-1

F.2 ANNUAL RADIOACTIVEEFFLUENT REPORT F-2

5556555

5

8/31/948/31/948/31/9412/31/948/31/948/31/948/31/94

8/31/94

3

3

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8/31/94

I

A-4D403XW.091

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F. REPORT CONTENT

F.1 Annual Radiological Environmental Operating Report

The Annual Radiological Environmental Operating Report shall includesummaries, interpretations, and statistical evaluation of the results of theradiological environmental surveillance activities for the report period,including a comparison with previous environmental surveillance reports andan assessment of the observed impacts of the plant operation on theenvironment. The report shall also include the results of the land usecensus required by Section E.2 of this manual. If levels of radioactivity aredetected that result in calculated doses greater than 10CFR50 Appendix IGuidelines, the report shall provide an analysis of the cause and a plannedcourse of action to alleviate the cause.

The report shall include a summary table of all radiological environmentalsamples which shall include the following information for each pathwaysampled and each type of analysis:

(1) Total number of analyses performed at indicator locations.

(2) Total number of analyses performed at control locations.

(3) Lower limit of detection (LLD).

(4) Mean and range of all indicator locations together.

(5) Mean and range of all control locations together.

(6) Name, distance and direction from discharge, mean and range for thelocation with the highest annual mean (indicator or control).

(7) Number of nonroutine reported measurements as defined in thesespecifications.

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 themissing results. The missing data shall be submitted in the next annualreport.

The report shall also include a map of sampling locations keyed to a tablegiving distances and directions from the discharge; the report shall alsoinclude a summary of the Interlaboratory Comparison Data required bySection E.3 of this manual.

F-1D483XW.091

OFFSITE DOSE

CALCULATION MANUAL


UNIT NOs. 1, 2, & 3

DOCKET NOs. 50-245, 50-336, 50-423

D483XW.091March 1995Revision 7 I

3/15/95Revision 7

MILLSTONE STATION


TABLE OF CONTENTS


A. INTRODUCTION A-1 3 12/31/94

B. RESPONSIBILITIES B-1 3 12/31/94


C.A QUARTERLY - TOTAL BODY DOSEa. Method 1 - Any Unit C-1 3 12/31/94b. Method 2 - Any Unit C-1 3 12/31/94

C.2 QUARTERLY - MAXIMUM ORGAN DOSEa. Method 1 - Any Unit C-1 3 12/31/94b. Method 2 - Any Unit C-2 3 12/31/94

C.3 ANNUAL - TOTAL BODY DOSE C-2 3 12/31/94C.4 ANNUAL - MAXIMUM ORGAN

DOSE C-2 3 12/31/94C.5 MONTHLY DOSE PROJECTIONS

a. Unit 1 C-3 3 12/31/94b. Unit 2 and Unit 3 C-4 3 12/31/94

C-5 3 12/31/94C.6 QUARTERLY DOSE CALCULATIONS

FOR ANNUAL RADIOACTIVEEFFLUENT REPORT C-5 3 12/31/94


D.1 10CFR20 LIMITS ("INSTANTANEOUS")a. Noble Gases - All Units D-1 3 5/1/93b. lodines, Particulates and D-1 3 5/1/93

Other - All Units D-2 4 3/15/95D-3 3 3/15/95

D.2 10CFR50 APPENDIX I - NOBLE GAS LIMITS

a. Quarterly Air Dose - Method 1All Units D-4 2 7/1/90

b. Quarterly Air Dose - Method 2All Units D-5 3 12/31/94

c. Annual Air Dose - All Units D-5 3 12/31/94

D483XW.0ol T of C1

3/15/95Revision 7

MILLSTONE STATION





D.3 10CFR50 APPENDIX I- IODINE ANDPARTICULATE DOSESa. Quarterly Doses - Unit 1 D-6 3 5/1/93



D.4 GASEOUS EFFLUENT MONTHLYDOSE PROJECTIONSa. Unit I D-11 2 7/1/90

D-12 2 7/1/90b. Unit 2 D-12 2 7/1/90c. Unit 3 D-13 3 12/31/94

D.5 QUARTERLY DOSE CALCULATIONSFOR ANNUAL RADIOACTIVEEFFLUENT REPORT D-13 3 12/31/94


E. LIQUID MONITOR SETPOINT CALCULATIONS










ED483XW.091 T of C1

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TION SUMP EFFLUENT LINE E-8 3 12/31/94E.11 UNIT 3 STEAM GENERATOR

BLOWDOWN E-8 3 12/31/94E-9 3 12/31/94











D483XW.091 T of C2

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MILLSTONE STATION



FIGURE NAME

FIGURE NUMBER

G-1

G-2

G-3

REV. NO.


Outer Terrestrial MonitoringStations


2

2

2

DATE

7/1/90

7/1/90

7/1/90

D483XW.091 T of C3

MILLSTONE STATION



APPENDICES

REV. NO.

3/15/95Revision 7

DATE

APPENDIX A

APPENDIX B

APPENDIX C

APPENDIX D

APPENDIX E

APPENDIX F

APPENDIX G








2

2

2

4

2

2

6

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7/1/90

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I

D483XW.091 T of C4

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(1) The release rate limit of 1-131, 1-133, and tritium from the site shall be:

5.5 x 104 131Qll + 5.1 x 10.2 131Q012 + 5.1 x 10.2 131 Q13 + 1.33 x 104

133% + 1.25 x 10-2 133Q12 + 1.25 x 10.2 i33Q13 + 4.4 x 10-8 QH1+ 4.2 x 10"6 QH2 + 4.2 x 10"6 QH3 < 1

(2) The release rate limit of particulates with half-lives greater than 8 days andtritium from the site shall be:

5.5 x 104 QP1 + 5.1 x 10.2 Qn + 5.1 x 10.2 QP3 + 4.4 x 10"8 QH1+ 4.2 x 10"6 QH2 + 4.2 x 10"6 QH3 < 1

Method 2

Above methods assume a conservative nuclide mix. If necessary, utilize theGASPAR code to estimate the dose rate limit.

With releases within the above limits, the dose rate to the maximum organ will beless than 1500 mrem/year from the inhalation pathway.

D-2D483XW.091

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INTENTIONALLY LEFT BLANK

D-3D483XW.091

3/15/95Revision 4

It would also be unnecessarily restrictive to assume the worst possiblemixture and use that as the limit for all situations. Therefore, a practicalsolution is to use a conservatively determined empirical method as givenabove.

3. Section D.1.b - Release Rate Limit - Iodine and Particulates

Doses are calculated using the methods of NUREG-0133 dated October 1978 andNRC Regulatory Guide 1.109, Revision 1. Note that the equation of page 27 ofNUREG-0133 (for all radionuclides, except tritium) has been corrected for theelemental iodine fraction, as in Regulatory Guide 1.109, Revision 1. For theinstantaneous release rate limit, only the inhalation pathway needs to beconsidered.

Method 1

DTz =[xIQOP0... 13 Q1]+[XIQoJi *

where: DT, = thyroid dose rate from inhalation due to iodine releases, mrem/yr

131Qz = release rate of 1-131, IuCi/sec133Q. = release rate of 1-133, R.Ci/secX/Q = meterological dispersion factor, sec/im3

Pi values derived from NUREG-0133.and Regulatory Guide 1.109 (seeTable 1).

Dose formula for tritium is:

D7H = [X /Q' S e QH]

where: DTH = thyroid (or any other organ) dose rate from inhalation due tritium

releases, mrem/yr

QH = release rate of H-3, jiCi/sec

Dose formula for particulates is:

Do =[X/QO PeQp]

where: Do, = maximum organ dose rate from inhalation due to particulate

releases

QP = release rate of particulates, iRCi/sec

other parameters as described for iodine, above.

a. Thyroid Doses

Release rate limit is 1500 mrem/year, using the average worst case X/Q'sand Pi values from Table 1 results in:

D48&3.w09 -10-

3/15/95Revision 4

UNIT 1

DT = .82 131Q11 + .20 1 Q33 + 6.63 X 10-5 QH1

UNIT 2 AND UNIT 3

DT = 76.8 131Qi + 18.72 133QI + 6.24 x 10 3QH

Summing all three units, and setting < 1500 mrem/year results in:

0.82 131Q11 + .20 133QII + 6.63 x 10'5 QHI + 76.8 131Q12 +18.72 133Q,2 + 6.24 x 10-3 QH2 + 76.8 131Q13 + 18.72 133Q13 +6.24 x 10-3 QH3 < 1500 mrem/yr

dividing by 1500 gives:

5.5 x 104 131Qil + 5.1 X 10.2 131QI2 + 5.1 X 10.2 131Q13 +1.33 x 104 133Q1 + 1.25 x 10.2 133QI2 + 1.25 x 10-2 13Q13 +

4.4 x 10"8 QHI + 4.2 x 10"6 QH2 + 4.2 x 10"6 QH3_< 1

b. Maximum Oracan (other than thyroid) Doses

Release rate limit is 1500 mrem/year. Using the 1980-1987 worst case X/Q's and Pivalues (conservative mix)* from Table 1 results in:

UNIT 1

Do = (5.1 x 10-. 1.6 x 107 . QP) + (5.1 x 1048 1.3 x 103. QH)

Do = (0.82 Qp) + (6.63 x 10"5 QH)

*Sr-90 values are too conservative. Review of the 1978-1988 effluent data shows Sr-90

has never contributed to greater than 2% of the total curies; use the next limiting nuclide(other than Sr-89). Comparison with ODCM Rev. 0, shows this results in a conservativecalculation.

UNIT 2 and UNIT 3

Do = (4.8 x 10'. 1.6 x 107. Qp) + (4.8 x 10-6 -1.3 x 103. QH)

= 76.8 Qp + 6.24 x 10.3 QH

D4a3XW.091 -11 -

3/15/95Revision 4

Summing all three units, and setting < 1500 mrem/year results in:

0.82 QP1 + 76.8 Qp2 + 76.8 QP3 + 6.63 x 10s QH1+

6.24 x 10-3 QH2 + 6.24 x 10-3 QH3 < 1500 mrem/yr

dividing by 1500 gives:

5.5 x 104 Qp, + 5.1 x 10-2 QP2 + 5.1 x 10.2 QP3 + 4.4 x 10"8 QH1 +

4.2 x 10"6 QH2 + 4.2 x 10"6 QH3_< 1

D483XW.091 -12 -

3/15/95Revision 4

INTENTIONALLY LEFT BLANK

D483XW.01 -13-17-

RADIOLOGICAL ENVIRONMENTAL REVIEW FOR ODCM CHANGE #94-9

In Section E the following radiation monitor alarm and alert setpoint requirements were removedfrom the manual:

1. li alarm for Unit 1 Service Water Effluent (E.2),2. Alert for Unit 2 Clean Liquid Radwaste Effluent (E.3),3. Alert for Unit 3 Liquid Waste Discharge (E.8), and4. Alert for Unit 3 Regenerant Evaporator Effluent (E.9).

The hi-alarm and alert are being eliminated from the manual to provide additional flexibility insetting the setpoint values to avoid nuisance annunciations in the control rooms. They are notneeded because each radiation monitor has another setpoint (hi-hi-alarm at Unit I and alarms atUnits 2/3) which, although higher than the hi-alarm or alert setpoints, are set conservativelyenough to terminate releases well before discharge limits are reached. Only this higher setpointneeds to be addressed in the ODCM. Because of the general accuracy of Chemistry samples andbecause of required monitor surveillances which provide sufficient confidence in the functionalityof the hi-hi-alarm and alarms the elimination of the hi-alarm and alerts is acceptable. Therefore aRadiological Environmental Impact Review is not needed.

On page D-14 a statement was added to step (7). The statement discusses the direct radiation thatpossibly will be received in the future due to the required onsite storage of low level radioactivewastes. In compliance with 40CFRI90, all sources from the station must be considered whendetermining the total dose to a real individual. The Radiological Assessment Branch will performan annual assessment of the maximum site boundary dose from all radwaste storage locations. Thedose expected from these radwaste sources may be comparable to the dose already being receivedform MPI Skyshine. The results of this assessment will be accounted for when determining theannual maximum site boundary dose. A Radiological Environmental Impact Review is not neededsince the dose expected is no more than what already exists form MP I Skyshine and will beassessed yearly by the Radiological Assessment Branch.

The goat farm at 14 miles NE sampled as a control sample location as part of the RadiologicalEnvironmental Monitoring Program went out of business. In Appendix G, a replacement locationhas been chosen at 14 miles NNW. The replacement farm is sufficiently distant from the station tobe considered an adequate control location. Because the sample is only used as a control and notfor the direct monitoring of environmental effects from the station a Radiological EnvironmentalImpact Review is not necessary.

- Prepared by:1)/

jýAI-ý,AdA44L 40004r-T

•#,//•_ •-•/•I'- Reviewed by: ___ ______

Approved by: A

Docket Nos.: 50-24550-33650-423

DPR-21DPR-65NPF-44

License Nos.:


MONITORING AND

OFFSITE DOSE

CALCULATION MANUAL

MILLSTONE UNIT NOs 1, 2, & 3

Northeast Nuclear Energy Company

0483XW.091December 1994Revision 7


MONITORING MANUAL


UNIT NOs. 1, 2, & 3

DOCKET NOs. 50-245, 50-336, 50-423

D4•6W.lDecember 1994Revision 6

12/31/94Revision 6

MILLSTONE STATION


TABLE OF CONTENTS

SECTION PAGE NO.

A. INTRODUCTION

B. RESPONSIBILITIES

C. LIQUID EFFLUENTS

C.1 LIQUID EFFLUENTS SAMPLINGAND ANALYSIS PROGRAM


TREATMENT




A-1

B-1

REV. NO.

3

3

DATE

12/1/94

12/31/94

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E. RADIOLOGICAl. lNVIRONMENTAL MONITORING

E.1 SAMPLING AND ANALYSIS E-1 5 8/31/94E-2 5 8/31/94E-3 5 8/31/94E-4 6 12/31/94E-5 5 8/31/94E-6 5 8/31/94

E.2 LAND USE CENSUS E-7 5 8/31/94E.3 INTERLABORATORY

COMPARISON PROGRAM E-8 5 8/31/94

F. REPORT CONTENT

F.1 ANNUAL RADIOLOGICALENVIRONMENTAL OPERATINGREPORT F-1 2 7/1/92

F.2 ANNUAL RADIOACTIVEEFFLUENT REPORT F-2 3 8/31/94

DQ&Xw.Ml A-4

12/31/94Revision 3

A. INTRODUCTION

The purpose of this manu,-;. to provide the sampling and analysis programswhich provide input to the O(CM for calculating liquid and gaseous effluentconcentrations and offsite Ldo es. Guidelines are provided for operatingradioactive waste treatment 2ystems in order that offsite doses are kept As-Low-As-Reasonably-Achievable (ALARA).

The Radiological Environmer:.al Monitoring Program outlined within this manualprovides confirmation that the; measurable concentrations of radioactive materialreleased as a result of opera' ions at the Millstone Site are not higher thanexpected.

In addition, this manual outlines the information required to be submitted to theNRC in both the Annual Radiological Environmental Operating Report and theAnnual Radioactive Effluent Report.

D483w.091 A-1

12/31/94Revision 3

B. RESPONSIBILITIES

All changes to this manual shall be r- iewed and approved by the StationOperations Review Committee prior to implementation.

All changes and their rationale sha# be documented in the AnnualRadioactive Effluent Report.

It shall be the responsibility of thl'•e41tion Senior Vice President to ensurethat this manual is used in performance of the surveillance requirementsand administrative controls of the Technical Specifications.

D483XW.091 B-1

.ievision 3

TA -1(Con,.-"-

TABLE NOTATIONS

a. The LLD is the smallest concentration of radioactive mate-s a sample that will be detectedwith 95% probability with 5% probability of falsely concludi. at a blank observation representsa "real" signal.

For a particular measurement system (which may include: ," emical separation):LLD = 4.66 Sb

E. V . 2.22 x 106 .Y. exp (- Z At)where:

LLD is the lower limit of detection as defined above (as pCi per unit mass or volume)

Sb is the standard deviation of the background counting rate or of the counting rate of ablank sample as appropriate (as counts per minute)

E is the counting efficiency (as counts per transformation)

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

2.22 x 106 is the number of transformations per minute per microcurie

Y is the fractional radiochemical yield (when applicable)

X. is the radioactive decay constant for the particular radionuclide

At is the elapsed time between midpoint of sample collection and midpoint of counting time

It should be recognized that the LLD is defined as an a pori (before the fact) limit representingthe capability of a measurement system and not as an a posteriori (after the fact) limit for aparticular measurement.

Analyses shall be performed in such a manner that the stated LLDs will be achieved underroutine conditions. Occasionally background fluctuations, unavoidably small sample sizes, thepresence of interfering nuclides, or other uncontrollable circumstances may render these LLDsunachievable. In such cases, the contributing factors will be identified and recorded on theanalysis sheet for the particular sample.

b. The LLD will be 5 x 10"7 lCi/ml. The principal gama emitters for which this LLD applies areexclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-134, and Cs-137.

This list does not mean that only these nuclides are to be detected and reported. Other peakswhich are measurable and identifiable, together with the above nuclides, shall also be identifiedand reported. Nuclides which are below the LLD for the analyses should not be reported asbeing present at the LLD level. When unusual circumstances result in a Driod LLDs higher thanrequired, the reasons shall be documented in the Annual Radioactive Effluent Report.

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;2/31/94Revision 3

TABLE C-2 (Cont'd.)

TABLE NOTATIONS

This list does not mean that only nuclides are to be detected and reported. Otherpeaks which are measurable and identifiable, together with the above nuclides,shall also be identified and reported. Nuclides which are below the LLD for theanalyses should not be reported as being present at the LLD level. When unusualcircumstances result in a orioi LLDs higher than required, the reasons shall bedocumented n the Annual Radioactive Effluent Report.

d For the Condensate Polishing Facility (CPF) - Waste Neutralization Sump, theseanalyses are only required if the gamma analysis of the CPF - WasteNeutralization Sump indicates a gamma activity greater than 5 x 10"7 RCi/mI.

e. For the Condensate Polishing Facility - Waste Neutralization Sump, theseanalyses are only required when the steam generator gross activity (sampled andanalyzed three times per week as per Table 4.7-2 of the Safety TechnicalSpecifications) exceeds 1 x 10"5 tCi/ml.

f. A composite sample is one in which the quantity of liquid sampled is proportionalto the quantity of liquid waste discharged and in which the method of samplingemployed results in a specimen which is representative of the liquids released.

g. Prior to analysis, all samples taken for the composite shall be thoroughly mixed inorder for the composite sample to be representative of the effluents released.

h. For the Steam Generator Blowdown and the Turbine Building Sump, theseanalyses are only required when the steam generator gross activity (sampled andanalyzed three times per week as per Table 4.7-2 of the Safety TechnicalSpecifications) exceeds 5 x 10-7 RCi/ml.

i. Daily grab samples shall be taken at least five days per week.

j. For the Service Water, these analyses are only required if a weekly gammaanalysis indicates a gamma activity greater than 5 x10.7 RCi/ml.

D48&XW.0S1 C-7

Revisio

TA1BLF Q-3 (Cont'd.)

TABLE NOTATIONS

This list does not mean that only these nuclides are to be detected and reported. Otherpeaks which are measurable and identifiable, together with the above nuclides, shallalso be identified and reported. Nuclides which are below the LLD for the analysesshould not be reported as being present at the LLD level. When unusual circumstancesresult in a Dflori LLDs higher than required, the reasons shall be documented in theAnnual Radioactive Effluent Report.

d. For the Condensate Polishing Facility (CPF) - Waste Neutralization Sump, theseanalyses are only required if the gamma analysis of the CPF - WasteNeutralization Sump indicates a gamma activity greater than 5 x 10-7 ilCi/ml.

e. For the Condensate Polishing Facility (CPF) - Waste Neutralization Sump, theseanalyses are only required when the steam generator gross activity (sampled andanalyzed three times per week as per Table 4.7-1 of the Safety TechnicalSpecifications) exceeds 1 x 105 OLCi/ml.

f. A composite sample is one in which the quality of liquid sampled is proportional tothe quantity of liquid waste discharged and in which the method of samplingemployed results in a specimen which is representative of the liquids released.

g. Prior to analysis, all samples taken for the composite shall be thoroughly mixed inorder for the composite sample to be representative of the effluents released.

h. For the Steam Generator Blowdown and the Turbine Building Sump, theseanalyses are only required when the steam generator gross activity (sampled andanalyzed three ties per week as per Table 4.7-1 of the Safety TechnicalSpecifications) exceeds 5 x 10"7 R.Ci/ml. Steam Generator Blowdown samples arenot required when blowdown is being recovered.

i. Daily grab samples shall be taken at least five days per week.

j. For the Service Water, these analyses are only required if a weekly gammaanalysis indicates a gamma activity greater than 5 x 10.7 RCi/ml.

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TRFQ1(Cont'd.)

TABLE NOTATIONS


b. For gaseous samples, the LLD will be 1 x 10'4 iCi/cc and for particulate samples,the LLD will be I x 10-11 pCi/cc. The principal gamma emitters for which theseLLDs apply are exclusively the following radionuclides: Kr-87, Kr-B8, Xe-133, Xe-133m, Xe-1 35, and XE-138 for gaseous emission and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions.The list does not mean that only these nuclides are to be detected and reported.Other peaks which are measurable and identifiable, together with the abovenuclides, shall also be identified and reported. Nuclides which are below the LLDfor the analyses should not be reported as being present at the LLD level for thatnuclide. When unusual circumstances result in a Drio LLDs higher than required,the reasons shall be documented in the Annual Radioactive Effluent Report.

c. Sampling and analysis shall also be performed within 24 hours following anincrease, as indicated by the steam jet air ejector off-gas monitor, of greater than50%, after factoring out increases due to changes in THERMAL POWER level.



f. Samples shall be changed at least once per seven days and analyses shall becompleted within 48 hours after changing. Special sampling and analysis ofiodine and particulate filters shall also be performed whenever subsequent reactorcoolant 1-131 samples show an increase of greater than a factor of 5 afterfactoring out increases due to changes in thermal power level. These filters shallbe changed following such a five-fold increase in coolant activity and every24 hours thereafter until the reactor coolant 1-131 levels are less than a factor of 5greater than the original coolant levels or until seven days have passed,whichever is shorter. Sample analyses shall be completed within 48 hours ofchanging. The LLDs may be increased by a factor of 10 for these samples.

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TAILE Q- (Cont'd.)

TABLE NOTATIONS

a. The lower limit of detection (LLD) is defined in Table Notations, Item a, of Tables C-1, C-2, orC-3.

b. For gaseous samples, the LLD will be 1 x 10"4 gCVcc and for particulate samples, the LLD will beI x 10"11 giCVcc. The principal gamma emitters for which these LLDs apply are exclusively thefollowing radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and XE-138 for gaseousemission and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144for particulate emissions. The list does not mean that only these nuclides are to be detectedand reported. Other peaks which are measurable and identifiable, together with the abovenuclides, shall also be identified and reported. Nuclides which are below the LLD for theanalyses should not be reported as being present at the LLD level for that nuclide. Whenunusual circumstances result in a priori LLDs higher than required, the reasons shall bedocumented in the Annual Radioactive Effluent Report.

c. Sampling and analysis shall also be performed within 24 hours following an unexplainedincrease, as indicated by the Unit 2 stack noble gas monitor, of greater than 50%, after factoringout increases due to changes in THERMAL POWER levels, containment purges, or otherexplainable increases.


e. Analyses for 1-133 will not be performed on each charcoal sample. Instead, at least once permonth, the ratio of 1-133 to 1-131 will be determined from a charcoal sample changed after 24hours of sampling. This ratio, along with the routine 1-131 activity determination will be used todetermine the release rate of 1-133.

f. Samples shall be changed at least once per seven days and analyses shall be completed within48 hours after changing. Special sampling and analysis of iodine and particulate filters shallalso be performed whenever subsequent reactor coolant 1-131 samples, which are taken two tosix hours following a THERMAL POWER change exceeding 15% of RATED THERMAL POWERin one hour, show an increase of greater than a factor of 5. These filters shall be changedfollowing such a five-fold increase in coolant activity and every 24 hours thereafter until thereactor coolant 1-131 levels are less than a factor of 5 greater than the original coolant levels oruntil seven days have passed, whichever is shorter. Sample analyses shall be completed within48 hours of changing. The LLDs may be increased by a factor of 10 for these samples.

g. Grab samples for tritium shall be taken weekly whenever the refueling cavity is flooded and thereis fuel in the cavity. The grab sample shall be taken from the stack (Units 1 and 2) where thecontainment ventilation is being discharged at the time of sampling.

h. Waste Gas Storage Tanks are normally released on a batch basis. However, for the purpose oftank maintenance, inspection, or reduction of oxygen concentration, a waste gas tank may becontinuously purged with nitrogen provided the following conditions are met:

(1) The previous batch of radioactive waste gas has been discharged to a final tank pressureof less than 5 PSIG.

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TABL D-3(Cont'd.)

TABLE NOTATIONS


b. For gaseous samples, the LLD will be 1 x 10'4 iCVcc and for particulate samples,the LLD will be 1 x 10-11 iLCi/cc. The principal gamma emitters for which theseLLDs apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and XE-138 for gaseous emission and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions.The list does not mean that only these nuclides are to be detected and reported.Other peaks which are measurable and identifiable, together with the abovenuclides, shall also be identified and reported. Nuclides which are below the LLDfor the analyses should not be reported as being present at the LLD level for thatnuclide. When unusual circumstances result in a Driori LLDs higher than required,the reasons shall be documented in the Annual Radioactive Effluent Report.

c. Sampling and analysis shall also be performed within 24 hours following anunexplained increase, as indicated by the Unit 3 vent noble gas monitor, ofgreater than 50%, after factoring out increases due to changes in THERMALPOWER levels, containment purges, or other explainable increases.



f. Samples shall be changed at least once per seven days and analyses shall becompleted within 48 hours after changing. Special sampling and analysis ofiodine and particulate filters shall also be performed whenever reactor coolant1-131 samples (which are taken two to six hours following a THERMAL POWERchange exceeding 15% of RATED THERMAL POWER in one hour perTable 4.4-4 of the Safety Technical Specifications) show an increase of greaterthan a factor of 5. These filters shall be changed following such a five-foldincrease in coolant activity and every 24 hours thereafter until the reactor coolant1-131 levels are less than a factor of 5 greater than the original coolant levels oruntil seven days have passed, whichever is shorter. Sample analyses shall becompleted within 48 hours of changing. The LLDs may be increased by a factorof 10 for these samples.

g. Grab samples for tritium shall be taken weekly from the ventilation vent wheneverthe refueling cavity is flooded and there is fuel in the cavity.

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TABLE E-2

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONSIN ENVIRONMENTAL SAMPLES

Reporting Levels

AirborneParticulate Fish Shellfish(c) Vegetables

Water or Gases (pCilg, (pCi/g, Milk (pCilg,Analysis (pCill) (pCiIm 3) wet) wet) (pCi/I) wet)

H-3 20,000(a)Mn-54 1,000 30 140Fe-59 400 10 60Co-58 1,000 30 130Co-60 300 10 50Zn-65 300 20 80Zr-95 400Nb-95 400Ag-110m 8 301-131 (b) 0.9 0.2 1 -3 0.1Cs-134 30 10 1 5 60 1Cs-137 50 20 2 8 70 2Ba-140 200 _ 1 300La-140 200 1 1 300

(a) For drinking water samples. This is 40 CFR Part 131 value.

(b) Level for 1-131 not included since no radioactivity discharged to any drinking waterpathways; other reporting levels are included for trending of long-lived isotopesonly.

(c) For on-site samples, these values can be multiplied by 3 to account for the nearfield dilution factor.

DU&xw.W E-4

I

OFFSITE DOSE

CALCULATION MANUAL


UNIT NOs. 1, 2, & 3

DOCKET NOs. 50-245, 50-336, 50-423

D483XW.001December 1994Revision 6

12/31/94Revision 6

MILLSTONE STATION

-• •OFFSITE DOSE CALCULATION MANUAL

TABLE OF CONTENTS


A. INTRODUCTION A-1 3 12/31/94

B. RESPONSIBILITIES B-1 3 12/31/94


C.A QUARTERLY - TOTAL BODY DOSEa. Method 1 - Any Unit C-1 3 12/31/94b. Method 2 - Any Unit C-1 3 12/31/94

C.2 QUARTERLY - MAXIMUM ORGAN DOSEa. Method 1 - Any Unit C-1 3 12/31/94b. Method 2 - Any Unit C-2 3 12/31/94

C.3 ANNUAL- TOTAL BODY DOSE C-2 3 12/31/94C.4 ANNUAL - MAXIMUM ORGAN

DOSE C-2 3 12/31/94C.5 MONTHLY DOSE PROJECTIONS

a. Unit I C-3 3 12/31/94b. Unit 2 and Unit 3 C-4 3 12/31/94

C-5 3 12/31/94C.6 QUARTERLY DOSE CALCULATIONS

FOR ANNUAL RADIOACTIVEEFFLUENT REPORT C-5 3 12/31/94


D.1 10CFR20 LIMITS ("INSTANTANEOUS")a. Noble Gases - All Units D-1 3 5/1/93b. Iodines, Particulates and D-1 3 5/1/93

Other - All Units D-2 3 5/1/93D-3 2 7/1/90

D.2 10CFR50 APPENDIX I - NOBLE GAS LIMITSa. Quarterly Air Dose - Method 1

All Units D-4 2 7/1/90b. Quarterly Air Dose - Method 2

All Units D-5 3 12/31/94c. Annual Air Dose - All Units D-5 3 12/31/94

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MILLSTONE STATION

.:SITE DOSE CALCULATION MANUAL




D.3 IOCFR50 APPENDIX I - IODINE ANDPARTICULATE DOSESa. Quarterly Doses - Unit I D-6 3 5/1/93



D.4 GASEOUS EFFLUENT MONTHLYDOSE PROJECTIONSa. Unit 1 D-11 2 7/1/90

D-12 2 7/1/90b. Unit 2 D-12 2 7/1/90c. Unit 3 D-13 3 12/31/94

D.5 QUARTERLY DOSE CALCULATIONSFOR ANNUAL RADIOACTIVEEFFLUENT REPORT D-13 3 12/31/94 I


E. LIQUID MONITOR SETPOINT CALCULATIONS










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,?iIL,&.. ONE STATION







TION SUMP EFFLUENT LINE E-8 3 12/31/94Ell UNIT 3 STEAM GENERATOR

BLOWDOWN E-8 3 12/31/94E-9 3 12/31/94











091wool T of C3

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MILLSTON(W: 70N

OFFSITE DOSE CALCUL ION MANUAL


FIGURE NAMe

FIGURE NUMBER

G-1

G-2

G-3

REV. NO.


Outer Terrestrial Monitoring

Stations


2

2

2

DATE

7/1/90

7/1/90

7/1/90

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MILLSTONE STATION I



APPENDICES

REV. NO.

APPENDIX A

APPENDIX B

APPENDIX C

APPENDIX D

APPENDIX E

APPENDIX F

APPENDIX G


Derivation of Factors for Section

C.2 - Uquid Doses






2

2

2

3

2

2

6

DATE

7/1/90

7/1/90

7/1/90

5/1/93

7/1/90

7/1/90

12/31/94

0483XW.o01 T of C5

/31/94wvision 3

A. INTRODUCTION

The purpose of this manual is to provide the parameters and methods to be usedin calculating offsite doses and effluent monitor setpoints at the Millstone NuclearPower Station. Included are methods for determining maximum individual wholebody and organ doses due to liquid and gaseous effluents to assure compliancewith the dose limitations in the Technical Specifications. Also included aremethods for performing dose projections to assure compliance with the liquid andgaseous treatment system operability sections of the Radiological EffluentMonitoring Manual. The manual also includes the methods used for determiningquarterly individual and population doses for inclusion in the Annual RadioactiveEffluent Report.

Another section of this manual discusses the methods to be used in determiningeffluent monitor alarm/trip setpoints to be used to ensure compliance with theinstantaneous release rate limits in the Technical Specifications.

The basis for some of the factors in this manual are included as appendices to thismanual. Supplemental information on environmental sample locations is providedin an additional appendix.

This manual does not include the surveillance procedures and forms required todocument compliance with the surveillance requirements in the TechnicalSpecifications. All that is included here are the methods to be used inperformance of the surveillance requirements.

Most of the calculations in this manual have several methods given for thecalculation of the same parameter. These methods are arranged in order ofsimplicity and conservatism, Method 1 being the easiest and most conservative.As long as releases remain low, one should be able to use Method I as a simpleestimate of the dose. If release calculations approach the limit, however, moredetailed yet less conservative calculations may be used At any time a moredetailed calculation may be used in lieu of a simple calculation. NUSCORadiological Assessment Branch may perform these more detailed calculations.

This manual is written common to all three units since some release pathways areshared and there are also site release limits involved. These facts make itimpossible to completely separate the three units.

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B. RESPONSIBILITIES

All changes to this manual shall be reviewed and approved by the SiteOperations Review Committee prior to implementation.

All changes and their rationale shall be documented in the AnnualRadioactive Effluent Report.

It shall be the responsibility of the Station Senior Vice President to ensurethat this manual is used in performance of the surveillance requirementsand administrative controls of the Technical Specifications.

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C.1 Quarterly - Whole Body Dose


Step I Determine CF which is total gross curies of fission andactivation products, excluding tritium and dissolved noblegases, released during the calendar quarter.

SteD 2 Determine CH which is total curies of tritium releasedduring the calendar quarter.

Step 3 Determine Dow which is the quarterly dose to the whole

body in mrem.

For Unit 1:

Dow = 1.0 CF + 5.6 x 107 CH

For Units 2 and 3:

Dow = 2 x 10.2 CF + 5.6 x 10.7 CH

Step 4 If DQw is greater than 0.5 mrem, go to Method 2.

Note: See Appendix A for derivation of these factors. ForUnit 1, the dose contribution from tritium can be neglectedsince it has never contributed to more than 2% of thewhole body doses.


If the calculated dose using Method 1 is greater than 0.5 mrem, usethe NRC computer code LADTAP II to calculate the liquid doses. Theuse of this code and the input parameters are given in RadiologicalAssessment Branch Procedure, Liquid Dose Calculations -LADTAP I1.

C.2 Quarterly - Maximum Orman Dose

a. Method 1 - Any Unit

Step I Determine CF which is total gross curies of fission andactivation products, excluding tritium and dissolved noblegases, released during the calendar quarter - same asStep C. I.a.

Steo 2 Determine DOO which equals the quarterly dose to the

maximum organ in mrem.

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For Unit 1:

Doo = 2.1 CF

For Units 2 and 3:

Doo = 0.2 CF

(See Appendix B for derivation of these factors)

Step 3 If D~o is greater than 2 mrem, go to Method 2.


If the calculated dose using Method 1 is greater than 2 mrem, use theNRC computer code LADTAP to calculate the liquid doses. The useof this code and the input parameters are given in RadiologicalAssessment Branch Procedure, Liquid Dose Calculations -LADTAP I1.

C.3 Annual - Whole Body Dose - Any Unit

Determine Dyw which equals dose to the whole body for the calendar yearas follows:

Dw = I Dow, where the sum is over the first quarter through the present

quarter whole body doses.

The following should be used as Dow:


(2) If the detailed calculations are not complete for a particular quarter,use the results as determined in Section C. 1.

(3) If Dyw is greater than 3 mrem and any DQw determined as inSection C. I was not calculated using Method 2 of that section,recalculate Dow using Method 2 if this could reduce Dyw to less than3 mrem.

C.4 Annual - Maximum Oraqan Dose - Any Unit

Determine Dyo which is the dose to the maximum organ for the calendaryear as follows:

Iyo = T. Doo, where the sum is over the first quarter through the presentquarter maximum organ doses.

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The following guidelines should be used:


(2) If the detailed calculations are not complete for a particular quarter,use the results as determined in Section C.2.

(3) If different organs are the maximum for different quarters, they maybe summed together and Dyo can be recorded as a less than value aslong as the value is less than 10 mrem.

(4) If Dyo is greater than 10 mrem and any value used in its determinationwas calculated as in Section C.2, but not with Method 2, recalculatethat value using Method 2 if this could reduce DIo to less than 10mrem.

C.5 Monthly Dose Proiections

a. Whole Body and Maximum Organ - Unit I

Step I Determine D'mw which is the whole body dose from thelast typical* previously completed month as calculated perthe methods in Section C. 1.

Ste_ 2 Determine D'MO which is the maximum organ dose fromthe last typical* previously completed month as calculatedper the methods in Section C. 1.

Stev 3 Estimate R, which is the ratio of the total estimatedvolume of liquid batches to be released in the presentmonth to the volume released in the past month.

Step4 Estimate R2 which is the ratio of estimated primary coolantactivity for the present month to that for the past month.

Step 5 Determine F which is the factor to be applied to theestimated ratio of final curies released if there areexpected differences in treatment of liquid waste for thepresent month as opposed to the past month (e.g.,bypass of filters or demineralizers). NUREG-0016 or pastexperience should be used to determine the effect ofeach form of treatment which will vary. F = 1 if there areno expected differences.

Step 6 Determine DE MW which is the estimated monthly wholebody dose as follows:

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DEMw = D'MW - R1 - R2 * F

Step 7 Determine DEMo which is the estimated monthly maximumorgan dose as follows:

DE M = D'MO * R, * R2 * F

S - The last typical month should be one without significant

operational differences from the projected month.

For example, if the plant was down for refueling the entiremonth of February and startup is scheduled for March 3,use the last month of operation as the base month toestimate March's dose.

Or, if there were no releases during September, do notuse September as the base month for October if it isestimated that there will be releases in October.

If the last typical month's doses were calculated usingLADTAP II (or similar methodology), also multiply theLADTAP doses by R5 where R5 = total dilution flow fromLADTAP run divided by estimated total dilution flow.

b. Whole Body and Maximum Organ - Unit 2 and Unit 3

Step I Determine D'mw which is the whole body dose from thelast typical* previously completed month as calculated perthe methods in Section C. 1.

Step 2 Determine D'MO which is the maximum organ dose fromthe last typical* previously completed month as calculatedper the methods in Section C.2.

S - See footnote in Section C.5.a.

Step 3 Estimate R1 which is the ratio of the total estimatedvolume of liquid batches to be released in the presentmonth to the volume released in the past month.

Step 4 Estimate R2 which is the ratio of estimated volume ofsteam generator blowdown to be released in presentmonth to the volume released in the past month.

Step 5 Determine F, which is the fraction of curies released lastmonth coming from steam generator blowdown.

i.e. F = curies from blowdowncuries from blowdown + curies from batch tanks

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Step 6 Estimate R3 which is the ratio of estimated secondarycoolant activity for the present month to that for the pastmonth.

Step 7 Estimate R4 which is the ratio of estimated primary coolantactivity for the present month to that for the past month.

Step 8 Determine F2 which is the factor to be applied to theestimated ratio of final curies released if there areexpected differences in treatment of liquid waste for thepresent month as opposed to the past month (e.g.,bypass of filters or demineralizers). NUREG-0017 or pastexperience should be used to determine the effect ofeach form of treatment which will vary. F2 = 1 if there areno expected differences.

SteD 9 Determine DEMw which equals estimated monthly totalbody dose as follows:

D EMW = D'Mw [(1 - F1) R1 R4 F2 + F1 R2 R3]

Step 10 Determine DE Mo which equals estimated monthlymaximum organ dose as follows:

DEMO = D'MO [(1 - F1) R1 R4 F2 + F1 R2 R3]

C.6 Quarterly Dose Calculations for Annual Radioactive Effluent Report

Detailed quarterly dose calculations required for the Annual RadioactiveEffluent Report shall be done using the NRC computer code LADTAP II.The use of this code, and the input parameters are given in RadiologicalAssessment Branch Procedure Liquid Dose Calculations - LADTAP II.

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Determine DQ. which equals the quarterly beta air dose fromUnit 3 (mrad).

D0B = 1.7 x 10"3 CN3

.t.. O If DoI, DOG2, or DOW are greater than 1.6 mrad; or Dmel, Dos2,or DB are greater than 3.3 mrad, go to Method 2.

S - See Appendix D for derivation of factors.

b. . Q•tarZ irly Air Dose - Method 2 - All Units

Unit 1 For MIVl dose calculations, use the AIREM computer code to determine thecritical location air doses.

The 3rd quarter 1980 joint frequency data should be used asinput for the AIREM code. The reason for this is given inAppendix D.

If the calculated air dose exceeds one-half the TechnicalSpecification limit, use real-time meteorology.

Units 2. 3 For MP2 and MP3 dose calculations use the GASPAR computer code todetermine the critical site boundary air doses.

For the Special Location, enter the following worst casequarterly average meteorology:

X/Q = 0.81 x 10"5 sec/m3

(See Appendix D)D/Q=O0.15 x10"4 m2z

If the calculated air dose exceeds one-half the quarterlyTechnical Specification limit, use real-time meteorology.

c. Annual Air Dose Limit Due to Noble Gases - All Units

Determine DYG1, DYG2, DYG3, D'y 1, Dym and Dym = gamma air dose and betaair dose for the calendar year for Unit 1, 2, or 3 as follows:

DyG1 = T.Da~l; DyB1 = T.Do 1 ; DYG2 = I£DoG2; Dm5 = EDcz; DYG3 = Y.Dow;

Dym = IDm.

where the sum is over the first quarter through the present quarter doses.

The following should be used as the quarterly doses:

(1) If the detailed quarterly dose calculations required per Section D.5 forthe Annual Radioactive Effluent Report are complete for any calendarquarter, use those results.

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b. Quarterly Doses - Uni. •.ad Unit 3

(1) Method I - Unit 2 and Unit 3

Step I Determine 131C,, which is the total curies of 1-131, and 13C3, which is thetotal curies of 1-133 in gaseous effluents from Unit 2 or 3 during thequarter.

SteD 2 Determine C,, which is the total curies of particulates with half-livesgreater than eight days released in gaseous effluents from Unit 2 or 3during the calendar quarter.

Step 3 Determine CH, which is the total curies of tritium released in gaseouseffluents from Unit 2 or 3 during the calendar quarter.

Step 4 Determine DOT which equals the quarterly thyroid dose as follows:

DOT = 3.1 X 10 3 131CI + 29.53 133C, + 2.6 x 10-3 CM

Step 5 Determine D0o which equals the quarterly dose to the maximum organother than the thyroid:

Doo = 1.1 x 103 Cp + 2.6 x 10"3 CH

Step 6 The maximum organ dose is the greater of DOT or D0o. If greater than2.5 mrem, go to Method 2.

(2) Method 2 - Unit 2 and Unit 3

Doses from vegetation consumption can be neglected during the 1st and 4thquarters and doses from milk consumption can be neglected during the 1st quarter.These time frames can be extended for short term releases (batch releases andweekly continuous, if necessary) if it can be verified that the milk animals were noton pasture and/or vegetation was not available for harvest. Therefore, calculatedoses to the thyroid and maximum organ for pathways that actually exist. Sumpathways, if necessary.

Perform Steps I through 3 as in Method 1, then:

Step 4

I. Inhalation PathwayDOT = 4.1 131CI + 1.0 13CI + 3.3 x 10" CmDoo = 4.1 CP + 3.3 x 10"4 CH

ii. Vegetation PathwayDOT = 105 131CI + 1.9 M=CI + 1.0 x 10-3 CHDoo = 124 Cp + 1.0 x 10"3 CH

iii. Milk PathwayDOT = 3000 131CQ + 26.6 13C,+ 1.3 x 104 CHDoo = 951 Cp + 1.3 x 10-3 Cm

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D.3 10 CFR5O Appendix I - Iodine no' ':iculate Doses

c. Annual Doses - All Units

Determine DyTj, Dr 2, DYT3, Iy1, LDY•2, and DY03 which are the thyroid andmaximum organ doses for the c .!-3ndar year for Units 1, 2, and 3respectively, as follows:

DyT, 2, or 3 = 2DQT = sum of the -.uarterly thyroid doses where the sum isover the first quarter through the ,)resent quarter.

DYO1, 2, or 3 = Woo = sum of the quarterly maximum organ doses where the

sum is over the first quarter through the present quarter.

The following guidelines should be used for DOT and DOO:

(1) If the detailed quarterly dose calculations required per Section D.5 forthe Annual Radioactive Effluent Report are complete for any calendarquarter, use those results.

(2) If the detailed calculations are not complete for a particular quarter,use the results as determined above in Section D.3.a or D.3.b.

(3) If Dy, and/or Dyo are greater than 15 mrem and quarterly dose wasnot calculated using Method 3 of Section D.3.a or D.3.b, recalculatethe quarterly dose using Method 3.

(4) If different organs are the maximum organ for different quarters, theycan be summed together and Dyo recorded as a less-than value aslong as the value is less than 15 mrem. If it is not, the sum for eachorgan involved should be determined.

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ii. Method 2

If necessary, estimate the curies expected to be released forthe next month and applicable method for dose calculation fromSection D.3.b.

(3) Due to Ventilation Releases*

Use the same method as given in Section D.4.a.(2).

c. Unit 3

(1) Due to Radioactive Gaseous Waste System

Use the same method as given in Section D.4.b.(1). However,instead of waste storage tanks use the reactor plant gaseous vents.The activity from this pathway increases when the process waste gassystem is out of service.

(2) Due to Steam Generator Blowdown Tank Vent

Use the same method as given in Section D.4.b.(2).

(3) Due to Ventilation Releases*

Use the same method as given in Section D.4.a.(2).

D.5. Quarterly Dose Calculations for Annual Radioactive Effluent Report

Detailed quarterly dose calculations required for the Annual Radioactive EffluentReport shall be done using the computer codes GASPAR and AIREM.

*Since dose projections are only required if the treatment specified in Section D of theRadiological Effluent Monitoring Manual are not operating, the monthly gamma and betaair dose projections are not required for ventilation releases.

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D.6. Compliance with 40CFRI 90

The following sources should be considered in determining the total dose to a real individual fromuranium fuel cycle sources:

a. Gaseous Releases from Units 1, 2, and 3.b. Liquid Releases from Units 1, 2, and 3.c. Direct Radiation from the Site.

Calculations and detailed surveys* indicate that typically the only significant contributor tooff-site exposure from this source is "Skyshine" from the Unit 1 Turbine Building. Thelocation of maximum dose is that of the critical fisherman. Listed below are theassumptions used for the calculation of these doses:

"CALCULATION OF SKYLINE CONTRIBUTION TO CRITICAL FISHERMAN(1) Based upon data obtained by Don Landers (MP Env. Lab) from the State of CT

Department of Environmental Protection (DEP) records on lobster catches:

Annual average of 3.5-4.5 days between trips to each lobster basket.

(2) Therefore, there are 104 trips per year.

(3) Conservatively, assuming it takes one hour in the area to check all the baskets, thisresults in 104 hours around the intake structures areas.

(4) Maximum dose rate in the area is normally 65 plR/hr.

(5) Average dose rate is approximately one-half of the maximum.

(6) Therefore, annual dose to critical lobsterman is approximately 104 hours/year x65 p.R/h x 1/2 = 3.4 mrem.

(7) Therefore, dose/month =

3.4 mrem x year x Unit 1 Capacity Factor = 0.3 mrem x Unit 1 Capacity Factoryear 12 months month

With the closing of the low level waste burial sites in 1994, the required onsitestorage may result in site boundary doses due to solid waste storage comparable toMP1 Skyshine. The Radiological Assessment Branch will perform an annualassessment of the maximum site boundary dose from radwaste storage.

d. Since all other uranium fuel cycle sources are greater than 5 miles away, they need not beconsidered.

References:

Memo to P. L. Tirinzoni from J. W. Doroski and C. A. Flory, Skyshine Evaluation at Millstone Unit#1, NE-87-RA-1033, December 8,1987This should be the most limiting individual since it is expected that even though fishermen mayspend more time near the area, they normally fish in an area of -1 p.R/hr.

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E. LIQUID MONITOR SETPOINTS

E.1 Unit I Liquid Radwaste Effluent Line

The trip/alarm setting on the Unit I liquid radwaste discharge line dependson dilution water flow, radwaste discharge flow, the isotopic composition ofthe liquid, the background count rate of the monitor and the efficiency of themonitor. Due to the variability of these parameters, an alarm/trip setpoint willbe determined prior to the release of each batch. The following method willbe used:

Step I From the tank isotopic analysis and the MPC values for eachidentified nuclide (including noble gases*) determine therequired reduction factor, i.e.:

For Nuclides Other Than Noble Gases*:

R, = Required Reduction Factor = 1/Z uCi/ml of nuclide iMPC of nuclide i

For Noble Gases:

R2 = Required Reduction Factor = 1/Z u-Ciml of noble gases2 x 104 pCilml

= 2 x 1O4//Z iCilml of noble gases

R = the smaller of R, or R2

*In lieu of determining the required reduction factor for noble gases,conservatism is allowed. For example, calculate the maximumconcentration of noble gases that can be discharged from any tank.

Assuming:Maximum discharge rate = 350 gpm

Minimum dilution flow = 110,000 gpm (1 circulating pump &I service water pump)

Maximum noble gas x 350 apm = 2 x 10-4 iCimlConcentration 110,000 gpm

Therefore,Maximum concentration = 0.063 IiCi/ml

Step 2 Determine the existing dilution flow which is DD = #circulating water pumps x 100,000 gpm + #service waterpumps x 10,000 gpm

Step 3 Determine the allowable discharge flow which is FF=0.1 xRxD

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Note that discharging at this flow rate would yield a dischargeconcentration corresponding to 10% of the TechnicalSpecification Limit due to the safety factor of 0. 1.

Steg4 Determine the total giCi/ml in the tank.

Steop 5 Using the current calibration response factor, determine the"cps" corresponding to two times the total jLCi/ml determined in

Step 4 (Note 1). This value or that corresponding to9.4 x 10"5 RCi/ml (Note 2), whichever is greater, plusbackground is the trip setpoint For the latter setpoint,independent valve verification should be performed.

Note 1: If discharging at the allowable discharge rate asdetermined in Step 3, this would yield a dischargeconcentration corresponding to 20% of theTechnical Specification limit.

Note 2: This value is based upon worst case conditions,assuming maximum discharge flow (350 gpm),minimum dilution water flow (110,000 gpm for MP1)and an assumed worst case mix of nuclides(3 x 10-). This value may be increased by factorsto account for the actual discharge flow and actualdilution flow. This will assure that low level releasesare not terminated due to small fluctuations inactivity. However, to verify that the correct tank isbeing discharged when using this value,independent valve verification should be performed.Administrative controls should be established toensure that the allowable discharge flow is notexceeded and the dilution flow is maintained.

Step 6 This allowable discharge flow rate calculated in Step 3 may beincreased by up to a factor of 5 with appropriate administrativecontrols (e.g., for Unit 2, insure other release points may notcause MPCs to be exceeded).

E.2 Unit I Reactor Building Service Water Effluent Line

The MP1 Reactor Building Service Water Monitor is approximately two timesthe ambient background reading on the monitor in counts per second.

E.3 Unit 2 Clean Liquid Radwaste Effluent Line

Similar to the Unit I liquid discharge line, the setpoints on the Unit 2 liquidwaste effluent line depend on dilution water flow, radwaste discharge flow,the isotopic composition of the liquid, the background count rate of themonitor and the efficiency of the monitor. Due to the variability of theseparameters, an alarm/trip setpoint will be determined prior to the release ofeach batch. The following method will be used:

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Step 1 From the tank isotopic analysis and the MPC values for eachidentified nuclide (including noble gases*) determine therequired reduction factor, i.e.:


R, = Required Reduction Factor = 1/Z uCi/ml of nuclide iMPC of nuclide i

For Noble Gases:

R2= Required Reduction Factor = 1/Z uCilmi of noble oases2 x 10' p.ci/ml

= 2 x 10-41z pCi/ml of noble gases

R = the smaller of R1 or R2

*In lieu of determining the required reduction factor for noble gases,conservatism is allowed. For example, calculate the maximumconcentration of noble gases that can be discharged from any tank.


Minimum dilution flow = 274,000 gpm (2 circulating pumps &I service water pump)

Maximum noble gas x 350 ,pm =2 x 10"4 itCi/mlConcentration 274,000 gpm

Therefore,Maximum concentration = 0.15 i.Ci/ml


Step 3 Determine the allowable discharge flow which is FF=0.1xRxDNote that discharging at this flow rate would yield a dischargeconcentration corresponding to 10% of the TechnicalSpecification Limit due to the safety factor of 0.1.

Step 4 Determine the total igCi/ml in the tank.

Step 5 Using the current calibration response factor, determine the"cpm" corresponding to two times the total iCi/ml determined inStep 4 (Note 1). This value or that corresponding to 2.2 x10"4 igCi/ml (Note 2), whichever is greater, plus background is

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the trip setpoint. For the latter setpoint, independent valveverification should be performed.

Note 1: If discharging at the allowable discharge rate asdetermined in Step 3, this would yield a dischargeconcentration corresponding to 20% of theTechnical Specification limit.

Note 2: This value is based upon worst case conditions,assuming maximum discharge flow (350 gpm),minimum dilution water flow (274,000 gpm for MP2)and an assumed worst case mix of nuclides(3 x 10i7 - 1-131 MPC). This will assure that lowlevel releases are not terminated due to smallfluctuations in activity. However, to verify that thecorrect tank is being discharged when using thisvalue, independent valve verification should beperformed. This value may be increased by factorsto account for the actual discharge flow and actualdilution flow; however, controls should beestablished to ensure that the allowable dischargeflow is not exceeded and the dilution flow ismaintained.

Step 6 This allowable discharge flow rate calculated in Step 3 may beincreased by up to a factor of 5 with appropriate administrativecontrols (e.g., for Unit 2, insure other release points may notcause MPCs to be exceeded).

E.4 Unit 2 Aerated Liquid Radwaste Effluent Line and Condensate PolishingFacility Waste Neutralization Sump Effluent Line

Same as E.3 for Clean Liquid Monitor and the Condensate Polishing Facility(CPF) Waste Neutralization Sump monitor except the CPF monitor has thecapability to readout in CPM or i±Ci/ml. For the CPF Waste NeutralizationSump monitor, use a default setpoint if no chemistry grab samples arerequired. This default should be the lower of: two times background or thevalue as specified in E.3.

E.5 Unit 2 Steam Generator Blowdown

Assumptions used in determining the Alarm setpoint for this monitor are:

a. Maximum possible total S.G. blowdown flow rate = 600 gpm

b. Minimum possible circulating water dilution flow during periods ofblowdown = 270,000 gpm (2 circulating water pumps) + 4,000 gpm(1 service water pump) = 274,000 gpm.

c. The release rate limit is conservatively set at 50% of the IOCFRPart 20 limit for 1-131 (0.5 x 3 x 10.7 piCi/ml = 1.5 x 10.7 p.Ci/ml)*.

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d. Background can be added after above calculations are performed.

Therefore, the alarm setpoint should correspond to a concentration of:

Alarm (p±Ci/ml) = 274.000 x 1.5 x 10' + background" = 6.8 x 10"s tCi/ml + background600

The latest monitor calibration curve should be used to determine the alarmsetpoint in cpm corresponding to 6.8 x 10s pICi/ml.

This setpoint may be increased through proper administrative controls if thesteam generator blowdown rate is maintained less than 600 gpm and/ormore than 2 circulating water pumps and 1 service water pump areavailable. The percent increase would correspond to the ratio of flows tothose assumed above or:

Alarm (pCi/ml) = 6.8 x 10"5 pCi/mix circulating & service water flow (aem) x 600274,000 SIG Blowdown (gpm)

+ Background = 1.5 x 10.7 LCi/m x circulatina & service water flow (aom) + Backgroundtotal S/G Blowdown (gpm)

Note: The Steam Generator Blowdown alarm criteria is in practice based onsetpoints required to detect allowable levels of primary to secondaryleakage. This alarm criteria is typically more restrictive than that required tomeet discharge limits. This fact should be verified, however, whenever thealarm setpoint is recalculated.

*In lieu of using the 1-131 MPC value, the identified MPC values forunrestricted area may be used.

"Background of monitor at monitor location (i.e., indication provided by

system monitor with no activity present in the monitored system)

E.6 Unit 2 Condenser Air Eiector

This monitor is included as a liquid monitor since the reason it's in theTechnical Specifications is for control of the steam generator blowdownliquid activity. In 1993 it was realized that the air ejector monitor cannotadequately fulfill this function. A Technical Specification Change Requesthas been submitted to delete this monitor from the Technical Specifications.Until that time, this monitor should be considered inoperable as the backupfor the S.G. Blowdown Monitor and, hence, the action statement forsampling would be required if the blowdown monitor is out of service.Therefore, there is no need to specify the Alarm setpoint requirements inthis section.

E.7 Unit 2 Reactor Building Closed Cooling Water

Assumptions used in determining the Alarm setpoint for this monitor are:

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a. Maximum flow from primary makeup water is 400 gpm.

b. Minimum circulating water dilution flow is 135,000 gpm (1 circulatingwater pump).

c. The release rate is conservatively set at 50% of the IOCFR Part 20limit for 1-131 (0.5 x 3 x 10i7 p Ci/ml = 1.5 x 10-7 pCi/ml).

d. Background can be added after the above calculations are performed.

Therefore, the alarm setpoint (using the latest monitor calibration curve)should correspond to a concentration of:

Alarm (jiCi/ml) = 135,000/400 x 1.5 x 10-7 + background*= 5 x 10'5 RCi/ml + background

Note that the purpose of this monitor is to detect high activity that may occurbetween the weekly composite RBCCW samples. Hence, the maximumundetected dose consequence, assuming an unlikely 400 gpm leak, is:

5 x 10-5 giCi/mo x 400 gal/min x 168 hr/week x 60 min/hr x

3785 cc/gal x Ci/10 6 1±Ci = 0.8 Ci

0.8 Ci x 0.2 mrem/Ci = 0.16 mrem maximum organ

This dose is below limits and is an event that should barely, if ever, happenover the life of the plant.

Monitor background at monitor location.

E.8 Unit 3 Liquid Waste Monitor

Similar to the Unit 1 liquid discharge line, the setpoints on the Unit 3 liquidwaste monitor depend on dilution water flow, radwaste discharge flow, theisotopic composition of the liquid, the background count rate of the monitorand the efficiency of the monitor. Due to the variability of these parameters,the alert and alarm setpoints will be determined prior to the release of eachbatch. The following method will be used:

Step I From the tank isotopic analysis and the MPC values for eachidentified nuclide (including noble gases*) determine therequired reduction factor, i.e.:


R, = Required Reduction Factor = 1I/ uaCi/ml of nuclide iMPC of nuclide i

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For Noble Gases*:

R= Required Reduction Factor = 1/E uCi/mi of noble aases2 x 10-4 pCi/ml

= 2 x 104/1 pCi/ml of noble gases

R = the smaller of R1 or R2

*In lieu of determining the required reduction factor for noble gases,

conservatism is allowed. For example, calculate the maximumconcentration of noble gases that can be discharged from any tank.


Minimum dilution flow = 300,000 gpm (2 circulating pumps)

Maximum uallowable" x 150 am = 2 x 10-4 I.Ci/mIConcentration 300,000 gpm

Therefore,Maximum "allowablen concentration = 0.4 g.Ci/ml


Step 3 Determine the allowable discharge flow which is FF = 0.1 x R x DNote that discharging at this flow rate would yield a dischargeconcentration corresponding to 10% of the TechnicalSpecification Limit due to the safety factor of 0.1.

Step 4 Determine the total ipCi/mI in the tank.

Step 5 The alarm setpoint will be two times the total gCi/ml determinedin Step 4 (Note 1) or 2.2 x 104 RCi/ml (Note 2), whichever isgreater, plus background.

Note 1: If discharging at the allowable discharge rate asdetermined in Step 3, this Alarm setpoint wouldyield a discharge concentration corresponding to20% of the Technical Specification limit.

Note 2: This value is based upon worst case conditions,assuming maximum discharge flow (150 gpm),minimum dilution water flow (2 circulating pumpsand 2 service water pumps = 334,000 gpm), and anassumed mix of nuclides as specified for anunidentified liquid release in IOCFR20

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(1 x 10-7 gCi/ml). This will assure that low levelreleases are not terminated due to smallfluctuations in activity. However, to verify that thecorrect tank is being discharged when using thisvalue, independent valve verification should beperformed. This value may be increased by factorsto account for the actual discharge flow and actualdilution flow; however, controls should beestablished to ensure that the allowable dischargeflow is not exceeded and the dilution flow ismaintained.

Step 6 The allowable discharge flow rate calculated in Step 3 may beincreased by up to a factor of 5 with appropriate administrativecontrols (to ensure other release points do not cause MPCs tobe exceeded).

E.9 Unit 3 Reqenerant Evaporator Effluent Line

The MP3 Regenerant Evaporator Monitor alarm setting is two times thenormal reading.

E.10 Unit 3 Waste Neutralization Sump Effluent Line

Same as Section E.8. Note that for this monitor, even though grab samplesmay not be required, setpoints still have to be utilized. In such cases, thedefault for low activity samples can be used.

E.11 Unit 3 Steam Generator Blowdown

The alarm setpoint for this monitor assumes:

a. Steam generator blowdown rate of 400 gpm (maximum blowdowntotal including weekly cleaning of generators - per 3-Part Memo fromMP3 Reactor Engineering).

b. The release rate limit is conservatively set at 10% of the IOCFRPart 20 limit (0.1 times the 1-131 MPC* for unrestricted areas whichequals 0.1 x 3 x 10"• Ci/ml).

c. Minimum possible circulating and service water dilution flow duringperiods of blowdown = 304,000 gpm (2 circulating water pumps) +30,000 gpm (2 service water pumps) = 334,000 gpm.

d. Background can be added after above calculations are performed.

Therefore, the alarm setpoint should correspond to a concentration of:

Alarm (I•Ci/m) = 334.000 x 3 x 108 background = 2.5 x 10"1 pCi/ml + background400

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This setpoint may be increased through proper administrative controls if thesteam generator blowdown rate is maintained less than 400 gpm and/ormore than 2 circulating and 2 service water pumps are available. Theamount of the increase would correspond to the ratio of flows to thoseassumed above or:

Alarm (iCi/ml) : 2.5 x 10"s pCi/ml x circulating & service water flow (aom) x 400334,000 gpm S/G Blowdown (gpm)

+ Background = 3 x 10-8 pLCi/ml x circulating & service water flow (aom) + Backgroundtotal S/G Blowdown (gpm)

Note: The Steam Generator Blowdown alarm criteria is in practice based onsetpoints required to detect allowable levels of primary to secondaryleakage. This alarm criteria is typically more restrictive than that required tomeet discharge limits. This fact should be verified, however, whenever thealarm setpoint is recalculated.

*In lieu of using the 1-131 MPC value, the identified MPC values forunrestricted area may be used.

E.12 Unit 3 Turbine Building Floor Drains Effluent Line

The alarm setpoint for this monitor assumes:

a. Drinking water is not a real pathway at this site. Therefore, the NRCcode, LADTAP, is used to calculate the dose to the maximumindividual.

b. The average annual discharge flow is 1.11 x 10-2 ft3/sec (process flowduring sump pump operation is 50 gpm and pump normally operatesless than 10% of the time for a conservative average flow of 5 gpm).There is no continuous additional dilution, therefore, discharge flow isconservatively assumed to equal dilution flow.

c. Near field dilution factor = 13,000.Far field dilution factor = 32,000.(Reference: Millstone 3 FSAR, Section 2.4.13)

d. Isotopic concentrations were taken from the Millstone 3 FSAR,Table 11.2-4 (See column under Turbine Building).

e. Each concentration above was multiplied by the total annual flow(9.95 x 109 cm 3, conservatively assuming 5 gpm continuous asdiscussed in item b).

f. The maximum individual organ dose is set equal to 1% of 75 mrem(40CFR190 limit). The limiting individual is the child; maximum organis the thyroid. This value is approximately one quarter of the valuerequiring operation of the liquid treatment system as specified inSection C.2 of the Radiological Effluent Monitoring Manual.

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The setpoint corresponding to 0.75 mrem to the child's thyroid is3.8 x 10.5 ACi/ml.

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ENVIRONMENTAL MONITORING PROGRAM

Samplina Locations

The following lists the environmental sampling locations and the types of samplesobtained at each location. Sampling locations are also shown on Figures G-1, G-2,and G-3.

LocationNumber

Direction & Distance FromRelease Point"*Name

1 _I*

2-13-14-15-16-17-18-19-110-111-112-C13-C14-C15-C16-C17-118-119-120-C21-I22-I23-I24-C25-126-C27-128-129-130-C31-1

On-site - Old Millstone RoadOn-site - Weather ShackOn-site - Bird SanctuaryOn-site - Aibacore DriveMP3 DischargeQuarry DischargeEnvironmental Lab DockEnvironmental LabBay Point BeachPleasure BeachNew London Country ClubFisher's Island, NYMystic, CTLedyard, CTNorwich, CTOld Lyme, CTSite BoundaryPleasure BeachCow Location #1Cow Location #2Goat Location #1Goat Location #2Goat Location #3Goat Location #4Fruits & VegetablesFruits & VegetablesNianticTwo Tree IslandWest Jordan CoveGolden SpurNiantic Shoals

0.6 Mi, NNW0.3 Mi, S

0.3 Mi, NE1.0 Mi, N

0.1 Mi, SSE0.3 Mi, SSE0.3 Mi, SE0.3 Mi, SE0.4 Mi, W1.2 Mi, E

1.6 Mi, ENE8.7 Mi, ESE

11.5 Mi, ENE12.0 Mi, NE14.0 Mi, N8.8 Mi, W0.5 Mi, NE1.2 Mi, E

9.5 Mi, WNW16.0 Mi, NNW

2.0 Mi, N5.2 Mi, NNE2.0 Mi, ENE

14.0 Mi, NNWWithin 10 MilesBeyond 10 Miles

1.7 Mi, WNW0.8 Mi, SSE0.4 Mi, NNE4.7 Mi, NNW1.8 Mi, NW

1.5 Mi, NNW

1.8 Mi, ESE4.0 Mi, ENE

0.3 Mi, WNW3.0 Mi, WSW3.5 Mi, WSW1.0 Mi, NW

Sample Tvyes

TLD, Air Particulate, Iodine, VegetationTLD, Air Particulate, IodineTLD, Air Particulate, IodineTLD, Air Particulate, IodineTLDTLDTLDTLDTLDTLD, Air Particulate, IodineTLD, Air Particulate, IodineTLDTLDTLDTLD, Air Particulate, IodineTLDVegetationVegetationMilkMilkMilkMilkMilkMilkVegetationVegetationTLD, Air Particulate, IodineMusselsClams

Bottom Sediment, OystersMusselsBottom Sediment, Oysters, Lobster,Fish, SeawaterBottom SedimentBottom SedimentLobster, FishOystersBottom Sediment, Oysters, SeawaterClams

32-1 Vicinity of Discharge

33-134-I35-136-137-C38-1

Seaside PointThames River Yacht ClubNiantic BayBlack PointGiant's NeckWaterford Shellfish Bed #1

*1 = Indicator **C = Control*"The release points are the MPI stack for terrestrial locations and the end of the quarry for aquatic locations.

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