ML101300353

Submittal of 2009 Radioactive Effluent Release Report

ML101300353
Person / Time
Site:	Millstone
Issue date:	04/27/2010
From:	Macmanus R Dominion Nuclear Connecticut
To:	Document Control Desk, NRC/FSME, Office of Nuclear Reactor Regulation
References
10-196, FOIA/PA-2011-0115
Download: ML101300353 (252)

Text

Dominion Nuclear Connecticut, Inc. .

Millstone Power Station Rope Ferry Road Waterford, CT 06385 APR 2 7 2010 U.S. Nuclear Regulatory Commission Serial No.10-196 Attention: Document Control Desk MPS Lic/GJC RO Washington, DC 20555-0001 Docket Nos. 50-245 50-336

50-423 License Nos. DPR-21

-DPR-65 NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNITS 1, 2. AND 3 2009 RADIOACTIVE EFFLUENT RELEASE REPORT In accordance with 10 CFR 50.36a, this letter transmits the annual Radioactive Effluent Release Report for the period January 2009 through December 2009. This report meets the provisions of Section 5.7.3 of the Millstone Power Station Unit 1 Permanently Defueled Technical Specifications (PDTS), and Sections 6.9.1.6b and 6.9.1.4 of the Millstone Power Station Units 2 and 3 Technical Specifications,. respectively. transmits Volume -1of the 2009 Radioactive Effluent Release Report, in accordance with Regulatory Guide 1.21. Volume 1 contains information regarding airborne, liquid, and solid radioactivity released from Millstone Power Station, off-site dose from airborne and liquid effluents, and a description of changes made to the Radioactive Effluent Monitoring and Off-Site Dose Calculation. Manual (REMODCM) during the year 2009. transmits Volume 2 of the report, which contains changes made to the REMODCM through 2009. Volume 2 consists of a complete copy of the REMODCM as of December 31, 2009, which satisfies the requirements of Sections 5.6.lc of the Millstone Power Station Unit 1 PDTS, and Sections 6:15c and 6.9.13c of the Millstone Power Station Units 2 and 3 Technical Specifications, respectively.

If you have any questions or require additional information, please contact Mr. William D. Bartron at (860) 444 4301.

Sincerely, R. K. MacManus Director, Nuclear Station Safety and Licensing A1,1556(

Serial No.10-196 2009 Radioactive Effluent Release Report Page 2 of 4 Attachments: 2 Commitments made in this letter: None.

cc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. J. B. Hickman NRC Project Manager Millstone Unit 1 U.S. Nuclear Regulatory Commission Mail Stop T-7E18 Washington, DC 20555 Ms. L. A. Kauffman NRC Inspector U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Mr. T. A. Moslak NRC Inspector U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406-1415 Ms. C. J. Sanders NRC Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8B3 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring & Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No.10-196 2009 Radioactive Effluent Release Report Page 3 of 4 Mr. James Chernaick Regional Radiation Representative (EPA Region 1, Boston)

U. S. Environmental Protection Agency (Region 1)

.John F. Kennedy Federal Building (ATR-2311)

Boston, MA 02203 Mr. Ronald E. Bernacki

• Department of Health and Human Services U. S. Food and Drug Administration Office of Regulatory Affairs Mortheast Region Field Office (NER-FO) 158-15 Liberty Ave.

.Jamaica, NY 11433 Mr. Daniel F. Caruso Chairman Connecticut Siting Council 10 Franklin Square New Britain, CT 06051 Mr. Pat Kelley Waterford-East Lyme Shellfish Commission Waterford Town Hall Waterford, CT 06385 Mr. Jason A. Martinez American Nuclear Insurers 95 Glastonbury Blvd.

Glastonbury, CT 06033 Mr. Dave Lamoureux Connecticut Department of Agriculture Aquaculture Division P. 0. Box 97 Millford, CT 06460 Mr. Dan Steward First Selectman Town of Waterford Waterford Town Hall Waterford, CT 06385 Mr. Paul Formica First Selectman Town of East Lyme PO Box 519 Niantic, CT 06357

Serial No.10-196 2009 Radioactive Effluent Release Report Page 4 of 4 Mr. Martin Berliner City Manager 181 State Street New London, CT 06320 University Of Connecticut Library Serials Department Storrs, CT 06268

Serial No.10-196 Docket Nos. 50-245 50-336 50-423 License Nos. DPR-21 DPR-65 NPF-49 ATTACHMENT I 2009 RADIOACTIVE EFFLUENTS RELEASE REPORT VOLUME I MILLSTONE POWER STATION UNITS 1, 2, AND 3 DOMINION NUCLEAR CONNECTICUT, INC. (DNC)

2009 Radioactive Effluents Release Report Volume 1 Dominion Nuclear Connecticut, Inc.

MILLSTONE UNIT LICENSE DOCKET 1 DPR-21 50-245 2 DPR-65 50-336

,Dominion 3 NPF-49 50-423

Table of Contents Volume 1 Table of Contents ................................................................................................... 1..

List of Tables.......................................................................................................... 2 References............................................................................................................. 3 Introduction............................................................................................................ 4 1.0 Off-Site Doses .................................................................................................... S 1.1 Dose Calculations.......................................................................................... 5 1.1.1 Airborne Effluents................................................................................ 5 1.1.2 Liquid Effluents.................................................................................. 6 1.2 Dose Results................................................................................................ 7 1.2.1 Airborne Effluents................................................................................ 7 1.2.2 Liquid Effluents.................................................................................. 7 1.2.3 Analysis of Results............................................................................... 7 2.0 Effluent Radioactivity ......................................................................................... 12 2.1 Airborne Effluents........................................................................................ 12

2. 1.1 Measurement of Airborne Radioactivity...................................................... 12 2.1.2 Estimate of Errors............................................................................... 14 2.1.3 Airborne Batch Release Statistics ............................................................. 14 2.1.4 Abnormal Airborne Releases .................................................................. 14 2.2 Liquid Effluents.......................................................................................... 34 2.2.1 Measurement of Liquid Radioactivity ........................................................ 34 2.2. 1.1 Continuous Liquid Releases ........................................................... 34 2.2.1.2 Liquid Tanks/Sumps..................................................................... 34 2.2.2 Estimate of Errors .............................................................................. 35 2.2.3 Liquid Batch Release Statistics................................................................ 35 2.2.4 Abnormal Liquid Releases ..................................................................... 35 2.2.5 Liquid Release Tables.......................................................................... 35 2.3 Solid Waste............................................................................................... 50 2.4 Groundwater Monitoring ................................................................................. 62 3.0 Inoperable Effluent Monitors ................................................................................. 66 4.0 Operating History............................................................................................... 67 5.0 Errata............................................................................................................ 70 6.0 REMODCM Changes'..'............",.... ................... 70 6.1 REMODCM Description of Changes................................................................. 71 Volume 2 2009 REMODCM Revision 26-01

List of Tables Table 1-1 Off-Site Dose Summary from Airborne Effluents - Units 1, 2, 3........................... ................... 8 Table 1-2 Off-Site Dose Summary from Liquid Effluents - Units 1, 2, 3 .............. ............................................................. 9 Table 1-3 Off-Site Dose Comparison to Limits - Units 1, 2, 3 ......................................................................................... 10 Table 1-4 Off-Site Dose Comparison - Units 1, 2, 3 ...................................................................................................... II Table 2. I-Al Unit I Airborne Effluents - Release Summary ................................................................................................ 16 Table 2.1 -A2 Unit I Airborne Effluents - Ground Continuous - BOP Vent & SFPI Vent .............................................................. 17 Table 2.2-Al Unit 2 Airborne Effluents - Release Summary ................................................................................................ 18 Table 2.2-A2 Unit 2 Airborne Effluents - Mixed Continuous - Aux Bldg Vent, SGBD Tank Vent & Spent Fuel Pool Evaporation .............. 19 Table 2.2-A3 Unit 2 Airborne Effluents - Mixed / Elevated Batch - Containment Purges ......................................................... .... 20 Table 2.2-A4 Unit 2 Airborne Effluents - Elevated Batch - WGDT ...................................................................................... 21 Table 2.2-A5 Unit2 Airborne Effluents - Elevated Continuous - Containment Vents/Site stack ..................................................... 22 Table 2.2-A6 Unit 2 Airborne Effluents - Ground Batch - Containment Equipment Hatch ............................................................ 23 Table 2.2-A7 Unit 2 Airborne Effluents - Ground Batch - RWST Vent .................................................................................. 24 Table 2.3-Al Unit 3 Airborne Effluents - Release Summary ............................................................................................... 25 Table 2.3-A2 Unit 3 Airborne Effluents - Mixed Continuous - Normal Ventilation & Spent Fuel Pool Evaporation ............................. 26 Table 2.3-A3 Unit 3 Airborne Effluents - Ground Continuous - ESF Building Ventilation ............................................................ 27 Table 2.3-A4 Unit 3 Airborne Effluents - Mixed Batch - Containment Drawdowns .................................................................... 28 Table 2.3-A5 Unit 3 Airborne Effluents - Mixed Batch - Containment Purges ............................................................................. 29 Table 2.3-A6 Unit 3 Airborne Effluents - Elevated Continuous - Gaseous Waste System ............................................................ 30 Table 2.3-A7 Unit 3 Airborne Effluents - Elevated Batch - Containment Vents ........................................................................ 31 Table 2.3-A8 Unit 3 Airborne Effluents - Ground Batch - Containment Equipment Hatch ............................................................ 32 Table 2.3-A9 Unit 3 Airborne Effluents - Ground Batch- RWST Vent ................................................................................. 33 Table 2. I-LI Unit I Liquid Effluents - Release Summary - (Release Point - Quarry) ..................................................................... 36 Table 2.l1--L2 Unit I Liquid Effluents - Batch - (Release Point - Quarry) ............................... ............................................ 37 Table 2.2-LI Unit 2 Liquid Effluents - Release Summary - (Release Point - Quarry) ..................................................................... 38 Table 2.2-L2 Unit 2 Liquid Effluents - Continuous - SGBD, SW, RBCCW - (Release Point - Quarry) ............................................ 39 Table 2.2-L3 Unit 2 Liquid Effluents - Batch - LWS - (Release Point - Quarry) ........................................................................ 40 Table 2.2-L4 Unit 2 Liquid Effluents - Release Summary - (Release Point - Yard Drain DSN 006) .................................. 41 Table 2.2-L5 Unit 2 Liquid Effluents - Continuous - Turbine Building Sump - (Release Point - Yard Drain DSN 006) ............................ 42 Table 2.3-LI Unit 3 Liquid Effluents - Release Summary - (Release Point - Quarry) ..............................................  :...................... 43 Table 2.3-L2 Unit 3 Liquid Effluents - Continuous - SGBD, SW - (Release Point - Quarry) .......................................................... 44 Table 2.3-L3 Unit 3 Liquid Effluents - Batch - LWS - (Release Point - Quarry) ........................................................................ 45 Table 2.3-L4 Unit 3 Liquid Effluents - Batch - CPF Waste Neut Sumps, Hotwell, S/G Bulk - (Release Point - Quarry) ............................ 46 Table 2.3-L5 Unit 3 Liquid Effluents - Release Summary - (Release Point- Yard Drain DSN 006) ................................................. 47 Table 2.3-L6 Unit 3 Liquid Effluents - Continuous - T B Sump, WTT Berm - (Release Point - Yard Drain DSN 006) ............................. 48 Table 2.3-L7 Unit 3 Liquid Effluents - Continuous - Foundation Drain Sumps - (Release Point - Yard Drain DSN 006) ........................ 49 Table 2. 1-S Unit I Solid Waste & Irradiated Component Shipments ................................................................................... 51 Table 2.2-S Unit 2 Solid Waste & Irradiated Component Shipments .................................................................................... 54 Table 2.3-S Unit 3 Solid Waste & Irradiated Component Shipments ................................................................................... 58 Table 2.4-GWI Environmental Well Monitoring Results ......................................................................................................... 63 Table 2.4-GW2 Catch Basin/Underdrain Monitoring Results ................................................................................................. 64 Table 2.4-G W3 U nderdrain M onitoring Results ................................................................................................................... 65 2

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 Nuclear Reactor Regulation, US Nuclear Regulatory Commission, Washington, DC 20555, manuscript completed January 1980, published June 1980.

2. Memo - MP-CHEM-10-008, 2009 Radioactive Effluent Da April 6, 2010.

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

4. Technical Evaluation M3-EV-09-0003, Impact of Faulty Flow Indication from HVR-FT10, April 1, 2009.

5. NRC Regulatory Guide 1.111 Methods for Estimating Atmospheric Transport and Dispersion 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 Computer Program for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents, DB Simpson, BL McGill, prepared by Oak Ridge National Laboratory, Oak Ridge, TN 37830, for Office of Administration, US Nuclear Regulatory Commission, manuscript completed 17 March 1980.

7. 10 CFR Part 50 Domestic Licensing of Production and Utilization Facilities, Appendix I Numerical Guides for Design Obiectives and Limiting Conditions for Operation to Meet the Criterion "As Low As Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents.

8. 40 CFR Part 190 Environmental Radiation Protection Standard for Nuclear Power Operation.

9. DOSLIQ-Dose Excel Code for Liquid Effluents, Software Document File, Rev 1, February 2002.

10. DOSAIR-Dose Excel Code for Airborne Effluents, Software Document File, Rev 0, February 2002.

11, GASPAR II - Technical Reference and User Guide (NUREG/CR-4653), March 1987.

12, NEI 07-07 Industry Ground Water Protection Initiative - Final Guidance Document, August 2007.

13. Memo - NSE-10-002, Millstone Effluent Monitors Out Of Service in 2009, February 11, 2010.

14. Memo - MP-HPO-10014, 2009 Report on Solid Waste and Irradiated Component Shipments March 15, 2010.

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Introduction This report, for the period of January through December of 2009, is being submitted by Dominion Nuclear Connecticut, Inc. for Millstone Power Station's Units 1, 2, and 3, in accordance with IOCFR50.36a, the REMODCM, and the Station's Technical Specifications. A combined report, written in the US NRC Regulatory Guide 1.21 format, is submitted for all three units.

Volume 1 contains radiological and volumetric information on airborne and liquid effluents, shipments of solid waste &

irradiated components, calculated offsite radiological doses, all changes to the REMODCM, information on effluent monitors inoperable for more than 30 consecutive days, and corrections to previous reports. Volume 2 contains a full copy of each of the complete revisions to the REMODCM effective during the calendar year.

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1.0 Off-Site Doses This report provides a summary of the 2009 off-site radiation doses from releases of radioactive materials in airborne and liquid effluents from Millstone Units 1, 2, and 3. This includes the annual maximum dose (mrem) to any real member of the public as well the maximum gamma and beta air doses.

To provide perspective, these doses are compared to the regulatory limits and to the annual average dose that a member of the public could receive from natural background and other sources.

1.1 Dose Calculations The off-site dose to humans from radioactive airborne and liquid effluents have been calculated using measured radioactive effluent data, measured meteorological data, and the dose computer models DOSAIR and DOSLIQ, which were developed by Millstone. The methodology and input parameters for DOSAIR are those used in GASPAR I1 (Reference 12) and NRC Regulatory Guide 1.109 (Reference 3). The methodology and input parameters for DOSLIQ are those used in LADTAP II (Reference 6) and NRC Regulatory Guide 1.109 (Reference 3). The calculated doses generally tend to be conservative due to the conservative model assumptions. More realistic estimates of the off-site dose can be obtained by analysis of environmental monitoring data. A comparison of doses estimated by each of the above methods is presented in the Annual Radiological Environmental Operating Report.

Doses are based upon exposure to the airborne and liquid effluents over a one-year period and an associated dose commitment over a 50-year period from initial exposure. The portion of the doses due to inhalation and ingestion take into account radioactive decay and biological elimination of the radioactive materials.

Maximum individual dose is defined as the dose to the individual who would receive the maximum dose from releases of airborne and liquid effluents. Although the location of the maximum individual may vary each quarterly period, the annual dose is the sum of these quarterly doses. This conservatively assumes that the individual is at the location of maximum dose each quarter.

The dose calculations are based upon three types of input: radioactive source term, site-specific data, and generic factors. The radioactive source terms (Curies) are characterized in Section 2, Effluent Radioactivity, of this report. The site-specific data includes: meteorological data (e.g. wind speed, wind direction, atmospheric stability, etc.)- to calculate the transport and dispersion of airborne effluents, and dilution factors for liquid effluents. The generic factors include the average annual consumption rates (for inhalation of air and ingestion of fruits, vegetables, leafy vegetables, grains, milk, poultry, meat, fish, and shellfish) and occupancy factors (for air submersion and ground irradiation, shoreline activity, swimming, boating, etc.). All these inputs are used in the appropriate dose models to calculate the maximum individual dose from radioactive airborne and liquid effluents.

1.1.1 Airborne Effluents Maximum individual doses due to the release of noble gases, radioiodines, and particulates were calculated using the computer code DOSAIR (Reference 11). This is equivalent to the NRC code, GASPAR I1,which uses a semi-infinite cloud model to implement the NRC Regulatory Guide 1.109 (Reference 3) dose models.

The values of average relative effluent concentration (x/Q) and average relative deposition (D/Q) used in the DOSAIR code were generated using EDAN4, a meteorological computer code which implements the assumptions cited in NRC Regulatory Guide 1.111 (Reference 5), Section C. The annual summary of hourly meteorological data (in 15-minute increments), which includes wind speed, direction, atmospheric stability, and joint frequency distribution, is not provided in the report but can be retrieved from computer storage.

5

Millstone Stack (375 ft) releases are normally considered elevated with Pasquill stability classes determined based upon the temperature gradient between the 33 ft and 374 ft meteorological tower levels. The doses were conservatively calculated using mixed mode 142 ft meteorology since DOSAIR may underestimate the plume exposure (prior to plume touchdown) for elevated releases from the Millstone Stack. All three units previously had the ability to discharge effluents to the Millstone Stack. However, in March 2001, Unit I was separated from releasing to the stack and modifications were made to add two new Unit I release points, the Spent Fuel Pool Island (SFPI) Vent and the Balance of Plant (BOP) Vent.

Unit 1 Spent Fuel Pool Island Vent (73 ft) and the Balance of Plant Vent (80 ft) releases are considered ground level; therefore these doses were calculated using the 33 ft meteorology. Continuous ventilation of the Spent Fuel Pool Island and the evaporation from the spent fuel pool water (H-3) release through the Spent Fuel Pool Island Vent. Continuous ventilation from other Unit 1 buildings and airborne releases from the reactor building evaporator are discharged to the BOP Vent. Doses from these release points were summed to determine the total Unit 1 airborne effluent dose.

Unit 2 Auxiliary Building Ventilation, Steam Generator Blowdown Tank Vent, and Containment Purge, (through the Unit 2 Vent, -159 ft) releases are considered mixed mode (partially elevated and partially ground) releases. The first two of these are continuous releases while the Containment Purge is typically a batch release.

Containment Purges can also be released via the Millstone Stack. Because doses for releases from the Unit 2 Vent and from the Millstone Stack are calculated using the same meteorology, the Containment Purge releases are not divided between Unit 2 Vent and Millstone Stack. Batch releases from the Waste Gas Decay Tanks and Containment Vents are typically discharged via the Millstone Stack. The doses for these elevated releases were conservatively calculated using mixed mode 142 ft meteorology for which the Pasquill stability classes are determined based upon the temperature gradient between the 33 ft and 142 ft meteorological tower levels. The Containment Equipment Hatch and the RWST Tank Vent releases are considered ground level where the 33 ft meteorology was used for the dose calculations. Each of the doses for the various release points were summed to determine the total Unit 2 airborne effluent dose.

The Unit 3 Vent (142.5 ft) is considered a mixed mode (partially elevated and partially ground) release point.

The Pasquill stability classes are determined based upon the temperature gradient between the 33 ft and 142 ft meteorological tower levels. Auxiliary Building Ventilation is a mixed mode continuous release while Containment Purge and the "initial" Containment Drawdown (released at the roof of the Auxiliary Building) are considered mixed mode batch releases. Gaseous waste and operational containment drawdowns (also called containment vents) are released through the Unit 3 Supplementary Leak Collection and Recovery System (SLCRS) system to the Millstone Stack (375 ft). The doses for these elevated releases were conservatively calculated using mixed mode 142 ft meteorology. The Engineered Safety Features Building (ESF) Ventilation, the Containment Equipment Hatch, and Refueling Water Storage Tank (RWST) Vent releases are considered ground level where the doses were calculated using 33 ft meteorology. Similar to Unit 2, each of the doses for the various release points were summed to determine the total Unit 3 airborne effluent dose.

1.1.2 Liquid Effluents Maximum individual doses from the release of radioactive liquid effluents were calculated using the DOSLIQ program (Reference 10). This program uses the dose models and parameters cited in NRC Regulatory Guide 1.109 with site-specific inputs to produce results similar to the LADTAP II code, (Reference 6).

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1.2 Dose Results The calculated maximum off-site doses are presented in Table 1-1 for airborne effluents and Table 1-2 for liquid effluents.

1.2.1 Airborne Effluents For the dose to the maximum individual, DOSAIR calculates the dose to the whole body, GI-tract (GI-LLI),

bone, liver, kidney, thyroid, lung, and skin from each of the following pathways: direct exposure from noble gases in the plume and from ground deposition, inhalation, and ingestion of vegetation, cow or goat milk, and meat. The values presented are a total from all pathways. However, only the whole body, skin, thyroid and maximum organ (other than thyroid) doses are presented.

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

For ground deposition, the maximum individual dose is calculated at both the off-site maximum land location of the highest X/Q and highest D/Q where a potential for dose exists.

For the vegetation pathway, the maximum individual dose is calculated at the vegetable garden of the highest D/Q (or highest X/Q when only tritium is released). For the vegetation pathway, the calculated dose is included in the maximum individual's dose only at locations and times where these pathways actually exist. Similarly, for meat, cow's milk, and goat's milk pathways, the calculated dose is included in the maximum individual's dose only at locations and times where these pathways actually exist.

To determine compliance with IOCFR50, Appendix I (Reference 7), the maximum individual whole body and organ doses include all applicable external pathways (i.e. plume and ground exposure) as well as the internal pathways (inhalation and ingestion).

1.2.2 Liquid Effluents The DOSLIQ code performs calculations for the following pathways: fish, shellfish, shoreline activity, swimming, and boating. Doses are calculated for the whole body, skin, thyroid, and maximum organ (GI-LLI, bone, liver, kidney, and lung).

1.2.3 Analysis of Results Table 1-3 provides a quantitative dose comparison with the limits specified in the REMODCM. The data indicates that the total whole body and organ doses to the maximum offsite individual from Millstone Station including all sources of the fuel cycle are well within the limits of 40CFRI90 (Reference 8). On-site radioactive waste storage during this year was within storage criteria and the maximum dose to a member of the public was approximately 0.18 mrem/yr. The doses from airborne and liquid effluents were added to the estimated dose from on-site radioactive waste storage to show compliance compared to 40CFR 190.

The Offsite Dose Comparison, Table 1-4, provides a perspective on the maximum offsite individual dose received from Millstone Station with the natural background radiation dose received by the average Connecticut resident. The total dose to the maximum individual received from Millstone Station is small (< 0.1%) in comparison to the dose received from natural background radiation.

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Table 1-1 2009 Off-Site Dose Commitments from Airborne Effluents Millstone Units 1, 2, 3 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Air (mrad) (mrad) (mrad) (mrad) (mrad)

Beta O.OOE+00 O.OOE+00 0.00E+00 O.OOE+00 O.OOE+00 Gamma O.OOE+00 O.OOE+00 O.OOE+0o O.OOE+00 O.OOE+00 Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 3.14E-05 1.16E-04 5.18E-05 1.16E-05 2.11 E-04 Skin 3.14E-05 1.18E-04 5.18E-05 1.16E-05 2.13E-04 Thyroid 3.14E-05 1.09E-04 5.18E-05 1.16E-05 2.04E-04 Max organ+ 3.14E-05 1.55E-04 5.18E-05 1.16E-05 2.50E-04 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total MaxAir (mrad) (mrad) (mrad) (mrad) (mrad)

Beta 1.90E-05 7.33E-04 6.89E-03 4.18E-03 1.18E-02 Gamma 3.70E-05 1.78E-03 4.26E-04 3.01 E-04 2.55E-03 Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 3.28E-04 4.51 E-03 1.25E-03 1.13E-03 7.21 E-03 Skin 3.46E-04 4.79E-03 5.33E-03 3.76E-03 1.42E-02 Thyroid 3.90E-04 5.44E-03 1.94E-03 1.17E-03 8.93E-03 Max organ+ 3.28E-04 4.51 E-03 1.29E-03 1.17E-03 7.30E-03 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Air (mrad) (mrad) (mrad) (mrad) (mrad)

Beta 8.97E-05 2.52E-04 1.84E-04 7.08E-04 1.23E-03 Gamma 8.31E-06 3.21 E-05 1.46E-05 1.21E-04 1.76E-04 Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 1.17E-03 2.49E-03 1.90E-03 6.21 E-04 6.19E-03 Skin 1.24E-03 2.55E-03 1.96E-03 1.OOE-03 6.75E-03 Thyroid 1,17E-03 2.50E-03 1-.90E-03 6.34E-04 6.21 E-03 Max organ+ 1,18E-03 2.50E-03 1.90E-03 6.28E-04 6.20E-03

+ Maxdmum of the following organs (not including thyroid): Bone, GI-LLI, Kidney, Liver, Lung 8

Table 1-2 2009 Off-Site Dose Commitments from Liquid Effluents Millstone Units 1, 2, 3 1tst Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 0.OOE+00 0.OOE+00 1.41E-07 0.OOE+00 1.41E-07 Thyroid 0.OOE+00 0.00E+00 4.04E-08 0.OOE+00 4.04E-08 Max Organ O.OOE+00 O.OOE+00 2.OOE-07 O.OOE+00 2.OOE-07 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total IMax Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 3.18E-05 3.43E-05 6.67E-04 9.56E-05 8.29E-04 Thyroid 2.66E-05 2.81E-05 1.94E-04 3.37E-05 2.82E-04 Max Organ 8.43E-05 9.96E-05 5.02E-03 2.56E-03 7.76E-03 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total iMax Individual ] (mrem) (mrem) (mrem) (mrem) (iner)

Whole Body 4.33E-05 5.83E-05 1.65E-04 1.33E-04 4.00E-04 Thyroid 3.13E-05 3.89E-05 1.10E-04 1.14E-04 2.95E-04 Max Organ 7.47E-04 6.16E-04 6.04E-04 2.94E-04 2.26E-03 9

Table 1-3 2009 Off-Site Dose Comparison to Limits Millstone Units 1, 2, 3 Airborne Effluents Dose Max Individual Dose vs REMODCM & 10CFR50 Appendix I Limits Whole Body Thyroid Max Organ* Skin Beta Air Gamma Air (mrem) (mrem) (Mrero) (mrem) (mrad) (mrad)

Unit 1 2.11E-04 2.04E-04 2.50E-04 2.13E-04 O.OOE+00 O.OOE+00 Unit 2 7.21E-03 8.93E-03 7.30E-03 1.42E-02 1.18E-02 2.55E-03 Unit 3 6.19E-03 6.21E-03 6.20E-03 6.75E-03 1.23E-03 1.76E-04 Millstone Station 1.36E-02 1.53E-02 1.38E-02 2.12E-02 1.31E-02 S.

2.72E-03 Dw~D*

Liquid Effluents Dose Max Individual Dose vs REMODCM & IOCFR50 Appendix I Limits Whole Body Thyroid Max Organ**

(mrem) (toem) (mrern)

Unit I 1.41E-07 4.04E-08 2.OOE-07 Unit 2 8.29E-04 2.82E-04 1.82E-02 Unit 3 6.65E-04 5.33E-04 7.76E-03 Millstone Station 1.49E-03 8.16E-04 2.60E-02 Total Off-Site Dose from Millstone Station Max Individual Dose vs REMODCM & 40CFR190 Limits Whole Body Thyroid Max Organ *

(mrem) (mrem) (mrem)

Airborne Effluents 1.36E-02 1.53E-02 1.38E-02 Liquid Effluents 1.49E-03 8.16E-04 2.60E-02 Radwaste Storage 1.80E-01 1.80E-01 1.80E-01 Millstone Station 1.95E-01 1.96E-01 2.20E-01 Note: RBEMODCM linits are listed in 1 OCFR50, Appendix I w hich contains additional linits not listed in the RBVEDCM

• Mxinum of the following organs (not including Thyroid): Bone, GI-LLI, Kidney, Liver, Lung 10

Table 1-4 2009 Offsite Dose Comparison Natural Background vs Millstone Station Average Resident Natural Background Radiation Dose Cosmic 27 mrem Cosmogenic 1 mrem Terrestial (Atlantic and Gulf Coastal Plain) 16 mrem Inhaled 200 mrem Inthe Body 40 mrem

- 284 mrem Courtesy NCRP Report 94 (1987)

Maximum Off-Site Individual Millstone Station Whole Body Dose Airborne Effluents 0.0136 mrem Liquid Effluents 0.0015 mrem On-site RadWaste Storage 0.1800 mrem 0.1951 mremi 1I

2.0 Effluent Radioactivity 2.1 Airborne Effluents 2.1.1 Measurement of Airborne Radioactivity 2.1.1.1 Continuous Releases The following pathways have continuous radiation monitors that include particulate filters and, except for Unit 1, charcoal cartridges for monitoring the activity being released:

Unit I Spent Fuel Pool (SFPI) Island (no charcoal cartridge)

Unit I Balance of Plant (BOP) Vent (no charcoal cartridge)

Unit 2 Ventilation Vent Unit 2 Wide Range Gas Monitor (WRGM) to Site Stack Unit 3 Ventilation Vent Unit 3 Supplementary Leak Collection and Recovery System (SLCRS) to Site Stack Unit 3 Emergency Safeguards Facility (ESF) Building Vent Charcoal cartridges and particulate filters are used to collect iodines and particulates, respectively. These filters are periodically replaced (typically weekly, except every two weeks for Unit 1) and then analyzed for isotopic content using a gamma spectrometer. Particulate filters are also analyzed for Sr-89 (for all but Unit 1), Sr-90 and gross alpha. At least monthly, gaseous grab samples are taken and analyzed for noble gasses and tritium.

The gas washing bottle (bubbler) method is utilized for tritium collection. This sample is counted on a liquid scintillation detector. Isotopic concentrations at the release point are multiplied by the total flow to obtain the total activity released for each isotope.

Since a major source of tritium is evaporation of water from the spent fuel pools, tritium releases were also estimated based upon amount of water lost and measured concentrations of the pool water. Grab samples from the Unit I SFPI Vent and the Unit 2 and 3 Vents are compared to the measured evaporation technique and the higher amount from either the vent or the measured evaporation technique is used to determine the amount of tritium released.

Another continuous airborne pathway is the Unit 2 Steam Generator Blowdown Tank (SGBD) Vent. A decontamination factor (DF) across the SGBD Tank vent was determined for iodines by comparing the results of gamma spectrometry, HPGe, analysis of the Steam Generator Blowdown water and grab samples of the condensed steam exiting the vent. This DF was applied to the total iodine releases via the Steam Generator Blowdown water to calculate the iodine release out the vent. An additional factor of 0.33 was utilized to account for the fraction of blowdown water actually flashing to steam in the Steam Generator Blowdown Tank.

12

2.1.1.2 Batch Releases The following pathways periodically have releases that are considered batches:

Unit I Reactor Building Evaporator (via BOP Vent)

Unit 2 Waste Gas Decay Tanks (via Unit 2 WRGM to Millstone Stack)

Unit 2 and 3 Containment Purges (via Unit Ventilation Vents, except for Unit 2 if using Enclosure Building Filtration System (EBFS) via WRGM to Millstone Stack)

Unit 2 and 3 Containment Vents (via EBFS to Millstone Stack for Unit 2 and via SLCRS to Millstone Stack for Unit 3)

Unit 2 and 3 Containment Equipment Hatch Openings Unit 2 and 3 Refueling Water Storage Tank (RWST) Vents Unit 3 Containment Drawdown Prior to processing each batch from the Unit I Reactor Building Evaporator a sample is collected and counted on a liquid scintillation detector. Concentration is multiplied by volume to determine the total activity released.

Waste Gases from the Unit 2 Gaseous Waste Processing System are held for decay in waste gas decay tanks (6) prior to discharge through the Millstone Site Stack. Each gas decay tank is analyzed prior to discharge for noble gas and tritium. Calculated volume discharged is multiplied by the isotopic concentrations (noble gas and tritium) from the analysis of grab samples to determine the total activity released.

Containment air is sampled periodically for gamma and tritium to determine the activity released from containment venting. The measured concentrations are multiplied by the containment vent volume to obtain the total activity released. Unit 2 typically performs this process of discharging air from containment to maintain pressure approximately once per week while at Unit 3 it is more often (typically every day or two). Any iodines and particulates discharged would be detected by the continuous monitoring discussed in section 2.1.1.1.

Containment air is sampled prior to each purge for gamma and tritium to determine the activity released from containment purging. Similar to containment venting, the measured concentrations are multiplied by the containment volume to obtain the total activity released. Any iodines and particulates discharged would be detected by the continuous monitoring discussed in section 2. 1. 1.1.

Samples of air near the Containment Equipment Hatch openings are analyzed for particulates and iodines, during refueling outages for the period that the equipment hatch is open. An estimated flow out of the hatch and sample results are used to determine the radioactivity released.

When water is transferred to Refueling Water Storage Tank (RWST) there is a potential for a release of radioactivity through the tank vent. A decontamination factor (DF) was applied to the total iodine contained in the water transferred to the RWST to estimate the iodine released. All noble gases are assumed to be released through the tank vent.

Unit 3 containment is initially drawn down prior to startup. This is accomplished by using the containment vacuum steam jet ejector which releases through an unmonitored vent on the roof of the Auxiliary Building.

Grab samples are performed prior to drawdown to document the amount of radioactivity released during this evolution.

13

2.1.2 Estimate of Errors Estimates of errors associated with radioactivity measurements were made using the following guidelines:

Radioactivity Measurement Calibration 10% Calibration to NBS standards Sampling/Data Collection 10% - 20% Variation in sample collection Sample Line Loss 20% - 40% Deposition of some nuclides Sample Counting 10% - 40% Error for counting statistics Flow & Level Measurements 10% - 20% Error for release volumes 2.1.3 Airborne Batch Release Statistics Unit 1 - None Unit 2 Ctmt Purges Ctmt Vents WGDT Number of Batches 4 40 6 Total Time (min) 3586 6232 3416 Maximum Time (min) 2514 243 820 Average Time (min) 896 156 569 Minimum Time (min) 74 55 215 Unit 3 Ctmt Purges Ctmt Vents* Ctmt Drawdowns Number of Batches 0 248 0 Total Time (min) 0 0 Maximum Time (min) 0 0 Average Time (min) 0 0 Minimum Time (min) 0

• 0

• 2-3 hrs per Vent 2.1.4 Abnormal Airborne Releases An abnormal airborne release of radioactivity is defined as an increase in airborne radioactive material released to the environment that was unplanned or uncontrolled due to an unanticipated event. These do not include normal routine effluent releases from anticipated operational and maintenance occurrences such as power level changes, reactor trip, opening primary system loops, degassing, letdown of reactor coolant or transferring spent resin and do not include non-routine events such as minor leakages from piping, valves, pump seals, tank vents, etc.

2.1.4.1 Unit 1 - None 2.1.4.2 Unit 2-None 2.1.4.3 Unit 3-None 14

2.1.5 Airborne Release Tables The following tables provide the details of the airborne radioactivity released from each of the Millstone units obtained from Reference 2. They are categorized by type of release, source(s), and by release point of discharge to the environment.

15

Table 2.1-Al Millstone Unit I Airborne Effluents - Release Summary 1 2009 Units I Ist Otr 2nd Otr 3rd Otr 4th Otr Total I A. Fission & Activation Gases D. Gross Alpha I. Total Activity Ci ReleasedII I IIIII E. Tritium "na" denotes Not Required to be Analyzed

+ "Total Activity Released" + Seconds in Quarter (or year)

-" measurements below detectable levels 16

Table 2.1-A2 Millstone Unit I Airborne Effluents - Ground Continuous - BOP Vent+ & SFPI Vent Nuclides II, Released Units I st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases Kr-85 Ci Total Activity Ci B. lodines / Halogens Ci na na na na na Total Activity Ci na na na na na C. Particulates Cs-137 Ci - 9.43E-07 - - -

Sr-90 Ci -..

Mn-54 Ci -..

Co-60 Ci -...

Zn-65 Ci -...

Cs-134 Ci - - -

Ce- 144 Ci ---

Total Activity Ci 9.43E-07 -

D. Gross Alpha Gross Alpha Ci - - [ - -

E. Tritium H-3 Ci 1.13E-01 6.70E-02 4.60E-02 6.43E-02 2.91E-01 I

+ includes releases from Unit I evaporator "na" denotes Not Required to be Analyzed

"-" measurements below detectable levels 17

Table 2.2-Al Millstone Unit 2 Airborne Effluents - Release Summary IUnits I st Qtr 2nd Qtr I 3rd Qtr 4th Qtr Total ,

A. Fission & Activation Gases D. Gross Alpha

1. Total Activity Ci Released

+ "Total Activity Released" - Seconds in Quarter

- measurements below detectable levels 18

Table 2.2-A2 Millstone Unit 2 Airborne Effluents - Mixed Continuous - Aux Bldg Vent & SGBD Tank Vent

& Snent Fuel Pool Evaporation Nuclides Released Units I lstQtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Gases Ar-41 Ci 5.951E-01 5.95E-01 Kr-85 Ci - 5.70E-01 9.11E+00 9.68E+00 Kr-87 Ci Kr-85m Ci Kr-88 Ci Xe-133 Ci - 1.40E-01 6.17E+00 6.3 1E+00 Xe-133m Ci -

Xe-135 Ci - 1.52E-01 1.52E-01 Xe-138 Ci -

Total Activity Ci - 7.47E-01 7.10E-01 1.53E+01 1.67E+O1 B. Iodines / Halo ens 1-131 Ci 1.53E-05 2.02E-05 1.01E-05 1.84E-05 6.39E-05 1-132 Ci - 2.42E-05 2.42E-05 1-133 Ci 6.33E-05 7.68E-05 4.13E-05 2.36E-05 2.05E-04 1-135 Ci 1.29E-05 1.29E-05 Br-82 Ci 1.80E-06 1.80E-06 Total Activity Ci 7.85E-05 1.1OE-04 5.14E-05 6.79E-05 3.08E-04 C. Particulates Mn-54 Ci - -

Co-58 Ci -

Fe-59 Ci -

Co-60 Ci - -

Zn-65 Ci -

Mo-99 Ci - -

Cs-134 Ci Cs-137 Ci - -

Ce-141 Ci Ce-144 Ci Sr-89 Ci - -

Sr-90 Ci -

Total Activity Ci - -

D. Gross Alpha Gross Alpha Ci - - -

E. Tritium

• H-3 Ci 2.38E+00 i7E+O1

1. 1.72E+00 1.23E+01 2.81E+0l measurements below detectable levels 19

Table 2.2-A3 Millstone Unit 2 Airborne Effluents - Mixed Batch - Containment Purges IReleased Unt Nuclides Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total I A. Fission & Activation Gases Kr-85 Ci *

• 2.04E+00 1.00E+00 3.04E+00 Kr-87 Ci *

• Kr-88 Ci *

• Xe-133 Ci *

• 7.84E-01 1.OOE-01 8.84E-01 Xe-133m Ci *

• Xe-135 Ci *

• Xe-138 Ci *

• Total Activity Ci 2.82E+00 1.1OE+00 3.92E+00 B. lodines / Halogens TotalActivity Ci + + + + +

C. Particulates TotalActivity Ci + + + + +

D. Gross Alpha Gross Alpha Ci + + + + +

E. Tritium H-3 Ci *

• 4.63E-01 3.03E-01 7.66E-01

+ lodines, Particulates and Gross Alpha appear in Tables 2.2-A2 or 2.2-A5 measurements below detectable levels "na" denotes Not Required to be Analyzed

• No activity released 20

Table 2.2-A4 Millstone Unit 2 Airborne Effluents - Elevated Batch - WGDT N uc l id es U is ,h trT a Released lst Qtr 1Units I 2ndQtr I 3rd Qtr W 4thQtr Total A. Fission & Activation Gases Kr-85 Ci *

• 2.89E+00 4.54E+00 7.44E+00 Kr-87 Ci *

• Kr-88 Ci *

• Xe-131m Ci *

• 6.49E-04 1.18E-03 1.83E-03 Xe-133 Ci *

• 1.65E-03 2.40E-03 4.05E-03 Xe-133m Ci *

• Xe-135 Ci *

• Xe-138 Ci *

• Total Activity Ci 2.90E+00 4.55E+00 7.44E+00 B. Iodines / Halogens Total Activity Ci na na na na na C. Particulates Total Activity Ci na na na na na D. Gross Alpha Gross Alpha Ci na na na na na E. Tritium H-3 Ci [

• 1.08E-03 8.30E-04 1.91E-03

• No activity released "na" denotes Not Required to be Analyzed

+ lodines, Particulates and Gross Alpha appear in Tables 2.2-A2 or 2.2-A5 measurements below detectable levels 21

Table 2.2-A5 Millstone Unit 2 Airborne Effluents - Elevated Continuous - Containment Vents/Site Stack Nuclides Released Units 1ist Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Gases Ar-41 Ci 2.51E-02 2.54E-02 2.37E-02 1.71E-02 9.13E-02 Kr-85 Ci L.OOE-02 3.01E-02 5.76E+00 6.52E-02 5.86E+00 Kr-87 Ci -

Kr-88 Ci Xe-133 Ci 1.79E-02 2.41E-02 2.24E+00 1.13E-01 2.39E+00 Xe-133m Ci Xe-135 Ci 1.79E-03 2.30E-03 5.63E-04 1.78E-04 4.83E-03 Xe-138 Ci -

Total Activity Ci 5.48E-02 8.18E-02 8.02E+00 1.95E-01 8.35E+00 B. lodines / Halogens 1-131 Ci 5.28E-07 5.28E-07 1-133 Ci 2.11E-07 2.1IE-07 Br-82 Ci 3.64E-07 3.64E-07 Total Activity Ci 1.IOE-06 [1.1OE-06 C. Particulates Be-7 Ci 1.40E-07 - 1.40E-07 Mn-54 Ci Co-58 Ci Fe-59 Ci Co-60 Ci Zn-65 Ci Mo-99 Ci Cs-134 Ci Cs-137 Ci Ce-141 Ci Ce-144 Ci Sr-89 Ci Sr-90 Ci Total Activity Ci 1.40E-07 1-.40E-07 D. Gross Alpha Gross Alpha Ci - - - - -

E. Tritium H-3 Ci 3.71E-01 I 5.27E-01 1.18E+00 3.88E-01 2.47E+00 measurements below detectable levels 22

Table 2.2-A6 Millstone Unit 2 Airborne Effluents - Ground Batch - Containment Equipment Hatch Nuclides I Released Units 1lst Qtr 2nd Qtr [ 3rdQtr 4th2Total Qtr A. Fission & Activation Gases y,Emitters Ci Total Activity Ci B. lodines / Halo ens 1-131 Ci * *

• 1.90E-07 1.90E-07 1-133 Ci * * *-

Total Activity Ci 1.90E-07 1.90E-07 C. Particulates Co-58 Ci * *

• 2.70E-07 2.70E-07 Cs-137 Ci * *

• 9.50E-08 9.50E-08 Total Activity Ci 3.65E-07 3.65E-07 D. Gross Alpha Gross Al ha Ci * *

• E. Tritium H-3 Ci na na na na I na

- measurements below detectable levels "na" denotes Not Required to be Analyzed

• No activity released 23

Table 2.2-A7 Millstone Unit 2 Airborne Effluents - Ground Batch - RWST Vent Nuclides  !

Released Units I 1st Qtr 2nd Qtr I 3rd Otr 4th Qtr Total A. Fission & Activation Gases Xe-133 Ci 2.10E-01 2.10E-01 Total Activity Ci * *

• 2.10E-01 2.10E-01 B. lodines / Halogens 1-131 Ci * *

• 1-133 Ci * *

• Total Activity Ci * *

• C. Particulates IYEmitters I Cii Total Activity Ci * *

• D. Gross Alpha Gross Alpha Ci * *

• na na ]

E. Tritium H-3 Ci * *

• na na measurements below detectable levels "na" denotes Not Required to be Analyzed

• No activity released 24

Table 2.3-Al Millstone Unit 3 Airborne Effluents - Release Summary Units 1st QtrIr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gas es D. Gross Alpha

1. Total Activity Ci Released

+ "Total Activity Released" - Seconds in Quarter

.-" measurements below detectable levels 25

Table 2.3-A2 Millstone Unit 3 Airborne Effluents - Mixed Continuous - Normal Ventilation &

Spent Fuel Pool Evaporation Nuclides 1ndQ Released Units 1st Qtr 2nd Qtr I3rd Qtr 4th Qtr Total I A. Fission & Activation Gases Kr-87 Ci Kr-88 Ci Xe-133 Ci - 3.76E-01 3.76E-01 Xe-1 33m Ci - - -

Xe-135 Ci - - -

Xe-138 Ci - - -

Total Activity Ci 3.76E-01 3.76E-01 B. lodines / Halogens 1-131 Ci 9.64E-06 9.64E-06 1-133 Ci -

Total Activity i 9.64E-06 9.64E-06 C. Particulates Be-7 Ci - -

Mn-54 Ci ....

Co-58 Ci .....

Fe-59 Ci .....

Co-60 Ci .....

Zn-65 Ci ....

Mo-99 Ci ....

Cs-134 Ci ....

Cs-137 Ci ....

Ce-141 Ci ....

Ce-144 Ci ....

Sr-89 Ci ....

Sr-90 Ci ....

Total Activity Ci D. Gross Alpha Gross Alpha Ci I - - -

E. Tritium H-3 Ci 1.58E+01 1.07E+01 8.68E+00 9.81E+00 4.50E+01 measurements below detectable levels 26

Table 2.3-A3 Millstone Unit 3 Airborne Effluents - Ground Continuous - ESF Building Ventilation Nuclides ,s Released Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases y Emitters Ci Total Activity Ci -

B. lodines I Halo ens 1-131 Ci 6.54E-07 6.54E-07 1-133 Ci - -

Total Activity Ci 6.54E-07 6.54E-07 C. Particulates Be-7 Ci 5.85E-07 - 5.85E-07 Cr-51 Ci 1.39E-07 1.39E-07 Mn-54 Ci -

Co-58 Ci -

Fe-59 Ci -

Co-60 Ci - -

Zn-65 Ci -

Mo-99 Ci Cs-1 34 Ci - -

Cs-137 Ci ....

Ce-141 Ci -

Ce-144 Ci -

Sr-89 Ci Sr-90 Ci -

Total Activity Ci 5.85E-07 1.39E-07 7.24E-07 D. Gross Alpha Gross Alpha Ci - - I - -

E. Tritium H-3 Ci 1.62E-01 I 2.73E-02 1.65E-01 3.54E-01

- measurements below detectable levels 27

Table 2.3-A4 Millstone Unit 3 Airborne Effluents - Mixed Batch - Containment Drawdowns Nuclides Released Units 1st Qtr I 2nd Qtr ! 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases y Emitters Ci .....

Total Activity Ci .....

B. lodines I Halogens 1-131 Ci .....

1-133 Ci .....

Total Activity Ci .

C. Particulates y Emitters Ci * * * *

• Total Activity Ci .....

D. Gross Alpha

[Gross Alp ha Ci na na na na E. Tritium H-3 Ci * * * * *

"-" measurements below detectable levels "na" denotes Not Analyzed

• No activity released 28

Table 2.3-A5 Millstone Unit 3 Airborne Effluents - Mixed Batch - Containment Purges Nuclides 1 2009 1 Released Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases y Emitters Ci Total Activity Ci * * * *

• B. lodines / Halogens Total Activity I Ci * *

• I C. Particulates Total Activity ci I * *

• D. Gross Alpha Gross Alpha Ci na na na na na E. Tritium H-3 Ci * * * * *

"na" denotes Not Required to be Analyzed

• No activity released

. 29

Table 2.3-A6 Millstone Unit 3 Airborne Effluents - Elevated Continuous - Gaseous Waste System Nuclides I Released Unitsl 1st Qtr I 2ndQtr I 3rdQtr I 4thQtr I Total I A. Fission & Activation Gases Kr-85 Ci 3.50E-01 3.73E-01 6.21E-01 2.73E+00 4.07E+00 Kr-87 Ci - - -

Kr-88 Ci - - -

Xe-131m Ci 9.58E-02 4.21E-01 2.60E-03 4.88E-02 5.68E-01 Xe-1 33 Ci - 3.90E+00 3.90E+00 Xe-1 33m Ci - - 7.58E-02 7.58E-02 Xe-135 Ci - - 1.66E-02 1.66E-02 Xe-138 Ci - -

Total Activity Ci 4.46E-01 7.94E-01 6.24E-01 6.77E+00 8.63E+00 B. lodines / Halo _ _ns 1-131 Ci 5.63E-08 1.1OE-07

. - 6.86E-07 8.52E-07 1-133 Ci - 4.76E-07 - - 4.76E-07 Br-82 Ci 5.15E-06 4.85E-06 5.44E-06 4.58E-06 2.OOE-05 Total Activity Ci 5.21 E-06 5.44E-06 5.44E-06 5.27E-06 2.13E-05 C. Particulates Mn-54 Ci - - -

Co-57 Ci - 1.25E-08 - 1.25E-08 Fe-59 Ci - - -

Co-58 Ci - -

Co-60 Ci 1.1OE-07 7.57E-08 - 1.86E-07 Zn-65 Ci - - -

Nb-95 Ci - - - 1.81E-08 1.81E-08 Mo-99 Ci - - -

Cs-134 Ci - - -

Cs-137 Ci - - -

Ce-141 Ci - - -

Ce-144 Ci - - -

Sr-89 Ci - -- -

Sr-90 Ci - - -

Total Activity Ci 1.10E-07 8.82E-08 1.81E-08 2.16E-07 D. Gross Alpha JGrossAlpha CI I - - - I -

E. Tritium 1H-3 I Ci 2.38E-01 12.71E-O1 12.51E-01 4.95E-01 I 1.26E+00 measurements below detectable levels 30

Table 2.3-A7 Millstone Unit 3 Airborne Effluents - Elevated Batch - Containment Vents Nuclides Released Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total A. Fission & Activation Gases Ar-41 Ci 6.55E-03 7.58E-03 8.30E-03 7.58E-03 3.OOE-02 Kr-85 Ci 5.83E-02 1.36E-02 - 3.81 E-03 7.57E-02 Kr-87 Ci - -

Kr-88 Ci - -

Xe-131 m Ci 7.11E-04 - 7.11E-04 Xe-133 Ci 9.50E-03 6.67E-02 3.16E-02 1.30E-01 2.38E-01 Xe-133m Ci - 1.48E-03 1.48E-03 Xe-135 Ci 1.49E-04 4.37E-04 7.62E-04 1.35E-03 2.70E-03 Xe-135m Ci 7.41E-05 - - 7.41E-05 Xe-138 Ci - -

Total Activity Ci 7.45E-02 8.84E-02 4.14E-02 1.44E-01 3.48E-01 B. lodines / Halogens TotalActivity I Ci + + + + +

C. Particulates TotalActivity Ci + + + +

D. Gross Alpha GrossAlpha Ci + + + +

E. Tritium H-3 Ci 1.34E-02 2.13E-02 2.38E-02 2.83E-02 8.68E-02

+ lodines, Particulates and Gross Aipha included in Table 2.3-A6 measurements below detectable levels 31

Table 2.3-A8 Millstone Unit 3 Airborne Effluents - Ground Batch - Containment Equipment Hatch Nuclides SI.

Released Units I1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases y Emitters Ci .....

Total Activity Ci .....

B. lodines / Halogens 1-131 Ci .....

1-133 Ci .....

Total Activity Ci .....

C. Particulates y Emitters Ci ....

Total Activity Ci .....

D. Gross Alpha Gross Alpha Ci na na na na na E. Tritium H-3 Ci na na na na na

• No activity released "na" denotes Not Required to be Analyzed 32

Table 2.3-A9 Millstone Unit 3 Airborne Effluents - Ground Batch - RWST Vent Nuclides Released Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases yEmitters Ci ......

Total Activity Ci ...

B. lodines I Halogens 1-131 Ci .....

1-133 Ci . I...

Total Activity Ci .....

C. Particulates y Emitters Ci .....

Total Activity Ci .....

D. Gross Alpha Gross Alpha Ci na na na na na E. Tritium H-3 ci na na na na F na I

• No activity released "na" denotes Not Required to be Analyzed 33

2.2 Liquid Effluents 2.2.1 Measurement of Liquid Radioactivity 2.2.1.1 Continuous Liquid Releases Grab samples are taken for continuous liquid release pathways and analyzed on the HPGe gamma spectrometer and liquid scintillation detector (for tritium) if required by the conditional action requirements of the REMODCM. Total estimated volume is multiplied by the isotopic concentrations (if any) to determine the total activity released. A proportional aliquot of each discharge is retained for composite analysis for Sr-89, Sr-90, Fe-55 arid gross alpha if required by the conditional action requirements of the REMODCM. Pathways for continuous liquid effluent releases include, Steam Generator Blowdown, Service Water Effluent, and Turbine Building Sump discharge from Units 2 & 3.

2.2.1.2 Liquid Tanks/Sumps There are numerous sources from which liquids containing radioactivity are discharged to the environs.

These are the primary liquid sources:

Unit 1 Underground ventilation duct Unit 2 Clean Waste Monitor Tanks (2)

Aerated Waste Monitor Tanks (2)

CPF Waste Neutralization Sump & Turbine Building Sump Steam Generator Bulk CST Pipe Trench Site Stack Sump (From Units 1, 2 and 3 but accounted for by Unit 2)

Unit 3 High Level Waste Test Tanks (2)

Low Level Waste Drain Tanks (2)

Boron Test Tanks CPF Waste Neutralization Sump & Turbine Building Sump Steam Generator Bulk Foundation Drain Sumps WTT Berm Water Prior to release, a tank is re-circulated for at least two equivalent tank volumes, a sample is drawn and then analyzed on the HPGe gamma spectrometer and liquid scintillation detector (H-3) for individual radionuclide composition. Isotopic concentrations are multiplied by the volume released to obtain the total activity released. For bulk releases, several samples are taken during the discharge to verify the amount of radioactivity released. A proportional aliquot of each discharge is retained for composite analysis for Sr-89, Sr-90, Fe-55, and gross alpha.

34

2.2.2 Estimate of Errors Estimates of errors associated with radioactivity measurements were made using the following guidelines:

Radioactivity Measurement Calibration 10% Calibration to NBS standards Sampling/Data Collection 10% - 20% Variation in sample collection Sample Line Loss 20% - 40% Deposition of some nuclides Sample Counting 10%-30% Error for counting statistics Flow & Level Measurements 10% - 20% Error for release volumes 2.2.3 Liquid Batch Release Statistics Unit I Unit 2 Unit 3 Number of Batches 3 72 41 Total Time (min) 345 9915 6102 Maximum Time (min) 345 1008 240 Average Time (min) 345 138 149 Minimum Time (min) 345 1 60 Average Stream Flow Not Applicable - Ocean Site 2.2.4 Abnormal Liquid Releases An abnormal release of radioactivity is the discharge of a volume of liquid radioactive material to the environment that was unplanned or uncontrolled.

In 2009, the following abnormal liquid releases occurred:

2.2.4.1 Unit 1 - None 2.2.4.2 Unit 2 - None 2.2.4.3 Unit 3 - None 2.2.5 Liquid Release Tables The following tables provide the details of the liquid radioactivity released from each of the Millstone units obtained from Reference 2. They are categorized by type of release, source(s), and by release point of discharge to the environment (i.e. Quarry, Yard Drain DSNO06).

35

Table 2.1-Li Millstone Unit I Liquid Effluents - Release Summary (Release Point - Quarry)

Units 1stQtr I 2ndQtr I 3rd Qtr I 4th Qtr Total A. Fission and Activation Products D. Gross Alpha

1. Total Activity Ci * *

• Released E. Volume

1. Released Waste Liters

• 9.71E+04 9.71E+04 Volume

2. Dilution Volume Liters

• 7.08E+08 7.08E+08 During Releases+'+

3. Dilution Volume Liters

• 2.72E+1 I 2.72E+1 I During Period- I I

- measurements below detectable levels

• No activity released

+ "Total Activity Released" - ("Released Waste Volume" + "Dilution Volume During Period")

++ Unit 2 E.3 quarterly dilution used because there is no more Unit I dilution

+++ E.3 quarterly dilution x (Total release time + Total quarter time) 36

Table 2.1-L2 Millstone Unit I Liquid Effluents - Batch (Release Point - Quarry)

Nuclides II' Released Units I 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Products Cs-137 Ci *

• 1.22E-05

• 1.22E-05 Sr-90 Ci *

• Fe-55 Ci *

• Total Activity Ci

• 1.22E-05 1.22E-05 B. Tritium HI-3 I ci

• C. Dissolved & Entrained Gases Kr-85 Ci *

• Ci Ci Total Activity Ci D. Gross Alpha Gross Alpha Ci I *

• measurements below detectable levels

• No activity released 37

Table 2.2-Li Millstone Unit No. 2 Liquid Effluents - Release Summary (Release Point - Quarry)

I Units 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total A. Fission and Activation Products D. Gross Alpha

1. Total Activity Ci - - I -

Released C E. Volume

1. Released Waste Volume Primary Liters 2.27E+05 2.54E+05 1.88E+06 5.83E+05 2.94E+06 Secondary Liters 2.33E+07 1.65E+07 ++ 5.60E+03 3.98E+07

2. Dilution Volume During Releases Primary Liters 1.49E+09 1.86E+09 7.93E+09 2.84E+09 1.41E+10 Secondary Liters

3. Dilution Volume During Period Liters 2.73E+1l 2.79E+11 2.72E+11 1.78E+11 L.OOE+12

- measurements below detectable levels

+ "Total Activity Released" -' (Primary "Released Waste Volume" + "Dilution Volume During Period")

++ No release of secondary waste 38

Table 2.2-L2 Millstone Unit 2 Liquid Effluents - Continuous - SGBD, SW, RBCCW (Release Point - Quarry)

Nuclides nt' Released I 1Units 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total A. Fission & Activation Products y Emitters Ci -

• Sr-89 Ci na na

• na na Sr-90 Ci na na

• na na Fe-55 Ci na na

• na na Total Activity Ci -

• B. Tritium H-3 I Ci 4.05E-02 1.51E-02

• 1.02E-04 5.57E-02 C. Dissolved & Entrained Gases y Emitters Ci Total Activity Ci D. Gross Alpha Gross Alpha Ci na na

• na na measurements below detectable levels

• No activity released "na" denotes not required to be analyzed 39

Table 2.2-L3 Millstone Unit 2 Liquid Effluents - Batch - LWS (Release Point - Quarry)

Nuclides II Released Units 1st Qtr I 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Products Be-7 Ci 9.75E-05 9.75E-05 Na-24 Ci 1.04E-06 1.04E-06 Cr-51 Ci 3.58E-03 5.68E-04 4.15E-03 Mn-54 Ci 2.23E-05 I. 17E-05 3.49E-04 8.52E-05 4.68E-04 Co-58 Ci 4.61E-05 3.30E-05 8.73E-03 2.96E-03 I. 18E-02 Fe-59 Ci 1.62E-03 1.59E-04 1.78E-03 Co-60 Ci 2.03E-04 2.43E-04 7.52E-03 1.86E-03 9.82E-03 Zn-65 Ci Nb-95 Ci 1.52E-05 1.97E-05 7.39E-04 6.28E-04 1.40E-03 Zr-95 Ci 2.69E-04 2.74E-04 5.43E-04 Mo-99 Ci Ru-103 Ci 1.57E-05 1.70E-06 1.74E-05 Ru-105 Ci 2.04E-04 7.91E-05 2.83E-04 Ag-ll0m Ci 5.29E-06 7.50E-06 3.51E-04 2.93E-04 6.57E-04 Sn-I 13 Ci 1.42E-05 2.87E-04 7.05E-06 3.09E-04 Sn-I 17m Ci 6.10E-06 - 2.58E-04 - 2.64E-04 Sb-122 Ci 1.48E-05 - 1.48E-05 Sb-124 Ci 2.78E-04 1.42E-04 4.20E-04 Sb-125 Ci 1. I1E-03 7.75E-04 1.03E-02 1.78E-03 1.40E-02 Sb-126 Ci 5.23E-06 - 5.23E-06 1-131 Ci 2.11E-05 - 2.1IE-05 1-132 Ci 4.33E-05 - 4.33E-05 Te-132 Ci 1.55E-04 - 1.55E-04 Cs-134 Ci 2.80E-06 1.36E-05 3.17E-06 5.55E-05 7.50E-05 Cs-137 Ci 7.83E-05 8.90E-05 1.04E-04 2.03E-04 4.74E-04 Ba-139 Ci 1.39E-04 1.39E-04 La-140 Ci 1.25E-03 1.25E-03 Ce-141 Ci Fe-55 Ci 2.98E-04 4.33E-04 2.81E-02 2.64E-03 3.15E-02 Ni-63 Ci 4.53E-04 2.58E-04 3.50E-03 4.86E-04 4.70E-03 Sr-89 Ci -

Sr-90 Ci Total Activity Ci 2.34E-03 1.90E-03 6.78E-02 1.22E-02 8.43E-02 B. Tritium H-3 ] Ci [ 5.76E+01 8.53E+01 3.93E+02 11.97E+01 5.56E+02 C. Dissolved & Entrained Gases Kr-85 Ci 5.51E-02 6.91E-02 8.93E-01 2.05E-01 1.22E+00 Xe-131m Ci 1.57E-02 1.57E-02 Xe-133 Ci 3.86E-04 8.13E-04 3.50E-01 3.78E-03 3.55E-01 Xe-133m Ci - 3.16E-03 3.16E-03 Xe-135 Ci - 5.1OE-04 1.27E-05 5.23E-04 Total Activity Ci 5.54E-02 6.99E-02 1.26E+00 2.09E-01 1.60E+00 D. Gross Alpha Gross Alpha Ci - I - -

measurements below detectable levels 40

Table 2.2-L4 Millstone Unit 2 Liquid Effluents - Release Summary (Release Point - Yard Drain - DSN 006)

Units I Ist Qtr I 2nd Qtr 3rd Qtr I 4th Qtr Total Gross Alpha Total Activity C Released C Volume Released Waste Liters 3.03E+06 6.22E+05 0 2.07E+06 5.72E+06 Volume

+4+ +--I +4+ +4 + Dilution Volume Liters During Releases Dilution Volume Liters 2.92E+07 3.64E+07 3.19E+07 3.05E+07 1.28E+08 During Period ++

measurements below detectable levels

• No activity released

+ "Total Activity Released" + ("Released Waste Volume" + "Dilution Volume During Period")

++ Includes all station dilution sources via Yard Drain - DSN 006

+++Continuous "Dilution Volume During Releases" is not quantified 41

Table 2.2-L5 Millstone Unit 2 Liquid Effluents -Continuous-Turbine Building Sump/CST Aux Steam Tunnel (Release Point - Yard Drain - DSN 006)

Nuclides II Released I 1Units 1st Qtr 2nd Qtr I 3rd Qtr I 4th Qtr Total A. Fission & Activation Products y Emitters Ci

• Sr-89 Ci na na

• na na Sr-90 Ci na na

• na na Fe-55 Ci na na

• na na Total Activity Ci *-

B. Tritium H-3 [ Ci 1 3.35E-02 3.36E-03

• 3.92E-02 7.61E-02 C. Dissolved & Entrained Gases y Emitters Ci --

Total Activity Ci D. Gross Alpha Gross Alpha Ci na na na na

"-" measurements below detectable levels

• No activity released "na" denotes Not Required to be Analyzed 42

Table 2.3-Li Millstone Unit 3 Liquid Effluents - Release Summary (Release Point - Quarry) 2009 I

IUnits I 1st Qtr 1 I 2nd Qtr I 3rd Qtr I 4thQtr Total D. Gross Alpha I1. Total Activity Ci I--

Released C E. Volume

1. Released Waste Volume Primary Liters 4.25E+05 5.08E+05 1.09E+06 7.OOE+05 2.72E+06 Secondary Liters 8.48E+06 9.32E+06 8.76E+06 9.93E+06 3.65E+07

2. Dilution Volume During Releases Primary Liters 1.52E+09 2.25E+09 5.27E+09 3.12E+09 1.22E+10 Secondary Liters 1.33E+10 1.66E+10 1.65E+10 1.64E+10 6.28E+10

3. Dilution Volume During Period Liters 4.30E+1I 4.65E+I1 4.72E+11 4.38E+11 1.8 IE+12

- measurements below detectable levels

+

"Total Activity Released" - (Primary "Released Waste Volume" + "Dilution Volume During Period")

43

Table 2.3-L2 Millstone Unit 3 Liquid Effluents - Continuous - SGBD & SW (Release Point - Quarry)

Nuclides U sl Released Units 1stQtr 2nd Qtr 3rd Qtr 4th Qtr I Total A. Fission & Activation Products y Emitters Ci - -

Sr-89 Ci na na na na na Sr-90 Ci na na na na na Fe-55 Ci na na na na na Total Activity Ci B. Tritium H-3 Ci 4.77E-02 1.19E-01 1.45E-01 1.39E-0 [ 4.51E-01 C. Dissolved & Entrained Gases y Emitters Ci -

Total Activity Ci D. Gross Alpha Gross Alpha Ci na na na na na measurements below detectable levels "na" denotes Not Required to be Analyzed 44

Table 2.3-L3 Millstone Unit 3 Liquid Effluents - Batch - LWS (Release Point - Quarry)

Nuclides Units i ll 1 Released Units sQtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Products Mn-54 Ci 3.38E-04 1.11E-04 4.59E-04 3.45E-05 9.43E-04 Co-57 Ci 9.68E-06 9.68E-06 Co-58 Ci 7.07E-04 1.67E-04 5.30E-04 3.54E-05 1.44E-03 Fe-59 Ci - -

Co-60 Ci 1.54E-03 1.21E-03 3.95E-03 1.03E-03 7.73E-03 Zn-65 Ci - - -

Nb-95 Ci 9.23E-05 2.33E-05 1.16E-04 Mo-99 Ci - -

Ag-I 10ml Ci i.63E-04 1.75E-04 7.44E-05 2.61E-05 4.39E-04 Sn-117m Ci 1.14E-05 3.63E-06 1.50E-05 Sb-125 Ci 1.46E-03 7.48E-03 6.22E-04 6.30E-03 1.59E-02 1-131 Ci -

Cs-134 Ci - 3.15E-04 - 6.96E-05 3.85E-04 Cs-137 Ci - 2.05E-04 5.61E-05 1.17E-04 3.78E-04 Ce-141 Ci -.....

Fe-55 Ci 3.70E-04 8.33E-04 4.28E-03 9.45E-04 6.43E-03 Ni-63 Ci - -

Sr-89 Ci - - -

Sr-90 Ci - -

Total Activity Ci 4.68E-03 1.05E-02 1.OOE-02 8.56E-03 3.37E-02 B. Tritium H-3 Ca I 3.58E+01 2.94E+O1 2.21E+02 2.39E--02 5.25E+02 C. Dissolved & Entrained Gases Xe-133 - I - 1.80E-05 1.80E-05 Total Activity Ci 1.80E-05 1.80E-05 D. Gross Alpha Gross Alpha I Ci -

measurements below detectable levels 45

Table 2.3-L4 Millstone Unit 3 Liquid Effluents - Batch - CPF Waste Neutralization Sumps, Hotwell, S/G Bulk (Release Point - Quarry)

Nuclides ii' Released Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total A. Fission & Activation Products y Emitters Ci ....

Fe-55 Ci na na na na na Sr-89 Ci na na na na na Sr-90 Ci na na na na na Total Activity Ci -

B. Tritium H-3 I Ci I 1.1IE-02 1.97E-02 2.29E-02 1.84E-02 7.21E-02 C. Dissolved & Entrained Gases y Emitters Ci -

Total Activity Ci D. Gross Alpha Gross Alpha Ci na na na na na

- measurements below detectable levels "na" denotes Not Required to be Analyzed 46

Table 2.3-L5 Millstone Unit 3 Liquid Effluents - Release Summary (Release Point - Yard Drain - DSN 006)

IUnits IlstQtr I2nd Qtr I 3rd Qtr I4th Qtr I -To~tall A. Fission and Activation Products C. Dissolved and Entrained Gases Total Activity Released Average Period Diluted Activity +

D. Gross Alpha Total Activity Ci ----

Released E. Volume Released Waste Liters 4.88E+06 5.98E+06 5.03 E+06 5.49E+06 2.14E+07 Volume

++ 4-4--+ +4-++/--4 Dilution Volume Liters During Releases Dilution Volume Liters 2.74E+07 3.1OE+07 2.69E+07 2.71E+07 1.12 E+08 During Period ++

"-" measurements below detectable levels

+ "Total Activity Released" + ("Released Waste Volume" + "Dilution Volume During Period")

++ Includes all station dilution sources via Yard Drain -

DSN 006

+++Continuous "Dilution Volume During Releases" is not quantified 47

Table 2.3-L6 Millstone Unit 3 Liquid Effluents - Continuous - TB Sump, WTT Berm (Release Point - Yard Drain - DSN 006)

Nuclides II' Released I lUnits 1st Qtr I 2nd Qtr 1 3rd Qtr 4th Qtr I Total A. Fission & Activation Products y Emitters Ci a na n -

Sr-89 Ci na na na na na Sr-90 Ci na na na na na Fe-55 Ci na na na na na Total Activity Ci B. Tritium H-3 Ci 3.71E-02 8.56E-02 8.32E-02 6.99E-02 2.76E-01 C. Dissolved & Entrained Gases y Emitters Ci -

Total Activity Ci D. Gross Alpha Gross Alpha Ci na. na na na na measurements below detectable levels "na" denotes Not Required to be Analyzed 48

Table 2.3-L7 Millstone Unit 3 Liquid Effluents - Continuous - Foundation Drain Sumps (Release Point - Yard Drain - DSN 006)

Nuclides Released Units 1lst Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Products

, Emitters Ci -

Sr-89 Ci na na na na na Sr-90 Ci na na na na na Fe-55 Ci na na na na na Total Activity Ci - I I I I B. Tritium H-3 ] Ci I 3.50E-03 2.50E-03 2.30E-03 2.70E-03 l. 1'OE-02 C. Dissolved & Entrained Gases

, Emitters Ci Total Activity Ci D. Gross Alpha Gross Alpha Ci na na na na na.

- measurements below detectable levels "na" denotes Not Required to be Analyzed 49

2.3 Solid Waste Solid waste shipment summaries for each unit are given in the following tables (Reference 14):

Table 2.1 -S Unit I Solid Waste and Irradiated Component Shipments Table 212-S Unit 2 Solid Waste and Irradiated Component Shipments Table 2.3-S 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 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 95 ft3 122 ft3 Steel Container 202.1 ft3 Steel "Sea Van" 1280 ft3 Polyethylene High Integrity Containers 120.3 ft 3

132.4 Wf 3

173.4 ft 3

202.1 Wf 50

Table 2.1-S Solid Waste and Irradiated Component Shipments Millstone Unit 1 January 1, 2009 through December 31, 2009 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

Resins, Filters, and Evaporator Bottoms Volume Curies Shipped Waste Class ft3 m3 Curies A N/A N/A N/A B N/A N/A N/A C N/A N/A N/A ALL N/A N/A N/A Major Nuclides for the Above Table:

Radionuclide  % of Total Curies CURIES (TOTAL) 0 Dry Active Waste Volume Curies Shipped Waste Class ft3 m3 Curies A 9.OOOOE+01 2.5488E+00 1.0511 E-03 B N/A N/A N/A C N/A N/A N/A ALL 9.0000E+01 2.5488E+00 1.0511 E-03 Major Nuclides for the Above Table:

Radionuclide I %of Total ] Curies Fe-55 46.01% 4.8363E-04 Co-60 19.85% 2.0863E-04 Ni-63 6.68% 7.0251 E-05 Sr-90 0.05% 5.4639E-07 Cs-1 37 27.37% 2.8763E-04 Pu-238 0.01% 1.1608E-07 Pu-239 < 0.01% 8.6569E-09 Am-241 0.02% 1.6853E-07 Cm-244 < 0.01% 7.0635E-08 CURIES (TOTAL) 1.0511E-03 51

Table 2.1-S (continued)

Solid Waste and Irradiated Component Shipments Millstone Unit 1 Irradiated Components Volume Curies Shipped Waste Class ft3 m3 Curies A N/A N/A N/A B N/A N/A N/A C N/A N/A N/A ALL N/A N/A N/A Major Nuclides for the Above Table:

tF Radionuclide  % of Total Curies CURIES (TOTAL) 0 Other Waste Volume Curies Shipped Waste Class ft 3 m3 Curies A 2.8201 E+01 7.9865E-01 1.3294E-03 B N/A N/A N/A C N/A N/A N/A ALL 2.8201 E+01 7.9865E-01 1.3294E-03 Major Nuclides for the Above Table:

Radionuclide [  % of Total ] Curies H-3 72.21% 9.5994E-04 C-14 0.03% 3.7500E-07 Mn-54 < 0.01% 8.1332E-09 Fe-55 2.24% 2.9719E-05 Co-58 < 0.01% 6.1678E-08 Co-60 5.03% 6.6926E-05 Ni-63 2.24% 2.9733E-05 Sr-90 0.09% 1.2400E-06 Tc-99 < 0.01% 5.8749E-10 Sb-124 < 0.01% 1.3991 E-08 Sb-125 < 0.01% 1.1205E-07 Cs-134 < 0.01% 1.2019E-08 Cs-1 37 17.95% 2.3860E-04 Th-230 < 0,01% 4.8958E-13 Np-237 < 0.01% 1.8359E-13 Pu-238 < 0.01% 1.2700E-07 Pu-239 < 0.01% 6.7187E-08 Pu-241 0.14% 1.8600E-06 Am-241 0.03% 4.1200E-07 Cm-244 0.02% 2.1000E-07 CURIES (TOTAL) 1.3294E-03 52

Table 2.1-S (continued)

Solid Waste and Irradiated Component Shipments Millstone Unit 1 Sum of All Low-Level Waste Shipped from Site Volume Curies Shipped Waste Class ft 3 m3 Curies A 1 .1820E+02 3.3474E+00 2.3805E-03 B N/A N/A N/A C N/A N/A N/A ALL 1.1 820E+02 3.3474E+00 2.3805E-03 Major Nuclides for the Above Table:

Radionuclide  % of Total Curies H-3 40.33% 9.5994E-04 C-14 0.02% 3.7500E-07 Mn-54 <0.01% 8.1332E-09 Fe-55 21.57% 5.1335E-04 Co-58 < 0.01% 6.1678E-08 Co-60 11.58% 2.7556E-04 Ni-63 4.20% 9.9983E-05 Sr-90 0.08% 1.7864E-06 Tc-99 < 0.01% 5.8749E-10 Sb-124 < 0.01% 1.3991E-08 Sb-125 < 0.01% 1.1205E-07 Cs-134 < 0.01% 1.2019E-08 Cs-137 22.11% 5.2623E-04 Th-230 < 0.01% 4.8958E-13 Np-237 < 0.01% 1.8359E-13 Pu-238 0.01% 2.4308E-07 Pu-239 < 0.01% 7.5844E-08 Pu-241 0.08% 1.8600E-06 Am-241 0.02% 5.8054E-07 Cm-244 0.01% 2.8064E-07 CURIES (TOTAL) 2.3805E-03 53

Table 2.2-S Solid Waste and Irradiated Component Shipments Millstone Unit 2 January 1, 2009 through December 31, 2009 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

Resins, Filters, and Evaporator Bottoms Volume Curies .Shipped Waste Class ft3 m3 Curies A 4.3684E+01 1.2371 E+00 3.4241 E-02 B 2.4060E+02 6.8138E+00 8.2278E+01 C 5.6140E+0 I 1.5899E+00 4.2685E+00 ALL 3.4043E+02 9.6408E+00 8.6580E+01 Maior Nuclides for the Above Table:

Radionuclide  % of Total Curies H-3 0.03% 2.8337E-02 C-14 0.02% 1.3896E-02 Cr-51 < 0.01% 4.9198E-04 Mn-54 1.42% 1.2282E+00 Fe-55 25.23% 2.1847E+01 Fe-59 < 0.01% 1.5595E-04 Co-57 0.04% 3.8402E-02 Co-58 0.08%. 6.6364E-02 Co-60 5.92% 5.1278E+00 Ni-63 45.67% 3.9540E+01 Zn-65 < 0.01% 2.4944E-04 Sr-89 < 0.01% 4.7298E-03 Sr-90 0.14% 1.2065E-01 Nb-94 < 0.01% 2.3575E-04 Zr-95 0.01% 9.9886E-03 Nb-95 < 0.01% 5.3488E-03 Ru-103 < 0.01% 1.3171 E-07 Ru-106 < 0.01% 2.9013E-04 Ag-110m < 0.01% 4.1804E-03 Sn-1 13 0.06% 4.8598E-02 Sb-124 < 0.01% 5.6180E-07 Sb-125 0.11% 9.7362E-02 Cs-134 5.48% 4.7423E+00 Cs-137 15.72% 1.3613E+01 Ce-141 < 0.01% 1.1040E-11 Ce-144 < 0.01% 1.9452E-03 Hf-181 < 0.01% 1.9001 E-05 Pu-238 < 0.01% 1.1284E-03 Pu-239 < 0.01% 3.8922E-04 Pu-241 0.04% 3.7334E-02 Am-241 < 0.01% 5.8004E-04 Cm-242 < 0.01% 2.0087E-04 Cm-244 < 0.01% 1.3496E-03 CURIES (TOTAL) 8.6580E+01 54

Table 2.2-S (continued)

Solid Waste and Irradiated Component Shipments Millstone Unit 2 Dry Active Waste Volume Curies Shipped Waste Class ft3 m3 Curies A 8.9951E+03 2.5474E+02 3.1575E+00 1B N/A N/A N/A C N/A N/A N/A ALL 8.9951E+03 2.5474E+02 3.1575E+00 Major Nuclides for the Above Table:

I Radionuclide  % of Total Curies H-3 0.12% 3.9183E-03 Cr-51 < 0.01% 5.6739E-08 Mn-54 0.40% 1.2662E-02 Fe-55 39.72% 1.2542E+00 Fe-59 < 0.01% 2.3499E-07 Co-57 < 0.01% 5.3904E-05 Co-58 0.14% 4.4032E-03 Co-60 14.34% 4.5289E-01 Ni-63 37.76% 1.1924E+00 Sr-89 < 0.01% 2.6704E-07 Zr-95 < 0.01% 1.4495E-04 Nb-95 < 0.01% 2.2104E-06 Sn-113 < 0.01% 6.0054E-05 Sb-125 0.44% 1.3796E-02 Cs-134 1.49% 4.7123E-02 Cs-137 5.45% 1.7210E-01 Hf-181 < 0.01% 8.1444E-08 Pu-238 < 0.01% 5.8173E-05 Pu-239 < 0.01% 2.9968E-05 Pu-241 0.11% 3.4883E-03 Am-241 < 0.01% 7.8377E-05 Cm-242 < 0.01% 1 .5957E-06 Cm-244 < 0.01% 6.9885E-05 CURIES (TOTAL) 3.1575E+00 55

Table 2.2-S (continued)

Solid Waste and Irradiated Component Shipments Millstone Unit 2 Irradiated Components Volume Curies Shipped 3 3 Waste Class ft m Curies A N/A N/A N/A B N/A N/A N/A C N/A N/A N/A ALL N/A N/A N/A Major Nuclides for the Above Table:

Radionuclide  % of Total Curies CURIES (TOTAL) 0 Other Waste Volume Curies Shipped Waste Class it 3 m3 Curies A 2.3855E+03 6.7556E+0 1 2.4222E-02 B N/A N/A N/A C N/A N/A N/A ALL 2.3855E+03 6.7556E+01 2.4222E-02 Major Nuclides for the Above Table:

Radionuclide J  % of Total Curies H-3 94.61% 2.2916E-02 C-14 < 0.01% 3.4700E-12 Cr-51 0.26% 6.2070E-05 Mn-54 0.13% 3.2502E-05 Fe-55 0.40% 9.7043E-05 Fe-59 < 0.01% 1.6465E-06 Co-57 < 0.01% 6.3118E-09 Co-58 0.68% 1.6399E-04 Co-60 0.10% 2.4414E-05 Ni-63 0.22% 5.2391 E-05 Zn-65 < 0.01% 2.9268E-10 Zr-95 0.10% 2.4893E-05 Nb-95 0.15% 3.6980E-05 Tc-99 < 0.01% 8.0856E-10 Sn-113 < 0.01% 6.8643E-09 Sb-124 < 0.01% 1.7377E-06 Sb-125 0.06% 1.5310E-05 1-129 < 0.01% 3.3016E-10 1-131 < 0.01% 1.6159E-06 Cs-134 0.12% 2.8422E-05 Cs-137 3.15% 7.6260E-04 Th-230 < 0.01% 6.7380E-13 Np-237 < 0.01% 2.5268E-13 Pu-239 < 0.01% 8.0856E-10 Am-241 < 0.01% 1.4824E- 12 CURIES (TOTAL) 2.4222E-02 56

Table 2.2-S (continued)

Solid Waste and Irradiated Component Shipments Millstone Unit 2 Sum of All Low-Level Waste Shipped from Site Volume Curies Shipped Waste Class mft Curies A 1.1424E+04 3.2353E+02 3.2159E+00 B 2.4060E602 6.8138E,00 8.2278E+01 C 5.6140E+01 1.5899E+00 4.2685E.00 ALL 1.1721E÷04 1 3.3194E+02 8.9762E+01 Major Nuclides for the Above Table:

Radionuclide  % of Total - Curies H-3 0.06% 5.5171 E-02 C-14 0.02% 1.3896E-02 Cr-51 < 0.01% 5.5411E-04 Mn-54 1.38% 1.2409E+00 Fe-55 25.74% 2.3101E+01 Fe-59 < 0.01% 1.5783E-04 Co-57 0.04% 3.8456E-02 Co-58 0.08% 7.0931 E-02 Co-60 6.22% 5.5807E+00 Ni-63 45.38% 4.0733E+01 Zn-65 < 0.01% 2.4944E-04 Sr-89 < 0.01% 4.7300E-03 Sr-90 0.13% 1.2065E-01 Nb-94 < 0.01% 2.3575E-04 Zr-95 0.01% 1.0158E-02 Nb-95 < 0.01% 5.3880E-03 Tc-99 < 0.01% 8.0856E-10 Ru-103 < 0.01% 1.3171E-07 Ru-106 < 0.01% 2.9013E-04 Ag-110m < 0.01% 4.1804E-03 Sn- 113 0.05% 4.8658E-02 Sb-124 < 0.01% 2.2995E-06 Sb-125 0.12% 1.1117E-01 1-129 < 0.01% 3.3016E-10 1-131 < 0.01% 1.6159E-06 Cs-134 5.34% 4.7894E+00 Cs-137 15.36% 1.3786E+01 Ce-141 < 0.01% 1.1040E-11 Ce-144 < 0.01% 1.9452E-03 Hf-181 < 0.01% 1.9082E-05 Th-230 < 0.01% 6.7380E-13 Np-237 < 0.01% 2.5268E-13 Pu-238 < 0.01% 1.1866E-03 Pu-239 < 0.01% 4.1919E-04 Pu-241 0.05% 4,0822E-02 Am-241 < 0.01% 6.5842E-04 Cm-242 < 0.01% 2.0247E-04 Cm-244 < 0.01% 1.4195E-03 CURIES (TOTAL) 8.9762E+01 57

Table 2.3-S Solid Waste and Irradiated Component Shipments Millstone Unit 3 January 1, 2009 through December 31, 2009 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR.DISPOSAL (Not irradiated fuel)

Resins, Filters, and Evaporator Bottoms Volume Curies Shipped Waste Class ft 3 m3 Curies A 6.5714E+01 1.861 OE+00 8.0971 E+00 B 3.4860E+02 9.8724E+00 1.3385E+02 C 6.4160E+01 1.8170E+00 1.0876E+01 ALL 4.7848E+02 1.3550E+01 1.5282E+02 Major Nuclides for the Above Table:

Radionuclide  % of Total Curies H-3 0.19% 2.9081E-01 Cr-51 < 0.01% 8.9145E-06 Mn-54 4.34% 6.6254E+00 Fe-55 26.99% 4.1251E+01 Fe-59 < 0.01% 4.8268E-06 Co-57 0.10% 1.5668E-01 Co-58 1.62% 2.4805E+00 Co-60 6.93% 1.0590E+01 Ni-63 37.77% 5.7725E+01 Zn-65 < 0.01% 1.4834E-07 Sr-89 < 0.01% 4.9689E-03 Sr-90 < 0.01% 6.9617E-03 Zr-95 < 0.01% 9.0412E-03 Nb-95 < 0.01% 6.3211E-03 Sn-113 < 0.01% 5.1787E-05 Sb-125 0.44% 6.6687E-01 Cs-134 11.41% 1.7439E+01 Cs-137 10.19% 1.5566E+01 Ce-144 < 0.01% 1.8370E-07 Pu-238 < 0.01% 8.7091E-05 Pu-239 < 0.01% 3.5815E-05 Pu-241 < 0.01% 3.9247E-03 Am-241 < 0.01% 7.8240E-05 Cm-242 < 0.01% 1.6885E-05 Cm-244 < 0.01% 1.5361 E-04 CURIES (TOTAL) 1.5282E+02 58

Table 2.3-S (continued)

Solid Waste and Irradiated Component Shipments Millstone Unit 3 Dry Active Waste Volume Curies Shipped Waste Class ft 3 m3 Curies A 3.2344E+03 9.1598E+01 3.1300E+00 B N/A N/A N/A C N/A N/A N/A ALL 3.2344E+03 9.1598E+01 3.1300E+00 Major Nuclides for the Above Table:

Radionuclide j  % of Total Curies H-3 3.65% 1.1415E-01 Mn-54 1.65% 5.1507E-02 Fe-55 60.63% 1.8979E+00 Co-58 0.62% 1.9502E-02 Co-60 13.82% 4.3271E-01 Ni-63 18.15% 5.6797E-01 Cs-134 0.81% 2.5337E-02 Cs-137 0.67% 2.1014E-02 CURIES (TOTAL) 3.1300E+00 Irradiated Components Volume Curies Shipped Waste Class ft3 m3 Curies A N/A N/A N/A B N/A N/A N/A C N/A N/A N/A ALL N/A N/A N/A Major Nuclides for the Above Table:

Fladionuclide  % of Total Curies CURIES (TOTAL) 0 59

Table 2.3-S (continued)

Solid Waste and Irradiated Component Shipments Millstone Unit 3 Other Waste Volume Curies Shipped Waste Class ft 3 m3 Curies A 1.9058E+03 5.3972E+01 3.5724E-02 B N/A N/A N/A C N/A N/A N/A ALL 1.9058E+03 5.3972E+01 3.5724E-02 Major Nuclide s for the Above Table:

Radionuclide  % of Total J Curies H-3 98.16% 3.5067E-02 C-14 < 0.01% 1.6799E-12 Cr-51 0.13% 4.5674E-05 Mn-54 0.06% 2.3194E-05 Fe-55 0.19% 6.7691 E-05 Fe-59 < 0.01% 1.2116E-06 Co-57 <0.01% 6.3118E-09 Co-58 0.33% 1.1899E-04 Co-60 0.05% 1.6936E-05 Ni-63 0.10% 3.5806E-05 Zn-65 < 0.01% 2.1167E-10 Zr-95 0.05% 1.8318E-05 Nb-95 0.08% 2.7214E-05 Tc-99 < 0.01% 1.4205E-08 Sn-113 < 0.01% 6.8643E-09 Sb-124 < 0.01% 1.6077E-06 Sb- 125 0.04% 1.3122E-05 1-129 < 0.01% 1.5984E-10 1-131 <0.01% 1.1686E-06 Cs-134 0.05% 1.8226E-05 Cs-137 0.75% 2.6730E-04 Th-230 < 0.01% 1.1837E-11 Np-237 < 0.01% 4.4389E-12 Pu-239 < 0.01% 1.4205E-08 Am-241 < 0.01% 2.6040E-11 CURIES (TOTAL) 3.5724E-02 60

Table 2.3-S (continued)

Solid Waste and Irradiated Component Shipments Millstone Unit 3 Sum of All Low-Level Waste Shipped from Site Volume Curies Shipped Waste Class ft3 m3 Curies A 5.2059E+03 1.4743E+02 1.1263E+01 B 3.4860E+02 9.8724E+00 1.3385E+02 C 6.4160E+01 1.8170E+00 1.0876E+01 ALL 5.6187E+03 1.5912E+02 1.5599E+02 Maior Nuclides for the Above Table:

Radionuclide  % of Total Curies H-3 0.28% 4.4003E-01 C-14 < 0.01% 1.6799E-12 Cr-51 < 0.01% 5.4589E-05 Mn-54 4.28% 6.6769E+00 Fe-55 27.66% 4.3149E+01 Fe-59 < 0.01% 6.0384E-06 Co-57 0.10% 1.5668E-01 Co-58 1.60% 2.5002E+00 Co-60 7.07% 1.1023E+01 Ni-63 37.37% 5.8293E+01 Zn-65 < 0.01% 1.4855E-07 Sr-89 < 0.01% 4.9689E-03 Sr-90 < 0.01% 6.9617E-03 Zr-95 < 0.01% 9.0596E-03 Nb-95 < 0.01% 6.3483E-03 Tc-99 < 0.01% 1.4205E-08 Sn-113 < 0.01% 5.1794E-05 Sb-124 < 0.01% 1.6077E-06 Sb- 125 0.43% 6.6688E-01 1-129 < 0.01% 1.5984E-10 1-131 < 0.01% 1.1686E-06 Cs-134 11.20% 1.7465E+01 Cs- 137 9.99% 1.5588E+01 Ce-144 < 0.01% 1.8370E-07 Th-230 < 0.01% 1.1837E-11 Np-237 < 0.01% 4.4389E-12 Pu-238 < 0.01% 8.7091E-05 Pu-239 < 0.01% 3.5829E-05 Pu-241 < 0.01% 3.9247E-03 Am-241 < 0.01% 7.8240E-05 Cm-242 < 0.01% 1.6885E-05 Cm-244 < 0.01% 1.5361 E-04 CURIES (TOTAL) 1.5599E+02 61

2.4 Groundwater Monitoring The Groundwater Protection Program (GPP) describes the means by which Millstone Station implements the actions cited in the Nuclear Energy's Institute's (NEI) Groundwater Protection Initiative. The purpose of the GPP is to establish a program to assure timely and effective management of situations involving potential releases of radioactive material to groundwater. A key element in the GPP is on-site groundwater monitoring.

The results of the onsite monitoring programs required by the Radiological Environmental Monitoring Program are documented in the Annual Radiological Environmental Monitoring Report; the remaining monitoring programs are documented on the following pages (Tables 2.4-GWI -2.4-GW3).

Another key element in the GPP is site hydrological characterization. The general trend of groundwater flow at the station is toward the Long Island Sound. Although positive measurements of plant related activity are noted in the Tables 2.4-GW 2.4-GW3, there are no pathways to any offsite drinking water supplies. The under-drain (foundation drains) system effectively captures groundwater in the area around Unit 3 and channels this water via the storm drain system to the Long Island Sound. The consequences of these measurements have been used to determine releases listed in Section 2.2 and the dose calculations listed in Section 1.0.

62

Table 2.4-GW1 Environmental Well Monitorina Results Laboratory Analysis Location Well ID March-09 June-09 Sept.-09 Dec.-09 MW-9B *

• TI - Unit I Tank Farm MW-9D *

• TI-MW-I *

• T5 - Abandoned Heating Oil UST - T5-MW-1 * * *

• Building 512 T5-MW-2 * * *

• T6 - Former ROB Heating Oil UST ME-5 *

• T7-MW-I *

• T7 - Former Stone & Webster USTs T7-MW-2 *

• T7-MW-3 *

• SI - Unit 1 Transformer Switchyard SI-MW-1 *

• MW-7C *

• S2 - Unit 2 Transformer Switchyard MW-7D *

• S2-MW-1 *

• ME-9 * * *

• S5 - Former Batch Plant S5-MW-I * * *

• MW-I * * *

• S1-WI *

• S11- Fueling Station S11-MW-i SII-MW-2 *

• S13-MW-1 *

• S13 - Recycling Area Waste Oil AST S13-MW-2 *

• MW-6B *

• M2 - Settling Pond ME-2 * * *

• M5-MW-7 *

• M5-MW-8 *

• M5 - Excavation Pile M5-MW-9S *

• M5-MW-9,D * *

• sampled, gamma and H-3 were LLD 63

Table 2.4-GW2 Catch BasinlUnderdrain Monitoring Results

• Type Location Identification Frequency Results Yard Drains Catch Basin 1-3 CB 1-3 Monthly Gamma and H-3 < LLD Catch Basin 1-5 CB 1-5 Monthly Gamma and H-3 < LLD Catch Basin 1-7 CB 1-7 Monthly Gamma and H-3 < LLD Catch Basin 1-13 CB 1-13 Monthly Gamma and H-3 < LLD Catch Basin 1-14 CB 1-14 Monthly Gamma and H-3 < LLD Catch Basin 1-22 CB 1-22 Monthly Gamma and H-3 < LLD Catch Basin 2-9 CB 2-9 Monthly Gamma and H-3 < LLD Gamma < LLD and, NPDES Discharge DSN 006 Monthly occasionally H-3 at -2000 pCi/liter

• ROB Yard Drain Monthly Gamma and H-3 < LLD ISFSI Yard Drain DMH# II Monthly Gamma and H-3 < LLD Sumps Unit 3 Containment Underdrains Weekly Gamma and H-3 < LLD Gamma < LLD, See next Unit 3 Foundation Underdrains** 3 SRW Sump 2 Quarterly page for H-3 data Gamma < LLD, See next 3 SRW Sump 3 Quarterly page for H-3 data

• Turbine building sumps are discharged via DSN-006. These sumps normally have detectable H-3, which is monitored and reported in the effluent section of this report. Unit 3 Foundation Underdrains (3SRW Sumps 2

& 3) are also discharged via DSN-006

• • New locations added in 2007. See Table 2.3 - L7 for the effluent release results for these locations.

64

Table 2.4-GW3 Under-drain Monitoring Results Sample Foundation Foundation Date Drain Sump 3 Drain Sump 2 (pCi/liter) (pCi/liter) 1/13/09 3660 2/02/09 < 1730 2/26/09 2980 3/10/09 < 1730 3/11/09 2740 4/2/09 2260 < 1730 4/30/09 2310 5/19/09 1880 6/3/09 < 1730 7/1/09 < 1730 < 1730 7/22/09 2050 7/22/09 < 1730 9/3/09 2210 9/3/09 1900 10/1/09 2060 < 1730 i0/30/09 < 1730 12/10/09 2160 65

3.0 Inoperable Effluent Monitors During the period January 1 through December 31, 2009, the following effluent monitors were inoperable for more than 30 consecutive days:

3.1 Unit 1 -None 3.2 Unit 2 - None 3.3 Unit 3 - Vent Flow Rate Monitor (Reference 13)

The Process Vent Flow Rate Monitor (HVR-FE 10) was declared inoperable for 199 days (03/24/09 -

10/09/09) due to erroneous flow readings, subsequent vendor repair and reinstallation. On March 24 the process vent flow rate monitor HVR-FE10 was identified as reading significantly below actual flow.

Both the flow instrument and the associated vent radiation monitor (HVR-RE1OB) were declared inoperable. Based on an engineering evaluation (Reference 4) the radiation monitor was declared operable on April 1 while the flow rate monitor remained inoperable. The flow rate monitor sensor and transmitter were sent to the vendor for repair and calibration. After return from the vendor the sensor and transmitter were re-installed and tested. On October 9 the vent flow rate monitor was declared operable. During the inoperability period, Unit 3 Vent total released radioactivity was determined using estimated flows. First quarter releases were adjusted by factors ranging from 1.6 to 2.0 based on ratios of assumed flows to measured flows and releases for the remainder of the year were determined using flows based on number of running ventilation fans. Compensatory measures were undertaken to ensure releases were appropriately characterized and quantified.

66

4.0 Operating History The operating history of the Millstone Units during this reporting period was as follows:

Unit I was shut down November 11, 1995 with a cessation of operation declared in July 1998.

Unit 2 operated with a DER capacity factor of 81.8% and Unit 3 operated with a DER capacity factor of 96.5%

The power histograms for 2009 are on the following pages.

67

MP2 - CYCLE 19 POWER HISTORY YEAR 2009 Note: Data at 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> intervals 110 100 90 80 70 60 I.U L50 40 30 20 10 0

0) 0) m) , ) C) C) 0) m) C) C) 0) m) ) m) m) m) m) m ) C) m) m) ) C) m C 0 0) 0 0 0D 0 0 0 0 0 0 0 0D 0 0 0 0 0 0 0 0 0 0

(.C0 C-V m0 - 0

• e)

C>

0-- 0-

~ 0 0U m)

CC r--

mN C)

U c) a - C)

NN i-M

\

0 0 o C0 0 0 00D 0 0 0D 0- - - - - -

68

MP3 - CYCLE 13 POWER HISTORY YEAR 2009 Note: Data at 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> intervals 110 100 90 80 70 60 LU.

a. 50 40 40 30 20 10 0

m) 0) 0) 0) 0) 0) 0) 0) 0) C) C) 0) C) 0) m) C) C) C) C) C) (D C) C) C)

S0 0 0 0o0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 C~~~)~ -; 0M ~ ~;69

- C14 0

6 C~

CN a 7-a QN( .

a)

(N

-~ Z- 4' L-0 L-0 zg CD N-

-t- N iz C)

5 ;5 B5 0 0 . N (

o oo 000 0 0 0 0 0 0 0 0 0 0 0 0 -- '

69

5.0 Errata None 6.0 REMODCM Changes The description and the bases of the change(s) for each REMODCM revision are included here in Volume I of the Radioactive Effluent Release Report. In addition, a complete copy of the REMODCM revision(s) for the calendar year 2009 is provided to the Nuclear Regulatory Commission as Volume 2 of the Radioactive Effluent Release Report.

In 2009, there was one revision made to the Millstone REMODCM:

Rev Effective Date 26-01 September 1, 2009 70

6.1 REMODCM - Description of Change(s)

I. Description of changes (include markup pages)

Section/Page Section Title and Description of Change with Basis Table I.D-1/23 In column 'Type of Activity Analysis,' delete the word particulate from the phrase

'Principal Particulate Gamma Emitter' in two places. The word particulate is not needed.

Table I.D-2/25

• In the column 'Type of Activity Analysis,' delete the word 'particulate' for Batch Release sample. This corrects a typographical error.

• To make table consistent with Table I.D-3, change the four release points in Section B, "Containment & Aux Building Releases," to two release points, "Containment" and "Spent Fuel Pool."

" To make table consistent with Table I.D-3, Add word "containment" before "equipment hatch" in three places in 2nd column.

" In column 'Type of Activity Analysis,' delete the word particulate from the phrase 'Particulate Gamma Emitter' for weekly containment equipment hatch and continuous particulate samples. The word particulate is not needed.

" In column 'Sample Type and Frequency,' change word purge to pump Table I.D-3/28 where describing the gaseous grab sample to maintain sub-atmospheric pressure. This is a more accurate description of the process being sampled.

• In column 'Minimum Analysis Frequency,' change 'Monthly for all release sources except Equipment Hatch' to 'Monthly for purges, vents and drawdowns.' This clarifies to which sample types the requirement is applicable.

" In column 'Type of Activity Analysis,' delete the word particulate from the phrase 'Particulate Gamma Emitter' for drawdown samples, weekly containment equipment hatch and continuous particulate samples. The word particulate is not needed.

• Delete the word 'particulate' in 'Type of Activity Analysis' column for monthly gaseous grab sample of continuous releases. This corrects a typographical error.

71

Section/Page Section Title and Description of Change with Basis Tables I.E-1 1. Change number of milk and pasture grass samples from three to two in

& I.E-2/39-41 Table I.E-1 and delete Location 22-1, "Goat Location No. 2" in Table I.E-2.

This sample is no longer available and there are no other milk sample locations within ten miles of Millstone according to the most recent survey.

2. Change the oyster sample Location 36-1. Black Point, 3.0 miles WSW to Location 88-1, DEP Dock near barge slip, 0.2 miles WNW. Collection of oysters at Black Point has been difficult. The new location will be closer to the liquid effluent release point.

Figs. I.E-1 & Change figures to match revisions to sample locations as identified in Table I.E-2/42-43 I.E-2.

ll.D.3/68&70 1. Change dose factors in organ dose for DTS and Dos from 4.66x1 0-5 CHS to 6.33x1 0-6 CHS and from 10.2 Cps to 9.0 Cps. An error in the calculation of these factors was discovered and correction of the error caused the factors to decrease.

2. Change reference to "Method 2a" to "Method 2." Method 2a was changed to Method 2 in a prior REMODCM revision.

3. At top of page 70 in 1st %,change "Section II.D.3.b" to "Method 2." This corrects an internal reference error.

II.E.3/80&81 1. Change noble gas concentration threshold from 0.011 to 5.7x"10 3 uCi/mI and add bases for the threshold. Change is being made because dilution water flow basis is being changed from 200,000 to 100,000 gpm to provide sufficient margin during VFD operation.

2. Delete the note in definition for D and add wording specifying that dilution flow is from circulating and service water pumps. This change is made to eliminate the inference that each circulating water pump provides 100,000 gpm of flow. After implementation of VFDs, this may not always be the case.

3. In Note 2 on page 81 delete the words "worst case conditions." This wording could be misleading.

II.E.4/81 Change setpoint for Condensate Polishing Facility Waste Neutralization Sump Effluent Line from 1.7 x 10-4 to 2.8 x 10-5 uCi/mI and the corresponding reading in CPM from 38,000 to 6,300. Setpoint needed to be lowered to accounted for possible lower flows after installation of VFDs on the circulating water pumps.

72

Section/Page Section Title and Description of Change with Basis II.E.5/82 1. Change setpoint for Unit 2 Steam Generator Blowdown from 8.5 x 10b to 4.3 x 106 uCi/ml and the basis for the setpoint from 200,000 to 100,000 gpm circ water pump flow. Setpoint needed to be lowered to accounted for possible lower flows after installation of VFDs on the circulating water pumps.

2. In last paragraph change words "2 circulating water pumps" to "100,000 gpm circulating water flow." This change is made to eliminate the inference that each circulating water pump provides 100,000 gpm of flow.

After implementation of VFDs, this may always be the case.

II.E.8/85&87 1. Change noble gas concentration threshold from 0.026 to 0.013 uCi/mI and add bases for the threshold. Change is being made because dilution water flow basis is being changed from 200,000 to 100,000 gpm to provide sufficient margin during VFD operation.

2. Delete the note in definition for D and add wording specifying that dilution flow is from circulating and service water pumps. This change is made to eliminate the inference that each circulating water pump provides 100,000 gpm of flow. After implementation of VFDs, this may always be the case.

3. In Note 2 on page 87 delete the words "worst case conditions." This wording could be misleading.

II.E.1 1/87-88 1. In part c of setpoint assumptions, delete the words "minimum possible" and reference to pumps and pump flows and change flow from 230,000 to Iuu,uuu gpm.

2. In setpoint formulae, change flow from 230,000 to 100,000 gpm and setpoint from 1.7x1 05 to 7.5x1 06 uCi/ml.

3. In paragraph on page 88 replace words "2 circulating and 2 service water pumps" with "100,000 gpm dilution flow."

These changes are made to establish setpoint based on operations of the VFDs.

73

SectionfPage Section Title and Description. of Change with Basis 11.F/90-92 The setpoint sections for the five major gaseous effluent release points are being revised. They include:

• Unit 1 SFPI Vent (Section II.F.1 on p. 90)

• Unit 2 releases to stack (Wide Range Gas Monitor in Section II.F.2 on p. 90)

• Unit 3 releases to stack (SLCRS in Section II.F.4 on p. 91)

• Unit 2 Vent (Section II.F.5 on p. 91)

• Unit 3 Vent (Section II.F.7 on p. 92)

In each section the following changes are being made:

2nd & 3rd paragraphs are being swapped so that the 2nd paragraph will address the allocated portion of the site instantaneous release rate limit for that release

" For each release point, the allocated portion of the site limit is being added.

For Unit 1 SFPI Vent this is 7%, for Units 2 and 3 Vents this is 33% each, and for Units 2 and 3 releases to the stack this is 13% each. There is a remaining 1% of site limit which is allocated to Unit 3 ESF Vent. These allocations are not new, but are now being added to the REMODCM.

• Bases information about process flow rates at each release point are being deleted. This information will be transferred to the REMODCM Technical Information Document (MP-22-REC-REF03).

Changes specific to each release point are discussed below.

II.F. 1/90 1. In NOTE in 2nd paragraph (new 3rd paragraph) add "and OU 1" after "OA 1."

the note is applicable to both EAL classifications.

2. Change the Unit 1 portion of site limit from 29,000 uCi/sec to 30,000 uCi/sec.

A recalculation of the release rate value resulted in a slight increase in the value.

However the conservatively lower setpoint based on 29,000 uCi/sec is being maintained.

Change the Unit 2 release to stack portion of site limit from 74,000 uCi/sec to II.F.2/90 72,000 uCi/sec and change setpoint from 1.31 E-2 uCi/cc to 1.3E-2 uCi/cc.

These changes are based on a recalculation of the release rate value.

• Change the Unit 3 release to stack portion of site limit from 74,000 uCi/sec 11.F.4/91 to 72,000 uCi/sec. This changes are based on a recalculation of the release rate value.

• Delete reference to Calculation RERM-01946R3 which is being voided.

74

Section[Page Section Title and Description of Change with Basis II.F.6/91-92 In the 1st paragraph, replace the value of 74,000 uCi/sec with 72,000 uCi/sec. In the 2nd and 3rd paragraphs (p. 92), replace 7,400 uCi/sec with 7,200 uCi/sec. A recalculation of the release rate value resulted in a slight decrease in the value.

II.F.7/92 and Delete reference to Calculation RERM-01976R3 which is being voided.

II.F.8/92 I1.F.9/92 In middle of paragraph delete "(as discussed above for each monitor pathway)." This phrase will not be needed once basis information on release point process flows are removed from the REMODCM.

75

fjI' Serial No.10-196 Docket Nos. 50-245 50-336 50-423 License Nos. DPR-21 DPR-65 NPF-49 ATTACHMENT 2 2009 RADIOACTIVE EFFLUENT RELEASE REPORT VOLUME 2 MILLSTONE POWER STATION UNITS 1, 2, AND 3 DOMINION NUCLEAR CONNECTICUT, INC. (DNC)

MDffeon Po©wer &afioa 2009 Radioactive Effluents Release Report Volume 2 Dominion Nuclear Connecticut, Inc.

MILLSTONE UNIT LICENSE DOCKET 1 DPR-21 50-245 2 DPR-65 50-336

~Domillnionw :3 NPF-49 50-423

Table of Contents Volume II 2009 REMODCM Revision 26-01

Millstone Station REMODCM Revision 26-01 Effective September 1, 2009

MILLSTONE POWER STATION STATION PROCEDURE AA Radiological EffMuent Monitorin g and Off- Site Dose Calculation Manual (REMODCM)

MP-22-REC-BAPO1 Rev. 026-01 Approval Date: 08/18/09 Effective Date: 09/01/09

Millstone All Units Station Procedure Radiological Effluent Monitoring and Off-Site Dose Calculation Manual (REMODCM)

TABLE OF CONTENTS I. Radiological Effluent Monitoring Manual (REMM) ............................ 7 I.A . Introduction .......................... .............................. 7 I.B . Responsibilities ............................................... 7 I.C . Liquid Effluents ..................................................... 7

1. Liquid Effluent Sampling and Analysis Program ..................... 7

2. Liquid Radioactive Waste Treatment .............................. 17

3. Basis for Liquid Sampling, Analysis and Radioactive Treatment System U se .......................................................... 19 I.D . G aseous Effluents ................................................... 23

1. Gaseous Effluent Sampling and Analysis Program ................... 23

2. Gaseous Radioactive Waste Treatment ............................ 31

3. Basis for Gaseous Sampling, Analysis, and Radioactive Treatment System U se .......................................................... 33 I.E. Radiological Environmental Monitoring ................................ 37

1. Sam pling and Analysis .......................................... 37

2. Land U se Census .............................................. 47

3. Interlaboratory Comparison Program ............................. 47

4. Bases for the Radiological Environmental Monitoring Program ....... 48 I.E Report Content ...................................................... 49

1. Annual Radiological Environmental Operating Report ............... 49

2. Radioactive Effluent Release Report ............................... 50 II Off-Site Dose Calculation Manual (ODCM) ................................. 53 II.A. Introduction ........................................................ 53 II.B. Responsibilities ..................................................... 53 II.C. Liquid Dose Calculations ....................................... 54

1. Whole Body Dose from Liquid Effluents ........................... 54

2. Maximum Organ Dose from Liquid Effluents ....................... 55

3. Estimation of Annual Whole Body Dose (Applicable to All Units) ..... 56

4. Estimation of Annual Maximum Organ Dose (Applicable to All Units) . 57

5. Monthly Dose Projections ....................................... 58

6. Quarterly Dose Calculations for Radioactive Effluent Release Report .. 60

7. Bases for Liquid Pathway Dose Calculations........................ 61 MP-22-REC-BAP01 STOP' 'THINK ACT REVIIEW Rev. 026-01 1 of 166

II.D. Gaseous Dose Calculations ........................................... 62

1. Site Release Rate Limits ("Instantaneous") ........................ 62

2. 10 CFR50 Appendix I - Noble Gas Limits ......................... 65

3. 10 CFR50 Appendix I - Iodine, Tritium and Particulate Doses ........ 68

4. Gaseous Effluent Monthly Dose Projections ............... 73

5. Quarterly Dose Calculations for Radioactive Effluent Release Report .. 75

6. Compliance with 40CFR190 ..................................... 76

7. Bases for Gaseous Pathway Dose Calculations ...................... 76 II.E. Liquid Discharge Flow Rates And Monitor Setpoints ...................... 78

1. Unit 1 Reactor Cavity Water Discharge Line ........................ 78

2. R eserved ............................ ......................... 79

3. Unit 2 Clean Liquid Radwaste Effluent Line - RM9049 and Aerated Liquid Radwaste Effluent Line -, RM9116 ......................... 79

4. Condensate Polishing Facility Waste Neutralization Sump Effluent Line - CND245 ......................................... 81

5. Unit 2 Steam Generator Blowdown - RM4262 and Unit 2 Steam Generator Blowdown Effluent Concentration Limitation ...................... 82

6. Unit 2 .Condenser Air Ejector - RM5099 .......................... 83

7. Unit 2 Reactor Building Closed Cooling Water RM6038 and Unit 2 Service Water, and RBCCW Sump and Turbine Building Sump Effluent Concentration Limitation ................................ 83

8. Unit 3 Liquid Waste Monitor - LWS-RE70 ....................... 85

9. Unit 3 Regenerant Evaporator Effluent Line - LWC-RE65 ......... 87

10. Unit 3 Waste Neutralization Sump Effluent Line - CND-RE07 ...... 87

11. Unit 3 Steam Generator Blowdown - SSR-RE08 and Unit 3 Steam Generator Blowdown Effluent Concentration Limitation ............. 87

12. Unit 3 Turbine Building Floor Drains Effluent Line - DAS-RE50 and Unit 3 Service Water and Turbine Building Sump Effluent Concentration Lim itation ....................................... 89

13. Bases for Liquid Monitor Setpoints ............................... 89 II.E Gaseous M onitor Setpoints ........................................... 90

1. Unit 1 Spent Fuel Pool Island Monitor - RM-SFPI-02 ............. 90

2. Unit 2 Wide Range Gas Monitor (WRGM) - RM8169 .............. 90

3. Reserved ..................................................... 90

4. Unit 3 SLCRS - HVR-RE19B .................................. 91

5. Unit 2 Vent - Noble Gas Monitor - RM8132B ..................... 91

6. Unit 2 Waste Gas Decay Tank Monitor RM9095 .................... 91

7. Unit 3 Vent Noble Gas Monitor - HVR-REIOB ................... 92

8. Unit 3 Engineering Safeguards Building Monitor - HVQ-RE49 ...... 92

9. Bases for Gaseous Monitor Setpoints .............................. 92 MP-22-REC-BAP01

STOP'I* .THINK .,ACT REVIEW Rev. 026-01 2 of 166

III REMODCM Unit One Controls ...................................... 98 III.A. Introduction ................................................ 96 III.B. Definitions and Surveillance Requirement (SR) Applicability .............. 96 III.C. Radioactive Effluent Monitoring Instrumentation ........................ 99

1. Radioactive Liquid Effluent Monitoring Instrumentation ............. 99

2. Radioactive Gaseous Effluent Monitoring Instrumentation .......... 102 III.D. Radioactive Effluents Concentrations And Dose Limitations .............. 106

1. Radioactive Liquid Effluents .................................... 106

2. Radioactive Gaseous Effluents .................................. 108 III.E. Total Radiological Dose From Station Operations Controls ............... 112 III.E Bases .................................................... 112 IV REMODCM Unit Two Controls ........................................... 117 IV.A . Introduction ...................................................... 117 IVB. Definitions, Applicability and Surveillance Requirements ................ 117 IV.C. Radioactive Effluent Monitoring Instrumentation.................... 121

1. Radioactive Liquid Effluent Monitoring Instrumentation ............ 121

2. Radioactive Gaseous Effluent Monitoring Instrumentation .......... 125 IV.D. Radioactive Effluents Concentrations And Dose Limitations .............. 129

1. Radioactive Liquid Effluents .................................... 129

2. Radioactive Gaseous Effluents .................................. 131 IV.E. Total. Radiological Dose From Station Operation ....................... 135 IV E B ases .............................................................. 135 V REMODCM Unit Three Controls ......................................... 139 V.A . Introduction ...................................................... 139 VB. Definitions and Applicability and Surveillance Requirements ............. 139 VC. Radioactive Effluent Monitoring Instrumentation ....................... 142

1. Radioactive Liquid Effluent Monitoring Instrumentation ............ 142

2. Radioactive Gaseous Effluent Monitoring Instrumentation .......... 147 VD. Radioactive Effluents Concentrations And Dose Limitations .............. 152

1. Radioactive Liquid Effluents ........................... ......... 152

2. Radioactive Gaseous Effluents .................................. 154 VME. Total Radiological Dose From Station Operations ....................... 158 V E B ases ............................................................ 158 MP-22-REC-BAP01 sTOP THINK AT RE;VIEW Rev 026-01 3 of 166

TABLES AND FIGURES TABLES Table I.C.-1, "Millstone Unit 1 Radioactive Liquid Waste Sampling and A nalysis Program " . ..... ......................................... 9 Table I.C.-2, "Millstone Unit 2 Radioactive Liquid Waste Sampling and A nalysis Program " . ..... ........................................ 11 Table I.C.-3, "Millstone Unit 3 Radioactive Liquid Waste Sampling and A nalysis Program " . ............................................. 14 Table I.D.- 1, "Millstone Unit 1 Radioactive Gaseous Waste Sampling and Analysis Program" ........................................ 23 Table I.D. -2, "Millstone Unit 2 Radioactive Gaseous Waste Sampling and A nalysis Program " . ............................................. 25 Table I.D.-3, "Millstone Unit 3 Radioactive Gaseous Waste Sampling and A nalysis Program " . ............................................. 28 Table I.E.-1, "Millstone Radiological Environmental Monitoring Program" ... 39 Table I.E. -2, "Environmental Monitoring Program Sampling Locations"....... 40 Table I.E.-3, Reporting Levels For Radioactivity Concentrations In Environmental Sam ples ....................................................... 44 Table I.E. -4, Maximum Values For Lower Limits Of Detection (LLD) ........ 45 Table App. II.A. -1, "Millstone Effluent Requirements and Methodology Cross Reference" ......................................................

Table III.C.-1, "Radioactive Liquid Effluent Monitoring Instrumentation" ... 100 Table III.C.-2, "Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements" . .................................... 101 Table III.C.-3, "Radioactive Gaseous Effluent Monitoring Instrumentation" . 103 Table III.C.-4, "Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements" . .................................... 105 Table IVC.-1, "Radioactive Liquid Effluent Monitoring Instrumentation" 122 Table IV.C.-2, "Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements" . .................................... 124 Table IV.C.-3, "Radioactive Gaseous Effluent Instrumentation" .. ........... 126 Table IVC.-4, "Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements" . .................................... 128 Table V.C.-1, "Radioactive Liquid Effluent Monitoring Instrumentation" .... 143 Table VC.-2, "Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements" . .................................... 146 Table V.C.-3, "Radioactive Gaseous Effluent Monitoring Instrumentation" .. 148 Table V.C.-4, "Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements" . .......................................... 150 MP-22-REC-BAP01 STOP THINK ACT REVI EW Rev. 026-01 4 of 166

FIGURES Figure I.C.- 1, "Reserved" . ............................................ 20 Figure I.C.-2, "Simplified Liquid Effluent Flow Diagram Millstone Unit 2 .... 21 Figure I.C.-3, "Simplified Liquid Effluent Flow Diagram Millstone Unit 3 .... 22 Figure I.D.- 1, "Simplified Gaseous Effluent Flow Diagram Millstone U nit O ne" . ..... .............................................. 34 Figure I.D.-2, "Simplified Gaseous Effluent Flow Diagram Millstone Unit Two" .............................................. 35 Figure I.D.-3, "Simplified Gaseous Effluent Flow Diagram Millstone U nit Three" . ..... ............................................. 36 Figure I.E.-1, "Inner Air Particulate And Vegetation Monitoring Stations" ... 42 Figure I.E.-2, "Outer Terrestrial Monitoring Stations ...................... 43 Figure III.D.-1, "Site Boundary for Liquid and Gaseous Effluents" . ........ 111 Figure IV.D. -1, "Site Boundary for Liquid and Gaseous Effluents"........... 134 Figure V.D.-1, "Site Boundary for Liquid and Gaseous Effluents" . ......... 157

-I MP-22-REC-BAP01 STOP 'THINK ACT IFREVIEW Rev. 026-01 5 of 166

SECTION I.

Radiological Effluent Monitoring Manual (REMM)

For the Millstone Nuclear Power Station Nos. 1, 2, & 3 Docket Nos. 50-245, 50-336, 50-423

,. MP REC-BAP01 STOP THINK' ACT REVIEW Rev. 026-01 6 of 166

SECTION I. RADIOLOGICAL EFFLUENT MONITORING MANUAL (REMM)

I.A. Introduction The purpose of Section I of this manual is to provide the sampling and analysis programs which provide input to Section II for calculating liquid and gaseous effluent concentrations and offsite doses. Guidelines are provided for operating radioactive waste treatment systems in order that offsite doses are kept As- Low-As - Reasonably- Achievable (ALARA).

The Radiological Environmental Monitoring Program outlined within this manual provides confirmation that the measurable concentrations of radioactive material in the environment as a result of operations at the Millstone Site are not higher than expected.

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

MP REC- REF03, "REMODCM Technical Information Document (TID)," has additional bases and technical information. It also contains a list of exceptions to Regulatory Guide 1.21 (see Section 2 of the TID).

I.B. Responsibilities All changes to the Radiological Effluent Monitoring Manual (REMM) shall be reviewed and approved by the Facility Safety Review Committee prior to implementation.

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

It shall be the responsibility of the Site Vice President Millstone to ensure that this manual is used as required by the administrative controls of the Technical Specifications. The delegation of implementation responsibilities is delineated in MP REC- PRG, "Radiological Effluent Program."

I.C. Liquid Effluents

1. Liquid Effluent Sampling and Analysis Program Radioactive liquid wastes shall be sampled and analyzed in accordance with the program specified in Table I.C. -1 for Millstone Unit No. 1, Table I.C. -2 for Millstone Unit No. 2, and Table I.C.-3 for Millstone Unit No. 3. The results of the radioactive analyses shall be input to the methodology of Section II to assure that the concentrations at the point of release are MP-22-REC-BAP01 STOP THINK ACT REVEW Rev. 026--01 7 of 166

maintained within the limits of Radiological Effluent Controls (Section III.D.l.a. for Millstone Unit No. 1, Section IVD.I.a. for Millstone Unit No.

2, and Section V.D.I.a. for Millstone Unit No. 3).

MP-22-REC-BAP01 STOP" -THINKý AC.T R.EIEW Rev. 026-01 8 of 166

Table I.C.-1 Millstone Unit 1 Radioactive Liquid Waste Sampling and Analysis Program Liquid Release Sample Type and Minimum Analysis 1Tpe of Activity Lower Limit Source Frequency Frequency Analysis of Detection (LLD)A (giCi/ml)

Any Batch Grab sample prior to Prior to each batch re- Principal Gamma. 5 x 10-7 Release from any each batch releaseB lease Emitters source Kr-85 I x 10-5 Prior to initial batch re- H-3 1 x 10-5 lease from any one source and monthly composite thereafterC Grab sample prior to Prior to initial batch re- Gross alpha 1 x 10-7 initial batch release lease from any one from any one source source and quarterly Sr-90 5 x 10-8 and quarterly compos- thereafter ite thereafter Fe-55 1 x 10-6 Table I.C.-1 TABLE NOTATIONS A, The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

Tr

£L,U~.

=-

-)

4.66 Sb (E)(V)(2.22x10 6)(ye -WI)

Where:

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

- Sb is the standard deviation of the background counting rate or of the counting rate of a blank 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 .tCi Y is the fractional radiochemical yield (when applicable)

SX, is the radioactive decay constant for the particular radionuclide

• At is the elapsed time between midpoint of sample collection and midpoint of counting time MP-22-REC-BAPO1 STOP, TH.INK ACT REVIEW Rev. 026-01 9 of 166

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

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

B. Prior to the sampling, each batch shall be isolated and at least two tank/sump volumes shall be recirculated or equivalent mixing provided.

C. A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. Prior to analysis, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluents released.

.t-MP-22-REC-BAP01 STOP" 'THINK ACT REVIEW Rev 026-01 10 of 166

Table I.C.-2 Millstone Unit 2 Radioactive Liquid Waste Sampling and Analysis Program Liquid Release Sample Iype and Minimum Analysis Type of Activity Lower Limit Source Frequency Frequency Analysis of Detection (LLD)A (RCi/ml)

A.Batch ReleaseB-I.Clean Waste Moni- Grab sample prior to Prior to each batch Principal Gamma 5 x 10-7 tor Tank, Aerated each batch release release EmittersC" Waste Monitor 1 X 10-6 Tank and Steam 1-131 Generator BulkD1 Ce- 144 5 x 10-6 Dissolved & I x 10-5 Entrained GasesK-2.Condensate Monthly H-3 1 x 10-5 Polishing Facility CompositeF-,G.

- Waste Quarterly Gross alpha 1x 10-7 SumpEl CompositeF"G. Sr-89, Sr-90 5 x 10-8 Fe-55 1 x 10-6 B.Continuous Release l.Steam Generator Daily Grab Sample'& Weekly Principal Gamma 5 x 10-7 BlowdownH. prior to aligning to CompositeEG. EmittersC.

2.Service Water Long Island Sound for 1-131 1 x 10-6 EffluentJ- RBCCW sump Ce-144 5 x10-6

_-145x1-3.Turbine SumpsL. Monthly Grab Monthly Dissolved & I x 10-5 Sample Entrained GasesK" 4.RBCCW SumpM" Weekly Grab or Com- Monthly H-3N. 1 x 10-5 posite CompositeF,G.

Weekly Composite Quarterly Gross alpha 1 x 10-7 CompositeF.G Sr-89, Sr-90 5 x 10-8 Fe-55 1 x 10-6 TABLE I.C.-2 TABLE NOTATIONS A. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

4.66 Sb LLD =

(E)(V)(2.22x106)(Ye-,'t-'J1)

MP-22-REC-BAPO1 STOP THINK ACT REVIEW Rev. 026-01 11 of 166

Where:

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

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

• E is the counting efficiency (as counts per transformation) 0 V is the sample size (in units of mass or volume)

• 2.22 x 106 is the number of transformations per minute per [tCi

• Y is the fractional radiochemical yield (when applicable)

• k 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 priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

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

B. A batch release is the discharge of liquid wastes of a discrete volume from the tanks listed in this table. Prior to the sampling, each batch shall be isolated and at least two tank/sump volumes shall be recirculated or equivalent mixing provided. Ifthe steam generator bulk can not be recirculated prior to batch discharge, samples will be obtained by representative compositing during discharge.

C. The LLD will be 5 x 10-7 [Ci/ml. The principal gamma emitters for which this LLD applies are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 10-6

[tCi/ml. This list does not mean that only these nuclides are to be detected and reported.

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in a priori LLDs higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.

D. For the Steam Generator Bulk:

IF the applicable batch gamma activity is not greater than 5 x 10-7 pCi/ml, THEN the sampling and analysis schedule for gross alpha, Sr-89, Sr-90, Fe-55 are not required.

MP-22-REC-BAP01 STOP THIiNK ACT RE.VIEW Rev. 026-01 12 of 166

E. For the Condensate Polishing Facility (CPF) waste neutralization sump:

IF there is no detectable tritium in the steam generators, THEN tritium sampling and analyses is not required.

IF the gross gamma activity in the grab sample taken prior to release does not exceed 5 x 10-7 0Ci/ml, THEN the sampling and analysis schedule for gross alpha, Sr-89, Sr-90 and Fe-55 are not required.

F. For Batch Releases and Steam Generator Blowdown only, a composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

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

H. For the Steam Generator Blowdown:

IF the steam generator gross gamma activity does not exceed 5 x 10-7 p.Ci/ml, THEN the sampling and analysis schedule for all principal gamma, 1-131, Ce-144, noble gases, gross alpha, Sr-89, Sr-90 and Fe-55 are not required.

1. Daily grab samples shall be taken at least five days per week. For service water, daily grabs shall include each train that is in-service.

J. For the Service Water:

LE a weekly gamma analysis does not indicate a gamma activity greater than 5 xl 0-7 pCi/ml, THEN the sampling and analysis schedule for gross alpha, Sr-89, Sr-90, Fe-55 are not required.

K. LLD applies exclusively to the following radionuclides: Kr-87, Kr-88, Xe-133, Xe- 133m, Xe- 135, and Xe- 138. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in a priori LLDs higher than required, the reasons shall be documented in the "Radioactive Effluent Release Report."

L. For the Turbine Building Sump:

IF there is no detectable tritium in the steam generators, THEN tritium sampling and analyses is not required.

IF the steam generator gross gamma activity does not exceed 5 x 10- 7 0Ci/ml, OR sump is directed to radwaste treatment, THEN the sampling and analysis schedule for all principal gamma, 1-131, Ce-144, noble gases, gross alpha, Sr-89, Sr-90 and Fe-55 are not required.

IF the release pathway is directed to yard drains, THEN the LLD for 1-131 shall be 1.5 x 10-7

[tCi/ml and for gross alpha 1 x 10-8 ItCi/ml.

M. For the RBCCW Sump:

IF the RBCCW Sump is directed to radwaste treatment or is not aligned to Long Island Sound, THEN sampling is not required.

IF the applicable batch gamma activity is not greater than 5 x 10-7 Ci/ml, THEN sampling and analysis schedule for gross alpha, Sr-89, Sr-90, Fe-55 are not required.

N. Detectable tritium shall be used to estimate tritium releases to the atmosphere via the blowdown tank vent.

MP-22-REC-BAPO1 STOP THINK ACT REVIEW Rev. 026-01 13 of 166

Table I.C.-3 Millstone Unit 3 Radioactive Liquid Waste Sampling and Analysis Program Liquid Release Sample Type and Minimum Analysis Type of Activity Lower Limit Source Frequency Frequency Analysis of Detection (LLD)A (RCi/ml)

A.Batch ReleaseB.

1.Condensate Polish- Grab sample prior Prior to each batch Principal Gamma 5 x 10-7 ing Facility Waste to each batch release release EmittersC" Neutralization SumpE. 1-131 1 x 10-6 Ce-144 5 x 10-6 Dissolved & I x 10-5 Entrained GasesK-2.Waste Test Tanks, Monthly H-3 1 x 10-5 Low Level Waste CompositeF.,G.

Tank, Boron Test Tanks and Steam Quarterly Gross alpha 1x Generator Bulk D. CompositeF"G6 Sr-89, Sr-90 5 x 10-8 Fe-55 1 x 10-6 B.Continuous Release 1.Steam Generator Daily Grab SampleI- Weekly Principal Gamma 5 x 10-7 BlowdownH" CompositeF.,G. EmittersC-2.Service Water Ef- 1-131 1 x 10-6 fluentJ. Ce-144 5 x 10-6 3.Turbine Building Monthly Grab Monthly Dissolved & 1 x 10-5 SumpsL" Sample Entrained GasesK-Weekly Grab or Coin- Monthly H-3M- 1 x 10-5 posite CompositeF.,G.

Weekly Composite Quarterly Gross alpha I x 10.7 CompositeF"G" Sr-89, Sr-90 5 x 10-8 Fe-55 1 x 10-6 TABLE INOAI.-ON TABLE NOTATIONS A. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

MP REC- BAP01 STOP THINK ACT REVIEW Rev. 026-01 14 of 166

For a particular measurement system (which may include radiochemical separation):

LLD = .4.66 Sb (E)(TV)( Z.Z2x I O6)( ye-A)

Where:

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

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

• E is the counting efficiency (as counts per transformation) o V is the sample size (in units of mass or volume)

• 2.22 x 106 is the number of transformations per minute per ýtCi V is the fractional radiochemical yield (when applicable)

Y

• *, 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 priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

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

B. A batch release is the discharge of liquid wastes of a discrete volume from the tanks listed in this table. Prior to the sampling, each batch shall be isolated and at least two tank/sump volumes shall be recirculated or equivalent mixing provided. Ifthe steam generator bulk can not be recirculated prior to batch discharge, samples will be obtained by representative compositing during discharge.

C. The LLD will be 5 x 10-7 0Ci/ml. The principal gamma emitters for which this LLD applies are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 10-6 pCi/ml. This list does not mean that only these nuclides are to be detected and reported.

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in a priori LLDs higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.

D. For the Steam Generator Bulk:

IF the applicable batch gamma activity is not greater than 5 x 10-7 pCi/ml, THEN the sampling and analysis schedule for gross alpha, Sr-89, Sr-90, Fe-55 are not required.

MP-22-REC-BAPO1 STOP THINK ACT REVIEW Rev 026-01 15 of 166

E. For the Condensate Polishing Facility (CPF) waste neutralization sump:

IF there is no detectable tritium in the steam generators, THEN tritium sampling and analyses is not required.

IF the gross gamma activity in the grab sample taken prior to release does not exceed 5 x 10- [.Ci/ml, THEN the sampling and analysis schedule for gross alpha, Sr-89, Sr-90 and Fe-55 are not required.

F. For Batch Releases and Steam Generator Blowdown only, a composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

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

H. For the Steam Generator Blowdown:

IF the steam generator gross gamma activity does not exceed 5 x 10-7 [tCi/ml, THEN the sampling and analysis schedule for all principal gamma, 1- 131, Ce- 144, noble gases, gross alpha, Sr-89, Sr-90 and Fe-55 are not required.

Steam Generator Blowdown samples are not required when blowdown is being recovered.

Daily grab samples shall be taken at least five days per week. For service water, daily grabs shall include each train that is in-service.

J. For the Service Water:

IF a weekly gamma analysis does not indicate a gamma activity greater than 5 x10- 7 [tCi/ml, THEN the sampling and analysis schedule for gross alpha, Sr-89, Sr-90, Fe-55 are not required.

K. LLD applies exclusively to the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level. When unusual circumstances result in a priori LLDs higher than required, the reasons shall be documented in the "Radioactive Effluent Release Report."

L. For the Turbine Building Sump:

IF there is no detectable tritium in the steam generators, THEN tritium sampling and analyses is not required.

IF the steam generator gross gamma activity does not exceed 5 x 10-7 [iCi/ml, OR sump is directed to radwaste treatment, THEN the sampling and analysis schedule for all principal gamma, 1-131, Ce- 144, noble gases, gross alpha, Sr-89, Sr-90 and Fe-55 are not required.

IF the release pathway is directed to yard drains, THEN the LLD for 1-131 shall be 1.5 x 10-7

!iCi/ml and for gross alpha 1 x 10-8 [tCi/ml.

M. Detectable tritium shall be used to estimate tritium releases to the atmosphere via the blowdown tank vent.

MP-22-REC-BAP01' STOP THJNK ACT REVIEW Rev. 026-01 16 of 166

2, Liquid Radioactive Waste Treatment

a. Dose Criteria for Equipment Operability Applicable to All Millstone Units The following dose criteria shall be applied separately to each Millstone unit.

1) IF the radioactivity concentration criteria for the Unit 3 steam generator blowdown is exceeded with blowdown recovery not available to maintain releases to as low as reasonably achievable; or, 1F any of the other radioactive waste processing equipment listed in Section b. are not routinely operating, THEN doses due to liquid effluents from the applicable waste stream to unrestricted areas shall be projected at least once per 31 days in accordance with the methodology and parameters in Section II.C.5.

2) IF any of these dose projections exceeds 0.006 mrem to the total body or 0.02 mrem to any organ, THEN best efforts shall be made to return the processing equipment to service,.or to limit discharges via the applicable waste stream.

3) IF an actual dose due to liquid effluents exceeds 0.06 mrem to the total body or 0.2 mrem to any organ AND the dose from the waste stream with processing equipment not operating exceeds 10% of one of these limits, THEN prepare and submit to the Commission a Special Report within 30 days as specified in Section 2.c.

b. Required Equipment for Each Millstone Unit Best efforts shall be made to return the applicable liquid radioactive waste treatment system equipment specified below for each unit to service or to limit discharge via the applicable waste stream if the projected doses exceed any of the doses specified above.

MP-22-REC-BAP01

STOP 'THINK' ACT REVE Rev. 026-01 17 of 166

1. Millstone Unit No. 1 Waste Stream Processing Equipment Reactor cavity water One filter and one demineralizer

2. Millstone Unit No. 2 Waste Stream Processing Equipment Clean liquid Deborating ion exchanger (T1l) OR Purification ion exchanger (T10A or T10B) OR Equivalent ion exchanger Primary demineralizer (T22 A or B) OR Equivalent demineralizer Secondary demineralizer (T23 A or B) OR Equivalent demineralizer//Aerated liquid Aerated liquid Demineralizer (T24) OR Equivalent demineralizer

3. Millstone Unit No. 3 Waste Stream Processing Equipment or Radioactivity Concentration High level Demineralizer filter (LWS-FLT3) and Demineralizer (LWS-DEMN2) OR Demineralizer (LWS-DEMN1) and Demineralizer filter (LWS-FLT1)

Boron recovery Cesium ion exchanger (DEMN A or B)

Boron evaporator (EV- 1)

Low level High level processing equipment Steam generator Blowdown recovery when total gamma activity exceeds 5E-7 [tCi/ml blowdown or tritium activity exceeds 0.02 gCi/ml.

c. Report Requirement For All Three Millstone Units If required by Section 2.a.3), prepare and submit to the Commission a Special Report within 30 days with the following content:

" Explanation of why liquid radwaste was being discharged without treatment, identification of any equipment not in service, and the reason for the equipment being out of service,

• Action(s) taken to restore the equipment to service, and

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

MP-22-REC-BAPO1 P THIN.K 'ACT REVIEW Rev. 026-01 18 of 166

3. Basis for Liquid Sampling, Analysis and Radioactive Treatment System Use Paragraph (a)(2) of Part 50.36a provides that licensee will submit an annual report to the Commission which specifies the quantity of each of the principal radionuclides released to unrestricted areas in liquid effluents during the past 12 months of plant operation. The indicated liquid surveillance programs (as directed by surveillance requirements for Radiological Effluent Controls in Sections III.D.l.a., IV.D.1.a., and V.D.1.a.)

provides the means to quantify and report on liquid discharges from release pathways. As specified in Regulatory Guide 1.21, this program monitors all major and potentially significant paths for release of radioactive material in liquid effluents during normal reactor operations, including anticipated operational occurrences. There are many minor release pathways which are not routinely monitored. The Millstone Effluent Control Program includes, as needed, evaluations to determine if any release point should be added to the REMODCM surveillance program. This information also provides for the assessment of effluent concentiations and environmental dose impacts for the purpose of demonstration compliance with the effluent limits of 10 CFR 20, and dose objectives of 10 CFR 50, Appendix I. The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of Lower Limits of Detection (LLDs) and are selected such that the detection of radioactivity in effluent releases will occur at levels below which effluent concentration limits and off-site dose objectives would be exceeded. The LLDs are listed in Table 4.11-1 of NUREG-1301 except for the LLD for Ce -144 which is contained in Footnote (3) of Table 4.11-1 of NUREG-1301.

The indicated liquid radwaste treatment equipment for each Unit have been determined, using the GALE code, to be capable to minimize radioactive liquid effluents such that the dose objectives of Appendix I can be met for expected routine (and anticipated operational occurrence) effluent releases.

This equipment is maintained and routinely operated to treat appropriate liquid waste streams without regards to projected environmental doses.

If not already in use, the requirement that the appropriate portions of the liquid radioactive waste treatment system for each Unit be returned to service when the specified effluent doses are exceeded provides assurance that the release of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This condition of equipment usage implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50, and the design objective given in Section II.D.

of Appendix I to 10 CFR 50. The specified dose limits governing the required use of appropriate portions of the liquid radwaste treatment system were selected as a suitable fraction of the dose design objectives set forth in Section II.A. of Appendix 1, 10 CFR 50 for liquid effluents following the guidance given in NUREG- 1301.

MP-22-REC-BAPO1 STOP THINK AC:T, REVIEWRev.026-01 19 of 166

Figure I.C.-1, "Reserved MP-22-REC-BAP01 STOp THNK "ACT REVIEW Rev. 026-01 20 of 166

U)

-1 C.

1, mN.

M Sampling Drains I r t WAerated Waste Equipment Drains Aerated Waste Ion Exchanger Montor Mon 0M oM Drains Tanks (T24) 0 Decontamination (T20A/B) Tank (T21)

":M (Note 2) "............

Portable Demins Turbine Bldg. Floor Drains (Note 1) LE..............................

TI 41 A/B/C 7

Note 1: Turbine Bldg floor drains normally to storm drains.

Drains may be diverted to the radwaste treatment system. . . . .=..........

Component included in radwaste Note 2: HP Decon facility sump is collected in totes and shipped offsite. ......... treatment equipment requirement

Containment Bldg Sump*

High Level .  : .................... ...............

Aux Bldg Sump*

Waste Drain

• Filter Demin WTeTest D TakTank Demin .+: Filter Reactor Plant Sample Sinks Tank (LWS-FLT3) - (LWS-DEMN2): (LWS-TK3A/B) (LWS-EMNI) (LW.-FLT") .I

.w......)........ . ...... ................. .. ..................

Lab Wastes* t Misc High-Level Waste Volume To Primary Grade Water I.

)rain System I Control Cl) 0, Tank Boron Recovery Boron Boron Tank Test Tank ioron B

S(BS-TKtA/B) (BRS-TK2A/B) D*EmNeralize,

,z (G," GI '

Degassifier ... . . . . . . . . . t. . . .

j GG ) Cesium *

. . . . . . . . L.:

Boron Filte Ion Exchanger Evaporator 0° Reactor Plant Gaseous Drains (CHS-DEMNIA/B) (EVI) (BRStFLT3)

.... I......

Filter (FLT4A/B)

-s.

Low-Level Miscellaneous Low-Level Waste Waste Drain Tank (LWS-TK4A/B) m5 Turbine Plant Leakage to Sump Niatc Bay -3'Wrý0 Vjia DS 06 0'

(IDAR-RE5O) IeR\

<~k CPF Waste Neut Sump (Tkl0 & Tkl 1)

Steam Generator Blowdown (Onen Cygle) (vCND-RF07c 4

......... Component included in radwaste treatment OSSR-RF081

= equipment requirement Quarry

I.D. Gaseous Effluents

1. Gaseous Effluent Sampling and Analysis Program Radioactive gaseous wastes shall be sampled and analyzed in accordance with the program specified in Table I.D.- 1 for Millstone Unit No. 1, Table I.D.-2 for Millstone Unit No. 2, and Table I.D.-3 for Millstone Unit No. 3.

The results of the radioactive analyses shall be input to the methodology of Section II to assure that offsite dose rates are maintained within the limits of Radiological Effluent Controls (Section III.D.2.a. for Millstone Unit No. 1, Section IVD.2.a. for Millstone Unit No. 2, and Section V.D.2.a. for Millstone Unit No. 3).

Table I.D. -1 Millstone Unit 1 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Release Sample T'ype and Minimum Ty(pe of Activity Lower Limit Point or Source Frequency Analysis Analysis of Detection Frequency (LLD)A (RCi/ml)

A.Spent Fuel Pool Monthly) - Gaseous Monthly Kr-85 1 x 10-4 Island Vent Grab Sample H-3 1 x 10-6 Twice per month Principal Gamma 1 x 10-11 (0

ContinuousB',E' EmittersC- - (with half Particulate Sample lives greater than 8 days)

ContinuousB.,E. Quarterly Sr-90, Gross alpha I x 10-z1 Particulate Sample Composite ContinuousB',E- Continuous Kr-85 1 x 10-6 Noble Gas Monitor (0 B.Balance of Twice per month Principal Gamma 1 x 10-11 Plant Vent EmittersC" - (with half ContinuousB.,'. lives greater than 8 days)

Particulate Sample Quarterly Sr-90, Gross alpha 1 x 10-11 Composite Grab sample of Reactor Prior to processing H-3 1 x 10-5 Bldg evaporator staging of each batch tank prior to processing Table I.D.-1 TABLE NOTATIONS A. The lower limit of detection (LLD) is defined in Table Notations, Item a, of Tables I.C.-1, I.C.-2, or I.C.-3.

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

MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 23of 166

C. For particulate samples, the LLD will be 1 x 10-11 [tCi/cc. The principal gamma emitters for which this LLD applies are exclusively the following radionuclides: Mn-54, Co-60, Zn-65, Cs-134, Cs-137, and Ce-144. 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 above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in a priori LLDs higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.

D. IFthere is an unexplained increase of the SFPI Vent noble gas monitor of greater than a factor of ten, OR the monitor reads 8.8E-5 [tCi/cc or greater, THEN sampling and analysis shall also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

E. Continuous when exhaust fans are in operation.

STOP. MP-22-REC-BAP01 THINK ACT REVIEW Rev. 026-01 24 of 166

Table I.D.-2 Millstone Unit 2 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Sample Type and Minimum Type of Activity Lower Limit Release Point Frequency Analysis Analysis of Detection or Source Frequency (LLD)A (ftCi/ml)

A.Batch Release Principal Gamma 1 x 10-4 0O Each Tank Discharge EmittersB 1.Waste Gas Stor- Gaseous Grab Prior to each age TankH Waste Gas Tank Discharge H-3 1 x 10-6 B.Containment&Aux Building Releases 1.Containment Gaseous Grab of purges 1. Prior to purge Principal Gamma I X 10-4 and vents 2. Same as sample EmittersB

1. Prior to Each Purges frequency for vent

2. Every two weeks for samples.

Ventingi Monthly H-3 1 x 10-6 (0 2.Spent Fuel Pool Continuous Particulate for Weekly Gamma emitters NA Open Containment Equip- for 1/2 hr count ment Hatch during Outage (I- 131, others with half-life greater than 8 days)

Continuous Charcoal for Weekly 1-131 and 1-133 for NA Open Containment Equip- one hour count ment Hatch during Outage

& Aux Bldg Rollup DoorL Gaseous Grab at Contain- Daily Noble Gases - 1 x 10-4 ment Equipment Hatch & Gross Activity Aux Bldg Rollup DoorL C.Continuous Release 1.Vent Monthly - Gaseous Grab Monthly(t, K Principal Gamma I x 10-4 (RM8132B) SampleCK" EmittersB H-3(j I x 10-6 2.Millstone Continuous Charcoal Sam- Weekly 1-131 1 x 10-1, Stack pleDF. 1-133 1 x 10-1 0 (RM8169-1) Continuous Particulate Weekly Principal Gamma 1 x 10-11 SampleD',E EmittersB -

(I- 131, others with half lives greater than 8 days)

Continuous Particulate Quarterly Composite Sr-89, Sr I x 10-1l SampleD. Gross alpha I x 10-tO Continuous Noble Gas. Continuous Monitor Noble Gases - I x 10-6 Gross Activity TABLE I.D.-2 TABLE NOTATIONS A. The lower limit of detection (LLD) is defined in Table Notations, Item a, of Tables I.C.-1, I.C.-2, or I.C.-3.

MP-22-REC-BAPO1

'STOP' THINK' ACT REVIEW Rev. 026-01 25 of 166

B. For gaseous samples, the LLD will be 1 x 10-4 [tCi/cc and for particulate samples, the LLD will be 1 x 10-11 [tCi/cc. The principal gamma emitters for which these LLDs 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, 1-131, 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 above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in a priori LLDs higher than required, the reasons shall be documented in the "Radioactive Effluent Release Report."

C. Sampling and analysis shall also be performed 24 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after:

(1) reactor shutdown or startup, or (2) reactor power change greater than 15% of maximum power within a one hour period. If power change is part of a series of step changes, the sample may be collected 24 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after last power change step.

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

E. RESERVED F. Samples shall be changed at least once per seven days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing.

For Unit 2 vent only Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup, or thermal power change exceeding 15% of rated thermal power within a 1 -hour period and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of

10. This requirement does not apply if: (1) analysis shows that the Dose Equivalent 1- 131 concentration in the reactor coolant has not increased more than a factor of three; and '(2) the noble gas monitor shows that effluent activity has not increased more than a factor of three.

G. IF the refueling cavity is flooded and there is fuel in the cavity, THEN grab samples for tritium shall be taken weekly. The grab sample shall be taken from the Millstone Stack or vent where the containment ventilation is being discharged at the time of sampling.

H. Waste Gas Storage Tanks are normally released on a batch basis via the Millstone Stack.

However, for the purpose of tank maintenance, inspection, or reduction of oxygen concentration, a waste gas tank may be vented or purged with nitrogen and released to the environment via the normal or alternate pathway using one of the following methods:

Method A: Without a permit provided the following conditions are met:

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

(2) No radioactive gases have been added to the gaseous processing system since the previous discharge.

MP-22-REC-BAPO1 STQP. THINK ACT: REVIEW Rev. 026-01 26 of 166

(3) Valve lineups are verified to ensure that no radioactive waste gases will be added to the tank.

(4) Prior to initiation of the vent or purge, a sample of the gas in the tank will be taken and analyzed for any residual gamma emitters and tritium. The tank may be released if:

a) Tank activity is less than 1% of the activity released in the previous batch release from the tank, or less than 1% of the activity released to date for the calendar year, and b) the activity of Kr-85 and Xe-133 is less than 0.01 Ci and the activity of all other gases is less than 0.001 Ci.

Method B: With a permit provided valve lineups are verified to ensure that no radioactive waste gases will be added to the tank.

I. IE compared to the radioactivity at the time of the air sample, a Radiation Monitor RM8123 or RM8262 gas channel or a particulate channel increases by a factor of two, THEN a new containment air sample shall be taken.

IF containment noble gas activity exceeds 1 E-6 [tCi/cc as indicated by the last grab sample, THEN sampling frequency shall be increased to weekly until such.time that the activity is less than 1 E-6 [tCi/cc.

J. During an outage a sample is only required prior to the initial purge.

K. IF there is an increase of the Millstone Stack or Unit 2 Vent noble gas monitor of greater than 50%, THEN sampling and analysis shall also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except for the following conditions:

(1) the increase is already accounted for, or (2) the monitor has returned to within 20% of the average reading prior to the.increase.

IF the Millstone Stack or Unit 2 Vent noble gas monitor increased greater than 50% for more than one hour and has decreased prior to collecting a sample representative of the elevated reading, THEN an estimate of radioactivity released during the period of elevated reading shall be made.

L. Continuous charcoal sample at Aux Bldg Rollup Door and daily gas sample at Equipment Hatch and Aux Bldg Rollup Door are only required when moving fuel. Sampling at the Equipment Hatch is not required if the Enclosure Building is intact. Sampling at the Aux Bldg Rollup Door is not required if the rollup door is closed.

MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 27 of 166

Table I.D.-3 Millstone Unit 3 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Sample Type and Minimum Type of Activity Lower Limit Release Point Frequency Analysis Analysis of Detection or Source Frequency J (LLD)A- (LsCi/ml)

A. Containment and Fuel Building Release 1.Containment Gaseous, particulate and Prior to each purge or Principal gamma 1 X10-4 charcoal grab prior to drawdown. Same as emittersB.

each drawdown (via air sample frequency for ejector) releases to maintain sub- atmospheric.

2.Fuel Building Gaseous grab prior to Prior To Each Draw- 1-131 1'X 10- 1 each purgeH. down I- 133 1 x 10-10 Gaseous Grab every two Principal gamma I x 10-t1 weeks for releases to emittersB. - (I-131, oth-maintain ers with half lives greater sub-atmospheric pres- than 8 days) sure (via containment vacuum pump)l' Monthly for purge, H-3 i x 10-6 vents, and drawdowns Continuous particulate Weekly Gamma emitters for 1/2 NA at open containment hour count (1-131, others equipment hatch with half-life greater than 8 days)

Continuous charcoal at Weekly 1- 131 and 1- 133 for one NA containment equipment hour count hatch and fuel building rollup doorst" Gaseous grab at Daily Noble Gases - Gross 1 x 10-4 containment equipment Activity hatch & fuel building rollup doorsK.

B.Continuous Release 1.Unit 3 Ventila- Monthly - Gaseous Monthlyc.,J. Principal gamma 1 X 10-4 tion Vent Grab SampleC.J. emittersB (HfVR-RE10B) H-3(i I x 10-2.Engineered Continuous charcoal Weekly 1-131 i x 10-12 0

Safeguards sampleD.,F. 1-133 1 x 10- 1 0 Building Continuous particulate Weekly Principal gamma 1 x 10-11 (HVQ- RE49) sampleD.,F emittersB. - (1-131, others 3.Millstone with half lives greater than 8 Stack via days)

SLCRS Continuous particulate Quarterly composite Sr-89, Sr-90 1 x 10-11 (HVR- sampleD. Gross alpha 1 x 10-11 RE19B) Continuous noble gasD- Continuous Noble gases - gross 1 x 10-7 1monitor activity I TABLE I.D.-3 TABLE NOTATIONS MP-22-REC-BAPO1 STOP THINK ACT- REVIEW Rev. 026-01 28 of 166

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

B. For gaseous samples, the LLD will be 1 x 10-4 [iCi/cc and for particulate samples, the LLD will be 1 x 10-11 lLCi/cc. The principal gamma emitters for which these LLDs 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, 1-131, 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 above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in a priori LLDs higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.

C. For the ventilation vent and SLCRS, sampling and analysis shall also be performed 24 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after:

(1) reactor shutdown or startup, or (2) reactor power change greater than 15% of maximum power within a one hour period. If power change is part of a series of step changes, the sample may be collected 24 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after the last power change step.

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

E. RESERVED F. Samples shall be changed at least once per seven days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing.

For Unit 3 Vent only:

Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup, or thermal power change exceeding 15% of rated thermal power within a 1 -hour period and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of

10. This requirement does not apply if: (1) analysis shows that the Dose Equivalent 1- 131 concentration in the reactor coolant has not increased more than a factor of three; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of three.

G. IF the refueling cavity is flooded and there is fuel in the cavity, THEN grab samples for tritium shall be taken weekly from the ventilation vent.

H. During an outage a sample is only required prior to the initial purge.

1. IF compared to the radioactivity at the time of the air sample, Radiation Monitor CMS22 gas channel or particulate channel increases by a factor of two, THEN a new containment air sample shall be taken.

IF. containment noble gas activity exceeds 1 E-6 [Ci/cc as indicated by the last grab sample, THEN sampling frequency shall be increased to weekly until such time that the activity is less than 1E-6 ICi/cc.

MP-22-REC-BAP01 STOP THINK ACT RREVIEW Rev. 026-01 29 of 166

J. IF there is an unexplained increase of the Unit 3 ventilation vent or SLCRS noble gas monitor of greater than 50%, THEN appropriate sampling and analysis shall also be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except for the following conditions:

(1) the increase is already accounted for, or (2) the monitor has returned to within 20% of the reading prior to the increase.

IF the SLCRS or Unit 3 Vent noble gas monitor increased greater than 50% for more than one hour and has decreased prior to collecting a sample representative of the elevated reading, THEN an estimate of radioactivity released during the period of elevated reading shall be made.

K. Continuous charcoal sample at any open Fuel Bldg Rollup Door and daily gas sample at Equipment Hatch and at any open Fuel Bldg Rollup Door are only required when moving fuel.

Sampling at a Fuel Bldg Rollup Door is not required if the rollup door is closed.

,.MP-22-REC- BAP01 STOP THINK ACT: REVIEW Rev. 026-01 30 of 166

2. Gaseous Radioactive Waste Treatment

a. Dose Criteria for Equipment Operability Applicable to All Millstone Units The following dose criteria shall be applied separately to each Millstone unit.

1) IF any of the radioactive waste processing equipment listed in Section 2.b. are not routinely operating or are being bypassed, THEN doses due to gaseous effluents from the untreated waste stream to unrestricted areas shall be projected at least once per 31 days in accordance with the methodology and parameters in Section II.D.4. For each waste stream, only those doses specified in Section II.D.4. need to be determined for compliance with this section.

2) IF any of these dose projections exceed 0.02 mrad for gamma radiation, 0.04 mrad for beta radiation or 0.03 mrem to any organ due to gaseous effluents, THEN best efforts shall be made to return the processing equipment to service.

3) IF actual doses exceed 0.2 mrad for gamma radiation, 0.4 mrad for beta radiation or 0.3 mrem to any organ AND the dose from a waste stream with equipment not operating exceed 10% any of these limits, THEN prepare and submit to the Commission a report as specified in Section I.D.2.c.

b. Required Equipment for Each Millstone Unit Best efforts shall be made to return the gaseous radioactive waste treatment system equipment specified below for each unit to service if the projected doses exceed any of doses specified above. For the Unit 2 gas decay tanks, the tanks shall be operated to allow enough decay time of radioactive gases to ensure that the Radiological Effluent Control dose limits are not exceeded.

MP-22-REC-BAP01

'STOP THINK ACT REVIEW Rev. 026-01 31 of 166

1. Millstone Unit No. 1 Waste Stream Processing Equipment None Specified None required

2. Millstone Unit No. 2 Waste Stream Processing Equipment Gaseous Radwaste Five (5) gas decay tanks Treatment System One waste gas compressor Ventilation Exhaust Auxiliary building ventilation HEPA filter (L26 or L27)

Treatment System Containment purge HEPA filter (L25)

Containment vent HEPA/charcoal filter (L29 A or B)

3. Millstone Unit No. 3 Waste Stream Processing Equipment or Radioactivity Concentration Gaseous Radwaste Charcoal bed adsorbers Treatment System One HEPA filter

c. Report Requirement For All Three Millstone Units If required by Section I.D.2.a.3), prepare and submit to the Commission a Special Report within 30 days with the following content:

Explanation of why gaseous radwaste was being discharged without treatment, identification of any equipment out of service, and the reason for being out of service,

• Action(s) taken to restore the inoperable equipment to service, and

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

MP-22-REC-BAP01 STOP` THINK ACT REVIEW Rev. 026-01 32 of 166

3. Basis for Gaseous Sampling, Analysis, and Radioactive Treatment System Use Paragraph (a)(2) of Part 50.36a provides that licensee will submit an annual report to the Commission which specifies the quantity of each of the principal radionuclides released to unrestricted areas in gaseous effluents during the past 12 months of plant operation. The indicated gaseous surveillance programs (as directed by surveillance requirements for Radiological Effluent Controls in Sections III.D.2.a., IV.D.2.a. and V.D.2.a.)

provides the means to quantify and report on radioactive materials released to the atmosphere. As specified in Regulatory Guide 1.21, this program monitors all major and potentially significant paths for release of radioactive material in gaseous effluents during normal reactor operations, including anticipated operational occurrences. There are many minor release pathways which are not routinely monitored. The Millstone Effluent Control Program includes, as needed, evaluations to determine if any release point should be added to the REMODCM surveillance program. This information also provides for the assessment of effluent dose rates and environmental dose impacts for the purpose of demonstration compliance with the effluent limits of 10 CFR 20, and dose objectives of 10 CFR 50, Appendix I. The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of lower limits of detection (LLDs) and are selected, based on NUREG-1301, such that the detection of radioactivity in releases will occur at levels below which effluent offsite dose objectives would be exceeded. The indicated gaseous radwaste treatment equipment for each Unit have been determined, using the GALE code, to be capable to minimize radioactive gaseous effluents such that the dose objectives of Appendix I can be met for expected routine (and anticipated operational occurrence) effluent releases. This equipment is maintained and routinely operated to treat appropriate gaseous waste streams without regards to projected environmental doses.

If not already in use, the requirement that the appropriate portions of the gaseous radioactive waste treatment system for each Unit be returned to service when the specified effluent doses are exceeded provides assurance that the release of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This condition of equipment usage implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50, and the design objective given in Section II.D.

of Appendix I to 10 CFR 50. The specified dose limits governing the required use of appropriate portions of the gaseous radwaste treatment system were selected as a suitable fraction of the dose design objectives set forth in Section II.A. of Appendix I, 10 CFR 50 for gaseous effluents following the .guidance in NUREG- 1301.

MP REC-BAP01 SToP 'TINK 'ACT REVIEW Rev. 026-01 33 of 166

Figure I.D.-1, "Simplified Gaseous Effluent Flow Diagram Millstone Unit One" 4 ..................... = generalized air flow-Reactor Balance of.

Plant Vent Building S

Turbine Building Radwaste Building MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 34 of 166

Figure I.D.-2, "Simplified Gaseous Effluent Flow Diagram Millstone Unit Two" Ma in w Air,,EjectorUnit I 2 Condensr -d- -onclosure Building Roaf Vea Waste Gas Decay Processing Waste Gas Tanks Systeml Conpressos -0 I// (T1.A17t WRGM RM-8 169 Enclosure Building Containment Purge (L75):

Steam Generator Fuel Blowdown QuenclRuilding Tank Vent (Aux HEPA _________

Building (1.27)!,*

38g6)

Steam Generator Blowdowti Tank Vent 30Enclosure Bldg Roof Vent Aux Bldg EANf Vn (1-26) Rad Monitor Turbine Bldg

• Turbine Bldg Roof Vent

• These flow paths used during an accident.

MP-22-REC-BAPO1 STOP THINK .ACT REVIEW Rev. 026-01 35 of 166

Figure I.D.-3, "Simplified Gaseous Effluent Flow Diagram Millstone Unit Three"

• These flow paths used during an accident.

MP-22-REC-BAP01 STOP 'THINK ACT, REVIEW Rev. 026-01 36 of 166

I.E. Radiological Environmental Monitoring

1. Sampling and Analysis The radiological sampling and analyses provide measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from plant operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

Program changes may be made based on operational experience.

The sampling and analyses shall be conducted as specified in Table I.E. -1 for the locations shown Table I.E.-2. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment or other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period.

All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Section I.E1. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice (excluding milk) at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathways in questions and appropriate substitutions made within 30 days in the radiological environmental monitoring program.

If milk samples are temporarily unavailable from any one or more of the milk sample locations required by Table I.E.-2, a grass sample shall be substituted during the growing season (Apr. - Dec.) and analyzed for gamma isotopes and 1-131 until milk is again available. Upon notification that milk samples will be unavailable for a prolonged period (>9 months) from any one or more of the milk sample locations required by Table I.E. -2, a suitable replacement milk location shall be evaluated and appropriate changes made in the radiological environmental monitoring program.

Reasonable attempts shall be made to sample the replacement milk location prior to the end of the next sampling period. Any of the above occurrences shall be documented in the Annual Radiological Environmental Operating Report, which is submitted to the U. S. Nuclear Regulatory Commission prior to May 1 of each year.

Changes to sampling locations shall be identified in a revised Table I.E. -2 and, as necessary, Figure(s) I.E. -1 through I.E. -3.

MP-22-REC-BAP01 STOP THINK A ACT 'REVIEW Rev. 026-01 37 of 166

If the level of radioactivity in an environmental sampling medium at one or more of the locations specified in Table I.E.-2 exceeds the report levels of Table I.E.-3 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from receipt of sample results, a Special Report which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table I.E.-3 to be exceeded. When more than one of the radionuclides in Table I.E.-3 are detected in the sampling medium, this report shall be submitted if:

concentration (1)

+ concentration (2) + 1.0 reporting level (1) reporting level (2)

When radionuclides other thanthose in Table I.E.-3 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the appropriate calendar year limit of the Radiological Effluent Controls (Sections III.D.1.b.,

III.D.2.b., or III.D.2.c. for Unit 1; Sections IVD.1.b., IV.D.2.b., or IVD.2.c.

for Unit 2; and Sections VD.L.b., V.D.2.b., or V.D.2.c. for Unit 3). This report is not required if the measured level of radioactivity was not the result of plant effluents, however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

The detection capabilities required by Table I.E.-4 are state-of-the-art for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement. All analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

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Table I.E.-1 Millstone Radiological Environmental Monitoring Program Exposure Pathway and/or No. of Sampling and Collection Type and Frequency of Sample Locations Frequency Analysis 1.Gamma Dose - 40 (a) Quarterly Gamma Dose - Quarterly Environmental TLD 2.Airborne Particulate 8 Continuous sampler - Gross Beta - Weekly Gamma Spec-weekly filter change trum - Quarterly on composite (by location), and on individual sampleif gross beta is greater than 10x the mean of the weekly control station's gross beta results 3.Airborne Iodine 8 Continuous sampler - 1- 131 - Weekly weekly canister change __

4.Vegetation 5 One sample near middle and Gamma Isotopic on each sample one near end of growing season 5.Milk 2 Semimonthly when animals Gamma Isotopic and 1-131 on each are on pasture; monthly at sample; Sr-89 and Sr-90 on Quarterly other times Composite 0 5.a.Pasture Grass 2 Sample as necessary to sub- Gamma Isotopic and 1-131 stitute for unavailable milk' 6.Sea Water 2 Continuous sampler with a Gamma Isotopic & Tritium on each monthly collection at indica- sample tor location. Quarterly at control location - Compos-ite of 6 weekly grab samples 7.Well Water 6 Semiannual Gamma Isotopic & Tritium on each sample 8.Bottom Sediment 5 Semiannual Gamma Isotopic on each sample 9.Soil 3 Annually Gamma Isotopic on each sample 10.Fin Fish-Flounder and 2 Quarterly Gamma Isotopic on each sample one other type of edible fin fish (edible portion)

I t.Mussels (edible portion) 2 Quarterly Gamma Isotopic on each sample 12.Oysters (edible portion) 4 Quarterly Gamma Isotopic on each sample 13.Clams (edible portion) 2 Quarterly Gamma Isotopic on each sample 14.Lobsters (edible portion) 2 Quarterly Gamma Isotopic on each sample (a) Two or more TLDs or TLD with two or more elements per location.

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Table I.E.-2 Environmental Monitoring Program Sampling Locations The following lists the environmental sampling locations and the types of samples obtained at each location. Sampling locations are also shown on Figures I.E.-1 and I.E.-2:

Location Direction & Dis- Sample Types tance from Re-No* Name lease Point*"

1-I Onsite - Old Millstone Road 0.6 Mi, NNW TLD, Air Particulate, Iodine, Vegetation 2-1 Onsite - Weather Shack 0.3 Mi, S TLD, Air Particulate, Iodine 3-I Onsite - Bird Sanctuary 0.3 Mi, NE TLD, Air Particulate, Iodine, Soil 4-I Onsite - Albacore Drive 1.0 Mi, N TLD, Air Particulate, Iodine, Soil 5-1 Onsite - MP3 Discharge 0.1 Mi, SSE TLD 6-I Onsite - Quarry Discharge 0.3 Mi, SSE TLD 7-I Onsite - Environmental Lab Dock 0.3 Mi, SE TLD 8-I Onsite - Environmental Lab 0.3 Mi, SE TLD 9-I Onsite - Bay Point Beach 0.4 Mi, W TLD 10-I Pleasure Beach 1.2 Mi, E TLD, Air Particulate, Iodine, Vegetation 11-I New London Country Club 1.6 Mi, ENE TLD, Air Particulate, Iodine.

12-C Fisher's Island, NY 8.0 Mi, ESE TLD 13-C Mystic, CT 11.5 Mi, ENE TLD 14-C Ledyard, CT 12.0 Mi, NE TLD, Soil 15-C Norwich, CT 14.0 Mi, N TLD, Air Particulate, Iodine 16-C Old Lyme, CT 8.8 Mi, W TLD 17-1 Site Boundary 0:5 Mi, NE Vegetation 21-I Goat Location No. 1 2.0 Mi., N Milk 24-C Goat Location No. 3 29 Mi, NNW Milk 25-1 Fruits & Vegetables Within 10 Miles Vegetation 26-C Fruits & Vegetables Beyond 10 Mi Vegetation 27-I Niantic 1.7 Mi, WNW TLD, Air Particulate, Iodine 28-I Two Tree Island 0.8 Mi, SSE Mussels, Fish' 29-I West Jordan Cove 0.4 Mi, NNE Clams, Fish1 30-I Niantic Shoals 1.5 Mi, NNW Mussels 31-I Niantic Shoals 1.8 Mi, NW Bottom Sediment, Oysters 32-I Vicinity of Discharge 2 Bottom Sediment, Oysters, Lobster, Fish',

Seawater 33-I Seaside Point 1.8 Mi, ESE Bottom Sediment 34-I Thames River Yacht Club 4.0 Mi, ENE Bottom Sediment 35-I Niantic Bay 0.3 Mi, WNW Lobster, Fish 37-C Giant's Neck 3.5 Mi, WSW Bottom Sediment, Oysters, Seawater 38-I Waterford Shellfish Bed No. 1 1.0 Mi, NW Clams 41-I Myrock Avenue 3.2 Mi, ENE TLD 42-I Billow Road 2.4 Mi, WSW TLD MP-22-REC-BAP01 STO P THINK ACT RREVIEW Rev. 026-01 40 of 166

Table I.E.-2, Cont.

Location Direction & Dis- Sample Types tance from Re-No* Name lease Point**

43-I Black Point 2.6 Mi, SW TLD 44-1 Onsite - Schoolhouse 0.1 Mi, NNE TLD 45-I Onsite Access Road 0.5 Mi, NNW TLD 46-I Old Lyme - Hillcrest Ave. 4.6 Mi, WSW TLD 47-I East Lyme - W. Main St. 4.5 Mi, W TLD 48-I East Lyme - Corey Rd. 3.4 Mi, WNW TLD 49-I East Lyme - Society Rd. 3.6 Mi, NW TLD 50-I East Lyme - Manwaring Rd. 2.1 Mi, W TLD 51-I East Lyme - Smith Ave. 1.5 Mi, NW TLD 52-1 Waterford - River Rd. 1.1 Mi, NNW TLD 53-I Waterford - Gardiners Wood Rd. 1.4 Mi, NNE TLD 55-I Waterford - Magonk Point 1.8 Mi, ESE TLD 56-I New London - Mott Ave. 3.7 Mi, E TLD 57-1 New London - Ocean Ave. 3.6 Mi, ENE TLD 59-I Waterford -Miner Ave. 3.4 Mi, NNE TLD 60-I Waterford - Parkway South 4.0 Mi, N TLD 61-I Waterford - Boston Post Rd. 4.3 Mi, NNW TLD 62-I East Lyme - Columbus Ave. 1.9 Mi, WNW TLD 63-I Waterford - Jordon Cove Rd. 0.8 Mi, NE TLD 64-I Waterford - Shore Rd. 1.1 Mi, ENE TLD 65-I Waterford - Bank St. 3.2 Mi, NE TLD 71-I Onsite well Onsite Well water 72-I Onsite well Onsite Well water 79-I Onsite well Onsite Well water 80-I Onsite well Onsite Well water 81-1 Onsite well Onsite Well water 82-I Onsite well Onsite Well water 88-1 DEP dock near barge slip 0.2 Mi, WNW Oysters 0 1Fish to be sampled from one of three locations -28, 29, or 32.

2Vicinity of discharge includes the Quarry and shoreline area from Fox Island to western point of Red Barn Recreation Area and offshore out to 500 feet.

• I = Indicator; C = Control.

• • = The release points are the Millstone Stack for terrestrial locations and the end of the quarry for aquatic location.

NOTE: Environmental TLDs also function as accident TLDs in support of the Millstone Emergency Plan.

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Figure I.E.- 1, "It iner Air Particulate And Vegetation Monitoring Stations" L A-.R gig Z Is 4npr.nip;.~

,TN4,N6Sk *A 0"

~ 4Y~7.1i Li -POR. -IN lrt IL A .k '

~

. -Vi ~ *'9A 0, UN'~ -.

'.ff4Al ggg .

,~bl~~AFam MPR2 S2, RECBA THINK STOP ACT REVIE Re5260

.42 ofp16

Figure I.E.-2, "Outer Terrestrial Monitoring Stations" MP-22-REC-BAP01 STOP 'THINK ACT REVIEW Rev 026-01 43 of 166

Table I.E.-3 Reporting Levels For Radioactivity Concentrations In Environmental Samples Analysis Water Airborne- Fish ShellfishC. Milk Vegetables (pCi/I) Particulate (pCi/g, wet) (pCi/g, wet) (pCi/I) (pCi/g, wet) or Gases (pCi/m 3)

H-3 20,000A-Mn-54 1,000 30 140 Fe-59 400 10 60 Co-58 1,000 30 130 Co-60 300 10 50 Zn-65 300 20 80 Zr-95 400 Nb-95 400 Ag-110m 8 30.

1-131 2 0 B_ 0.9 0.2 1 3 0.1 Cs- 134 30 10 1 5 60 1 Cs-137 50 20 2 8 70 2 Ba-140 200 300 La- 140 200 300 A. 20,000 pCi/I for drinking water samples. (This is 40 CFR Part 141 value.) For non-drinking water pathways, a value of 30,000 pCi/I may be used. I B. Reporting level for 1-131 applies to non-drinking water pathways (i.e., seawater). If drinking water pathways are sampled, a value of 2 pCi/I is used.

C. For on-site samples, these values can be multiplied by 3 to account for the near field dilution factor MP-22-REC-BAP01 STOP THINK ACT, REVIEW Rev 026-01 44 of 166

Table I.E.-4 Maximum Values For Lower Limits Of Detection (LLD)A-Analysis Water Airborne- Fish Milk Food Sediment (pCi/l) Particulate Shellfish (pCi/I) Products (pCi/g, dry) or Gases (pCi/g, wet) (pCi/g, wet)

(pCi/m 3 )

gross beta 1 x 10- 2 H-3 20 0 0 D.

Mn-54 15 0.130 Fe-59 30 0.260 Co-58,60 15 0.130 Zn-65 30 0.260 Zr-95 30 Nb-95 15 2 0.06B.

1-131 15C' 7 x 10- 1 Cs-134 15 5 x 10-2 0.130 15 0.060 0.150 Cs-137 18 6 x 10- 2 0.150 18 0.080 0.180 Ba- 140 6 0C- 70 La-140 1 5 c- 25 TABLE NOTATIONS Table I.E.-4 A. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

LLD = (E)( V)(4.66 Sb 2.22 )( Ye-At)

Where:

• LLD is the lower limit of detection as defined above (as l.Ci per unit mass or volume)

• Sb is the standard deviation of the background counting rate or of the counting rate of a blank 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 is the number of transformations per minute per pCi

• Y is the fractional radiochemical yield (when applicable)

MP-22-REC-BAP01 ISTOP THINK -ACT REVIEW Rev. 026-01 45 of 166

0 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 (or end of the sample collection period) and time of counting.

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

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified in the Annual Radiological Environmental Operating Report.

B. LLD for leafy vegetables.

C. From end of sample period.

D. If no drinking water pathway exists (i.e., seawater), a value of 3,000 pCi/I may be used.

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2. Land Use Census The land use census ensures that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. This census satisfies the requirements of Section IVB.3 of Appendix I to 10 CFR 50. The land use census shall be maintained and shall identify the location of the nearest resident, nearest garden*, and milk animals in each of the 16 meteorological sectors within a distance of five miles.

The validity of the land use census shall be verified within the last half of every year by either a door-to-door survey, aerial survey, consulting local agriculture authorities, or any combination of these methods.

With a land use census identifying a location(s) which yields a calculated dose or dose commitment greater than the doses currently being calculated in the off-site dose models, make the appropriate changes in the sample locations used.

With a land use census identifying a location(s) which has a higher D/Q than a current indicator location the following shall apply:

1) If the D/Q is at least 20% greater than the previously highest D/Q, replace one of the present sample locations with the new one within 30 days if milk is available.

2) If the D/Q is not 20% greater than the previously highest D/Q, consider direction, distance, availability of milk, and D/Q in deciding whether to replace one of the existing sample locations. If applicable, replacement shall be within 30 days. If no replacement is made, sufficient justification shall be given in the annual report.

Sample location changes shall be noted in the Annual Radiological Environmental Operating Report.

• Broad leaf vegetation (a composite of at least 3 different kinds of vegetation) may be sampled at the site boundary in each of 2 different direction sectors with high D/Qs in lieu of a garden census.

3. Interlaboratory Comparison Program The Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

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Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.

With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.

4. Bases for the Radiological Environmental Monitoring Program Federal regulations (10 CFR Parts 20 and 50) require that radiological environmental monitoring programs be established to provide data on measurable levels of radiation and radioactive materials in the site environs.

In addition, Appendix I to 10 CFR 50 requires that the relationship between quantities of radioactive material released in effluents during normal operation, including anticipated operational occurrences, and the resultant radiation doses to individuals from principal pathways of exposure be evaluated. The Millstone Environmental Radiological Monitoring Program (REMP) has been established to verify the effectiveness of in-plant measures used for controlling the release of radioactive materials from the plant, as well as provide for the comparison of measurable concentrations of radioactive materials found in the environment with expected levels based on effluent measurements and the modeling of the environmental exposure pathways.

The REMP detailed in Table I.E.- 1 provides measurements of radioactive materials or exposures in the environment along all principal exposure pathways to man that could be impacted by plant effluents. These include direct radiation exposure, inhalation exposure, and ingestion of food products (both aquatic and land grown). In addition, intermediate media such as vegetation and bottom sediments are included as potential early indicators of radioactive material buildup. The selections of sample locations include areas subject to plant effluents that would be expected to exhibit early indication of any buildup of plant related radioactive materials.

The required detection capabilities for environmental sample analyses are tabulated in terms of lower limits of detection (LLDs). Except for Ba- 140 and La- 140 in milk, the required LLDs are from NUREGs- 1301 and 1302.

The NUREGs specify an LLD of 15 pCi/l for the parent-daughter combination of Ba-La- 140. An LLD of 25 pCi/l is specified for the daughter La-140 and 70 pCi/l for the parent Ba- 140.

- MP-22-REC-BAP01 STOP0:. TH'i"NKk .....ýACTf" REVIEWRev.026-01 48 of 166

Annual reports of environmental radiation monitoring summaries are filed with the NRC in accordance with the requirements of 10 CFR 50.36b and the guidance contained in Regulatory Guide 4.8, Environmental Technical Specifications for Nuclear Power Plant," and NUREG-0472 (NUREG-0473) Revision 3, "Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors (Boiling Water Reactors)."

I.E Report Content

1. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The report shall also include the results of the land use census required by Section I.E.2. of this manual. If levels of radioactivity are detected that result in calculated doses greater than IOCFR50 Appendix I Guidelines, the report shall provide an analysis of the cause and a planned course of action to alleviate the cause.

The report shall include a summary table of all radiological environmental samples which shall include the following information for each pathway sampled 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 the location with the highest annual mean (indicator or control).

7) Number of non-routine reported measurements as defined in these specifications.

In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in the next annual report.

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This report shall include a comparison of dose assessments of the measured environmental results of the calculated effluent results to confirm the relative accuracy or conservatism of effluent monitoring dose calculations.

The report shall also include a map of sampling locations keyed to a table giving distances and directions from the discharge; the report shall also include a summary of the Interlaboratory Comparison Data required by Section I.E.3. of this manual.

2. Radioactive Effluent Release Report The Radioactive Effluent Release Report (RERR) shall include quarterly quantities of and an annual summary of radioactive liquid and gaseous effluents released from the unit in the Regulatory Guide 1.21 (Rev. 1, June 1974) format. Radiation dose assessments for these effluents shall be provided in accordance with 10 CFR 50.36a and the Radiological Effluent Controls. An annual assessment of the radiation doses from the site to the most likely exposed REAL MEMBER OF THE PUBLIC shall be included to demonstrate conformance with 40 CFR 190. Gaseous pathway doses shall use meteorological conditions concurrent with the quarter of radioactive gaseous effluent releases. Doses shall be calculated in accordance with the Offsite Dose Calculation Manual. The licensee shall maintain an annual summary of the hourly meteorological data (i.e., wind speed, wind direction and atmospheric stability) either in the form of an hour-by-hour listing on a magnetic medium or in the form of a joint frequency distribution. The licensee has the option of submitting this annual meteorological summary with the RERR or retaining it and providing it to the NRC upon request.

The RERR shall be submitted prior to May 1 of each year for the period covering the previous calendar year.

The RERR shall include a summary of each type of solid radioactive waste shipped offsite for burial or final disposal during the report period and shall include the following information for each type:

" type of waste (e.g., spent resin, compacted dry waste, irradiated components, etc.)

• solidification agent (e.g., cement)

• total curies

• total volume and typical container volumes

" principal radionuclides (those greater than 10% of total activity)

• types of containers used (e.g., LSA, Type A, etc.)

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The RERR shall include a list of all abnormal releases of radioactive gaseous and liquid effluents (i.e., all unplanned or uncontrolled radioactivity releases, including reportable quantities) from the site to unrestricted areas.

Refer To MP-22-REC-REF03, "REMODCM Technical Information Document (TID)," for guidance on classifying releases as normal or abnormal. The following information shall be included for each abnormal release:

• total number of and curie content of releases (liquid and gas)

• a description of the event and equipment involved

• cause(s) for the abnormal release

• actions taken to prevent recurrence

" consequences of the abnormal release Changes to the MP-22-REC-BAPO1, "Radiological Effluent Monitoring And Offsite Dose Calculation Manual (REMODCM)," shall be submitted to the NRC as appropriate, as a part of or concurrent with the RERR for the period in which the changes were made.

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SECTION II.

Offsite Dose Calculation Manual (ODCM)

For the Millstone Nuclear Power Station Nos. 1,:2, & 3 Docket Nos. 50-245, 50-336, 50-423 MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 52 of 166

SECTION II. OFF-SITE DOSE CALCULATION MANUAL (ODCM)

II.A. Introduction The purpose of the Off-Site Dose Calculation Manual (Section II of the REMODCM) is to provide the parameters and methods to be used in calculating offsite doses and effluent monitor setpoints at the Millstone Nuclear Power Station. Included are methods for determining maximum individual whole body and organ doses due to liquid and gaseous effluents to assure compliance with the regulatory dose limitations in 10 ýCFR 50, Appendix I. Also included are methods for performing dose projections to assure compliance with the liquid and gaseous treatment system operability sections of the Radiological Effluent Monitoring Manual (REMM - Section I of the REMODCM). The manual also includes the methods used for determining quarterly and annual doses for inclusion inthe Radioactive Effluent Release Report.

The bases for selected site-specific factors used in the dose calculation methodology are provided in MP-22-REC-REF03, "REMODCM Technical Information."

Another section of this manual discusses the methods to be used in determining effluent monitor alarm/trip setpoints to be used to ensure compliance with the instantaneous release rate limits in Sections III.D.2.a., IVD.2.a., and VD.2.a.

This manual includes the methods to be used in performance of the surveillance requirements in the Radiological Effluent Controls of Sections III, IV, and V.

Appendix A, Tables App.A-1 provide -a cross-reference of effluent requirements and applicable methodologies contained in the REMODCM.

Most of the calculations in this manual have several methods given for the calculation of the same parameter. These methods are arranged in order of simplicity and conservatism, Methodý 1 being the easiest and most conservative.

As long as releases remain low, one should be able to use Method I as a simple estimate of the dose. If release calculations approach the limit, however, more detailed yet less conservative calculations may be used. At any time a more detailed calculation may be used in lieu of a simple calculation.

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

II.B. Responsibilities All changesto the Off-Site Dose Calculation Manual (ODCM) shall be reviewed and approved by the Facilities Safety Review Committee prior to implementation.

*:' , ,.

• MP REC -BAP01 STOP. THINK ACT REVIEW Rev. 026-01 53 of 166

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

It shall be the responsibility of the Site Vice President Millstone to ensure that this manual is used as required by the administrative controls of the Technical Specifications. The delegation of implementation responsibilities is delineated in the MP-22-REC-PRG, "Radiological Effluent Control."

II.C. Liquid Dose Calculations The determination of potential doses from liquid effluents to the maximum exposed member of the public is divided into two methods. Method 1 is a simplified calculation approach that is used as an operational tool to ensure that effluent releases as they occur are not likely to cause quarterly and annual offsite dose limits to be exceeded. Effluent doses are calculated at least once every 31 days. Method 2 is a more detailed computational calculation using accepted computer models to demonstrate actual regulatory dose compliance. Method 2 is used whenever the Method 1 estimation begins to approach a regulatory limit, and for preparation of the Radioactive Effluent Release Report, which includes the quarterly and annual dose impacts for all effluents recorded discharged to the environment during the year of record.

1. Whole Body Dose from Liquid Effluents Radiological Effluent Controls (Sections III, IV, and V) limit the whole body dose to an individual member of'the public to 1.5 mrem per calendar quarter and 3 mrem per year from liquid effluents released from each unit. (See Appendix A, Table App.A-1 forl cross-reference effluent control requirements and applicable sections in the REMODCM which are used to determine compliance). In addition, installed portions of liquid radwaste treatment system are required to be operated to reduce radioactive materials in liquid effluents when' the projected whole body dose over 31 days from applicable waste streams exceeds 0.006 mrem. This part of the REMODCM provides the calculation methodology for determining the whole body dose from radioactive materials released into liquid pathways of exposure associated with routine discharges. This includes the liquid pathways which contribute to the 25 mrem annual total dose limit (40 CFR190) to any real individual member of the public from all effluent sources (liquids, gases, and direct).

a. Method 1 (Applicable to Units 1, 2, and 3)

For Unit 1: No Method 1, use Method 2 (Section II.C.1.b.)

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For Units 2 and 3:

Dw= 0.2 CF+ 5.6 x 10- 7 CH Where:

Dw= The estimated whole body dose to a potentially maximum exposed individual (in mrem) due to fission and activation products released in liquid effluents during a specified time period.

CF = total gross curies of fission and activation products, excluding tritium and dissolved noble gases, released during the period of interest.

CH - total curies of tritium released during the period of interest.

If Dw, within a calendar quarter is greater than 0.5 mrem, go to Method 2.

b. Method 2 (Applicable to Units 1, 2, and 3)

If the calculated dose using;Method 1 is greater than 0.5 mrem within a calendar quarter, or if a more accurate determination is desired, use the NRC computer code LADTAP II, or a code which uses the methodology given in Regulatory Guide 1.109, to calculate the liquid whole body doses. Method 2 (LADTAP II) is also used in the performance of dose calculations for the Radioactive Effluent Release Report. The use of this code is given in MP-22-REC-GDLO2, "Liquid Dose Calculations

- LADTAP-HI." Additional information on LADTAPII is contained in MP-22-REC-REF03, "REMODCM Technical Information Manual."

I

2. Maximum Organ Dose from Liquid Effluents Radiological Effluent Controls (Sections 1II, IV, and V) limit the maximum organ dose to an individual member of the public to 5 mrem per calendar quarter and 10 mrem per year from liquid effluents released from each unit.

(See Appendix A, Table App.A- 1 for cross-reference effluent control requirements and applicable sections in the REMODCM which are used to determine compliance). In addition, installed portions of liquid radwaste treatment system are required to be operated to reduce radioactive materials in liquid effluents when the projected maximum organ dose over 31 days from applicable waste streams exceeds 0.02 mrem. This part of the REMODCM provides the calculation methodology for determining the maximum organ dose from radioactive materials released into liquid pathways of exposure associated with routine discharges. This includes the

_.,- MP-22-REC-BAP01 STOP' THINK ACT REVIEW Rev. 026-01 55 of 166

liquid pathways which contribute to the 25 mrem annual organ (except 75 mrem thyroid) dose limit (40 CFR190) to any real individual member of the public from all effluent sources (liquids, gases, and direct).

a. Method 1 (Applicable to Units 1, 2, and 3)

For Unit 1: No Method 1, use Method 2 (Section II.C.2.b.)

For Units 2 and 3:

Do = 1.5 CF Where: Do =The estimated maximum organ dose to the potentially maximum exposed individual (in mrem) due to fission and activation products released in liquid effluents during a specified time period.

CF =total gross curies of fission and activation products, excluding tritium and dissolved noble gases, released during the period of interest - same asSection II.C.l.a.

If Do, within a calendar quarter is greater than 2 mrem, go to Method 2.

b. Method 2 (Applicable to Units 1, 2, and 3)

If the calculated dose using Method 1 is greater than 2 mrem, or if a more accurate determination is desired, use the NRC computer code LADTAP II, or a code which uses the methodology given in Regulatory Guide 1.109, to calculate the liquid maximum organ doses. Method 2 (LADTAP II) is also used in the performance of dose calculations for the Radioactive Effluent Release Report. The use of this code and the input parameters are given in MPý-22-REC-GDLO2, "Liquid Dose Calculations - LADTAP-II." Additional information on LADTAPII is contained in MP REC- REF03, "REMODCM Technical Information Manual."

3. Estimation of Annual Whole Body Dose (Applicable to All Units)

An estimation of annual (year-to-date) whole body dose (Dyw) from liquid effluents shall be made every month to determine compliance with the annual dose limits for each Unit which releases any radioactivity in liquid effluents. Annual doses will be determined as follows:

DYW= ZDw Where the sum of the doses include the whole body dose contribution from all effluent releases for each Unit recorded to-date. For estimation of the MP-22-REC-BAPO1 STOP THINK ACT REVIEW Rev. 026-01 56 of 166

Total Dose requirements of 40CFR190, the effluent releases from all three Units combined are used.

The following shall be used as Dw:

1) If the detailed quarterly dose calculations required per Section II.C.6.

for the Radioactive Effluent Release Report are completed for any calendar quarter, use that result.

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

3) If the annual dose estimate, Dyw, is greater than 3 mrem and any Dw determined as in Section IIC.1. was not calculated using Method 2 (i.e., LADTAP II computer code or a Regulatory Guide 1.109 code),

recalculate Dw using Method 2 if this could reduce Dyw to less than 3 mrem.

4. Estimation of Annual Maximum Organ Dose (Applicable to All Units)

An estimation of annual (year-to-date) maximum organ dose (Dyo) from liquid effluents shall be made every month to determine compliance with the annual dose limits for each Unit which releases any radioactivity in liquid effluents. Annual doses will be determined as follows:

Dyo= IDo Where the sum of the doses include the maximum organ dose contribution from all effluent releases for each Unit recorded to-date. For estimation of the Total Dose requirements of 40CFR190, the effluent releases from all three Units combined are used.

The following guidelines shall be used:

1) If the detailed quarterly dos'e calculations required per Section II.C.6.

for the Radioactive Effluent Release Report are completed for any calendar quarter, use that result.

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

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

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4) If Dyo is greater than 10 mrem and any value used in its determination was calculated as in Section, II.C.2., but not with Method 2 (i.e.,

LADTAP II computer code or a Regulatory Guide 1.109 code),

recalculate that value using Method 2 if this could reduce Dyo to less than 10 mrem.

5. Monthly Dose ProjectionsSection I.C.2.a. of the REMM requires that certain portions of the liquid radwaste treatment equipment be used to reduce radioactive liquid effluents when the projected doses for each Unit (made at least once per 31 days) exceeds 0.006 mrem whole body or 0.02 mrem to any organ. The following methods are applied in the estimation of monthly dose projections:

a. Whole Body and Maximum Organ (Applicable to Unit 1 Only)

In the dose code DOSLIQ use concentrations of radionuclides in reactor cavity water and estimates of projected volumes and discharge rates for the following 31 days to estimate dose from liquid discharge of reactor cavity water in the following 31 days.

I

b. Whole Body and Maximum Organ when Steam Generator Total Gamma Activity is less than 5E-7 [iCi/ml and Steam Generator Tritium is less than 0.02 [tCi/ml (Applicable to Units 2 and 3)

The projected monthly whole body dose (Units 2 or 3) is determined from:

DEMw = D'Mw [R1 R4 F 2]

The monthly projected maximum organ dose (Units 2 or 3) is determined from:

DEMo = D'Mo [R 1 R 4 F 2]

Where:

D'MW = the whole body' dose from the last typical previously completed month as calculated per the methods in Section II.C.1.

D'MO = the maximum organ dose from the last typical previously completed month as calculated per the methods in Section II.C.2.

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R, = the ratio of the total estimated volume of liquid batches to be released in the present month to the volume released in the past month.

R4 = the ratio of estimated primary coolant activity for the present month, to that for the past month.

F2 = the factor to be applied to the estimated ratio of final curies released if there are expected differences in treatment of liquid waste for the present month as opposed to the past month (e.g., bypass of filters or demineralizers).

NUREG-0017'or past experience shall be used to determine the effect of each form of treatment which will vary. F2 = 1 if there are no expected differences.

Notes:

1) The last month should be typical without significant operational differences from the projected month. If there were no releases during last month, do not use that month as the base month if it is estimated that thlere will be releases for the coming month.

2) If the last typical month's doses were calculated using LADTAP II (or similar methodology), also multiply the LADTAP (or similar methodology) doses'by R 5 where R 5 =

total dilution flow from LADTAP run divided by estimated total dilution flow.

c. Whole Body and Maximum Organ when Steam Generator Total Gamma Activity Exceeds 5E-7 [tCi/ml or Steam Generator Tritium Exceeds 0.02 tCi/ml (Applicable to Units 2 and 3)

The projected monthly whole body dose (Units 2 or 3) is-determined from:

DEMw = D'MW [(1 - FI) R 1 R 4 F 2 + F 1 R 2 R 3]

The monthly projected maximum organ dose (Units 2 or 3) is determined from:

DEMo = D'MO [(I - F1 ) R, R4 F2 + F 1 R2 R3]

STP HI REV :

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Where:

D'Mw=the whole body dose from the last typical previously completed month as calculated per the methods in Section II.C.1.

D'MO= the maximum organ dose from the last typical previously completed month as calculated per the methods in Section II.C.2.

R= the ratio of the total estimated volume of liquid batches to be released in the present month to the volume released in the past month.

R2= the ratio of estimated volume of steam generator blowdown to be released in present month to the volume released in the past month!

F1 = the fraction of curies released last month coming from steam generator blowdown calculated as:

I curies from blowdown curies from blowdown + curies from batch tanks R3= the ratio of estimated secondary coolant activity for the present month to that for the past month.

R4= the ratio of estimated primary coolant activity for the present month to that for the past month.

F2 = the factor to beý applied to the estimated ratio of final curies released if there are expected differences in treatment of liquid waste for the present month as opposed to the past month (e.g., bypass of filters or demineralizers).

NUREG-0017 or past experience shall be used to determine the effect of each form of treatment which will vary. F2 '= 1 if there are no expected differences.

6. Quarterly Dose Calculations for Radioactive Effluent Release Report Detailed quarterly dose calculations required for the Radioactive Effluent Release Report shall be done using the NRC computer code LADTAP II, or a code which uses the methodology given in Regulatory Guide 1.109. The use of LADTAP II code, and the input parameters are given in MP-22-REC-GDLO2, "Liquid Dose Calculations - LADTAP II." Use of a code using the methodology given in Regulatory Guide 1.109 is described in MP-22-REC-GDLO3, "Liquid Dose Calculations -DOSLIQ."

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Additional information on liquid dose calculations is contained in MP-22-REC-REF03, "REMODCM Technical Information Manual."

7. Bases for Liquid Pathway Dose Calculations The dose calculation methodology and parameters used in Section II of the REMODCM implement the requirements in Section III.A of Appendix I (10CFR50) which states that conformance with the dose objectives of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a member of the public through appropriate

• pathways is unlikely to be substantially underestimated. The dose estimations calculated by both Method 1 and Method 2 are based on the liquid models presented in Regulatory Guide 1.109, Rev.1; "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I". These equations are implemented via the use of the NRC sponsored computer code LADTAP II. Input parameter values typically used in the dose models are listed in MP-22-REC-REF03, "REMODCM Technical Information Document." This same methodology is used in the determination of compliance with the 40CFR190 total dose standard for the liquid pathways.

The conversion constants in the Method 1 equations are based on the maximum observed comparison of historical effluent releases for each unit and corresponding whole body or critical organ doses to a maximum individual. The dose conversion factors are calculated based on the ratio of the observed highest dose (whole body and organ) and the curies of fission and activation products released during the period. This ratio results in the Method 1 equation conversion factor in mrem/Ci released. This same approach was repeated separately for tritium (as a different radionuclide class) discharged in liquids wastes. MP-22-REC-REF03 describes the derivation of the Method 1 constants and list the historical whole body and maximum organ doses calculated for each unit operation.

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II.D. Gaseous Dose Calculations The determination of potential release rates and doses from radioactive gaseous effluents to the maximum off-site receptor are divided into two methods. Method 1 provides simplified operational tools to ensure that effluent releases are not likely to cause quarterly and annual off-site dose or dose rate limits to be exceeded. Effluent doses are calculated at least once every 31 days. Method 2 provides for a more detailed computational calculation using accepted computer models to demonstrate actual regulatory compliance. Method 2 is used whenever the Method 1 estimation approaches a regulatory limit, and for preparation of the Radioactive Effluent Release Report which includes the quarterly and annual dose impacts for all effluents recorded discharged to the atmosphere during the year of record.

1. Site Release Rate Limits ("Instantaneous")

Radiological Effluent Controls (Sections III, IV, and V) for each unit require that the instantaneous off-site dose rates from noble gases released to the atmosphere be limited such that they do not exceed 500 mrem/year at any time to the whole body or 3000 mrem/year to the skin at any time from the external cloud. For iodine- 131, 133, tritium, and particulates (half-lives >

8 days), the inhalation pathway critical organ dose rate from all units shall not exceed 1500 mrem/year at any time. These limits apply to the combination of releases from all three Units on the site, and are directly related to the radioactivity release rates measured for each Unit. By limiting gaseous release rates for both classes of radionuclides (i.e., noble gases; and iodines, tritium, and particulates) to within values which correlate to the above dose rate limits, assurance is provided that the Radiological Effluent Controls dose rate limits are not exceeded.

a. Method 1 for Noble Gas Release Rate Limits The instantaneous noble gas release rate limit from the site shall be:

QIV/90,000 +Q2s/560,000 + Q2v/290,000 + Q3S/ 5 6 0 ,0 0 0 + Q3v/290,000 < 1 Where:

Qiv = Noble gas release rate from Spent Fuel Pool Island Vent

([Ci/sec)

Q2S = Noble gas release rate from MP2 to Millstone Stack (ýtCi/sec)

Q2V = Noble gas release rate from MP2 Vent ([tCi/sec)

Q3V = Noble gas release rate from MP3 Vent ([tCi/sec)

Q3S = Noble gas release rate from MP3 to Millstone Stack ([tCi/sec)

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As long as the above is less than or equal to 1, the doses will be less than or equal to 500 mrem to the total body and less than 3000 mrem to the skin. The limiting factor for the Unit 1 SFPI vent of 90,000 is based on the skin dose limit of 3,000 mrem/year, while all the other factors are based on the whole body dose limit of 500 mrem/year.

b. Method 1 Release Rate Limit 131,1-133, H-3 and Particulates Half Lives Greater Than 8 Days With releases satisfying the following limit conditions, the dose rate to the maximum organ will be less than 1500 mrem/year from the inhalation pathway:

1) The site release rate limit of 1-131, 1-133, and tritium (where the thyroid is the critical organ for these radionuclides) shall be:

DRthyl + DRthy2 + DRthy3 < 1 Where the contribution from each Unit is calculated from:

Unit 1: DRthyl = 9.36 x 10-6 QHIV Unit 2: DRthy2 = 5.1 x 10-2 131QI2V + 2.38 x 10-3 1 3 1Q1 2s +

1.25 x 10-2 133Q12V+5. 7 5 x 10- 4 133Q12S +

4.2 x 10-6 QH2V+ 1.9 x 10-7 QH2S Unit 3: DRthy3 = 5.1 x 10-2 131Q13V + 2.38 x 10- 3 3IQI3s +

1.25 x 10-2 133Q[3V +5.75 x 10-4 133Q13s +

4.2 x 10-6 QH3V + 1.9 x 10-7 QH3S

2) The site release rate limit of particulates with half-lives greater than 8 days and tritium (where the critical organ is a composite of target organs for a mix of radionuclides) shall be:

DRorg1 + DRorg2 + DRorg3 <1 Where the contribution from each Unit is calculated from:

Unit l:DRorgi = 1.05 x 10-1 [Qplv + QPIB ]+9.36 x 10-6 QHIV Unit 2:DRorg2 = 2.38 x 10-3 QP2S + 5.1 x 10-2 QP2V + 1.9 x 10-7 QH2S + 4.2 x 10-6 QH2V Unit 3:DRorg3 = 2.38 x 10-3 QP3S + 5.1 x 10-2 QP3V +

1.9 x 10- 7 QH3S + 4.2 x 10-6 QH3V MP-22-REC-BAPO1 STQP THINK ACT RVIFEW Rev. 026-01 63 of 166

Each of the release rate quantities in the above equations are defined as:

131QI2V = Release rate of 1-131 from MP2 Vent ([tCi/sec)*

131Q12S = Release rate of 1-131 from MP2 to Millstone Stack

([tCi/sec) 133Q12V = Release rate of 1-133 from MP2 Vent ([tCi/sec)*

133Q12S = Release rate of 1-133 from MP2 to Millstone Stack (gtCi/sec) 131Q13v = Release rate of 1-131 from MP3 Vents (Normal and ESF)

([Ci/sec)*

131Q13S = Release rate of 1-131 from MP3 to Millstone Stack

([tCi/sec) 133Q13V = Release rate of 1- 133 from MP3 Vents (Normal and ESF)

([tCi/sec)*

133QB3s = Release rate of 1-133 from MP3 to Millstone Stack

([tCi/sec)

QH1V = Release rate of tritium from the Spent Fuel Pool Island and Balance of Plant Vents (!tCi/sec)

QH2V = Release rate of tritium from MP2 Vent ([tCi/sec)*

QH2S = Release rate of tritium from MP2 to Millstone Stack (uCi/sec)

QH3V = Release rate of tritium from MP3 Vents (Normal and ESF)

([iCi/sec)

• QH3S = Release rate of tritium from MP3 to Millstone Stack

([tCi/sec)

QpIv = Release rate of total particulates with half-lives greater than 8 days from the Spent Fuel Pool Island Vent (jxCi/sec)

QPIB = Release rate of total particulates with half-lives greater than 8 days from the Balance of Plant Vent (jxCi/sec)

QP2V = Release rate of total particulates with half-lives greater than 8 days from the MP2 Vent ([tCi/sec)

QP2S = Release rate of total particulates with half-lives greater than 8 days from MP2 to Millstone Stack (ltCi/sec)

QP3V = Release rate of total particulates with half-lives greater than 8 days from MP3 Vents (Normal and ESF) (,uCi/sec)

QP3S = Release rate of total particulates with half-lives greater than 8 days from MP3 to Millstone Stack ([tCi/sec)

• includes releases via the steam generator blowdown tank vent.

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c. Method 2 The above Method 1 equations assume a conservative nuclide mix. If necessary, utilize the GASPAR, or a code which uses the methodology given in Regulatory Guide 1.109, code to estimate the dose rate from either noble gases or iodines, tritium, and particulates with half-lives greater than 8 days. The use of the code is described in MP-22-REC-GDLO4, "Gaseous Dose Calculations - GASPARII."

Additional information on GASPAR is contained in the MP-22-REC-REF03, "REMODCM Technical Information Manual."

2. 10 CFR50 Appendix I - Noble Gas Limits Radiological Effluent Controls (Sections III, IV, and V) limit the off-site air dose from noble gases released in gaseous effluents to 5 mrad gamma and 10 mrad beta for a calendar quarter (10 and 20 mrad gamma and beta, respectively, per calendar year). Effluent dose calculations are calculated at least once every 31 days. In addition, installed portions of the gaseous radwaste treatment system are required to be operated to reduce radioactive materials in gaseous effluents when the projected doses over 31. days from the applicable waste stream exceed 0.02 mrad air gamma or 0.04 mrad air beta. (See Appendix A, Tables App.A- 1 for a cross reference of effluent control requirements and applicable sections of the REMODCM which are used to determine compliance.) This part of the REMODCM provides the calculation methodology for determining air doses from noble gases.

a. Method I Air Dose* (Applicable to Units 1, 2, and 3)

For Unit 1: DG1 = 3.3 x 10-6 CNlV*

DB1 = 1.49 x 10-3 CNiV*

For Unit 2: DG2 = 6.3 x 10-4 CN2V + 1.81 x 10-4 CN2S

• DB2 = 1.7 x 10-3 CN2V + 1.81 x 10-6 CN2S

• For Unit 3: DG3 = 6.3 x 10-4 CN3V+ 1.81 x 10-4 CN3S

• DB3 = 1.7 x 10-3 CN3V+ 1.81 x 10-6 CN3S If DG1, DG2, or DG3 are greater than 1.6 mrad or DBI, DB2, or DB3 are greater than 3.3 mrad within a calendar quarter, go to Method 2 below.

Where:

DG1 = the gamma air dose from Unit 1 for the period of interest (mrad).

DB1 = the beta air dose from Unit lfor the period of interest (mrad).

MP-22-REC-BAP01 STOP TH.INK ACT REVIEEW Rev. 026-01 65 of 166

DG2 = the gamma air dose from Unit 2 for the period of interest (mrad).

DB2 = the beta air dose from Unit 2 for the period of interest (mrad).

DG3 = the gamma air dose from Unit 3 for the period of interest (mrad).

DB3 = the beta air dose from Unit 3 for the period of interest (mrad).

CN1V= the total curies of noble gas released from Spent Fuel Pool Island Vent during the period of interest.

CN2V = the total curies of noble gas released from Unit 2 Vent during the period of interest. Include containment releases to Unit 2 Vent CN2S = the total curies of noble gas released from Unit 2 to Millstone Stack during the period of interest.

CN3V = the total curies of noble gas released from Unit 3 vents during the period of interest. Include containment releases to Unit 3 Vent and ESF Building Vent.

CN3S = the total curies of noble gas released from Unit 3 to Millstone Stack during the period of interest.

• See MP-22-REC-REF03, "REMODCM Technical Information Document," Section 4.2, for the derivation of air dose Method I factors.

b. Method 2 Air Dose (Applicable to Units 1, 2, and 3)

Use the GASPAR computer code, or a code which uses the methodology given in Regulatory Guide 1.109, to determine the critical site boundary air doses.

For the Special Location, enter the following worst case quarterly average meteorology based on the Unit 2 vent eight-year history for 1980 to 1987:

3 x/Q = 8.1 x 10-6 sec/m D/Q = 1.5 x 10-7 m-2 (See the MP-22-REC-REF03, (Att) "Determination of Maximum x/Q and D/Q.")

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If the calculated air dose exceeds one half the quarterly Radiological Effluent Control limit, use meteorology concurrent with quarter of release.

c. Estimation of Annual Air Dose Limit Due to Noble Gases (Applicable to Units 1, 2, and 3)

An estimation of annual (year-to-date) beta and gamma air doses (DYB and DyG, respectively) from noble gases released from Units 1, 2 and 3 shall be made every month to determine compliance with the annual dose limits for each Unit. Annual air doses will be determined as follows:

Unit 1 Unit 2 Unit 3 DYGL = ZDGI DYG2 = YDG2 DYG3 = ZDG3 DYB1 = Y.DBI DYB2 = YDB2 DYB3 = EDB3 Where the sums are over the first quarter (i.e., summation of the all release periods within the quarter) through the present calendar quarter doses.

Where: DYG1, DYG2, DYG3, DYBI, DYB2 and DyB3 = gamma air dose and beta air dose for the calendar year for Unit 1, 2, or 3.

The following shall be used as the quarterly doses:

(1) If the detailed quarterly dose calculations required per Section II.D.5. for the Radioactive Effluent Release Report are complete for any calendar quarter, use those results.

(2) If the detailed calculations are not complete for a particular quarter, use the results as determined above in Sections II.D.2.a. or II.D.2.b.

If DyG1, YG2 or YG3 are greater than 10 mrad or DyB1, YB2 or YB3 are greater than 20 mrad and any corresponding quarterly dose was not calculated using Method 2 (Section II.D.2.b.), recalculate the quarterly dose using meteorology concurrent with quarter of release.

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3. 10 CFR50 Appendix I - Iodine, Tritium and Particulate Doses Radiological Effluent Controls (Section III, IV, and V) limit the off-site dose to a critical organ from radioiodines, tritium, and particulates with half-lives greater than 8 days released in gaseous effluents to 7.5 mrem for a calendar quarter (15 mrem per calendar year). Effluent dose calculations are performed at least once every 31 days. In addition, installed portions of the gaseous radwaste treatment system are required to be operated to reduce radioactive materials in gaseous effluents when the projected doses over 31 days from the applicable waste stream exceed 0.03 mrem. (See Appendix A, Table App.A- 1 for a cross reference of effluent control requirements and applicable sections of the REMODCM which are used to determine compliance.) This part of the REMODCM provides the calculation methodology for determining critical organ doses from atmospheric releases of iodines, tritium and particulates.

a. Critical Organ Doses (Applicable to Millstone Stack and Unit 1 releases)

1) Method 1 - Millstone Stack and Unit 1 Releases Calculate organ doses for DTS and Dos:

For Unit 2 or 3: DTS = 30 1 3 1CIS + 0.29 13 3 Cis + 6.33 x 10-6 CHS 0 DOS = 9.0Cps + 6.33 x 1 0 - CHS Sum critical organ doses from stack with critical organ doses from vent in Section II.D.3.b.1) below:

If either dose is greater than 2.5 mrem within a calendar quarter go to Method 2 below For Unit 1: DTS = 1.97 x 10-3 CHV DOS = 94.8[Cpv+ CpB]+ 1.97 x 10-3 CHV If either dose is greater than 2.5 mrem within a calendar quarter go to Method 2 below Where:

DTS = the thyroid dose for the period of release of gaseous effluents.

DOS = the dose to the maximum organ other than the thyroid for the period of gaseous effluent release.

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131CIS =The total curies of 1- 131 released in gaseous effluents from Unit 2 or 3 to Millstone Stack during the period of interest.

133CIS = The total curies of 1-133 released in gaseous effluents from Unit 2 or 3 to Millstone Stack during the period of interest.

Cps the total curies of particulates with half-lives greater than 8 days released in gaseous effluents from Millstone Stack during the period of interest.

Cpv the total curies of particulates with half-lives greater than 8 days released in gaseous effluents from the SFPI vent during the period of interest.

CPB = the total curies of particulates with half-lives greater than 8 days released in gaseous effluents from the BOP vent during the period of interest.

CHS = the total curies of tritium released in gaseous effluents from Millstone Stack during period of interest.

CHV = the total curies of tritium released in gaseous effluents from the SFPI and BOP vents during period of interest.

2) Method 2 - Millstone Stack and Unit 1 Releases Use the GASPAR code, or a code which uses the methodology given in Regulatory Guide 1.109, with actual locations, real-time meteorology and the pathways which actually exist at the time at those locations:

Sum critical organ doses from stack with critical organ doses from vent in Section II.D.3.b. below.

b. Critical Organ Doses (Applicable to Units 2 and 3 vent releases)

1) Method 1 - Unit 2 and Unit 3 releases For Unit 2 and Unit3, separately, calculate organ doses DT and DO:

DTV = 230 131CIV + 4.0 133 CIV + 2.6 x 10-3 CHV Dov= 1.1 X 0 3 CPV + 2.6 x 10-3 CHV Sum with organ doses for releases from the stack from Section II.D.3.a.1):

DT = DTS + DTV Do = Dos + Dov MP-22-REC-BAPO1 STOP THINK 'ACT REVIEW Rev. 026-01 69 of 166

If either dose is greater than 2.5 mrem within a calendar quarter go to Section II.D.3.a. and recalculate any organ dose greater than 2.5 mrem for releases from the stack and go to Method 2 and recalculate any organ dose greater than 2.5 mrem for releases from the vent, where:

DT = the total thyroid dose for the period of gaseous effluents releases.

Do= the total dose to the maximum organ other than the thyroid for the period of gaseous effluent releases.

DTV = the thyroid dose for the period of gaseous effluents releases from the vent.

Dov= the dose to the maximum organ other than the thyroid for the period of gaseous effluent releases from the vent.

131Civ=The total curies of 1-131 in gaseous effluents from Unit 2 other than to the Millstone Stack (Unit 2 Vent, containment releases to vent, and Steam Generator Blowdown Tank Vent) or from Unit 3 other than to the Millstone Stack (Unit 3 Vent, ESF Building Vent, containment releases to vent, Steam Generator Blowdown Tank Vent, and Containment Drawdown using mechanical vacuum) during the period of interest.

133CIV=The total curies of 1-133 in gaseous effluents from Unit 2 other than to the Millstone Stack (Unit 2 Vent, containment releases to vent, and Steam Generator Blowdown Tank Vent) or from Unit 3 other than to the Millstone Stack (Unit 3 Vent, ESF Building Vent, containment releases to vent, Steam Generator Blowdown Tank Vent, and Containment Drawdown using mechanical vacuum) during the period of interest.

Cps The total curies of particulates with half-lives greater than eight days released in gaseous effluents from Unit 2 other than to -the Millstone Stack (Unit 2 Vent and containment releases to vent) or from Unit 3 other than to the Millstone Stack (Unit 3 Vent, ESF Building Vent, containment releases to vent, and Containment .

Drawdown using mechanical vacuum) during the period of interest.

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CHV = The total curies of tritium released in gaseous effluents from Unit 2 other than to the Millstone Stack (Unit 2 Vent, Steam Generator Blowdown Tank Vent and containment releases to vent) or from Unit 3 other than to the Millstone Stack (Unit 3 Vent, ESF Building Vent, Steam Generator Blowdown Tank Vent containment releases to vent, and Containment Drawdown using mechanical vacuum) during the period of interest.

2) Method 2 - Unit 2 and Unit 3 releases Use the GASPAR code, or a code which uses the methodology given in Regulatory Guide 1.109, with the actual locations, real-time meteorology and the pathways which actually exist at the time at these locations. For Unit 2, the code shall be run separately for steam generator blowdown tank vents and ventilation releases, containment purges and waste gas tank releases. For Unit 3, the code shall be run separately for ventilation, process gas, containment vacuum system, ESF ventilation and containment purges.

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c. Estimation of Annual Critical Organ Doses Due to lodines, Tritium and Particulates (Applicable to Units 1, 2, and 3)

An estimation of annual (year-to-date) critical organ doses (DyT and DyO for thyroid and maximum organ other than thyroid, respectively) from radioiodine, tritium and particulates with half- lives greater than 8 days released from Units 1, 2 and 3 shall be made every month to determine compliance with the annual dose limits for each Unit. Annual critical organ doses will be determined as follows:

Unit I Unit 2 Unit 3 DYTI = IDT1 DYG2 = EDT2 DYT3 = YDT3 Dyo 1 = EDo 1 Dy0 2 = ED0 2 Dy03 = EDo 3 Where the sums are over the first quarter (i.e., summation of the all release periods within the quarter) through the present calendar quarter doses.

Where:

DyT1, DyT2, DYT3, Dyol, Dy0 2 and Dy03 = thyroid (T) dose and maximum organ (0) dose (other than the thyroid) for the calendar year for Unit 1, 2, or 3.

The following guidelines shall be used for DT and DO:

(1) If the detailed quarterly dose calculations required per Section II.D.5. for the Radioactive Effluent Release Report are complete for any calendar quarter, use those results.

(2) If the detailed calculations are not complete for a particular quarter, use the results as determined above in Section II.D.3.a. or II.D.3.b.

(3) If DyT and/or Dyo are greater than 15 mrem and quarterly dose was not calculated using Method 2 of Section II.D.3.a. or II.D.3.b.,

recalculate the quarterly dose using Method 2.

(4) If different organs are the maximum organ for different quarters, they can be summed together and Dyo recorded as a less-than value as long as the value is less than 15 mrem. If it is not, the sum for each organ involved shall be determined.

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4. Gaseous Effluent Monthly Dose ProjectionsSection I.D.2.a. of the REMM requires that certain portions of the gaseous radwaste treatment equipment be returned to service to reduce radioactive gaseous effluents when the projected doses for each Unit (made at least once per 31 days) exceed 0.02 mrad gamma air, 0.04 mrad beta air, or 0.03 mrem to any organ from gaseous effluents. The following methods are applied in the estimation of monthly dose projections.

a. Unit 1 Projection Method None required.

b. Unit 2 Projection Method

1) Due to Gaseous Radwaste Treatment System (Unit 2)

Determine the beta and gamma monthly air dose projection from noble gases from the following:

DEMG (mrad) = 1.81 x 10-4 CEN DEMB (mrad) = 1.81 x 10-6 CEN Where:

CEN = the number of curies of noble gas estimated to be released from the waste gas storage tanks during the next month.

DEMG = the estimated monthly gamma air dose.

DEMB= the estimated monthly beta air dose.

(The dose conversion factor is from MP-22-REC-REF03, "REMODCM Technical Information Document," Section 4.2, for the Millstone Stack releases since the Unit 2 waste gas tanks are discharged via the Millstone Stack. This factor is conservative because the isotopic mix assumed for the dose conversion factor consists of shorter-lived noble gases which have higher dose conversion factors than the typical mix from Unit 2 waste gas tank discharges.)

.2) (Reserved)

3) Due to Ventilation Releases (Unit 2)

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If portions of the ventilation treatment system are expected to be out of service during the month, determine the monthly maximum organ dose projection (DEMO) from the following:

Method 1 Determine DEMO which is the estimated monthly dose to the maximum organ from the following:

DEMO = 1/3 Rl,(1.01- R 2 ) (R 3 + 0.01) Do For the last quarter of operation, determine Do as determined per Section II.D.3.b.

R, = the expected reduction factor for the HEPA filter.

Typically this should be 100 (see NUREG-0016 or 0017 for additional guidance).

R2 = the fraction of the time which the equipment was inoperable during the last quarter.

R3 = the fraction of the time which the equipment is expected to be inoperable during the next month.

ii. Method 2 If necessary, estimate the curies expected to be released for the next month and applicable method for dose calculation from Section II.D.3.b.

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c. Unit 3 Projection Method

1) Due to Radioactive Gaseous Waste System (Unit 3)

Determine the beta and gamma monthly air dose projection from noble gases from the following:

DEMG (mrad) = 1.81 x 10-4 CEN DEMB (mrad) = 1.81 x 10-6 CEN Where:

CEN = the number of curies of noble gas estimated to be released from the reactor plant gaseous vents to the Millstone stack (the activity from this pathway increases when the process waste gas system is out of service.) during the next month.

DEMG =the estimated monthly gamma air dose.

DEMB = the estimated monthly beta air dose.

(The dose conversion factor is from the MP-22-REC-REF03, "REMODCM Technical Information Document," for the Millstone Stack releases since the Unit 3 reactor plant gaseous vents are discharged via the Millstone Stack.)

5. Quarterly Dose Calculations for Radioactive Effluent Release Report Detailed quarterly gaseous dose calculations required for the Radioactive Effluent Release Report shall be done using the computer code GASPAR, or a code which use the methodology given in Regulatory Guide 1.109. The use of LADTAP II code and the input parameters are given in MP-22-REC-GDLO4, "Gaseous Dose Calculations - GASPARII." Use of a code using the methodology given in Regulatory Guide 1.109 is described in MP-22-REC-GDLO5," Gaseous Dose Calculations - DOSAIR."

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6. Compliance with 40CFR190 The following sources shall be considered in determining the total dose to a real individual from uranium fuel cycle sources:

a. Gaseous Releases from Units 1, 2, and 3.

b. Liquid Releases from Units 1, 2, and 3.

c. Direct and Scattered Radiation from Radioactive Material on Site.

d. Since all other uranium fuel cycle sources are greater than 5 miles away, they need not be considered.

The Radiological Effluent Controls in Sections III.E. (Unit 1), IVE. (Unit 2), and V.E. (Unit 3) contain specific requirements for ensuring compliance with 40CFR190 based on gaseous and liquid doses (sources a and b).

Doses to source c are controlled by design and operations to ensure the off-site dose from each radwaste storage facility is less than one mrem per year. Potential doses from each facility are evaluated in Radiological Environmental Reviews (RERs) where total off-site doses from all sources are considered to ensure compliance with 40CFR190.

7. Bases for Gaseous Pathway Dose Calculations The dose calculation methodology and parameters used in Section II of the REMODCM implement the requirements in Section III.A. of Appendix I (10CFR50) which states that conformance with the ALARA dose objectives of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated.

Operational flexibility is provided by controlling the instantaneous release rate of noble gas (as well as iodines and particulate activity) such the maximum off-site dose rates are less than the equivalent of 500 mrem/year to the whole body, 3000 mrem/year to the skin from noble gases, or 1500 mrem/year to a critical organ from the inhalation of iodines, tritium and particulates. The dose rate limits are based on the 10CFR20 annual dose limits, but applied as an instantaneous limit to assure that the actual dose over a year will be well below these numbers.

The equivalent instantaneous release rate limits for Millstone Stack were determined using the EPA AIREM code. For Units 2 & 3, these doses were calculated using the NRC GASPAR code. The AIREM code calculates cloud gamma doses using dose tables from a model that considers the finite extent of the cloud in the vertical direction. Beta doses are calculated assuming semi-infinite cloud concentrations, which are based upon a MP-22-REC-BAP01 STOP' THINK ACT REVIEW Rev 026-01 76 of 166

standard sector averaged diffusion equation. The GASPAR code implements the models of NRC Regulatory Guide 1.109, Rev. 1, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I." Input parameter values typically used in the dose models are listed in MP-22-REC-REF03, "REMODCM Technical Information Document." This same methodology is used in the determination of compliance with the 40CFR190 total dose standard for the gaseous pathways.

In the determination of compliance with the dose and dose rate limits, maximum individual dose calculations are performed at the nearest land site boundary with maximum decayed X/Q, and at the nearest vegetable garden (assumed to be nearest residence) and cow and goat farms with maximum D/Qs. The conversion constants in the Method 1 equations for maximum air doses, organ and whole body doses, and dose rates are based on the maximum observed comparison of historical effluent releases and corresponding calculated maximum doses. The dose conversion factors are calculated based on the ratio of the observed highest dose and the curies of fission and activation products released during the period. This ratio results in the Method 1 equation conversion factor in mrem/Ci released.

MP-22-REC-REF02 describes the derivation of the Method 1 constants and list the historical maximum doses calculated for the maximum organ.

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II.E. Liquid Discharge Flow Rates And Monitor Setpoints

1. Unit 1 Reactor Cavity Water Discharge Line The limit on discharge flow rate and setpoint on the Unit 1 liquid waste monitor depend on dilution water flow, radwaste discharge flow, the isotopic composition of the liquid, the background count rate of the monitor and 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 each batch. The following method will be used:

STEP 1:

From the isotopic analysis and the Effluent Concentration (EC) values for each identified nuclide determine the required reduction factor, i.e.:

R = Required Reduction Factor 0 1

-of nuclide i 10 x EC of nuclide i STEP 2:

Determine the allowable discharge flow (F)

F=0.1xRxD Where:

D = The existing dilution flow which is the any dilution flow from Millstone Unit 2 and/or Unit 3 not being credited for any other radioactivity discharge during discharge of Unit 1 water.

0.1 = safety factor to limit discharge concentration to 10% of the Radiological Effluent Control Limit.

STEP 3:

Calculate the monitor setpoint as follows:

Rset = 2 x AC x RCF Where:

Rset = The setpoint of the monitor.

AC = The total radwaste effluent concentration ([tCi/ml) in the tank.

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RCF = The response correction factor for the effluent line monitor using the current calibration factor or isotopic-specific responses.

2 = Tolerance limit which brings the setpoint at twice the expected response of the monitor based on sample analysis. With the safety factor of 0.1 the setpoint would be at 20% of the Radiological Effluent Control Limit.

Option setpoint:

A setpoint based upon worst case conditions may be used. Assume the maximum possible discharge flow, a minimum dilution flow not to exceed 100,000 gpm, and a limit of 1 x 10-7 *tCi/ml which is lower than any 10CFR20 EC limit except for transuranics. This will assure that low level releases are not terminated due to small fluctuations in activity. When using this option setpoint independent verification of discharge lineup shall be performed.

The optional setpoint may be adjusted (increased or decreased) by factors to account for the actual discharge flow and actual dilution flow; however, controls shall be established to ensure that the allowable discharge flow is not exceeded and the dilution flow is maintained.

2. Reserved

3. Unit 2 Clean Liquid Radwaste Effluent Line - RM9049 and Aerated Liquid Radwaste Effluent Line - RM9116 The setpoint on the Unit 2 clean and aerated liquid waste effluent lines depend on dilution water flow, radwaste discharge flow, the isotopic composition of the liquid, the background count rate of the monitor and the efficiency of the monitor. Due to the variability of these parameters, an alarm/trip setpoint will be determined prior to the release of each batch.

The following method will be used:

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STEP 1:

From the tank isotopic analysis and the Effluent Concentrations (EC) in 10CFR20, App. B, Table 2, Col. 2 for each identified nuclide determine the required reduction factor, i.e.:

For Nuclides Other Than Noble Gases:

R = Required Reduction Factor =

Z C

,,,, of nuclide i 10 x EC of nuclide i For Noble Gases: If the noble gas concentration is less than 5.7 x 10-3 tCi/ml, the reduction factor need not be determined. This 0 concentration is based on 100,000 gpm dilution flow and a safety factor of 0.1 (See Note below.)

2 x 10-41"' ml R2= Required Reduction Factor = o E,*of noble gases 2 x 10 n oble gases R = the smaller of R1 or R2 STEP 2:

Determine the allowable discharge flow (F) in gpm:

F =0.1xRxD Where:

D = the existing dilution flow (D) from circulating and service water pumps. I NOTE Note that discharging at this flow rate would yield a discharge concentration corresponding to 10% of the Radiological Effluent Control Limit due to the safety factor of 0.1.

With this condition on discharge flow rate met, the monitor setpoint can be calculated:

Rset = 2 x AC x RF (See Note 1 below.)

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Where:

Rset the setpoint of the monitor (cpm).

AC = the total radwaste effluent concentration (tCi/ml) in the tank.

RF = the response factor for the effluent line monitor using the current calibration factor or isotopic-specific responses.

2 the multiple of expected count rate on the monitor based on the radioactivity concentration in the tank.

This value or that corresponding to 2.8 x 10-5 ýtCi/ml (Note 2 below),

whichever is greater, plus background is the trip setpoint. For the latter setpoint, independent valve verification shall be performed and minimum dilution flow in Note 2 shall be verified and if necessary, appropriately adjusted.

Note 1: If discharging at the allowable discharge rate (F) as determined in above, this setpoint would correspond to 20% of the Radiological Effluent Control limit.

Note 2: This value is based upon assuming maximum discharge flow (350 gpm), dilution water flow of 100,000 gpm and a limit of I x 10-7 which is lower than any Technical Specification limit (ten times 10CFR20 EC values) except for transuranics. This will assure that low level releases are not terminated due to small fluctuations in activity.

However, to verify that the correct tank is being discharged when using this value, independent valve verification shall be performed.

This value may be adjusted (increased or decreased) by factors to account for the actual discharge flow and actual dilution flow; however, controls shall be established to ensure that the allowable discharge flow is not exceeded and the dilution flow is maintained.

4. Condensate Polishing Facility Waste Neutralization Sump Effluent Line -

CND245 When the grab sample prior to release required by Table I.C.-2 is greater than 5 x 10-7 [Ci/ml, the setpoint shall be determined as for the Clean and Aerated Liquid Monitors in Section II.E.3. except the CPF monitor has the capability to readout in CPM or ptCi/ml. If the grab sample is less than 5 x 10-7 tCi/ml, use a setpoint of the lower of ten times background or the value as specified in II.E.3. A setpoint based on ten times background shall not exceed a reading corresponding to 2.8 x 10-5 [tCi/ml, which is approximately 6,300 CPM based on recent calibration data.

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5. Unit 2 Steam Generator Blowdown - RM4262 and Unit 2 Steam Generator Blowdown Effluent Concentration Limitation 5a. Unit 2 Steam Generator Blowdown - RM4262 Assumptions used in determining the Alarm, setpoint for this monitor are:

a. Total S.G. blowdown flow rate = 700 gpm.

b. Minimum circulating water dilution flow during periods of blowdown = 100,000 gpm. 1I0

c. The release rate limit is conservatively set at 3 x 10-8 [tCi/ml which is lower than any 10CFR20 Effluent Concentration (EC) limit except for some transuranics *

d. Background can be added after above calculations are performed.

Therefore, the alarm setpoint corresponds to a concentration of:

100,000 x 3x10- + background**= .4.3x10 6 Mi/mi + background Alarm (uCi/ml) =

The latest monitor calibration curve shall be used to determine the alarm setpoint in cpm corresponding to 4.3 x 10-6 [tCi/ml.

This setpoint may be adjusted (increased or decreased) through proper administrative controls if the steam generator blowdown rate is maintained other than 700 gpm and/or other than 100,000 gpm circulating water flow are available. The adjustment would correspond to the ratio of flows to those assumed above or:

I circulating & service water flow (gpm) 700 Alarm (uCi/ml) = 4.3xlO-6Ci/ml x + 10 I uU, ttUV SIG bfowd.,v (gPlm)

Background = 3xlO-ýu~i/il x circulating & service water flow (gpm) + Backgriound total S/G blowdown (gpm)

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NOTE The Steam Generator Blowdown alarm criteria is in practice based on setpoints required to detect allowable levels of primary to secondary leakage. This alarm criteria is typically more restrictive than that required to meet discharge limits. This fact shall be verified, however, whenever the alarm setpoint is recalculated.

• In lieu of using 3 x 10-8 ItCi/ml, the identified EC limits from 10CFR20 may be used.

• • Background of monitor at monitor location (i.e., indication provided by system monitor with no activity present in the monitored system).

5b. Unit 2 Steam Generator Blowdown Effluent Concentration Limitation The results of analysis of blowdown samples required by Table I.C.-2 of Section I of the REMODCM shall be used to ensure that blowdown effluent releases do not exceed ten times the concentration limits in 10CFR20, Appendix B.

6. Unit 2 Condenser Air Ejector - RM5099 N/A since this monitor is no longer a final liquid effluent monitor.

7. Unit 2 Reactor Building Closed Cooling Water RM6038 and Unit 2 Service Water, and RBCCW Sump and Turbine Building Sump Effluent Concentration Limitation 7a. Unit 2 Reactor Building Closed Cooling Water RM6038 The purpose of the Reactor Building Closed Cooling Water (RBCCW) radiation monitor is to give warning of abnormal radioactivity in the RBCCW system and to prevent releases to the Service Water system which, upon release to the environment, would exceed ten times the concentration values in 10CFR20. According to Calculation RERM-02665-R2, radioactivity in RBCCW water which causes a monitor response of greater than the setpoint prescribed below could exceed ten times thel0CFR20 concentrations upon release to the Service Water system.

SETPOINT DURING POWER OPERATIONS:

To give adequate warning of abnormal radioactivity, the setpoint shall be two times the radiation monitor background reading, provided that the MP-22-REC-BAP01 STOP' THINK ACT REVIEWWRev. 026-01 83 of 166

background reading does not exceed 2,000 cpm. The monitor background reading shall be the normal monitor reading. If the monitor background reading exceeds 2,000 cpm, the setpoint shall be set at the background reading plus 2,000 cpm and provisions shall be made to adjust the setpoint if the background decreases.

SETPOINT DURING SHUTDOWN:

1) During outages not exceeding three months the setpoint shall be two times the radiation monitor background reading, provided that the background reading does not exceed 415 cpm. If the monitor background reading exceeds 415 cpm, the setpoint shall be set at the background reading plus 415 cpm and provisions shall be made to adjust the setpoint if the background decreases.

2) During extended outages exceeding three months, but not exceeding three years, the setpoint shall be two times the radiation monitor background reading, provided that the background reading does not exceed 80 cpm. If the monitor background reading exceeds 80 cpm, the setpoint shall be set at the background reading plus 80 cpm and provisions shall be made to adjust the setpoint if the background decreases.

PROVISIONS FOR ALTERNATE DILUTION FLOWS:

These setpoints are based on a dilution flow of 4,000 gpm from one service water train. If additional dilution flow is credited, the setpoint may be adjusted proportionately. For example, the addition of a circulating water pump dilution flow of 100,000 gpm would allow the setpoint to be increased by a factor of 25.

7b. Unit 2 Service Water, and RBCCW Sump and Turbine Building Sump Effluent Concentration Limitation Results of analyses of service water, RBCCW sump and turbine building sump samples taken in accordance with Table I.C.-2 of Section I of the REMODCM shall be used to limit radioactivity concentrations in the service water, RBCCW sump and turbine building sump effluents to less than ten times the limits in 10CFR20, Appendix B.

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8. Unit 3 Liquid Waste Monitor - LWS-RE70 The setpoints on the Unit 3 liquid waste monitor depend on dilution water flow, radwaste discharge flow, the isotopic composition of the liquid, the background count rate of the monitor and 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 each batch. The following method will be used:

Step 1:

From the tank isotopic analysis and the Effluent Concentration (EC) values for each identified nuclide determine the required reduction factor, i.e.:

For Nuclides Other Than Noble Gases:

R Required Reduction Factor = 1

,, of nuclide i 10 x EC of nuclide i For Noble Gases: If the noble gas concentration is less than 0.013 [Ci/ml, the reduction factor need not be determined. This concentration is based on 100,00 gpm dilution flow and a safety factor of 0.1 (See Note Below.) 0 2 x 10-41*/

M/C R2 = Required Reduction Factor - o e

'gi, f noble gases Z-lnoble gases 2 x 10-4"o R = the smaller of R1 or R2 Step 2:

Determine the allowable discharge flow (F)

F = 0.1xRxD Where:

D = The existing dilution flow (P) from circulating and service water pumps. 0 MP-22-REC-BAP01 STO P THINK ACT1 REVIEW Rev. 026-01 85 of 166

NOTE Note that discharging at this flow rate would yield a discharge concentration corresponding to 10% of the Radiological Effluent Control Limit due to the safety factor of 0.1.

With this condition on discharge flow rate met, the monitor setpoint can be calculated:

Rset = 2 x AC x RCF (see Note 1)

Where:

Rset= The setpoint of the monitor.

AC= The total radwaste effluent concentration ([tCi/ml) in the tank.

RCF= The response correction factor for the effluent line monitor using the current calibration factor or isotopic-specific responses.

2 = The multiple of expected count rate on the monitor based on the radioactivity concentration in the tank.

This value, or that corresponding to 6.6 x 10-5 [tCi/ml (Note 2 below),

whichever is greater, plus background is the trip setpoint. For the latter setpoint, independent valve verification shall be performed and minimum dilution flow in Note 2 shall be verified and if necessary, appropriately adjusted.

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NOTE

1. If discharging at the allowable discharge rate (F) as determined above, this Alarm setpoint would yield a discharge concentration corresponding to 20% of the Radiological Effluent Control limit.

2. This value is based upon assuming maximum discharge flow (150 gpm), dilution water flow of 100,000 gpm, and a limit of 1 x 10-7 JtCi/ml which is lower than any Technical Specification limit (ten times 10CFR20 EC values) except for transuranics. This will assure that low level releases are not terminated due to small fluctuations in activity.

However, to verify that the correct tank is being discharged when using this value, independent valve verification shall be performed. This value may be adjusted (increased or decreased) by factors to account for the actual discharge flow and actual dilution flow; however, controls shall be established to ensure that the allowable discharge flow is not exceeded and the dilution flow is maintained

9. Unit 3 Regenerant Evaporator Effluent Line - LWC-RE65 The MP3 Regenerant Evaporator has been removed from service with DCR M3-97-041. Therefore a radiation monitor alarm is not needed.

10. Unit 3 Waste Neutralization Sump Effluent Line - CND-RE07 Same asSection II.E.8.

11. Unit 3 Steam Generator Blowdown - SSR-RE08 and Unit 3 Steam Generator Blowdown Effluent Concentration Limitation 11a. Unit 3 Steam Generator Blowdown - SSR-RE08 The alarm setpoint for this monitor assumes:

a. Steam generator blowdown rate of 400 gpm (maximum blowdown total including weekly cleaning of generators - per ERC 25212-ER-99-0133).

b. The release rate limit is conservatively set at 3 x 10-8 uCi/ml which is well below any 10CFR20 Effluent Concentration except for transuranics*.

c. Circulating and service water dilution flow during periods of blowdown = 100,000 gpm. I

d. Background can be added after above calculations are performed.

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Therefore, the alarm setpoint corresponds to a concentration of: I Alarm*Ci/l) -100, 000 Alarm uCi/m = 100400 x 3x10- 8 + background = 7.5x10- 6 /uCi/ml + background This setpoint may be increased through proper administrative controls if the steam generator blowdown rate is maintained less than 400 gpm and/or more than 100,000 gpm dilution flow are available. The amount I(

of the increase would correspond to the ratio of flows to those assumed above or:

Alarmn (uCi/l) = 7.5x10 AmlCi/l 5 x circulating & service 100,000water flow (gpmn) x S/G blowdown 400 (gpm) I +

Background = 3xlO-1uCi/ml x circulating & service water flow (gwn) + Background total S/G blowdown (gpm)

NOTE The Steam Generator Blowdown alarm criteria is in practice based on setpoints required to detect allowable levels of primary to secondary leakage. This alarm criteria is typically more restrictive than that required to meet discharge limits. This fact shall be verified, however, whenever the alarm setpoint is recalculated.

• In lieu of using 3 x 10-8 ,tCi/ml, ten times the identified 10CFR20 EC values may be used.

1lb. Unit 3 Steam Generator Blowdown Effluent Concentration Limitation The results of analysis of blowdown samples required by Table I.C.- 3 of

• Section I of the REMODCM shall be used to ensure that blowdown effluent releases do not exceed ten times the concentration limits in 10CFR20, Appendix B.

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12. Unit 3 Turbine Building Floor Drains Effluent Line - DAS-RE50 and Unit 3 Service Water and Turbine Building Sump Effluent Concentration Limitation 12a. Unit 3 Turbine Building Floor Drains Effluent Line - DAS-RE50 The alarm setpoint for this monitor shall be set to four times (4X) the reading of the monitor when there is no gamma radioactivity present in the turbine building sumps. As determined in Calculation RERM-04101R3, the setpoint shall not exceed 1.4 x 10-5 Ci/ml.

12b. Unit 3 Service Water and Turbine Building Sump Effluent Concentration Limitation Results of analyses of service water and turbine building sump samples taken in accordance with Table I.C.-3 of Section I of the REMODCM shall be used to limit radioactivity concentrations in the service water and turbine building sump effluents to less than ten times the limits in 10CFR20, Appendix B.

13. Bases for Liquid Monitor Setpoints Liquid effluent monitors are provided on discharge pathways to control, as applicable, the release of radioactive materials in liquid effluents during actual or potential releases of liquid waste to the environment. The alarm /

trip setpoints are calculated to ensure that the alarm / trip function of the monitor will occur prior to exceeding ten times the Effluent Concentration (EC) limits of 10 CFR 20 (Appendix B, Table 2, Column 2), which applies to the release of radioactive materials from all units on the site. This limitation also provides additional assurance that the levels of radioactive materials in bodies of water in Unrestricted Areas will result in exposures within the Section II.A. design objectives of Appendix I to 10CFR50 to a member of the public.

In application, the typical approach is to determine the expected concentration in a radioactive release path and set the allowable discharge rate past the monitor such the existing dilution flow will limit the effluent release concentration to 10% of the limit for the mix. The setpoint is then selected to be only 2 times the expected concentration, or 20% of the limit.

As a result, considerable margin is included in the selection of the setpoint for the monitor to account for unexpected changes in the discharge concentration or the contribution from other potential release pathways occurring at the same time as the planned effluent release. For those monitors on systems that are not expected to be contaminated, the alarm point is usually selected to be two times the ambient background to give notice that normal conditions may have changed and should be evaluated.

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II.E Gaseous Monitor Setpoints

1. Unit 1 Spent Fuel Pool Island Monitor - RM-SFPI-02 The instantaneous release rate limit from the site shall be set in accordance with the conditions given in Section II.D.l.a. in order to satisfy Unit 1 Radiological Effluent Controls III.C.2. and III.D.2.a.

The Unit 1 allocated portion of the site instantaneous release rate limit is (

30,000 [tCi/sec. This assumes that 7% of the site limit for skin dose of 3000 mrem per year is assigned to the Unit 1 Spent Fuel Pool Island vent. If effluent conditions from the Unit 1 Spent Fuel Pool Island vent reach 30,000

ýtCi/sec, releases from Units 2 and 3 vents and from the Millstone Stack shall be determined to ensure that the sum of the individual noble gas release rates do not cause the site skin dose limit to be exceeded. Use Section II.D.1.a. and Section 4.2 of MP-22-REC-REF03. "REMODCM Technical Information Document," in making this determination.

The alarm setpoint shall be set at or below the monitor reading in pCi/cc corresponding to the Unit 1 portion of the limit. The setpoint shall be set at or below 1.71E-3 ýtCi/cc. NOTE: This setpoint is the basis for emergency classification in Unit 1 EAL Table (OA- 1 and OU- 1). A change to this setpoint would require a concurrent change to the EAL.

2. Unit 2 Wide Range Gas Monitor (WRGM) - RM8169 The instantaneous release rate limit from the site shall be set in accordance with the conditions given in Section II.D.l.a. in order to satisfy Units 2 Radiological Effluent Controls IVC.2. and IVD.2.a.

For releases from Unit 2 to the Millstone Stack, the allocated portion of the (

site instantaneous release rate limit is 72,000 [tCi/sec. This assumes that 13%

of the site limit is assigned to Unit 2 releases to the Millstone Stack. If effluent conditions from Unit 2 releases to the Millstone Stack reach 72,000 [tCi/sec, releases from Units 1, 2, and 3 vents and from Unit 3 releases to the Millstone Stack shall be determined to ensure that the sum of the individual noble gas release rates do not cause the site limit to be exceeded.

Use Section II.D.l.a. and Section 4.2 of MP-22-REC-REF03, "REMODCM Technical Information Document,"

in making this determination.

The alarm setpoint shall be set at or below the monitor reading in [tCi/cc GD corresponding to the Unit 2 release to the, stack portion of the limit. The setpoint shall be set at or below 1.3E-2 IACi/cc.

3. Reserved

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4. Unit 3 SLCRS - HVR-RE19B The instantaneous release rate limit from the site shall be set in accordance with the conditions given in Section II.D.l.a. in order to satisfy Unit 3 Radiological Effluent Controls VC.2. and V.D.2.a.

For releases from Unit 3 to the Millstone Stack, the allocated portion of the 0 site instantaneous release rate limit is 72,000 [tCi/sec. This assumes that 13%

of the site limit is assigned to Unit 3 releases to the Millstone Stack. If effluent conditions from Unit 3 releases to the Millstone Stack reach 72,000 [tCi/sec, releases from Units 1 and 2 vents, Unit 3 ESF vent and from Unit 2 releases to the Millstone Stack shall be determined to ensure that the sum of the individual noble gas release rates do not cause the site dose limit to be exceeded. Use Section II.D.l.a. and Section 4.2 of MP-22-REC-REF03, "REMODCM Technical Information Document,"

in making this determination.

The alarm setpoint shall be set at or below the monitor reading in tCi/cc corresponding to the Unit 3 release to the stack portion of the limit. The setpoint shall be set at or below 1.16 E-2 [tCi/cc.

5. Unit 2 Vent - Noble Gas Monitor - RM8132B The instantaneous release rate limit from the site shall be set in accordance with the conditions given in Section II.D.l.a. in order to satisfy the Unit 2 Radiological Effluent Controls in Sections IV.C.2. and IVD.2.a.

For releases from Unit 2 vent, the allocated portion of the site instantaneous )

release rate limit is 95,000 [tCi/sec. This assumes that 33% of the site limit is assigned to Unit 2 vent releases. If effluent conditions from Unit 2 vent releases reach 95,000 [tCi/sec, releases from Units 1 and 3 vents and from Units 2 and 3 releases to the Millstone Stack shall be determined to ensure that the sum of the individual noble gas release rates do not cause the site limit to be exceeded. Use Section II.D.1.a. and Section 4.2 of MP REC- REF03, "REMODCM Technical Information Document,"

in making this determination.

The alarm setpoint shall be set at or below the monitor reading in cpm corresponding to the Unit 2 vent portion of the limit. The setpoint shall be set at or below 42,000 CPM.

6. Unit 2 Waste Gas Decay Tank Monitor RM9095 Administratively all waste gas decay tank releases are via the Millstone Stack. Unit 2 has a release rate limit to the Millstone Stack of 72,000 pCi/sec I (see the MP REC - REF03, "REMODCM Technical Information Document," Section 4.2 for bases).

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Batch releases of waste gas shall be limited to less than 10% of the Unit 2 releases to the Millstone Stack release rate limits. Therefore, the waste gas (

decay tank monitor setpoint should be set not to exceed 7,200 [tCi/sec.

The MP2 waste gas decay tank monitor (given [tCi/cc per cpm) calibration curve and the tank discharge rate is used to assure that the concentration of gaseous activity being released from a waste gas decay tank does not cause the setpoint of 7,200 [tCi/sec to be exceeded.

7. Unit 3 Vent Noble Gas Monitor - HVR-RE10B The instantaneous release rate limit from the site shall be set in accordance with the conditions given in Section II.D.l.a. in order to satisfy Unit 3 Radiological Effluent Controls in Sections V.C.2. and VD.2.a.

For releases from Unit 3 vent, the allocated portion of the site instantaneous release rate limit is 95,000 [tCi/sec. This assumes that 33% of the site limit is assigned to Unit 3 vent releases. If effluent conditions from Unit 3 vent releases reach 95,000 [tCi/sec, releases from Units 1 and 2 vents, Unit 3 ESF vent and from Units 2 and 3 releases to the Millstone Stack shall be determined to ensure that the sum of the individual noble gas release rates do not cause the site limit to be exceeded. Use Section II.D.l.a. and Section 4.2 of MP-22-REC-REF03, "REMODCM Technical Information Document," in making this determination.

The alarm setpoint shall be set at or below the monitor reading in cpm (

corresponding to the Unit 3 vent portion of the limit. The setpoint shall be set at or below 8.4 x 10-4 [LCi/cc.

8. Unit 3 Engineering Safeguards Building Monitor - HVQ-RE49 The Alarm setpoint shall be set at or below the value of 5.9E-4 [tCi/cc I(

9. Bases for Gaseous Monitor Setpoints Gaseous effluent monitors are provided on atmospheric release pathways to control, as applicable, the release of radioactive materials in gaseous effluents to the environment. The alarm / trip setpoints are calculated to ensure that the alarm / trip function of the monitor will occur prior to exceeding the dose rate limits required by the Technical Specifications (Units 2 and 3) or Radiological Effluent Controls (Sections III. IV, and V) requirements for each unit. Monitor setpoint selection is based on a conservative set of conditions for each release pathway such that the dose rate at any time at and beyond the site boundary from all gaseous effluents from all units on the site will be within the numerical values of the annual dose limits of 10 CFR 20 in Unrestricted Areas. Since the Radiological Effluent Controls are constructed such that the numerical values of the MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 92 of 166

annual dose limits of 10 CFR 20 be applied on an instantaneous basis (i.e.,

no time averaging over the year), and the integrated dose objectives of 10 CFR 50, Appendix I provide for corrective actions to reduce effluents if the ALARA dose values are exceeded, assurance is obtained that compliance with the revised annual dose limits of 10 CFR 20.1301 (100 mrem total effective dose equivalent to a member of the public) will also be met. The use of the stated instantaneous release rate values, which equate to the site dose rate limits, also provides operational flexibility to accommodate short periods of higher than normal effluent releases that may occur during plant operations.

APPENDIX II.A REMODCM METHODOLOGY CROSS-REFERENCES Radiological effluent controls (Sections III, IV, and V) identify the requirements for monitoring and limiting liquid and gaseous effluents releases from the site such the resulting dose impacts to members of the public are kept to "As Low As Reasonably Achievable" (ALARA). The demonstration of compliance with the dose limits is by calculational models that are implemented by Section II of the REMODCM.

Table App. II.A- 1 provides a cross-reference guide between liquid and gaseous effluent release limits and those sections of the REMODCM, which are used to determine compliance. It also shows the administrative Technical Specifications which reference the REMODCM for operation of radioactive waste processing equipment. This table also provides a quick outline of the applicable limits or dose objectives and the required actions if those limits are exceeded. Details of the effluent control requirements and the implementing sections of the REMODCM should be reviewed directly for a full explanation of the requirements.

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Table II.A.-1 Millstone Effluent Requirements and Methodology Cross Reference Radiological REMODCM Applicable Limit or Exposure Required Action Effluent Controls & Methodology Objective Period Technical Section Specifications IV/V.E.I.a Tables l.C.-2 Ten times 10CFR20App.B, Instantaneous Restore concentration towithin lim-4 Liquid Effluent and I.C.-3 Table 2, Column 2, & 2x10- its within 15 mins.

Concentration ptCi/mL for dissolved noble gases*

IV/V.E.I.b II.C.1. !s 1.5 mrem TB. Calendar Quar- 30-day report if exceeded. Relative Dose- Liquids II.C.2. -*5 mrern Organ ter** accuracy or conservatism of the cal-II.C.3.

• 3 mrem TB. Calendar Year culations shall be confirmed by per-1I.C.4. *< 10 mrem Organ formance of the REMP in Section L.

TS. 6.16 (Unit 2) I.C.2.

• 0.06 mrem TB. Projected for 31 Return to operation Liquid Waste T.S. 6.14 (Unit 3) II.C.5. _*0.2 mrem Organ days (if system Treatment System Liquid Radwaste Treat- not in use) ment Ill.D.2.a Tables I.D.- 1, *5 500 mrem/yr TB. from Instantaneous Restore release rates to within spec-IV/V.D.2.a I.D.-2, & noble gases* ifications within 15 minutes Gaseous Effluents Dose I.D-3 Rate II.D.1.a.  :!53000 mremyr skin from noble gases*

II.D.l.b.

• 1500 mrem/yr organ from particulates with Tip > 8d.,

1-131, 1-133 & tritium*

IlI.D.2.b II.D.2.

• 5 mrad gamma air Calendar Quar- 30-day report if exceeded IV/V.D.2.bDose Noble *_ 10 mrad beta air ter**

Gases G s10 mrad gamma air Calendar Year

• _20 mrad beta air III.D.2.c ll.D.3.

• 7.5 mrem organ Calendar Quar- 30-day report if exceeded. Relative 1V/VD.2.c ter** accuracy or conservatism of the cal-Dose 1-131, 1-133, Par- culations shall be confirmed by per-ticulates, H-3 < 15 mrem organ Calendar Year formance of the REMP in Section 1.

TS. 5.6.4 (Unit 1) II.D.2. > 0.02 mrad gamma air Projected for 31 Return to operation Gaseous Rad-T.S. 6.14 (Unit 2) 11.D.4. > 0.04 mrad beta air Days (if system waste Treatment System TS 6.16 (Unit 3) > 0.03 mrem organ not in use)

Gaseous Radwaste Treatment IM.E II.D.6. *<25 mrem TB.* 12 Consecutive 30-day report if Unit 1 Effluent IVI N.F _< 25 mrem organ* Months** Control III.D.1.2, lII.D.2.2, or Total Dose s75 mrem thyroid* JII.D.2.3 or Units 2/3 Effluent Con-trol IV/V.E.1.2, IV/V.E.2.2, or IV/

V.E.2.3 aie exceeded by a factor of

2. Restore dose to public to within the applicable EPA limit(s) or ob-tain a variance NOTE: T.B. means total or whole body.

• Applies to the entire site (Units 1, 2, and 3) discharges combined.

• • Cumulative dose contributions calculated once per 31 days.

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SECTION III.

Millstone Unit 1 Radiological Effluent Controls Docket Nos. 50-245 MP-22-REC-BAP01 STOP THINK .ACT REVIEw Rev 026- 01 95 of 166

SECTION III. REMODCM UNIT ONE CONTROLS III.A. Introduction The purpose of this section is to provide the following for Millstone Unit One:

a. the effluent radiation monitor controls and surveillance requirements,

b. the effluent radioactivity concentration and dose controls and surveillance requirements, and

c. the bases for the controls and surveillance requirements.

Definitions of certain terms are provided as an aid for implementation of the controls and requirements.

Some surveillance requirements refer to specific sub-sections in Sections I and II as part of their required actions III.B. Definitions and Surveillance Requirement (SR) Applicability III.B.1 - Definitions The defined terms of this sub-section appear in capitalized type and are applicable throughout Section III.

1. ACTION - that part of a Control that prescribes remedial measures required under designated conditions.

2. INSTRUMENT CALIBRATION - the adjustment, as necessary, of the instrument output such that it responds within the necessary range and accuracy to know values of the parameter that the instrument monitors. The INSTRUMENT CALIBRATION shall encompass those components, such as sensors, displays, and trip functions, required to perform the specified safety function(s). The INSTRUMENT CALIBRATION shall include the INSTRUMENT FUNCTIONAL TEST and may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

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-STOP THINK ACT REVIEW Rev. 026-01 96 of 166

3. INSTRUMENT FUNCTIONAL TEST - the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify that the instrument is OPERABLE, including all components inthe channel, such as alarms, interlocks, displays, and trip functions, required to perform the specified safety function(s). For digital instruments, the computer database may be manipulated, in lieu of a signal injection, to verify operability of alarm and/or trip functions. The INSTRUMENT FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is tested.

4. INSTRUMENT CHECK - the qualitative determination of operability by observation of behavior during operation. This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

5. OPERABLE - An instrument shall be OPERABLE when it is capable of performing its specified functions(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the instrument to perform its functions(s) are also capable of performing their related support function(s).

6. REAL MEMBER OF THE PUBLIC - an individual, not occupationally associated with the Millstone site, who is exposed to existing dose pathways at one particular location. This does not include employees of the utility or utilities which own a Millstone plant and utility contractors and vendors.

Also excluded are persons who enter the Millstone site to service equipment or to make deliveries. This does include persons who use portions of the Millstone site for recreational, occupational, or other purposes not associated with any of the Millstone plants.

7. SITE BOUNDARY - that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

8. SOURCE CHECK - the qualitative assessment of channel response when the channel is exposed to radiation.

9. RADIOACTIVE WASTE TREATMENT SYSTEMS - Radioactive Waste Treatment Systems are those liquid, gaseous, and solid waste systems which are required to maintain control over radioactive materials in order to meet the controls set forth in this section.

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III.B.2 - Surveillance Requirement (SR) Applicability

1. SRs shall be met during specific conditions in the Applicability for individual LCOs unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in III.B.2 3.

Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

2. The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the frequency is met.

3. If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the LCO not met may be delayed from the time of discovery up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified frequency, whichever is less. This delay period is permitted to allow performance of the surveillance. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met and the applicable Condition(s) must be entered. The Completion Times of the Required Actions begin immediately upon expiration of the delay period. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met and the applicable Condition(s) must be entered. The Completion Times of the Required Actions begin immediately upon failure to meet the Surveillance.

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III.C. Radioactive Effluent Monitoring Instrumentation

1. Radioactive Liquid Effluent Monitoring Instrumentation CONTROLS The radioactive liquid effluent monitoring instrumentation channels shown in Table III.C.- 1 shall be OPERABLE with applicable alarm/trip setpoints set to ensure that the limits of Specification III.D.l.a. are not exceeded. The setpoints shall be determined in accordance with methods and parameters described in Section II.

APPLICABILITY: As shown in Table III.C.- 1 ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels OPERABLE requirement, take the action shown in Table III.C.- 1.

Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Releases need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENTS Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table III.C.-2.

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TABLE III.C.-1 Radioactive Liquid Effluent Monitoring Instrumentation Instrument 1

1.Radioactivity Monitor

• Liquid Effluent Line

• Whenever the pathway is being used except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling.

Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases may continue provided that best efforts are made to repair the instrument and that prior to initiating a release:

(1) At least two independent samples are analyzed in accordance with the first Surveillance Requirement of Specification II.D.1 .a. and; (2) The original release rate calculations and discharge valving are independently verified by a second individual.

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TABLE III.C.-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Instrument Channel Source Channel Channel Check Check Calibration Functional Test 1.Radioactivity Monitor Liquid Effluent D* P T(1) Q Line D = Daily P = Prior to each batch release T = Once every two years Q = Once every 3 months During releases via this pathway and when the monitor is required OPERABLE per Table 1I1.C.-1. The CHANNEL CHECK should be done when the discharge is in progress.

(1) Calibration shall includethe use of a radioactive liquid or solid source which is traceable to an NIST source.

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2. Radioactive Gaseous Effluent Monitoring Instrumentation CONTROLS The radioactive gaseous effluent monitoring instrumentation channels shown in Table III.C.-3 shall be OPERABLE with applicable alarm setpoints set to ensure that the limits of Control III.D.2.a. are not exceeded. The setpoints shall be determined in accordance with methods and parameters described in Section II.E1.

APPLICABILITY: As shown in Table III.C.-3 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm setpoint less conservative than required by the above Control, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels operable requirements, take the action shown in Table III.C.-3. Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radiological Effluent Release Report why the inoperability was not corrected in a timely manner. Release need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENT Each r'adioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the INSTRUMENT CHECK, INSTRUMENT CALIBRATION, INSTRUMENT FUNCTIONAL TEST, and SOURCE CHECK operations at the frequencies shown in Table III.C.-4.

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TABLE III.C.-3 Radioactive Gaseous Effluent Monitoring Instrumentation Instrument Minimum Alarm Setpoints Applicability Action

1. Operable Required 1.Spent Fuel Pool Island Vent (a) Noble Gas Activity Monitor 1 Yes

• A (b) Particulate Sampler 1 No

• B (c) Vent Flow Rate Monitor 1 No

• C (d) Sampler Flow Rate Monitor 1 Yes

• D 2.Balance of Plant Vent (a) Particulate Sampler 1 No

• B (b) Sampler Flow Monitor 1 Yes D

• Channels are OPERABLE and in service on a continuous, uninterrupted basis when exhaust fans are operating, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required tests, checks, calibrations, and sampling associated with the instrument or any system or component which affects functioning of the instrument.

ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that grab samples are taken daily when fuel is being moved, or during any evolution or event which would threaten fuel integrity, and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Action B With the number of samplers OPERABLE less than required by the Minimum number OPERABLE requirement, effluent releases via this pathway may continue provided that the best efforts are made to repair the instrument and that a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> sample is collected with auxiliary sampling equipment once every seven (7) days, or anytime significant generation of airborne radioactivity is expected, and analyzed for principal gamma emitters with half lives greater than 8 days within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the end of the sampling period. Operation of the auxiliary sampling equipment shall be verified every twelve (12) hours. I Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument.

Action D With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated once during the MP-22-REC-BAPO1 STOP THINK ACT REVIEW Rev. 026-01 103 of 166

Chemistry compensatory sampling time period as specified in Action A or Action B. Sample flow rate need not be estimated ifthe auxiliary sampling equipment of Action B is in use. I MP-22-REC-BAP01 StOP " THINK ACT REVIEW Rev. 026-01 104 of 166

TABLE III.C.-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Instrument Instrument Instrument Functional Source Check Calibration Test Check 1.Spent Fuel Pool Island Vent (a) Noble Gas Activity Monitor D(3 ) T(6) Q(7) M (b) Particulate Sampler TM NA NA NA (c) Vent Flow Rate Monitor D T NA NA (d) Sampler Flow Rate Monitor D T NA NA 2.Balance of Plant Vent (a) Particulate Sampler TM NA NA NA (b) Sampler Flow Monitor D T NA NA D = Daily W = Weekly TM = Twice per month M = Monthly Q = Once every 3 months T Once every two years NA= Not Applicable Table IIl.C.-4 TABLE NOTATION (1) RESERVED (2) RESERVED (3) Instrument check daily only when there exist releases via this pathway.

(4) RESERVED (5) RESERVED (6) Calibration shall include the use of a known source whose strength is determined by a detector which has been calibrated to a source which is traceable to the NIST. These sources shall be in a known reproducible geometry.

(7) The INSTRUMENT FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Instrument indicates a downscale failure.

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III.D. Radioactive Effluents Concentrations And Dose Limitations

1. Radioactive Liquid Effluents

a. Radioactive Liquid Effluents Concentrations LIMITING CONDITIONS OF OPERATIONS The concentration of radioactive material released from the site (see Figure III.D. -1) shall not exceed ten times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall not exceed 2 x 1 0 -4 [tCi/ml total activity.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released from the site exceeding the above limits, restore the concentration to within the above limits within 15 minutes.

SURVEILLANCE REQUIREMENT Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Section I.

The results of the radioactive analysis shall be used in accordance with the methods of Section II to assure that the concentrations at the point of release are maintained within the limits of Specification III.D.1.a.

MP-22-REC-BAPO1 STOP THINK ACT REVIEW Rev. 026- 01 106 of 166

b. Radioactive Liquid Effluents Doses LIMITING CONDITIONS OF OPERATIONS The dose or dose commitment to any REAL MEMBER OF THE PUBLIC from radioactive materials in liquid effluents from Unit 1 released from the site (see Figure III.D. -1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ; and,

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 3 mrem to the total body and 10 mrem to any organ.

SURVEILLANCE REOUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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2. Radioactive Gaseous Effluents

a. Radioactive Gaseous Effluents Dose Rate CONTROLS The dose rate, at any time, offsite (See Figure III.D.-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

a. The dose rate limit for noble gases shall be less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin; and,

b. The dose rate limit for Tritium and for all radioactive materials in particulate form with half lives greater than 8 days shall be less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, decrease the release rate to comply with the limit(s) given in Control III.D.2.a. within 15 minutes.

SURVEILLANCE REOUIREMENT

1) The instantaneous release rate corresponding to the above dose rate shall be determined in accordance with the methodology of Section II.

2) The instantaneous release rate shall be monitored in accordance with the requirements of Section III.C.2.

3) Sampling and analysis shall be performed in accordance with Section I to assure that the limits of Control III.D.2.a. are met.

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b. Radioactive Gaseous Effluents Noble Gas Dose CONTROLS The air dose offsite (see Figure III.D. -1) due to noble gases released in gaseous effluents from Unit 1 shall be limited to the following:

a. During any calendar quarter, to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation;

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

APPLICABILITY: At all times.

ACTION:

With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and the calendar year so that the cumulative dose during the calendar year is within 10 mrad for gamma radiation and 20 mrad for beta radiation.

SURVEILLANCE REQUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section 1.

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c. Gaseous Effluents - Dose from Radionuclides Other than Noble Gas CONTROLS The dose to any REAL MEMBER OF THE PUBLIC from Tritium and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents released offsite from Unit 1 (see Figure III.D.-1) shall be limited to the following:

a. During any calendar quarter to less than or equal to 7.5 mrem [to any organ];

b. During any calendar year to less than or equal to 15 mrem [to any organ].

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of Tritium and radioactive materials in particulate form exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 15 mrem to any organ.

SURVEILLANCE REOUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be' confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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Figure III.D.-1, "Site Boundary for Liquid and Gaseous Effluents" MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 111 of 166

III.E. Total Radiological Dose From Station Operations Controls CONTROLS The annual dose or dose commitment to any REAL MEMBER OF THE PUBLIC, beyond the site boundary, from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls III.D.l.b., III.D.2.b. or III.D.2.c. prepare and submit a Special Report to the Commission within 30 days and limit the subsequent releases such that the dose commitment from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months.

This Special Report shall include an analysis which demonstrates that radiation exposures from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site (including all effluent pathways and direct radiation) are less than the 40 CFR 190 Standard.

If the estimated doses exceed the above limits, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190.

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

SURVEILLANCE REQUIREMENTS Cumulative dose contributions from liquid and gaseous effluents and direct radiation from the Millstone Site shall be determined in accordance with Section II once per 31 days.

III.E BasesSection III.C.1 - Radioactive Liquid Effluent Monitoring Instrumentation No controls required; Unit 1 is not currently releasing radioactivity in liquid effluentsSection III.C.2 - Radioactive Gaseous Effluent Monitoring Instrumentation MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 112 of 166

The Spent Fuel Pool Island Vent is the only gaseous pathway currently requiring radiation monitoring for Unit 1.

Section III.D.l.a. - Radioactive Liquid Effluents Concentrations This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within: (1) the Section II.A. design objectives of Appendix I, 10 CFR 50, to an individual and (2) the limits of 10 CFR 20 to the population. The concentration limit for noble gases is based upon the assumption that Xe- 135 is the controlling radioisotope and its concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

Section III.D.l.b. - Radioactive Liquid Effluents Doses This specification is provided to implement the requirements of Sections II.A.,

III.A, and IVA of Appendix I, 10 CFR 50. The specification implements the guides set forth in Section II.A of Appendix I. The Action statements provide the required operating flexibility and at the same time implement the guides set forth in Section III.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

Section III.D.2.a. - Radioactive Gaseous Effluents Dose Rate This control is provided to ensure that the dose rate at anytime from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR 20 for all areas offsite. The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table 2. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual offsite to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR MP-22-REC-BAP01 STOP THIN.K ACT' REVI EW Rev. 026,--01 113 of 166

20. For individuals who may, at times, be within the site boundary, the occupancy of that individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to less than or equal to 1500 mrem/year for the nearest cow to the plant.

Section III.D.2.b. - Radioactive Gaseous Effluents Noble Gas Dose This control is provided to implement the requirements of Sections II.B., III.A.,

and IVA. of Appendix I, 10 CFR 50. The control implements the guides set forth in Section II.B of Appendix I. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IVA of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculational of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.

The ODCM equations provided for determining the air doses at the site boundary were based upon utilizing successively more realistic dose calculational methodologies. More realistic dose calculational methods are used whenever simplified calculations indicate a dose approaching a substantial portion of the regulatory limits. The methods used are, in order, previously determined air dose per released activity ratio, historical meteorological data and actual radionuclide mix released, or real time meteorology and actual radionuclides released.

Section III.D.2.c. - Radioactive Gaseous Effluents, Particulates, and Gas Other Than Noble Gas Doses These controls is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR 50. The controls are the guides set forth in Section II.C of Appendix I. The action statements provide the required MP-22-REC-BAPO1 STOP THINK ACT REVIEW Rev. 026-01 114 of 166

operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides for Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials will to be consistent with the methodology provided in Regulatory Guide 1.109, "Calculating of Annual Dose to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I,"

Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision I, July 1977. These equations provide for determining the doses based upon either conservative atmospheric dispersion and an assumed critical nuclide mix or using real time meteorology and specific nuclides released. The release rate specifications for radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

Section III.E. - Total Radiological Dose from Station Operations This control is provided to meet the reporting requirements of 40 CFR 190. For the purpose of the Special Report, it may be assumed that the dose commitment to any REAL MEMBER OF THE PUBLIC from other fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered.

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SECTION IV Millstone Unit 2 Radiological Effluent Controls Docket Nos. 50-336 MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 J116 of 166

SECTION IV. REMODCM UNIT TWO CONTROLS IV.A. Introduction The purpose of this section is to provide the following for Millstone Unit Two:

a. the effluent radiation monitor controls and surveillance requirements,

b. the effluent radioactivity concentration and dose controls and surveillance requirements, and

c. the bases for the controls and surveillance requirements.

Definitions of certain terms are provided as an aid for implementation of the controls and requirements.

Some surveillance requirements refer to specific sub-sections in Sections I and II as part of their required actions.

IVB. Definitions, Applicability and Surveillance Requirements IV.B.1 - Definitions The defined terms of this sub-section appear in capitalized type and are applicable throughout Section IV.

1. ACTION - Those additional requirements specified as corollary statements to each principal control and shall be part of the control.

2. OPERABLE / OPERABILITY - An instrument shall be OPERABLE or have OPERABILITY when it is capable of performing its specified functions(s) and when all necessary attendant instrumentation, controls, normal and emergency electrical power sources, or other auxiliary equipment that are required for the instrument to perform its functions(s) are also capable of performing their related support function(s).

3. CHANNEL CALIBRATION - A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to know values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

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4. CHANNEL CHECK - A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

5. CHANNEL FUNCTIONAL TEST - A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions. For digital instruments, the computer database may be manipulated, in lieu of a signal injection, to verify operability of alarm and/or trip functions.

6. SOURCE CHECK - A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

7. MEMBER(S) OF THE PUBLIC - MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location.

8. MODE - Refers to Mode of Operation as defined in Safety Technical Specifications.

9. SITE BOUNDARY - The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

10. UNRESTRICTED AREA - Any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional and/or recreational purposes, MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 118 of 166

11. DOSE EQUIVALENT 1- 131 - DOSE EQUIVALENT 1- 131 shall be that concentration of 1-131 ([tCi/gram) which alone would produce the same CDE-thyroid dose as the quantity and isotopic mixture of 1-131-, 1-132, 1- 133, 1-134, and 1- 135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under Inhalation in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

IV.B.2 - Applicability IV.B.2a - LIMITING CONDITIONS FOR OPERATION

1. Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

2. Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals, except as provided in Condition IVB.2.a(6). If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3. NOT USED.

4. NOT USED.

5. When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system(s), subsystem(s), train(s), component(s) and device(s) are OPERABLE, or likewise satisfy the requirements of this specification.

6. Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to Condition IVB.2.a(2) for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

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IVB2.b - SURVEILLANCE REQUIREMENTS

1. Surveillance Requirements shall be applicable during any condition specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

2. Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance time interval.

3. Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Condition IV.B2.b(2), shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance-when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.

4. Entry into any specified condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.

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IV.C. Radioactive Effluent Monitoring Instrumentation

1. Radioactive Liquid Effluent Monitoring Instrumentation LIMITING CONDITIONS OF OPERATIONS The radioactive liquid effluent monitoring instrumentation channels shown in Table IV.C.- 1 shall be OPERABLE with applicable alarm/trip setpoints set to ensure that the limits of Specification IV.D.1.a. are not exceeded. The setpoints shall be determined in accordance with methods and parameters described in Section II.

APPLICABILITY: As shown in Table IVC.- I ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels OPERABLE requirement, take the action shown in Table IV.C.- 1.

Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Releases need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENTS Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table IV.C.-2.

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TABLE IV.C.-1 Radioactive Liquid Effluent Monitoring Instrumentation Instrument Minimum Alarm Setpoints Applicability Action

1. Operable I Required 1.Gross Radioactivity Monitors Providing Automatic Termination Of Release (a) Clean Liquid Radwaste Effluent 1 Yes

• A Line (b)Aerated Liquid Radwaste Effluent 1 Yes *** A Line (c) Steam Generator Blowdown 1 Yes **** B Monitor (d)Condensate Polishing Facility 1 Yes *** E Waste Neut Sump 2.Gross Radioactivity Monitors Not Providing AutomaticTermination Of Release (a) Reactor Building Closed Cooling 1 Yes

• C Water Monitor#

3.Flow Rate Measurements (a) Clean Liquid Radwaste Effluent 1 No D Line (b)Aerated Liquid Radwaste No D Effluent Line (c) Condensate Polishing Facility 1 No D TABLE IV.C.- 1 TABLE NOTES At all times - which means that channels shall be OPERABLE and in service on a continuous, uninterrupted basis, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

• • Deleted.

Whenever the pathway is being used except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

MODEs 1 -4, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

1. Since the only source of service water contamination is the reactor building closed cooling water, monitoring of the closed cooling water and conservative leakage assumptions will provide adequate control of service water effluents.

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ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases may continue provided that best efforts are made to repair the instrument and that prior to initiating a release:

(1) At least two independent samples are analyzed in accordance with the first Surveillance Requirement of Specification IV.D.1 .a. and; (2) The original release rate calculations and discharge valving are independently verified by a second individual.

Action B With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, either:

(1) Suspend all effluent releases via this pathway, or (2) Make best efforts to repair the instrument and obtain grab samples and analyze for gamma radioactivity at lower limits of detection as specified in Table I.C.-2; a) Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 [tCi/gm DOSE EQUIVALENT 1-131.

b) Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 0Ci/gm DOSE EQUIVALENT 1-131.

Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that once per .12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples of the service water effluent are collected and analyzed for gamma radioactivity at LLD as specified in Table I.C.-2; Action D With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves may be used to estimate flow.

Action E With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases may continue provided that best efforts are made to repair the instrument and that prior to initiating a release:

(1) At least two independent samples are analyzed in accordance with the first Surveillance Requirement of Specification IV.D.1 .a., and; (2) If one of the samples has gamma radioactivity greater than any of the LLDs in Table I.C.-2, the original release rate calculations and discharge valving are independently verified by a second individual.

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TABLE IV.C.-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Instrument Channel Source Channel Channel Check Check Calibration Functional Test 1.Gross Radioactivity Monitors Providing Alarm and Automatic Termination Of Release

a. Clean Liquid Radwaste Effluent Line D* P R(1) Q(2)

b. Aerated Liquid Radwaste Effluent D* P R(1) Q(2)

Line

c. Steam Generator Blowdown Monitor D* M R(1) Q(2) d.Condensate Polishing Facility Waste D* P R(1) Q(2)

Neut Sump 2.Gross Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination Of Release

a. Reactor Building Closed Cooling D* M R(1) 0(2)

Water Monitor I 3.Flow Rate Measurements

a. Clean Liquid Radwaste Effluent Line

b. Aerated Liquid Radwaste Effluent Line

c. Condensate Polishing Facility Waste Neut Sump D = Daily R = Once every 18 months M = Monthly 0 = Once every 3 months P = Prior to each batch release N/A= Not Applicable TABLE IV.C.-2 TABLE NOTATION During releases via this pathway and when the monitor is required OPERABLE per Table IV.C.-1. The CHANNEL CHECK should be done when the discharge is in progress.

(1) Calibration shall include the use of a radioactive liquid or solid source which is traceable to an NIST source.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

a) Instrument indicates measured levels above the alarm/trip setpoint.

b) Instrument indicates a downscale or circuit failure.

- Automatic isolation of the discharge stream shall also be demonstrated for this case for each monitor except the reactor building closed cooling water monitor.

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2. Radioactive Gaseous Effluent Monitoring Instrumentation LIMITING CONDITIONS OF OPERATIONS The radioactive gaseous effluent monitoring instrumentation channels shown in Table IV.C.-3 shall be OPERABLE with applicable alarm/trip setpoints set to ensure that the limits of Specifications IV.D.2.a. are not exceeded. The setpoints shall be determined in accordance with methods and parameters described in Section II.

APPLICABILITY: As shown in Table IVC.-3 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels OPERABLE requirement, take the action shown in Table IV.C.-3.

Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Release need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENTS Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table IVC.-4.

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TABLE IV.C.-3 Radioactive Gaseous Effluent Instrumentation Instrument Minimum Alarm Applicability Action Channels Setpoints Operable Required 1.MP2 Vent (normal range, RM-8132 only; high range monitor, RM-8168, requirements are in the TS)

a. Noble Gas Activity Monitor 1 Yes*** ** A

b. Iodine Sampler 1 No ** B

c. Particulate Sampler 1 No ** B

d. Vent Flow Rate Monitor 1 No C

e. Sampler Flow Rate Monitor 1 No *C*

2.Millstone Stack - applicable to the WRGM (RM-8169, normal range, channel 1,only; mid range channel 2 and high range channel 3 requirements are contained in TRM LCO 3.3.3.8)

a. Noble Gas Activity Monitor 1 Yes***

• E

b. Iodine Sampler 1 No B

c. Particulate Sampler 1 No B

d. Stack Flow Rate Monitor 1 No

• C

e. Sampler Flow Rate Monitor 1 No ** C 3.Waste Gas Holdup System

a. Noble Gas Monitor Providing 1 Yes

• D Automatic Termination of Release

• During waste gas holdup system discharge.

• • At all times when air is being released to the environment by the pathway being monitored, which means that channels be OPERABLE and in service on a continuous, uninterrupted basis, except that outages are permitted for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of maintenance and performance of required tests, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

• No automatic isolation features.

ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that grab samples are taken once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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Action B With the number of samplers OPERABLE less than required by the Minimum number OPERABLE requirement, effluent releases via this pathway may continue provided that the best efforts are made to repair the instrument and that samples are continuously collected with auxiliary sampling equipment for periods of seven (7) days and analyzed for principal gamma emitters with half lives greater than 8 days within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the end of the sampling period. Auxiliary sampling must be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of initiation of this action statement. Operation of the auxiliary sampling equipment shall be verified every twelve (12) hours. Auxiliary sampling outages are permitted for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of maintenance and performance of required tests, checks, calibrations, or sampling.

Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated once per 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Sample flow rate need not be estimated if the auxiliary sampling equipment of Action B is in use.

Action D With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement:

Releases from the Millstone Unit 2 waste gas system may continue provided that best efforts are made to repair the instrument and that prior to initiating the release:

a) At least two independent samples of the tank's contents are analyzed; and b) The original release rate calculations and discharge valve lineups are independently verified by a second individual. Otherwise, suspend releases from the waste gas holdup system.

Action E With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, Millstone Unit 2 releases via the Millstone Stack may continue provided that best efforts are made to repair the instrument and that grab samples are taken once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 127 of 166

TABLE IV.C.- 4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Instrument Channel Source Channel Channel Check Check Calibration Functional

_ I Test 1.MP2 Vent (normal range, RM-8132 only; high range monitor, RM-8168, requirements are in the TS)

a. Noble Gas Activity Monitor D M R(1) Q(2)

b. Iodine Sampler W NA NA NA

c. Particulate Sampler W NA NA NA

d. Vent Flow Rate Monitor D NA R Q

e. Sampler Flow Rate Monitor D NA R NA 2.Millstone Stack - applicable to the WRGM (RM-8169, normal range, channel 1,only; mid range channel 2 and high range channel 3 requirements are contained in TRM LCO 3.3.3.8)

a. Noble Gas Activity Monitor D M R(1) Q( 2 )

b. Iodine Sampler W NA NA NA

c. Particulate Sampler W NA NA NA

d. Stack Flow Rate Monitor D NA R Q(2)

e. Sampler Flow Rate Monitor D NA R NA 3.Waste Gas Holdup System

a. Noble Gas Monitor 1D* P R( 1) Q(2).

• During releases via this pathway and when the monitor is required OPERABLE per Table IV.C.-3.

The CHANNEL CHECK should be performed when the discharge is in progress.

P = Prior to discharge R = Once every 18 months D = Daily Q = Once every 3 months W = Weekly NA= Not Applicable M = Monthly TABLE IV.C. - 4 TABLE NOTATION (1) Calibration shall include the use of a known source whose strength is determined by a detector which has been calibrated to a source which is traceable to the NIST. These sources shall be in a known, reproducible geometry.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation* occurs ifany of the following conditions exist:

a) Instrument indicates measured levels above the alarm/trip setpoint.

b) Instrument indicates a downscale failure.

• - Also demonstrate automatic isolation for the waste gas system noble gas monitor.

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IV.D. Radioactive Effluents Concentrations And Dose Limitations

1. Radioactive Liquid Effluents

a. Radioactive Liquid Effluents Concentrations LIMITING CONDITIONS OF OPERATIONS The concentration of radioactive material released from the site (see Figure IV.D.-1) shall not exceed ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrainednoble gases. For dissolved or entrained noble gases, the concentration shall not exceed 2 x 104 ýtCi/ml total activity.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released from the site exceeding the above limits, restore the concentration to within the above limits within 15 minutes.

SURVEILLANCE REOUIREMENTS

1) Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Section I.

2) The results of the radioactive analysis shall be used in accordance with the methods of Section II to assure that the concentrations at the point of release are maintained within the limits of Specification IV.D.I.a.

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b. Radioactive Liquid Effluents Doses LIMITING CONDITIONS OF OPERATIONS The dose or dose commitment to any REAL MEMBER OF THE PUBLIC from radioactive materials in liquid effluents from Unit 2 released from the site (see Figure IV.D. -1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ; and,

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 3 mrem to the total body and 10 mrem to any organ.

SURVEILLANCE REOUIREMENTS

1) Dose Calculations. Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in Section II at least once per 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

MP-22-REC-BAPO1 STOP THNK ACT' REVIEW Rev. 026-01 130 of 166

2. Radioactive Gaseous Effluents

a. Radioactive Gaseous Effluents Dose Rate LIMITING CONDITIONS OF OPERATIONS The dose rate, at any time, offsite (see Figure IV.D.-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

a. The dose rate limit for noble gases shall be less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin; and,

b. The dose rate limit for Iodine- 131, Iodine- 133, Tritium, and for all radioactive materials in particulate form with half lives greater than 8 days shall be less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, decrease the release rate to comply with the limit(s) given in Specification IVD.2.a. within 15 minutes.

SURVEILLANCE REOUIREMENTS

1) The release rate, at any time, of noble gases in gaseous effluents

.shall be controlled by the offsite dose rate as established above in Specification IVD.2.a. The corresponding release rate shall be determined in accordance with the methodology of Section II.

2) The noble gas effluent monitors of Table IV.C. -3 shall be used to control release rates to limit offsite doses within the values established in Specification IV.D.2.a.

3) The release rate of radioactive materials in gaseous effluents shall be determined by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Section I. The corresponding dose rate shall be determined using the methodology given in Section II.

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b. Radioactive Gaseous Effluents Noble Gas Dose LIMITING CONDITIONS OF OPERATIONS The air dose offsite (see Figure IV.D.- 1) due to noble gases released in gaseous effluents from Unit 2 shall be limited to the following:

a. During any calendar quarter, to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation;

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and the calendar year so that the cumulative dose during the calendar year is within 10 mrad for gamma radiation and 20 mrad for beta radiation.

SURVEILLANCE REOUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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c. Gaseous Effluents - Doses from Radionuclides Other than Noble Gas LIMITING CONDITIONS OF OPERATIONS The dose to any REAL MEMBER OF THE PUBLIC fromIodine-131, Iodine -133, Tritium, and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents released offsite from Unit 2 (see Figure IV.D.-1) shall be limited to the following:

a. During any calendar quarter to less than or equal to 7.5 mrem to any organ;

b. During any calendar year to less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 15 mrem to any organ.

SURVEILLANCE REOUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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Figure IV.D.- 1, "Site Boundary for Liquid and Gaseous Effluents" MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-0*1 134 of 166

IV.E. Total Radiological Dose From Station Operation CONTROLS The annual dose or dose commitment to any REAL MEMBER OF THE PUBLIC, beyond the site boundary, from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls IVD.2.a., IV.D.1.b., or IVD.2.c. prepare and submit a Special Report to the Commission within 30 days and limit the subsequent releases such that the dose commitment from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months.

This Special Report shall include an analysis which demonstrates that radiation exposures from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site (including all effluent pathways and direct radiation) are less than the 40 CFR 190.

If the estimated doses exceed the above limits, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190.

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

SURVEILLANCE REQUIREMENTS Cumulative dose contributions from liquid and gaseous effluents and direct radiation from the Millstone Site shall be determined in accordance with Section II once per 31 days.

IV.E BasesSection IV.C.1. - Radioactive Liquid Effluent Monitoring Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated in accordance with the approved methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding ten times the limits of 10 CFR 20. The OPERABILITY and use of this MP-22-REC-BAPO1 STOP THINK .ACT REVIEW Rev. 026-01 135 of 166

instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50. Monitoring of the turbine building sumps and condensate polishing facility floor drains is not required due to relatively low concentrations of radioactivity possible.

Section IV.C.2. - Radioactive Gaseous Effluent Monitoring Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with the approved methods in the REMODCM to ensure that the alarm/trip will occur prior to exceeding the dose rate limits, at any time, as specified in Section IV.D.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50.

Two types of radioactive gaseous effluent monitoring instrumentation, monitors and samplers, are being used at MP2 vent and Millstone Stack. Monitors have alarm/trip setpoints and are demonstrated operable by performing one or more of the following operations: CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST. Samplers are strictly collection devices made of canisters and filters. The CHANNEL CHECK surveillance requirements are met through (1) documented observation of the in-service rad monitor sample flow prior to filter replacement; (2) documented replacement of in-line iodine and particulate filters; and (3) documented observation of sample flow following the sampler return to service.

The flow indicator is the only indication available for comparison. These observations adequately provide assurance that the sampler is operating and is capable of performing its design function.

There are a number, of gaseous release points which could exhibit very low concentrations of radioactivity. For all of these release paths, dose consequences would be insignificant due to the intermittent nature of the release and/or the extremely low concentrations of radioactivity. Since it is not cost-beneficial (nor in many cases practical due to the nature of the release (steam) or the impossibility of detecting such low levels), to monitor these pathways, it has been determined that these release paths require no monitoring or sampling.

Section IVD.1.a. - Radioactive Liquid Effluents Concentrations This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within: (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to an individual and (2) the limits of 10 CFR 20 to the population. The concentration limit for noble gases is MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 136 of 166

based upon the assumption that Xe- 135 is the controlling radioisotope and its concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

Section IV.D.I.b. - Radioactive Liquid Effluents Doses This specification is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR 50. The specification implements the guides set forth in Section II.A of Appendix I. The Action statements provide the required operating flexibility and at the same time implement the guides set forth in Section III.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that. the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

Section IV.D.2.a..- Radioactive Gaseous Effluents Dose Rate This specification is provided to ensure that the dose rate at anytime from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR 20 for all areas offsite. The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table 2. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual offsite to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR

20. For individuals who may, at times, be within the site boundary, the occupancy of that individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid or any other organ dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year.

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SECTION V Millstone Unit 3 Radiological Effluent Controls Docket Nos. 50-423 MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 138 of 166

SECTION V REMODCM UNIT THREE CONTROLS V.A. Introduction The purpose of this section is to provide the following for Millstone Unit Three:

a. the effluent radiation monitor controls and surveillance requirements,

b. the effluent radioactivity concentration and dose controls and surveillance requirements, and

c. the bases for the controls and surveillance requirements.

Definitions of certain terms are provided as an aid for implementation of the controls and requirements.

Some surveillance requirements refer to specific sub-sections in Sections I and II as part of their required actions.

V.B. Definitions and Applicability and Surveillance Requirements V.B.1 - Definitions The defined terms of this sub-section appear in capitalized type and are applicable throughout Section V.

1. ACTION - ACTION shall be that part of the control which prescribes remedial measures required under designated conditions.

2. CHANNEL OPERATIONAL TEST - A CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. For digital instruments, the computer database may be manipulated, in lieu of a signal injection, to verify operability of alarm and/or trip functions.

The CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or trip setpoints such that the setpoints are within the required range and accuracy.

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3. CHANNEL CALIBRATION - A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

4. CHANNEL CHECK - A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

5. DOSE EQUIVALENT 1-131 - DOSE EQUIVALENT 1-131 shall be that concentration of 1- 131 ([tCi/gram) which alone would produce the same CDE-thyroid dose as the quantity and isotopic mixture of 1- 131, 1- 132, 1-133, 1- 134, and 1- 135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under Inhalation in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

6. MEMBER(S) OF THE PUBLIC - MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the licensee, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location.

7. MODE - Refers to Mode of Operation as defined in Safety Technical Specifications.

8. OPERABLE - OPERABILITY - An instrument shall be OPERABLE or have OPERABILITY when it is capable of performing its specified functions(s) and when all necessary attendant instrumentation, controls, electrical power, or other auxiliary equipment that are required for the instrument to perform its functions(s) are also capable of performing their related support function(s).

9. SITE BOUNDARY - The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

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10. SOURCE CHECK - A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

11. UNRESTRICTED AREA - Any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or industrial, commercial, institutional and/or recreational purposes.

V.B.2 - Applicability V.B.2.a - LIMITING CONDITIONS FOR OPERATION

1. Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

2. Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

V.B.2.b - SURVEILLANCE REQUIREMENTS

1. Surveillance Requirements shall be applicable during any condition specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

2. Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance time interval.

3. Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Condition V.B.2.b(2), shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.

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4. Entry into any specified condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.

V.C. Radioactive Effluent Monitoring Instrumentation

1. Radioactive Liquid Effluent Monitoring Instrumentation LIMITING CONDITIONS OF OPERATION The radioactive liquid effluent monitoring instrumentation channels shown in Table VC.- 1 shall be OPERABLE with their Alarm/Trip setpoints set to ensure that the limits of Specification V.D.1.a are not exceeded. The alarm/trip setpoints shall be determined in accordance with methodology and parameters as described in Section II.

APPLICABILITY: As shown in Table VC.- 1 ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the action shown in Table VC.- 1. Exert best efforts to restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Releases need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENTS Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST at the frequencies shown in Table VC.-2.

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TABLE V.C.- 1 Radioactive Liquid Effluent Monitoring Instrumentation Instrument Minimum Applicability Action I_#OperableI I 1.Radioactivity Monitors Providing Alarm and Automatic Termination Of Release

a. Waste Neutralization Sump Monitor Condensate 1* ## D Polishing Facility

b. Turbine Building Floor Drains 1 # B

c. Liquid Waste Monitor 1 # A

d. RESERVED

e. Steam Generator Blowdown Monitor 1 ### B 2.Flow Rate Measurement Devices - No Alarm Setpoint Requirements

a. Waste Neutralization Sump Effluents 1" # C

b. RESERVED

c. Liquid Waste Effluent Line 1 # C

d. RESERVED

e. Steam Generator Blowdown Effluent Line 1 # C

• NA if tritium in the steam generators is less than detectable, or gamma radioactivity in the steam generators is less than 5 x 10-7 [tCi/ml, or the sump is being directed to radwaste.

1. At all times - which means that channels shall be OPERABLE and in service on a continuous, uninterrupted basis, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

1. 1. MODEs 1 -5, and MODE 6 when pathway is being used, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

The monitor must be on-line with no unexpected alarms. When the affected discharge path is isolated in MODE 6, the applicable LCO and Surveillance Requirements are not applicable.

1. 1. 1. MODEs 1-5, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument. The monitor must be on-line with no unexpected alarms.

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SECTION VI. Summary of Changes 1.1 Revision 025-02 1.1.1 Created Summary of Changes Section.

1.1.2 Corrected Table I.D-3 Title to Unit 3 from Unit 2.

1.1.3 Re-numbered Table I.E-I and changed number of locations from 3 to 2 under Well Water.

1.1.4 Changed referenced document V.B.1 definition 5. DOSE EQUIVALENT 1-131 to Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion".

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.TABLE V.C.-1 ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with the first Surveillance Requirement of Specification V.D.1 .a. and;

2. The original release rate calculations and discharge line valving are independently verified by a second individual.

Action B With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided best efforts are made to repair the instrument and that grab samples are analyzed for gamma radioactivity at the lower limits of detection specified in Table I.C.-3:

1. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 pCi/gram DOSE EQUIVALENT 1-131, or

2. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 pCi/gram DOSE EQUIVALENT 1-131.

Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves may be used to estimate flow.

Action D With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases may continue provided that best efforts are made to repair the instrument and that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with the first Surveillance Requirement of Specification V.D.1 .a., and;

2. If one of the samples has gamma radioactivity greater than any of the lower limits of detection specified in Table I.C.-3, the original release rate calculations and discharge valving are independently verified by a second individual.

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TABLE V.C.- 2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Instrument Channel Source Channel Channel Check Check Calibration Functional I I__ Test 1.Radioactivity Monitors Providing Alarm and Automatic Termination Of Release

a. Waste Neutralization Sump Monitor D P R(2 ) Q()

Condensate Polishing Facility

b. Turbine Building Floor Drains D M R(2 ) QM

c. Liquid Waste Monitor D P R( 2 ) QM

d. Deleted

e. Steam Generator Blowdown Monitor D M R(2 ) Q(M) 2.Flow Rate Measurements

a. Waste Neutralization Sump Effluents D( 3 ) NA R Q

b. RESERVED

c. Liquid Waste Effluent Line D( 3 ) N/A R Q

d. Deleted

e. Steam Generator Blowdown Effluent D( 3 ) N/A R Q Line D = Daily R = Once every 18 months M = Monthly Q = Once every 3 months P = Prior to each batch release N/A= Not Applicable TABLE VC.-2 TABLE NOTATION (1) The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur ifany of the following conditions exists:

a) Instrument indicates measured levels above the alarm/trip setpoint, or b) Circuit failure (Alarm only), or Instrument indicates a downscale failure (Alarm only).

(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities of NIST These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release.

CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.

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2. Radioactive Gaseous Effluent Monitoring Instrumentation LIMITING CONDITIONS OF OPERATION The radioactive gaseous effluent monitoring instrumentation channels shown in Table V.C.-3 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Specification V.D.2.a. are not exceeded. The Alarm/Trip Setpoints of these channels shall be determined in accordance with the methodology and parameters in Section II.

APPLICABILITY: As shown in Table V.C.-3.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the Minimum Channels OPERABLE, take the ACTION shown in Table V.C.-3. Exert best efforts to restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Release need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REOUIREMENT Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST at the frequencies shown in Table V.C.-4.

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TABLE V.C.-3 Radioactive Gaseous Effluent Monitoring Instrumentation Instrument Minimum Applicability Action Channels Operable 1.Millstone Unit 3 Ventilation Vent (Turbine Building - HVR-RE10B, normal range only; high range monitor, HVR-RE10A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing Alarm 1

• A

b. Iodine Sampler 1

• B

c. Particulate Sampler 1

• B

d. Vent Flow Rate Monitor 1

• C

e. Sampler Flow Rate Monitor 1

• C 2.Millstone Stack - applicable to SLCRS (HVR-RE19B, normal range only; high range monitor, HVR-REI9A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing Alarm

b. Iodine Sampler

c. Particulate Sampler

d. Process Flow Rate Monitor

e. Sampler Flow Rate Monitor 3.Engineered Safeguards Building Monitor (HVQ-RE49)

a. Noble Gas Activity Monitor

b. Iodine Sampler

c. Particulate Sampler

d. Discharge Flow Rate Monitor

e. Sampler Flow Rate Monitor TABLE V.C.-3 Table Notations Whenever the release path is in service. Outages are permitted for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of maintenance and performance of required tests, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

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TABLE V.C.-3 ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that a) best efforts are made to repair the instrument and that grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, OR b) ifthe cause of the inoperability is solely due to a loss of annunciation in the control room and the Remote Indicating Controller (RIC) remains OPERABLE, perform a channel check at the RIC at least once per twelve hours and verify the indication has not alarmed.

Action B With the number of samplers OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that the best efforts are made to repair the instrument and that samples are continuously collected with auxiliary sampling equipment for periods of seven (7) days and analyzed for principal gamma emitters with half lives greater than 8 days within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the end of the sampling period. Auxiliary sampling must be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after initiation of this ACTION statement. Operation of the auxiliary sampling equipment shall I be verified every twelve (12) hours. Auxiliary sampling outages are permitted for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of maintenance and performance of required tests, checks, calibrations, or sampling.

Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated at least once per 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Sample flow rate need not be estimated if the auxiliary sampling equipment of Action B is in use.

Action D With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Action E With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument.

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TABLE V.C.-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Instrument Check Source- Channel Channel When Check Calibra- Opera- Surveil-tion tional lance is Check Re-

. I quired 1.Millstone Unit 3 Ventilation Vent (Thrbine Building - HVR-RE10B, normal range only; high range monitor, HVR- RE10A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing D M R( 1) Q( 2 )

Alarm

b. Iodine Sampler W NA NA NA *

c. Particulate Sampler W NA NA NA

d. Vent Flow Rate Monitor D NA R Q *

e. Sampler Flow Rate Monitor D NA R Q

• 2.Millstone Stack - applicable to SLCRS (HVR-RE19B, normal range only; high range monitor, HVR-RE19A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing D M R( 3 ) Q( 2 )

Alarm

b. Iodine Sampler W NA NA NA *

c. Particulate Sampler W NA NA NA *

d. Process Flow Rate Monitor D NA R Q *

e. Sampler Flow Rate Monitor D NA R Q

• 3.Engineered Safeguards Building Monitor (HVQ - RE49)

a. Noble Gas Activity Monitor Providing D M R( 1) Q(2)

Alarm

b. Iodine Sampler W NA NA NA *

c. Particulate Sampler W NA NA NA *

d. Discharge Flow Rate Monitor D NA R Q *

e. Sampler Flow Rate Monitor D NA R Q

• At all times except when the vent path is isolated.

D = Daily a Once every 18 months W = Weekly Q= Once every 3 months M = Monthly N/A= Not Applicable MP REC- BAP01 STOP THINK' ACT REVIEW Rev. 026-01 150 of 166

TABLE V.C. - 4 Table Notations (1) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities of NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(2) The CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

a) Instrument indicates measured levels above the Alarm Setpoint, or b) Circuit failure, or c) Instrument indicates a downscale failure.

(3) The CHANNEL CALIBRATION shall include the use of a known source whose strength is determined by a detector which has been calibrated to an NIST source. These sources shall be in know, reproducible geometry.

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VD. Radioactive Effluents Concentrations And Dose Limitations

1. Radioactive Liquid Effluents

a. Radioactive Liquid Effluents Concentrations LIMITING CONDITIONS OF OPERATION The concentration of radioactive material released from the site (see Figure V.D.-1) shall be limited to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall not exceed 2 x 10-4 [tCi/ml total activity.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released from the site exceeding the above limits, restore the concentration to within the above limits within 15 minutes.

SURVEILLANCE REQUIREMENTS

1) Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Section I.

2) The results of the radioactive analysis shall be used in accordance with the methods of Section II to assure that the concentrations at the point of release are maintained within the limits of Specification V.D.I.a.

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b. Radioactive Liquid Effluents Doses LIMITING CONDITIONS OF OPERATION The dose or dose commitment to any REAL MEMBER OF THE PUBLIC from radioactive materials in liquid effluents from Unit 3 released from the site (see Figure V.D.-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and to less than or equal to 5 mrem to any organ; and,

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and the remainder of the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 3 mrem to the whole body and 10 mrem to any organ.

SURVEILLANCE REOUIREMENTS

1) Dose Calculations. Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in Section II at least once per 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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2. Radioactive Gaseous Effluents

a. Radioactive Gaseous Effluents Dose Rate LIMITING CONDITIONS OF OPERATION The dose rate, at any time, offsite (see Figure VD.- 1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

a. The dose rate limit for noble gases shall be less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin; and,

b. The dose rate limit due to inhalation for Iodine-131, Iodine-133, Tritium, and for all radioactive materials in particulate form with half lives greater than 8 days shall be less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, decrease the release rate to comply with the limit(s) given in Specification V.D.2.a. within 15 minutes.

SURVEILLANCE REOUIREMENTS

1) The release rate, at any time, of noble gases in gaseous effluents shall be controlled by the offsite dose rate as established in Specification V.D.2.a. The corresponding release rate shall be determined in accordance with the methodology of Section II.

2) The noble gas effluent monitors of TableV.C.-3 shall be used to control release rates to limit offsite doses within the values established in Specification VD.2.a.

3) The release rate of radioactive materials in gaseous effluents shall be determined by obtaining representative samples and performing analyses in accordance with the sampling and analysis program, specified in Section 1. The corresponding dose rate shall be determined using the methodology given in Section II.

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b. Radioactive Gaseous Effluents Noble Gas Dose LIMITING CONDITIONS OF OPERATION The air dose offsite (see Figure VD.- 1) due to noble gases released from Unit 3 in gaseous effluents shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose during the calendar year is within 10 mrad for gamma radiation and 20 mrad for beta radiation.

SURVEILLANCE REOUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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c. Gaseous Effluents - Doses from Radionuclides Other than Noble Gas LIMITING CONDITIONS OF OPERATION The dose to any REAL MEMBER OF THE PUBLIC from Iodine-131, Iodine- 133, Tritium, and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents released offsite from Unit 3 released offsite (see Figure VD.-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and,

b. During any calendar year: Less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

c. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 15 mrem to any organ.

SURVEILLANCE REQUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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Figure V.D.-1, "Site Boundary for Liquid and Gaseous Effluents" MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 157 of 166

V.E. Total Radiological Dose From Station Operations CONTROLS The annual dose or dose commitment to any REAL MEMBER OF THE PUBLIC, beyond the site boundary, from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls VD.1.b., VD.2.b., or V.D.2.c. prepare and submit a Special Report to the Commission within 30 days and limit the subsequent releases such that the dose commitment from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over consecutive months.

This Special Report shall include an analysis which demonstrates that radiation exposures from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site (including all effluent pathways and direct radiation) are less than the 40 CFR 190 Standard.

If the estimated doses exceed the above limits, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190.

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

SURVEILLANCE REOUIREMENTS Cumulative dose contributions from liquid and gaseous effluents and direct radiation from the Miilstone Site shall be determined in accordance with Section II once per 31 days.

VE BasesSection V.C.1. - Radioactive Liquid Effluent Monitoring Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section II to ensure that the alarm/trip will occur prior to exceeding ten times the limits of 10 CFR 20. The OPERABILITY and MP-22-REC-BAP01 STOP THINK, -ACT REVIEW Rev. 026-01 158 of 166

use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 20 Part 50.

OPERABILITY of a radiation monitor is determined by its ability to perform its specified function. The specified function of the radioactivity monitors listed in Table VC. -1 is to provide Control Room alarm annunciation and automatic termination of release. The monitor must be on-line with no unexpected alarms in order to perform its specified function.

Definition B.7 states a component is OPERABLE when it is capable of performing its specified function. The monitors are described in Tables VC.- 1 and V.C.-2 as "Radioactivity Monitors Providing Alarm and Automatic Termination of Releases." Table VC.-2 Note 1 requires that the Analog Channel Operational Test (ACOT) demonstrate that automatic isolation and "control room annunciation" occur. Control room annunciation cannot occur unless the monitor is on line (i.e. in communication with the RMS computer.).

Section VC.1. Surveillance Requirement requires that the ACOT be performed to demonstrate OPERABILITY. General Design Criteria 64 states in part:

"Means shall be provided for monitoring effluent discharge paths for radioactivity." Regulatory Guide 1.21 Appendix A describes a monitor program that is acceptable to the Regulatory staff. Under Section B of Appendix A, "LIQUID EFFLUENTS," the first paragraph states in part: "During the release of radioactive wastes, the effluent control monitor should be set to alarm and to initiate automatic closure..."

Certain of the monitors listed in Table VC. -1 are designed to operate without sample pumps. These monitors utilize pressure in the effluent line during discharge to provide sample flow and sample pressure. Low sample flow and/or low sample pressure alarms may be received when no discharge is in progress.

These are expected alarms. Sample flow and/or sample pressure will return to normal when the discharge is initiated. These alarms will clear when discharge begins. The monitors are OPERABLE since they are able to perform their specified function with the expected alarms in.

Table VC. -1 note ## requires entry into the radioactive liquid effluent monitoring action statements whenever the radiation monitors are not available in the required MODE. This note applies to items l.a (3CND-RE07, Waste Neutralization Sump), 1.d (3LWC-RE65, Regenerate Evaporator Monitor), and i.e (3SSR-RE08, Steam Generator Blowdown Monitor) in Table VC.- 1. The original issue of this requirement (as a Technical Specification) in January 1986 stated the applicability was "At all times." Technical Specification Amendment 22 added "APPLICABILITY" to Table VC.- 1 (then Tech Spec Table 3.3- ). The applicability added in Amendment 22 is the present wording. The letter 1312821, MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 159 of 166

dated February 24, 1988, in the following sections discusses the change request and are quoted below:

" In "Discussion": "The proposed changes will now explicitly allow a monitor to be taken out of service for up to hours for maintenance/testing without entering the ACTION statement."

• In "Significant Hazards Consideration" item 1: "the proposed changes would only allow these radiation monitors to be out of service for a short period of time (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)."

In "Significant Hazards Consideration" item 2: "The proposed changes also have no effect on alarm setpoints or control functions. Further, no operator actidns that are required to mitigate any accident rely solely on these monitors, and these monitors provide no protective functions."

Technical Specification Amendment 22 provided for the following:

• allowance for planned inoperability of monitoring instrumentation for up to hours for the purpose of maintenance and performance of required test, check, calibration or sampling a requirement to initiate auxiliary sampling within hours after inoperability of certain gaseous effluent monitors allowance for inoperability of certain effluent monitoring instrumentation, during MODE 6 (refueling) when the effluent pathway is not being used.

Section V.C.2.- Radioactive Gaseous Effluent Monitoring Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section II to ensure that the alarm/trip will occur prior to exceeding the limits of SectionV.D.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 20 Part 50.

The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of Specification V.C.2.a shall be such that concentrations as low as 1 x 10-6 itCi/cc are measurable.

The vent normal range radiation monitor, HVR* 10B, satisfies the requirements of SectionV.C.2. for Unit 3 releases to the vent which is located on the turbine MP-22-REC-BAP01 STOPR THiNK ACT REVIEW Rev. 026-01 160 of 166

building. There are no requirements in the REMODCM associated with the vent high range radiation monitor, HVR*10A.

The SLCRS normal range radiation monitor, HVR*19B, satisfies the requirements of SectionV.C.2. for Unit 3 releases to the Millstone Stack. There are no requirements in the REMODCM associated with the SLCRS high range radiation monitor, HVR*19A.

Section VD.1.a. - Radioactive Liquid Effluents Concentrations This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than ten times the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within: (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR 20.1301 to the population. The concentration limit for noble gases is based upon the assumption that Xe- 135 is the controlling radioisotope and its concentrations in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

Section V.D.1.b. - Radioactive Liquid Effluents Doses This specification is provided to implement the requirements of Sections II.A.,

III.A., and IV.A. of Appendix I, 10 CFR Part 50. The specification implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section III.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." The dose calculation methodology and parameters in Section II implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in Section II for calculating the doses due to the actual release rates of radioactive materials in liquid effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

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Section V.D.2.a.- Radioactive Gaseous Effluents Dose Rate This specification will ensure that the dose from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for all areas offsite. The annual dose limits specified in this section are the dose limits from the version of 10 CFR Part 20 prior to 1994. Annual dose limit in the current version of 10CFR20 were reduced from 500 to 100 mrem. But REMODCM restrictions will not allow the current annual dose limit to be exceeded because the REMODCM requires termination, within fifteen minutes, of any release which exceed the setpoint and much lower annual dose limits from 10CFR50, Appendix I are implemented. For individuals who may, at times, be within the SITE BOUNDARY, the occupancy of that individual will be usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the whole body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid or any other organ dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year.

Section VD.2.b. - Radioactive Gaseous Effluents Noble Gas Dose This specification is provided to implement the requirements of Sections II.B.,

III.A., and IVA. of Appendix I, 10 CFR Part 50. The specification implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section VA. of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A. of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in Section II for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculational of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.

The Section II equations provided for determining the air doses at the site boundary are based upon utilizing successively more realistic dose calculational methodologies. More realistic dose calculational methods are used whenever simplified calculations indicate a dose approaching a substantial portion of the MP-22-REC-BAP01 STOP THINK ACT REVIEW Rev. 026-01 162 of 166

regulatory limits. The methods used are, in order, previously determined air dose per released activity ratio, historical meteorological data and actual radionuclide mix released, or real time meteorology and actual radionuclides released.

Section V.D.2.c. - Radioactive Gaseous Effluents for Radionuclides Other Than Noble Gas These specifications are provided to implement the requirements of Sections II.C., III.A., and IV.A. of Appendix I, 10 CFR Part 50. The specifications are the guides set forth in Section II.C. of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A. of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."

The Section II calculational methods specified in the surveillance requirements implement the requirements in Section III.A. of Appendix I that conformance with the guides for Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The Section II calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculating of Annual Dose to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"

Revision I, July 1977. The release rate specifications for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man. The pathways that are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

Section V.E.- Total Radiological Dose from Station Operations This specification is provided to meet the dose limitations of 40 CFR 190. For the purpose of the Special Report, it may be assumed that the dose commitment to any REAL MEMBER OF THE PUBLIC from other fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered.

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Summary of Changes Summary of Changes Rev 026-01

1. Revised Table I.D. -1 to removed term "Particulate" from "Type of Activity" column.

2. Revised Table I.D.-2 Section B. to change "Containment Purge" to "Containment" and to remove "Containment Venting" and "Open Equipment Hatch During Outage," in the "Gaseous Release Point or Source" column, revised the "Sample Type and Frequency,"

removed the word "particulate" from "Type of Activity Analysis" column

3. Revised Table I.D. -3 Section B to removed the word "particulate" from "Type of Activity Analysis" column, and modified Fuel Building Minimum Analysis Frequency to read "Monthly for surge, vents, and drawdowns."

4. Revised Table I.E.-1 to reduce the number of locations for 5.Milk and 5.a.Pasture Grass from 3 to 2.

5. Removed sampling location 22-I from Table I.E.-2.

6. Deleted sampling location 36-T in Table I.E. -2 and added location 88-I "DEP Dock Near Barge Slip."

7. Revised Section II.D.3.a Method 1 calculation.

8. Revised Figure 1.E.-1, "Inner Air Particulate And Vegetation Monitoring Stations."

9. Revised Figure 1.E.-2, "Outer Terrestrial Monitoring Stations."

10. Revised Section II.E.3. Step 1 to reduce the noble gas concentration that requires reduction factor determination.

11. Revised Section II.E.8. Step 1 to reduce the noble gas concentration that requires reduction factor determination.

12. Revised Section II.E.4 "Condensate Polishing Facility Waste Neutralization Sump Effluent Line - CND245," to modify the value for a setpoint based on ten times background.

13. Revised Section II.E.5 "Unit 2 Steam Generator Blowdown - RM4262 and Unit 2 Steam Generator Blowdown Effluent Concentration Limitation," for a normal minimum circulating water dilution flow of 100,000 gpm.

14. Revised Section II.E.11.11a.c to modifiy the circulating and service water dilution flow assumption for the alarm setpoint for SSR-RE08 Unit 3 Steam Generator Blowdown.

15. Revised Section II.E.11.lla.d to substitute the phrase "230,000 gpm dilution flow," for "2 circulating and 2 service water pumps."

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16. Revised Section II.F, "Gaseous Monitor Setpoints."

Summary of Changes Rev 026-00

1. Corrected Figure I.C-2, "Simplified Liquid Effluent Flow Diagram Millstone Unit 2"

2. Corrected Figure I.C-3, "Simplified Liquid Effluent Flow Diagram Millstone Unit 3"

3. In Table I.D.-1, "Millstone Unit 1 Radioactive Gaseous Waste Sampling and Analysis Program," under B. Balance of Plant M-nt, Minimum Analysis Frequency is changed to Quarterly Composite.

4. Table I.D.-2, "Millstone Unit 2 Radioactive Gaseous Waste Sampling and Analysis Program," under B. Containment & Aux Building Releases, Minimum Analysis Frequency is modified.

5. Table I.D.-3, "Millstone Unit 3 Radioactive Gaseous Waste Sampling and Analysis Program," Section A is modified to include frequency change and other enhancements.

6. Clarifying text is added to Gaseous Radioctive Waste Treatment on page 32.

7. Additional information is added to Figure I.D.-2, "Simplified Gaseous Effluent Flow Diagram Millstone Unit Two."

8. Additional information is added to Figure I.D.-3, "Simplified Gaseous Effluent Flow Diagram Millstone Unit Three."

9. In Section I.E. Radiological Environmental Monitoring, under 1. Sampling and Analysis, the following language is included; "..., prepare and submit to the Commission within 30 days from receipt of sample results, a Special Report which includes...". This is a change from "...,

prepare and submit to the Commission within 30 days from the end of the affected calendar quarter, a Special Report which includes...".

10. In Table I.E.-1, "Millstone Radiological Environmental Monitoring Program," the number of locations listed under 7. Well Water is changed from 2 to 6.

11. In table I.E.-2 four new locations are identified.

12. On page 61 under 6. "Quarterly Dose Calculations for Radioactive Effluent Release Report," additional guidance/clarification is added.

13. On page 76 under c. "Unit 3 Projection Method," sections 2, "Due to Steam Generator Blowdown Tank Vent (Unit 3), and section 3) "Due to Ventillation Releases (Unit 3)," are deleted.

14. On page 76 under 5. "Quarterly Dose Calculations for Radioactive Effluent Release Report," additional guidance/clarification is added.

MP-22-REC-BAP01 STOP THINK .ACT REVIEW Rev. 026-01 165 of 166

15. On page 92, under 5. "Unit 2 Vent - Noble Gas Monitor - RM8132B," additional guidance/clarification is added.

16. On page 103 under ACTION STATEMENTS, the following is added to Action B, "Operation of the auxiliary sampling equipment shall be verified every twelve (12) hours."

17. On page 103 under ACTION STATEMENTS, the following is added to Action D, "Sample flow rate need not be estimated if the auxiliary sampling equipment of Action B is in use."

18. On page 107 under SURVEILLANCE REQUIREMENTS, 1) Dose Calculations has the frequency "once every 31 days," added.

19. On page 119, 11. "DOSE EQUIVALENT 1-131," has the following text added, "The thyroid dose conversion factors used for this calculation shall be those listed under Inhalation in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

20. On page 127 under 'Action B," the following wording is added; "Operation of the auxiliary sampling equipment shall be verified every twelve (12) hours."

21. On page 127 under 'Action C," the following wording is added; "Sample flow rate need not be estimated if the auxiliary sampling equipment of Action B is in use."

22. On page 149 under "Action B," the following wording is added; "Operation of the auxiliary sampling equipment shall be verified every twelve (12) hours."

21. On page 149 under 'Action C," the following wording is added; "Sample flow rate need not be estimated if the auxiliary sampling equipment of Action B is in use."

. IMP-22-REC-BAP01 STOP THINK' ACT REVIEW Rev. 026-01 166 of 166