ML091330240

2008 Radioactive Effluent Release Report

ML091330240
Person / Time
Site:	Millstone
Issue date:	04/30/2009
From:	Griffin R Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
09-200, MPS Lic/GJC R0, FOIA/PA-2011-0115
Download: ML091330240 (269)

Text

Dominion Nuclear Connecticut, Inc.

Millstone Power Station Dominion-Rope Ferry Road Waterford, CT 06385 APR 3 0 2009 U.S. Nuclear Regulatory Commission Serial No.09-200 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 2008 RADIOACTIVE EFFLUENT RELEASE REPORT In accordance with 10 CFR 50.36a, this letter transmits the annual Radioactive Effluent Release Report for the period January 2008 through December 2008. 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 1 of the 2008 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 2008. transmits Volume 2 of the report, which contains changes made to the REMODCM through 2008. Volume 2 consists of a complete copy of the REMODCM as of December 31, 2008, which satisfies the requirements of Sections 5.6.1c 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. T. Griffi Director, clear Station Safety and Licensing

Serial No.09-200 2008 Radioactive Effluent Release Report Page 2 of 3 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 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

Serial No.09-200 2008 Radioactive Effluent Release Report Page 3 of 3 Director Bureau of Air Management Monitoring & Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127 Mr. Robert Oliveria American Nuclear Insurers 95 Glastonbury Blvd.

Glastonbury, CT 06033

Serial No.09-200 Docket Nos. 50-245 50-336 50-423 License Nos. DPR-21 DPR-65 NPF-49 ATTACHMENT 1 2008 RADIOACTIVE EFFLUENT RELEASE REPORT VOLUME 1 MILLSTONE POWER STATION UNITS 1, 2, AND 3 DOMINION NUCLEAR CONNECTICUT, INC. (DNC)

Millstone Power Station 2008 Radioactive Effluents Release Report Volume 1 Dominion Nuclear Connecticut, Inc.

MILLSTONE UNIT LICENSE DOCKET 1 DPR-21 50.245 2 DPR-65 50-336 FDominion 3 NPF-49 50-423

Table of Contents Volume 1 Table of Contents ...................................................................... 1 List of Tables .................................... ............. ................................................ 2 Re fe re nce s ............................................................................................................................. ....... 3 Introd uctio n ........................... ......................................................................................................... 4 1.0 Off-Site Doses ................................................ ........... ...... .............................. 5 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 Measurem ent of Airborne Radioactivity .......................................................... 12 2.1.2 Estim ate of Errors ......................................................................................... 14 2.1.3 Airborne Batch Release Statistics ................................................................ 14 2.1.4 Abnorm al 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/Sum ps ............................................................................ 34 2.2.2 Estim ate of Errors .......................................................................................... 35 2.2.3 Liquid Batch Release Statistics ................................. 35 2.2.4 Abnorm al Liquid Releases ............................................................................ 35 2.2.5 Liquid Release Tables ....................................................................................... 35 2.3 Solid W aste ................................................................................................................... 49 2.4 Groundwater Monitoring ........................................................................................... .. 74 3.0 Inoperable Effluent Monitors .............................................................................................. 86 4.0 Operating History ............................................................. ................................................... 87 5 .0 E rra ta .................................................................................................................................... 90 6.0 REM O DCM Changes ............................... ....................................................................... 92 6.1 REM O DCM Description of Changes........................................................................ 93 Volume 2 2008 REMODCM Revision 26-00 1

List of Tables Table 1-1 Off-Site Dose Summary from Airborne Effluents - Units 1, 2, 3 Table 1-2 Off-Site Dose Summary from Liquid Effluents - Units 1, 2, 3 Table 1-3 Off-Site Dose Comparison to Limits - Units 1, 2, 3 Table 1-4 Off-Site Dose Comparison - Units 1, 2,3 Table 2.1-A1 Unit 1 Airborne Effluents - Release Summary Table 2.1-A2 Unit 1 Airborne Effluents - Ground Continuous - B 0 P Vent & SFPI Vent Table 2.1-L1 Unit 1 Liquid Effluents - Release Summary (Release Point - Quarry)

Table 2.1-L2 Unit 1 Liquid Effluents - Batch - (Release Point - Quarry)

Table 2.2-Al Unit 2 Airborne Effluents - Release Summary Table 2.2-A2 Unit 2 Airborne Effluents - Mixed Continuous - Aux Bldg Vent, SGBD Tank Vent

& Spent Fuel Pool Evaporation Table 2.2-A3 Unit 2 Airborne Effluents - Mixed / Elevated Batch - Containment Purges Table 2.2-A4 Unit 2 Airborne Effluents - Elevated Batch - WGDT Table 2.2-A5 Unit 2 Airborne Effluents - Elevated Continuous - Containment Vents/Site stack Table 2.2-A6 Unit 2 Airborne Effluents - Ground Batch - Containment Equipment Hatch Table 2.2-A7 Unit 2 Airborne Effluents - Ground Batch - RWST Vent Table 2.2-L1 Unit 2 Liquid Effluents - Release Summary - (Release Point - Quarry)

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

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

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

Table 2.2-L5 Unit 2 Liquid Effluents - Continuous - Turbine Building Sump - (Release Point - Yard Drain DSN 006)

Table 2.3-Al Unit 3 Airborne Effluents - Release Summary Table 2.3-A2 Unit 3 Airborne Effluents - Mixed Continuous - Normal Ventilation & Spent Fuel Pool Evaporation Table 2.3-A3 Unit 3 Airborne Effluents - Ground Continuous - ESF Building Ventilation Table 2.3-A4 Unit 3 Airborne Effluents - Mixed Batch - Containment Drawdowns Table 2.3-A5 Unit 3 Airborne Effluents - Mixed Batch - Containment Purges Table 2.3-A6 Unit 3 Airborne Effluents - Elevated Continuous - Gaseous Waste System Table 2.3-A7 Unit 3 Airborne Effluents - Elevated Batch - Containment Vents Table 2.3-A8 Unit 3 Airborne Effluents - Ground Batch - Containment Equipment Hatch Table 2.3-Ag Unit 3 Airborne Effluents - Ground Batch- RWST Vent Table 2.3-L1 Unit 3 Liquid Effluents - Release Summary - (Release Point - Quarry)

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

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

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

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

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

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

Table 2.1 -S Unit 1 Solid Waste & Irradiated Component Shipments Table 2.2-S Unit 2 Solid Waste & Irradiated Component Shipments Table 2.3-S Unit 3 Solid Waste & Irradiated Component Shipments Table 2.4-GW1 Environmental Well Monitoring Results Table 2.4-GW2 Catch Basin/Underdrain Monitoring Results Table 2.4-GW3 Underdrain Monitoring Results Table 2.4-GW4 RWST Yard Sample Results 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. Intentionally left blank

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. Intentionally left blank

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 Energy, 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 Liqht-Water-Cooled Nuclear Power Reactor Effluents.

8. 40 CFR Environmental Protection Agency, 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 - MP-CHEM-09011, 2008 Radioactive Effluent Data, April 7, 2009.

14. Memo - MP-CHEM-09012, 2008 Radioactive Effluent Miscellaneous Data, April 8, 2009.

15. Memo - MP-CHEM-09013, 2008 Unplanned Radioactive Releases, April 8, 2009.

16. Memo - MP-CHEM-09015, GPI Tables for 2008 RERR, April 9,.2009.

17. Memo - NSE-09-007, 2008 Effluent Monitor Performance, April 8, 2009.

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18. Memo - MP-HPO-09025, 2008 Report on Solid Waste and Irradiated Component Shipments, April 9, 2009.

Introduction This report, for the period of January through December of 2008, is being submitted by Dominion Nuclear Connecticut, Inc. for Millstone Power Station's Units 1, 2, and 3, in accordance with 10CFR50.36a, 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 2008 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 II (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 definedas 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 maximurmindividual 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 codeDOSAIR (Reference 11). This is equivalent to the NRC code, GASPAR II, which'uses a semi-infinitecloud model to implement the NRC Regulatory Guide 1.109 (Reference 3) dose modes. .

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 5

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.

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 1 was separated from releasing to the stack and modifications were made to add two new 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, 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 10CFR50, 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 40CFR190 (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 40CFR190.

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 2008 Off-Site Dose Commitments from Airborne Effluents Millstone Units 1, 2, 3 Ist Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Air (mrad) (mrad) (mrad) (mrad) (mrad)

Beta O.OOE+00 O.OOE+00 0.O0E+00 O.OOE+00 O.OOE+00 Gamma O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 8.11E-05 1.76E-04 1.42E-04 5.45E-05 . 4.53E-04 Skin 8.89E-05 1.76E-04 1.42E-04 5.45E-05 4.61 E-04 Thyroid 8.10E-05 1.76E-04 1.42E-04 5.45E-05 4.53E-04 Max organ+I 8.13E-05 1.76E-04 1.42E-04 5.45E-05 4.53E-04 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Air (mrad) (mrad) (mrad) (mrad) (mrad)

Beta 4.93E704 5.33E-03 8.60E-04 2.79E-04 6.96E-03 Gamma 3.75E-05 1.33E-03 1.34E-04 6.89E-05 1.57E-03 Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 1.54E-04 2.40E-03 6.10E-04 4.46E-04 3.61 E-03 Skin 4.99E-04 4.95E-03 9.82E-04 6.12E-04 7.04E-03 Thyroid 4.53E-04 6.41 E-02 9.38E-04 4.95E-04 6.60E-02 Max organ+I 1.60E-04 2.58E-03 6.17E-04 4.48E-04 3.81 E-03 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Air (mrad) (mrad) (mrad) (mrad) (mrad)

Beta 4.29E-04 1.62E-03 6.48E-04 7.60E-03 1.03E-02 Gamma 2.53E-05 4.53E-04 1.53E-04 2.56E-03 3.19E-03 Max Individual (mrem) (mrem). (intem) (imrem) (mrem)

Whole Body 1.70E-03 2.56E-03 2.67E-03 1.11E-02 1.80E-02 Skin 1.99E-03 2.80E-03 2.81 E-03 1.55E-02 2.31 E-02 Thyroid 1.70E-03 2.56E-03 2.67E-03 8.64E-02 9.34E-02 Max organ+ 1.71E-03 2.56E-03 2.67E-03 1.14E-02 1.83E-02

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

Table 1-2 2008 Off-Site Dose Commitments from Liquid Effluents Millstone Units 1, 2, 3 Max Individual 1st Quarter (mrem) 2nd Quarter (mrem) 3rd Quarter (mrem) 4th Quarter Annual Total (mrem) (mrem)

Whole Body 2.41 E-07 O.OOE+00 O.OOE+00 O.OOE+00 2.41 E-07 Thyroid 6.91 E-08 O.OOE+00 O.OOE+00 O.OOE+00 6.91 E-08 Max Organ 3.42E-07 O.OOE+O0 O.OOE+O0 O.OOE+00 3.42E-07 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 2.69E-04 7.06E-04 2.27E-05 3.28E-05 1.03E-03 Thyroid 1.74E-04 1.74E-04 6.04E-06 1.45E-05 3.69E-04 Max Organ 5.72E-03 1.14E-02 1.02E-03 1.36E-04 1.82E-02 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 3.37E-05 7.39E-05 1.25E-04 4.32E-04 6.65E-04 Thyroid 2.29E-05 5.48E-05 1.03E-04 3.52E-04 5.33E-04 Max Organ 2.42E-04 1.92E-04 2.41 E-04 4.85E-03 5.52E-03 9

Table 1-3 2008 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) (mrem) (mrem) (mrad) (mrad)

Unit I 4.53E-04 4.53E-04 4.53E-04 4.61E-04 O.OOE+00 O.OOE+00 Unit 2 3.61E-03 6.60E-02 3.81 E-03 7.04E-03 6.96E-03 1.57E-03 Unit 3 1.80E-02 9.34E-02 1.83E-02 2.31E-02 1.03E-02 3.19E-03 Millstone Station 2.21 E-02 1.60E-01 2.26E-02 3.06E-02 1.73E-02 4.76E-03 a 1 V0 V@ 10 ff  % _

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

(mrem) (mrem) (mrem)

Unit 1 2.41E-07 6.91 E-08 3.42E-07 Unit 2 1.03E-03 3.69E-04 1.82E-02 Unit 3 6.65E-04 5.33E-04 5.52E-03 Millstone Station 1.70E-03 9.02E-04 2.37E-02 Total Off-Site Dose from Millstone Station Max Individual Dose vs REMODCM & 40CFRI90 Limits Whole Body Thyroid Max Organ *

(mrem) (mrem) (mrem)

Airborne Effluents 2.21 E-02 1.60E-01 2.26E-02 Liquid Effluents 1.70E-03 9.02E-04 2.37E-02 Radwaste Storage 1.80E-01 1.80E-01 1.80E-01 Millstone Station 2.04E-01 3.41 E-01 2.26E-01 9=h33 _N_

IZZZ F__9_ _9 Note: REMODCM limits are listed in 10CFR50, Appendix I which contains additional limits not listed in the REMODCM

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

Table 1-4 2008 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 In the Body 40 mrem

.1 - 284 mreml Courtesy NCRP Report 94 (1987)

Maximum Off-Site Individual Millstone Station Whole Body Dose Airborne Effluents 0.0221 mrem Liquid Effluents 0.0017 mrem On-site RadWaste Storage 0.1800 mrem 1 0.2038 mreml 11

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 1 Spent Fuel Pool (SFPI) Island (no charcoal cartridge)

Unit 1 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 1 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 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 1 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 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 at least daily). 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 vent 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 these evolutions.

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% - 30% 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 1 37 5 Total Time (min) 841 5518 1929 Maximum Time (min) 841 230 612 Average Time (min) 841 149 386 Minimum Time (min) 841 48 145 Unit 3 Ctmt Purges Ctmt Vents* Ctmt Drawdowns Number of Batches 1 230 1 Total Time (min) 301 427 Maximum Time (min) 301 427 Average Time (min) 301 427 Minimum Time (min) 301 427

• - 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 14

2.1.4.3 Unit 3-

Description:

The 'B' RHR (Residual Heat Removal) suction valve, 3RHS*MV8702A, was identified (CR-07-12301) to have a known active pressure seal leak. Operability Determination OD MP3-020-07 evaluated the leak and determined that the leakage outside containment, .

including the assumed 3RHS*MV8702A pressure seal leakage, is below the allowable limit of 5000 cc/hr. The valve was operable but not fully qualified with the valve to be overhauled during 3R12. During the shutdown and once RHR cool down was initiated on the RCS (approximately 10/12/08 12:30), the 'B' RHR suction valve began leaking (CR113568) at a rate of 1-2 gpm resulting in elevated noble gas and iodine airborne activity in the ESF building RHR cubicles (CR1 13701). Normal ESF building ventilation is unfiltered and was in-service during this time.

Although attempts were made to filter the RHR cubicle areas, elevated levels existed for several days. These elevated levels resulted in additional releases to the environment as shown below.

These levels were well within regulatory limits. Table 2.3-A3 includes these values.

Cause: Leaking suction valve as described above.

Resolution: Valve isolated and repaired.

Radioactive Source: RHR System (Reactor coolant)

Pathway to the Environment/Release Point: ESF (Emergency Safeguards Facility) Ventilation, ground level release Release Start/Stop Dates/Times: Start: 10/12/08 12:30 Stop: 11/24/08 6:53 Most of the releases of noble gases, iodines and particulates occurred in the first few days of the event, however significant releases occurred several days later during repair activities. Releases continued for much of 3R12. Therefore, the total time of the event encompasses all the releases from this pathway.

Total Curies: Isotope Curies Calculated Doses:

Xe-133 7.09e-1 H-3 1.00e-2 Maximum Air (mrad) Beta: 1.87E-04 1-131 1.88e-3 Gamma: 6.29E-05 I-132 1-133 2.67e-4 2.67e-4 Maximum Individual (mrem) 1-133 2.00e-3 1-135 2.11 e-4 Whole Body: 1.20E-04 Br-82 1.22e-6 Skin: 1.99E-04 Co-58 2.59e-6 Thyroid: 2.07E-02 Co-60 7.76e-8 Maximum 1.61 E-04.

Ce-141 7.79e-8 Organ:

Eu-152 9.66e-7 These curies are also included in the 4th quarter of Table 2.3-A3 ESF Building Ventilation.

2.1.5 Airborne Release Tables The following tables provide the details of the airborne radioactivity released from each of the Millstone units. 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 1 Airborne Effluents - Release Summary Units 1st Qtr I 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases B. lodines / Halogens

1. Total Activity Ci Released

2. Average Period uCi/sec Release Rate

• D. Gross Alpha

1. Total Activity Ci Released E. Tritium

• "Total Activity Released" - Seconds in Quarter denotes less than Minimum Detectable Activity (MDA) 16

Table 2.1-A2 Millstone Unit 1 Airborne Effluents - Ground Continuous - BOP Vent & SFPI Vent

[ Nuclides Released Uni st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Gases y Emitters Ci Ci Ci Ci Ci Ci Ci Total Activity Ci I B. lodines / Halogens Ci na na na na Ci Ci Ci Ci Ci Total Activity Ci C. Particulates Cs-137 Ci, 1.19E-06 1.19E-06 Sr-90 Ci Ci Ci Ci Ci Ci Total Activity Ci 1.19E-06 1.19E-06 D. Gross Alpha Gross Alpha Ci - - - -

E. Tritium H-3 Ci 1.40E-01 1.32E-01 1.13E-01 2.35E-01 6.20E-01

"-" denotes less than Minimum Detectable Activity (MDA) 17

Table 2.2-Al Millstone Unit 2 Airborne Effluents - Release Summary Units ", st Qtr 2nd Qtr I 3rd Qtr I 4th Qtr Total C. Particulates D. Gross Alpha

1. Total Activity C Released

• "Total Activity Released" + Seconds in Quarter denotes less than Minimum Detectable Activity (MDA) 18

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

& Spent Fuel Pool Evaporation

• Nuclides Released I Units "lstQtr 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission& Activation Gases Ar-41 Ci -

Kr-85 Ci - 5.00E-01 5.OOE-01 1.OOE+00 Kr-85m Ci Xe-133 Ci 1.02E+01 1.OOE+00 5.OOE-01 1.17E+01 Xe-133m Ci 9.99E-02 9.99E-02 Xe-135 Ci 6.06E-02 6.06E-02 Ci Ci Ci Ci Total Activity Ci 1.04E+01 1.50E+00 1.OOE+00 1.29E+01 B. lodines / Halogens 1-131 Ci 7.49E-05 1.17E-03 6.74E-06 1.06E-05 1.26E-03 1-132 Ci 3.78E-05 7.70E-05 - 1.15E-04 1-133 Ci 2.72E-04 8.07E-05 2.71 E-05 4.06E-05 4.20E-04 1-134 Ci - -

1-135 Ci 2.03E-04 - 2.03E-04 Br-82 Ci 4.82E-06 - 4.82E-06 Total Activity Ci 5.88E-04 1.33E-03 3.38E-05 5.12E-05 2.01 E-03 C. Particulates Co-58 Ci 2.24E-06 - - 2.24E-06 Co-60 Ci 1.44E-06 - - 1.44E-06 Nb-95 Ci 3.27E-07 - - 3.27E-07 Zr-95 Ci 5.18E-07 - - 5.18E-07 Cs-137 Ci 2.16E-07 7.26E-07 - - 9.42E-07 Ci Sr-89 Ci Sr-90 Ci Ci Ci Total Activity Ci 2.16E-07 5.25E-06 5.47E-06 D. Gross Alpha IGross Alpha Ci - - - -

E. Tritium

• IH-3 Ci 1.14E+00 5.02E+00 1.59E+00 2.32E+00 1.01E+01

• Includes estimated Spent Fuel Pool evaporation

"-" denotes less than Minimum Detectable Activity (MDA) 19

Table 2.2-A3 Millstone Unit 2 Airborne Effluents - Mixed Batch - Containment Purges

. Nuclides S0 Released I UnitsI 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total A. Fission & Activation Gases Kr-85 Ci na 1.82E+00 na na 1.82E+00 Xe-133 Ci na 2.OOE-01 na na 2.OOE-01 Xe-135 Ci na 4.OOE-03 na na 4.OOE-03 Ci Ci Ci Ci Ci Ci Ci Ci Ci Ci Total Activity Ci 2.02E+00 2.02E+00 B. lodines / Halogens*

IT* a I

• Total Activity C. Particulates*

I I al ****

Total Activity Ci D. Gross Alpha

• IGross Alpha I Ci I * * *

• E. Tritium IH-3 Ci na 9.40E-02 na na 9.40E-02 lodines, Particulates and Gross Alpha appear in Tables 2.2-A2 or 2.2-A5

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 20

Table 2.2-A4 Millstone Unit 2 Airborne Effluents - Elevated Batch - WGDT Nuclides n20I8; Released Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases Kr-85 Ci 5.16E-01 1.81E+00 8.02E-01 na 3.13E+00 Xe-133 Ci 4.27E-05 - na 4.27E-05 Ci Ci Ci Ci Ci Ci Total Activity Ci 5.16E-01 1.81E+00 8.02E-01 3.13E+00 B. lodines / Halogens_*

S

• Ci [ ....

• Total Activity I Ci C. Particulates

• Itota, ctivit ICi I D. Gross Alpha

• IGross Alpha Ci * * *

• E. Tritium IH-3 Ci 3.42E-04 3.37E-04 4.44E-04 - 1.12E-03

• lodines, Particulates and Gross Alpha appear in Tables 2.2-A2 or 2.2-A5

"-" denotes less than Minimum Detectable Activity (MDA) 21

Table 2.2-A5 Millstone Unit 2 Airborne Effluents - Elevated Continuous - Containment Vents/Site Stack Nuclides I Released Units I 1st Qtr I 2ndQtr I 3rdQtr I1 4thQtr I Total I A. Fission & Activation Gases Ar-41 Ci 2.66E-02 1.52E-02 2.11E-02 2.15E-02 8.44E-02 Kr-85 Ci 2.39E-01 3.85E-02 1.49E-02 2.92E-01 Xe-131m Ci Xe-133 Ci 3.04E-02 1.94E-01 7.1OE-03 1.06E-02 2.42E-01 Xe-133m Ci Xe-135 Ci 2.65E-04 2.63E-04 6.84E-04 1.02E-03 2.23E-03 Xe-138 Ci 3.13E-04 3.13E-04 Ci Total Activity Ci 2.96E-01 2.48E-01 4.41E-02 3.31E-02 6.21E-01 B. lodines / Halogens 1-131 Ci 1.01E-07 7.08E-05 - 6.68E-07 7.16E-05 1-132 Ci -

1-133 Ci 3.82E-07 1.05E-05 - 1.09E-05 Br-82 Ci 1.35E-06 - 1.35E-06 Ci Ci Total Activity Ci 4.83E-07 8.27E-05 6.68E-07 8.38E-05 C. Particulates Co-58 Ci 1.21E-07 - 1.21E-07 Cs-137 Ci 2.64E-08 1.12E-07 1.38E-07 Ci Sr-89 Ci Sr-90 Ci Ci Ci Ci Total Activity Ci 1.47E-07 1.12E-07 2.59E-07 D. Gross Alpha Gross Alpha Ci - I - -

E. Tritium LH-3 Ci 1.41E-01 2.03E-01 1.35E-01 8.35E-01 1-31E+OO

"-" denotes less than Minimum Detectable Activity (MDA) 22

Table 2.2-A6 Millstone Unit 2 Airborne Effluents - Ground Batch - Containment Equipment Hatch I- Nuclides Released I Unt 1st =Qtr 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Gases y Emitters Ci na na na na Ci Ci Ci Ci Total Activity Ci B. lodines / Halogens 1-131 Ci na 2.40E-06 na na 2.40E-06 1-132 Ci na 5.80E-06 na na 5.80E-06 1-133 Ci na 1.80E-07 na na 1.80E-07 Ci Ci Total Activity Ci 8.38E-06 _ 8.38E-06 C. Particulates Cr-51 Ci na 5.60E-06 na na 5.60E-06 Mn-54 Ci na 1.80E-07 na na 1.80E-07 Co-58 Ci na 3.70E-06 na na 3.70E-06 Co-60 Ci na 6.20E-07 na na 6.20E-07 Zr-95 Ci na 1.70E-06 na na 1.70E-06 Nb-95 Ci na 2.1OE-06 na na 2.1OE-06 Cs-137 Ci na 3.30E-07 na na 3.30E-07 Ci Ci Ci Ci Ci Total Activity Ci I 1.42E-05 1.42E-05 D. Gross Alpha IGross Alpha Ci na na na na E. Tritium IH-3 Ci na na na na

"-" denotes less than Minimum Detectable Activity (MDA)

'na" denotes Not Analyzed 23

Table 2.2-A7 Millstone Unit 2 Airborne Effluents - Ground Batch - RWST Vent Nuclides "t:

I Released I Units 1 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases Xe-1 33 Ci na 2.60E-02 na na 2.60E-02 Ci Ci Ci Ci Ci Ci Ci Total Activity Ci I 2.60E-02 2.60E-02 B. lodines / Halogens 1-131 Ci na - na na 1-132 Ci na - na na 1-133 Ci na - na na Total Activity Ci C. Particulates y Emitters Ci na - na na Total Activity Ci D. Gross Alpha IGross Alpha I Ci I na I na I na I na I E. Tritium IH-3 I Ci na na na na

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 24

Table 2.3-Al Millstone Unit 3 Airborne Effluents - Release Summary

" s--6 0:

Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases D. Gross Alpha

1. Total Activity Ci Released C "Total Activity Released" ÷ Seconds in Quarter denotes less than Minimum Detectable Activity (MDA) 25

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

Spent Fuel Pool Evaporation

• Nuclides Released Units 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases Xe-133 Ci 2.92E+00 2.63E+01 2.92E+01 Ci Ci Ci Ci Ci Ci Ci Total Activity Ci I 2.92E+00 I 2.63E+01 j292E+O1 B. lodines / Halogens 1-131 Ci 1.31E-03 1.31E-03 1-133 Ci 2.49E-04 2.49E-04 Ci Ci Ci Ci Total Activity Ci 1.56E-03 1.56E-03 C. Particulates Be-7 Ci - 1.30E-05 1.30E-05 Cr-51 Ci - 1.26E-04 1.26E-04 Mn-54 Ci - 1.53E-05 1.53E-05 Co-58 Ci - 1.10E-04 1.1OE-04 Co-60 Ci - 6.81 E-05 6.81 E-05 Nb-95 Ci - 1.18E-05 1.18E-05 Zr-95 Ci - 6.15E-06 6.15E-06 Ba-140 Ci - 2.25E-05 2.25E-05 Ci Sr-89 Ci Sr-90 Ci Ci Ci Ci Total Activity Ci 1.30E-05 3.60E-04 3.73E-04 D. Gross Alpha lGross Al ha I Ci I - - -

E. Tritium

• H-3 I Ci I 2.09E+01 9.94E+00 1.05E+01 2.97E+01 7.10E+01 Includes estimated Spent Fuel Pool evaporation

"-" denotes less than Minimum Detectable Activity (MDA) 26

Table 2.3-A3 Millstone Unit 3 Airborne Effluents - Ground Continuous - ESF Building Ventilation Nuclides Released 1stQtr 1Units I 2nd Qtr I- 3rd Qtr 4rthtr Ttal A. Fission & Activation Gases Xe-133 Ci 7.09E-01 7.09E-01 Ci Ci Ci Ci Ci Ci Ci Total Activity Ci I 7.09E-01 7.09E-01 B. lodines / Halogens 1-131 Ci 1.88E-03 1.88E-03 1-132 Ci 2.67E-04 2.67E-04 1-133 Ci 2.OOE-03 2.OOE-03 1-135 Ci 2.11E-04 2.11E-04 Br-82 Ci 1.22E-06 1.22E-06 Ci Total Activity Ci I 4.36E-03 4.36E-03 C. Particulates Be-7 Ci 3.69E-07 3.85E-07 3.89E-07 1.14E-06 Co-58 Ci 2.59E-06 2.59E-06 Co-60 Ci 7.76E-08 7.76E-08 Ce-141 Ci 7.79E-08 7.79E-08 Eu-152 Ci 9.66E-07 9.66E-07 Ci Sr-89 Ci Sr-90 Ci Ci Ci Ci Ci Ci Ci Total Activity Ci 4.47E-07 3.85E-07 3.89E-07 3.63E-06 4.85E-06 D. Gross Alpha Gross Alpha Ci I - - II E. Tritium 1H-3 Ci I 2.57E-01 I 1.09E-01 1.88E-01 I 1.03E-01 6.57E-01

"-" denotes less than Minimum Detectable Activity (MDA) 27

Table 2.3-A4 Millstone Unit 3 Airborne Effluents - Mixed Batch - Containment Drawdowns Nuclides I Released I Unitsj 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Gases Xe-133 Ci na na na Ci Ci Ci Ci Ci Total Activity Ci B. Iodines / Halogens 1-131 Ci na na na 1.42E-07 1.42E-07 Ci Ci Total Activity Ci 1.42E-07 1.42E-07 C. Particulates y Emitters Ci na na na Ci Ci Ci Ci Ci Total Activity Ci D. Gross Alpha Gross Alpha Ci na na na na E. Tritium H-3 Ci na na na 1.76E-03 1.76E-03

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 28

Table 2.3-A5 Millstone Unit 3 Airborne Effluents - Mixed Batch - Containment Purges Nuclides" Released Units i 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases Ar-41 Ci na na na Kr-85m Ci na na na 8.90E-04 8.90E-04 Xe-133 Ci na na na 5.27E-01 5.27E-01 Xe-133m Ci na na na 8.90E-03 8.90E-03 Xe-135 Ci na na na 2.35E-02 2.35E-02 Ci Ci Ci Ci Ci Ci Ci Total Activity Ci 5.60E-01 5.60E-01 B. lodines / Halogens

• I I C i II ....

Total Activity Ci C. Particulates

• I

• Cil Total Activity Ci D. Gross Alpha

• IGross Alpha ci * * *

• E. Tritium H-3 Ci na na na I 1.39E-02 1.39E-02

• lodines, Particulates and Gross Aipha included in Table 2.3-A2

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 29

Table 2.3-A6 Millstone Unit 3 Airborne Effluents - Elevated Continuous - Gaseous Waste System Nuclides Released 1Units st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission &Activation Gases .,, _ , .- ..

Kr-85 Ci 1.88E+00 7.45E-O1 7.72E-01 2.54E+00 5.94E+00 Kr-85m Ci - 1.88E-02 1.88E-02 Xe-1 31m Ci - 2.45E-01 2.45E-01 Xe-133 Ci - 2.07E+00 1.88E+00 1.69E+01 2.09E+01 Xe-133m Ci - 1.94E-03 4.01E-01 4.03E-01 Xe-135 Ci - - - 1.60E+00 1.60E+00 Ci - - -

Ci - - - -

Ci - - - -

Ci - - - -

Ci - - - -

Total Activity Ci 1.88E+00 2.82E+00 2.65E+00 2.17E+01 2.91E+01 B. lodines I Halogens 1-131 Ci 1.01E-07 - 1.01E-07 1.03E-03 1.03E-03 1-133 Ci - 8.90E-05 8.90E-05 Br-82 Ci 6.40E-06 4.99E-06 6.19E-06 2.12E-06 1.97E-05 Ci Ci Ci ITotal Activity Ci 6.50E-06 4.99E-06 6.29E-06 I1.12E-03 i.14E-03 C. Particulates Cr-51 Ci 5.50E-07 5.50E-07 Mn-54 Ci 7.47E-08 - 7.47E-08 Co-58 Ci - 9.48E-06 9.48E-06 Co-60 Ci 3.66E-07 - - 9.99E-08 4.66E-07 Cs-1 34 Ci - - - 5.89E-07 5.89E-07 Cs-137 Ci - - - 5.31E-07 5.31E-07 Ba-140 Ci - - - 1.94E-06 1.94E-06 Ci Sr-89 Ci - - -

Sr-90 Ci - - -

Ci Ci Total Activity Ci 4.41 E-07 1.32E-05 1.36E-05 D. Gross Alpha Gross Alpha Ci oilI E. Tritium H-3 Ci 3.36E-01 I 8.06E-01i 6.67E-01 I1.17E+00 2.98E+00

"-" denotes less than Minimum Detectable Activity (MDA) 30

Table 2.3-A7 Millstone Unit 3 Airborne Effluents - Elevated Batch - Containment Vents SNuclides [00 Released I Units 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases Ar-41 Ci 7.57E-03 7.44E-03 8.57E-03 3.77E-03 2.74E-02 Kr-85 Ci 1.44E-02 1.40E-02 9.24E-03 5.60E-02 9.36E-02 Kr-85m Ci 3.51E-05 2.02E-04 1.24E-04 3.61E-04 Xe-131m Ci 1.91E-03 2.50E-03 5.11E-03 2.59E-03 1.21E-02 Xe-133 Ci 3.69E-01 6.31E-01 6.82E-01 5.40E-02 1.74E+00 Xe-133m Ci 4.27E-03 8.92E-03 6.98E-03 6.18E-04 2.08E-02 Xe-135 Ci 4.46E-03 7.08E-03 4.83E-03 7.25E-04 1.71E-02 Ci - -

Ci

____ ____ Ci -.--

Ci Ci Total Activity Ci 4.02E-01 6.71E-01 7.17E-01 1.18E-01 1.91E+00 B. lodines / Halogens *

• Cii Total Activity Ci C. Particulates

• I Cii Total Activity Ci D. Gross Alpha

• IGross Alpha Ci * * *

• E. Tritium 1H-3 Ci 3.24E-02 2.17E-02 1.83E-02 4.13E-03 7.65E-02 lodines, Particulates and Gross Aipha included in Table 2.3-A6

"-" denotes less than Minimum Detectable Activity (MDA) 31

Table 2.3-A8 Millstone Unit 3 Airborne Effluents - Ground Batch - Containment Equipment Hatch Nuclides uI; Released Units 1st Qtr .2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases y Emitters Ci na na na Ci Ci Ci Ci Ci Ci Total Activity Ci I I B. lodines / Halogens 1-131 Ci na na na 1.42E-04 1.42E-04 Ci Ci Ci Ci

,Total Activity Ci 1.42E-04 1.42E-04 C. Particulates Cr-51 Ci na na na 1.58E-05 1.58E-05 Mn-54 Ci na na na 1.86E-06 1.86E-06 Co-58 Ci na na na 2.15E-05 2.15E-05 Fe-59 Ci na na na 5.14E-07 5.14E-07 Co-60 Ci na na na 4.03E-06 4.03E-06 Nb-95 Ci na na na 9.84E-07 9.84E-07 Zr-95 Ci na na na 1.01 E-06 1.01 E-06 Cs-137 Ci na na na 1.12E-06 1.12E-06 Ci Ci Ci Ci Ci Ci Total Activity Ci 4.68E-05 4.68E-05 D. Gross Alpha IGross Alpha I Oi na I na I na na E. Tritium IH-3 Ci na na na na

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 32

Table 2.3-A9 Millstone Unit 3 Airborne Effluents - Ground Batch - RWST Vent Nuclides I i Released nits 1st Qtr 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases Kr-85 Ci na na na Xe-133 Ci na na na 3.32E-01 3.32E-01 Ci Ci Total Activity Ci 3.32E-01 3.32E-01 B. lodines / Halogens 1-131 Ci na na na 2.33E-04 2.33E-04 1-132 Ci na na na 9.73E-06 9.73E-06 1-133 Ci na na na -

Ci Total Activity Ci 2.43E-04 2.43E-04 C. Particulates Cr-51 Ci na na na 5.79E-05 5.79E-05 Mn-54 Ci na na na 5.79E-06 5.79E-06 Co-57 Ci na na na 3.13E-07 3.13E-07 Co-58 Ci na na na 1.37E-04 1.37E-04 Fe-59 Ci na na na 1.55E-06 1.55E-06 Co-60 Ci na na na 5.47E-06 5.47E-06 Zr-95 Ci na na na 2.96E-06 2.96E-06 Nb-95 Ci na na na 4.64E-06 4.64E-06 Mo-99 Ci na na na 2.67E-08 2.67E-08 Tc-99m Ci na na na 1.32E-08 1.32E-08 Ag-110m Ci na na na 1.37E-07 1.37E-07 Sn-113 Ci na na na 1.32E-08 1.32E-08 Sn-117m Ci na na na 8.86E-09 8.86E-09 Sb-124 Ci na na na 1.60E-07 1.60E-07 Sb-125 Ci na na na 2.03E-06 2.03E-06 Cs-134 Ci na na na 1.46E-05 1.46E-05 Cs-136 Ci na na na 5.18E-07 5.18E-07 Cs-1 37 Ci na na na 7.89E-06 7.89E-06 Ba-140 Ci na na na 4.85E-08 4.85E-08 Ci

___I__I__

Total Activity Ci 2.41 E-04 2.41E-04 D. Gross Alpha

[Gross Alpha I na I na I na I na I E. Tritium IH-3 I Ci na I na na I na I

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not 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 and 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 tanks & sumps that are used to discharge liquids containing radioactivity to the environs; they are:

Unit 1 Reactor Cavity Water Unit 2 Clean Waste Monitor Tanks (2)

Aerated Waste Monitor Tanks (2)

CPF Waste Neutralization Sump & Turbine Building Sump Steam Generator Bulk 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 Prior to release, a tank is re-circulated for 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, Ni-63, 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 1 64 56 Total Time (min) 373 6558 7340 Maximum Time (min) 373 258 1412 Average Time (min) 373 103 131 Minimum Time (min) 373 1 13 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 2008, 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. They are categorized by type of release, source(s), and by release point of discharge to the environment.

35

Table 2.1-L1 Millstone Unit 1 Liquid Effluents - Release Summary (Release Point - Quarry)

Units 1 1st Qtr 2nd Qtr 3rd Qtr I 4th Qtr I Total A. Fission and Activation Products C. Dissolved and Entrained Gases

1. Total Activity Ci Released

2. Average Period uCi/ml Diluted Activity

• D. Gross Alpha

1. Total Activity Ci Released E. Volume

1. Released Waste Liters 1.06E+05 1.06E+05 Volume

2. Dilution Volume Liters 7.24E+08 7.24E+08 During Releases"_

3. Dilution Volume Liters 2.54E+ 11 2.54E+11 During Period'

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

+ Unit 2 E.3 quarterly dilution used because there is no more Unit 1 dilution

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

Table 2.1-L2 Millstone Unit 1 Liquid Effluents - Batch (Release Point - Quarry)

-Nuclides fflg Released I Units 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total A. Fission & Activation Products Cs-137 Ci 2.OOE-05 na na na 2.OOE-05 Ci Ci Total Activity Ci 2.OOE-05 2.OOE-05 B. Tritium

[H-3 Ici] - na [ na na C. Dissolved & Entrained Gases y Emitters Ci na na na Ci Ci Total Activity Ci D. Gross Alpha IGrossAlpha I ci I na I na I na

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 37

Table 2.2-L1 Millstone Unit No. 2 Liquid Effluents - Release Summary (Release Point - Quarry)

IUnits I 11st Qtr 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission and Activation Products

1. Total Activity Ci 2.

Released

[. Average Period Diluted Activity*

B. Tritium C.

1.

2.

D. Gross Alpha

1. Total Activity Ci Released E. Volume

1. Released Waste Volume Primary Liters 6.21 E+05 1.25E+06 1.69E+05 1.11 E+05 2.15E+06 Secondary J Liters 4.08E+04 1.40E+05 7.35E+06 7.53E+06

2. Dilution Volume During Releases Primary Liters 2.99E+09 3.18E+09 1.68E+09 1.08E+09 8.93E+09 Secondary Liters

3. Dilution Volume During Period I Liters I 2.54E+11 1.87E+11 2.87E+11 2.76E+11 1.0OE+12 "Total Activity Released" + (Primary "Released Waste Volume" + "Dilution Volume During Period")

38

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

NuclidesI ,20 Released I Units 1st Qtr I 2ndQtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Products Cs-137 Ci 1.59E-07 1.59E-07 Ci Ci Total Activity Ci 1.59E-07 1.59E-07 B. Tritium IH-3 &Ci 3.26E-04 ! 1.09E-03 - 1 1"21E-02 ] 1.35E-02 ]

C. Dissolved & Entrained Gases y Emitters Ci Ci Ci Total Activity Ci D. Gross Alpha Gross Alpha I Oi na na na na

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 39

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

Nuclides Released I Units 1st Qtr 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Products Be-7 Ci .... - 1.76E-05 1.76E-05 Cr-51 Ci 1.23E-02 2.30E-04 1.25E-02 Mn-54 Ci 5.56E-04 9.31 E-04 3.83E-05 1.53E-03 Co-57 Ci 3.12E-05 2.1OE-06 3.33E-05 Co-58 Ci 2.57E-04 9.34E-03 2.84E-04 2.65E-05 9.91 E-03 Fe-59 Ci 1.56E-03 9.62E-06 1.57E-03 Co-60 Ci 9.78E-03 1.27E-02 3.51 E-04 1.47E-05 2.28E-02 Nb-95 Ci 5.59E-05 2.37E-03 1.89E-04 - 2.61 E-03 Zr-95 Ci 1.47E-03 8.35E-05 - 1.55E-03 Ru-103 Ci 1.34E-04 - 1.34E-04 Ru-105 Ci 5.80E-05 1.27E-05 - 7.07E-05 Ag-11Orn Ci 1.66E-03 1.15E-03 2.38E-04 8.11E-06 3.06E-03 Sn-113 Ci 2.45E-05 1.02E-03 1.29E-04 2.23E-05 1.20E-03 Sn-117m Ci 6.51E-04 8.52E-06 6.60E-04 Sb-124 Ci 2.41 E-04 3.47E-05 1.68E-05 2.93E-04 Sb-125 Ci 2.82E-03 7.33E-03 1.55E-03 1.93E-03 1.36E-02 1-132 Ci 2.08E-04 2.08E-04 Te-132 Ci 1.82E-04 1.82E-04 Cs-1 34 Ci 8.87E-05 8.87E-05 Cs-137 Ci 9.52E 2.31E-04 1.71 E-06 3.54E-04 6.82E-04 La-1 40 Ci 3.34E-06 3.34E-06 Fe-55 Ci 4.65E-03 2.36E-02 1.22E-03 1.17E-03 3.06E-02 Ni-63 Ci 1.45E-03 7.66E-03 1.14E-04 2.06E-04 9.43E-03 Sr-89 Ci Sr-90 Ci I Total Activity Ci 2.14E-02 8.31 E-02 4.49E-03 3.85E-03 1.13E-01 B. Tritium IH-3 [ Ci I 3.12E+02 1.28E+02 [ 7.11E+00 [ 2.94E+01 [ 4.77E+02 C. Dissolved & Entrained Gases Kr-85 Ci 9.06E-02 4.48E-01 2.16E-02 - 5.60E-01 Xe-1 31m Ci 5.95E-03 - 5.95E-03 Xe-133 Ci 1.01E-02 3.50E-01 7.73E-05 - 3.60E-01 Xe-133m Ci 3.93E-03 - - 3.93E-03 Xe-135 Ci 7.09E-04 - - 7.09E-04 I Ci Total Activity Ci 1.01E-01 8.09E-01 2.17E-02 9.31E-01 D. Gross Alpha IGross Al ha I - i - -I

"-" denotes less than Minimum Detectable Activity (MDA) 40

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

Units " st Qtr I 2nd Qtr I3rd Qtr I 4th Qtr J Total A. Fission and Activation Products

1. Total Activity Released

2. Average Period Diluted Activity B. Tritium 1.

2.

C. Di 1.

2.

D. Gross Alpha

1. Total Activity C Released E. Volume

1. Released Waste Liters 2.40E+06 9.76E+05 1.95E+05 5.OOE+05 4.07E+06 Volume

2. Dilution Volume Liters During Releases

3. Dilution Volume Liters 3.39E+07 2.89E+07 3.08E+07 3.50E+07 1.29E+08 During Period **

• "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 (Release Point - Yard Drain - DSN 006)

Nuclides fill; Released I Units st Qtr I 2nd Qtr I 3rd Qtr 4th QtrI Total A. Fission & Activation Products y Emitters Ci Ci Ci Total Activity Ci B. Tritium IH-3 Ci 1.39E-02 ] 6.91E-03 [ 4.46E-04 1.08E-03 ] 2.23E-02 ]

C. Dissolved & Entrained Gases y Emitters Ci Ci Ci Total Activity Ci D. Gross Alpha Gross Alpha I i na na na na

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 42

Table 2.3-L1 Millstone Unit 3 Liquid Effluents - Release Summary (Release Point - Quarry)

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

1.

11.

2.

C. Dissolved and Entrained Gases 1.

2.

D. Gross Alpha

1. Total Activity Ci Released E. Volume

1. Released Waste Volume Primary Liters 1.30E+05 4.39E+05 8.27E+05 6.82E+06 8.22E+06 Secondary Liters 1.92E+07 8.90E+06 8.82E+06 3.72E+06 4.07E+07

2. Dilution Volume During Releases Primary Liters 5.37E+08 2.42E+09 3.97E+09 7.88E+09 1.48E+10 Secondary Liters 1.26E+10 1.16E+10 1.12E+10 3.13E+09 3.85E+10

3. Dilution Volume During Period Liters 4.52E+11 4.59E+11 4.71E+11 3.26E+11 1.71E+12 "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 Released I Units 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total A. Fission & Activation Products y Emitters Ci - "

Ci Ci Total Activity Ci B. Tritium H-3 I Ci 3.27E-01 1.75E-01 [ 9.03E-02 1.85E-02 6.11E-01 C. Dissolved & Entrained Gases y Emitters Ci Ci Ci Total Activity Ci D. Gross Alpha IGross Alpha I ci na I na I na I na I

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 44

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

Nuclides Released I Units IlstQtr 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Products Cr-51 Ci - - - 1.01E-02 1.01E-02 Mn-54 Ci 5.62E-05 1.02E-04 2.39E-04 1.93E-03 2.33E-03 Co-58 Ci 4.56E-05 - 7.39E-06 1.33E-02 1.34E-02 Fe-59 Ci - - - 5.98E-04 5.98E-04 Co-60 Ci 3.30E-04 5.47E-04 1.73E-03 8.62E-03 1.12E-02 Nb-95 Ci - 7.67E-06 - 1.47E-03 1.48E-03 Zr-95 Ci - - 8.43E-04 8.43E-04 Ag-i10m Ci 6.21E-05 2.OOE-05 8.11E-06 1.31E-04 2.21E-04 Sn-117m Ci 8.15E-05 8.15E-05 Sb-122 Ci 1.07E-05 1.07E-05 Sb-124 Ci 2.20E-05 2.20E-05 Sb-125 Ci 5.11E-05 3.44E-03 1.55E-02 4.55E-03 2.35E-02 1-131 Ci - 1.06E-03 1.06E-03 Cs-134 Ci 6.53E-05 2.30E-04 2.11E-04 5.06E-04 Cs-137 Ci 5.14E-05 1.95E-04 3.54E-04 6.OOE-04 Ci Fe-55 Ci 8.37E-04 1.08E-03 9.92E-04 7.31E-03 1.02E-02 Ni-63 Ci 5.93E-05 3.92E-03 3.98E-03 Sr-89 Ci Sr-90 Ci Ci Ci Ci Total Activity Ci 1.56E-03 5.62E-03 1.91 E-02 5.39E-02 8.02E-02 B. Tritium

[H-3 [Ci 8.47E+00 [ 8.12E+01 [ 2.34E+02 [ 3.71E+02 6.95E+02 C. Dissolved & Entrained Gases V Emitters Ci Ci Ci Ci Ci Ci Total Activity Ci I I I D. Gross Alpha IGross Alpha I Ci III -

"-" denotes less than Minimum Detectable Activity (MDA) 45

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

Nuclides III:

Released I units 1st Qtr Qtr I 3rd QtrI 4th QtrI Total A. Fission & Activation Products Cr-51 Ci 1.55E-05 1.55E-05 Mn-54 Ci - 6.22E-06 6.22E-06 Co-58 Ci - 5.49E-05 5.49E-05 Co-60 Ci - 2.37E-06 2.37E-06 Sb-122 Ci - 2.47E-06 2.47E-06 1-131 Ci - 3.92E-05 3.92E-05 1-133 Ci - 2.45E-06 2.45E-06 Cs-1 34 Ci - 5.38E-05 5.38E-05 Cs-1 36 Ci - 6.02E-06 6.02E-06 Cs-137 Ci - 2.41E-05 2.41E-05 Ce-144 Ci - 4.27E-06 4.27E-06 Ci Fe-55 Ci na na na

• Ni-63 Ci na na na

• Sr-89 Ci na na na

• Sr-90 Ci na na na

• Total Activity Ci 2.11 E-04 2.11 E-04 B. Tritium

[H-3 Cio 1.64E-02 I 2.12E-02 1.04E-02 ] 1.24E-03 { 4.92E-02 C. Dissolved & Entrained Gases Xe-133 Ci 2.78E-05 2.78E-05 Ci Ci Ci Total Activity Ci 2.78E-05 2.78E-05 D. Gross Alpha Gross Alpha I oi na na na na Nuclide results included in Table 2.3-L3 denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 46

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

SUnits 1st Qtr nd Qtr 3rd Qtr 4th Qtr I Total D. Gross Alpha

1. Total Activity Ci Released E. Volume

1. Released Waste Liters 1.12E+07 5.32E+06 5.18E+06 3.12E+06 2.48E+07 Volume

2. Dilution Volume Liters

• During Releases

3. Dilution Volume Liters 2.51 E+07 2.46E+07 2.58E+07 3.24E+07 1.08E+08 During Period I* I II _

• "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 Released I Units 1st Qtr I 2nd Qtr I 200 3rd Qtr I 4th Qtr I Total A. Fission & Activation Products y Emitters Ci Ci Ci Total Activity Ci B. Tritium IH-3 Ci 5.24E-02 I 8.88E-02 I 5.OOE-02 I 5.07E-03 I 1.96E-01 ]

C. Dissolved & Entrained Gases y Emitters Ci Ci Ci Total Activity Ci D. Gross Alpha Gross Alpha I ci na na na na

"-"denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 48

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

'.Nuclides Released IUnits 1st Qtr I 2nd Qtr I 3rdiiI;Qtr I 4th Qtr I Total A. Fission & Activation Products y Emitters Ci Ci Ci Total Activity Ci B..Tritium

[H-3 Ci 5.OOE-02 1.90E-03 1.50E-03 [ 3.20E-03 5.66E-02 C. Dissolved & Entrained Gases V Emitters Ci Ci Ci Total Activity Ci D. Gross Alpha

[Gross Alpha I Ci na na na na

"-" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed 49

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

Table 2.1-S Unit 1 Solid Waste and Irradiated Component Shipments Table 2.2-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 3

87 ft 3 95 ft3 122 ft Steel Container 202.1 ft 3 Steel "Sea Van" 1280 ft3 Polyethylene High Integrity Containers 120.3 ft 3 132.4 ft 3 173.4 ft 3 202.1 ft3 50

Table 2.1-S Solid Waste and Irradiated Component Shipments Millstone Unit 1 January 1, 2008 through December 31, 2008 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

1. Type of Waste

a. Spent resins, Filter sludges, Evaporator bottoms, etc.

j. Disposition Units AnnualTotals Est. Total

b. Dry compressible waste, Contaminated equipment, etc.

Disposition -Units Est. Total

.. .. - Annual Totals

• .. Error Error From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Super- mr 2.7180E+00 Compaction, Incineration, etc.. Ci 1.1677E-03

c. Irradiated components, Control rods, etc.

Units AnnuaiTtals Est. Total DisoitionI oal Error

d. Other - (Grease, Oil, Oily waste)

Disposition Units' II Annual Totals Esit.Totalrr ErrorA d.

51

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

b. Dry compressible waste, Contaminated.equipment, etc.

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Super-Compaction, Incineration, etc..

Radionuclide  % of Total Curies H-3 0.29 3.3445E-06 C-14 Cr-51 Mn-54 0.29 3.4290E-06 Fe-55 47.89 5.5919E-04 Fe-59 Co-57 Co-58 0.14 1.6801E-06 Co-60 18.60 2.1717E-04 Ni-59.

Ni-63 10.52 1.2280E-04 Zn-65 Sr-89 Sr-90 0.04 4.4916E-07 Zr-95 Nb-95 Tc-99 Ru-1 03 Ru-1 06 Ag-1 1Om Sn-113 Sb-122 Sb-124 Sb-125 1-129 Cs-1 34 0.22 2.5564E-06 Cs-137 21.99 2.5673E-04 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 1.0257E-07 Pu-239 <0.01 7.0774E-09 Pu-241 Am-241 0.01 1.4872E-07 Cm-242 Cm-243 Cm-244 <0.01 5.8252E-08 CURIES (TOTAL) 1.1677E-03 52

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

d. Other- (Mixed Waste)

From Millstone Nuclear Power Station to Perma-Fix of Florida, Inc., Gainesville, FL for Stabilization, Fuel Blending, etc..

Radionuclide  % of Total Curies H-3 0.04 2.6735E-07 C-14 Cr-51 Mn-54 0.89 6.1168E-06 Fe-55 51.43 3.5295E-04 Fe-59 Co-57 Co-58 1.04 7.1186E-06 Co-60 14.42 9.8935E-05 Ni-59 Ni-63 27.00 1.8531 E-04 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Ag-110m Sn-113 Sb-122 Sb-124 Sb-125 1-129 Cs-134 1.36 9.3462E-06 Cs-137 3.71 2.5436E-05 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 1.2499E-08 Pu-239 <0.01 3.4687E-09 Pu-241 0.12 7.9208E-07 Am-241 <0.01 1.6906E-08 Cm-242 Cm-243 Cm-244 <0.01 8.2257E-09 CURIES (TOTAL) 6.8631 E-04 53

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

d. Other- (Water)

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Incineration.

Radionudide I% of Total CUries.

H-3 96.11 8.8497E-03 C- 14 Cr-51 0.01 9.7651E-07 Mn-54 0.02 1.3815E-06 Fe-55 1.80 1.6535E-04 Fe-59 <0.01 7.4708E-08 Co-57 <0.01 7.8209E-09 Co-58 0.04 3.5466E-06 Co-60 0.43 3.9810E-05 Ni-59 Ni-63 0.54 5.0061 E-05 Zn-65 <0.01 1.2132E-08 Sr-89 <0.01 1.7507E-11 Sr-90 <0.01 2.9054E-10 Zr-95 <0.01 9.4060E-08 Nb-95 <0.01 1.5516E-07 Tc-99 <0.01 1.4402E-07 Ru-103 <0.01 2.8587E-08 Ru-106 <0.01 5.0277E-08 Ag-11im <0.01 5.7820E-09 Sn-113 Sb-1 22 <0.01 6.6087E-09 Sb-1 24 Sb-125 0.14 1.3172E-05 1-129 <0.01 1.7003E-11 Cs-134 0.17 1.5201 E-05 Cs-1 37 0.73 6.7656E-05 La- 140 <0.01 2.4823E-10 Ce- 141 <0.01 6.3800E-09 Ce-144 <0.01 3.1195E-08 Hf-1 81 Ta-1 82 Th-230 <0.01 1.2001E-10 Np-237 <0.01 4.5002E-1 1 Pu-238 <0.01 1.5382E-08 Pu-239 <0.01 1.5258E-07 Pu-241 <0.01 1.0031 E-09 Am-241 <0.01 4.6916E-08 Cm-242 Cm-243 Cm-244 <0.01 2.4547E-08 CURIES (TOTAL) 9.2077E-03 54

3. Solid Waste Disposition (Shipments from Millstone)

Number of Shipments* Mode of Transportation I Destination Truck (Sole Use Vehicle) jChem-Nuclear Systems, LLC, Barnwell, SC Truck (Sole Use Vehicle) IDiversified Scientific Services, Inc., Kingston, TN Truck (Sole Use Vehicle) IDuratek, Inc., Kingston, TN 4 I Truck (Sole Use Vehicle) IDuratek, Inc., Oak Ridge, TN 1 Truck (Sole Use Vehicle) IPerma-Fix of Florida, Inc., Gainesville, FL Truck (Sole Use Vehicle) IStudsvik Processing Facility, Erwin, TN

• Indicates the number of shipments in this category which contained any unit-1 waste.

(Example: A shipment containing wastes from units 1, 2 and 3 will be counted once on each of the three unit-specific sections of this report.) 5 physical shipments were made from this station in 2008.

B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments Mode of Transportation Destination INo Shipments in 2008 N/A N/A 55

Table 2.2-S Solid Waste and Irradiated Component Shipments Millstone Unit 2 January 1, 2008 through December 31, 2008 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel) b.

c. Irradiated components, Control rods, etc.

FDisposition Units I Annual Totals Est. Total Error I

d. Other - (Grease, Oil, Oily waste) I

d. Other - (Mixed Waste)

Disposition Units Annual Totals Error Error From Millstone Nuclear Power Station to Diversified Scientific Services, Inc., m3 7.5048E-04 Kingston, TN for Fuel Blending, Incineration. Ci 7.1515E-06 2 From Millstone Nuclear Power Station to Perma-Fix of Florida, Inc., Gainesville, FL for Stabilization, Fuel Blending, etc..

m3 Ci J 1.1438E-01 6.8631 E-04 1 25%

2 d.

56

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

a. Spent resins, Filter sludges, Evaporator bottoms, etc.

From Millstone NucIlear Power Station to Duratek Inc O.ak Ridne TN fnr Sijnpr-Conmaction Incineration etc Radionuclide %of Total Curies H-3 0.25 1.8799E-03 C-14 Cr-51 0.93 7.0385E-03 Mn-54 0.79 5.9319E-03 Fe-55 44.16 3.3262E-01 Fe-59 0.04 2.9486E-04 Co-57 0.09 6.8341E-04 Co-58 6.60 4.9689E-02 Co-60 12.04 9.0703E-02 Ni-59 Ni-63 26.59 2.0026E-01 Zn-65 0.03 1.9440E-04 Sr-89 <0.01 6.6614E-05 Sr-90 0.04 3.2504E-04 Zr-95 2.65 1.9931 E-02 Nb-95 2.79 2.1030E-02 Tc-99 Ru-103 <0.01 3.4917E-05 Ru-106 0.28 2.0938E-03 Ag-110m 0.17 1.2626E-03 Sn-113 0.14' 1.0851E-03 Sb-122 Sb-124 <0.01 3.8299E-05 Sb-125 .1.35 1.0141E-02 1-129 Cs-134 0.13 9.5072E-04 Cs-137 0.37 2.7878E-03 La-140 Ce-141 Ce-144 0.09 6.6398E-04 Hf-181 <0.01 5.7224E-05 Ta-182 Th-230 Np-237 Pu-238 0.01 9.3122E-05 Pu-239 <0.01 3.4627E-05 Pu-241 0.42 3.1500E-03 Am-241 <0.01 3.9476E-05 Cm-242 <0.01 3.4766E-05 Cm-243 Cm-244 0.02 1.1813E-04 CURIES (TOTAL) 7.5323E-01 57

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

a. Spent resins, Filter sludges, Evaporator bottoms, etc.

From Millstone Nuclear Power Station to Studsvik Processino Facilitv Frwin TN for Thermal..............

Destruction p Incineration

............... t etc Radionuclide  % of Total Curies H-3 0.01 2.4052E-02 C-14 Cr-51 Mn-54 0.17 3.4695E-01 Fe-55 4.65 9.3888E+00 Fe-59 Co-57 0.13 2.5667E-01 Co-58 0.56 1.1311E+00 Co-60 5.15 1.0387E+01 Ni-59 Ni-63 53.41 1.0776E+02 Zn-65 Sr-89 <0.01 2.1254E-03 Sr-90 0.10 2.0054E-01 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Ag-110m Sn-113 Sb-122 Sb-124 Sb-125 0.72 1.4455E+00 1-129 Cs-134 9.91 1.9997E+01 Cs-137 25.16 5.0762E+01 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 1.4408E-03 Pu-239 <0.01 4.7578E-04 Pu-241 0.02 4.2023E-02 Am-241 <0.01 5.5789E-04 Cm-242 <0.01 9.0127E-05 Cm-243 Cm-244 <0.01 9.3022E-04 CURIES (TOTAL) 2.0175E+02 58

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

b. Dry compressible waste, Contaminated equipment, etc.

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Super-Compaction, Incineration, etc..

Radionuclide,  % of Total Curies C-14 Cr-51 Mn-54 0.74 1.2767E-02 Fe-55 44.68 7.7205E-01 Fe-59 Co-57 <0.01 6.9900E-05 Co-58 0.29 5.0369E-03 Co-60 13.93 2.4074E-01 Ni-59 Ni-63 33.44 5.7784E-01 Zn-65 Sr-89 Sr-90 Zr-95 <0.01 1.8500E-05 Nb-95 <0.01 8.7500E-07 Tc-99 Ru-103 Ru-106 Ag-110m Sn-113 Sb-122 Sb-124 Sb-125 0.56 9.7167E-03 1-129 Cs-134 1.40 2.4164E-02 Cs-137 4.46 7.7140E-02 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 1.9710E-05 Pu-239. <0.01 9.9624E-06 Pu-241 0.07 1.2523E-03 Am-241 <0.01 2.5607E-05 Cm-242 <0.01 1.2500E-07 Cm-243 Cm-244 <0.01 2.3471 E-05 CURIES (TOTAL) 1.7278E+00 59

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

d. Other - (Grease, Oil, Oily waste)

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Super-Compaction, Incineration, etc..

Radionuclide  % of Total J Curies H-3 0.03 5.9878E-10 C-14 Cr-51 Mn-54 2.18 4.0923E-08 Fe-55 56.33 1.0588E-06 Fe-59 Co-57 Co-58 Co-60 13.79 2.5921 E-07 Ni-59 Ni-63 24.37 4.5813E-07 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Ag-110m Sn-113 Sb-122 Sb-124 Sb-125 1.60 3.0001E-08 1-129 Cs-134 Cs-137 1.63 3.0548E-08 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 2.3510E-11 Pu-239 Pu-241 0.08 1.4959E-09 Am-241 <0.01 3.1784E-11 Cm-242 Cm-243 Cm-244 CURIES (TOTAL) 1.8797E-06 60

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

d. Other- (Mixed Waste)

From Millstone Nuclear Power Station to Diversified Scientific Services, Inc., Kingston, TN for Fuel Blending, Incineration.

Radionuclide  % of Total Curies H-3 99.28 7.1000E-06 C-14 0.72 5.1500E-08 Cr-51 Mn-54 Fe-55 Fe-59 Co-57 Co-58 Co-60 Ni-59 Ni-63 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Ag-110m Sn-113 Sb-122 Sb-124 Sb-125 1-129 Cs-134 Cs-137 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 Pu-239 Pu-241 Am-241 Cm-242 Cm-243 Cm-244 CURIES (TOTAL) 7.1515E-06 61

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

d. Other - (Mixed Waste)

From Millstone Nuclear Power Station to Perma-Fix of Florida, Inc., Gainesville, FL for Stabilization, Fuel Blending, etc..

Radionuclide  %,of Total Curies H-3 0.04 2.6735E-07 C-14 Cr-51 Mn-54 0.89 6.1168E-06 Fe-55 51.43 3.5295E-04 Fe-59 Co-57 Co-58 1.04 7.1186E-06 Co-60 14.42 9.8935E-05 Ni-59 Ni-63 27.00 1.8531E-04 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Aq-110m Sn-113 Sb-122 Sb-124 Sb-125 1-129 Cs-134 1.36 9.3462E-06 Cs-137 3.71 2.5436E-05 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 1.2499E-08 Pu-239 <0.01 3.4687E-09 Pu-241 0.12 7.9208E-07 Am-241 <0.01 1.6906E-08 Cm-242 Cm-243 Cm-244 <0.01 8.2257E-09 CURIES (TOTAL) 6.8631E-04 62

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

d. Other - (Water)

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for incineration.

Radionuclide  :,%of Total Curies, H-3 96.06 8.6927E-02 C-14 <0.01 1.7236E-08 Cr-51 <0.01 8.2212E-06 Mn-54 0.02 2.0244E-05 Fe-55 1.62 1.4685E-03 Fe-59 <0.01 8.3029E-07 Co-57 <0.01 6.7008E-08 Co-58 0.21 1.9238E-04 Co-60 0.39 3.4878E-04 Ni-59 <0.01 3.2933E-10 Ni-63 0.49 4.4465E-04 Zn-65 <0.01 3.2415E-07 Sr-89 <0.01 1.3464E-07 Sr-90 <0.01 1.8881 E-06 Zr-95 <0.01 8.0057E-07 Nb-95 <0.01 1.3876E-06 Tc-99 <0.01 7.2029E-07 Ru-103 <0.01 2.4004E-07 Ru-106 <0.01 4.1981E-07 Ag-110m <0.01 4.8301E-08 Sn-113 Sb-122 <0.01 2.1672E-07 Sb-124 Sb-125 0.12 1.0897E-04 1-129 <0.01 1.4224E-10 Cs-134 0.15 1.3569E-04 Cs-137 0.92 8.2828E-04 La-140 <0.01 3.3106E-09 Ce-141 <0.01 5.3648E-08 Ce-144 <0.01 5.1123E-07 Hf-181 Ta-182 <0.01 1.9684E-07 Th-230 <0.01 6.0015E-10 Np-237 <0.01 2.2505E-10 Pu-238 <0.01 1.2266E-07 Pu-239 <0.01 7.8786E-07 Pu-241 <0.01 9.1520E-08 Am-241 <0.01 3.7128E-07 Cm-242 <0.01 6.5423E-10 Cm-243 <0.01 3.3973E-10 Cm-244 <0.01 1.9381E-07 CURIES (TOTAL) 9.0493E-02 63

3. Solid Waste Disposition (Shipments from Millstone)

Number of Shipments*, Mode of Transportation Destination Truck (Sole Use Vehicle) Chem-Nuclear Systems, LLC, Barnwell, SC 1 Truck (Sole Use Vehicle) IDiversified Scientific Services, Inc., Kingston, TN Truck (Sole Use Vehicle) IDuratek, Inc., Kingston, TN 23 Truck (Sole Use Vehicle) IDuratek, Inc., Oak Ridge, TN 1 Truck (Sole Use Vehicle) jPerma-Fix of Florida, Inc., Gainesville, FL 3 Truck (Sole Use Vehicle) IStudsvik Processing Facility, Erwin, TN Indicates the number of shipments in this category which contained any unit-2 waste.

(Example: A shipment containing wastes from units .1, 2 and 3 will be counted once on each of the three unit-specific sections of this report.) 28 physical shipments were made from this station in 2008.

B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments Mode of Transportation . Destination No Shipments in 2008 N/A N/A 64

Table 2.3-S Solid Waste and Irradiated Component Shipments Millstone Unit 3 January 1, 2008 through December 31, 2008 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

1. Type of Waste

a. Spent resins, Filter sludges, Evaporator bottoms, etc.

Disposition Units Annual Totals Est. Total F I I I Error From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Super-Compaction, Incineration, etc..

[ From Millstone Nuclear Power Station for Thermal to Studsvik Destruction, Processing Incineration, etc..I Facility, Erwin, TN I CiJ m3 3.6759E+00 4.7317E+01 25 25%

b.

c. Irradiated components, Control rods, etc.

Disposition Units I Annual Totals I Est.Totalr d.

d. Other - (Mixed Waste)

Disposition Units Annual Totals Error Error From Millstone Nuclear Power Station to Diversified Scientific Services, Inc., m3 7.6478E-02 Kingston, TN for Fuel Blending, Incineration. "Ci 7.1634E-06 25%

From Millstone Nuclear Power Station to Perma-Fix of Florida, Inc., Gainesville, ( m3 1.1438E-01 25%

FL for Stabilization, Fuel Blending, etc..j Ci 6.8631 E-04 25%

d.

65

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

a. Spent resins, Filter sludges, Evaporator bottoms, etc.

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Super-Compaction, Incineration, etc..

Radionuclide  % of Total Curies H-3 2.54 8.8080E-02 C-14 Cr-51 Mn-54 1.74 6.0161 E-02 Fe-55 65.49 2.2664E+00 Fe-59 <0.01 1.2799E-07 Co-57 0.04 1.4449E-03 Co-58 0.72 2.4856E-02 Co-60 14.59 5.0502E-01 Ni-59 Ni-63 14.00 4.8451E-01 Zn-65 <0.01 2.0609E-05 Sr-89 Sr-90 <0.01 8.8358E-05 Zr-95 0.08 2.8887E-03 Nb-95 0.07 2.4148E-03 Tc-99 Ru-103 Ru-106 <0.01 3.3093E-04 Ag-11Orn <0.01 1.5803E-04 Sn-113 <0.01 -1.0017E-05 Sb-122 Sb-124 Sb-125 0.51 1.7674E-02 1-129 Cs-134 <0.01 1.8880E-04 Cs-137 0.12 4.2305E-03 La-140 Ce-141 Ce-144 <0.01 9.0335E-05 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 4.2003E-05 Pu-239 <0.01 1.6217E-05 Pu-241 0.06 2.2071E-03 Am-241 <0.01 1.6735E-05 Cm-242 <0.01 5.1168E-06 Cm-243 Cm-244 <0.01 5.0303E-05 CURIES (TOTAL) 3.4609E+00 66

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

a. Spent resins, Filter sludges, Evaporator bottoms, etc.

From Millstone Nuclear Power Station to Studsvik Processing Facility, Erwin, TN for Thermal Destruction, Incineration, etc..

Radiohuclide J  % of Total Curies H-3 0.03 1.3311E-02 C-14 <0.01 1.7173E-03 Cr-51 Mn-54 4.35 2.0562E+00 Fe-55 22.76 1.0769E+01 Fe-59 Co-57 0.09 4.4275E-02 Co-58 1.40 6.6292E-01 Co-60 7.11 3.3642E+00 Ni-59 Ni-63 52.91 2.5038E+01 Zn-65 Sr-89 <0.01 3.6940E-04 Sr-90 <0.01 3.1128E-03 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Ag-110m Sn-113 Sb-122 Sb-124 Sb-125 0.74 3.4980E-01 1-129 Cs-1 34 6.00 2.8367E+00 Cs-137 4.60 2.1772E+00 La-140 Ce-141 Ce-144 Hf-1 81 Ta- 182 Th-230 Np-237 Pu-238 <0.01 2.9135E-05 Pu-239 <0.01 7.1375E-06 Pu-241 <0.01 5.9703E-04 Am-241 <0.01 1.1643E-05 Cm-242 <0.01 1.2187E-05 Cm-243 Cm-244 <0.01 2.0071 E-05 CURIES (TOTAL) 4.7317E+01 67

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

b. Dry compressible waste, Contaminated equipment, etc.

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Super-Compaction, Incineration, etc..

Radionuclide j  % of Total Curies H-3 1.27 3.5073E-02 C-14 Cr-51 Mn-54 1.93 5.3115E-02 Fe-55 61.09 1.6834E+00 Fe-59 Co-57 Co-58 1.11 3.0667E-02 Co-60 12.95 3.5680E-01 Ni-59 Ni-63 20.05 5.5266E-01 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 0.06 1.7382E-03 Tc-99 Ru-103 Ru-106 AgI-110m Sn-113 Sb-122 Sb-124 Sb-125 0.54 1.4962E-02 1-129 Cs-134 0.31 8.4478E-03 Cs-137 0.69 1.8877E-02 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 1.5030E-08 Pu-239 Pu-241 Am-241 <0.01 2.1804E-08 Cm-242 Cm-243 Cm-244 CURIES (TOTAL). 2.7558E+00 68

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

d. Other - (Grease, Oil, Oily waste)

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Super-Compaction, Incineration, etc..

Radionuclide  % of Total Curies H-3 3.02 2.9759E-07 C-14 Cr-51 Mn-54 2.08 2.0515E-07 Fe-55 53.94 5.3078E-06 Fe-59 Co-57 Co-58 Co-60 13.90 1.3681E-06 Ni-59 Ni-63 23.34 2.2966E-06 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Ag-110m Sn-113 Sb-122 Sb-1 24 Sb-125 1.53 1.5040E-07 1-129 Cs-1 34 Cs-137 2.10 2.0679E-07 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 1.1786E-10 Pu-239 Pu-241 0.08 7.4991E-09 Am-241 <0.01 1.5934E-10 Cm-242 Cm-243 Cm-244 CURIES (TOTAL) 9.8402E-06 69

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

d. Other- (Mixed Waste)

From Millstone Nuclear Power Station to Diversified Scientific Services, Inc., Kingston, TN for Fuel Blending, Incineration.

Radionuclide  % of Total Curies H-3 99.19 7.1053E-06 C- 14 0.72 5.1500E-08 Cr-51 Mn-54 Fe-55 0.05 3.7900E-09 Fe-59 Co-57 Co-58 Co-60 0.01 8.1800E-10 Ni-59 Ni-63 0.01 1.0200E-09 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Ag- 11Om Sn-113 Sb-122 Sb-124 Sb-125 1-129 Cs-134 Cs-137 0.01 9.5800E-10 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 Pu-239 Pu-241 Am-241 Cm-242 Cm-243 Cm-244 CURIES (TOTAL) 7.1634E-06 70

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

d. Other- (Mixed Waste)

From Millstone Nuclear Power Station to Perma-Fix of Florida, Inc., Gainesville, FL for Stabilization, Fuel Blending, etc..

I Radionuclide  % Of Total I Curies H-3 0.04 2.6735E-07 C-14 Cr-51 Mn-54 0.89 6.1168E-06 Fe-55 51.43 3.5295E-04 Fe-59 Co-57 Co-58 1.04 7.1186E-06 Co-60 14.42 9.8935E-05 Ni-59 Ni-63 27.00 1.8531 E-04 Zn-65 Sr-89 Sr-90 Zr-95 Nb-95 Tc-99 Ru-103 Ru-106 Ag-110m Sn-1 13 Sb-122 Sb-124 Sb-125 1-129 Cs-134' 1.36 9.3462E-06 Cs-137 3.71 2.5436E-05 La-140 Ce-141 Ce-144 Hf-181 Ta-182 Th-230 Np-237 Pu-238 <0.01 1.2499E-08 Pu-239 <0.01 3.4687E-09 Pu-241 0.12 7.9208E-07 Am-241 <0.01 1.6906E-08 Cm-242 Cm-243 Cm-244 <0.01 8.2257E-09 CURIES (TOTAL) 6.8631E-04 71

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

d. Other - (Water)

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Incineration.

Radionuclide  % of Total Curies H-3 96.12 3.1828E-02 C-14 Cr-51 <0.01 2.1703E-06 Mn-54 0.03 8.6199E-06 Fe-55 1.50 4.9555E-04 Fe-59 <0.01 4.3978E-07 Co-57 <0.01 1.7382E-08 Co-58 0.08 2.7301E-05 Co-60 0.33 1.0926E-04 Ni-59 Ni-63 0.37 1.2133E-04 Zn-65 <0.01 3.2094E-07 Sr-89 <0.01 3.8615E-08 Sr-90 <0.01 3.3498E-07 Zr-95 <0.01 2.0905E-07 Nb-95 <0.01 4.2946E-07 Tc-99 <0.01 1.5359E-06 Ru-103 <0.01 6.3536E-08 Ru-106 <0.01 1.1174E-07 Ag-110m <0.01 1.2851E-08 Sn-113 Sb-122 <0.01 7.9100E-08 Sb-124 Sb-125 0.09 3.0697E-05 1-129 <0.01 3.7785E-11 Cs-134 0.12 3.9612E-05 Cs-137 1.34 4.4531E-04 La-140 <0.01 5.5199E-10 Ce-141 <0.01 1.4180E-08 Ce-144 <0.01 2.1106E-07 Hf-181 Ta-182 Th-230 <0.01 1.2800E-09 Np-237 <0.01 4.7998E-10 Pu-238 <0.01 3.4473E-08 Pu-239 <0.01 1.5550E-06 Pu-241 <0.01 2.0785E-08 Am-241 <0.01 1.0708E-07 Cm-242 <0.01 8.8649E-10 Cm-243 <0.01 4.6033E-10 Cm-244 <0.01 5.4548E-08 CURIES (TOTAL) 3.3113E-02 72

3. Solid Waste Disposition (Shipments from Millstone)

Number of Shipments* - Mode of Transportation I Destination Truck (Sole Use Vehicle) IChem-Nuclear Systems, LLC, Barnwell, SC 1 Truck (Sole Use Vehicle) IDiversified Scientific Services, Inc., Kingston, TN Truck (Sole Use Vehicle) IDuratek, Inc., Kingston, TN 20 Truck (Sole Use Vehicle) IDuratek, Inc., Oak Ridge, TN 1 Truck (Sole Use Vehicle) jPerma-Fix of Florida, Inc., Gainesville, FL 1 Truck (Sole Use Vehicle) IStudsvik Processing Facility, Erwin, TN Indicates the number of shipments in this category which contained any unit-3 waste.

(Example: A shipment containing wastes from units 1, 2 and 3 will be counted once on each of the three unit-specific sections of this report.) 23 physical shipments were made from this station in 2008.

B. IRRADIATED FUEL SHIPMENTS (Disposition) of Shipments

-Number Mode of Transportation Destination No Shipments in 2008 1 N/A N/A 73

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-GW1 - 2.4-GW4).

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-GW4, there are no pathways to any offsite drinking water supplies. The underdrain 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.

74

Table 2.4-GW1 Environmental Well Monitoring Results Laboratory Analysis Location Well ID March-08 June-08 Sept.-08 Dec.-08 MW-9B

• TI - Unit I Tank Farm MW-9D *

• TI-MW-] *

• T3 - Former Waste Oil UST T3-MW-1 * * *

• T4 - Abandoned Heating Oil UST - T4-MW-2 Building 511 T5 - Abandoned Heating Oil UST - T5-MW-1 * * *

• Building 512 T5-MW-2 * * *

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

• T7-MW-1 *

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

• T7-MW-3 *

• S I - 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-1 *

• MW-1 ** * *

• S11-MW-1 SI 1- Fueling Station S11-MW-2 *

• S13-MW-1 *

• S 13 - 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-9D *

• S12-MW-1

• Environmental Compliance S12-MW-2

• Sampling S 12-MW-3 **

MW-4A

- sampeJ~u garrunak nuLiU 3 were L~LD not samipled - notIicateU inI the rieiu 75

Table 2.4-GW2 Catch Basin/Underdrain 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 ISFS1 Yard Drain DMH#1 1 Monthly Gamma and H-3 < LLD Sumps Unit 3 Containment Underdrains Weekly Gamma and H-3 < LLD Unit 3 Foundation Underdrains** 3 SRW Sump 2 Quarterly See next page 3 SRW Sump 3 Quarterly See next page

• 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 (3SRW2 & 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.

76

Table 2.4-GW3 Underdrain Monitoring Results Sample Foundation Foundation Unit 3 Date Drain Sump 3 Drain Sump 2 RWST Pit (pCi/liter) (pCi/liter) (pCi/liter 01/01/08 12800 01/02/08 12800 01/03/08 13200 01/04/08 12000 60000 01/05/08 9980 71400 01/06/08 10600 77400 01/07/08 8040 81900 01/08/07 7650 86700 01/09/08 9050 92500 01/10/08 8280 01/11/08 8370 99000 01/12/08 7990 25000 01/13/08 7080 38200 01/14/08 7170 13300 01/15/08 7520 16800 01/16/08 7010 25700 01/17/08 6800 38400 01/18/08 8330 13200 01/19/08 6640 23200 01/20/08 7040 24700 01/21/08 6900 29000 01/22/08 8420 35000 01/23/08 5100 39200 01/24/08 4770 43000 01/25/08 4480 49200 01/26/08 4670 46200 01/27/08 4730 54900 01/28/08 5130 55500 01/29/08 3720 61700 01/30/08 4420 48800 01/31/08 4770 56200 02/01/08 4130 18400 02/02/08 4090 18900 02/03/08 3370 30200 02/04/08 3050 24200 02/05/08 4050 48100 02/06/08 3340 21100 02/07/08 2900 14400 02/08/08 2720 14500 02/09/08 3340 14400 02/10/08 3130 17000 02/11/08 2240 11600 02/12/08 2950 18200 02/13/08 2540 22600 02/14/08 3120 27200 77

Table 2.4-GW3 Underdrain Monitoring Results (cont)

Sample Foundation Foundation Unit 3 Date Drain Sump 3 Drain Sump 2 RWST Pit (pCi/liter) (pCi/liter) (pCi/liter) 02/15/08 3350 27900 02/16/08 2110 27400 02/17/08 3050 28400 02/18/08 2190 25900 02/19/08 1970 27500 02/20/08 2460 27200 02/21/08 2920 28700 02/22/08 2180 27300 02/23/08 3210 23900 02/24/08 2590 24500 02/25/08 2210 24900 02/26/08 2270 25000 02/27/08 2740 36100 02/28/08 2350 02/29/08 3490 03/01/08 2240 03/02/08 3140 03/03/08 2360 03/04/08 2310 03/05/08 3000 8370, 8720, 7970 35900 03/06/08 2820 7450 03/07/08 2850 7530 03/08/08 <1750 03/09/08 2160 03/10/08 2400 7710 03/11/08 2640 9160 03/12/08 5960 10350 03/13/08 7055 9940 13600 03/14/08 7120 8900 03/15/08 5630 8860 03/16/08 4850 7290 03/17/08 5550 6590 03/18/08 4360 6380 03/19/08 4200 4570 03/20/08 4610 4700 03/21/08 5440 4320 03/22/08 10600 4270 03/23/08 5160 3740 03/24/08 4270 3070 03/25/08 4640 3090 03/26/08 4330 2880 16400 03/27/08 3340 2950 03/28/08 3700 2560 03/29/08 3080 2610 03/30/08 3230 2810 78

Table 2.4-GW3 Underdrain Monitoring Results (cont)

Sample Foundation Foundation Unit 3 Date Drain Sump 3 Drain Sump 2 RWST Pit (pCi/liter) (pCi/liter) (pCi/liter) 03/31/08 3100 2150 04/01/08 2820 1770 04/02/08 3310 <1750 04/03/08 1930 2050 17400 04/04/08 3230 1900 04/05/08 2970 1990 04/06/08 .2530 <1750 04/07/08 3240 <1750 04/08/08 2290 <1750 04/09/08 3220 <1750 18600 04/10/08 <1750 <1750 04/11/08 2920 <1750 04/12/08 2840 1940 04/13/08 3060 <1750 04/14/08 1810 <1750 04/15/08 2380 <1750 04/16/08 2710 <1750 39800 04/17/08 <1750 <1750.

04/18/08 1900 <1750 04/19/08 2150 <1750 04/20/08 2650 <1750 04/21/08 2400 <1750 04/23/08 <1750 <1750 35600 04/25/08 2340 <1750 04/28/08 <1750 <1750 04/30/08 <1750 <1750 152000 05/02/08 2480 <1750 05/05/08 2350 <1750 05/07/08 2610 <1750 140000 05/09/08 2270 <1750 05/12/08 2310 <1750 05/14/08 1910 <1750 05/16/08 1960 <1750 05/19/08 2270 <1750 05/21/08 2310, <1750 225000 05/27/08 1830 <1750 79

Table 2.4-GW3 Underdrain Monitoring Results (cont)

Sample Foundation Foundation Unit 3 Date Drain Sump 3 Drain Sump 2 RWST Pit (pCi/liter) (pCi/liter) (pCi/liter) 05/28/09 <1750 50000 05/30/08 1970 <1750 06/03/08 <1750 <1750 06/09/08 2150 <1750 06/11/09 <1750 7510000 06/16/08 <1750 <1750 06/19/09 <1750 4180000 06/23/08 <1750 <1750 06/25/09 <1750 4760000 06/30/08 1790 <1750 17700000 07/03/08 1760 <1750 07/07/08 1940 <1750 19400000 07/10/08 <1750 <1750 07/14/08 2210 <1750 07/16/09 <1750 27000000 07/17/09 <1800 07/21/08 <1800 <1800 07/23/09 9100000 07/28/08 <1750 <1750 3870000 07/31/08 <1750 <1750 08/04/08 <1750 <1750 08/05/09 18600000 08/08/08 <1750 <1750 08/11/09 <1750 8570000 08/14/08 <1750 08/18/08 <1750 <1750 11500000 08/22/08 <1750 08/25/08 <1750 <1750 18500000 09/02/09 <1750 15000000 09/05/08 <1750 09/08/08 <1750 <1750 09/09/09 1680000 09/15/08 <1750 <1750 09/18/09 <1750 2520000 09/24/09 <1750 7770000 i;

80

Table 2.4-GW3 Underdrain Monitoring Results (cont)

Sample Foundation Foundation Unit 3 Date Drain Sump 3 Drain Sump 2 RWST Pit (pCi/liter) (pCi/liter) (pCi/liter) 09/29/08 2120 <1750 878000 10/01/08 2580 <1750 10/06/08 2280 <1750 10/13/08 1940 <1750 10/17/09 <1750 5390000 10/20/08 3740 <1800 5640000 10/21/08 3530 10/22/08 3430 10/23/08 4330 10/24/08 3410 10/27/08 2870 <1750 2070000 10/29/08 3120 10/31/08 3530 <1750 11/03/08 3640 <1750 11/06/08 2930 <1750 11/07/08 614000 11/18/08 436000 12/01/08 2450 <1750 12/04/08 2250 <1750 12/17/08 3200 81

Table 2.4-GW4 RWST Yard Sample Results DBG Onsite Results Vendor Results Location Sample

• Date (pCi/L) (pCi/L)

Type (ft) Sampled gamma H-3 gamma H-3 DP 001 soil 0 01/15/2008 <LLD <LLD DP 001a soil 0 01/16/2008 < LLD <LLD DP 001b soil 0 01/22/2008 < LLD <LLD soil 5 01/22/2008 < LLD <LLD soil 10 01/22/2008 < LLD <LLD water 11 01/23/2008 < LLD 2800 <LLD 3150 water 02/04/2008 < LLD 2930 water 02/14/2008 < LLD 3280 <LLD 3190 water 02/21/2008 3240 water 03/03/2008 <LLD 5780 water 03/04/2008 <LLD 6040 6730 water 03/05/2008 <LLD 5910 water 03/10/2008 8740 water 03/11/2008 7920 water 03/12/2008 7080 water 03/13/2008 6570 water 03/14/2008 6550 water 03/17/2008 6720 water 03/20/2008 8460 water 03/24/2008 10900 water 03/25/2008 9380 water 03/26/2008 9840 water 03/28/2008 7880 water 03/31/2008 7840 water 04/02/2008 6080 water 04/14/2008 10300 water 04/16/2008 8730 water 04/18/2008 8260 water 04/21/2008 7410 water 04/30/2008 14200 water 05/02/2008 NSA water 05/07/2008 NSA water 05/15/2008 NSA water 05/30/2008 NSA water 06/13/2008 NSA water 07/01/2008 NSA water 07/10/2008 14300 water 09/04/2008 6810 water 10/22/2008 5980 DP 002 soil 0 01/14/2008 <LLD <LLD DP 002a soil 0 01/18/2008 <LLD <LLD soil 5 01/18/2008 <LLD <LLD soil 10 01/18/2008 <LLD <LLD water 14.5 01/24/2008 <LLD 2600 <LLD 2690 water 02/04/2008 <LLD 3350 82

Table 2.4-GW4 RWST Yard Sample Results (cont)

DBG Onsite Results Vendor Results Location Sample

• Date (pCi/L) (pCi/L)

Type (ft) Sampled gamma H-3 gamma H-3 water 02/14/2008 < LLD 4310 3700 water 02/21/2008 3380 water 03/03/2008 <LLD 2650 water 03/10/2008 2430 water 03/17/2008 2740 water 04/02/2008 4160 water 04/14/2008 5360 water 04/21/2008 5820 water 04/30/2008 5900 water 05/07/2008 5690 water 05/15/2008 NSA water 05/30/2008 NSA water 06/13/2008 5690 water 07/01/2008 6530 water 07/10/2008 5730 water 08/01/2008 8880 water 09/04/2008 7020 water 10/21/2008 6240 DP 003 soil 0 01/14/2008 <LLD <LLD soil 5 01/15/2008 <LLD <LLD DP 003a soil 0 01/22/2008 <LLD <LLD soil 5 01/22/2008 <LLD <LLD soil 10 01/22/2008 <LLD <LLD water 16.5 01/24/2008 <LLD <LLD 1870 water 02/04/2008 2100 water 02/14/2008 2350 water 02/21/2008 2480 water 03/03/2008 <LLD 2460 water 03/10/2008 <LLD water 03/17/2008 <LLD water 04/02/2008 <LLD water 04/14/2008 1920 water 04/21/2008 2760 water 04/30/2008 3500 water 05/07/2008 3080 water 05/15/2008 3450 water 05/30/2008 3490 water 06/13/2008 3550 Ed water 07/10/2008 1940 water 08/01/2008 <LLD water 09/04/2008 2010 water 10/21/2008 3940 DP 004 soil 0 01/15/2008 <LLD <LLD soil 5 01/16/2008 <LLD <LLD DP 004a soil 0 01/18/2008 <LLD <LLD soil 5 01/18/2008 <LLD <LLD soil 10 01/18/2008 <LLD <LLD OP 005 soil 0 01/15/2008 <LLD <LLD 83

Table 2.4-GW4 RWST Yard Sample Results (cont)

DBG Onsite Results Vendor Results Location Sample

• Date (pCi/L) (pCi/L)

Type (ft) Sampled gamma H-3 gamma H-3 DP 006 soil 0 01/17/2008 <LLD <LLD DP 006a soil 5 01/22/2008 <LLD <LLD soil 10 01/22/2008 <LLD <LLD DP 007 soil 0 01/17/2008 <LLD <LLD soil 5 01/17/2008 <LLD <LLD water 9 01/18/2008 <LLD 1970 <LLD 1660 water 02/14/2008 <LLD 2010 <LLD 3400 water 1 02/21/2008 <LLD 2280 water 03/03/2008 <LLD <LLD water 03/10/2008 1910 water 03/17/2008 <LLD water 04/02/2008 3320 water 04/14/2008 3450 water 04/21/2008 3190 water 04/30/2008 2470 water 05/07/2008 2960 water 05/15/2008 2300 water 05/30/2008 NSA water 06/13/2008 NSA water 06/13/2008 NSA water 08/01/2008 3210 water 09/04/2008 NSA DP 008 soil 0 01/16/2008 <LLD <LLD DP 008a soil 0 01/17/2008 <LLD <LLD soil 5 01/17/2008 <LLD <LLD DP 009 soil 0 01/15/2008 <LLD <LLD soil 5 01/16/2008 <LLD <LLD DP 010 soil 0 01/22/2008 <LLD soil 5 01/22/2008 <LLD <LLD DP 011 soil 0 01/23/2008 <LLD <LLD soil 5 01/23/2008 <LLD <LLD soil 10 01/23/2008 <LLD <LLD water 16 01/25/2008 <LLD <LLD 1780 water 02/06/2008 <LLD water 02/14/2008 <LLD water 02/21/2008 <LLD water 03/03/2008 <LLD water 03/10/2008 <LLD 8660 <LLD 10100 water 03/11/2008 <LLD 10600 water 03/12/2008 7010 water 03/13/2008 7460 water 03/14/2008 9360 water 03/17/2008 4790 water 03/20/2008 13900 water 03/24/2008 17900 84

Table 2.4-GW4 RWST Yard Sample Results (cont)

Onsite Results Vendor Results Location Sample DBG* Date (pCi/L) (pCi/L)

Type (ft) Sampled gamma H-3 gamma H-3 water 03/25/2008 17700 water 03/26/2008 13800 water 03/28/2008 11100 water 03/31/2008 5900 water 04/02/2008 3340 water 04/14/2008 1840 water 04/21/2008 2240 water 04/30/2008 <LLD water 05/07/2008 <LLD water 05/15/2008 <LLD water 05/30/2008 <LLD water 06/13/2008 NSA water 07/01/2008 NSA water 08/01/2008 NSA water 09/04/2008 <LLD water 10/24/2008 <LLD MW-GPI-6 water 10/24/2008 <LLD water 12/02/2008 <LLD water 12/04/2008 <LLD

• DBG = depth below grade ** NSA = no sample available 85

3.0 Inoperable Effluent Monitors During the period January 1 through December 31, 2008, 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 - None 86

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

Unit 1 was shut down November 11, 1995 with a cessation of operation declared in July 1998.

Unit 2 operated with a DER capacity factor of 85.6% and Unit 3 operated with a DER capacity factor of 87.6%

The power histograms for 2008 are on the following pages.

87

MP2 - CYCLE 18 & 19 POWER HISTORY YEAR 2008 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 2 U Ic 90 + 7 8 9 80 +

-1 70 +

iz 60 0

0.

50 40 -1

1. POWER REDUCED TO 97% TO PLACE THE 'D' CW PUMP IN-SERVICE

2. POWER REDUCTION TO REMAIN BELOW THE NPDES LIMIT FOR UNIT DELTA-T 30+ 3. SHUTDOWN FOR 2R18 REFUEL OUTAGE (4/6108)

4. AUTO TRIP FROM 100% POWER (5/22/08) - LIGHTING STRIKE ON TRANSMISSION LINES TRIPPED OUTPUT BREAKERS

5. AUTO TRIP DURING PLANT STARTUP (5/24108) - RSST BREAKERS OPENED UNEXPECTEDLY 20+ 6. MANUAL TRIP (6/28/08) - BOTH MAIN FEED PUMPS TRIPPED DUE TO LEVEL OSCILLATIONS IN THE FEEDWATER HEATER SYSTEM

7. POWER REDUCTION TO 90% (9/13108) TO FACILITATE TURBINE VALVE TESTING

8. POWER REDUCTION TO 95% (10/9108) FOR 'B' CWýPP EDUCTOR PIPING REPAIRS

9. DOWN POWER TO 98% (12/6/08) TO FACILITATE TURBINE VALVE TESTING 10+ 10. DOWN POWER TO 94.5% (12/27/08) DUE TO 'A' CW PP TRIP - NPDES PERMIT LIMITED 0 CN - -

00 0D C) CCY) C ("r -) Cc)

CO 1* 4 c"J m m 6)(

Cý3~ Z?

t I)

Z(N I)

Z?5 (.( r-- rz 00 00 0 0) 0 C -(N (

0 0 C) CD C C C 0 CD C C C) 0 0 c) 0 88

MP3 - CYCLE 12 & 13 POWER HISTORY

-YEAR 2008 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 I

I 90 +

5 'U 8

2 45 6 1 3 80 3R12 70 o 60 0

50 40-30 1. POWER REDUCED TO 90% TO FACILITATE MAIN TURBINE CONTRL VALVE TESTING AND EHC TROUBLESHOOTING.

2. POWER REDUCED TO 90% TO FACILITATE MAIN TURBINE CONTRL VALVE TESTING 20 -- 3. DOWN POWER TO 90% (4/5/08) TO SECURE THE 'B' TDFP (LEAKING SEAL) AND PLACE MDFP IN-SERVICE

4. DOWN POWER TO 92% (617/08) TO REMOVE 'A' HEATER DRAIN PUMP TO REPLACE CONTROLLER FOR THE 4A NLCV

5. DOWN POWER TO 93% (6/8/08) WHEN THE 'B' MOISTURE SEPARATOR DRAIN PUMP TRIPPED ON LOW TANK LEVEL 6&DOWN POWER TO 90% (7/19/08) TO FACILITATE TCV TESTING & PLACE THE 'B' TDFP IN-SERVICE (REMOVE MDFP FROM SERVICE) 10-

7. RX TRIP FROM 30% POWER (10/11108) DUE TO SG LEVEL - WHILE TAKING THE UNIT OFF-LINE FOR 3R12

8. DOWN POWER to 94% TO FACILITATE REPAIRS TO A & B DSM NLCV(S) 0 I mam-A-Im I- - --- Rpm 00 oo 0

IC) u2 U) ý2 U) 0 Ct v) C1 c) ' )

0-C "T 0) m' 00 cy) C)

C.)

= C\J C'o C' 4' U ) U) CO

) CO iN- 05 00 aO) a) CD C (J CD C) C C C C C C C Cz C) C C C C) C CD 89

5.0 Errata 5.1 Table 2.3-Al from the 2007 Radioactive Effluents Release Report had missing numbers for Particulates Total Activity Released and the Average Period Release Rate for the 1 st quarter 2007. A revised table on the following page shows the correct numbers highlighted.

90

Table 2.3-Al Millstone Unit No. 3 Airborne Effluents - Release Summary Units ", st Qtr I 2nd Qtr I 3rd Qtr

• 4th Qtr Total D. Gross Alpha

1. Total Activity Ci Released

"-" denotes less than Minimum Detectable Activity (MDA) 91

6.0 REMODCM Changes The description and the bases of the change(s) for each REMODCM revision are included here in Volume 1 of the Radioactive Effluent Release Report. In addition, a complete copy of the REMODCM revision(s) for the calendar year 2008 is provided to the Nuclear Regulatory Commission as Volume 2 of the Radioactive Effluent Release Report.

In 2008, there was 1 revision made to the Millstone REMODCM:

Rev Effective Date 26-00 June 2008 92

6.1 REMODCM - Description of Change(s)

Change Request #: 08-01 I. Description of changes (include markup pages)

Originator name (Print):

Section/Page Section Title and Description of Change with Basis Fig. I.C-2/21 "Simplified Liquid Effluent Flow Diagram Millstone Unit 2" Revise figure as shown in markup.

1) Deletes Degasifier which is no longer usable.

2) Adds more detail to show:

a) cross connection between T21 and T14A/B, b) use of temporary filters between T14A/b and T23 and T15A/B, c) connection between T23 and T15A/B and between T14A/B and Ti 5A/B, d) portable demin T141 A/B/C with option of bypassing ion exchanger T24, and e) cross connection between T21 and T24.

3) Adds footnote that HP decon facility sump is collected in totes and Fig. I.C-3/22 shipped offsite.

Revise figure as shown in markup.

1) Adds more detail to show:

a) optional filter after LWS-TK1 A/B, b) cross connections between Boron Recovery System and High Level Waste System, and c) Boron Recovery Tank.

2) Corrects 'CVCS Ion Exchanger (CHS DEMN 1A/B)' to 'Cesium Ion Exchanger (BRS DEMN 1A/B).'

3) Deletes LWS-FLT 2A/B.

4) Deletes arrow showing direct discharge from Filter LWS-FLT1 to discharge tunnel.

5) Replaces the "Chemical and Volume Control System" box with additional detail to show more components of letdown flow to the degasifier and Table I.D- volume control tank.

1/23 Change frequency of analyses for Balance of Plant Vent Sr-90, Gross alpha from 'Twice per month' to 'Quarterly composite.' This corrects a copy error from a previous REMODCM change. The frequency of analysis for the gamma emitters from the block above was inadvertently copied.

93

Change Request #: 08-01 Section/Page Section Title and Description of Change with Basis Table I.D- Change frequency of analyses of gaseous grabs of containment vent to "Same as 2/25 sample frequency for vent samples." In a prior revision of the REMODCM the Table LD-3/28 frequency had been changed from weekly to every two weeks but the analysis frequency was not changed. to correspond to sampling.

Delete the name "Hogger" because it is a slang term which is no longer used for the Table I.D-3/28 containment vacuum pump and it is not needed in this context.

Fig. I.D-3/36 Combine the four containment means of release in column one into one source Table I.D-3/28 release point, as 'Containment.' Drawdown, purge, and vent modes of release are already listed in the 2 "d column, "Sample Type and Frequency." In the 2 "d column, the words 'via. air ejector' are added in parentheses to clarify the drawdown mode of release and venting is redefined as "releases to maintain sub-atmospheric pressure (via containment vacuum pumps)." The redefinition of venting makes it technically more accurate as this mode of release is not a vent process.

In 2.a. 1 after the words "not routinely operating" add the words "or being bypassed.",

I.D.2/31 This is needed because some of the processing equipment could be operational while still being bypassed.

From Unit 3 gaseous processing equipment delete the Fuel Building Ventilation I.D.2/32 Filter. This filter is not normally in service for the purpose of extending filter life in case of an accident or condition leading to a large release from the Fuel Building.

Releases from the Fuel Building are normally below detectability. By removing the Fuel Building Ventilation Filter from list of processing equipment it would avoid an REMODCM requirement to use the filter in order to prevent a very small effluent dose.

Added a footnote to each diagram to indicate which flow paths are used during an Fig. I.D-2/35 accident. This is to distinguish use of these flow paths from normal effluent use.

Fig. I.D-3/36 Changing reporting requirement from "30 days from the end of the affected calendar!

I.E.1/38 quarter" to "30 days from receipt of sample results." Some samples are submitted to the contractor analysis lab at the end of the quarter and take 30 days for analyses.

Change No. of Locations of Well Water from 2 to 6. This change will support the Table I.E- 1/39 Groundwater Protection Initiative Program as recommended by NEI.

94

Change Request #: 08-01 Section/Page Section Title and Description of Change with Basis Table I.E-2/41 Add four onsite wells Nos. 79-I, 80-I, 81-1, and 82-I. This change will support the Groundwater Protection Initiative Program as recommended by NET.

Table I.E-3/44 In Footnote A delete the words "(i.e., seawater)." This would allow the exception to use a value of 30,000 pCi/l for non-drinking water samples taken from on-site wells. This change will support the Groundwater Protection Initiative Program as recommended by NEI.

II.C.6/60 1) Add the sentence "Use of a code using the methodology given in Regulatory Guide 1.109 is described in MP-22-REC-GDL03, 'Liquid Dose Calculations -

DOSLIQ' " and

2) In the last sentence of this section replace 'LADTAPII' with 'liquid dose calculations.'

This change is needed to provide a complete description of the methodology presently being used to calculate dose from release of radioactivity in liquid effluents.

II.D.4.c. 1/75 In definition for CEN add the words "to the Millstone Stack" after "from the reactor plant gaseous vents" to distinguish the pathway of the release from release to the Unit 3 vent.

II.D.4.c.2 & c.3 Delete these two sections. Dose projections methodologies shown in

/75-76 these sections are for the purpose of implementing the requirements in Section I.D.2 when certain processing equipment is not operating. There is no processing for Steam Generator Blowdown Tank Vent; therefore, Section I1.D.4.c.2 is not needed. With deletion of the Fuel Building ventilation filter (see change to Section I.D.2, p.32 above) there is no processing of ventilation releases; therefore, Section ll>D.c.3 is not needed.

II.D.5/76 At the end of Section ll.D.5 (p. 76), add the sentences:

1) "The use of this code and the input parameters are given in MP REC-GDL04, 'Gaseous Dose Calculations - GASPARII" and

2) "Use of a code using the methodology given in Regulatory Guide 1.109 is described in MP-22-REC-GDLO5, 'Gaseous Dose Calculations -

DOSAIR.'"

This change is needed to provide a complete description of the methodology presently being used to calculate dose from release of radioactivity in gaseous effluents.

95

Change Request #: 08-01 Section/Pae Section Title and Description of Change with Basis IJ.F.5/92 Add the following sentence at end of 2nd paragraph: "In Calculation RERM-02906R2 a maximum allowable setpoint of 42,000 cpm is determined using vent flow of 100,000 ft3/min." This would identify the setpoint in the REMODCM without having to refer to a calculation.

Tables I1.C-3, 1) In Action B add the sentence "Operation of the auxiliary sampling IV.C-3, V.C-3 equipment shall be verified every twelve hours." This is needed to

/104, 127, 149 ensure that the sampling is maintained during the monitor inoperability.

2) Add the following sentence at end of Action D: "Sample flow rate need not be estimated if the auxiliary sampling equipment of Action B is in use." This will relieve operators of having to estimate flow rate when not needed (see CR-07-10892).

III.D. I .b/107 At the end of the first surveillance requirement add the words "once every 31 days." This would make the requirement consistent with the same requirement for Unit 2 (see page 130) and for Unit 3 (see page 153).

IV.B.2/119 Change definition of "Dose Equivalent 1-131" to support AST licensing change. This definition would be worded the same as for Unit 3 (see page 140).

11/53 Correct the following typographical errors:

II.C.5/58 1) On top of page, change "CDCM" TO "ODCM" in title line.

2) On bottom of page, delete asterices which appear in definitions for D'MW II.F.8/93 and D'Mo.

3) Change 'Rev. 0' to 'Rev. 1' in reference to Calculation RERM-01946-R3.

96

Serial No.09-200 Docket Nos. 50-245 50-336 50-423 License Nos. DPR-21 DPR-65 NPF-49 ATTACHMENT 2 2008 RADIOACTIVE EFFLUENT RELEASE REPORT VOLUME 2 MILLSTONE POWER STATION UNITS 1, 2, AND 3 DOMINION NUCLEAR CONNECTICUT, INC. (DNC)

Millstone Power Station 2008 Radioactive Effluents Release Report Volume 2 Gaseous Effluents VEi Dominion Nuclear Connecticut, Inc.

MILLSTONE UNIT LICENSE DOCKET 1 DPR-21 50-245 2 DPR-65 50-336 5Dominion 3 NPF-49 50-423

MILLSTONE POWER STATION STATION PROCEDURE A

• ~Th~A~S I VV

-~Al RadiologicalE ent Monitor-ng and Off-Site Dose Calulation Manual. (RMODCM)

MP-22-REC-BAPO1 Rev. 026-00 Approval Date: 6/10/08 Effective Date: Effctie Dte:6/19/08

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

TABLE OF CONTENTS

1. Radiological Effluent Monitoring Manual (REMM) .......................... 7 I.A. Introduction ............................................... 7 I.B . R esponsibilities ........................... .......................... 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 Use ................. ................................ 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 Use ................................................. 33 I.E. Radiological Environmental Monxitoring ........................... 37

1. Sampling and Analysis ............................ 37

2. Land Use Census ............................................... 47

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

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

1. Annual Radiological Environmental OperatingReport ............. 49

2. Radioactive Effluent Release Report ................... .......... 50 II Off-Site Dose Calculation Manual (CDCMV1) ................................. 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-BAPOI s TO T-HIK `A..REVIEW'Rev.026.-00 I of 165

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. Reserved .................... ............................. 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 - CN D 245 ......... ..................................... 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 Limitation ..................................... 89

13. Bases for Liquid Monitor Setpoints ............................. 89 11.E Gaseous Monitor 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.- R eserved ...................................................... 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-RE1OB ................... 92

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

9. Bases for Gaseous Monitor Setpoints ........ ................... 92 MP-22-REC-BAPO1 STOP T.ACT' EV!EW Rev. 026-00 2 of 165

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 Monitoiing 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 IL.F Bases ....................................................... 112 IV REMODCM Unit Two Controls .......................................... 117 IV.A. Introduction ............ " . ......... *......... ........... 117 IV.B. Definitions, Applicability and Surveillance Requirements ...... 117 IVC. 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 1.E. Total Radiological Dose From Station Operation ........................ 135 IV E B ases .............................. .............................. 135 V REMODCM Unit Three Controls " ... 139 V.A. Introduction.................................. ................. 139 V.B. 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 VE. Total Radiological Dose From Station Operations ..................... 158 V .F B ases ......... ............. ................................... 1,58

. MP-22-REC-BAPO1 STOP THINK AOT1 REVIEWi Rev. 026 -00 3 of 165

TABLES AND FIGURES TABLES Table I.C.- 1, "Millstone Unit I Radioactive Liquid Waste Sampling and Analysis Program " . .... ................................ ............ 9 Table I.C.-2, "Millstone Unit 2 Radioactive Liquid Waste Sampling and Analysis Program "............................. ................... 11 Table I.C.-3, "Millstone Unit 3 Radioactive Liquid Waste Sampling and Analysis Program" ........................................... 14 Table I.D. -1, "Millstone Unit 1 Radioactive Gaseous Waste Sampling and Analysis Program" .......................................... 23 Table ID.-2, "Millstone Unit 2 Radioactive Gaseous Waste Sampling and Analysis Program". ..................................... 25 Table I.D.-3, "Millstone Unit 3 Radioactive Gaseous Waste Sampling and Analysis Program ...................... ..................... 28 Table I.E.-i, "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 R eference .. .. .................................................

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 IV.C.-4, "Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements" .. ................. ............. 128 Table V.C.-I, "Radioactive Liquid Effluent Monitoring Instrumentation" ... 143 Table VC.-2, "Radioactive Liquid Effltient Monitoring Instrumentation Surveillance Requirements" ......... ........................... 146 Table VC.-3, "Radioactive Gaseous Effluent Monitoring Instrumentation" .. 148 Table VC.-4, "Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirem ents" . .................. ........................ 150 MP REC-BAPOl STOP .. IE- ;EVIEW A Rev. 026-00 4 of 165

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 Unit One" ................................................. 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 IIID.- 1, "Site Boundary for Liquid and GaseousEffluents".. ........ 111 Figure IVD.-1, "Site Boundaiy for Liquid and Gaseous Effluents"........... 134 Figure V.D.- 1, "Site Boundary for Liquid and Gaseous Effluents"............ 157 MP-22-REC-BAPO1 Rev. 026-00 5 of 165

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-22-REC-BAPO1 ST Q THINK ACT .. EV\ Rev. 026- 00 6 of 165

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 arid the Radioactive Effluent Release Report.

MP-22-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 Site Operations 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-22-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 TFable 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 11 to assure that the concentrations at the point of release are MP REC-BAvP01 STOP THI-ACT RIEViEW Rev. 026-00 7 of 1.65

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

2, and Section VD.1.a. for Millstone Unit No. 3).

MP-22-REC-BAPOI SýTOP !THINK ACT REVIEW Rev. 026-00 8 of 165

- 1. -, --- ' . -

Table I.C.-1 Millstone Unit I 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

([lCi/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 1x 10-5 Prior to initial batch re- H-3 1 x io5 lease from any one source and monthly composite thereafterc Grab sample prior to Prior to initial batch re- Gross alpha Ix 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 6 ite thereafter Fc-55 . 1x10 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):

LLD 4.66 Sb (E)(T1)(2.22xlO 6)(ye - ,t)

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)*

" 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)

• 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 ni. MP-22-REC-BAPO1 STOP T1 INK Rev 026-06 firV"ffA 9 of 165

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.

MP-22-REC-BAP01 T iHINK :ACT R.EVIE W*Rev. 026-00 7..1 lof 165

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

A.Batch ReleaseB.

1.Clean Waste Moni- Grab sample prior to Prior to each batch Principal Gamma 5 X10-7 tor Tank, Aerated each batch release release Emittersc.

Waste Monitor Tank and Steam 1-131 1x 106 Generator BulkD. Ce- 144 5 x 10--6 Dissolved & 1 x 10-.

Entrained GasesK 2.Condensate Monthly H-3 I x 10-5 Polishing Facility CompositeE,G-

- Waste Quarterly Grossalpha l""

Neutralization urterlh x 10-7 SumpE. CompsiteFG Sr-89, Sr-90 5 x 10-8 Fe-55 1x0-6 B.Continuous Release 1.Steam Generator Daily Grab Sample'-& Weekly Principal Gamma 5 x 10-7 BlowdownH- prior to aligning to CompositeE,, Emittersc.

2.Service Water Long Island Sound for 1-131 1 x 10-6 EffluentW Ce- 144 5 x 10-6 3.Turbine SumpsL Monthly Grab Monthly Dissolved & I x 10-5 Sample Entrained GasesK 4.RBCCW SumpM Weekly Grab or Corn- Monthly H-3N_ 1 x'10-5 posite CompositeEo-Weekly Composite Quarterly Gross alpha I x 10-1 CompositeE.G Sr-89, Sr-90 5 x 10.8 Fe-55 I x 10-6 TABLE !.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)(T/)(2.22x I 06)(Ye MP-22-REC-BAPO1 STI' THINK A REVIE, Rev. 026-00 I1I of 165

it Where:

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

" Sb is the standard deviation of the background counting rate or of the counting rate of 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 4Ci

• Y.is the fractional radiochemical yield (when applicable)

• X,is the radioactive decay constant for the particular radionuclide At is the elapsed time between midpoint of sample collection and midpoint of counting time A

It should be recognized that the LLD is defined as an priori (before the fact) limit representing the capability of a measurement system and not as an a posterioei (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 berecirculated 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 pCi/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 1xCi/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/mI,-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. 026-00 12 of 165

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 [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 IACi/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.

I. 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 l.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 [tCi/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 liCi/ml and for gross alpha 1 x 10-8 pCi/ml.

M. For the RBCCW Sumj:

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 [tCi/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.

IMP-22-REC-BAP01.

"-SQP 1] AC REVIEW Rev. 026-00 13 of 165

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

A.Batch ReleaseB-I.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 & 1 x 10-5 Entrained GasesK_

2.Waste Test Tanks, Monthly H-3 1 x 30-5 Low Level Waste Composite',G.

Tank, Boron Test Tanks and Steam Quarterly . Gross alpha 1 x 10.'7 Generator Bulk D_ CompositeE'G Sr-89, Sr-90 5x 0.8 6

Fe-55 1 x 10-B.Continuous Release I.Steam Generator Daily Grab SampleL Weekly Principal Gamma 5 x 10-7 BlowdownH- .CompositeG. ErmittersC.

2.Service Water Ef-' 1-1.31 1 x 10-6 fluentJ 6 Ce-144 5 x 10-3.Thrbine Building Monthly Grab Monthly Dissolved & 1 x 10-5 SumpsL Sample Entrained GasesK-Weekly Grab or Corn- Monthly H-3m- I x 10.5 posite CompositeF.,G.

Weekly Composite Quarterly Gross alpha 1 x 10-7 CompositeEG. 8 Sr-89, Sr-90 5 x 10.

6 Fe-55 1 x 10-

• TABLE I.C. 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-22-REC-BAPO1 STOP  :'IIIK -'ACT REViEV Rev. 026- 00 14 of 165

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

LLD= 4.66 Sb (E)( V)(2.22x106)( Ye6- ,I)

Where:

• LLD is the lower limit of detection as defined above (as ttCi 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 RCi

• Y is the fractional radiochemical yield (when applicable) a 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 Itshould 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. If the 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 [tCi/ml. The principal gamma emitters for which this LLD applies are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-6O, 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 lpCi/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 OCi/ml, THEN the sampling and analysis schedule for gross alpha, Sr-89, Sr-90, Fe-55 are not required.

MP-22-REC-BAP01 S-T.oi:KINK 'ý' -ACT VRev. 026-00 RE VIEW*I i15 of 165

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 [tCi/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 [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.

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

Daily grabl 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:

7 IF a weekly gamma analysis does not indicate a gamma activity greater than 5 x100-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 BuildingSmp 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 [Ci/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 x10-7 ltCi/ml and for gross alpha 1 x 10-8 pCi/ml.

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

MP-22-REC--BAP01 STO.P T... .N.K A V1EW Rev. 026-00 16 of 165

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, IF 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 Ic.

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.

.. TO.~.. .MP-22-REC-BAP01 THINPIK ACT EEWI Rev. 026-00 lof 17 165

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 (T11) OR Purification ion exchanger (T10A or TIOB) 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 [LCi/ml blowdown or tritium activity exceeds 0.02 RCi/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.

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.*;[ Rev. 026-00 18 of 165

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.1.a., IV.D.I.a., and V.D.I.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 effluentconcentrations 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 I, 10 CFR 50 for liquid effluents following the guidance given in NUREG - 1301.

MP-22-REC-BAPO1 STOP THINK .'- ACT REi*.VIEW Rev. 026-00 19 of 165

Figure I.C.-1, "Reserved MP-22-REC-BAPOI

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Figure 1.C.-3, "Simplified Liquid Effluent Flow Diagram Millstone Unit MP-22-REC--BAPO1 ST.O THINK

• iP AICT :RE Rev. 026-00 22 of 165

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

Table 1J.). -I Millstone Unit 1 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Release Sample Type and Minimtun Type of Activity Lower Limit Point or Source Frequency. Analysis Analysis of Detection Frequency (LLD)A (1lCi/mi)

A.Spent Fuel Pool Monthlyt" - Gaseous Monthly Kr-85 Ix 10-4 Island Vent Grab Sample H-3 1 x 10-6 Twice per month Principal Particulate 1x -11 ContinuousB-,E Gamma EmittersC -.

Particulate Sample (with half lives greater than 8 days)

ContinuousB-,E. Quarterly Sr-90, Gross alpha .1x 10-11 Particulate Sample Composite ContinuousB',E Continuous Kr -85 -1x 10-6 Noble Gas Monitor B.Balance of Twice per month Principal Particulate 1 x 10-11 Plant Vent Gamma EmittersC -

ContinuousC.'L. (with half lives greater Particulate Sample than 8 days)

Quarterly Sr-90, Gross alpha I 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.

STOP;* . MP-22-REC-BAPOI 4N-IK AC VR ev. 026-00 23 of 165

C. For particulate samples, the LLD will be 1 x 10-11 pCi/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. IF there 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 IlCi/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.

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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 (PQCi/il)

A.Batch Release Principal Particulate Y1 1-4

• 1.Waste Gas Stor- Gaseous Grab Prior to each Each Tank Discharge Gamma Emitters13 age TankH Waste Gas Tank Discharge H-3 "" TX 1(-6 B.Containruent&Aux Building Releases

.Containment Gaseous Grab of purges 1. Prior to purge Principal Gamma i X 10.4 Purge and vents 2. Same as sample EmittersB 2.Containment 1. Prior to Each Purge' frequency for vent Venting 2. Ever), two weeks for samples.

Venting -

3.Open Equip- Monthly H-3 I X 10-6 ment Hatch During Outages 4.Spent Fuel Pool Continuous Particulate for Weekly Particulate Gamma NA Open Equipment Hatch emitters for 1/2 hr during Outage count (I- 131, others with half-life great-er than 8 days)

Continuous Charcoal for Weekly 1-131 and 1-133 for NA Open Equipment Hatch one hour count during Outage & Aux Bldg Rollup DoorL Gaseous Grab at Equip- Daily Noble Gases - I X10-4 ment Hatch & Aux Bldg Gross Activity Rollup DoorL C.Continuous Release 1.Vent Monthly - Gaseous Grab MonthlyC, K Principal Gamma 1 X10-4 (RM8132B) SampleC.,K EmittersB H-3U I X 10-6 2.Millstone Continuous Charcoal Sam- Weekly 1-131 1 X10-12 Stack pleD'F 1-133 1 x 10-10 (RM8169- 1) Continuous Particulate Weekly Principal Particu- I, x 10-1 SampleD.E late Gamma Emit-tersB - (I-131, oth-ers with half lives greater than 8 days)

Continuous Particulate Quarterly Composite Sr-89, Sr 1x10-SampleD. Gross alpha l x 10-11 Continuous Noble Gas - Continuous Monitor Noble Gases - 1 x 10-6.

I_ 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.-I, I.C.-2, or I.C.-3.

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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 [,Ci/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.

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(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.

1F 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 [LCi/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 [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 increaseis 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.

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

'A. Containment and Fuel Building Release 1.Containment Gaseous, particulate and Prior to each purge or Principal gamma 1 X 10-4 charcoal grab prior to drawdown. Same as emittersB-each drawdown (via air sample frequency for ejector) releases to maintain -

sub- atmospheric.

2.Fuet Building Gaseous grab prior to Prior To Each Draw- 1-131 1 x 10-12 each purgen. down 1-133. I x 10-10 Gaseous Grab every two Principal particulate I x-it weeks for releases to gamma emittersB- -

maintain (1-131, others with half lives sub.-atmospheric pres- greater than 8 days) sure (via containment vacuum purge)--

Monthly for all re- H-3 1 x 10-,

lease sources except Equipment Hatch Continuous particulate Weekly Particulate gamma emit- NA at open containment ters for 1/2 hour count equipment hatch (1-131, others 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 doorsK' Gaseous grab at Daily Noble Gases - Gross 1 X 10-4 containment equipment Activity hatch & fuel building rollup doorsK, B.Continuous Release L.Unit 3 Ventila- Monthly - Gaseous MonthlyC-,J . Priticipal particulate 1 x 10-ý tion Vent' Grab SampleC-,J. gamma emnittersB (HVR- ..

RElOB) H-3G I X 10-6 2.Engineered' Continuous charcoal Weekly 1-131 1 X 10-12 Safeguards samnpleD.,E i-133 1 x 10-10 Building Continuous particulate Weekly Principal particulate 1 x 10-11 (HVQ-RE49) sampleDl.,F gamma emittersB- -

3.Millstone (1-131, others with half lives Stack via greater than 8 days)

SLCRS Continuous particulate Quarterly composite Sr-89, Sr-90 1 x 10-It (HVR- sampleD. Gross alpha 1 x 10- 1 1 RE19B) Continuous noble gas- Continuous Noble gases - gross I x 10"6

_monitor activity MP-22-REC-BAP01 STOP'iL T'H NK ACTI VIE.. Rev. 026 - 00

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TABLE I.D.-3 TABLENOTATIONS .

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 ýCi/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. 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.

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.

"** ... ,* .*i)i ....... *i. MP-*22--REC- BAP01 4INK .SO

AT R.EV

, IE Rev. 026 -00 29 of 165

.F containment noble gas activity exceeds 1E-6 i[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 1E-6 iCi/cc.

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-BAPO1 STOP EHINK ACT REVIEW Rev. 026-00 30 of 165

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 mnem to any organ AND the dose from a waste stream with equipment not operating exceed 10% any of these limits, THEN prepare and sutbmit 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.

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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.

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' 32 of 165

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 1. 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 1, 10 CFR 50 for gaseous effluents following the guidance in NUREG-1301.

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Figure I.D.- 1, "Simplified Gaseous Effluent Flow Diagram Millstone Unit One" A..... I............... = generalized air flow Reactor Balance of Plant Vent Building S

- VcnI plenum Turbine uildng Radwaste Building MP-22-REC-BAP01 s~top 'T;H i4K:' ACT RE;-. Re 0216-00 34 of 165

Figure I.D.-2, "Simplified Gaseous Effluent Flow Diagram Millstone Unit Two"

-Rad Monilor (Rh4132A/B]

Turbine Bldg

) Turbine Bldg Roof'Vent 1These flow paths used during an accident.

MP-22-REC-BAP01

-P fINK' ACT RE VIEW Rev. 026-00 35 of 165

Figure I.D.-3, "Simplified GaseousEffluent Flowy Diagram Millstone Unit Three" t

Atmosphere ReacIor Plant Ventilation Vcnt

.I Ventilation Vent Rad Monitor

("RRl8 Service Building Atmosphcre HEP harcoa Fulel I* VR'FLT2A )

Auxiliar Millstone

.Building AuxiiaryStack B~uilding HEPACaca Gaseous Waste (5HVRJLlf3 System Waste D~isposal

• l iIilI!l Building Containment PrnML1.

SIuidt Contaiurnct, *. HEPA i SLCRS n

hCR Rad Monitor (SLCRS) 4F (3HVR taHVk&EIgu)

CPF 1urtHin Gland AtmoSphcre 5nA, CSc.,S' Cun Exh-u

• These flow paths used during an accident.

MP-22-REC-BAPO1 STQP THINK ACT REVB'I Rev. 026-00 36 of 165

I.E. Radiological Environmental Monitoring L 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 unobtainiable 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 IL.. 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.

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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 than those 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.I.b., 1V.D.2.b., or IV.D.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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C.'-' C ~ *1.

Table I.E.-I 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 - 407a) 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 sample if gross beta is greater than lIx 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 3 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 5.a.Pasture Grass 3 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 I

S.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 l.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 LE.-2 Environmental Monitoring Program SamplingLocations 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*"

I-i Onsite - Old Millstone Road 0.6 Mi, NNW TLD, Air Particulate, Iodine, Vegetation 2-I Onsite - Weather Shack 03 Mi, S TLD, Air Particulate, Iodine 3-1 Onsite - Bird Sanctuary 0.3 Mi, NE TLD, Air Particulate, Iodine, Soil 4-1 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 7-1 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-I Site Boundary 0.5 Mi, NE Vegetation 21-1 Goat Location No. 1 2.0 Mi., N Milk 22-I Goat Location No. 2 2.7 Mi, NE Milk 24-C Goat Location No. 3 29 Mi, NNW Milk 25-I 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 1 29-I West Jordan Cove 0.4 Mi, NNE Clams, Fish 1 30-1 Niantic Shoals 1.5 Mi, NNW Mussels 31-I Niantic Shoals 1.8 Mi, NW Bottom Sediment, Oysters 32-1 Vicinity of Discharge 2 Bottom Sediment, Oysters, Lobster; Fish 1, Seawater 33-1 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 36-1 Black Point 3.0 Mi, WSW Oysters 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

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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 45-1 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 7 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 51-1 East Lymne - 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 55-1 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 59-1 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 - :lordon Cove Rd. 0.8 Mi, NE TLD 64-1 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 72-1 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

'Fish to be sampled from one of three locations -28, 29, or 32.

2 Vicinity 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, "Inner Air Particulate And Vegetation Monitoring Stations"

. .,:.

• t5,,--.* * *.,.,- ,, . .. . . .,r .. . .. *:  ::M S

'O~tiOw W A. 54AV.X RI TLMonitonng(on,---)' ...

fl TL-D and Air Montm'lo'ring(aitolates & Iodine) -a'  ;.. #.?.* * .. *:* -:: ::*&.-*,*. .. .. .. .,.

[*$j 4 'errestriai Monitoring (milk grass fruit &, vegetables. soi or leaves) -S" nr.

ti .Aqua tic.Monitoring,(seawater, sediment, flora fish mussel; oyster, cana or.ilisier) '.-;' '.. A ...J.v . ':-"

'-M.n-eluts,..;-

.'-*::J*"** * e . ':t:**.* 9 *',:

,.-'--*..-e-.'..

--- '* '..,.::. . ',* .--.. :.:;' ':*'5

':-:

• MP-22-REC-BAP01

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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 )

14-3 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-ll0m 8 30 1-131 20B- 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 REC-BAP01

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Table L.E.-4 Maximum Values For Lower Limits Of Detection (LLD)A.

Analysis Water Airborne- Fish Milk Food Sediment (pCi/i) 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 2000D Mn-54 15 0.130 Fe-59 30 0.260 Co-58, 60 1.5 0.130 Zn- 65 30 0.260 Zr-95 30 Nb-95 1.5 2 1 0.06B-

[-131 15C- 7 x 10-2 Cs- 134 15 5 x 10- 0.130 15 0.060 0.150 Cs-.137 18 6 x 10.-2 0.150 18 0.080 0.180 Ba- 140 60 c. 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 4.66 Sb (E)(P)(2.22)(Ye-')9 Where:

• LLD is the lower limit of detection as defined above (as [LCi 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)

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. I, is the radioactive decay constant for the particular radionuclide SAt 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.

A > _ '-'ID 1Dn1I P 0- Q A

• l* JL.JX'*Y3LU J.

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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/O 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 theiresults 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 modelingof 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/I 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-BAPOI TOP TiINK ACT . VIE:'VRev. 026-00 48 of 165

Annual reports of environmental radiation monitoring summaries are filed with the NRC in accordancewith 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.F 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 dosesgreater than 10CFR50 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 ihclude a suimmary 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-routihe 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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!I 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 CER 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 (iFe., 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 yearfor 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 REC- BAPFO, "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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I; 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

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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 CFR 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 in the 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 V.D.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, Method 1 being the easiest and most conservative.

As long as releases remain low, one should be able to use Method 1 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 bompletely separate the three units.

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

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

It shall be the responsibility of the Vice President and Senior Nuclear Executive

- 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 - REC - PRG, "Radiological Effluent Control."

ILC. 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 II, 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 for cross- reference effluent control requirements and applicable sections in the REMODCM which are used to determine compliance). In addition, installed portionsof 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 rnrem 'annual total dose limit (40 CFR190) to any real individual member of the public from all effluent sources (liquids, gases, and direct).

a. Method I (Applicable to Units1, 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-GDL02, "Liquid Dose Calculations

- LADTAP-II." Additional information on LADTAPII is contained in MP-22-REC-REF03, "REMODCM Technical Information Manual."

2. Maximum Organ Dose from Liquid Effluents Radiological Effluent Controls (Sections III, 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, Thble 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

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liquid pathways which contribute to the 25 mnrem 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 mrero) 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 ReleaseReport. The use of this code and the input parameters are given in MP-22-REC-GDL02, "Liquid Dose Calculations - LADTAP-IH." Additiona l information on LADTAPII is contained in MP REC-REF03, "REMODCM/1 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 closes include the whole body dose contribution from all effluent releases for each Unit recorded to-date. For estimation of the

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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 II.C.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 =E Do.

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 close 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 estimat~es 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.

b. Whole Body and Maximum Organ when Steam Generator Total Gamma Activity is less than 5E-7 lLCi/ml and Steam Generator Tritium is less than' 0.02 ptCi/ml (Applicable to Units 2 and 3)

The projected monthly whole body dose (Units 2 or 3) is determined from:

Dow D'Aih [R R14 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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RI = 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 thepresent month as opposed to the past month (e.g., bypass of filters or deminreralizers).

NUREG-0017 or past experience shall be used to determine the effect of each form of treatment which will vary. F 2 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 there 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 Maxinium Organ when Steam Generator Total Gamma Activity Exceeds 5E-7 liCi/ml or Steam Generator Tritium Exceeds 0.02

ýtCi/ml (Applicable to Units 2 and 3)

The projected monthly whole body dose (Units 2 or 3) is determined from:

• DEMTV = D D'p,,py [(I - F1 ) R 1 4 F2 + F, R 2 R3 ]

The monthly projected maximum organ dose (Units 2 or 3) is determined from:

DEMo = D'MO [(1 - F1 ) R, R 4 F 2 + F, R 2 R 3]

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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 calcLilated per the methods in Section II.C.2.'

RI = 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:

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 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. F 2 = 1 if there are no expected differences.

6. Quarterly Dose Calculatiohfs 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 LI code, and the input parameters are given in MP-22-REC-GDLO2, "Liquid Dose Calculations - LADTAP TI." 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 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 imnplement 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 implementedvia 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 I 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:

Ov /90,000 +Q2s/560,000 + Q2Vi 2 9 0 ',0 0 0 + 03S/560,000 + 0 3V/ 2 9 0 ,0 0 0 < 1 Where:

Qv = Noble gas release rate from Spent Fuel Pool Island Vent (ICi/sec)

Q2S = Noble gas release rate from MP2 to Millstone Stack (tCi/sec) 0 2V = Noble gas release rate from MP2 Vent (itCi/sec) 03V = Noble gas release rate from MP3 Vent (pCi/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 I SFPI vent of 90,000 is based on

• theskin dose limit of 3,000 tnrem/year, while all the other factors are based on the whole body dose limit of 500 mremiyear.

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 than1500 mrem/year from the inhalation pathway:

1) The site release rate.limit of,I-131, 1-133, and tritium (where the thyroid is the critical organ for these radionuclides) shall be:

DRthyl + DRthyi2 + DRthy3 < 1 Where the contribution from each Unit is calculated from:

Unit 1: DRthyl 9.36x 10-6. QH1V Unit 2: DRthy2 = 5.1 x 10-2131012/ + 2.38 x 1073 131OI2s +

1.25 x 10-2 133Q12V+5.75 x 10- 4. 133 Q1 2s +

4.2 x 10-6 QH2V+l. 9 x 10- 7 QH2S Unit 3: DRthy3 = 5.1 x 10-2 131 0 13V + 2.38 x 10-3 131Q13s +

1.25 x 10-2 133Q13V +5.75 x 10-4133013s +

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 mlx of radionuclides) shall be:

DRorgI + DRorg2 + DRorg3 <1 Where the contribution, from each Unit is calculated from:

Unit l:Dlorgi = 1.05 x 10-1 [QP1v + 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 3

Unit 3:DRorg3 = 2.38 x'10- Qp's3 + 5.1 x 10-2 QP3V +

1.9 x 10-7o QH3S + 4.2 x 10-6 QH3V

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Each of the release rate quantities in the above equations are defined as:

131Q12V = Release rate of 1- 131 from. MP2 Vent (gCi/sec)*

131 0 12S ='Release rate of 1-131 from MP2 to Millstone Stack (gCi/sec) 133Q12V =Release rate of 1-133 from MP2 Vent (gCi/sec)*

133Q12S = Release rate of I- 133 from MP2 to Millstone Stack

• (p.9i/sec)*

131 0 13V Release rate of 1-131 from MP3 Vents (Normal and ESF)

(.tCi/sec)*

131QI3S= Release rate of 1-131 from MP3 to Millstone Stack (gCi/sec) 133Q13v =Release rate of 1-133 from MP3.Vents (Normal and ESF)

• (p.Ci/sec)*

33QI3s = Release rate of.I-133 from MP3to Millstone Stack (p.Ci/sec)

QHIV = Release rate of tritium from the Spent Fuel Pool Island and Balance of Plant Vents (piCi/sec)

QH2V = Release rate of tritium from MP2 Vent (p.Ci/sec)*

QH2S = Release rate of tritium from MP2 to Millstone Stack (p.Ci/sec)

QIH3V = Release rate of tritium from MP3 Vents (Normal and ESF)

(p.Ci/sec)*

QHq3s = Release rate of tritium from MP3 to Millstone Stack (Uci/sec)

QpiV Release rate of total particulates with half-lives greater than 8 days from the Spent Fuel Pool Island Vent (gLCi/see)

QP1B = Release rate of total particulates with half-lives greater than 8 days from the Balance of Plant Vent (p.Ci/sec)

QP2v= Release rate of totalparticulates with half-lives greater than 8 days from the MP2 Vent (9tCi/sec)

QP2S Release rate of total particulates with half-lives greater than 8 days from MP2 to Millstone Stack (p9Ci/sec)

QP3V = Release rate of total particulates withhalf-lives greater than 8 days from MP3 Vents (Normal and ESF) (p.Ci/sec)

QP3S = Release rate of total: particulates with half-lives greater than 8 days from MP3 to Millstone Stack (ptCi/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-GDL04, "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 1 Air Dose* (Applicable to Units 1, 2, and 3)

For Unit 1: D0 1 = 3.3 x 10-6 CNIV*

DB1 = 1.49 x 10-3 CN1V*

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 DG 3 are greater than 1.6 mrad or DB1, DB2, or DB3 are greater than 3.3 mradwithin a calendar quarter, go to Method 2 below.

Where:

DG1 = the gamma air dose from Unit 1 for the period of interest (mrad).

DBT = the beta air dose from Unit ifor the period of interest (mrad).

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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).

CN4v= the. total curies of noble gas released from Spent Fuel Pool 4 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 1 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 thedUnit 2 vent eight-year history for 1980 to 1987:

3 JQ = 8.1 x 10-6 sec/m D/Q 1.5 x 10-7 m-2 (See the MP-22-REC-REF03, (Aft) "Determination of Maximum x/O 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 DyG1 = IDGI DyG2 =ZDG2 DyG3 = XDG 3 DYBI = YDBI DYB2 =IDB2 DYB3 = YDB3 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: DyGI, DyG2, DYG3, DyBj, 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 DYBI, 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 11, IV, and V) limit the off-site dose to a critical organ from radiolodines, 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 = 303jCIs +/- 0.29133CIs + 4.66 x 10-5 CHS DOS 10.2Cps + 4.66 x 10 5 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 2a below:

For Unit 1: DTS = 1.97 x 10- 3 CHV DOS = 94.8[Cpv+ CPBI+ 1.97x i0-3 CHV If either.dose is greater than 2;5 mrem within a calendar quarter go to Method 2a 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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1 31 CIS =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 curie-s 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.

CiV = the total curies of tritium released in gaseous effluents from the SFPJ 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 TI.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 131Clv + 4.0 133CIv + 2.6 x 10-3 CHV DOV = 1.1 x 103 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-BAP01

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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 Section II.D.3.b. below 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 thyroiddose 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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77! e 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 U~se 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 ev6ry month to determine compliaince with the annual dose limits for each Unit. Annual critical organ doses will be determined as follows:

Unit 1 Unit 2 Unit 3 DYTI = ZDTI DYG 2 = -DT2 DyT3 = XDT3 Dy 0 1 = XDo1 Dy0 2 = XD 0 2 Dy0 3 = YDo3 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, Dyo1 , Dy02 and DYO 3 = thyroid (T) dose and maximum organ (O) 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 0102 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 Tebchnical 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:

i Method 1 Determine DEM0 which is the estimated monthly dose to the maximum organ from the following:

DEMO = 1/3 R 1 (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.

SI. MP-22-REC-BAPOI T K -ACT REIE~**WRev. 026-00 74 of 165

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-GDL04, "Gaseous Dose Calculations - GASPARIL." 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), IV.E. (Unit 2), and V.E. (Unit 3) contain, specific requirements for ensuring compliance with 40CFRI90 based on gaseous and liquid doses (sources a and b).

Doses to source cý are controlled by design and operations to ensure the off-site dosefrom 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 IOCFR20 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-BAPO1 STOP N W Rev. 026 - 00 76 of 165

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 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 boundarywith 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.

= *

• -: MP REC- BAPO1 STOP TA-RN K ACT REVsiEWRev. 026-00 77 of 165,

ILE. Liquid Discharge Flow Rates And Monitor Setpoints L 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:

Fr'onm the isotopic analysis, and the Effluent Concentration (EC) values for each identified nuclide determine the required reduction factor, i.e.:

R = Required Reduction Factor = w o nudhde i~

of W-ýC 1-0 x--ECof idei 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 (p.Ci/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 10i7 gCi/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:

• :*. i :MP-22-REC-BAP01

.STOP " I' INK ACT REVI Rev. 026-00 79 of 165

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

• fcd Eofnuclide i 10o -E- -of"--dg,*Je i For Noble Gases: If the noble gas concentration is less than 0.011 jtCi/ml, the reduction factor need not be determined 1 2 x 10 4l R2 = Required Reduction Factor =An o@/f noble gases noble gases 2x 10 -**c@i R= the smaller of R1 or R 2 STEP 2:

Determine the allowable discharge flow (F) in gpm:

F=0.1xRxD Where:

D = .the existing dilution flow (D)

Note: D # circulating water, pumps x 100,000 gprn + # service water pumps x 4,000 gpm)

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 I below.)

.i* .*~ii:MP.-22-REC-B3AP01 ST1-p THINK 'ACT REVJIEW Rev. 026-00 80 of 165

Where:

Rset= the setpoint of the monitor (cpm).

AC = the total radwaste effluent concentration ([iCi/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 !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.

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 worst case conditions, assuming maximum discharge flow (350 gpm), dilution water flow of 100,000 gpm and a limit of 1 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 i0-7 RtCi/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 1 rCi/ml. If the grab sample is less than 5 x 10- 7 kCi/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 1.7 x 10-4 [tCi/ml, which is approximately 38,000 CPM based on recent calibration data.

MP-22-REC-BAP01 S'TOP{* T.IK ACT. VIWRev. 026-00 81lof 165

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. Normal minimum possible circulating water dilution flow during periods of blowdown = 200,000 gpm (2 circulating water pumps) 200,000 gpm..

c. The release rate limit is conservatively set at 3 x 10-8 [Ci/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:

Alarm (uCilml= 200,000 x 3xl0O8 + background*,*= 8.5x10- 6

'uCi/lrn + background The latest monitor calibration curve shall be used to determine the alarm setpoint in cpm corresponding to 8.5 x 10-6 ,Ci/ml.

This setpoint may be adjusted (increased or decreased) through proper administrative controls if the steam generator blowdown rate is maintained other than 700 gpin and/or other than 2 circulating water pumps are available. The adjustment would correspond to the ratio of flows to those assumed above or:

circulating & service waler flow. (gpm) 700 +

Alorm (u Ci/l)= 8.5xO-'IzCi/ml x 200,000 -x SGb,..d.,,. (gpm) circudating & se'rvice water flow (gpn) + Background Background = 3xRI0-Ci/ml x tackbround total SIG blowdown (gpmi)

MP-22-REC-BAPO1 SOTHHI- " AOT -**':KREVIEW Rev. 026-00 82 of 165

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-. Ci/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 Blowdowvn Effluent Concentration Limitation The results of analysis of blowdown samples required by Tlable 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 IOCFR20. 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 "S .... I AT ... REV IW Rev. 026 -00 83 of 165

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 cprn. 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.

.STOP THINK ACT .. ** MP-22-REC-BAP01 RVIEWRev. 026-00 84 of 165

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= I Required Reduction Factor S of1inudhde i of 10 x EC of nuclide For Noble Gases: If the noble gas concentration is less than 0.026 kLCi/ml, the reduction factor need not be determined 2 x 10-4,""

Ml R2 Required Reduction Factor = Ci noble gases 2*-2of 2x 10- 10§'noble .gases R the smaller of R1 or R2 Step 2-Determine the allowable, discharge flow (F)

F= 0.1xRxD Where:

D = The existing dilution flow. ()):

Note: D = # circulating water pumps x 100,000 gpm+ # service water pumps x 15,000 gpm MP-22-REC-BAP01

~1H]NKv a.5) . RIRev. 026-00 85 of 165

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 conditior on discharge flow rate met, the monitor setpoint can be calculated:

Rset 2 x AC xRCF (see Note 1)

Where:

Rset= The setpoint of the monitor.

AC= The total radwaste effluent concentration (ýtCi/mn) 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.

MP-22-REC-BAPOI T.op

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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 worst case conditions, assuming maximum discharge flow (150 gpm), dilution water flow of 100,000 gpmn, and a limit of 1 X 10-7 pCi/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-REO7 Same asSection II.E.8.

11. Unit 3 Steam Generator Blowdown - SSR-RE08 and Unit 3 Steam GeneratorBlowdown 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 [tCi/ml which is well below any IOCFR20 Effluent Concentration except for transuranics*.

c. Minimum possible circulating and service water dilution flow during periods of blowdown = 200,000 gpm (2 circulating water pumps) +

30,000 gpm (2 service water pumps) = 230,000 gpm.

.. MP-22-REC-BAPO1 STOF' -ThINK ACT1 RE--VIEW Rev. 026-00 87 of 165

d. Background can be added after above calculations are performed.

Therefore, the alarm setpoint corresponds to a concentration of:

Alr¢t im) 230, 000 -5.

Alarm (110/m)- 400 x 3x10- 8 + background = 1.7xl- 5 Cil/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 2 circulating and 2 service~water pumps are available.

The amount of the increase would correspond to the ratio of flows to those. assumed above or:

Alarm (zCi/ml[) = 1.7xlO-HCi/ml x circulating & service water flow (gpm) 230,000 400 SIG blowdown (gpm) circulating & service water flow (gprM)

Back&,rowi = 3x1 0-8uCi/ml x + Background total SIG blowdown. (gpm)

NOTE

" lthe sram Generator Blowdown alarm criteria is in practicebased 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 [Ci/ml, ten times the identified 10CFR20 EC values may be used.

11b. 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. -o

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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 1.4x 10-5 itCi/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 hpplicable, the release of radioactive materials in liquid effluents during actual or potential releases of liquid waste to the environment.. The alarm I 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.

MP-22-REC-BAPO1 JSO I H; K THN ACGT RIEW* R . 026-00 89 of 165

I.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.1.a. in order to satisfy Radiological Effluent Controls III.C.2. and III.D.2.a.

The alarm setpoint shall be set at or below the monitor reading in [tCi/cc corresponding to 29,000 gCi/sec assuming, a maximum ventilation flow of 36,000 cfm. The corresponding monitor reading is 1.71E-3 VCi/cc. NOTE:

This setpoint is the basis for emergency classification in Unit 1 EAL Table (OA- 1). A change to this setpoint would require a concurrent change to the EAL The release rate value of 29,000 vCi/sec assumes that 7% of the site limit for skin dose of 3000 mrem per year is assigned to the Unit I Spent Fuel Pool Island vent. If effluent conditions from the Unit 1 Spent Fuel Pool Island vent reach 29,000 ltCi/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.

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.1l.a. in order to satisfy Units 2 Radiological Effluent Controls IV.C.2. and IVD.2.a.

The alarm setpoint shall be set at or below the monitor reading in llCi/cc corresponding to 74,000 VCi/sec assuming a maximum ventilation flow of

.12,000 cfm. The corresponding monitor reading is 1.31E-2 [tCi/cc.

The release rate value of 74,000v[tCi/sec assumes that 13% of the site limit is assigned to Unit 2 releases to the Millstone Stack. If effluenf conditions from Unit 2 to the Millstone Stack reach 74,000 ptCi/sec, releases from Units 1, 2 and 3 vents and from Unit 3 to the Millstone Stack shall be determined to ensure that the sum of the individual noble gas release rates do not exceed the site limit as dictaited in Section II.D.l.a., and described in MP REC- REF03, "REMODCM Technical Informatio.n Document,"

Section 4.2:

3. Reserved MP-22-REC-BAPo1 STOP THINK ACTI HVEW Rev. 026- 00 90 of 1.65

4. Unit 3 SLCRS - HVR-RE19B The instantaneous release rate limit from the site shallbe set in accordance with the conditions given in Section ILD.l.a. in order to satisfy Unit 3 Radiological Effluent Controls VC.2. and V.D.2.a.

The alarm setpoint shall be set at or below the monitor reading in iiCicc corresponding to 74,000 [tCi/sec assuming a maximum ventilation flow of 12,000 cfrn. The corresponding monitoryreading is 1.1.6 E-2 [tCi/cc in accordance with Calculation RERM-019416R3, Rev.1.

If effluent conditions from Unit 3 to the Millstone Stack'reach.74,000 ptCi/sec, measures shall be taken to ensure that the sum'of the individual noble gas release-rates do not exceed the site limit as dictated in Section II.D.1.a., and described in MP-22--REC- REF03, "REMODCM Technical Information Document," Section 4.2.

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.1.a. in order to satisfy Radiological Effluent Controls in Sections IV.C.2. and IV.D.2.a.

The alarm setpoint shall be set at or below the "cpm" corresponding to 95,000 ptCi/sec from the MP2 vent noble gas monitor calibration curve. The calibration curve (given as gtCi/sec per cpm) is determined by assuming the maximum possible ventilation flow for various fan combinations. Curves for three different fan combinations are normally given. In calculation RERM-02906R2 a maximum allowable setpoint of 42,000 CPM is determined using vent flow of 100,000 ft3/min.

The release rate value of 95,000 ltCi/sec assumes that 33% of the site limit is assigned to the MP2 vent. If effluent conditions from the MP2 vent reach the alarm setpoint, releases from Units 1 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 exceed the site limit as dictated in Section ILD.1.a., and described in MP-22-REC-REF03, "REMODCM Technical Information Document," Section 4.2.:

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 74,000 ItCi/sec (see the MP-22-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 7,400 RCi/sec.

The MP2 waste gas decay tank monitor (given lCi/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 7,400 ptCi/sec to be exceeded.

7. Unit 3 Vent Noble GasMonitor - HVR-REIOB The instantaneous release rate limit from the site shall be set in accordance with the conditions'given in Section ILD.l.a. in order to satisfy Radiological Effluent Controls in Sections VC.2. and VD.2.a.

The alarm setpoint shall'be set at or below a value of 8.4 x 10-4 1tCi/cc for the MP3 vent in accordance with Calculation RERM-01946R3, Rev.1.

If effluent conditions from the MP3 vent reach the alarm setpoint, releases from Units 1 and 2 vents and from th'e Millstone Stack shall be determined to ensure that the sumn of the individual noble gas release rates do not exceed the site limit~as dictated in Section II.D.i.a., and described in MP REC- REF03, "REMODCM Technical Information Document,"

Section 4.2.

8. Unit 3 Engineering Safeguards Building Monitor - HVQ-RE49 The Alarm setpoint shall be set at or below the value of 5.9E-4 ýtCi/cc in accordance with Calculation RERM-01946-R3, Rev. 1.

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 (as discussed above for each monitor 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 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 MP-22-REC-BAPO1 STOP THINK

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• "*; ,,.z* ,,*,'*?.*1* ,* ,,.: * *d*,, * , * ,* ,*.=. *,;, *r-*rm* *,* *,* 2,*.* ;, ,,....... , ......

!* *:;4*."?'*"*?:'!:i* :*)*).*-*?:4.:*S: :.:'*" ;"'* * *5*"::**"*;"* *:-:.**.* b*;:'.*':.::*-'**':.* *!*:?.*';*:'::2 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 I.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.

MP-22-REC-BAPO1 S-TOP T.HINK R Rev. 026-00 93 of 1.65

Table II.A.-I 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 'rabies I.C.-2 Ten times 10CFR20App.B, Instantaneous Restore concentration towithin lint-4 Liquid Effluent and L.C.-3 Table 2, Column 2, & 2x10- its within 15 mins.

Concentration iCi/mL for dissolved noble gases*

IV/V.E.l.b II.C.1. < 1.5 mrem TB. Calendar Quar- 30-day report if exceeded. Relative Dose-Liquids II.C.2. -5 mrem Organ ter*" accuracy or conservatism of the cal-lI.C.4.

II.C.3. <*3 310mremTB.

mrem Organ Calendar Year ulations shall be confirmed by per-formance of the REMP in Section I.

T.S. 6.16 (Unit 2) 1j.C.2: " *-0.06 mrem TB. Projected for 31 Return to operation Liquid Waste TS. 6.14 (Unit 3) 11.C.5. *0.2 mrem Organ days(if system Treatment System Liquid Radwasle Treat- not in use) ment l1I.D.2.a Trables I.D.-1, *500 mremlyr 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.

• 3000 inrem/yr skin from noble gases*

II.D.l.b. -- 1500 mrem/yr organ from particulates with Tt12 > 8d.,

1-131, 1-133 & tritium*

tll.D.2.b I.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 -s 10 mrad gamma air Calendar Year

• <20 mrad beta-air If .D.2.c II.D.3. < 7.5 mrem organ Calendar Quar- 30-day report if exceeded. Relative JV/V.D.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 I.

TS. 5.6.4 (Unit 1) ILD.2. > 0.02 mrad gamma air Projected for 31 Return to operation Gaseous Rad-T.S. 6.14 (Unit 2) l1.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 II.E II.D.6. <25 snrem TB.* 12 Consecutive 30-day report if Unit 1 Effluent IV] .VF *<25 mrem organ* Months,* Control III.D.1.2, III.D.2.2, or Total Dose _75 mrem thyroid* III.D.2.3 or Units 213 Effluent Con-trol IViV.E.1.2, IV/VE.2.2, or IV/

VE.2.3 are 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.

...  :.. .. ':;MP-22-REC-BAPO1

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SECTION III.

Millstone Unit 1 Radiological Effluent Controls Docket Nos. 50-245

. MP-22REC-BAPO1 STP THIN ulIE ACT1 Re 2-00 95 of 165

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.

L. ACTION - that part of a Control that prescribes remedial measures required under designated conditionls.

2. INSTRUMENT CALIBRATION - the adjustment, as necessary, of the instrument output such that it respbnds 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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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 in the 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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I!

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 performanceor as measured from the time a specified condition of the frequency is met.

3. If it is discovered that a Surveillance frequency, then compliance with the was not performed requirement within to declare theitsLCO specified 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 isnot peiformed within the delay period, the'LCO must immirediately 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 delayperiod. 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 Tinmes of the Required Actionsbegin immediately uponfailure to meet the Surveillance.

.. MP-22-REC-BAP01

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III.C. Radioactive Effluent Monitoring Instrumentation

1. Radioactive Liquid Effluent Monitori ng Instrumentation CONtROL'S The radioactive liquid effluent monitoring instrumentation channels shown in Table II.C.- I shall be OPERABLE with applicable alarm/trip setpoints set to ensure that the limits Of Specification JII.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. -I

.ACTION:

a. With a radioactiveliquid 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.

,MP-22-REC-BAP01 STOP THNK ACT: ,. Rev.026-00 99 of 165

TABLE 11I.C.-1 Radioactive Liquid Effluent Monitoring Instrumentation Instrument Minimum # 'Alarm Setpoints Applicability Action Operable Required

'LRadioactivity Monitor i Yes

• A 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 III.D.1:a. and; (2) The original release rate calculations and discharge valving are independently verified by a second individual.

.. ;:;:?,

}> MP- 22--REC- BAP01 ISTOP TINI<:N ACT REVIEW 1"ev. 026-00 100 of 165

TABLE II.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 0 = Once every 3 months

• During releases via this pathway, and 'when the monitor is required OPERABLE per Table III.C.-1. The CHANNEL CHECK should be done whenthe discharge is in progress.

(1) Calibration shall include the use of a radioactive liquid or solid source which'is traceable to an NIST source.

. MP-22-REC-BAPO1 STOP -HINK ~AC~1 - E Rev 026-00 101 of 165

2. Radioactive Gaseous Effluent Monitoring Instrumentation CONTROLS The radioactive gaseous effluent monitoring instrumentation channels shown inTable III.C0-3 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 IIF.1.

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.

SUREFILLANCE RENUIREMENT Each radioactive 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 ItI.C.-4.

MP-22-REC-BAPO1 STOP THINK AC .REV.IE Rev. 026-00 102 of 165

TABLE 1II.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 2.3alance 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.hours.

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.

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--BAPOI STOP HlNK A**W*.' Rev. 026-00 103 of 165

Chemistry compensatory sampling time period as specified in Action A or Action B. Sample flow rate need not be estimated if the auxiliary sampling equipment of Action B is in use. I -

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TABLE III.C.-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements I 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 0 Once every 3 months T = Once every two years NA Not Applicable Table II1.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 st rength 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.

ACT REV MP-22-REC-BAPO1 STIOP ITANrK ,'T '* IEWRev. 026- 00 105 of 165

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 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 REOUIREMENT Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified, in Section 1.,

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-BAP01 STiOP IHINK ACT REVIEW Rev. 026-00 106 of 165

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 Figu-e III.D..- I) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the totalbody and toless than or equal to 5 mrem to any organ; and,

b. During any' calendar year to less than orequal 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 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.

M 22-*REC- BAN I STOP TIK AT RVE Rev. 026 -00 107 of 165

2. Radioactive Gaseous Effluents

a. Radioactive Gaseous Effluents Dose Rate CONTROLS The dose rate, at anyrtime, 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 mrcm/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.D12.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 df Control III.D.2.a. are met.

MP-22-REC-BAPO1 S;'TOP HV1K AT*! ... WRev 026-00 108 of 165

b. Radioactive Gaseous Effluents Noble Gas Dose CONTROLS The air dose offsite (see FigureII.D.-1) due tonoble 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) ahd 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 withSection 11 once6 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.

MP-22-REC-BAPO1 STQP THINK ACIT R'EIE' Rev. 026-00 109 of 165

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 th.

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 RE tUIREMENTS

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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4 4

/4 Figure III.D.-1, "Site Boundary for Liquid and Gaseous Effluents" MP-22-REC-BAP01 si:oP TINKi!ii Act RE!ViE-Rev. 026 -00 11of165

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 boundaiy, from the Millstone Site is limited to less than or equal to 25 m-em 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 torem) over 12 consecutive months.

This Special Report shallinclude an analysis which demonstrates that radiation exposures from the site to any REAL MEMBER OF THE PUBLIC.from the MillstoneSite (including all efflueht pathways and direct radiation) are less than the 40 CFR 1-90 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.

SUJRVEILLANCE REQUIREMEI-iTS' Cumulative dose contributions fromliquidand gaseous effluents and direct radiation from thie Millstone Site shall be determined in accordance-with Section If once per 31 days..

III.F. BasesSection III.C.I - Radioactive Liquid.Efluent Monitoring Instrumentation No controls required; Unit 1 is not currently releasing radioactivity in liquid effluents .

Section. II.C.2 - Radioactive Gaseous Effluent Monitoring Instrumentation MP-22-REC-BAP01 STOP .4..N ACT REYI LVV Rev. 026-00 112 of 165

The Spent Fuel Pool Island Vent is the only gaseous pathway currently requiring radiation monitoring for Unit 1.

Section III.D.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 basedupon the assumption that Xe- 135 is the controlling radioisotope and its concentration in air (submersion) was converted to an equivalent concentration in waterusing the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

Section IILD.Lb. - Radioactive Liquid Effluents Doses This specification is provided to implement the requirements of Sections II.A.,

111A, and IVA of Appendix 1, 10 CFR 50. The specification implements the guides set forth in Section JI.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 conformancewith the guides df 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-BAPOI STOP ...--- TINK A. R**E V Rev. 026-00

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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 oi~beyond the site boundary to less than or equal to 500 mrem/yeai 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 turen/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 operatinrg 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 theguides 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 i-adioactive noble gaises 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 I1.C, JII.A and IVA of Appendix I, 10 CFR 50. The controls are the guides set forth in Section ILC of Appendix 1. The action statements provide the required

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- "- . ' ° S * , C 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 proceduresbased on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calclulational 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 Lightt-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 iin 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 are'as where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) dep6sition on the ground with 'subsequent exposure of man.

Section II.E. - Total Radiological Dose from Station Operations This control is provided tomect-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 nucle'ar 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 S .MP-22-REC-BAP01

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SECTION IV. REMODCM UNIT TWO CONTROLS IVA. 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 prdvided as an aid for iniplementation of the controls and requirements.

Some surveillance requirements refer to specific sub-sections in Sections I and II as part of their required actions.

IV.B. Definitions, Applicability and Surveillance Requirements V.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 intp.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 r6sponse 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 equipmentor to make deliveries. This category does include persons who use portions of the site for recreational, occupatio.nM, 0or'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.

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11- DOSE EQUIVALENT 1-131 - DOSE EQUIVALENT 1-131 shall be that concentration of 1- 131 (RCi/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 thoselisted 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."

IVB.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 IV.B.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), traih(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 IV.3.2.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 Requii:ement 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'hours. 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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IVC. Radioactive EfflMUent Monitoring Instrumrentation

1. Radioactive Liquid Effluent Monitoring Instrumentation LIMITING CONDITIONS OF OPERATIONS The radioactive liquid effluent monitoring instriimentation 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.I.a. are not exceeded. The setpoints shall be determined in accordance with methods and.parameters described in Section' IL-APPLICABILITY: As shown in Table IV.C..- 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 IVC. -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 IVC.-1 Radioactive Liquid Effluent Monitoring Instrumentation Intuet#

Instrument Operable AlarmSetpoints Minimum Required Applicability* [Action 1.Gross Radioactivity Monitors Providing Automatic Termination Of Release (a) Clean Liquid RadwasteEffluent 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 1YesE Waste Neut Sumrp 2.Gross Radioactivity Monitors Not Providing Automatic Termination Of Release (a)Reactor Building Closed Cooling 1 Yes C Water.Monitor# J 3.Flow Rate Measurements (a) Clean Liquid Radwaste Effluent 1 No . D Line (b)Aerated Liquid Radwaste 1 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 pCi/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 [,Ci/gm DOSE EQUIVALENT.I.-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 inTable 1.07-2; 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 performancecurves 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) Ifone 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 1Channel Source Channel Channel Check Check

_"_ _ . J Calibration

_,Test 1.Gross Radioactivity Monitors Providing Alarm and Automatic Termination Of Release Functional

a. Clean Liquid Radwaste Effluent Line D* P R(I) 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 Water Monitor .j D 1 M " R(!)

R(1) 0(2)

Q(2) 3.Flow Rate Measurements

a. Clean Liquid Radwaste Effluent Line D* N/A R Q

b. Aerated Liquid Radwaste Effluent D N/A Q

'Line

c. Con~densate Polishing Facility Waste :D* .N/A R Q Neut Sump 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 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 IVC.- 3 shall be OPERABLE with applicable alarm/trip setpoints set to ensure that the limits of Specifications IVD.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 IV.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 channels less than the minimum channels OPERABLE requirement, take the action shown in Table IVC,.-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 IV.C.-4.

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TABLE IV.C.-3 Radioactive Gaseous Effluent Instrumentation Instrument Minimum Alarm Applicability Action

_ __ *Operable JRequired_

Channels Setpoints I.MP2 Vent (normal range, RM-8132 only; high range monitor, RM-8168,,requirements are in the TS)

a. Noble Gas Activity Monitor I y1e s##1 ** 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... .. I. . No , C 3.Waste Gas Holdup System a- Noble Gas Monitor Providing 1.1 Yes *  : D, Automatic Termination of Release . I

• 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 'maycontinue 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 repairthe 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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TABLE IV.C.- 4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Instrument Channel Source Channel *Channel Check Check Calibration Functional 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 , Q

a. Noble Gas Monitor DP R( Q(2 )

i*uring releases via this pathway and When the monitor is required OPERABLE per Table IlV.C.-3.

The CHANNEL CHECK should be performed when the discharge is in progress.

P = Prior.to discharge R = Once every 18 months D = Dailyý. 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 IIMITTTN CONDITIONS OF OPERATIONS The concentration of radioactive material released from the site (see Figure IVD. 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 entrained noble gases. For dissolved or-en trained noble gases, the concentration shall niot exceed 2 x 10-4 ,Ci/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 i5 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 IVD.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 T1HE 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 tuaiter 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

• bbdy 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) anid defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents- during the. rerffaindet 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 mreim to the total body and 10 mrem to any organ.

SURVEILLANCE REQUIREMENTS

,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 MMonitoring Program as det ailed in Section I.

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2. Radioactive Gaseous Effluents

a. Radioactive Gaseous Effluents Dose Rate LIMITING CONDITIONS OF OPERArIONS The dose rate, at any time, offsite (see Figure IV.D.-i) 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 mremryr 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) giyen' in Specification IV.D.2.a. within 15 Minutes.

SURVEILLANCE REQ UIREMENTS

1) The release rate, at any time, of noble gases in gaseous effluents shall becontrolled by the offsite dose rate asestablished above in Specification IV.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 Table IV.C.-3 shall be used to control release rates to limit offsite doses within the values established in Specification IVD.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 'tonoble 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 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 - Doses from Radionuclides Other than Noble Gas LIMITING CONDITIONS OF OPERATIONS The dose to any REAL MEMBER OF THE PUBLIC from Iodine- 131, Iodine'- 133, Tritium, and radioactive-rmaterials in particulate form with half lives greater than 8 days in gaseous effluents released offsite from Unit 2 (see Figure IVD.-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 dalendar year to less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

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 arid submit to the Commissidn 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 IV.D.-1, "Site Boundary for Liquid and Gaseous Effluents" MP-22--REC-BAPOI

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IVE. 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 IV.D.2.a., IVD.Ib., or IV.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.

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 Millstone Site shall be determined in accordance with Section II once per 31 days.

IVE Bases Section IVC.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 tines the limits of 10 CFR 20. The OPERABILITY and use of this MP-22-REC-BAPOI STOP " H K.

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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 dtains is not required due to relatively low concentrations of radioactivity possible.

Section IV.C.2. - Radioactive Gaseous Effluent MoniLoring 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 nmethods 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 df this instrumentation is consistent withthe'requirements of General Design Criteria 60, 63, arid 64 of Appendix A to 10 CFR 50..

fwvo types of riadioactive gaseous effluent monitoring instrumentatioi, monitors and samplers, are being used at MP2 vent and Millstone Stack. Monitors have alarm/trip setpoinits anid. ire demonstrated operable by performing one or more of the following operations: CHANNEL CHECK, SOURCE CHECK,

,.CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST Samplers ale strictly collection devices made of canisters and filters. The CHANNEL CHECK surveillance requirements are&.net through (1) docutented observation of the in-service rad monitor siimple flow prior to filter replacement; (2) documented replacement of in-liie' 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 isý 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 ca:es practical due to the nature 6f 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 IV.D.1.a. - Radioactive Liquid Effluents. Concentrations This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents f61n 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: (i) the Section II.A design objectives of Appendix 1, 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 K. T..HINK ACT RE-.VIE*(W:w Rev. 026 - 00 136 of 165

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 IVD.1.b. - Radioactive Liquid Effluents Doses This specification is provided to implement the requirements of Sections ILA, IILA, and IVA of Appendix I, 10 CFR 50. The specification implements the guides set forth in Section II.A of AppendixI. The Action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IILA 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 fi.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 Regulhtoiy 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 CER 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 besufficiently 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 releaserate limits also restrict, at all times, the corresponding thyroid of any other organ dose rate above background to a child via the inhalation pathkvay 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 TI THNSTO K '-ACT E I*"W Rev. 026--00 138 of 165

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.

VB. Definitions and Applicability and Surveillance Requirements VB.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. CI-HANNEL 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, sh all 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 sameparameter.

5. DOSE EQUIVALENT 1-131 -. DOSE. EQUIVALENT 1-131 shall be that concentration of I- 1-31. ([Ci/gram) which alone would *Producethe same CDE- thyroid dose as the quantity and isotopic mixture of I- 131, 1-132,.

1-133, !-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 peirsons 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 categoly 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 itHE 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'otheiwise 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 exposedto:

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 - Applicb- "

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 hot 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 VB.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 Opera'tion have been performed within the stated surveillance interval or as otherwise specified.

VC. Radioactive Effluent Monitoring Instrumentation

1. Radioactive Liquid Effluent Monitoring Instrumentation LIMITING CONDITIONS OF OPERATION*

• The radioactive liquid effluent monitoring instrumentation channels shown in Table V.C-! 1 shall be OPERABLE with thdr Alarm(frip setpoints set to ensure that the limitsof Specification VD.L.a are not exceeded. The alarm/trip setpoirits shallbe determined in accordance with methodology and parameters as described in Section IT.,

APPLICABILITY: As shown in Tahble VC.- 1 ACTION:

a. With a radioactive liquid effluentmonitoring instrumentation channel Alarm/Trip setpoint less conservative than requiredby 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 Tlhble VC.-1. Exert best-efforts to restorethe 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 day's provided the specified actions aie continued.

SURVEILLANCE REUIJREMENTS 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

1. Operable 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 Requir6ements

a. Waste Neutralization Sump Effluents [. # 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 IxCi/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 TBble I.E-7 1and changed number of locations from 3 to 2 under Well Water:.

-1.1.4 Changed referenced document VB.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..-l 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. arid;

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 IiCi/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. Ifone 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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TABLEV.C.-2 Radioactive Liquid. Effluent Monitoring Instrumentation Surveillance Requirements Instrument Channel Source. Channel Channel Check Check Calibration Functional

_ __

• Test 1.Radioactivity Monitors Providing Alarm and Automatic Termination .Of Release

a. Waste Neutralization Sump Monitor D P R( 2) Q(M Condensate Polishing Facility. ... _.__._,___

b. Turbine Building Floor Drains D M '2) O(1)

c. Liquid Waste Monitor D P (2) Q(1)

d. Deleted

'e. Steam Generator Blowdown Monitor D M R(2) Q(1) 2.Flow Rate Measurements

a. Waste Neutralization Sump Effluents D( 3) NA QR

b. RESERVED

c. Liquid Waste Effluent Line D(3 ) N/A R Q d, Deleted

e. Steam Generator Blowdown Effluent D(3 ) N/A Q Line D Daily R= Once every 18 months M PirMonthly

=Prior to each batch release a = Once every 3 months P N/A= Not Applicable TABLE V.C.-2 TABLE NOTATION (1) The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occur if any 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 VD.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.

APPL CABILITY: As shown in Table VC.-3.

ACTIýN:

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 EffluientiRelease 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 VC.-4; MP-22--REC-BAP01 STOP " NT* T REVIEW Rev. 026-00 147 of 165

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-REIOA, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing Alarm 1 A

b. Iodine Sampler 1 B

c. Particulate Samplei '1 B

d. Vent Flow Rate Monitor .1 C

e. Sampler Flow Rate Monitor 1 C 2.Millstone Stack - applicable to SLCRS (I-HVR-RE19B, normal range only; high range monitor, HVR-RE19A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing

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 fun'ctioning of the instrument.. ..

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". *:,.'.'5.)i¸ 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 samptes 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 toa loss of annunciation in the control room and the Remote Indicating Controller (RIC) remains OPERABL*E,- 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 beverified 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-

,,:,quired 1.Millstone Unit 3 Ventilation' Vent (Turbine Building - HVR-RE1OB, norinarrange'only; high range monitor, HVR-RE10A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing )D M R(Q) Q(2) 1 Alarm . [

b. Iodine Sampler

c. Particulate Sampler

d. Vent Flow Rate Monitor

e. Sampler Flow Rate Monitor 2.Millstone Stack - applicable to:SLCRS (HVR- RE19B, normal range only; high range monh HVR-RRE19A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providinj Alarm

b. Iodine Sampler

c. Particulate Sampler

d. Process Flow Rate Monitor

e. Sampler Flow Rate Monitor 3.Engineered Safeguards Building Monit(

a. Noble Gas Activity Monitor Providinj Alarm

b. Iodine Sampler

c. Particulate Sampler

d. Discharge Flow Rate Monitor

e. Sampler Flow Rate Monitor

• At all times except when the vent path is isolated.

D = Daily R = Once every 18 months W = Weekly Q = Once every 3 months M = Monthly N/A= Not Applicable MP REC-BAPOI

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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 it 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 includethe 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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V.D. Radioactive Effluents Concentrations And Dose Limitations

1. Radioactive Liquid Effluents a.., Radioactive LiquidEffluents Concentrations LIMITING CONDITIONS OF OPERATION .

The-concentration of radioactive material released from the site (see Figure VD.-1) shallbe 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 gasesi :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 concentratioin of radioactive material released from the site exceeding the above limits, restore the concentration to within the above limits within 15 minutes.*

SURVEILLAN CE RE U IRMENTS,

1) Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Section 1.

2) The results of the radioactive analysis shall be used in accordance with the methods of Section 1I to assure that the concentrations at the point of release are maintained within the limits of Specification VD.1.a.

,L.-

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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 VD.-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 6alendar 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 V.D.-l) 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 Io.dine-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 1.500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the (dose rate(s) exceeding the above limits, decrease the release ra te to comply with the limit(s) given in Specification VD.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'VD.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 V.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 doserate 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) dud 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 equalto 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

'APPLICABILITY: At all timesl 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 calendat 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 Progra fi 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 V.D.-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 radiolodines, radioactive materials in particulate form, or radionuclides other than noble gases ini 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"

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V.E. Total Radiological Dose From Station Operations CONTROLS The annual dose or dose c6mmitment 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 tothe total body orl any organ (except the thyroid, which is limited to less than or equal to 75 mremr).

APPLICABILITY: At all times.

ACTION:-

With the calculated dose fromn the, release of radioactive materials in liquid or gaseous effluents. exceeding twice the limits of Controls V.D.IJb., V.D.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 OFTHE 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)-ate 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 variante-is granted until staff action on the request is complete.

SURVEILLANCE REOUIREMENTS Cumulative dose contributions from liquid and .gaseous effluenits and direct radiation from the Millstone Site shall be determined in accordance with Section I! once per 31 days.

VF. 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 ofliquid 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

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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 V.C.-2 Note 1 requires that the Analog Channel Operational Test (ACOT) demonstrate that automatic isolation and "control roori: annunciation" occur. Control room annunciation cannot occur unless the monitor is on line (i.e. in communication with the RMS computer.).

Section V.C.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. Thesealarms will clear when-discharge begins. The monitors are OPERABLE since they are able to perform their

.specified function with the expected alarms in.

Table V.C.-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 V.C.- 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 Specl4ble 3.3- ). The applicability added in Amendment 22 is the present wording. The letter 1312821,

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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 (i2 hours)?.

SIfn "'Significa'ntHazards Consideration" iteinrh 2: "The proposed changes also have y no effect on alarm setpoints or control functions. Further, no operator actions that arerequired to mitigate any accident rely solely ont'hese 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 Theradioactive 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 11 to ensure that the alarm/trip will occur prior to exceedingthe limits of SectionV.D.2.a. The OPERABILITY and use of this instrumentation is consistent withthe 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 vCi/cc are measurable.

The vent normal range radiation monitor, HVR*10B, satisfies the requirements of SectionVC.2. for Unit 3 releases to the vent which is located on the turbine MP-22-REC-BAPOI STOP THINK ACT REVIEW Rev. 026-00 160 of 165

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, ITVR*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 in10 CFR Part 20, Appendix B, Table 2.

This limitation provides additional assurance that the ievels of radioactive materials in bodies of water outside the site will resfilt in exposhires within: (1) the Section ILA 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) Publication2.

Sectioni V.D.I.b. - Radioactive Liquid Effluents Doses This specification is provided to implement the requirements of Sections II.A.,

II.A., and IWA. of Appendix 1, 10 CFR Part 50- The specification implements the guides set forth in Section ILA of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section hII.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 beconsistent with the methodologyprovided 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 1," April 1977.

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Section VD.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 frdm 500 to 100 torero. 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 cxceed'the setpoint and much lower- ariiai .dose'limits from 10CFR50, Appendix. I areimplemented. For individuals who may, at times, be within the SITE BOUNDARY, the occupancy of thatindi{iidual will be usually be sufficiently'low to compensate for any inci.ease in the atniospheric diffusion factor above that for the SITE BOUNDARY. The specified release rate limits restrict, atall'timies, the corresponding gamrmanand,beta dose rates above background to an' indiidual at or beyond the SITE BOUNDARY to less than or equal to 500 mtrem/year to the whole body or to less than or.equal to 3000 "rnrem/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/yearL Section VD.2.b. - Radioactive Gaseous Effluents Noble Gas Dose This specification is provided to implenen .the requirements of Sections II.B.,

IIJ.A., and IVA. of Appendix 1,10 CFR. Part 50. The specification implements the guides set forth in Section 1I.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the g*ides set forth in Section VA. of Appendix I to assure that theereleases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section II.A. of Appendix I that conformance with the, guides of Appendix I be shown by calculational procedures bd'sed 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 thedoses 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

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regulatory limits. Themethods used are, in order, previously determined air dose per released activity ratio, historical meteorological data and actual radionuclide mix released, of real time meteorology and actual radionuclides released.

Section V.D.2.c. - Radioactive Gaseous Effluents for Radionuclides Other Than Noble Gas These specifications are prdvided to implement the requiremients of Sections II.C., I.A., and IVA,~ of Appendix 1, 10 CFR Part 50. The specifications are the guides set forth in Section TI.C. of Appendix I. The ACTION statements provide the required operating flexibility a nd at the same time implenment the, guides set forth in Section IVA. 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 It 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,", Revisiou 1,'October 1977 and Regulatory Guide 1.1.11, "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 e.xamined 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) depositioin 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.- iibtal Radiol0gical 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 026-00

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.-i, "Millstone Unit 1,Radioactive Gaseous Waste Sampling and Analysis Program," under B. Balance of Plant Vent, 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.-72, "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, anguage 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 fi-om 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.

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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, II. "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 1.1), "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 ifthe auxiliary sampling equipment of Action B is in use."

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