ML081290514

Radioactive Effluent Release Report

ML081290514
Person / Time
Site:	Millstone
Issue date:	04/28/2008
From:	Price J Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
08-0187, FOIA/PA-2010-0209
Download: ML081290514 (249)

Text

Dominion Nuclear Connecticut, Inc. D

• Millstone Power Station Rope Ferry Road, Waterford, CT 06385 APR 2 8 2008 U.S. Nuclear Regulatory Commission Serial No. 08-0187 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 2007 RADIOACTIVE EFFLUENT RELEASE REPORT In accordance with 10 CFR 50.36a, this letter transmits the annual Radioactive Effluent Release Report for the period January 2007 through December 2007. This report meets the provisions of Section 5.7.3 of the Millstone Unit 1 Permanently Defueled Technical Specifications (PDTS), and Sections 6.9.1.6b and 6.9.1.4 of the Millstone Units 2 and 3 Technical Specifications, respectively. transmits Volume 1 of the 2007 Radioactive Effluent Release Report, in accordance with Regulatory Guide 1.21. Volume 1 contains information regarding airborne, liquid, and solid radioactivity released from Millstone, 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 2007. transmits Volume 2 of the report, which contains changes made to the REMODCM through 2007. Volume 2 consists of a complete copy of the REMODCM as of December 31, 2007, which satisfies the requirements of Sections 5.6.1c of the Millstone-Unit 1 PDTS, and Sections 6.15c and 6.9.13c of the Millstone 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, J ji~#~ric 4ite) ice President - Millstone 4K~r,

Serial No. 08-0187 2007 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 Mr. J. D Hughey Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 8B3 Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station

Serial No. 08-0187 2007 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. 08-0187 Docket Nos. 50-245 50-336 50-423 License Nos. DPR-21 DPR-65 NPF-49 ATTACHMENT 1 2007 RADIOACTIVE EFFLUENT RELEASE REPORT VOLUME 1 MILLSTONE POWER STATION UNITS 1, 2, AND 3 DOMINION NUCLEAR CONNECTICUT, INC. (DNC)

Millstone Power Station 2007 Radioactive Effluents Release Report Volume 1 Dominion Nuclear Connecticut, Inc.

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

Table of Contents Volume 1 T a b le of C o nte n ts ......................................................................................................................... 1 List o f T a b le s ............................................................................................................................... 2 Re fe re n c e s .................................................................................................................................. 3 Intro d u ctio 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 D o s e Re su lts .................................................................................................................. 7 1.2.1 Airborne Effluents ........................................................................................... 7 1.2.2 Liquid Effluents ............................................................................................. 7 1.2.3 Analysis of Results ......................................................................................... 7 2.0 Effluent Radioactivity ........................................................................................................... 12 2.1 Airborne Effluents .................................................................................................... 12 2.1.1 Measurement of Airborne Radioactivity ........................................................ 12 2.1.2 Estimate of Errors ......................................................................................... 14 2.1.3 Airborne Batch Release Statistics ................................................................. 15 2.1.4 Abnormal Airborne Releases ....................................................................... 15 2.2 Liquid Effluents ............................................................................................................ 33 2.2.1 Measurement of Liquid Radioactivity ........................................................... 33 2.2.1.1 Continuous Liquid Releases ............................................................... 33 2.2.1.2 Liquid Tanks/Sumps .......................................................................... 33 2.2.2 Estimate of Errors ......................................................................................... 33 2.2.3 Liquid Batch Release Statistics ................................................................... 34 2.2.4 Abnormal Liquid Releases ............................................................................ 34 2.2.5 Liquid Release Tables ................................................................................... 35 2 .3 S o lid W a ste ................................................................................................................. 48 2.4 Groundwater Monitoring ........................................................................................... 70 3.0 Inoperable Effluent Monitors ........................................................................................... 72 4.0 Operating History ................................................................................................................ 73 5 .0 E rra ta .................................................................................................................................. 75 6.0 REMODCM Changes .......................................................................................................... 76 6.1 REMODCM Description of Changes ..................................................................... 77 Volume 2 2007 REMODCM Revision 25-02 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-Al Unit 1 Airborne Effluents -. Release Summary Table 2.1-A2 Unit 1 Airborne Effluents - Ground Continuous - Balance of Plant Vent

& Spent Fuel Pool Island Vent Table 2.1-L1 Unit 1 Liquid Effluents - Release Summary 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 - Quarry Table 2.2-L2 Unit 2 Liquid Effluents - Continuous - SGBD Table 2.2-L3 Unit 2 Liquid Effluents - Batch - LWS Table 2.2-L4 Unit 2 Liquid Effluents - Continuous - Turbine Building Sump - Yard Drain DSN 006 Table 2.3-Al Unit 3 Airborne Effluents - Release Summary Table 2.3-A2 Unit 3 Airborne Effluents - Mixed Continuous - Vent & 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 & Containment Vents Table 2.3-A7 Unit 3 Airborne Effluents - Ground Batch - Containment Equipment Hatch Table 2.3-A8 Unit 3 Airborne Effluents - Ground Batch- RWST Vent Table 2.3-L1 Unit 3 Liquid Effluents - Release Summary - Quarry Table 2.3-L2 Unit 3 Liquid Effluents - Continuous - SGBD, SW & TK2 Table 2.3-L3 Unit 3 Liquid Effluents - Batch - LWS Table 2.3-L4 Unit 3 Liquid Effluents - Batch - CPF Waste Neutralization Sumps & Hotwell Discharge Table 2.3-L5 Unit 3 Liquid Effluents - Release Summary - Yard Drain DSN 006 Table 2.3-L6 Unit 3 Liquid Effluents - Continuous - Turbine Building Sump Table 2.3-L7 Unit 3 Liquid Effluents - Batch - WTT Berm Water Table 2.3-L8 Unit 3 Liquid Effluents - Continuous - Foundation Drain Sumps 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-GWl Groundwater Monitoring Results Table 2.4-GW2 Groundwater Monitoring 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, ORNUNUREG/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 Light-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 No. MP-HPO-08016, 2007 Report on Solid Waste and Irradiated Component Shipments, April 10, 2008.

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Introduction This report, for the period of January through December of 2007, 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 2007 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 defined as the dose to the individual who would receive the maximum dose from releases of airborne and liquid effluents. Although the location of the maximum individual may vary each quarterly period, the annual dose is the sum of these quarterly doses. This conservatively assumes that the individual is at the location of maximum dose each quarter.

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

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

The values of average relative effluent concentration (X/Q) and average relative deposition (D/Q) used in the DOSAIR code were generated using EDAN4, a meteorological computer code which implements the assumptions cited in NRC Regulatory Guide 1.111 (Reference 5), Section C. The 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 Vent (SFPI) and the Balance of Plant Vent (BOP).

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 evaporation from the spent fuel pool water (H-

3) release to 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. Some Containment Purges are 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 "initial" Containment Drawdown (through 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 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 xIQ 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 yIQ 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 2007 Off-Site Dose Commitments from Airborne Effluents Millstone Units 1, 2, 3 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Air (mrad) (mrad) (mrad) (mrad) (mrad)

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

Whole Body 3.17E-05 2.86E-04 3.23E-04 3.02E-05 6.71 E-04 Skin 3.17E-05 2.90E-04 3.23E-04 3.02E-05 6.75E-04 Thyroid 3.17E-05 2.73E-04 3.23E-04 3.02E-05 6.58E-04 Max organ+ 3.17E-05 3.63E-04 3.23E-04 3.02E-05 7.48E-04 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Air (mrad) (mrad) (mrad) (mrad) (mrad)

Beta 1.80E-04 8.52E-05 3.35E-04 1.15E-04 7.15E-04 Gamma 5.32E-05 1.17E-04 9.43E-05 3.26E-05 2.98E-04 Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 3.50E-04 6.84E-04 6.63E-04 2.16E-04 1.91 E-03 Skin 4.79E-04 7.52E-04 8.36E-04 3.01 E-04 2.37E-03 Thyroid 4.97E-04 1.68E-03 2.40E-03 3.05E-04 4.88E-03 Max organ+ 3.52E-04 6.86E-04 6.71 E-04 2.17E-04 1.93E-03 Ist Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Air (mrad) (mrad) (mrad) (mrad) (mrad)

Beta 1.61 E-05 1.37E-02 1.08E-03 3.35E-04 1.51 E-02 Gamma 2.OOE-05 3.57E-03 6.66E-05 2.31 E-05 3.68E-03 Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 7.56E-04 2.16E-02 2.89E-03 7.52E-04 2.60E-02 Skin 7.73E-04 2.70E-02 3.28E-03 9.74E-04 3.20E-02 Thyroid 7.56E-04 1.19E-01 2.86E-03 7.52E-04 1.23E-01 Max organ+ 7.56E-04 3.34E-02 2.91 E-03 7.56E-04 3.79E-02

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

Table 1-2 2007 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 (mrem)

Annual Total (mrem)

Whole Body 0.00E+00 0.OOE+00 0.00E+00 0.00E+00 0.OOE+00 Thyroid 0.00E+00 0.00E+00 O.OOE+00 0.OOE+00 0.OOE+00 Max Orqan O.OOE+00 I O.OOE+00 0.OOE+00 I 0.OOE+00 I 0.00E+00 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 2.36E-05 3.08E-05 4.69E-05 6.37E-05 1.65E-04 Thyroid 8.29E-06 9.20E-06 2.34E-05 3.88E-05 7.97E-05 Max Organ 1.01E-04 1.94E-04 3.14E-04 3.02E-04 9.11E-04 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Annual Total Max Individual (mrem) (mrem) (mrem) (mrem) (mrem)

Whole Body 4.19E-04 1.39E-04 1.36E-04 7.01 E-05 7.64E-04 Thyroid 3.41 E-04 2.35E-04 3.82E-05 3.74E-05 6.51 E-04 Max Organ 8.45E-04 4.14E-04 6.86E-04 1.96E-04 2.14E-03 9

Table 1-3 2007 Off-Site Dose Comparison to Limits Millstone Units 1, 2, 3 Airborne Effluents Max Individual Dose vs REMODCM Limits Whole Body Thyroid Max Organ** Skin Beta Air Gamma Air (mrem) (mrem) (mrem) (mrem) (mrad) (mrad)

Unit 1 6.71 E-04 6.58E-04 7.48E-04 6.75E-04 O.OOE+00 O.OOE+00 Unit 2 1.91 E-03 4.88E-03 1.93E-03 2.37E-03 7.15E-04 2.98E-04 Unit 3 2.60E-02 1.23E-01 3.79E-02 3.20E-02 1.51 E-02 3.68E-03 Millstone Station 2.85E-02 1.29E-01 4.05E-02 3.50E-02 1.58E-02 3.98E-03 REMODCrMeL-ImirtllJ -- 15" -- 1 [

Liquid Effluents Max IndividualDose vs REMODCM Limits Whole Body Thyroid Max Organ**

(mrem) (mrem) (mrem)

Unit 1 0.OOE+00 0.OOE+00 0 .OE+00 Unit 2 1.65E-04 7.97E-05 9.11 E-04 Unit 3 7.64E-04 6.51E-04 2.14E-03 Millstone Station 9.29E-04 7.31 E-04 3.05E-03 Millstone Station Total Off-Site Dose Max Individual Dose vs 40CFRI90 Limits Whole Body Thyroid Max Organ **

Airborne Effluents 2.85E-02 1.29E-01 4.05E-02 Liquid Effluents 9.29E-04 7.31 E-04 3.05E-03 Radwaste Storage 1.80E-01 1.80E-01 1.80E-01 Millstone Station 2.09E-01 3.09E-01 2.24FO1- I

• IOCFR50, Appendix I Guidelines Maximum of the following organs (not including Thyroid): Bone, GI-LLI, Kidney, Liver, Lung 10

Table 1-4 2007 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 I - 284 mreml Courtesy NCRP Report 94 (1987)

Maximum Off-Site Individual Millstone Station Whole Body Dose Airborne Effluents 0.0285 mrem Liquid Effluents 0.0009 mrem On-site RadWaste Storage .0.1800 mrem I 0.2095 mremJ 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.

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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).

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.

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:

Sampling/Data Collection 10% Variation in data collection Sample Line Loss 20% Deposition of some nuclides Calibration 10% Calibration to NBS standards Sample Counting 10% Error for counting statistics Flow & Level Measurements 10% Error for release volumes 14

2.1.3 Airborne Batch Release Statistics Unit I - None Unit 2 Ctmt Purges Ctmt Vents WGDT Number of Batches 0 44 7 Total Time (min) 0 6559 3791 Maximum Time (min) 0 320 768 Average Time (min) 0 223 542 Minimum Time (min) 0 67 340 Unit 3 Ctmt Purges Ctmt Vents Ctmt Drawdowns Number of Batches 1 239 1 Total Time (min) 409

• 239 Maximum Time (min) 409

• 239 Average Time (min) 409

• 239 Minimum Time (min) 409

• 239

• ~ 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 I - None 2.1.4.2 Unit 2- None 2.1.4.3 Unit 3- None 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 1 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total D. Gross Alpha

1. Total Activity Ci dash "-" denotes less than Minimum Detectable Activity (MDA) 16

Table 2.1-A2 Millstone Unit 1 Airborne Effluents Ground Continuous- Balance of Plant Vent & Spent Fuel Pool Island Vent Nuclides 0I Released I Units 1st Qtr I 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Gases jKr-85 I Ci I - - I - I - 1 Total Activity I - - -

B. lodines11-131 ai Total Activity Ci -

C. Particulates lCs-137 I Ci I - 1.78E-06 1

,.78E-06I ITotal Activity I - 1.78E-06 - - 1.78E-06 D. Gross Alpha IGross Alpha I c - - - I E. Tritium JH-3 Ci 1.02E-01 1.82E-01 2.68E-01 1.92E-01 7.44E-01 dash denotes less than Minimum Detectable Activity (MDA) 17

Table 2.2-Al Millstone Unit No. 2 Airborne Effluents - Release Summary Units 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total D. Gross Alpha I1. Total Activity Ci - - - I I Released . 1C

"-" denotes less than Minimum Detectable Activity (MDA) 18

Table 2.2-A2 Millstone Unit No. 2 Airborne Effluents - Mixed Continuous - Aux Bldg Vent & SGBD Tank Vent

& Spent Fuel Pool Evaporation

• Nuclides U200 Released Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases IXe-133 1 Ci I - 1.50E-01 Total Activity Ci - 1.50E-01 I 1.50E-01 B. lodines 1-131 Ci 2.83E-05 2.22E-05 2.32E-05 1.19E-05 8.55E-05 1-133 Ci 1.10E-04 7.69E-05 9.10E-05 5.24E-05 3.30E-04 1-135 Ci 4.39E-05 - 4.39E-05 Total Activity Ci 1.38E-04 1.43E-04 1.14E-04 6.43E-05 4.59E-04 C. Particulates 1-131 Ci -

Cs-1 37 Ci - 5.67E-07 5.67E-07 Mn-54 Ci - 2.92E-07 2.92E-07 Sr-89 Ci -

Sr-90 Ci Total Activity Ci 8.60E-07 8.60E-07 D. Gross Alpha IGross Alpha Ci - - -

E. Tritium

• H-3 Ci 2.07E+00 1.84E+00 1.31E+00 1.71E+00 6.94E+00

• Includes estimated Spent Fuel Pool evaporation

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

Table 2.2-A3 Millstone Unit No. 2 Airborne Effluents - Mixed Batch - Containment Purges I uclidesI i°i Released I Units 1st Qtr I 2nd Qtr 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases lGamma EmittersI Ci I na na na 1 na na Total Activity I Ci I na na na na na B. lodines

• S* ICi * * * ....

Total Activity Ci ......

C. Particulates

• I

• Cii Total Activity I.Ci I .

D. Gross Alpha

• IGrossAlpha Ci I* * * *

• E. Tritium IH-3 Ci na na na na na

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

,v, denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed I .no releases via this pathway for the quarter 20

Table 2.2-A4 Millstone Unit No. 2 Airborne Effluents - Elevated Batch - WGDT Nuclides Unt00 Released Units 1st Qtr I 2nd Qtr 3rd Qtr I 4th Qtr j Total j A. Fission & Activation Gases Ar-41 Ci 2.82E-04 2.82E-04 Kr-85 Ci 4.72E-01 4.91 E-02 1.26E-01 2.59E-01 9.06E-01 Kr-85m Ci 6.1 OE-05 6.1OE-05 Kr-88 Ci 7.19E-05 7.19E-05 Xe-131m Ci 2.01E-04 - 2.01E-04 Xe-1 33 Ci 2.77E-05 1.44E-03 1.23E-04 2.52E-03 4.11E-03 Xe-1 33m Ci - 1.35E-04 1.35E-04 Xe-135 Ci - 1.35E-03 1.35E-03 Total Activity Ci 4.72E-01 5.05E-02 1.26E-01 2.63E-01 9.12E-01 B. lodines

• IT c*

• I ** 1 "

• 1 Total Activity I.Ci I C. Particulates

• I

• Cii ..... I Total Activity Ci ..

D. Gross Alpha*

[GrossAlpha I Ci *

• E. Tritium IH-3 I Ci I 3.58E-04 2.40E-05 1.58E-04 4.53E-04 I 9.93E-04

• 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 No. 2 Airborne Effluents - Elevated - Containment Vents/Site Stack Nuclides ]

Released I Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases Ar-41 Ci 2.59E-02 2.81 E-02 2.76E-02 2.86E-02 1.10E-01 Kr-85 Ci 2.48E-03 1.94E-02 1.02E-01 3.56E-01 4.79E-01 Xe-1 33 Ci 4.84E-02 5.23E-02 5.67E-02 5.28E-02 2.10E-01 Xe-1 35 Ci 2.21E-04 1.97E-04 2.31E-04 2. 01E-04 8.50E-04 Total Activity Ci 7.70E-02 9.99E-02 1.86E-01 4.37E-01 8.00E-01 B. lodines11-131 Cii - I -- ' - -

ITotal Activity Ci I I C. Particulates 1-131 Ci Co-58 Ci 3.20E-07 - 3.20E-07 Sr-89 Ci -

Sr-90 Ci -

Total Activity Ci 3.20E-07 - 3.20E-07 D. Gross Alpha lGross Alpha Ci - - 7 7 E. Tritium IH-3 Ci 7.26E-02 1.54E-01 2.02E-01 4.92E-01 9"20E-01 denotes less than Minimum Detectable Activity (MDA) 22

Table 2.2-A6 Millstone Unit No. 2 Airborne Effluents - Ground Batch - Containment Equipment Hatch Nuclides 2I07 Released I Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr [ Total A. Fission & Activation Gases IGamma Emittersl Ci I na I na I na I na na Total Activity I Ci I na na na na na B. lodines11-131 1 Cii na na I na I na I na Total Activity Ci na na na na na C. Particulates T I Ci na na na na na Total Activity Ci na na na na na D. Gross Alpha IGross Alpha Ci na na na na na E. Tritium IH-3 Ci na na na na na

'v' denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed I Jno releases via this pathway for the quarter 23

Table 2.2-A7 Millstone Unit No. 2 Airborne Effluents - Ground Batch - RWST Vent Nuclides i Released I Units I 1,st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Gases IGamma Emitersl Ci I na na I na na na Total Activity Ci I na na na na na B. lodines11-131 1 naI I na na na I na Total Activity Ci na na na na na C. Particulates.

I Ci na na na na na Total Activity Ci na na na na na D. Gross Alpha IGross Alpha I Ci I na I na I na I na na E. Tritium IH-3 Ci na na na na nao

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

"na" denotes Not Analyzed I I no releases via this pathway for the quarter 24

Table 2.3-Al Millstone Unit No. 3 Airborne Effluents - Release Summary Units 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total D. Gross Alpha

1. Total Activity Ci Released I

"-" denotes less than Minimum Detectable Activity (MDA) 25

Table 2.3-A2 Millstone Unit No. 3 Airborne Effluents - Mixed Continuous - Normal Ventilation &

Spent Fuel Pool Evaporation Nuclides lI Released I Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases I~-13 iI - I1.78E+01 I - - 1.78E+01 Total Activity Ci - 1.78E+01 - 1.78E+01 B. lodines 1-131 Ci - 1.93E-03 1.93E-03 1-132 Ci - 2.33E-05 2.33E-05 o-133 Ci - 2.54E-05 2.54E.-05 Total Activity Ci - 1.98E-03 1.98E-03 C. Particulates 1-131 Ci - -

Co-57 Ci - - 6.03E-07 6.03E-07 Co-58 Ci 1.25E-04 6.13E-06 - 1.31E-04 Co-60 Ci 2.71 E-05 6.48E-06 - 3.35E-05 Cr-51 Ci 1.96E-04 - - 1.96E-04 Cs-1 34 Ci 2.84E-06 - - 2.84E-06 Mn-54 Ci 4.55E-06 - 4.55E-06 Nb-95 Ci 1.14E-05 - 1.14E-05 Sn-117m Ci - 1.38E-06 1.38E-06 Sr-89 Ci - -

Sr-90 Ci - - -

Zr-95 Ci - 7.19E-06 7.19E-06 Total Activity Ci 3.74E-04 1.26E-05 1.98E-06 3.89E-04 D. Gross Alpha IGross Alpha I Ci I - - I E. Tritium IH-3 Ci 6.07E+00 2.63E+01 8.32E+00 1.14E+01 5.21E+01

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

Table 2.3-A3 Millstone Unit No. 3 Airborne Effluents - Ground Continuous - ESF Building Ventilation Nuclides Released i Units II 1st Qtr I 2nd Qtr I 3rd][SQ Qtr I 4th Qtr I Total A. Fission -& Activation Gases _________ I_

Xe-1 31 m Ci 4.79E-02 4.79E-02 Xe-1 33 Ci 4.40E-03 4.40E-03 Total Activity Ci 5.23E-02 5.23E-02 B. lodines j_ 93 E _ 5 I 2 7 08 1-131 Ci 9.34E-05 2.27E-08 9.34E-05 1-133 Ci 2.19E-05 2.19E-05 Total Activity Ci 1.15E-04 2.27E-08 1.15E-04 C. Particulates 1-131 Ci -

Ba-140 Ci 2.30E-07 - 2.30E-07 Be-7 Ci 5.05E-07 1.75E-06 - 2.25E-06 Ce-141 Ci - 3.11E-08 3.11E-08 Co-58 Ci 2.69E-07 5.64E-07 8.33E-07 Co-60 Ci 9.19E-08 4.25E-07 1.03E-07 6.20E-07 Sr-89 Ci -

Sr-90 Ci - -

Total Activity Ci 5.05E-07 2.34E-06 1.02E-06 1.03E-07 3.96E-06 D. Gross Alpha IGross Alpha I Ci I - I I E. Tritium H-3 Ci 1 1.76E-01 1.05E-01 1.69E-01 1.79E-01 6.29E-01

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

Table 2.3-A4 Millstone Unit No. 3 Airborne Effluents - Mixed Batch - Containment Drawdowns Nuclides Us Released n ts I1st Qtr I 2nd Qtr 3rd Qtr

• 4th Qtr Total A. Fission & Activation Gases IGamma Emitters ICi I na. I na6 na -

Total Activity Ci na na I na B. lodines11-131 1 Ci I na I4.08E-08 I na I na, 4.08E-08 Total Activity Ci na 4.08E-08 na na 4.08E-08 C. Particulates I Ci na. I " na I na -

Total Activity Ci na na na D. Gross Alpha IGrossAlpha Ci I na I na na I na I na E. Tritium IH-3 Ci na 7.33E-04 na na 7.33E-04 denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed l, I no releases via this pathway for the quarter 28

Table 2.3-A5 Millstone Unit No. 3 Airborne Effluents - Mixed Batch - Containment Purges Nuclides Released I0 Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases ,.

Kr-85 Ci na 1.65E-01 na na 1.65E-01 Kr-85m Ci na 1.49E-03 na na 1.49E-03 Xe-131m Ci na 2.09E-02 na na 2.09E-02 Xe-1 33 Ci na 9.50E-01 na na 9.50E-01 Xe-1 33m Ci na 1.96E-02 na na 1.96E-02 Xe-I 35 Ci na 6.87E-02 na na 6.87E-02 Total Activity Ci na 1.23E+00 na na 1.23E+00 B. lodines*

I Cii Total Activity Ci..

C. Particulates

• I

• Cii Total Activiy Ci I I I I D. Gross Alpha*

IGrossAlpha I Ci I

• I * * *

• E. Tritium H-3 Ci na 2.64E-02 na na 2.64E-02

• lodines, Particulates and Gross Aipha included in Tables 2.3-A2 or 2.3-A6

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

"na" denotes Not Analyzed I J no releases via this pathway for the quarter 29

Table 2.3-A6 Millstone Unit No. 3 Airborne Effluents - Elevated Continuous - Gaseous Waste System

& Containment Vents Nuclides ilr Released I Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases Ar-41 Ci 1.1OE-02 5.77E-03 6.94E-03 7.31E-03 3.1OE-02 Kr-85 Ci 6.68E+00 2.93E+00 1.80E+00 1.14E+01 Kr-85m Ci 4.26E-04 5.71 E-05 1.84E-04 6.68E-04 Kr-87 Ci 5.82E-05 - 5.82E-05 Kr-88 Ci - 8.52E-03 8.52E-03 Xe-1 31 m Ci 1.82E-03 2.50E-01 7.48E-04 3.72E-03 2.57E-01 Xe-1 33 Ci 3.23E-02 1.34E+01 3.04E-01 4.66E-01 1.42E+01 Xe-133m Ci 2.22E-01 4.04E-03 7.14E-03 2.33E-01 Xe-135 Ci 2.16E-02 6.16E-01 1.32E-02 6.64E-03 6.58E-01 Total Activity Ci 6.68E-02 2.12E+01 3.26E+00 2.29E+00 2.68E+01 B. lodines 1-131 Ci 3.93E-08 8.08E-05 8.08E-05 1-132 Ci 4.26E-08 4.26E-08 1-133 Ci 2.54E-06 2.54E-06 Total Activity Ci 3.93E-08 8.34E-05 8.34E-05 C. Particulates ....

1-131 Ci Br-82 Ci 3.46E-06 - 6.04E-06 7.34E-06 1.68E-05 Co-57 Ci 1.74E-08 - 1.74E-08 Co-58 Ci 3.31 E-06 - 6.44E-08 3.37E-06 Co-60 Ci 3.34E-07 8.11E-07 - 1.33E-07 1.28E-06 Cr-51 Ci - 8.72E-07 - 8.72E-07 Mn-54 Ci - 3.35E-08 - 3.35E-08 Sb-1 25 Ci - 5.08E-08 5.08E-08 Sr-89 Ci -- -

Sr-90 Ci -...

Total Activity Ci 3.79E-06 5.09E-06 6.04E-06 7.54E-06 2.25E-05 D. Gross Alpha IGross Alpha I Ci I - - -

E. Tritium 1H-3 I Ci I 1.06E+00 I 9.49E-01 I 2.03E-01 I 2.71E-01 1 2.49E+O00

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

Table 2.3-A7 Millstone Unit No. 3 Airborne Effluents - Ground Batch - Containment Equipment Hatch MONuclides- W 2007 Released I Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases IGamma EmittersI Ci I na na na na I na Total Activity I Ci I na na na na na B. lodines11-131 cCi I na 9.60E-06 I na I na 1 9.60E-06 Total Activity Ci na 9.60E-06 na na 9.60E-06 C. Particulates 1-131 Ci na na na -

Co-58 Ci na 2.03E-05 na na 2.03E-05 Co-60 Ci na 3.89E-06 na na 3.89E-06 Cr-51 Ci na 2.37E-05 na na 2.37E-05 Fe-59 Ci na 7.75E-07 na na 7.75E-07 Mn-54 Ci na 1.84E-06 na na 1.84E-06 Nb-95 Ci na 4.87E-06 na na 4.87E-06 W-187 Ci na 1.78E-07 na na 1.78E-07 Zr-95 Ci na 2.90E-06 na na 2.90E-06 Total Activity Ci na 5.84E-05 na na 5.84E-05 D. Gross Alpha IGross Alpha Ci na na na na na E. Tritium 1H-3 Ci I na I na I na I na I na

"" denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed I I no releases via this pathway for the quarter 31

Table 2.3-A8 Millstone Unit No. 3 Airborne Effluents - Ground Batch - RWST Vent MNu clidesN

= 2 0'T t Released I Units I 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Gases IXe-1 33 Ci na 1.40E-01 na na 1.40E-01 Total Activity Ci na 1.40E-01 na na I1.40E-01 B. lodines 1-131 Ci na 1.14E-04 na na 1.14E-04 I-132 Ci na 2.64E-07 na na 2.64E-07 Total Activity Ci na 1.14E-04 na na 1.14E-04 C. Particulates 1-131 Ci na - na na Ag-I 1Om Ci na 1.62E-07 na na 1.62E-07 Co-57 Ci na 4.05E-07 na na 4.05E-07 Co-58 Ci na 1.41 E-04 na na 1.41 E-04 Co-60 Ci na 1.02E-05 na na 1.02E-05 Cr-51 Ci na 1.27E-04 na na 1.27E-04 Cs-134 Ci na 1.04E-04 na na 1.04E-04 Cs-1 36 Ci na 4.78E-06 na na 4.78E-06 Cs-1 37 Ci na 5.73E-05 na na 5.73E-05 Fe-59 Ci na 2.44E-06 na na 2.44E-06 Mn-54 Ci na 9.76E-06 na na 9.76E-06 Nb-95 Ci na 1.49E-05 na na 1.49E-05 Sb-1 24 Ci na 2.70E-07 na na 2.70E-07 Sb-1 25 Ci na 2.45E-06 na na 2.45E-06 Sn-113 Ci na 2.03E-07 na na 2.03E-07 Zn-65 Ci na 1.37E-07 na na 1.37E-07 Zr-95 Ci na 1.03E-05 na na 1.03E-05 Total Activity Ci na 4.85E-04 na na 4.85E-04 D. Gross Alpha IGross Alpha Ci na na na na na E. Tritium H-3 Ci na na na I na na

'v' denotes less than Minimum Detectable Activity (MDA)

"na" denotes Not Analyzed EZZZ no releases via this pathway for the quarter 32

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 I 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.

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

Sampling/Data Collection 10% Variation in data collection Calibration 10% Calibration to NBS standards Sample Counting 10% Error for counting statistics Flow & Level Measurements 10% Error for release volumes 33

2.2.3 Liquid Batch Release Statistics Unit I Unit 2 Unit 3 Number of Batches 0 44 100 Total Time (min) 0 3749 17874 Maximum Time (min) 0 231 235 Average Time (min) 0 85 179 Minimum Time (min) 0 18 65 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 2007, the following abnormal liquid releases occurred:

2.2.4.1 Unit I - None 2.2.4.2 Unit 2 - None 2.2.4.3 Unit 3 -

Unplanned Releases from Foundation Drain Sump 3 Several additional ground water monitoring locations have been phased in since the beginning of the Groundwater Protection Initiative (GPI). The Unit 3 Foundation drains (3 SRW Sumps 2 and 3) were added to the ground water monitoring plan in 2007. These sumps are part of the Unit 3 Foundation Underdrain System. Its purpose is to collect and move ground water away from building foundations thereby preventing the in-leakage of groundwater. Several samples were obtained in the first half of 2007. These samples showed low levels of tritium. In November, 2007, tritium was detected in 3SRW Sump No. 3 in excess of the EPA drinking water limit (20,000 pCi/liter). The highest measured level of activity was 34,000 pCi/liter. Exceedance of the EPA value caused a special report to station key stakeholders as listed in RAC 14, Non-Emergency Station Events. Notification of stakeholders is a condition of the recently invoked NEI Guideline 07-07, Groundwater Protection Initiative. Since this release of tritium to ground water was not expected, it is an unplanned release.

An investigation was performed to determine the source and the extent of the radioactive plume and to ensure that no sources of public drinking water were affected. This was accomplished by sampling existing groundwater monitoring wells and additional subterranean cavities (e.g. sumps, piping leading to these sumps, cable vaults, man ways, and storm drains). Three foundation drain pipes discharge into 3SRW Sump 3. These three pipes are described as follows: (1) a pipe at elevation 7.33 feet that runs northward from the sump toward the Unit 3 Waste Test Tank Berm; (2) a pipe at elevation 0.83 feet that runs westward toward the Unit 3 Fuel Pool and (3) a pipe that runs southward toward the Unit 3 RWST. Visual examinations of these pipes indicated that the water was flowing in from only pipes (1) and (3). Analyses of the water from these pipes indicated that pipe (1) did not have any detectable tritium and pipe (3) contained tritium at an approximate concentration of 40,000 pCi/liter. Pipe (2) was typically dry.

34

Further investigation into the source of this tritium led to examining the RWST pit. This pit contained approximately 200 gallons of water with tritium at approximately 4,000,000 pCi/liter. This was a sufficient source to produce the levels of tritium that were detected in 3SRW Sump No. 3. Efforts to contain this pit confirmed it as the source of tritium in this sump. Examination of potential inputs to the RWST pit indicated the RWST overflow line terminated here. Measurements of the water condensing in this line indicated that the source was RWST vapor (condensation). The corrective action to prevent reoccurrence is to perform a design change to the overflow line such that it will be directly piped to radwaste processing.

Review of the site hydrology and data from sampling other wells and several additional water sampling locations, indicated that the contamination was limited to the Unit 3 Refueling Water Storage Tank (RWST) yard. The Foundation Drain system limits the spread of any underground plume in this area by channeling water to the sumps which is pumped into the yard drain system. The releases via the yard drain system are listed in Table 2.3-L5 and the detailed measurements are listed in Table 2.4-GW. The calculated doses for 2007 resulting from these releases are as follows:

Total body dose: 4.8 E-06 mrem Max Organ dose: 4.8 E-06 mrem 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 No. 1 Liquid Effluents - Release Summary (Release Point - Quarry)

<< No Liquid Discharges>>

I Untsl 1sttr 2nd Qtr I 3rd Qtr I 4th Qtr Total B. Tritium C.

D. Gross Alpha

1. Total Activity ci na na na na na Released E. Volume

1. Released Waste Liters 0 0 0 0 0 Volume

2. Dilution Volume Liters 0 0 0 0 0 During Releases

3. DilutionVolume Liters 0 0 0 0 0 During Period(1 )

36

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

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

1. Total Activity 7 Ci Released E. Volume

1. Released Waste Liters 1.40E+05 2.75E+07 3.11E+07 1.06E+07 6.93E+07 Volume

2. Dilution Volume Liters 1.55E+09 1.20E+09 2.16E+09 2.27E+09 7.18E+09 During Releases

3. Dilution Volume Liters 2.68E+11 2.66E+11 2.86E+11 2.85E+11 1.11E+12 During Period

"-"denotes less than Minimum Detectable Activity (MDA) 37

Table 2.2-L2 Millstone Unit No. 2 Liquid Effluents - Continuous - SGBD, SW, RBCCW (Release Point - Quarry)

Nuclides=15 Released I UnitsI lstQtr I 2ndQtr I 3rdQtr I 4th Qtr I Total A. Fission & Activation Products IGamma EmittersI CiI - I -

Total Activi Ci I - I - I I B. Tritium IH-3 I Ci 2.05E-05I 6.42E-02 6.52E-02 1.11E-02 1.40E-01 C. Dissolved & Entrained Gases IGammaEmittersl CiI -

Total Activity I - - -

D. Gross Alpha IGross Alpha I Ci I na na I na na na

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

°na" denotes Not Analyzed 38

Table 2.2-L3 Millstone Unit No. 2 Liquid Effluents - Batch - LWS (Release Point - Quarry)

Nuclides Released IUnitsI 1st Qtr I 2nd Qtr I 3rd Qtr .I 4th Qtr I Total A. Fission & Activation Products Ag-11im Ci 1.08E-05 4.08E-05 8.18E-05 6.52E-05 1.99E-04 Co-57 Ci - 2.44E-06 1.26E-06 3.70E-06 Co-58 Ci 4.20E-04 1.01 E-04 1.27E-04 8.41E-05 7.32E-04 Co-60 Ci 9.12E-05 6.94E-05 4.96E-04 1.24E-03 1.90E-03 Cr-51 Ci - 1.74E-05 - 1.74E-05 Cs-1 34 Ci 2.64E-04 2.20E-04 2.49E-04 2.47E-04 9.80E-04 Cs-137 Ci 3.96E-04 3.85E-04 4.77E-04 5.01E-04 1.76E-03 Fe-55 Ci- 6.63E-04 1.05E-03 1.33E-03 -1.05E-03 4.09E-03 Fe-59 Ci 3.45E-06 - - 3.45E-06 1-131 Ci - 2.08E-06 - - 2.08E-06 Mn-54 Ci - 1.05E-06 1.33E-05 2.64E-05 4.08E-05 Nb-95 Ci 2.72E-06 2.96E-06 1.40E-06 7.08E-06 Ni-63 Ci 5.26E-04 8.54E-04 1.38E-03 1.54E-03 4.29E-03 Ru-1 05 Ci 4.86E-06 - - 4.86E-06 Sb-122 Ci 4.97E-06 2.1OE-06 7.08E-06 Sb-124 Ci 3.45E-05 3.11E-06 - - 3.76E-05 Sb-125 Ci 2.17E-03 1.88E-03 4.01E-03 2.66E-03 1.07E-02 Sn-113 Ci 3.06E-05 1.87E-05 1.18E-05 1.78E-05 7.89E-05 Sr-89 Ci - 6.45E-05 2.15E-05 - I 8.60E-05 Total Activity Ci 4.62E-03 4.69E-03 8.22E-03 7.43E-03 2.50E-02 B. Tritium IH-3 Ci 8.89E+00 9.78E+00 4.92E+01 8.12E+01 1.49E+02 C. Dissolved & Entrained Gases Kr-85 Ci 9.82E-03 1.23E-02 2.27E-02 3.88E-02 8.36E-02 Xe-133 Ci 1.38E-05 1.82E-04 4.61E-04 1.93E-03 2.59E-03 Total Activity Ci 9.84E-03 1.24E-02 2.31 E-02 4.08E-02 8.62E-02 D. Gross Alpha lGross Al ha I Ci I - -

"-"denotes less than Minimum Detectable Activity (MDA) 39

Table 2.2-L4 Millstone Unit No. 2 Liquid Effluents -Continuous-Turbine Building Sump (Release Point - Yard Drain - DSN 006)

Nuclides Us Released Units I 1st Qtr I 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission & Activation Products Gamaemitters Ci -

TotalActivity Ci[ I B. Tritium H-3 Ci 7.37E-03 3.81E-03 3.65E-03 1.16E-02 2.64E-02 uCi/m Average Period I 2.14E-07 1.14E-07 8.54E-08 3.11E-07 1.79E-07 Diluted Activity C. Dissolved & Entrained Gases Gamma/emitters Ci - I I Total Activit Ci ----

D. Gross Alpha Gross Alpha I ci na na na na na

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

"na' denotes Not Analyzed 40

Table 2.3-L1 Millstone Unit No. 3 Liquid Effluents - Release Summary (Release Point - Quarry)

Units 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr Total D. Gross Alpha

1. Total Activity C, - - "

Released T E. Volume

1. Released Waste Liters 8.03E+06 3.14E+06 6.67E+06 8.12E+06 2.60E+07 Volume

2. DilutionVolume Liters 2.33E+10 6.74E+09 1.95E+10 1.66E+10 6.61E+10

. During Releases

3. DilutionVolume Liters 4.40E+11 3.24E+11 4.72E+11 4.57E+11- 1.69E+12 During Period - , I denotes less than Minimum Detectable Activity (MDA) 41

Table 2.3-L2 Millstone Unit No. 3 Liquid Effluents - Continuous - SGBD & SW (Release Point - Quarry)

Nuclides '

Released Unitsl 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Products Gamma Emitters Ci - I - - -

Activi Ci - tTotal B. Tritium 1113 1 CiI 3.69E-02 I 6.37E-03 I 3.29E-02 I 1.02E-01 I 1.78E-01 I C. Dissolved & Entrained Gases jGamma Emiftersl Ci I - I TotalActivity I -

D. Gross Alpha IGrssAlpha I Ci na I na I na . I na I na

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

"na" denotes Not Analyzed 42

Table 2.3-L3 Millstone Unit No. 3 Liquid Effluents - Batch - LWS (Release Point - Qua/y)

. Nuclides I/11 1Releas~ed Initsl 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total A. Fission & Activation Products Ag-110m Ci 4.81E-05 8.37E-06 1.44E-05 1.15E-05 8.24E-05 Co-57 Ci 1.05E-05 - - 1.05E-05 Co-58 Ci 1.21 E-04 1.56E-03 1.54E-03 4.60E-04 3.68E-03 Co-60 Ci 6.01E-03 1.83E-03 1.31 E-03 6.99E-04 9.85E-03 Cr-51 Ci 1.10E-04 - 1.1OE-04 Cs-134 Ci - 1.63E-04 7.16E-04 8.79E-04 Cs-1 37 Ci 2.43E-05 9.60E-05 5.29E-04 6.49E-04 Fe-55 Ci 5.91 E-03 3.50E-03 8.40E-03 1.69E-03 1.95E-02 Fe-59 Ci - - 1.64E-05 - 1.64E-05 1-131 Ci 1.20E-03 - - 1.20E-03 Mn-54 Ci 6.03E-04 2.24E-04 2.26E-04 7.80E-05 1.13E-03 Nb-95 Ci 3.89E-05 2.12E-04 3.58E-05 2.87E-04 Sb-122 Ci 2.78E-05 -. 2.78E-05 Sb-124 Ci 2.98E-04 8.53E-06 - 3.07E-04 Sb-125 Ci 3.52E-03 6.OOE-03 1.95E-03 1.13E-03 1.26E-02 Sn-113 Ci - 6.58E-06 - 6.58E-06 Sn-117m Ci 8.75E-06 - 8.75E-06 Te-132 Ci - 7.31E-06 - 7.31E-06 Sr-89 Ci 1.07E-04 - 1.07E-04 Sr-90 .Ci - - -

Zr-95 Ci 7.14E-06 2.86E-05 - 3.57E-05 Total Activity Ci 1.62E-02 1.48E-02 1.41 E-02 5.35E-03 5.05E-02 B. Tritium H-3 ) Ci [ 7.70E+02 , 1.72E+02 5.64E+01 I 3.74E+01 [ 1.04E+03 i C. Dissolved & Entrained Gases Xe-131m Ci 1.89E-04 - 1.89E-04 Xe-133 Ci 1.36E-05 1.16E-02 4.47E-05 1.17E-02 Xe-133m Ci 1.82E-04 - 1.82E-04 Xe-135 Ci 1.32E-05 - 2.78E-05 3.87E-05 7.97E-05 Total Activity Ci 2.68E-05 1.20E-02 7.25E-05 3.87E-05 1.21 E-02 D. Gross Alpha, IGrossAlpha Ci I , I -"

"-" denotes less than Minimum Detectable Activity (MDA) 43

Table 2.3-L4 Millstone Unit No. 3 Liquid Effluents - Batch - CPF Waste Neutralization Sumps, Hotwell, S/G Bulk (Release Point - Quarry)

I Nuclides I ll Released I UnitsI 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total.

A. Fission & Activation Products IGammaEmitteral .. Ci I I -

TotalActivity I Ci I - I -"

B. Tritium IH-3 Ci 1.11E-02 I 4.74E-04 9.80E-03 1.75E-02 3.89E'02 C. Dissolved & Entrained Gases IGamma EmitteraI Ci I I - - - - I Total Activity I Ci I " I "

D. Gross Alpha IGrossAlpha I Ci I na I na I na I na I na

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

"na" denotes Not Analyzed 44

Table 2.3-L5 Millstone Unit No. 3 Liquid Effluents - Release Summary (Release Point - Yard Drain - DSN 006)

I Units P 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total C. Dissolved and Entrained Gases D. Gross Alpha

1. Total Activity Ci na na na na na Released E. Volume

1. Released Waste Liters 4.62E+06 3.86E+06 Volume

3. Dilution Volume Liters 3.15E+07 3.11 E+07 During Period

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

"na" denotes Not Analyzed 45

Table 2.3-L6 Millstone Unit No. 3 Liquid Effluents - Continuous - Turbine Building Sump (Release Point - Yard Drain - DSN 006)

Nuclides 1i1 Released I Units 1 1st Qtr I 2nd Qtr 3rd Qtr I 4th Qtr I Total A. Fission &Activation Products IGamma EmittersI Ci I Total Activity CI -

B. Tritium IH-3 ICi 3.60E-01 ] 8.68E-03 6.06E-02 9.29E-02 5.22E-01 C. Dissolved & Entrained Gases GammalEmittersl Ci I Total Activity Ci - - t D. Gross Alpha IGross Alpha I Ci I na na I na I na I na

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

"na" denotes Not Analyzed 46

Table 2.3-L7 Millstone Unit No. 3 Liquid Effluents - Continuous - WTT Berm Water (Release Point - Yard Drain - DSN 006)

INuclidesI 'I' Released Units 1st Qtr I 2nd Qtr 3rd Qtr 4th Qtr Total A. Fission &Activation Products IGamma Emittersl Ci I " - - - I Total Activity I Ci I " -

B. Tritium H-3 Ci 9.52E-04 2.60E-03 1.87E-04 7.68E-05 3.82E-03 C. Dissolved &Entrained Gases IGamma Emittersl Ci I - - - II Total Activity I Ci I I D. Gross Alpha IGross Alpha Ci na na na na na

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

"na" denotes Not Analyzed 47

Table 2.3-L8 Millstone Unit No. 3 Liquid Effluents - Continuous - Foundation Drain Sumps (Release Point - Yard Drain - DSN 006)

Abnormal Releases Nuclides Released IUnitsl 1st Qtr I 2nd Qtr I 3rd Qtr I 4th Qtr I Total I A. Fission & Activation Products lGammaEmittersl Ci I -

ITotal Activit I Ci I ---

B. Tritium H-3 Ci I 3.OE-03 3.OE-03 1.OE-02 1.5E-02 3.1E-02 C. Dissolved & Entrained Gases lGamma Emitters I- I -

Total Activity ICi D. Gross Alpha Gross Al ha Ci na I na I na I na I na

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

"na" denotes Not Analyzed 48

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 ft 3 Steel Boxes 45 ft3 3

87 ft 95 ft33 122 ft Steel Container 202.1 ft 3 Steel "Sea Van" 1280 ft 3 Polyethylene High Integrity Containers 120.3 ft3 132.4 ft33 173.4 ft 202.1 ft3 49

Table 2.1-S Solid Waste and Irradiated Component Shipments Millstone Unit 1 January 1, 2007 through December 31, 2007 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 r .. Units Annual Totals J Es t. T o ta l Error From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Resin r "3 2.3789E+00 Processing. Ci 3.1372E+00 25%

b. Dry compressible waste, Contaminated equipment, etc.

Disposition I

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Release, Super-Compaction, Incineration, Re-pack/Burial, etc..

c. Irradiated components, Control rods, etc.

Disposition Units Annual Totals E Error

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

F_ Disposition I Units I Annual Totals Est. Total Error

d. Other - (Mixed Waste)

I Disposition I Units I Annual Totals Est. Total Error 50

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 Ridae. TN for Resin Processina.

Radionuclide  % of Total Curies H-3 0.31 9.6596E-03 C-14 0.02 7.1166E-04 Cr-51 Mn-54 Fe-55 75.34 2.3635E+00 Fe-59 Co-57 Co-58 Co-60 0.14 4.4735E-03 Ni-63 19.20 6.0246E-01 Zn-65 Sr-89 Sr-90 0.03 9.8389E-04 Nb-94 Zr-95 Nb-95 Tc-99 Ru-103 Ag-1 10m Sn-113 Sn-1i17m Sb-1 24 Sb-125 1-129 Ba-1 31 1-131 Cs-1 34 Cs-137 3.37 1.0580E-01 Ba-140 Ce-144 Hf-1 81 Pu-238 0.06 1.9925E-03 Pu-239 0.03 1.0161E-03 Pu-241 1.35 4.2501E-02 Arm-241 0.09 2.7150E-03 Cm-242 Cm-243 Crm-244 0.05 1.4340E-03 CURIES (TOTAL) 3.1372E+00 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 Ridae. TN for Release. Super-Compaction.IIncineration.I Re-oack/Burial. Ietc..

Radionuclide  % of Total Curies H-3 C -14 . . .. .. . .... .

Cr-51 Mn-54 ....

Fe-55 53.82 1.4395E-02 Fe-59 "_"

Co-57 Co-58 Co-60 12.13 3.2436E-03 Ni-63 3.18 8.5102E-04 Zn-65 Sr-89 __"

Sr-90 Nb-94 Zr-95 Nb-95 .

Tc-99 Ru-1 03 Ag-1 1Om Sn-113 Sn-i 17m Sb-124 Sb-125 1-129 Ba-131 1-131 Cs-1 34 Cs-1 37 30.79 8.2368E-03 Ba-140 ....

Ce-144 Hf-1 81 Pu-238 0.01 3.2942E-06 Pu-239 <0.01 1.8665E-06 Pu-241 Am-241 0.04 1.0980E-05 Cm-242 Cm-243 Cm-244 0.02 5.4915E-06 CURIES (TOTAL) 2.6748E-02 52

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

d. Other - (Water)

From Millstone Nuclear Power Station to Duratek Inc. Oak Ridoe TN for Incineration Radionuclide  % of Total Curies H-3 95.44 1.0285E-02 C-14 <0.01 1.0120E-06 Cr-51 0.06 6.4973E-06 Mn-54 0.13 1.3480E-05 Fe-55 1.81 1.9523E-04 Fe-59 <0.01 7.1536E-07 Co-57 <0.01 3.5495E-07 Co-58 0.83 8.9595E-05 Co-60 0.54 5.8721 E-05 Ni-63 0.77 8.3444E-05 Zn-65 0.05 5.1864E-06 Sr-89 Sr-90 <0.01 6.2796E-08 Nb-94 Zr-95 <0.01 9.0428E-07 Nb-95 <0.01 8.5705E-07 Tc-99 Ru-1 03 Ag-110m <0.01 2.9588E-07 Sn-113 Sn-1 17m Sb-124 <0.01 6.4099E-07 Sb-125 <0.01 3.3987E-07 1-129 Ba-1 31 1-131 <0.01 2.1264E-07 Cs-1 34 0.04 4.4824E-06 Cs-1 37 0.27 2.8569E-05 Ba-140 Ce-144 <0.01 7.9802E-08 Hf-1 81 Pu-238 <0.01 4.4721 E-09 Pu-239 <0.01 9.1849E-11 Pu-241 <0.01 2.7129E-07 Am-241 <0.01 6.0201E-09 Cm-242 <0.01 1.7346E-11 Cm-243 Cm-244 <0.01 2.1798E-10 CURIES (TOTAL) 1.0776E-02 53

3. Solid Waste Disposition (Shipments from Millstone)

IINumber of Shipments* Mode of Transportation I Destination Truck Sole Use Vehicle) lChem-Nuclear Systems, LLC, Barnwell, SC Truck (Sole Use Vehicle) IDiversified Scientific Services, Inc., Kingston, TN Truck (Sole Use Vehicle) lDuratek, Inc., Kingston, TN 6 Truck (Sole Use Vehicle) IDuratek, Inc., Oak Ridge, TN Truck Sole Use Vehicle) IStudsvik Processing Facility, Erwin, TN

• Indicates the number of shipments in this category which contained any unit-i 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 2007.

B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments I Mode of Transportation I Destination No Shipments in 2007 r N/A N/A 54

Table 2.2-S Solid Waste and Irradiated Component Shipments Millstone Unit 2 January 1,2007 through December 31, 2007 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

1. Type of Waste

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

Units I r Annual Totals rEst. Total EsError Disposition Error 3

From Millstone Nuclear Power Station to Chem-Nuclear Systems, LLC, Barnwell, m 8.2447E+00 SC for Burial. Ci 6.7261E+01 2

b. Dry compressible waste, Contaminated equipment, etc.

Disposition Units Annual Totals Est. Total Duratek, Inc., Oak Ridge, TN for I 2.0330E+02 From Millstone Nuclear Power Station to 25%

Release, Super-Compaction, Incineration, Re-pack/Burial, etc.. Ci 1.9162E+00

c. Irradiated components, Control rods, etc.

Disposition I UnitsI II Annual Totals Est. Total I Error

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

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

d. Other - (Mixed. Waste)

Disposition From Millstone Nuclear Power Station to Diversified Scientific Services, Inc.,

Kingston, TN for Incineration, Fuel Blending, etc..

d. Other- (Water) 55

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

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

From Millstone Nuclear Power Station to Chem-Nuclear Systems, LLC, Barnwell, SC for Burial.

Radionuclide  % of Total Curies H-3 0.65 4.3546E-01 C-14 0.45 3.0369E-01 Cr-51 <0.01 2.5947E-05 Mn-54 2.80 1.8833E+00 Fe-55 23.31 1.5679E+01 Fe-59 <0.01 3.9430E-05 Co-57 0.08 5.5631E-02 Co-58 0.43 2.9238E-01 C0-60 7.77 5.2288E+00 Ni-63 21.20 1.4260E+01 Zn-65 <0.01 8.8874E-04 Sr-89 0.03 1.8138E-02 Sr-90 0.11 7.4397E-02 Nb-94 <0.01 2.7231E-04 Zr-95 0.03 2.2678E-02 Nb-95 <0.01 4.4121E-03 Tc-99 Ru-1 03 <0.01 4.4535E-06 AI-110m <0.01 3.0420E-03 Sn-113 0.06 4.3304E-02 Sn-1 17m Sb-124 <0.01 7.0059E-05 Sb-1 25 0.45 3.0530E-01 1-129 <0.01 3.0239E-04 Ba-131 1-131 Cs-134 19.03 1.2799E+01 Cs-137 23.40 1.5742E+01 Ba-140 <0.01 9.8205E-14 Ce-144 0.04 2.4151E-02 Hf-181 <0.01 2.2118E-07 Pu-238 <0.01 2.0975E-03 Pu-239 <0.01 1.1459E-03 Pu-241 0.11 7.7334E-02 Am-241 <0.01 1.2430E-03 Cm-242 <0.01 2.7032E-04 Cm-243 Cm-244 <0.01 2.5317E-03 CURIES (TOTAL) 6.7261E+01 56

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 Release, Super-Compaction, Incineration, Re-pack/Burial, etc..

Radionuclide  % of Total Curies H-3 0.03 5.0339E-04 C-14 <0.01 6.3155E-05 Cr-51 0.02 4.2518E-04 Mn-54 0.26 4.9938E-03 Fe-55 45.23 8.6662E-01 Fe-59 0.02 3.0039E-04 Co-57 0.01 2.3266E-04 Co-58 1.25 2.3956E-02 Co-60 11.03 2.1145E-01 Ni-63 22.47 4.3054E-01 Zn-65 <0.01 6.4809E-07 Sr-89 <0.01 1.0763E-04 Sr-90 <0.01 4.7565E-05 Nb-94 _

Zr-95 0.50 9.5933E-03 Nb-95 0.27 5.1974E-03 Tc-99 <0.01 9.2817E-06 Ru-1 03 <0.01 6.6240E-05 AI-!7 10m <0.01 5.7640E-05 Sn-113 0.09 1.7978E-03 Sn-117m <0.01 6.6023E-07 Sb-124 <0.01 1.6320E-06 Sb-1 25 0.43 8.2093E-03 1-129 Ba-131 <0.01 5.8647E-07 1-131 Cs-134 6.29 1.2060E-01 Cs-137 11.89 2.2775E-01 Ba-140 _

Ce-144 0.04 7.8020E-04 Hf-181 Pu-238 <0.01 1.6678E-04 Pu-239 <0.01 1.3068E-04 Pu-241 0.11 2.1308E-03 Am-241 <0.01 1.7966E-04 Cm-242 0.01 2.5032E-04 Cm-243 Cm-244 <0.01 6.6518E-05 CURIES (TOTAL) 1.9162E+00 57

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, Re-packing for Burial, etc..

Radionuclide  % of Total Curies H-3 1.96 5.4621 E-06 C-14 Cr-51 Mn-54 13.03 3.6279E-05 Fe-55 24.41 6.7981E-05 Fe-59 0.06 1.7437E-07

• Co-57 0.17 4.6386E-07 Co-58 14.71 4.0973E-05 Co-60 5.25 1.4623E-05 Ni-_63 6.83 1.9018E-05 Zn-65 Sr-89 Sr-90 Nb-94 Zr-95 Nb-95 Tc-99 Ru-103 Ag-1 1Om Sn-113 0.16 4.3391E-07 Sn-1 17m Sb-124 Sb-1 25 0.82 2.2962E-06 1-129 Ba-131 1-131 Cs-134 15.08 4.1979E-05 Cs-137 17.50 4.8720E-05 Ba-140 Ce-144 Hf-1 81 Pu-238 <0.01 11.5953E-08 Pu-239 <0.01 1.4085E-08 Pu-241 <0.01 7.7188E-09 Am-241 <0.01 1.8209E-08 Cm-242 Cm-243 Cm-244 CURIES (TOTAL) 2.7846E-04 58

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 Incineration, Fuel Blending, etc..

Radionuclide  % of Total Curies H-3 C-14 . . .. .. .. ..

Cr-51 Mn-54 Fe-55 8.29 5.8271E-08 Fe-59 .... ..

Co-57 Co-58 Co-60 2.44 1.7162E-08 Ni-63 4.09 2.8713E-08 Zn-65 Sr-89 Sr-90 Nb-94 Zr-95 Nb-95 Tc-99 Ru-1 03 Ag-1i1m Sn- 113 Sn-1 17m Sb-124 Sb-125 1-129 Ba-1 31 1-131 Cs-134 25.48 1.7903E-07 Cs-137 59.68 4.1934E-07 Ba-140 Ce-144 Hf-181 Pu-238 <0.01 4.2113E-11 Pu-239 <0.01 3.7475E-1i1 Pu-241 Am-241 <0.01 4.8053E-1i1 Cm-242 Cm-243 Cm-244 CURIES (TOTAL) 7.0265E-07 59

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 94.55 2.0686E-02 C-14 0.03 6.6105E-06 Cr-51 0.05 1.0558E-05 Mn-54 0.19 4.2549E-05 Fe-55 1.76 3.8583E-04 Fe-59 <0.01 1.1625E-06 Co-57 <0.01 6.2316E-07 Co-58 0.76 1.6655E-04 Co-60 0.95 2.0869E-04 Ni-63 0.70 1.5255E-04 Zn-65 0.06 1.3889E-05 Sr-89 <0.01 8.3759E-08 Sr-90 <0.01 5.2709E-07 Nb-94 Zr-95 <0.01 1.4695E-06 Nb-95 <0.01 1.3927E-06 Tc-99 <0.01 1.8135E-07 Ru-1 03 Ag-,11m <0.01 4.8080E-07 Sn-113.

Sn-117m Sb-124 <0.01 1.0416E-06 Sb-125 <0.01 2.0392E-06 1-129 <0.01 1.5406E-09 Ba-1 31 1-131 <0.01 3.4554E-07 Cs-134 0.09 1.9000E-05 Cs-137 0.80 1.7594E-04 Ba-140 Ce-144 <0.01 5.0902E-07 Hf-181 Pu-238 <0.01 1.3514E-08 Pu-239 <0.01 3.8446E-09 Pu-241 <0.01 5.8973E-07 Am-241 <0.01 1.5620E-08 Cm-242 <0.01 4.6192E-10 Cm-243 <0.01 9.7220E-10 Cm-244 <0.01 1.3079E-09 CURIES (TOTAL) 2.1879E-02 60

3. Solid Waste Disposition (Shipments from Millstone)

Nubr fShipments* -Mode of Transportation TDestination 3 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) Duratek, Ind.ý Kingston, TN 14 Truck (Sole Use Vehicle) IDuratek, Inc., Oak Rid1e, TN Truck (Sole Use Vehicle) IStudsvik Processing Facilit, 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 2007.

B. IRRADIATED FUEL SHIPMENTS (Disposition)

II Number of Shipments I Mode of Transportation I Destination Ii No Shipments in 2007 N/A N/A 61

Table 2.3-S Solid Waste and Irradiated Component Shipments Millstone Unit 3 January 1, 2007 through December 31, 2007 A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel)

b. Dry compressible waste, Contaminated equipment, etc.

Disposition I

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for Release, Super-Compaction, Incineration, Re-pack/Burial, etc..

I

c. Irradiated components, Control rods, etc.

Disposition Units Annual Totals Est. Total1 T Error

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

Disposition I

From Millstone Nuclear Power Station to Duratek, Inc., Oak Ridge, TN for I Compaction, Incineration, Re-packing for Burial, etc..

d. Other - (Mixed Waste)

Disposition I Units I Annual Totals IEst. Total III I Error_

62

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

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

From Millstone Nuclear Power Station to Chem-Nuclear Systems, LLC, Barnwell, SC for Burial.

Radionuclide  % of Total Curies H-3 1.58 1.0152E-01 C-14 0.08 51317E-03 Cr-51 <0.01 2.5179E-11 Mn-54 0.20 1.2687E-02 Fe-55 52.83 3.3980E+00 Fe-59 <0.01 1.1431E-09 Co-57 <0.01 2.7763E-04 Co-58 <0.01 1.1446E-05 Co-60 15.47 9.9516E-01 Ni-63 28.10 1.8071 E+00 Zn-65 <0.01 2.4780E-05 Sr-89 <0.01 9.9209E-1 1 Sr-90 <0.01 2.4086E-04 Nb-94 0.01 8.1694E-04 Zr-95 <0.01 6.1311E-07 Nb-95 <0.01 5.8751E-10 Tc-99 Ru-103 <0.01 4.1720E-11 A-11Orn <0.01 6.9359E-05 Sn-113 <0.01 3.9404E-06 Sn-1 17m _

Sb-1 24 <0.01 4.8164E-09 Sb-125 0.27 1.7177E-02 1-129 <0.01 8.0165E-08 Ba-131 1-131 Cs-134 0.29 1.8644E-02 Cs-137 1.08 6.9564E-02 Ba-140 <0.01 1.6181 E-28 Ce-144 0.01 8.7113E-04 Hf-181 <0.01 8.6218E-12 Pu-238 <0.01 9.1993E-05 Pu-239 <0.01 4.6384E-05 Pu-241 0.06 3.6773E-03 Am-241 <0.01 8.6293E-05 Cm-242 <0.01 7.9842E-07 Cm-243 Cm-244 <0.01 1.5917E-04 CURIES (TOTAL) 6.4313E+00 63

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 Resin Processing.

Radionuclide  % of Total Curies H-3 0.48 6.8747E-02 C-14 Cr-51 Mn-54 1.95 2.7948E-01 Fe-55 65.66 9.4263E+00 Fe-59 Co-57 0.04 5.4731E-03 Co-58 0.23 3.3110E-02 Co-60 18.24 2.6190E+00 Ni-63 12.92 1.8556E+00 Zn-65 Sr-89 Sr-90 Nb-94 Zr-95 <0.01 7.0734E-04 Nb-95 0.02 3.3646E-03 Tc-99 Ru-1 03 Ag-1 1Oin Sn-I13 Sn-1 17m Sb-124 Sb-125 0.37 5.3662E-02 1-129 _

Ba-131 1-131 Cs-134 Cs-1 37 0.07 9.4159E-03 Ba-140 Ce-144 <0.01 1.6116E-05 Hf-1 81 Pu-238 <0.01 5.7332E-05 Pu-239 <0.01 1.7019E-05 Pu-241 0.01 1.9924E-03 Am-241 <0.01 2.9830E-05 Cm-242 <0.01 1.8560E-05 Cm-243 Cm-244 <0.01 1.2143E-04 CURIES (TOTAL) 1.4357E+01 64

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..

Radionuclide  % of Total Curies H-3 0.02 2.1346E-02 C -14 ..... . . .

Cr-51 Mn-54 9.74 9.3697E+00 Fe-55 32.44 3.1199E+01 Fe-59 Co-57 0.29 2.7820E-01 Co-58 9.26 8.9043E+00 Co-60 11.10 1.0678E+01 Ni-63 36.21 3.4830E+01 Zn-65 Sr-89 <0.01 1.2806E-03 Sr-90 <0.01 3.5070E-03 Nb-94 Zr-95 Nb-95 Tc-99 Ru-1 03 Ag-1 1Oin Sn-113 Sn-1 17m Sb-124 Sb-1 25 0.88 8.4242E-01 1-129 Ba-1 31 1-131 Cs-1 34 Cs-1 37 0.06 5.3843E-02 Ba-140 Ce-144 Hf-181 Pu-238 Pu-239 Pu-241 Am-241 Cm-242 Cm-243 Cm-244 CURIES (TOTAL) 9.6182E+01 65

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 Release, Super-Compaction, Incineration, Re-pack/Burial, etc..

Radionuclide  % of Total [ Curies H-3 0.09 3.0742E-03 C-14 Cr-51 0.18 5.8562E-03 Mn-54 2.68 8.7342E-02 Fe-55 61.61 2.0056E+00 Fe-59 Co-57 <0.01 5.2478E-06 Co-58 4.20 1.3678E-01 Co-60 11.62 3.7819E-01 Ni-63 17.43 5.6756E-01 Zn-65 Sr-89 Sr-90 Nb-94 Zr-95 0.07 2.3462E-03 Nb-95 0.50 1.6209E-02 Tc-99 Ru-1 03 Ag-111Om, Sn-113 Sn-1 17m Sb-124 Sb-1 25 0.75 2.4359E-02 1-129 Ba-1 31 1-131 Cs-134 <0.01 1.2534E-04 Cs-1 37 0.86 2.8076E-02 Ba-140 Ce-144 Hf-181 Pu-238 <0.01 2.2839E-07 Pu-239 <0.01 1.8848E-07 Pu-241 <0.01 7.0379E-07 Am-241 <0.01 2.4261E-07 Cm-242 <0.01 1.0150E-08 Cm-243 Cm-244 <0.01 5.4645E-08 CURIES (TOTAL) 3.2555E+00 66

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, Re-packing for Burial, etc..

Radionuclide  % of Total Curies H-3 1.13 2.3632E-06 C-14 Cr-51 Mn-54 16.72 3.5075E-05 Fe-55 28.72 6.0245E-05 Fe-59 0.08 1.7437E-07 Co-57 0.22 4.6386E-07 Co-58 19.46 4.0823E-05 Co-60 6.18 1.2956E-05 Ni-63 8.06 1.6901 E-05 Zn-65 . ...

Sr-89 Sr-90 Nb-94 Zr-95 Nb-95 Tc-99 Ru-1 03 Ag-1 1Om Sn-113 0.21 4.3391E-07 Sn-1 17m Sb-124 Sb-125 0.36 7.6490E-07 1-129 Ba-1 31 1-131 Cs-134 8.60 1.8048E-05 Cs-137 10.24 2.1470E-05 Ba-140 Ce-144 Hf-1 81 Pu-238 <0.01 1.3939E-08 Pu-239 <0.01 1.2291E-08 Pu-241 <0.01 7.7188E-09 Am-241 <0.01 1.5911E-08 Cm-242 Cm-243 Cm-244 CURIES (TOTAL) 2.0977E-04 67

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

d. Other - (Water)

From Millstone Nuclear Power Station to 1liratek Inc l*ak Ridne TN for Incineratinn Radionuclide * % of Total Curies H-3 95.22 1.8151E-01 C-14 0.01 2.0512E-05 Cr-51 0.06 1.1149E-04 Mn-54 0.15 2.8574E-04 Fe-55 1.85 3.5205E-03 Fe-59 <0.01 1.2276E-05 Co-57 <0.01 6.2886E-06 Co-58 0.84 1 .6050E-03 Co-60 0.66 1.2664E-03 Ni-63 0.77 1.4741 E-03 Zn-65 0.05 9.6921E-05 Sr-89 <0.01 3.5708E-07 Sr-90 <0.01 1.7712E-06 Nb-94 Zr-95 <0.01 1.5517E-05 Nb-95 <0.01 1.4707E-05 Tc-99 <0701 7.7314E-07 Ru-1 03 Ag-110m <0.01 5.0772E-06 Sn-113 Sn-1 17m Sb-124 <0.01 1.0999E-05 Sb-125 <0.01 6.1177E-06 1-129 <0.01 6.5677E-09 Ba-1 31 1-131 <0.01 3.6489E-06 Cs-134 0.05 8.7819E-05 Cs-1 37 0.29 5.5494E-04 Ba-140 Ce-144 <0.01 1.5652E-06 Hf-181 Pu-238 <0.01 9.4693E-08 Pu-239 <0.01 1.2292E-08 Pu-241 <0.01 5.2229E-06 Am-241 <0.01 1.1779E-07 Cm-242 <0.01 1.8378E-09 Cm-243 <0.01 4.1447E-09 Cm-244 <0.01 3.9236E-09 CURIES (TOTAL) 1.9062E-01 68

3. Solid Waste Disposition (Shipments from Millstone)

El I_Nube oShpet IMod of TrnprainIetnto I Number of 1Ship~ments*ý Mode (Sole Truck of Tranpotaio Use Vehicle) IIChem-Nuclear Systems, Destination LLC, Barnwell, SC I

Truck (Sole Use Vehicle) IDiversified Scientific Services, Inc., Kingston, TN Truck (Sole Use Vehicle) IDuratek, InC., Kingston, TN 17 Truck (Sole Use Vehicle) IDuratek, Inc., Oak Ridge, TN 2 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.) 28 physical shipments were made from this station in 2007.

B. IRRADIATED FUEL SHIPMENTS (Disposition)

I Number of Shipments I Mode of TransportationI Destination No Shipments in 2007 r N/A I N/A 69

Table 2.4-GW2 Groundwater Monitoring Results Foundation Drain Sumo 2 12/03/2007Drain 12/03/2007 rnn

< LLD J na na na na n

na 12/04/2007 < LLD na na na na na 12/05/2007 < LLD na na na na na 12/07/2007 < LLD < LLD na na na na Foundation Drain Sump 3 01/02/2007 2,440 < LLD na na na na 06/06/2007 1,070 < LLD na na na na 06/26/2007 4,590 < LLD na na na na 08/16/2007 13,900 < LLD na na na na 10/05/2007 10,000 < LLD <LLD < LLD < LLD na 11/19//07 25,800 < LLD na na na na 11/27/2007 29,400 na na na na na 11/28/2007 34,000 na na na na na 11/29/2007 31,100 na na na na na 11/30/2007 29,000 < LLD <LLD < LLD <LLD < LLD 12/01/2007 28,200 na na na na na 12/02/2007 28,700 na na na na na 12/03/2007 25,400 na na na na na 12/04/2007 27,800 na na na na na 12/05/2007 29,800 na na na na na 12/06/2007 25,900 na na na na na 12/07/2007 30,800 na na na na na 12/08/2007 23,400 na na na na na 12/09/2007 31,700 na na na na na 12/10/2007 26,800 na na na na na 12/11/2007 28,900 na na na na na 12/12/2007 29,800 na na na na na 12/14/2007 21,400 na na na na na 12/15/2007 27,300 na na na na na 12/16/2007 27,300 na na na na na 12/17/2007 19,700 na na na na na 12/18/2007 20,100 na na na na na 12/19/2007 22,300 na na na na na 12/20/2007 22,400 na na na na na 12/21/2007 23,000 na na na na na 12/22/2007 21,300 na na na na na 12/23/2007 16,800 na na na na na 12/24/2007 13,700 na na na na na 12/25/2007 18,500 na na na na na 12/26/2007 12,400 na na na na na 12/27/2007 13,000 na na na na na 12/28/2007 16,200 na na na na na 12/29/2007 14,100 na na na na na 12/30/2007 14,100 na na na na na 12/31/2007 12,800 na na na na na Source for Foundation Drain Sump 3: RWST Pit 12/19/2007 4.28E+06 6.4 LLD <LLD 392 <LLD

• Transuranics: Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Cm-242, Cm-243, Cm-244 "na"denotes not analyzed 71

3.0 Inoperable Effluent Monitors During the period January 1 through December 31, 2007, the following effluent monitors were inoperable for more than 30 consecutive days:

3.1 Unit I - None 3.2 Unit 2 - None 3.3 Unit 3 - None 72

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 99.3% and the power history for 2007 is shown below.

MP2 - CYCLE 18 POWER HISTORY YEAR 2007 110 100*

90 80 70, s0 2,

50 40 30 20 10 0

i i; PS ý5 io o o o

• 73

Unit 3 operated with a DER capacity factor of 85.9% and the power history for 2007 is shown below.

MP3 - CYCLE 11 & 12 POWER HISTORY YEAR 2007 Note: Data at 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> intervals 110 100 90 80 70 60 50 40 30 20T 10 Roo~ IRSoIR 3aagg; $ g s;g; 5;5; 5k

- F1 0

Ae ' a 74

5.0 Errata 5.1 none 75

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 2007 is provided to the Nuclear Regulatory Commission as Volume 2 of the Radioactive Effluent Release Report.

In 2007, there were no changes made to the Millstone REMODCM:

Rev Effective Date None 76

6.1 REMODCM Description of Changes none 77

Serial No. 08-0187 Docket Nos. 50-245 50-336 50-423 License Nos. DPR-21 DPR-65 NPF-49 ATTACHMENT 2 2007 RADIOACTIVE EFFLUENT RELEASE REPORT VOLUME 2 MILLSTONE POWER STATION UNITS 1, 2, AND 3 DOMINION NUCLEAR CONNECTICUT, INC. (DNC)

Millstone Power Station 2007 Radioactive Effluents Release Report Volume 2 Dominion Nuclear Connecticut, Inc.

MILLSTONE UNIT LICENSE DOCKET I DPR-21 50-245 2 DPR-65 50-336 5%Dominioiw 3 NPF-49 50-423

Table of Contents Volume 1I 2007 Radiological Effluent Monitoring and Off-Site Dose Calculation Manual (REMODCM), MP-22-REC-BAPO1 Revision 25 - 02

Introduction This report, for the period of January through December of 2007, is being submitted by Dominion Nuclear Connecticut, Inc. for Millstone Power Station's Units 1, 2, and 3, in accordance with IOCFR50.36a, the REMODCM, and the Station's Technical Specifications. A combined report, written in the US NRC Regulatory Guide 1.21 format, is submitted for all three units.

Volume 1 contains radiological and volumetric information on airborne and liquid effluents, shipments of solid waste & irradiated components, calculated offsite radiological doses, all changes to the REMODCM, information on effluent monitors inoperable for more than 30 consecutive days, and corrections to previous reports.

Per Unit 1 Technical Specification 5.6.1, Unit 2 Technical Specification 6.15, and Unit 3 Technical Specification 6.13, a legible copy of the entire REMODCM, effective during the calendar year, shall be submitted as a part of or concurrent with the Radioactive Effluents Release Report.

Volume 2 contains a copy of the REMODCM effective during the calendar year.

No changes were made to the REMODCM for the calendar year 2007.

MILLSTONE POWER STATION STATION PROCEDURE AA 7NZ

/ K Radiological 1 Effluent Monitoring\and Off- Site Dose Calculation Manual (REMODCM)

MP REC - BAPOI Rev. 025-02 Approval Date: 11/28/06 Effective Date: 12/04/06

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

TABLE OF CONTENTS I. Radiological Effluent Monitoring Manual (REMM) ............................ 7 I.A . Introduction ........................................................ 7 I.B . 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 U se ............................................... ........... 19 I.D . G aseous Effluents ................................................... 23

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

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

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

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

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

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

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

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

2. Radioactive Effluent Release Report .............................. 50 II Off-Site Dose Calculation Manual (CDCM) ................................. 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 AC --- MP-22-REC-BAP01

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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 .. 76

6. Compliance with 40CFR190 ..................................... 77

7. Bases for Gaseous Pathway Dose Calculations ...................... 77 II.E. Liquid Discharge Flow Rates And Monitor Setpoints ..................... 79

1. Unit 1 Reactor Cavity Water Discharge Line ........................ 79

2. Reserved ..................................................... 80

3. Unit 2 Clean Liquid Radwaste Effluent Line - RM9049 and Aerated Liquid Radwaste Effluent Line - RM9116 ......................... 80

4. Condensate Polishing Facility Waste Neutralization Sump Effluent Line - CN D 245 ............................................... 82

5. Unit 2 Steam Generator Blowdown - RM4262 and Unit 2 Steam Generator Blowdown Effluent Concentration Limitation ...................... 83

6. Unit 2 Condenser Air Ejector - RM5099 .......................... 84

7. Unit 2 Reactor Building Closed Cooling Water RM6038 and Unit 2 Service Water, and RBCCW Sump and Turbine Building Sump Effluent Concentration Limitation ................................ 84

8. Unit 3 Liquid Waste Monitor - LWS-RE70 ....................... 86

9. Unit 3 Regenerant Evaporator Effluent Line - LWC-RE65 ......... 88

10. Unit 3 Waste Neutralization Sump Effluent Line - CND-RE07 ...... 88

11. Unit 3 Steam Generator Blowdown - SSR-RE08 and Unit 3 Steam Generator Blowdown Effluent Concentration Limitation ............. 88

12. Unit 3 Turbine Building Floor Drains Effluent Line - DAS-RE50 and Unit 3 Service Water and Turbine Building Sump Effluent Concentration Limitation ....................................... 90

13. Bases for Liquid Monitor Setpoints ............................... 90 II.F Gaseous Monitor Setpoints .......................................... .91

1. Unit 1 Spent Fuel Pool Island Monitor - RM-SFPI-02 ............. 91

2. Unit 2 Wide Range Gas Monitor (WRGM) - RM8169 .............. 91

3. R eserved ..................................................... 91

4. Unit 3 SLCRS - HVR-RE19B .................................. 92

5. Unit 2 Vent - Noble Gas Monitor - RM8132B ..................... 92

6. Unit 2 Waste Gas Decay Tank Monitor RM9095 .................... 92

7. Unit 3 Vent Noble Gas Monitor - HVR-RE10B ................... 93

8. Unit 3 Engineering Safeguards Building Monitor - HVQ-RE49 ...... 93

9. Bases for Gaseous Monitor Setpoints.............................. 93

III REMODCM Unit One Controls ............................................ 98 III.A . Introduction ....................................................... 97 III.B. Definitions and Surveillance Requirement (SR) Applicability ............. 97 III.C. Radioactive Effluent Monitoring Instrumentation ...................... 100

1. Radioactive Liquid Effluent Monitoring Instrumentation ............ 100

2. Radioactive Gaseous Effluent Monitoring Instrumentation .......... 103 III.D. Radioactive Effluents Concentrations And Dose Limitations .............. 106

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

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

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

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

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

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

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

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

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

2. Radioactive Gaseous Effluents .................................. 154 VE. Total Radiological Dose From Station Operations ....................... 158 V E B ases ............................................................ 158 MP-22-REC-BAP01 QST~rOnP ';TH 1ACT REVIEW Rev. 025-02

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

Table III.C.-1, "Radioactive Liquid Effluent Monitoring Instrumentation" ... 101 Table III.C.-2, "Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements" . .................................... 102 Table III.C.-3, "Radioactive Gaseous Effluent Monitoring Instrumentation" . 104 Table III.C.-4, "Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements" . .................................... 105 Table IV.C.-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. -1, "Radioactive Liquid Effluent Monitoring Instrumentation" .... 143 Table V.C.-2, "Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements" . .................................... 146 Table VC.-3, "Radioactive Gaseous Effluent Monitoring Instrumentation" .. 148 Table V.C.-4, "Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements" ........................................... 150

~A, _M-22-REC-BAiP01 STP , *HiNK' AT "EVIEW R Rev. 025-02

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FIGURES Figure I.C.- 1, "Reserved" . ............................................ 20 Figure I.C.-2, "Simplified Liquid Effluent Flow Diagram Millstone Unit 2 .... 21 Figure I.C.-3, "Simplified Liquid Effluent Flow Diagram Millstone Unit 3 .... 22 Figure I.D. -1, "Simplified Gaseous Effluent Flow Diagram Millstone U nit O ne" . ..... .............................................. 34 Figure I.D.-2, "Simplified Gaseous Effluent Flow Diagram Millstone U nit Tw o" . ...... .............................................. 35 Figure I.D.-3, "Simplified Gaseous Effluent Flow Diagram Millstone U nit Three" . ..... ............................................. 36 Figure I.E.- 1, "Inner Air Particulate And Vegetation Monitoring Stations" ... 42 Figure I.E.-2, "Outer Terrestrial Monitoring Stations ...................... 43 Figure III.D.- 1, "Site Boundary for Liquid and Gaseous Effluents" ......... 111 Figure IVD.- 1, "Site Boundary for Liquid and Gaseous Effluents"........... 134 Figure V.D.- 1, "Site Boundary for Liquid and Gaseous Effluents" . ......... 157 STP NTI1, ACTF; I-5 5of 163

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 A-op ACT EW Rev. 025-02 MP-22-REC-BAP01 6 of 163

SECTION I. RADIOLOGICAL EFFLUENT MONITORING MANUAL (REMM)

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

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

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

MP-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 REC- PRG, "Radiological Effluent Program."

I.C. Liquid Effluents

1. Liquid Effluent Sampling and Analysis Program Radioactive liquid wastes shall be sampled and analyzed in accordance with the program specified in Table I.C.-1 for Millstone Unit No. 1, Table I.C.-2 for Millstone Unit No. 2, and Table I.C.-3 for Millstone Unit No. 3. The results of the radioactive analyses shall be input to the methodology of Section II to assure that the concentrations at the point of release are j__ j/ _MP-22-REC-BAP01 NSTT N*A" T\ REVIEWRev. 025-02

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

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

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Table I.C.- 1 Millstone Unit 1 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 (p.Ci/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 I Kr-85 I x 10-5 Prior to initial batch re- H-3 1 x 10-5 lease from any one source and monthly composite thereafterC Grab sample prior to Prior to initial batch re- Gross alpha 1 x 10-7 initial batch release lease from any one from any one source source and quarterly Sr-90 5 x 10-8 and quarterly compos- thereafter ite thereafter Fe-55 Ix10 6 Table I.C.-1 TABLE NOTATIONS A. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 Sb (E)(V)(2.22x1O 6 )(ye- At)

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) 0 2.22 x 106 is the number of transformations per minute per [tCi 0 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

• *7"*

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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.

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Table I.C.-2 Millstone Unit 2 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 (gtCi/ml)

A.Batch ReleaseB 1.Clean Waste Moni- Grab sample prior to Prior to each batch Principal Gamma' 5 x 10-7 tor Tank, Aerated each batch release release EmittersC.

Waste Monitor Tank and Steam 1-131 1xx10-6 Generator BulkD- Ce-144 5 x 10-6 Dissolved & 1 x 10-5 Entrained GasesK-2.Condensate Monthly H-3 1 x 10-5 Polishing Facility CompositeF,G

- Waste Quarterly Gross alpha 1 x 10-7 SumpEl CompositeF"G" Sr-89, Sr-90 5 x 10-8 Fe-55 1 x 10-6 B.Continuous Release 1.Steam Generator Daily Grab Sample.& Weekly Principal Gamma 5 x 10-7 BlowdownH- prior to aligning to CompositeE,G. EmittersC-2.Service Water Long Island Sound for 1-131 1x 10-6 Effluenter RBCCW sump CI_-_144 Ce-144 x 10-66 55 x10-3.Turbine SumpsL. Monthly Grab Monthly Dissolved & 1 x 10-5 Sample Entrained GasesK.

5 4.RBCCW SumpM. Weekly Grab or Com- Monthly H-3N- 1 x 10-posite CompositeF',G.

Weekly Composite Quarterly Gross alpha 1 x 10- 7 CompositeFG. Sr-89, Sr-90 5 x 10-8 Fe-55 1x10-6 TABLE I.C.-2 TABLE NOTATIONS A. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD =4.66 Sb (E)(V)(2.22x1I06)( Ye -,W t)

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Where:

• LLD is the lower limit of detection as defined above (as RCi 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) 0 2.22 x 106 is the number of transformations per minute per 0Ci 0 Y is the fractional radiochemical yield (when applicable) 0 Xis the radioactive decay constant for the particular radionuclide

• At is the elapsed time between midpoint of sample collection and midpoint of counting time It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

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

B. A batch release is the discharge of liquid wastes of a discrete volume from the tanks listed in this table. Prior to the sampling, each batch shall be isolated and at least two tank/sump volumes shall be recirculated or equivalent mixing provided. 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-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 ttCi/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 [Ci/ml, THEN the sampling and analysis schedule for gross alpha, Sr-89, Sr-90, Fe-55 are not required.

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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 O tCi/ml, THEN the sampling and analysis schedule for all principal gamma, 1-131, Ce-144, noble gases, gross alpha, Sr-89, Sr-90 and Fe-55 are not required.

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

J. For the Service Water:

IF a weekly gamma analysis does not indicate a gamma activity greater than 5 xl0- 7 [0Ci/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 pCi/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 0Ci/ml and for gross alpha 1 x 10-8 [Ci/ml.

M. For the RBCCW Sump:

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

IF the applicable batch gamma activity is not greater than 5 x 10-7 [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.

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

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

Tank, Boron Test Tanks and Steam Quarterly Gross alpha 1 x i0.

Generator Bulk D. CompositeE,G. Sr-89, Sr-90 5 x 10-8 Fe-55 1 x 10-6 B.Continuous Release 1.Steam Generator Daily Grab Samplel Weekly Principal Gamma 5 x 10-7 BlowdownH. CompositeEG. EmittersC.

2.Service Water Ef- 1-131 1x 10-6 fluentJ.

Ce-144 5 x 10- 6 5

3.Turbine Building Monthly Grab Monthly Dissolved & 1 x 10-SumpsL. Sample Entrained GasesK-5 Weekly Grab or Com- Monthly H-3M. 1 x 10-posite CompositeF.,G.

Weekly Composite Quarterly Gross alpha 1 x 10-7 CompositeF.,G.

Sr-89, Sr-90 5 x 10-8 6

Fe-55 1 x 10-TABLE I.C.-3 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.

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

LL 4.66 Sb, 6 )(ye -At)

LLD =(E)(V)(2.22x10 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)

- 0 2.22 x 106 is the number of transformations per minute per 0Ci

" Y is the fractional radiochemical yield (when applicable)

• X is the radioactive decay constant for the particular radionuclide

• At is the elapsed time between midpoint of sample collection and midpoint of counting time It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

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

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

C. The LLD will be 5 x 10- 7 0Ci/ml. The principal gamma emitters for which this LLD applies are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 10-6

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

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

D. For the Steam Generator Bulk:

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

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

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

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

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

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

H. For the Steam Generator Blowdown:

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

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

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

J. For the Service Water:

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

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

L. For the Turbine Building Sump:

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

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

IF the release pathway is directed to yard drains, THEN the LLD for 1-131 shall be 1.5 x 10-7 0Ci/ml and for gross alpha 1 x 10-8 0Ci/ml.

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

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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 2.c.

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

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

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

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

Boron recovery Cesium ion exchanger (DEMN A or B)

Boron evaporator (EV- 1)

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

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

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

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

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

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

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

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Figure I.C. -1, "Reserved

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Figure I.C.-2, "Simplified Liquid Effluent Flow Diagram Millstone Unit 2" Aerated Liquid Rad neBldg. Floor Drains* . 'i [(RM9116)

• Turbine Bldg floor drains normally to storn drains.

Drains may be diverted to the radwastc treatment system.

[] = Component included in radwastc trcatmcnt equipment requirement STOP I4 NK7 AC ,,,,-AMP-22-REC-l3AP01 REVI EW Rev. 025 -02

ýCT ýýA21 of 163

Figure I.C.-3, "Simplified Liquid Effluent Flow Diagram Millstone Unit 3" SIc- Gcncmtor BIowdmn (OpenCycle)

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I.D. Gaseous Effluents

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

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

Table I.D.- 1 Millstone Unit 1 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Release Sample Type and Minimum 'Type of Activity Lower Limit Point or Source Frequency Analysis Analysis of Detection Frequency (LLD)A (pCi/ml)

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

(with half lives greater than 8 days)

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

(with half lives greater than 8 days)

Twice per month Sr-90, Gross alpha 1 x 10-11 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.

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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 ýiCi/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 (0Ci/ml)

A.Batch Release Principal Particulate 1 x 10-4 Each Tank Discharge Gamma EmittersB 1.Waste Gas Stor- Gaseous Grab Prior to each age TankH Waste Gas Tank Discharge H-3 1 x 10-6 B.Containment&Aux Building Releases 1.Containment Gaseous Grab of purges 1. Prior to purge Principal Gamma 1 x 10-1 Purge and vents 2. Weekly for venting EmittersB 2.Containment 1. Prior to Each Purges & prior to venting-Venting 2. Every two weeks for samples taken due to VentingI Footnote I 3.Open Equip- Monthly H-3 1x 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 - 1x 10-4 ment Hatch & Aux Bldg Gross Activity Rollup DoorL C.Continuous Release 1.Vent Monthly - Gaseous Grab Monthlyu, t- Principal Gamma 1 x 10-4 (RM8132B) SampleCK. EmittersB H-3j 1 x 10-6 2.Millstone Continuous Charcoal Sam- Weekly 1-131 1 x 10-12 Stack pleDF" 1-133 1 x 10- 10 (RM8169-1) Continuous Particulate Weekly Principal Particu- I x 10-11 SampleDF" late Gamma Emit-tersB - (I-131, oth-ers with half lives greater than 8 days)

Continuous Particulate Quarterly Composite Sr-89, Sr 1 x 10-11 SampleD. Gross alpha 1 x 10-11 Continuous Noble GasD. Continuous Monitor Noble Gases - 1 x 10-6 Gross Activity TABLE I.D.-2 TABLE NOTATIONS A. The lower limit of detection (LLD) is defined in Table Notations, Item a, of Tables I.C.-1, I.C.-2, or I.C.-3.

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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 0Ci/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 I) 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: I(

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.

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

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

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

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

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

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

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

A. Containment and Fuel Building Release 1.Containment Gaseous, particulate and Prior to each purge or Principal gamma 1X 10-4 Hogger Draw- charcoal grab prior to drawdown; weekly for emittersB-down each drawdown venting & prior to 2.Containment venting samples taken Purge due to Footnote I.

3.Containment Gaseous grab prior to Prior To Each Draw- 1-131 1 x 10-12 Vent each purgeH. down 1-133 1 x 10-10 4.Open Equin Gaseous Grab every two weeks for ventingI' Principal particulate gamma emittersB" -

1 x 10-11 IC Hatch During (I- 131, others with half lives Outages greater than 8 days) 5.Fuel Building Monthly for all re- H-3 1 x 10-6 lease sources except Equipment Hatch Continuous particulate Weekly Particulate gamma emit- NA at open equipment hatch ters for 1/2: hour count (1-131, others with half-life greater than 8 days)

Continuous charcoal at Weekly 1-131 and 1-133 for one NA equipment hatch and hour count fuel building rollup doorsK.

Gaseous grab at equip- Daily Noble Gases - Gross 1 x 10-4 ment hatch & fuel build- Activity ing rollup doorsK-B.Continuous Release 1.Unit 3 Ventila- Monthly - Gaseous MonthlyC',J" Principal particulate 1 x 10-4 tion Vent Grab SampleC.J. gamma emittersB (HVR-RE10B) H-3-' 1 x0-6 2.Engineered Continuous charcoal Weekly 1-131 1 x 10-11 Safeguards sampleD-,E 1-133 1 x 10-1 0 Building Continuous particulate Weekly Principal particulate 1 x 10-11 (HVQ-RE49) sampleD.,F, gamma emittersB" -

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

SLCRS Continuous particulate Quarterly composite Sr-89, Sr-90 1 x 10-11 (HVR- sampleD. Gross alpha l x 10-11 RE19B) Continuous noble gasD. Continuous Noble gases - gross 1 x 10-6 monitor activity TABLE I.D.-3 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.

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B. For gaseous samples, the LLD will be 1 x 10-4 [iCi/cc and for particulate samples, the LLD will be 1 x 10-11 [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.

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

IF, containment noble gas activity exceeds 1 E-6 pCi/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.

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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.

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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, THEN doses due to gaseous effluents from the untreated waste stream to unrestricted areas shall be projected at least once per 31 days in accordance with the methodology and parameters in Section II.D.4. For each waste stream, only those doses specified in Section II.D.4. need to be determined for compliance with this section.

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

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

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

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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 Building Ventilation Fuel building ventilation 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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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., IVD.2.a. and V.D.2.a.

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

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

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

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Figure I.D.- 1, "Simplified Gaseous Effluent Flow Diagram Millstone Unit One"

.4 .....................

= generalized air flow Reactor Balance of Plant Vent Building Turbine Building Radwaste Building

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Figure I.D.-2, "Simplified Gaseous Effluent Flow Diagram Millstone Unit Two" LHEPA F _____ _ ' j Vent (L26) [Rad Monitor (RM8132A1BV Turbine Bldg

) Turbine Bldg Roof Vent R 3MP-22-REC-BAPOI 7

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

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I.E. Radiological Environmental Monitoring

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

Program changes may be made based on operational experience.

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

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

If milk samples are temporarily unavailable from any one or more of the milk sample locations required by Table I.E.-2, a grass sample shall be substituted during the growing season (Apr. - Dec.) and analyzed for gamma isotopes and I- 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 the end of the affected calendar quarter, 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.l.b.,

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

for Unit 2; and Sections V.D.1.b., VD.2.b., or VD.2.c. for Unit 3). This report is not required if the measured level of radioactivity was not the result of plant effluents, however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

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

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Table I.E.- 1 Millstone Radiological Environmental Monitoring Program Exposure Pathway and/or No. of Sampling and Collection Type and Frequency of Sample Locations Frequency Analysis 1.Gamma Dose - 4 0(a) Quarterly Gamma Dose - Quarterly Environmental TLD 2.Airborne Particulate 8 Continuous sampler - Gross Beta - Weekly Gamma Spec-weekly filter change trum - Quarterly on composite (by location), and on individual sample if gross beta is greater than 10x the mean of the weekly control station's gross beta results 3.Airborne Iodine 8 Continuous sampler - 1-131 - Weekly weekly canister change 4.Vegetation 5 One sample near middle and Gamma Isotopic on each sample one near end of growing season 5.Milk 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 0 monthly collection at indica- sample tor location. Quarterly at control location - Compos-ite of 6 weekly grab samples 7.Well Water 2 Semiannual Gamma Isotopic & Tritium on each sample I

8.Bottom Sediment 5 Semiannual Gamma Isotopic on each sample 9.Soil 3 Annually Gamma Isotopic on each sample I

10.Fin Fish-Flounder and 2 Quarterly Gamma Isotopic on each sample one other type of edible fin fish (edible portion) 11.Mussels (edible portion) 2 Quarterly Gamma Isotopic on each sample 12.Oysters (edible portion) 4 Quarterly Gamma Isotopic on each sample 13.Clams (edible portion) 2 Quarterly Gamma Isotopic on each sample 14.Lobsters (edible portion) 2 Quarterly Gamma Isotopic on each sample (a) Two or more TLDs or TLD with two or more elements per location.

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

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

1-I Onsite - Old Millstone Road 0.6 Mi, NNW TLD, Air Particulate, Iodine, Vegetation 2-I Onsite - Weather Shack 0.3 Mi, S TLD, Air Particulate, Iodine 3-I Onsite - Bird Sanctuary 0.3 Mi, NE TLD, Air Particulate, Iodine, Soil 4-1 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-I Onsite - Environmental Lab Dock 0.3 Mi, SE TLD 8-1 Onsite - Environmental Lab 0.3 Mi, SE TLD 9-1 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 27-1 Niantic 1.7 Mi, WNW TLD, Air Particulate, Iodine 28-I Two Tree Island 0.8 Mi, SSE Mussels, Fish1 29-I West Jordan Cove 0.4 Mi, NNE Clams, Fish1 30-1 Niantic Shoals 1.5 Mi, NNW Mussels 31-I Niantic Shoals 1.8 Mi, NW Bottom Sediment, Oysters 32-I Vicinity of Discharge 2 Bottom Sediment, Oysters, Lobster, Fish 1 ,

Seawater 33-I Seaside Point 1.8 Mi, ESE Bottom Sediment 34-I Thames River Yacht Club 4.0 Mi, ENE Bottom Sediment 35-1 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 38-1 Waterford Shellfish Bed No. 1 1.0 Mi, NW Clams 41-I Myrock Avenue 3.2 Mi, ENE TLD 42-1 Billow Road 2.4 Mi, WSW TLD REC-BAP01

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-I Onsite - Schoolhouse 0.1 Mi, NNE TLD 45-I Onsite Access Road 0.5 Mi, NNW TLD 46-1 Old Lyme - Hillcrest Ave. 4.6 Mi, WSW TLD 47-1 East Lyme - W, Main St. 4.5 Mi, W TLD 48-I East Lyme - Corey Rd. 3.4 Mi, WNW TLD 49-1 East Lyme - Society Rd. 3.6 Mi, NW TLD 50-I East Lyme - Manwaring Rd. 2.1 Mi, W TLD 51-I East Lyme - Smith Ave. 1.5 Mi, NW TLD 52-1 Waterford - River Rd. 1.1 Mi, NNW TLD 53-I Waterford - Gardiners Wood Rd. 1.4 Mi, NNE TLD 55-I Waterford - Magonk Point 1.8 Mi, ESE TLD 56-1 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 63-1 Waterford - Jordon Cove Rd. 0.8 Mi, NE TLD 64-I Waterford - Shore Rd. 1.1 Mi, ENE TLD 65-I Waterford - Bank St. 3.2 Mi, NE TLD 71-I Onsite well Onsite Well water 72-1i Onsite well Onsite Well water 1

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" 1 6Teresi .Ia  :

M.

fwtw flTLD andAir Moniltoring (particulates & iodline? 4ff.-Ia-T~errestrial Monitoring (milk, grass, fruit &vegetables, soil, or leaves) 77 u Mti.oitanng sediment, flora, fish,

)ra~y mussel, oyster, clarn or lobster)

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Figure I.E.-2, "Outer Terrestrial Monitoring Stations" ntnn TLDM ( £" , ...

TLD and-Aronitoni n (particulates &iodine)

Terrestl ial n torimIk-grass .frit6&ý soil or leýves)

-betables Aquatic Monitoning (seaater;sedimAetf.; R' flora.. fish

-isy mu sseloyster clam

's. ,os. r cl or lb am,: or lobster)

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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 Shellfish (pCi/g, wet) (pCi/I) (pCi/g, wet) or Gases (pCi/g, wet)

(pCi/m 3 )

H-3 20,000A Mn-54 1,000 30 140 Fe-59 400 10 60 Co-58 1,000 30 130 Co-60 300 10 50 Zn-65 300 20 80 Zr-95 400 Nb-95 400 Ag-r110m 8 30 1-131 2 0 B. 0.9 0.2 1 3 0.1 Cs- 134 30 10 1 5 60 1 Cs- 137 50 20 2 8 70 2 Ba-140 200 300 La- 140 200 300 A. 20,000 pCi/I for drinking water samples. (This is 40 CFR Part 141 value.) For non-drinking water pathways (i.e., seawater), a value of 30,000 pCi/I may be used.

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

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Table I.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 15 0.130 Zn-65 30 0.260 Zr-95 30 Nb-95 15 2 1 0.06B.

1-131 1 5 C" 7 x 10-Cs- 134 15 5 x 10-2 0.130 15 0.060 0.150 Cs-137 18 6 x 10- 2 0.150 18 0.080 0.180 Ba- 140 6 0 C" 70 La- 140 15Cl 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)(V)(2.22)(Ye-* 9)

Where:

• LLD is the lower limit of detection as defined above (as [tCi per unit mass or volume) 0 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 0 Y is the fractional radiochemical yield (when applicable) 0MP-22-REC-BAP1 STOP, "'ýTHINKF ACT 7REEV\EW Rev. 025-02 45 of 163

0 ) is the radioactive decay constant for the particular radionuclide

• At is the elapsed time between midpoint of sample collection and midpoint of counting time (or end of the sample collection period) and time of counting.

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

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

B. LLD for leafy vegetables.

C. From end of sample period.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

The NUREGs specify an LLD of 15 pCi/1 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/1 for the parent Ba-140.

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

I.E Report Content

1. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The report shall also include the results of the land use census required by Section I.E.2. of this manual. If levels of radioactivity are detected that result in calculated doses greater than 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 include a summary table of all radiological environmental samples which shall include the following information for each pathway sampled and each type of analysis:

1) Total number of analyses performed at indicator locations.

2) Total number of analyses performed at control locations.

3) Lower limit of detection (LLD).

4) Mean and range of all indicator locations together.

5) Mean and range of all control locations together.

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

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

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

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

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

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

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

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

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

" solidification agent (e.g., cement)

• total curies

• total volume and typical container volumes

• principal radionuclides (those greater than 10% of total activity) types of containers used (e.g., LSA, Type A, etc.)

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

Refer To MP 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 Changesto the MP-22-REC-BAPO0, "Radiological Effluent Monitoring And Offsite Dose Calculation Manual (REMODCM)," shall be submitted to the NRC as appropriate, as a part of or concurrent with the RERR for the period in which the changes were made.

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SECTION II.

Offsite Dose Calculation Manual (ODCM)

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

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

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., IV.D.2.a., and VD.2.a.

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

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

Most of the calculations in this manual have several methods given for the calculation of the same parameter. These methods are arranged in order of simplicity and conservatism, 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 completely 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-22-REC-PRG, "Radiological Effluent Control."

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

1. Whole Body Dose from Liquid Effluents Radiological Effluent Controls (Sections III, IV, and V) limit the whole body dose to an individual member of the public to 1.5 mrem per calendar quarter and 3 mrem per year from liquid effluents released from each unit. (See Appendix A, Table App.A- 1 for cross-reference effluent control C requirements and applicable sections in the REMODCM which are used to determine compliance). In addition, installed portions of liquid radwaste treatment system are required to be operated to reduce radioactive materials in liquid effluents when the projected whole body dose over 31 days from applicable waste streams exceeds 0.006 mrem. This part of the REMODCM provides the calculation methodology for determining the whole body dose from radioactive materials released into liquid pathways of exposure associated with routine discharges. This includes the liquid pathways which contribute to the 25 mrem annual total dose limit (40 CFR190) to any real individual member of the public from all effluent sources (liquids, gases, and direct).

a. Method 1 (Applicable to Units 1, 2, and 3)

For Unit 1: No Method 1, use Method 2 (Section II.C.l.b.)

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For Units 2 and 3:

Dw = 0.2 CF + 5.6 x 10-7 CH Where:

Dw= The estimated whole body dose to a potentially maximum exposed individual (in mrem) due to fission and activation products released in liquid effluents during a specified time period.

CF = total gross curies of fission and activation products, excluding tritium and dissolved noble gases, released during the period of interest.

CH = total curies of tritium released during the period of interest.

If Dw, within a calendar quarter is greater than 0.5 mrem, go to Method 2.

b. Method 2 (Applicable to Units 1, 2, and 3)

If the calculated dose using Method 1 is greater than 0.5 mrem within a calendar quarter, or if a more accurate determination is desired, use the NRC computer code LADTAP II, or a code which uses the methodology given in Regulatory Guide 1.109, to calculate the liquid whole body doses. Method 2 (LADTAP II) is also used in the performance of dose calculations for the Radioactive Effluent Release Report. The use of this code is given in MP-22-REC-GDLO2, "Liquid Dose Calculations

- LADTAP-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 1II, IV, and V) limit the maximum organ dose to an individual member of the public to 5 mrem per calendar quarter and 10 mrem per year from liquid effluents released from each unit.

(See Appendix A, Table App.A- 1 for cross-reference effluent control I 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 k~' , `TPT'THINVK / ACT; 'VIEw Rev 025-02

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liquid pathways which contribute to the 25 mrem annual organ (except 75 mrem thyroid) dose limit (40 CFR190) to any real individual member of the public from all effluent sources (liquids, gases, and direct).

a. Method 1 (Applicable to Units 1, 2, and 3)

For Unit 1: No Method 1, use Method 2 (Section II.C.2.b.)

For Units 2 and 3:

Do= 1.5 CF Where: Do =The estimated maximum organ dose to the potentially maximum exposed individual (in mrem) due to fission and activation products released in liquid effluents during a specified time period.

CF =total gross curies of fission and activation products, excluding tritium and dissolved noble gases, released during the period of interest - same asSection II.C.l.a.

If Do, within a calendar quarter is greater than 2 mrem, go to Method 2.

b. Method 2 (Applicable to Units 1, 2, and 3)

If the calculated dose using Method 1 is greater than 2 mrem, or if a more accurate determination is desired, use the NRC computer code LADTAP II, or a code which uses the methodology given in Regulatory Guide 1.109, to calculate the liquid maximum organ doses. Method 2 (LADTAP II) is also used in the performance of dose calculations for the Radioactive Effluent Release Report. The use of this code and the input parameters are given in MP-22-REC-GDLO2, "Liquid Dose Calculations - LADTAP-H." Additional information on LADTAPII is contained in MP-22-REC-REF03, "REMODCM Technical Information Manual."

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

An estimation of annual (year-to-date) whole body dose (Dyw) from liquid effluents shall be made every month to determine compliance with the annual dose limits for each Unit which releases any radioactivity in liquid effluents. Annual doses will be determined as follows:

Dyw = ZDw Where the sum of the doses include the whole body dose contribution from all effluent releases for each Unit recorded to-date. For estimation of the S J 56o1MP-22-REC-BAP01

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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:

Do= ZDo 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 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.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 I Only)

In the dose code DOSLIQ use concentrations of radionuclides in reactor cavity water and estimates of projected volumes and discharge rates for the following 31 days to estimate dose from liquid discharge of reactor cavity water in the following 31 days.

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

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

DEMw = D'Mw [R 1 R 4 F 2]

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

DEMo = D-MO [R 1 R 4 F2]

Where:

D'MW =the whole body dose from the last typical* previously completed month as calculated per the methods in Section II.C.1.

D'MO = the maximum organ dose from the last typical* previously completed month as calculated per the methods in Section II.C.2.

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R, = the ratio of the total estimated volume of liquid batches to be released in the present month to the volume released in the past month.

R4 = the ratio of estimated primary coolant activity for the present month to that for the past month.

F2 = the factor to be applied to the estimated ratio of final curies released if there are expected differences in treatment of liquid waste for the present month as opposed to the past month (e.g., bypass of filters or demineralizers).

NUREG-0017 or past experience shall be used to determine the effect of each form of treatment which will vary. F2 = 1 if there are no expected differences.

Notes:

1) The last month should be typical without significant operational differences from the projected month. If there were no releases during last month, do not 'use that month as the base month if it is estimated that 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 R5 where R 5 =

total dilution flow from LADTAP run divided by estimated total dilution flow.

c. Whole Body and Maximum Organ when Steam Generator Total Gamma Activity Exceeds 5E-7 [tCi/ml or Steam Generator Tritium Exceeds 0.02

ýtCi/ml (Applicable to Units 2 and 3)

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

DEMw = D'MW [(1 - F1 ) R, R4 F 2 + F1 R 2 R 3]

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

DEMo = D'MO [(I - F1 ) R, R 4 F 2 + F 1 R2 R 3]

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Where:

D'Mw= the whole body dose from the last typical previously IC completed month as calculated per the methods in Section II.C.1.

D'MO= the maximum organ dose from the last typical previously le completed month as calculated per the methods in Section II.C.2.

R1 = 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. F2 = 1 if there are no expected differences.

6. Quarterly Dose Calculations for Radioactive Effluent Release Report Detailed quarterly dose calculations required for the Radioactive Effluent Release Report shall be done using the NRC computer code LADTAP II, or a code which uses the methodology given in Regulatory Guide 1.109. The use of this code, and the input parameters are given in MP-22-REC-GDLO2, "Liquid Dose Calculations - LADTAP 1." IG Additional information on LADTAPII is contained in MP REC - REF03, "REMODCM Technical Information Manual."

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7. Bases for Liquid Pathway Dose Calculations The dose calculation methodology and parameters used in Section II of the REMODCM implement the requirements in Section III.A of Appendix I (10CFR50) which states that conformance with the dose objectives of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The dose estimations calculated by both Method 1 and Method 2 are based on the liquid models presented in Regulatory Guide 1.109, Rev.1; "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I". These equations are implemented via the use of the NRC sponsored computer code LADTAP II. Input parameter values typically used in the dose models are listed in MP-22-REC-REF03, "REMODCM Technical Information Document." This same methodology is used in the determination of compliance with the 40CFR190 total dose standard for the liquid pathways.

The conversion constants in the Method 1 equations are based on the maximum observed comparison of historical effluent releases for each unit and corresponding whole body or critical organ doses to a maximum individual. The dose conversion factors are calculated based on the ratio of the observed highest dose (whole body and organ) and the curies of fission and activation products released during the period. This ratio results in the Method 1 equation conversion factor in mrem/Ci released. This same approach was repeated separately for tritium (as a different radionuclide class) discharged in liquids wastes. MP-22-REC-REF03 describes the derivation of the Method 1 constants and list the historical whole body and maximum organ doses calculated for each unit operation.

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II.D. Gaseous Dose Calculations The determination of potential release rates and doses from radioactive gaseous effluents to the maximum off-site receptor are divided into two methods. Method 1 provides simplified operational tools to ensure that effluent releases are not likely to cause quarterly and annual off-site dose or dose rate limits to be exceeded. Effluent doses are calculated at least once every 31 days. Method 2 provides for a more detailed computational calculation using accepted computer models to demonstrate actual regulatory compliance. Method 2 is used whenever the Method 1 estimation approaches a regulatory limit, and for preparation of the Radioactive Effluent Release Report which includes the quarterly and annual dose impacts for all effluents recorded discharged to the atmosphere during the year of record.

1. Site Release Rate Limits ("Instantaneous")

Radiological Effluent Controls (Sections III, IV, and V) for each unit require that the instantaneous off-site dose ratesfrom 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:

Q1v /90,000 + Q2S/ 5 6 0 ,00 0 + Q2V/ 2 9 0 ,0 00 + Q3s/ 5 6 0 ,0 0 0 + Q3v/ 2 9 0,000 < 1 Where:

Qv = Noble gas release rate from Spent Fuel Pool Island Vent

([tCi/sec)

Q2S = Noble gas release rate from MP2 to Millstone Stack (ItCi/sec)

Q2v = Noble gas release rate from MP2 Vent ([tCi/sec)

Q3V = Noble gas release rate from MP3 Vent ([tCi/sec)

Q3S = Noble gas release rate from MP3 to Millstone Stack ([tCi/sec)

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As long as the above is less than or equal to 1, the doses will be less than or equal to 500 mrem to the total body and less than 3000 mrem to the skin. The limiting factor for the Unit 1 SFPI vent of 90,000 is based on the skin dose limit of 3,000 mrem/year, while all the other factors are based on the whole body dose limit of 500 mrem/year.

b. Method 1 Release Rate Limit 131, 1-133, H-3 and Particulates Half Lives Greater Than 8 Days With releases satisfying the following limit conditions, the dose rate to the maximum organ will be less than 1500 mrem/year from the inhalation pathway:

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

DRthyl + DRthy2 + DRthy3 < 1 Where the contribution from each Unit is calculated from:

Unit 1: DRthyl = 9.36 x 10-6 QH1V Unit 2: DRthy2 = 5.1 X10-2 131Q12V + 2.38 x 10- 3 131Q12S +

1.25 x 10-2 133Qi2V+5. 7 5 x 10- 4 133Q12S +

4.2 x 10-6 QH2V+ 1. 9 X 10- 7 QH2S Unit 3: DRthy3 = 5.1 x 10-2 131Q13V + 2.38 x 10- 3 131Q13s +

1.25 x 10-2 133Q13V +5.75 x 10-4 133QI3s +

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

2) The site release rate limit of particulates with half-lives greater than 8 days and tritium (where the critical organ is a composite of target organs for a mix of radionuclides) shall be:

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

Unit l:DRorga = 1.05 x 10-1 [QP1V + QPTB ]+9.36 x 10-6 QH1V Unit 2:DRorg2 = 2.38 x 10-3 QP2S + 5.1 x 10-2 QP2V + 1.9 X 10-7 QH2S + 4.2 x 10-6 QH2V Unit 3:DRorg3 2.38 x 10-3 QP3S + 5.1 x 10-2 QP3V +

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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 (tCi/sec)*

131Q12S = Release rate of 1- 131 from MP2 to Millstone Stack (jiCi/sec) 133Q12v = Release rate of 1-133 from MP2 Vent (RCi/sec)*

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

([tCi/sec) 131Q13V = Release rate of 1-131 from MP3 Vents (Normal and ESF)

(RiCi/sec)*

131Q13S = Release rate of 1- 131 from MP3 to Millstone Stack

([tCi/sec) 133Q13V = Release rate of 1- 133 from MP3 Vents (Normal and ESF)

([tCi/sec)*

133Q13S = Release rate of 1-133 from MP3 to Millstone Stack (ltCi/sec)

QH1V = Release rate of tritium from the Spent Fuel Pool Island and Balance of Plant Vents ([tCi/sec)

QH2V = Release rate of tritium from MP2 Vent ([tCi/sec)*

QH2S = Release rate of tritium from MP2 to Millstone Stack (itCi/sec)

QH3V = Release rate of tritium from MP3 Vents (Normal and ESF)

([tCi/sec)

• QH3S = Release rate of tritium from MP3 to Millstone Stack (IACi/sec)

Qpjv = Release rate of total particulates with half-lives greater than 8 days from the Spent Fuel Pool Island Vent ([tCi/sec)

QPIB = Release rate of total particulates with half-lives greater than 8 days from the Balance of Plant Vent ([tCi/sec)

QP2V = Release rate of total particulates with half-lives greater than 8 days from the MP2 Vent (*ICi/sec)

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

QP3V - Release rate of total particulates with half-lives greater than 8 days from MP3 Vents (Normal and ESF) (itCi/sec)

QP3S = Release rate of total particulates with half-lives greater than 8 days from MP3 to Millstone Stack ([tCi/sec)

• includes releases via the steam generator blowdown tank vent.

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c. Method 2 The above Method 1 equations assume a conservative nuclide mix. If necessary, utilize the GASPAR, or a code which uses the methodology given in Regulatory Guide 1.109, code to estimate the dose rate from either noble gases or iodines, tritium, and particulates with half-lives greater than 8 days. The use of the code is described in MP-22-REC-GDLO4, "Gaseous Dose Calculations - GASPARII."

Additional information on GASPAR is contained in the MP-22-REC-REF03, "REMODCM Technical Information Manual."

2. 10 CFR50 Appendix I - Noble Gas Limits Radiological Effluent Controls (Sections III, IV, and V) limit the off-site air dose from noble gases released in gaseous effluents to 5 mrad gamma and 10 mrad beta for a calendar quarter (10 and 20 mrad gamma and beta, respectively, per calendar year). Effluent dose calculations are calculated at least once every 31 days. In addition, installed portions of the gaseous radwaste treatment system are required to be operated to reduce radioactive materials in gaseous effluents when the projected doses over 31 days from the applicable waste stream exceed 0.02 mrad air gamma or 0.04 mrad air beta. (See Appendix A, Tables App.A- 1 for a cross reference of effluent I 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: DG1 = 3.3 x 10-6 CN1V*

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 DG3 are greater than 1.6 mrad or DB1, DB2, or DB3 are greater than 3.3 mrad within a calendar quarter, go to Method 2 below.

Where:

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

DB1 = the beta air dose from Unit 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).

CN1V= the total curies of noble gas released from Spent Fuel Pool Island Vent during the period of interest.

CN2V = the total curies of noble gas released from Unit 2 Vent during the period of interest. Include containment releases to Unit 2 Vent CN2S = the total curies of noble gas released from Unit 2 to Millstone Stack during the period of interest.

CN3V = the total curies of noble gas released from Unit 3 vents during the period of interest. Include containment releases to Unit 3 Vent and ESF Building Vent.

CN3S = the total curies of noble gas released from Unit 3 to Millstone Stack during the period of interest.

• See MP-22-REC-REF03, "REMODCM Technical Information Document," Section 4.2, for the derivation of air dose Method 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 the Unit 2 vent eight-year history for 1980 to 1987:

3 x/Q = 8.1 x 10-6 sec/m D/Q = 1.5 x 10- 7 m-2 (See the MP-22-REC-REF03, (Att) "Determination of Maximum x/Q and D/Q.")

• /S*/V w-U*- "'-*-.

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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 DY6, 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 = YDG1 DYG2 = YDG2 DYG3 = YDG3 DYBI = YDB1 DYB2 = YDB2 DYB3 = YIDB3 Where the sums are over the first quarter (i.e., summation of the all release periods within the quarter) through the present calendar quarter doses.

Where: DyG1, DYG2, DYG3, DYBi, DYB2 and DYB3 = gamma air dose and beta air dose for the calendar year for Unit 1, 2, or 3.

The following shall be used as the quarterly doses:

(1) If the detailed quarterly dose calculations required per Section II.D.5. for the Radioactive Effluent Release Report are complete for any calendar quarter, use those results.

(2) If the detailed calculations are not complete for a particular quarter, use the results as determined above in Sections II.D.2.a. or II.D.2.b.

If DYG1, YG2 or YG3 are greater than 10 mrad or DyB1, YB2 or YB3 are greater than 20 mrad and any corresponding quarterly dose was not calculated using Method 2 (Section II.D.2.b.), recalculate the quarterly dose using meteorology concurrent with quarter of release.

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3. 10 CFR50 Appendix I - Iodine, Tritium and Particulate Doses Radiological Effluent Controls (Section III, IV, and V) limit the off-site dose to a critical organ from radioiodines, tritium, and particulates with half-lives greater than 8 days released in gaseous effluents to 7.5 mrem for a calendar quarter (15 mrem per calendar year). Effluent dose calculations are performed at least once every 31 days. In addition, installed portions of the gaseous radwaste treatment system are required to be operated to reduce radioactive materials in gaseous effluents when the projected doses over 31 days from the applicable waste stream exceed 0.03 mrem. (See Appendix A, Table App.A- 1 for a cross reference of effluent control IC 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 = 30131CIS + 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+ CPB]+ 1.97 x 10- 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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131CIS =The total curies of 1-131 released in gaseous effluents from Unit 2 or 3 to Millstone Stack during the period of interest.

133CIS -- The total curies of 1- 133 released in gaseous effluents from Unit 2 or 3 to Millstone Stack during the period of interest.

Cps = the total curies of particulates with half-lives greater than 8 days released in gaseous effluents from Millstone Stack during the period of interest.

Cpv = the total curies of particulates with half-lives greater than 8 days released in gaseous effluents from the SFPI vent during the period of interest. -

CPB = the total curies of particulates with half-lives greater than 8 days released in gaseous effluents from the BOP vent during the period of interest.

CHS = the total curies of tritium released in gaseous effluents from Millstone Stack during period of interest.

CHV = the total curies of tritium released in gaseous effluents from the SFPI and BOP vents during period of interest.

2) Method 2 - Millstone Stack and Unit 1 Releases C Use the GASPAR code, or a code which uses the methodology given in Regulatory Guide 1.109, with actual locations, real-time meteorology and the pathways which actually exist at the time at those locations:

Sum critical organ doses from stack with critical organ doses from vent in Section II.D.3.b. below.

b. Critical Organ Doses (Applicable to Units 2 and 3 vent releases)

1) Method 1 - Unit 2 and Unit 3 releases For Unit 2 and Unit3, separately, calculate organ doses DT and DO:

DTV = 230 131CIV + 4.0 133 CIV + 2.6 x 10-3 CHV Dov = 1.1 x 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 7

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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 thyroid dose for the period of gaseous effluents releases from the vent.

DOV= the dose to the maximum organ other than the thyroid for the period of gaseous effluent releases from the vent.

131CiV=The total curies of 1-131 in gaseous effluents from Unit 2 other than to the Millstone Stack (Unit 2 Vent, containment releases to vent, and Steam Generator Blowdown Tank Vent) or from Unit 3 other than to the Millstone Stack (Unit 3 Vent, ESF Building Vent, containment releases to vent, Steam Generator Blowdown Tank Vent, and Containment Drawdown using the hogger) 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 the hogger) 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 the hogger) during the period of interest.

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CHV = The total curies of tritium released in gaseous effluents from Unit 2 other than to the Millstone Stack (Unit 2 Vent, Steam Generator Blowdown Tank Vent and containment releases to vent) or from Unit 3 other than to the Millstone Stack (Unit 3 Vent, ESF Building Vent, Steam Generator Blowdown Tank Vent containment releases to vent, and Containment Drawdown using the hogger) during the period of interest.

2) Method 2 - Unit 2 and Unit 3 releases Use the GASPAR code, or a code which uses the methodology given in Regulatory Guide 1.109, with the actual locations, real-time meteorology and the pathways which actually exist at the time at these locations. For Unit 2, the code shall be run separately for steam generator blowdown tank vents and ventilation releases, containment purges and waste gas tank releases. For Unit 3, the code shall be run separately for ventilation, process gas, containment vacuum system, ESF ventilation and containment purges.

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c. Estimation of Annual Critical Organ Doses Due to lodines, Tritium and Particulates (Applicable to Units 1, 2, and 3)

An estimation of annual (year-to-'date) critical organ doses (DYT and DyO for thyroid and maximum organ other than thyroid, respectively) from radioiodine, tritium and particulates with half-lives greater than 8 days released from Units 1, 2 and 3 shall be made every month to determine compliance with the annual dose limits for each Unit. Annual critical organ doses will be determined as follows:

Unit 1 Unit 2 Unit 3 DyT1 = JDT1 DYG2 = JDT2 DYT3 = JDT3 Dyo1 = IDo 1 Dy02 = IDo 2 DY03 = ID03 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 Dy03 = thyroid (T) dose and maximum organ (0) dose (other than the thyroid) for the calendar year for Unit 1, 2, or 3.

The following guidelines shall be used for DT and DO:

(1) If the detailed quarterly dose calculations required per Section II.D.5. for the Radioactive Effluent Release Report are complete for any calendar quarter, use those results.

(2) If the detailed calculations are not complete for a particular quarter, use the results as determined above in Section II.D.3.a. or II.D.3.b.

(3) If DYT and/or Dy 0 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., I 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.

4. Gaseous Effluent Monthly Dose ProjectionsSection I.D.2.a. of the REMM requires that certain portions of the gaseous radwaste treatment equipment be returned to service to reduce radioactive gaseous effluents when the projected doses for each Unit (made at least once per 31 days) exceed 0.02 mrad gamma air, 0.04 mrad beta air, or 0.03 mrem to any organ from gaseous effluents. The following methods are applied in the estimation of monthly dose projections.

a. Unit 1 Projection Method None required.

b. Unit 2 Projection Method

1) Due to Gaseous Radwaste Treatment System (Unit 2)

Determine the beta and gamma monthly air dose projection from noble gases from the following:

DEMG (mrad) = 1.81X 10-4 CEN DEMB (mrad) = 1.81 x 10-6 CEN Where:

CEN = the number of curies of noble gas estimated to be released from the waste gas storage tanks during the next month.

DEMG = the estimated monthly gamma air dose.

DEMB= the estimated monthly beta air dose.

(The dose conversion factor is from MP-22-REC-REF03, "REMODCM Technical Information Document," Section 4.2, for the Millstone Stack releases since the Unit 2 waste gas tanks are discharged via the Millstone Stack. This factor is conservative because the isotopic mix assumed for the dose conversion factor consists of shorter-lived noble gases which have higher dose conversion factors than the typical mix from Unit 2 waste gas tank discharges.)

2) (Reserved)

3) Due to Ventilation Releases (Unit 2)

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If portions of the ventilation treatment system are expected to be out of service during the month, determine the monthly maximum organ dose projection (DEMO ) from the following:

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

DEMO = 1/3 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.

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c. Unit 3 Projection Method

1) Due to Radioactive Gaseous Waste System (Unit 3)

Determine the beta and gamma monthly air dose projection from noble gases from the following:

DEMG (mrad) = 1.81 x 10-4 CEN DEMB (mrad) = 1.81 x 10-6 CEN Where:

CEN = the number of curies of noble gas estimated to be released from the reactor plant gaseous vents (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.)

2) Due to Steam Generator Blowdown Tank Vent (Unit 3)

i. Method 1 Determine DEMO which is the estimated monthly dose to the maximum organ:

DEMO = 1/3 R1 X DT For the last quarter of operation, determine DT as determined per Section II.D.3.b.

Where:

R, = the expected ratio of secondary coolant iodine level for the coming month as compared with the average level during the quarter used in the determining D above.

ii. Method 2 If necessary, estimate the curies expected to be released for the next month and applicable method for dose calculation from Section II.D.3.b.

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3) Due to Ventilation Releases (Unit 3)**

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 DEMO which is the estimated monthly dose to the maximum organ.

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.

Where:

R1 = 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 estimated 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.

• • Since dose projections are only required if the treatment specified in Section I.D. of the Radiological Effluent Monitoring Manual are not operating, the monthly gamma and beta air dose projections are not required for ventilation releases.

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.

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6. Compliance with 40CFR190 The following sources shall be considered in determining the total dose to a real individual from uranium fuel cycle sources:

a. Gaseous Releases from Units 1, 2, and 3.

b. Liquid Releases from Units 1, 2, and 3.

c. Direct and Scattered Radiation from Radioactive Material on Site.

d. Since all other uranium fuel cycle sources are greater than 5 miles away, they need not be considered.

The Radiological Effluent Controls in Sections III.E. (Unit 1), IVE. (Unit 2), and VE. (Unit 3) contain specific requirements for ensuring compliance with 40CFR190 based on gaseous and liquid doses (sources a and b).

Doses to source c are controlled by design and operations to ensure the off-site dose from each radwaste storage facility is less than one mrem per year. Potential doses from each facility are evaluated in Radiological Environmental Reviews (RERs) where total off-site doses from all sources are considered to ensure compliance with 40CFR190.

7. Bases for Gaseous Pathway Dose Calculations The dose calculation methodology and parameters used in Section II of the REMODCM implement the requirements in Section III.A. of Appendix I (10CFR50) which states that conformance with the ALARA dose objectives of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated.

Operational flexibility is provided by controlling the instantaneous release rate of noble gas (as well as iodines and particulate activity) such the maximum off-site dose rates are less than the equivalent of 500 mrem/year to the whole body, 3000 mrem/year to the skin from noble gases, or 1500 mrem/year to a critical organ from the inhalation of iodines, tritium and particulates. The dose rate limits are based on the 10CFR20 annual dose limits, but applied as an instantaneous limit to assure that the actual dose over a year will be well below these numbers.

The equivalent instantaneous release rate limits for Millstone Stack were determined using the EPA AIREM code. For Units 2 & 3, these doses were calculated using the NRC GASPAR code. The AIREM code calculates cloud gamma doses using dose tables from a model that considers the finite extent of the cloud in the vertical direction. Beta doses are calculated assuming semi-infinite cloud concentrations, which are based upon a MP-22-REC-BAP01 NSTOPY AT zAHIK CK REVI EW Rev. 025 -02

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standard sector averaged diffusion equation. The GASPAR code implements the models of NRC Regulatory Guide 1.109, Rev. 1, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I." Input parameter values typically used in the dose models are listed in MP-22-REC-REF03, "REMODCM Technical Information Document." This same methodology is used in the determination of compliance with the 40CFR190 total dose standard for the gaseous pathways.

In the determination of compliance with the dose and dose rate limits, maximum individual dose calculations are performed at the nearest land site boundary with maximum decayed X/Q, and at the nearest vegetable garden (assumed to be nearest residence) and cow and goat farms with maximum D/Qs. The conversion constants in the Method 1 equations for maximum air doses, organ and whole body doses, and dose rates are based on the maximum observed comparison of historical effluent releases and corresponding calculated maximum doses. The dose conversion factors are calculated based on the ratio of the observed highest dose and the curies of fission and activation products released during the period. This ratio results in the Method 1 equation conversion factor in mrem/Ci released.

MP-22-REC-REF02 describes the derivation of the Method 1 constants and list the historical maximum doses calculated for the maximum organ.

,A-.,

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II.E. Liquid Discharge Flow Rates And Monitor Setpoints

1. Unit 1 Reactor Cavity Water Discharge Line The limit on discharge flow rate and setpoint on the Unit 1 liquid waste monitor depend on dilution water flow, radwaste discharge flow, the isotopic composition of the liquid, the background count rate of the monitor and the efficiency of the monitor. Due to the variability of these parameters, the alert and alarm setpoints will be determined prior to the release of each batch. The following method will be used:

STEP 1:

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

R = Required Reduction Factor = 1

.,offci nuclide i 10 x-EC of nucide i STEP 2:

Determine the allowable discharge flow (F)

F = 0.1 x RxD Where:

D= The existing dilution flow which is the any dilution flow from Millstone Unit 2 and/or Unit 3 not being credited for any other radioactivity discharge during discharge of Unit 1 water.

0.1 = safety factor to limit discharge concentration to 10% of the Radiological Effluent Control Limit.

STEP 3:

Calculate the monitor setpoint as follows:

Rset = 2 x AC x RCF Where:

Rset = The setpoint of the monitor.

AC = The total radwaste effluent concentration ([tCi/ml ) in the tank.

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RCF = The response correction factor for the effluent line monitor using the current calibration factor or isotopic-specific responses.

2= Tolerance limit which brings the setpoint at twice the expected response of the monitor based on sample analysis. With the safety factor of 0.1 the setpoint would be at 20% of the Radiological Effluent Control Limit.

Option setpoint:

A setpoint based upon worst case conditions may be used. Assume the maximum possible discharge flow, a minimum dilution flow not to exceed 100,000 gpm, and a limit of 1 x 10-7 [tCi/ml which is lower than any 10CFR20 EC limit except for transuranics. This will assure that low level releases are not terminated due to small fluctuations in activity. When using this option setpoint independent verification of discharge lineup shall be performed.

The optional setpoint may be adjusted (increased or decreased) by factors to account for the actual discharge flow and actual dilution flow; however, controls shall be established to ensure that the allowable discharge flow is not exceeded and the dilution flow is maintained.

2. Reserved

3. Unit 2 Clean Liquid Radwaste Effluent Line - RM9049 and Aerated Liquid Radwaste Effluent Line - RM9116 The setpoint on the Unit 2 clean and aerated liquid waste effluent lines depend on dilution water flow, radwaste discharge flow, the isotopic composition of the liquid, the background count rate of the monitor and the efficiency of the monitor. Due to the variability of these parameters, an alarm/trip setpoint will be determined prior to the release of each batch.

The following method will be used:

,z MP-22-REC-BAPO1 71SZT ýTTHI NK >AT 'EVIE-WRev. 025-02 80Oof 163

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:

Ri = Required Reduction Factor =

Z -M of nuclide i 10 x1 EC 9f nuclide i For Noble Gases: If the noble gas concentration is less than 0.011 [Ci/ml, the reduction factor need not be determined 2 x 10-41"Ci R2 = Required Reduction Factor = I ml Z

• of noble gases 2 x 10-4 L1i Z *C noble gases eml R = the smaller of R 1 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 gpm + # 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 1 below.)

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Where:

Ret= the setpoint of the monitor (cpm).

AC = the total radwaste effluent concentration ([tCi/ml) in the tank.

RF = the response factor for the effluent line monitor using the current calibration factor or isotopic-specific responses.

2= the multiple of expected count rate on the monitor based on the radioactivity concentration in the tank.

This value or that corresponding to 2.8 x 10- 5 [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 10- [tCi/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 [tCi/ml. If the grab sample is less than 5 x 10-7tCi/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 ýtCi/ml, which is approximately 38,000 CPM based on recent calibration data.

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5. Unit 2 Steam Generator Blowdown - RM4262 and Unit 2 Steam Generator Blowdown Effluent Concentration Limitation 5a. Unit 2 Steam Generator Blowdown - RM4262 Assumptions used in determining the Alarm setpoint for this monitor are:

a. Total S.G. blowdown flow rate = 700 gpm.

b. 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 [tCi/ml which is lower than any 10CFR20 Effluent Concentration (EC) limit except for some transuranics *

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

Therefore, the alarm setpoint corresponds to a concentration of:

Alar(p~ / ml) 200 ,000 6 uCi/ml Alarm (uCi/m0) 20,00 x 3x10- 8 + background**= 8.5x10- + background The latest monitor calibration curve shall be used to determine the alarm setpoint in cpm corresponding to 8.5 x 10-6 [tCi/ml.

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

A larm (uC i/ ,nl)= 8 .5 x l O -6uCi/ ml x 2 0 0water 0x S/ ho~ .. . ( g 6 circulating & service flow (gpm) 700 +

200,000 S/Gb1 m~1,do (gpm)

Background = 3xI0-'uCi/nl x circulating & service water flow (gpm) + Background total S/G blowdown (gpm)

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NOTE The Steam Generator Blowdown alarm criteria is in practice based on setpoints required to detect allowable levels of primary to secondary leakage. This alarm criteria is typically more restrictive than that required to meet discharge limits. This fact shall be verified, however, whenever the alarm setpoint is recalculated.

• In lieu of using 3 x 10-8 itCi/ml, the identified EC limits from 10CFR20 may be used.

• • Background of monitor at monitor location (i.e., indication provided by system monitor with no activity present in the monitored system).

5b. Unit 2 Steam Generator Blowdown Effluent Concentration Limitation The results of analysis of blowdown samples required by Table I.C. -2 of Section I of the REMODCM shall be used to ensure that blowdown effluent releases do not exceed ten times the concentration limits in 10CFR20, Appendix B.

6. Unit 2 Condenser Air Ejector - RM5099 N/A since this monitor is no longer a final liquid effluent monitor.

7. Unit 2 Reactor Building Closed Cooling Water RM6038 and Unit 2 Service Water, and RBCCW Sump and Turbine Building Sump Effluent Concentration Limitation 7a. Unit 2 Reactor Building Closed Cooling Water RM6038 The purpose of the Reactor Building Closed Cooling Water (RBCCW) radiation monitor is to give warning of abnormal radioactivity in the RBCCW system and to prevent releases to the Service Water system which, upon release to the environment, would exceed ten times the concentration values in 10CFR20. According to Calculation RERM -02665- R2, radioactivity in RBCCW water which causes a monitor response of greater than the setpoint prescribed below could exceed ten times thel0CFR20 concentrations upon release to the Service Water system.

SETPOINT DURING POWER OPERATIONS:

To give adequate warning of abnormal radioactivity, the setpoint shall be two times the radiation monitor background reading, provided that the

background reading does not exceed 2,000 cpm. The monitor background reading shall be the normal monitor reading. If the monitor background reading exceeds 2,000 cpm, the setpoint shall be set at the background reading plus 2,000 cpm and provisions shall be made to adjust the setpoint if the background decreases.

SETPOINT DURING SHUTDOWN:

1) During outages not exceeding three months the setpoint shall be two times the radiation monitor background reading, provided that the background reading does not exceed 415 cpm. If the monitor background reading exceeds 415 cpm, the setpoint shall be set at the background reading plus 415 cpm and provisions shall be made to adjust the setpoint if the background decreases.

2) During extended outages exceeding three months, but not exceeding three years, the setpoint shall be two times the radiation monitor background reading, provided that the background reading does not exceed 80 cpm. If the monitor background reading exceeds 80 cpm, the setpoint shall be set at the background reading plus 80 cpm and provisions shall be made to adjust the setpoint if the background decreases.

PROVISIONS FOR ALTERNATE DILUTION FLOWS:

These setpoints are based on a dilution flow of 4,000 gpm from one service water train. If additional dilution flow is credited, the setpoint may be adjusted proportionately. For example, the addition of a circulating water pump dilution flow of 100,000 gpm would allow the setpoint to be increased by a factor of 25.

7b. Unit 2 Service Water, and RBCCW Sump and Turbine Building Sump Effluent Concentration Limitation Results of analyses of service water, RBCCW sump and turbine building sump samples taken in accordance with Table I.C.-2 of Section I of the REMODCM shall be used to limit radioactivity concentrations in the service water, RBCCW sump and turbine building sump effluents to less than ten times the limits in 10CFR20, Appendix B.

"STO. 7

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8. Unit 3 Liquid Waste Monitor - LWS-RE70 The setpoints on the Unit 3 liquid waste monitor depend on dilution water flow, radwaste discharge flow, the isotopic composition of the liquid, the background count rate of the monitor and the efficiency of the monitor. Due to the variability of these parameters, the alert and alarm setpoints will be determined prior to the release of each batch. The following method will be used:

Step 1:

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

For Nuclides Other Than Noble Gases:

R1= Required Reduction Factor =1 Zc .1,f nuclide lOx1 EC of nuclide i For Noble Gases: If the noble gas concentration is less than 0.026 [tCi/ml, the reduction factor need not be determined 2 x 10 -4/UCi R2= Required Reduction Factor = 1 Zci

• --"2 ofxnoble gases 10- 4.1 7 *n/ noble gases 2O~ml R= the smaller of R1 or R2 Step 2:

Determine the allowable discharge flow (F)

F = 0.1xRxD Where:

D = The existing dilution flow (D):

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

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NOTE Note that discharging at this flow rate would yield a discharge concentration corresponding to 10% of the Radiological Effluent Control Limit due to the safety factor of 0.1.

With this condition on discharge flow rate met, the monitor setpoint can be calculated:

Rset = 2 x AC x RCF (see Note 1)

Where:

Rset= The setpoint of the monitor.

AC= The total radwaste effluent concentration (ýtCi/ml) in the tank.

RCF= The response correction factor for the effluent line monitor using the current calibration factor or isotopic-specific responses.

2 = The multiple of expected count rate on the monitor based on the radioactivity concentration in the tank.

This value, or that corresponding to 6.6 x 10-5 ýtCi/ml (Note 2 below), I whichever is greater, plus background is the trip setpoint. For the latter setpoint, independent valve verification shall be performed and minimum dilution flow in Note 2 shall be verified and if necessary, appropriately adjusted.

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NOTE

1. If discharging at the allowable discharge rate (F) as determined above, this Alarm setpoint would yield a discharge concentration corresponding to 20% of the Radiological Effluent Control limit.

2. This value is based upon worst case conditions, assuming maximum discharge flow (150 gpm), dilution water flow of 100,000 gpm, and a limit of 1 x 10-7 [Ci/ml which is lower than any Technical Specification limit (ten times 10CFR20 EC values) except for transuranics. This will assure that low level releases are not terminated due to small fluctuations in activity. However, to verify that the correct tank is being discharged when using this value, independent valve verification shall be performed. This value may be adjusted (increased or decreased) by factors to account for the actual discharge flow and actual dilution flow; however, controls shall be established to ensure that the allowable discharge flow is not exceeded and the dilution flow is maintained

9. Unit 3 Regenerant Evaporator Effluent Line - LWC-RE65 The MP3 Regenerant Evaporator has been removed from service with DCR M3-97-041. Therefore a radiation monitor alarm is not needed.

10. Unit 3 Waste Neutralization Sump Effluent Line - CND-RE07 Same asSection II.E.8.

11. Unit 3 Steam Generator Blowdown - SSR-RE08 and Unit 3 Steam Generator Blowdown Effluent Concentration Limitation 11a. Unit 3 Steam Generator Blowdown - SSR-RE08 The alarm setpoint for this monitor assumes:

a. Steam generator blowdown rate of 400 gpm (maximum blowdown total including weekly cleaning of generators - per ERC 25212-ER-99-0133).

b. The release rate limit is conservatively set at 3 x 10-8 [tCi/ml which is well below any 10CFR20 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.

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

Therefore, the alarm setpoint corresponds to a concentration of:

Alam (t~iml) 230, 000 -

Alarm (uCi/m) =2 400 x 3x10 8 + background = 1.7x10- 5 /UCi/ml + background This setpoint may be increased through proper administrative controls if the steam generator blowdown rate is maintained less than 400 gpm and/or more than 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:

circulating & service water flow (gpm) 400 Alarm (uCi/ml) = 1.7xlO-*tCi/ml x +

9lnI Mnl S/G blowdown (gpmn)

Background = 3xlO8 yCi/ml x circulating & service water flow (gpmn) + Background total S/G blowdown (gpm)

NOTE The Steam Generator Blowdown alarm criteria is in practice based on setpoints required to detect allowable levels of primary to secondary leakage. This alarm criteria is typically more restrictive than that required to meet discharge limits. This fact shall be verified, however, whenever the alarm setpoint is recalculated.

• In lieu of using 3 x 10-8 itCi/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.

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12. Unit 3 Turbine Building Floor Drains Effluent Line - DAS-RE50 and Unit 3 Service Water and Turbine Building Sump Effluent Concentration Limitation 12a. Unit 3 Turbine Building Floor Drains Effluent Line - DAS-RE50 The alarm setpoint for this monitor shall be set to four times (4X) the reading of the monitor when there is no gamma radioactivity present in the turbine building sumps. As determined in Calculation RERM-04101R3, the setpoint shall not exceed 1.4 x 10-5 [tCi/ml.

12b. Unit 3 Service Water and Turbine Building Sump Effluent Concentration Limitation Results of analyses of service water and turbine building sump samples taken in accordance with Table I.C.-3 of Section I of the REMODCM shall be used to limit radioactivity concentrations in the service water and turbine building sump effluents to less than ten times the limits in 10CFR20, Appendix B.

13. Bases for Liquid Monitor Setpoints Liquid effluent monitors are provided on discharge pathways to control, as applicable, the release of radioactive materials in liquid effluents during actual or potential releases of liquid waste to the environment. The alarm /

trip setpoints are calculated to ensure that the alarm / trip function of the monitor will occur prior to exceeding ten times the Effluent Concentration (EC) limits of 10 CFR 20 (Appendix B, Table 2, Column 2), which applies to the release of radioactive materials from all units on the site. This limitation also provides additional assurance that the levels of radioactive materials in bodies of water in Unrestricted Areas will result in exposures within the Section II.A. design objectives of Appendix I to 10CFR50 to a member of the public.

In application, the typical approach is to determine the expected concentration in a radioactive release path and set the allowable discharge rate past the monitor such the existing dilution flow will limit the effluent release concentration to 10% of the limit for the mix. The setpoint is then selected to be only 2 times the expected concentration, or 20% of the limit.

As a result, considerable margin is included in the selection of the setpoint for the monitor to account for unexpected changes in the discharge concentration or the contribution from other potential release pathways occurring at the same time as the planned effluent release. For those monitors on systems that are not expected to be contaminated, the alarm point is usually selected to be two times the ambient background to give notice that normal conditions may have changed and should be evaluated.

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II.E Gaseous Monitor Setpoints

1. Unit 1 Spent Fuel Pool Island Monitor - RM-SFPI-02 The instantaneous release rate limit from the site shall be set in accordance with the conditions given in Section II.D.l.a. in order to satisfy 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 [tCi/sec assuming a maximum ventilation flow of 36,000 cfm. The corresponding monitor reading is 1.71E-3 [tCi/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 [tCi/sec assumes that 7% of the site limit for skin dose of 3000 mrem per year is assigned to the Unit 1 Spent Fuel Pool Island vent. If effluent conditions from the Unit 1 Spent Fuel Pool Island vent reach 29,000 [tCi/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.l.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.L.a. in order to satisfy Units 2 Radiological Effluent Controls IVC.2. and IV.D.2.a.

The alarm setpoint shall be set at or below the monitor reading in ýtCi/cc corresponding to 74,000 ýtCi/sec assuming a maximum ventilation flow of 12,000 cfm. The corresponding monitor reading is 1.31E-2 ptCi/cc.

The release rate value of 74,000 ýtCi/sec assumes that 13% of the site limit is assigned to Unit 2 releases to the Millstone Stack. If effluent 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 dictated in Section II.D.l.a., and described in MP-22-REC-REF03, "REMODCM Technical Information Document,"

Section 4.2.

3. Reserved

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

The alarm setpoint shall be set at or below the monitor reading in [tCi/cc corresponding to 74,000 [Ci/sec assuming a maximum ventilation flow of 12,000 cfm. The corresponding monitor reading is 1.16 E-2 [tCi/cc in accordance with Calculation RERM-01946R3, Rev.1.

If effluent conditions from Unit 3 to the Millstone Stack reach 74,000

[tCi/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.l.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.l.a. in order to satisfy Radiological Effluent Controls in Sections IV.C.2. and IVD.2.a.

The alarm setpoint shall be set at or below the "cpm" corresponding to 95,000 RCi/sec from the MP2 vent noble gas monitor calibration curve. The calibration curve (given as [tCi/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.

The release rate value of 95,000 RCi/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 II.D.l.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 [Ci/sec (see the MP-22-REC-REF03, "REMODCM Technical Information Document," Section 4.2 for bases).

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 [tCi/sec.

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The MP2 waste gas decay tank monitor (given [tCi/cc per cpm) calibration curve and the tank discharge rate is used to assure that the concentration of gaseous activity being released from a waste gas decay tank does not cause the setpoint of 7,400 [tCi/sec to be exceeded.

7. Unit 3 Vent Noble Gas Monitor - HVR-RE10B The instantaneous release rate limit from the site shall be set in accordance with the conditions given in Section II.D.l.a. in order to satisfy Radiological Effluent Controls in Sections V.C.2. and VD.2.a.

4 [tCi/cc The alarm setpoint shall be set at or below a value of 8.4 x 10- 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 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 II.D.l.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. 0.

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 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 J __ MP-22-REC-BAP1 TACT -STOP 'REVIEW Rev. 025-02

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release rate values, which equate to the site dose rate limits, also provides operational flexibility to accommodate short periods of higher than normal effluent releases that may occur during plant operations.

APPENDIX II.A REMODCM METHODOLOGY CROSS-REFERENCES Radiological effluent controls (Sections III, IV, and V) identify the requirements for monitoring and limiting liquid and gaseous effluents releases from the site such the resulting dose impacts to members of the public are kept to 'As Low As Reasonably Achievable" (ALARA). The demonstration of compliance with the dose limits is by calculational models that are implemented by Section II of the REMODCM.

Table App. II.A- 1 provides a cross-reference guide between liquid and gaseous effluent release limits and those sections of the REMODCM, which are used to determine compliance. It also shows the administrative Technical Specifications which reference the REMODCM for operation of radioactive waste processing equipment. This table also provides a quick outline of the applicable limits or dose objectives and the required actions if those limits are exceeded. Details of the effluent control requirements and the implementing sections of the REMODCM should be reviewed directly for a full explanation of the requirements.

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Table II.A.- 1 Millstone Effluent Requirements and Methodology Cross Reference Radiological REMODCM Applicable Limit or Exposure Required Action Effluent Controls & Methodology Objective Period Technical Section Specifications IV/V.E.I.a Tables I.C.-2 Ten times 10CFR20App.B, Instantaneous Restore concentration to within lim-4 Liquid Effluent and I.C.-3 Table 2, Column 2, & 2x10- its within 15 mins.

Concentration [tCi/mL for dissolved noble gases*

IV/V.E.I.b II.C.1. !51.5 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-

• 3 mrem TB. Calendar Year culations shall be confirmed by per-II.C.. 1 Oformance of the REMP in Section I.

II.C.4. -<10 mrem Organ_______________

TS. 6.16 (Unit 2) I.C.2. *<0.06 mrem TB. Projected for 31 Return to operation Liquid Waste TS. 6.14 (Unit 3) II.C.5. -*0.2 mrem Organ days (if system Treatment System Liquid Radwaste Treat- not in use) ment III.D.2.a Tables I.D.- 1, !<500 mrem/yr TB. from Instantaneous Restore release rates to within spec-IV/V.D.2.a I.D.-2, & noble gases* ifications within 15 minutes Gaseous Effluents Dose I.D-3 Rate II.D.la. -*3000 mrem/yr skin from noble gases*

II.D.1.b. *<1500 mrem/yr organ from particulates with T 1/2 > 8d.,

1-131, 1-133 & tritium*

III.D.2.b II.D.2. -<5 mrad gamma air Calendar Quar- 30-day report if exceeded IV/V.D.2.bDose Noble *<10 mrad beta air ter**

Gases - 10 mrad gamma air Calendar Year

• <20 mrad beta air II1.D.2.c II.D.3. -<7.5 mrem organ Calendar Quar- 30-day report if exceeded. Relative IV/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 i 15 mrem organ Calendar Year formance of the REMP in Section I.

TS. 5.6.4 (Unit 1) II.D.2. > 0.02 mrad gamma air Projected for 31 Return to operation Gaseous Rad-TS. 6.14 (Unit 2) II.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 mrem TB.* 12 Consecutive 30-day report if Unit 1 Effluent IVI/V.F *<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 2/3 Effluent Con-trol IVV.E.1.2, IVV.E.2.2, or IV/

V.E.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: TB. means total or whole body.

• Applies to the entire site (Units 1, 2, and 3) discharges combined.

• • Cumulative dose contributions calculated once per 31 days.

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SECTION III.

Millstone Unit 1 Radiological Effluent Controls Docket Nos. 50-245 A MP-22-REC-BAP01 "9 7 " ~THINK 7 "ACT>/ "REVIEWRev. 025-02 7 7XN" 96 of 163

SECTION III. REMODCM UNIT ONE CONTROLS III.A. Introduction The purpose of this section is to provide the following for Millstone Unit One:

a. the effluent radiation monitor controls and surveillance requirements,

b. the effluent radioactivity concentration and dose controls and surveillance requirements, and

c. the bases for the controls and surveillance requirements.

Definitions of certain terms are provided as an aid for implementation of the controls and requirements.

Some surveillance requirements refer to specific sub -sections in Sections I and II as part of their required actions III.B. Definitions and Surveillance Requirement (SR) Applicability III.B.1 - Definitions The defined terms of this sub-section appear in capitalized type and are applicable throughout Section III.

1. ACTION - that part of a Control that prescribes remedial measures required under designated conditions.

2. INSTRUMENT CALIBRATION - the adjustment, as necessary, of the instrument output such that it responds within the necessary range and accuracy to know values of the parameter that the instrument monitors. The INSTRUMENT CALIBRATION shall encompass those components, such as sensors, displays, and trip functions, required to perform the specified safety function(s). The INSTRUMENT CALIBRATION shall include the INSTRUMENT FUNCTIONAL TEST and may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

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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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III.B.2 - Surveillance Requirement (SR) Applicability

1. SRs shall be met during specific conditions in the Applicability for individual LCOs unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in III.B.2 3.

Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

2. The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the frequency is met.

3. If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the LCO not met may be delayed from the time of discovery up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified frequency, whichever is less. This delay period is permitted to allow performance of the surveillance. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met and the applicable Condition(s) must be entered. The Completion Times of the Required Actions begin immediately upon expiration of the delay period. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met and the applicable Condition(s) must be entered. The Completion Times of the Required Actions begin immediately upon failure to meet the Surveillance.

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III.C. Radioactive Effluent Monitoring Instrumentation

1. Radioactive Liquid Effluent Monitoring Instrumentation CONTROLS The radioactive liquid effluent monitoring instrumentation channels shown in Table III.C.- 1 shall be OPERABLE with applicable alarm/trip setpoints set to ensure that the limits of Specification III.D.l.a. are not exceeded. The setpoints shall be determined in accordance with methods and parameters described in Section II.

APPLICABILITY: As shown in Table III.C. -1 ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels OPERABLE requirement, take the action shown in Table III.C.- 1.

Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Releases need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENTS Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table III.C.-2.

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TABLE III.C.-1 Radioactive Liquid Effluent Monitoring Instrumentation Instrument Minimum # Alarm Setpoints Applicability Action Operable Required 1.Radioactivity Monitor 1 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 II.D.1 .a. and; (2) The original release rate calculations and discharge valving are independently verified by a second individual.

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TABLE III.C.-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Instrument Channel Source Channel Channel Check Check Calibration Functional Test 1.Radioactivity Monitor Liquid Effluent D* P T(1) Q Line D = Daily P = Prior to each batch release T = Once every two years Q = Once every 3 months

• During releases via this pathway and when the monitor is required OPERABLE per Table 111.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.

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2. Radioactive Gaseous Effluent Monitoring Instrumentation CONTROLS The radioactive gaseous effluent monitoring instrumentation channels shown in Table III.C.-3 shall be OPERABLE with applicable alarm setpoints set to ensure that the limits of Control III.D.2.a. are not exceeded. The setpoints shall be determined in accordance with methods and parameters described in Section II.E1.

APPLICABILITY: As shown in Table III.C.-3 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm setpoint less conservative than required by the above Control, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels operable requirements, take the action shown in Table III.C.-3. Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radiological Effluent Release Report why the inoperability was not corrected in a timely manner. Release need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENT Each 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 III.C.-4.

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TABLE III.C.--3 Radioactive Gaseous Effluent Monitoring Instrumentation Instrument Minimum Alarm Setpoints Applicability Action

1. Operable Required 1.Spent Fuel Pool Island Vent (a) Noble Gas Activity Monitor 1 Yes

• A (b) Particulate Sampler 1 No

• B (c) Vent Flow Rate Monitor 1 No

• C (d) Sampler Flow Rate Monitor 1 Yes

• D 2.Balance of Plant Vent (a) Particulate Sampler 1 No

• B (b) Sampler Flow Monitor 1 Yes

• D Channels are OPERABLE and in service on a continuous, uninterrupted basis when exhaust fans are operating, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required tests, checks, calibrations, and sampling associated with the instrument or any system or component which affects functioning of the instrument.

ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that grab samples are taken daily when fuel is being moved, or during any evolution or event which would threaten fuel integrity, and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I 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.

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 Chemistry compensatory sampling time period as specified in Action A or Action B.

NST8P; THK A

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TABLE III.C.-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Instrument Instrument Instrument Functional Source Check Calibration Test Check 1.Spent Fuel Pool Island Vent (a) Noble Gas Activity Monitor D( 3 ) T(6) Q(7) M (b) Particulate Sampler TM NA NA NA (c) Vent Flow Rate Monitor D T NA NA (d) Sampler Flow Rate Monitor D T NA NA 2.Balance of Plant Vent (a) Particulate Sampler TM NA NA NA (b) Sampler Flow Monitor D T NA NA D = Daily W = Weekly TM = Twice per month M = Monthly Q = Once every 3 months T = Once every two years NA= Not Applicable Table III.C.-4 TABLE NOTATION (1) RESERVED (2) RESERVED (3) Instrument check daily only when there exist releases via this pathway.

(4) RESERVED (5) RESERVED (6) Calibration shall include the use of a known source whose strength is determined by a detector which has been calibrated to a source which is traceable to the NIST These sources shall be in a known reproducible geometry.

(7) The INSTRUMENT FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs ifany of the following conditions exist:

1. Instrument indicates measured levels above the alarm/trip setpoint.

2. Instrument indicates a downscale failure.

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III.D. Radioactive Effluents Concentrations And Dose Limitations

1. Radioactive Liquid Effluents

a. Radioactive Liquid Effluents Concentrations LIMITING CONDITIONS OF OPERATIONS The concentration of radioactive material released from the site (see Figure III.D. -1) shall not exceed ten times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall not exceed 2 x 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 REQUIREMENT Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program specified in Section I.

The results of the radioactive analysis shall be used in accordance with the methods of Section II to assure that the concentrations at the point of release are maintained within the limits of Specification III.D.l.a.

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b. Radioactive Liquid Effluents Doses LIMITING CONDITIONS OF OPERATIONS The dose or dose commitment to any REAL MEMBER OF THE PUBLIC from radioactive materials in liquid effluents from Unit 1 released from the site (see Figure III.D. -1) shall be limited:

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

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 3 mrem to the total body and 10 mrem to any organ.

SURVEILLANCE REQUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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2. Radioactive Gaseous Effluents

a. Radioactive Gaseous Effluents Dose Rate CONTROLS The dose rate, at any time, offsite (See Figure III.D.-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

a. The dose rate limit for noble gases shall be less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin; and,

b. The dose rate limit for Tritium and for all radioactive materials in particulate form with half lives greater than 8 days shall be less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, decrease the release rate to comply with the limit(s) given in Control III.D.2.a. within 15 minutes.

SURVEILLANCE REQUIREMENT

1) The instantaneous release rate corresponding to the above dose rate shall be determined in accordance with the methodology of Section II.

2) The instantaneous release rate shall be monitored in accordance with the requirements of Section III.C.2.

3) Sampling and analysis shall be performed in accordance with Section I to assure that the limits of Control III.D.2.a. are met.

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b. Radioactive Gaseous Effluents Noble Gas Dose I CONTROLS The air dose offsite (see Figure III.D. -1) due to noble gases released in gaseous effluents from Unit 1 shall be limited to the following:

a. During any calendar quarter, to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation;

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

APPLICABILITY: At all times.

ACTION:

With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and the calendar year so that the cumulative dose during the calendar year is within 10 mrad for gamma radiation and 20 mrad for beta radiation.

SURVEILLANCE REQUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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c. Gaseous Effluents - Dose from Radionuclides Other than Noble Gas CONTROLS The dose to any REAL MEMBER OF THE PUBLIC from Tritium and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents released offsite from Unit 1 (see Figure III.D. -1) shall be limited to the following:

a. During any calendar quarter to less than or equal to 7.5 mrem [to any organ];

b. During any calendar year to less than or equal to 15 mrem [to any organ].

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of Tritium and radioactive materials in particulate form exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 15 mrem to any organ.

SURVEILLANCE 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 III.D.-1, "Site Boundary for Liquid and Gaseous Effluents" MP-22-REC-BAP01

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III.E. Total Radiological Dose From Station Operations Controls CONTROLS The annual dose or dose commitment to any REAL MEMBER OF THE PUBLIC, beyond the site boundary, from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls III.D.l.b., III.D.2.b. or III.D.2.c. prepare and submit a Special Report to the Commission within 30 days and limit the subsequent releases such that the dose commitment from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months.

This Special Report shall include an analysis which demonstrates that radiation exposures from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site (including all effluent pathways and direct radiation) are less than the 40 CFR 190 Standard.

If the estimated doses exceed the above limits, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190.

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

SURVEILLANCE REQUIREMENTS Cumulative dose contributions from liquid and gaseous effluents and direct radiation from the Millstone Site shall be determined in accordance with Section II once per 31 days.

III.E BasesSection III.C.1 - Radioactive Liquid Effluent Monitoring Instrumentation No controls required; Unit 1 is not currently releasing radioactivity in liquid effluentsSection III.C.2 - Radioactive Gaseous Effluent Monitoring Instrumentation REC-BAP01

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

Section III.D.l.a. - Radioactive Liquid Effluents Concentrations This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within: (1) the Section II.A. design objectives of Appendix I, 10 CFR 50, to an individual and (2) the limits of 10 CFR 20 to the population. The concentration limit for noble gases is based upon the assumption that Xe -135 is the controlling radioisotope and its concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

Section III.D.l.b. - Radioactive Liquid Effluents Doses This specification is provided to implement the requirements of Sections II.A.,

III.A, and IV.A of Appendix I, 10 CFR 50. The specification implements the guides set forth in Section II.A of Appendix I. The Action statements provide the required operating flexibility and at the same time implement the guides set forth in Section III.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

Section 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 S N - / -MP-22-REC-BAP01 STOP K7 'REVIEW Rev. 025-02

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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 or beyond the site boundary to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to less than or equal to 1500 mrem/year for the nearest cow to the plant.

Section III.D.2.b. - Radioactive Gaseous Effluents Noble Gas Dose This control is provided to implement the requirements of Sections II.B., III.A.,

and IV.A. of Appendix I, 10 CFR 50. The control implements the guides set forth in Section II.B of Appendix I. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IVA of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculational of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.

The ODCM equations provided for determining the air doses at the site boundary were based upon utilizing successively more realistic dose calculational methodologies. More realistic dose calculational methods are used whenever simplified calculations indicate a dose approaching a substantial portion of the regulatory limits. The methods used are, in order, previously determined air dose per released activity ratio, historical meteorological data and actual radionuclide mix released, or real time meteorology and actual radionuclides released.

Section III.D.2.c. - Radioactive Gaseous Effluents, Particulates, and Gas Other Than Noble Gas Doses These controls is provided to implement the requirements of Sections II.C, III.A and IVA of Appendix I, 10 CFR 50. The controls are the guides set forth in Section II.C of Appendix I. The action statements provide the required

operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides for Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials will to be consistent with the methodology provided in Regulatory Guide 1.109, "Calculating of Annual Dose to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I,"

Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision I, July 1977. These equations provide for determining the doses based upon either conservative atmospheric dispersion and an assumed critical nuclide mix or using real time meteorology and specific nuclides released. The release rate specifications for radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man. The pathways which are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

Section III.E. - Total Radiological Dose from Station Operations This control is provided to meet the reporting requirements of 40 CFR 190. For the purpose of the Special Report, it may be assumed that the dose commitment to any REAL MEMBER OF THE PUBLIC from other fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or withina radius of 5 miles must be considered.

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SECTION IV Millstone Unit 2 Radiological Effluent Controls Docket Nos. 50-336

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SECTION IV. REMODCMUNIT 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 provided as an aid for implementation of the controls and requirements.

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

IVB. Definitions, Applicability and Surveillance Requirements IVB.1 - Definitions The defined terms of this sub-section appear in capitalized type and are applicable throughout Section IV

1. ACTION - Those additional requirements specified as corollary statements to each principal control and shall be part of the control.

2. OPERABLE / OPERABILITY - An instrument shall be OPERABLE or have OPERABILITY when it is capable of performing its specified functions(s) and when all necessary attendant instrumentation, controls, normal and emergency electrical power sources, or other auxiliary equipment that are required for the instrument to perform its functions(s) are also capable of performing their related support function(s).

3. CHANNEL CALIBRATION - A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to know values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

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4. CHANNEL CHECK - A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

5. CHANNEL FUNCTIONAL TEST - A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions. For digital instruments, the computer database may be manipulated, in lieu of a signal injection, to verify operability of alarm and/or trip functions.

6. SOURCE CHECK - A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

7. MEMBER(S) OF THE PUBLIC - MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the utility, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location.

8. MODE - Refers to Mode of Operation as defined in Safety Technical Specifications.

9. SITE BOUNDARY - The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

10. UNRESTRICTED AREA - Any area at or beyond the site boundary to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the site boundary used for residential quarters or industrial, commercial, institutional and/or recreational purposes.

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11. DOSE EQUIVALENT 1-131 - DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 ([tCi/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Regulatory Guide 1.109 Rev. 1, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluent for the Purpose of Evaluating Compliance with 10 CFR 50 Appendix I."

IV.B.2 - Applicability IVB.2a - LIMITING CONDITIONS FOR OPERATION

1. Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

2. Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation .and associated ACTION requirements are not met within the specified time intervals, except as provided in Condition IVB.2.a(6). If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3. NOT USED.

4. NOT USED.

5. When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system(s), subsystem(s), train(s), component(s) and device(s) are OPERABLE, or likewise satisfy the requirements of this specification.

6. Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to Condition IVB.2.a(2) for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

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IVB2.b - SURVEILLANCE REQUIREMENTS

1. Surveillance Requirements shall be applicable during any condition specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

2. Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance time interval.

3. Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Condition IVB2.b(2), shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.

4. Entry into any specified condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.

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IVC. Radioactive Effluent Monitoring Instrumentation

1. Radioactive Liquid Effluent Monitoring Instrumentation LIMITING CONDITIONS OF OPERATIONS The radioactive liquid effluent monitoring instrumentation channels shown in Table IVC.- 1 shall be OPERABLE with applicable alarm/trip setpoints set to ensure that the limits of Specification IVD.1.a. are not exceeded. The setpoints shall be determined in accordance with methods and parameters described in Section II.

APPLICABILITY: As shown in Table IV.C. -1 ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels OPERABLE requirement, take the action shown in Table IV.C.- 1.

Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Releases need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENTS Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table IV.C.-2.

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TABLE I.C.- 1 Radioactive Liquid Effluent Monitoring Instrumentation Instrument Minimum Alarm Setpoints Applicability Action I# Operable Required 1.Gross Radioactivity Monitors Providing Automatic Termination Of Release (a) Clean Liquid Radwaste Effluent 1 Yes

• A Line (b)Aerated Liquid Radwaste Effluent 1 Yes *** A Line (c) Steam Generator Blowdown 1 Yes B Monitor (d)Condensate Polishing Facility 1 Yes *** E Waste Neut Sump I 2.Gross Radioactivity Monitors Not Providing Automatic Termination Of Release (a) Reactor Building Closed Cooling 1 Yes Water Monitor#

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 tCi/gm DOSE EQUIVALENT 1-131.

b) Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 [tCi/gm DOSE EQUIVALENT 1-131.

Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> grab samples of the service water effluent are collected and analyzed for gamma radioactivity at LLD as specified in Table I.C.-2; Action D With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves may be used to estimate flow.

Action E With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases may continue provided that best efforts are made to repair the instrument and that prior to initiating a release:

(1) At least two independent samples are analyzed in accordance with the first Surveillance Requirement of Specification IV.D.1 .a., and; (2) If one of the samples has gamma radioactivity greater than any of the LLDs in Table I.C. -2, the original release rate calculations and discharge valving are independently verified by a second individual.

A ~A/

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TABLE IV.C.-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Instrument Channel Source Channel Channel Check Check Calibration Functional Test 1.Gross Radioactivity Monitors Providing Alarm and Automatic Termination Of Release

a. Clean Liquid Radwaste Effluent Line D* P R(1) Q(2)

b. Aerated Liquid Radwaste Effluent D* P R(1) Q(2)

Line

c. Steam Generator Blowdown Monitor D* M R(1) Q(2) d.Condensate Polishing Facility Waste D* P R(1) Q(2)

Neut Sump 2.Gross Radioactivity Monitors Providing Alarm But Not Providing Automatic Termination Of Release

a. Reactor Building Closed Cooling D* M R(1) Q(2)

Water Monitor 3.Flow Rate Measurements

a. Clean Liquid Radwaste Effluent Line D* N/A R 0

b. Aerated Liquid Radwaste Effluent D* N/A R Q Line

c. Condensate 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.

REC-BAP01

2. Radioactive Gaseous Effluent Monitoring Instrumentation LIMITING CONDITIONS OF OPERATIONS The radioactive gaseous effluent monitoring instrumentation channels shown in Table IV.C. -3 shall be OPERABLE with applicable alarm/trip setpoints set to ensure that the limits of Specifications 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 IVC.-3 ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of channels less than the minimum channels OPERABLE requirement, take the action shown in Table IV.C.-3.

Exert best efforts to restore the inoperable monitor to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Release need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENTS Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table IV.C.-4.

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TABLE IV.C.-3 Radioactive Gaseous Effluent Instrumentation Instrument Minimum Alarm Applicability Action Channels Setpoints Operable Required

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

a. Noble Gas Activity Monitor 1 Yes*** ** A

b. Iodine Sampler 1 No ** B

c. Particulate Sampler 1 No ** B

d. Vent Flow Rate Monitor 1 No ** C

e. Sampler Flow Rate Monitor 1 No ** C 2.Millstone Stack - applicable to the WRGM (RM-8169, normal range, channel 1, only; mid range channel 2 and high range channel 3 requirements are contained in TRM LCO 3.3.3.8)

a. Noble Gas Activity Monitor 1 Yes*** ** E IG

b. Iodine Sampler 1 No ** B

c. Particulate Sampler 1 No ** B

d. Stack Flow Rate Monitor 1 No ** C

e. Sampler Flow Rate Monitor 1 No ** C 3.Waste Gas Holdup System

a. Noble Gas Monitor Providing 1 Yes D Automatic Termination of Release

• During waste gas holdup system discharge.

• • At all times when air is being released to the environment by the pathway being monitored. which means that channels be OPERABLE and in service on a continuous, uninterrupted basis, except that outages are permitted for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of maintenance and performance of required tests, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

• No automatic isolation features.

ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that grab samples are taken once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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Action B With the number of samplers OPERABLE less than required by the Minimum number OPERABLE requirement, effluent releases via this pathway may continue provided that the best efforts are made to repair the instrument and that samples are continuously collected with auxiliary sampling equipment for periods of seven (7) days and analyzed for principal gamma emitters with half lives greater than 8 days within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the end of the sampling period. Auxiliary sampling must be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of initiation of this action statement. Auxiliary sampling outages are permitted for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of maintenance and performance of required tests, checks, calibrations, or sampling.

Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated once per 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Action D With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement:

Releases from the Millstone Unit 2 waste gas system may continue provided that best efforts are made to repair the instrument and that prior to initiating the release:

a) At least two independent samples of the tank's contents are analyzed; and b) The original release rate calculations and discharge valve lineups are independently verified by a second individual. Otherwise, suspend releases from the waste gas holdup system.

Action E With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, Millstone Unit 2 releases via the Millstone Stack may continue provided that best efforts are made to repair the instrument and that grab samples are taken once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> STOP THIK

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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

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

a. Noble Gas Activity Monitor D M RF) 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 I 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() Q(2)

b. Iodine Sampler W NA NA NA

c. Particulate Sampler W NA NA NA

d. Stack Flow Rate Monitor D NA R Q(2)

e. Sampler Flow Rate Monitor D NA R NA 3.Waste Gas Holdup System

a. Noble Gas Monitor D* P R1 2

• During releases via this pathway and when the monitor is required OPERABLE per Table IV.C.-3.

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

P = Prior to discharge R = Once every 18 months D = Daily Q = Once every 3 months W = Weekly NA= Not Applicable M = Monthly TABLE I.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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IVD. Radioactive Effluents Concentrations And Dose Limitations

1. Radioactive Liquid Effluents

a. Radioactive Liquid Effluents Concentrations LIMITING CONDITIONS OF OPERATIONS The concentration of radioactive material released from the site (see Figure IV.D. -1) shall not exceed ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall not exceed 2 x 10-4 [tCi/ml total activity.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released from the site exceeding the above limits, restore the concentration to within the above limits within 15 minutes.

SURVEILLANCE REQUIREMENTS

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

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

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b. Radioactive Liquid Effluents Doses LIMITING CONDITIONS OF OPERATIONS The dose or dose commitment to any REAL MEMBER OF THE PUBLIC from radioactive materials in liquid effluents from Unit 2 released from the site (see Figure IVD. -1) shall be limited:

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

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 3 mrem to the total body and 10 mrem to any organ.

SURVEILLANCE 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 Monitoring Program as detailed in Section I.

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2. Radioactive Gaseous Effluents

a. Radioactive Gaseous Effluents Dose Rate LIMITING CONDITIONS OF OPERATIONS The doserate, at any time, offsite (see Figure IVD.-1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

a. The dose rate limit for noble gases shall be less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin; and,

b. The dose rate limit for Iodine-131, Iodine-133, Tritium, and for all radioactive materials in particulate form with half lives greater than 8 days shall be less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, decrease the release rate to comply with the limit(s) given in Specification IV.D.2.a. within 15 minutes.

SURVEILLANCE REQUIREMENTS

1) The release rate, at any time, of noble gases in gaseous effluents shall be controlled by the offsite dose rate as established above in Specification 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 IVC.-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 to noble gases released in gaseous effluents from Unit 2 shall be limited to the following:

a. During any calendar quarter, to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation;

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and the calendar year so that the cumulative dose during the calendar year is within 10 mrad for gamma radiation and 20 mrad for beta radiation.

SURVEILLANCE 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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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 materials 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 calendar year to less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 15 mrem to any organ.

SURVEILLANCE 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" IMP-22-REC-BAP*1 KSOX'-'THINiK' ACTZ REVIEW Rev.05-2

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IV.E. Total Radiological Dose From Station Operation CONTROLS The annual dose or dose commitment to any REAL MEMBER OF THE PUBLIC, beyond the site boundary, from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls IVD.2.a., IV.D.1.b., or IVD.2.c. prepare and submit a Special Report to the Commission within 30 days and limit the subsequent releases such that the dose commitment from the site to any REAL MEMBER OF. THE PUBLIC from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over 12 consecutive months.

This Special Report shall include an analysis which demonstrates that radiation exposures from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site (including all effluent pathways and direct radiation) are less than the 40 CFR 190.

If the estimated doses exceed the above limits, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190.

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

SURVEILLANCE REQUIREMENTS Cumulative dose contributions from liquid and gaseous effluents and direct radiation from the Millstone Site shall be determined in accordance with Section II once per 31 days.

IV.E BasesSection IV.C.1. - Radioactive Liquid Effluent Monitoring Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated in accordance with the approved methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding ten times the limits of 10 CFR 20. The OPERABILITY and use of this

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

Section IVC.2. - Radioactive Gaseous Effluent Monitoring Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with the approved methods in the REMODCM to ensure that the alarm/trip will occur prior to exceeding the dose rate limits, at any time, as specified in Section IV.D.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50.

Two types of radioactive gaseous effluent monitoring instrumentation, monitors and samplers, are being used at MP2 vent and Millstone Stack. Monitors have alarm/trip setpoints and are demonstrated operable by performing one or more of the following operations: CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST Samplers are strictly collection devices made of canisters and filters. The CHANNEL CHECK surveillance requirements are met through (1) documented observation of the in-service rad monitor sample flow prior to filter replacement; (2) documented replacement of in-line iodine and particulate filters; and (3) documented observation of sample flow following the sampler return to service.

The flow indicator is the only indication available for comparison. These observations adequately provide assurance that the sampler is operating and is capable of performing its design function.

There are a number of gaseous release points which could exhibit very low concentrations of radioactivity. For all of these release paths, dose consequences would be insignificant due to the intermittent nature of the release and/or the extremely low concentrations of radioactivity. Since it is not cost-beneficial (nor in many cases practical due to the nature of the release (steam) or the impossibility of detecting such low levels), to monitor these pathways, it has been determined that these release paths require no monitoring or sampling.

Section 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 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 E R.MP-22-REC-BAP01

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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 II.A, III.A, and IVA of Appendix I, 10 CFR 50. The specification implements the guides set forth in Section II.A of Appendix I. The Action statements provide the required operating flexibility and at the same time implement the guides set forth in Section III.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

Section IVD.2.a. - Radioactive Gaseous Effluents Dose Rate This specification is provided to ensure that the dose rate at anytime from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR 20 for all areas offsite. The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table 2. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual offsite to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR

20. For individuals who may, at times, be within the site boundary, the occupancy of that individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the site boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid or any other organ dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year.

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SECTION V Millstone Unit 3 Radiological Effluent Controls Docket Nos. 50-423

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SECTION V REMODCM UNIT THREE CONTROLS VA. 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 V.B.1 - Definitions The defined terms of this sub-section appear in capitalized type and are applicable throughout Section V

1. ACTION - ACTION shall be that part of the control which prescribes remedial measures required under designated conditions.

2. CHANNEL OPERATIONAL TEST - A CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. For digital instruments, the computer database may be manipulated, in lieu of a signal injection, to verify operability of alarm and/or trip functions.

The CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or trip setpoints such that the setpoints are within the required range and accuracy.

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3. CHANNEL CALIBRATION - A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

4. CHANNEL CHECK - A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

5. DOSE EQUIVALENT 1-131 - DOSE EQUIVALENT 1-131 shall be that Q concentration of 1-131 ([tCi/gram) which alone would produce the same CDE-thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed under Inhalation in Federal Guidance Report No. 11 (FGR 11), "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion and Ingestion."

6. MEMBER(S) OF THE PUBLIC - MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the licensee, its contractors or its vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

The term "REAL MEMBER OF THE PUBLIC" means an individual who is exposed to existing dose pathways at one particular location.

7. MODE - Refers to Mode of Operation as defined in Safety Technical Specifications.

8. OPERABLE - OPERABILITY - An instrument shall be OPERABLE or have OPERABILITY when it is capable of performing its specified functions(s) and when all necessary attendant instrumentation, controls, electrical power, or other auxiliary equipment that are required for the instrument to perform its functions(s) are also capable of performing their related support function(s).

9. SITE BOUNDARY - The SITE BOUNDARY shall be that line beyond which the land is not owned, leased, or otherwise controlled by the licensee.

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10. SOURCE CHECK - A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

11. UNRESTRICTED AREA - Any area at or beyond the SITE BOUNDARY to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the SITE BOUNDARY used for residential quarters or industrial, commercial, institutional and/or recreational purposes.

V.B.2 - Applicability V.B.2.a - LIMITING CONDITIONS FOR OPERATION

1. Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.

2. Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

VB.2.b - SURVEILLANCE REQUIREMENTS

1. Surveillance Requirements shall be applicable during any condition specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.

2. Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance time interval.

3. Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Condition V.B.2.b(2), shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.

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4. Entry into any specified condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.

VIC. 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 their Alarm/Trip setpoints set to ensure that the limits of Specification VD.1.a are not exceeded. The alarm/trip setpoints shall be determined in accordance with methodology and parameters as described in Section Ii.

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

a. With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip setpoint less conservative than required by the above specification, without delay suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the action shown in Table VC.-1. Exert best efforts to restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Releases need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENTS Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST at the frequencies shown in Table VC.-2.

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TABLE VC.- 1 Radioactive Liquid Effluent Monitoring Instrumentation Instrument Minimum Applicability Action I# 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 Requirements

a. Waste Neutralization Sump Effluents 1* # C

b. RESERVED

c. Liquid Waste Effluent Line 1 # C

d. RESERVED

e. Steam Generator Blowdown Effluent Line 1 # C

• NA if tritium in the steam generators is less than detectable, or gamma radioactivity in the steam generators is less than 5 x 10-7 [tCi/ml, or the sump is being directed to radwaste.

1. At all times - which means that channels shall be OPERABLE and in service on a continuous, uninterrupted basis, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument.

1. 1. MODEs 1 -5, and MODE 6 when pathway is being used, except that outages are permitted, for a maximum of hours, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument. The monitor must be on-line with no unexpected alarms. When the affected discharge path is isolated in MODE 6, the applicable LCO and Surveillance Requirements are not applicable.

1. 1. 1. MODEs 1-5, except that outages are permitted, for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, for the purpose of maintenance and performance of required test, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument. The monitor must be on-line with no unexpected alarms.

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SECTION VI. Summary of Changes 1.1 Revision 025-02 1.1.1 Created Summary of Changes Section.

1.1.2 Corrected Table I.D-3 Title to Unit 3 from Unit 2.

1.1.3 Re-numbered Table I.E-I and changed number of locations from 3 to 2 under Well Water.

1.1.4 Changed referenced document 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.C.- 1 ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with the first Surveillance Requirement of Specification V.D.1 .a. and;

2. The original release rate calculations and discharge line valving are independently verified by a second individual.

Action B With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided best efforts are made to repair the instrument and that grab samples are analyzed for gamma radioactivity at the lower limits of detection specified in Table I.C.-3:

1. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 [tCi/gram DOSE EQUIVALENT 1-131, or

2. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 pCi/gram DOSE EQUIVALENT 1-131.

Action C With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that best efforts are made to repair the instrument and that the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump performance curves may be used to estimate flow.

Action D With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases may continue provided that best efforts are made to repair the instrument and that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with the first Surveillance Requirement of Specification V.D.1 .a., and;

2. If one of the samples has gamma radioactivity greater than any of the lower limits of detection specified in Table I.C.-3, the original release rate calculations and discharge valving are independently verified by a second individual.

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TABLE VC.-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(1)

Condensate Polishing Facility

b. Turbine Building Floor Drains D M R(2) Q(I)

c. Liquid Waste Monitor D p R(2) Q(1)

d. Deleted

e. Steam Generator Blowdown Monitor D M R(2 )

2.Flow Rate Measurements

a. Waste Neutralization Sump Effluents D( 3) NA R Q

b. RESERVED

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

d. Deleted

e. Steam Generator Blowdown Effluent D(3) N/A R Q Line D = Daily R Once every 18 months M = Monthly Q Once every 3 months P = Prior to each batch release N/A= Not Applicable TABLE 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 VC. -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.

APPLICABILITY: As shown in Table VC.-3.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above specification, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With the number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the Minimum Channels OPERABLE, take the ACTION shown in Table V.C.-3. Exert best efforts to restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner. Release need not be terminated after 30 days provided the specified actions are continued.

SURVEILLANCE REQUIREMENT 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.

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TABLE V.C.-3 Radioactive Gaseous Effluent Monitoring Instrumentation Instrument Minimum Applicability Action Channels Operable 1.Millstone Unit 3 Ventilation Vent (Turbine Building - HVR-RE10B, normal range only; high range monitor, HVR-RE10A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing Alarm

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 monitor, HVR-RE19A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing Alarm

b. Iodine Sampler c, Particulate Sampler

d. Process Flow Rate Monitor

e. Sampler Flow Rate Monitor 3.Engineered Safeguards Building Monitor (HVQ-

a. Noble Gas Activity Monitor

b. Iodine Sampler

c. Particulate Sampler

d. Discharge Flow Rate Monitor

e. Sampler Flow Rate Monitor TABLE V.C.-3 Table Notations Whenever the release path is in service. Outages are permitted for a maximum of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the purpose of maintenance and performance of required tests, checks, calibrations, or sampling associated with the instrument or any system or component which affects functioning of the instrument. I0 MP-22-REC-BAP01

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TABLE V.C.-3 ACTION STATEMENTS Action A With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that a) best efforts are made to repair the instrument and that grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, OR b) if the cause of the inoperability is solely due to a loss of annunciation in the control room and the Remote Indicating Controller (RIC) remains OPERABLE, perform a channel check at the RIC at least once per twelve hours and verify the indication has not alarmed.

Action B With the number of samplers OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that the best efforts are made to repair the instrument and that samples are continuously collected with auxiliary sampling equipment for periods of seven (7) days and analyzed for principal gamma emitters with half lives greater than 8 days within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after-the end of the sampling period. Auxiliary sampling must be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after initiation of this ACTION statement. 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 />.

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, normal range only; high range monitor, HVR-RE10A, requirements are in the TRM)

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

Alarm

b. Iodine Sampler W NA NA NA *

c. Particulate Sampler W NA NA NA *

d. Vent Flow Rate Monitor D NA R Q *

e. Sampler Flow Rate Monitor D NA R Q

• 2.Millstone Stack - applicable to SLCRS (HVR-RE19B, normal range only; high range monitor, HVR-RE19A, requirements are in the TRM)

a. Noble Gas Activity Monitor Providing D M R( 3 ) Q(2)

Alarm

b. Iodine Sampler W NA NA NA *

c. Particulate Sampler W NA NA NA *

d. Process Flow Rate Monitor D NA R Q *

e. Sampler Flow Rate Monitor D NA R Q

• 3.Engineered Safeguards Building Monitor (HVQ-RE49)

a. Noble Gas Activity Monitor Providing D M R(l) Q(2)

Alarm

b. Iodine Sampler W NA NA NA *

c. Particulate Sampler W NA NA NA *

d. Discharge Flow Rate Monitor D NA R Q *

e. Sampler Flow Rate Monitor D NA R Q

• At all times except when the vent path is isolated. I D = Daily R = Once every 18 months W = Weekly Q = Once every 3 months M = Monthly N/A= Not Applicable I./ MP-22-REC-BAP01 SACTREVIEW Rev. 025-02 V\ / \150 of 163

TABLE VC.-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 ifany of the following conditions exist:

a) Instrument indicates measured levels above the Alarm Setpoint, or b) Circuit failure, or c) Instrument indicates a downscale failure.

(3) The CHANNEL CALIBRATION shall include the use of a known source whose strength is determined by a detector which has been calibrated to an NIST source. These sources shall be in know, reproducible geometry.

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VD. Radioactive Effluents Concentrations And Dose Limitations

1. Radioactive Liquid Effluents

a. Radioactive Liquid Effluents Concentrations LIMITING CONDITIONS OF OPERATION The concentration of radioactive material released from the site (see Figure VD.-1) shall be limited to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall not exceed 2 x 10-4 [tCi/ml total activity.

APPLICABILITY: At all times.

ACTION:

With the concentration of radioactive material released from the site exceeding the above limits, restore the concentration to within the above limits within 15 minutes.

SURVEILLANCE REQUIREMENTS

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

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

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b. Radioactive Liquid Effluents Doses LIMITING CONDITIONS OF OPERATION The dose or dose commitment to any REAL MEMBER OF THE PUBLIC from radioactive materials in liquid effluents from Unit 3 released from the site (see Figure 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 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 Monitoring Program as detailed in Section I.

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2. Radioactive Gaseous Effluents

a. Radioactive Gaseous Effluents Dose Rate LIMITING CONDITIONS OF OPERATION The dose rate, at any time, offsite (see Figure VD. -1) due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

a. The dose rate limit for noble gases shall be less than or equal to 500 mrem/yr to the total body and less than or equal to 3000 mrem/yr to the skin; and,

b. The dose rate limit due to inhalation for Iodine-131, Iodine-133, Tritium, and for all radioactive materials in particulate form with half lives greater than 8 days shall be less than or equal to 1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTION:

With the dose rate(s) exceeding the above limits, decrease the release rate to comply with the limit(s) given in Specification VD.2.a. within 15 minutes.

SURVEILLANCE REQUIREMENTS

1) The release rate, at any time, of noble gases in gaseous effluents shall be controlled by the offsite dose rate as established in Specification V.D.2.a. The corresponding release rate shall be determined in accordance with the methodology of Section II.

2) The noble gas effluent monitors of TableV.C.-3 shall be used to control release rates to limit offsite doses within the values established in Specification 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 dose rate shall be determined using the methodology given in Section II.

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b. Radioactive Gaseous Effluents Noble Gas Dose LIMITING CONDITIONS OF OPERATION The air dose offsite (see Figure VD. -1) due to noble gases released from Unit 3 in gaseous effluents shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

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

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose during the calendar year is within 10 mrad for gamma radiation and 20 mrad for beta radiation.

SURVEILLANCE 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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c. Gaseous Effluents - Doses from Radionuclides Other than Noble Gas LIMITING CONDITIONS OF OPERATION The dose to any REAL MEMBER OF THE PUBLIC from Iodine-131, Iodine- 133, Tritium, and radioactive materials in particulate form with half lives greater than 8 days in gaseous effluents released offsite from Unit 3 released offsite (see Figure VD. -1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ and,

b. During any calendar year: Less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

c. With the calculated dose from the release of radioiodines, radioactive materials in particulate form, or radionuclides other than noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce the releases during the remainder of the current calendar quarter and during the remainder of the calendar year so that the cumulative dose or dose commitment to any REAL MEMBER OF THE PUBLIC from such releases during the calendar year is within 15 mrem to any organ.

SURVEILLANCE REQUIREMENTS

1) Dose Calculations - Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with Section II once every 31 days.

2) Relative accuracy or conservatism of the calculations shall be confirmed by performance of the Radiological Environmental Monitoring Program as detailed in Section I.

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Figure V.D. -1, "Site Boundary for Liquid and Gaseous Effluents" MP-22-REC-BAPOI ST TIN ACT REVIEW Rev. 025-02 "I /A 157of 163

VE. Total Radiological Dose From Station. Operations CONTROLS The annual dose or dose commitment to any REAL MEMBER OF THE PUBLIC, beyond the site boundary, from the Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which is limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Controls V.D.1.b., VD.2.b., or V.D.2.c. prepare and submit a Special Report to the Commission within 30 days and limit the subsequent releases such that the dose commitment from the site to any REAL MEMBER OF THE PUBLIC from the' Millstone Site is limited to less than or equal to 25 mrem to the total body or any organ (except thyroid, which is limited to less than or equal to 75 mrem) over consecutive months.

This Special Report shall include an analysis which demonstrates that radiation exposures from the site to any REAL MEMBER OF THE PUBLIC from the Millstone Site (including all effluent pathways and direct radiation) are less than the 40 CFR 190 Standard.

If the estimated doses exceed the above limits, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190.

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

SURVEILLANCE REQUIREMENTS Cumulative dose contributions from liquid and gaseous effluents and direct radiation from the Millstone Site shall be determined in accordance with Section II once per 31 days.

VE BasesSection V.C.1. - Radioactive Liquid Effluent Monitoring Instrumentation The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section II to ensure that the alarm/trip will occur prior to exceeding ten times the limits of 10 CFR 20. The OPERABILITY and

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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 room annunciation" occur. Control room annunciation cannot occur unless the monitor is on line (i.e. in communication with the RMS computer.).

Section VC.1. Surveillance Requirement requires that the ACOT be performed to demonstrate OPERABILITY. General Design Criteria 64 states in part:

"Means shall be provided for monitoring effluent discharge paths for radioactivity." Regulatory Guide 1.21 Appendix A describes a monitor program that is acceptable to the Regulatory staff. Under Section B of Appendix A, "LIQUID EFFLUENTS," the first paragraph states in part: "During the release of radioactive wastes, the effluent control monitor should be set to alarm and to initiate automatic closure..."

Certain of the monitors listed in Table VC.- 1 are designed to operate without sample pumps. These monitors utilize pressure in the effluent line during discharge to provide sample flow and sample pressure. Low sample flow and/or low sample pressure alarms may be received when no discharge is in progress.

These are expected alarms. Sample flow and/or sample pressure will return to normal when the discharge is initiated. These alarms will clear when discharge begins. The monitors are OPERABLE since they are able to perform their specified function with the expected alarms in.

Table VC. -1 note ## requires entry into the radioactive liquid effluent monitoring action statements whenever the radiation monitors are not available in the required MODE. This note applies to items L.a (3CND-RE07, Waste Neutralization Sump), 1.d (3LWC-RE65, Regenerate Evaporator Monitor), and i.e (3SSR-RE08, Steam Generator Blowdown Monitor) in Table VC.- 1. The original issue of this requirement (as a Technical Specification) in January 1986 stated the applicability was "At all times." Technical Specification Amendment 22 added "APPLICABILITY" to Table V.C.- 1 (then Tech Spec Table 3.3- ). The applicability added in Amendment 22 is the present wording. The letter 1312821, MP-22-REC-BAP01 T THINK ACT 025-02 REVW 15Vof26Rev.

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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 (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)."

" In "Significant Hazards Consideration" item 2: "The proposed changes also have no effect on alarm setpoints or control functions. Further, no operator actions that are required to mitigate any accident rely solely on these monitors, and these monitors provide no protective functions."

Technical Specification Amendment 22 provided for the following:

• allowance for planned inoperability of monitoring instrumentation for up to hours for the purpose of maintenance and performance of required test, check, calibration or sampling

" a requirement to initiate auxiliary sampling within hours after inoperability of certain gaseous effluent monitors

" allowance for inoperability of certain effluent monitoring instrumentation, during MODE 6 (refueling) when the effluent pathway is not being used.

Section V.C.2.- Radioactive Gaseous Effluent Monitoring Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section II to ensure that the alarm/trip will occur prior to exceeding the limits of SectionVD.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 20 Part 50.

The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of Specification V.C.2.a shall be such that concentrations as low as 1 x 10-6 [tCi/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 7 ~THINKACT 7 IMP-22-REC-BAPO1 STO * -EVIEW Rev. 025-02

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building. There are no requirements in the REMODCM associated with the vent high range radiation monitor, HVR*10A.

The SLCRS normal range radiation monitor, HVR*19B, satisfies the requirements of SectionV.C.2. for Unit 3 releases to the Millstone Stack. There are no requirements in the REMODCM associated with the SLCRS high range radiation monitor, HVR*19A.

Section V.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 Part 20, Appendix B, Table 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within: (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to an individual and (2) the limits of 10 CFR 20.1301 to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its concentrations in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

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

III.A., and IV.A. of Appendix I, 10 CFR Part 50. The specification implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section III.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable." The dose calculation methodology and parameters in Section II implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The equations specified in Section II for calculating the doses due to the actual release rates of radioactive materials in liquid effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I, Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

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Section V.D.2.a.- Radioactive Gaseous Effluents Dose Rate This specification will ensure that the dose from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for all areas offsite. The annual dose limits specified in this section are the dose limits from the version of 10 CFR Part 20 prior to 1994. Annual dose limit in the current version of 10CFR20 were reduced from 500 to 100 mrem. But REMODCM restrictions will not allow the current annual dose limit to be exceeded because the REMODCM requires termination, within fifteen minutes, of any release which exceed the setpoint and much lower annual dose limits from 10CFR50, Appendix I are implemented. For individuals who may, at times, be within the SITE BOUNDARY, the occupancy of that individual will be usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the whole body or to less than or equal to 3000 mrem/year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid or any other organ dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year.

Section V.D.2.b. - Radioactive Gaseous Effluents Noble Gas Dose This specification is provided to implement the requirements of Sections II.B.,

III.A., and IV.A. of Appendix I, 10 CFR Part 50. The specification implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section VA. of Appendix I to assure that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A. of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in Section II for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents will be consistent with the methodology provided in Regulatory Guide 1.109, "Calculational of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.

The Section II equations provided for determining the air doses at the site boundary are based upon utilizing successively more realistic dose calculational methodologies. More realistic dose calculational methods are used whenever simplified calculations indicate a dose approaching a substantial portion of the

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 VD.2.c. - Radioactive Gaseous Effluents for Radionuclides Other Than Noble Gas These specifications are provided to implement the requirements of Sections II.C., III.A., and IVA. of Appendix I, 10 CFR Part 50. The specifications are the guides set forth in Section II.C. of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A. of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."

The Section II calculational methods specified in the surveillance requirements implement the requirements in Section III.A. of Appendix I that conformance with the guides for Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The Section II calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculating of Annual Dose to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,"

Revision I, July 1977. The release rate specifications for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man. The pathways that are examined in the development of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

Section V.E. - Total Radiological Dose from Station Operations This specification is provided to meet the dose limitations of 40 CFR 190. For the purpose of the Special Report, it may be assumed that the dose commitment to any REAL MEMBER OF THE PUBLIC from other fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered.

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