Text

Part 1 of 2,Nine Mile Point, Unit 2, January-December 2001 Radioactive Effluent Release Report
	ML020720326
Person / Time
Site:	Nine Mile Point
Issue date:	02/28/2002
From:	Montgomery B; Constellation Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NMP2L 2051, RG-1.021, Rev 1
	Download: ML020720326 (175)
	v • d • e

P.O. Box 63 Lycoming, New York 13093 0 Constellation Nuclear Nine Mile Point Nuclear Station February 28, 2002 NMP2L 2051 A Member of the ConstellationEnergy Group U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RE: Nine Mile Point Unit 2 Docket No. 50-4 10 NPF-69

Subject:

January-December 2001 Radioactive Effluent Release Report Gentlemen:

In conformance with the Nine Mile Point Unit 2 (NMP2) Technical Specifications, enclosed is the Radioactive Effluent Report for the reporting period January through December 2001.

Included in this report is a summary of gaseous, liquid, and solid effluents released from the station during the reporting period (Attachments 1 - 6), a summary of revisions to the Offsite Dose Calculation Manual and the Radwaste Process Control Program during the reporting period (Attachments 7 and 8), and an explanation as to the cause and corrective actions regarding the inoperability of any station liquid and/or gaseous effluent monitoring instrumentation (Attachment 9). Attachments 10 and 11 provide a summary and assessment of radiation doses to members of the public within and outside the site boundary, respectively, from liquid and gaseous effluents as well as direct radiation in accordance with 40 CFR 190.

The format used for the effluent data is outlined in Appendix B of Regulatory Guide 1.21, Revision 1. Dose assessments were made in accordance with the NMP2 Offsite Dose Calculation Manual. Distribution is in accordance with 10 CFR 50.4(b)(1) and the Technical Specifications.

Attachment 12 to this report is an update of actual data for the third and fourth quarters 2000 used in the July through December 2000 Semi-Annual Radioactive Effluent Release Report.

Page 2 NMP2L 2051 Dose 3 is a Summary of Changes to the Environmental Monitoring and Calculation Locations. 4 is a copy of Revision 22 of the Offsite Dose Calculation Manual. 5 is a copy of Revision 5 of the Radwaste Process Control Program.

did not exceed During the reporting period from January through December 2001, NMP2 Manual any 10 CFR 20, 10 CFR 50, Technical Specification, or Offsite Dose Calculation limits for gaseous or liquid effluents.

If you have any questions concerning the attached report, please contact Mr. Anthony Salvagno, (315) 349-1456, Engineering Services, Nine Mile Point.

Very truly yours, ce . Montg mery Ge ral Manager Nuclear Egineering B SM/CLW/cld Enclosure xc: Mr. H. J. Miller, Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector, Region I Mr. P. S. Tam, Senior Project Manager, NRR (2 copies)

Records Management

NINE MILE POINT NUCLEAR STATION - UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT January- December 2001 SConstellat~on Nulea~r Nine Mile Point Nuclear Station A Member or Ihe Cllati# Ehrqy Grmfp

Page 1 of 2 NINE MILE POINT NUCLEAR STATION - UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER 2001 SUPPLEMENTAL INFORMATION Facility: Nine Mile Point Unit #2 Licensee: Nine Mile Point Nuclear Station, LLC Controls Program)

1. TECHNICAL SPECIFICATION PROGRAM - (ODCM Limits - Radioactive Effluent A) FISSION AND ACTIVATION GASES the

1. The dose rate limit of noble gases released in gaseous effluents from the site to areas at or beyond to the whole body and less than or equal to site boundary shall be less than or equal to 500 mrem/year 3000 mrem/year to the skin.

areas at or

2. The air dose from noble gases released in gaseous effluents from Nine Mile Point Unit 2 to quarter to less than or equal to 5 mrad for beyond the site boundary shall be limited during any calendar radiation, and during any calendar year to gamma radiation and less than or equal to 10 mrad for beta less than or equal to 20 mrad for beta radiation.

less than or equal to 10 mrad for gamma radiation and B&C) TRITIUM, IODINES AND PARTICULATES, HALF LIVES > 8 DAYS form with

1. The dose rate limit of Iodine-13 1, Iodine-133, Tritium and all radionuclides in particulate from the site to areas at or beyond the site half-lives greater than eight days, released in gaseous effluents boundary shall be less than or equal to 1500 mrem/year to any organ.

in

2. The dose to a member of the public from Iodine-13 1, Iodine-133, Tritium and all radionuclides gaseous effluents released from Nine Mile particulate form with half-lives greater than eight days in to less Point Unit 2 to areas at or beyond the site boundary shall be limited during any calendar quarter year to less than or equal to 15 mrem to than or equal to 7.5 mrem to any organ and, during any calendar any organ.

D) LIQUID EFFLUENTS shall be

1. The concentration of radioactive material released in the liquid effluents to unrestricted areas Appendix B, Table 2, limited to ten times the concentrations specified in 10CFR Part 20.1001-20.2402, For dissolved or entrained Column 2 for radionuclides other than dissolved or entrained noble gases.

noble gases, the concentration shall be limited to 2E-04 microcuries/ml total activity.

Page 2 of 2 in liquid effluents

2. The dose or dose commitment to a member of the public from radioactive materials areas shall be limited during any calendar quarter released from Nine Mile Point Unit 2 to unrestricted body and to less than or equal to 5 mrem to any organ, to less than or equal to 1.5 mrem to the whole equal to 3 mrem to the whole body and to less than or equal and during any calendar year to less than or to 10 mrem to any organ.

RADIOACTIVITY

2. MEASUREMENTS AND APPROXIMATIONS OF TOTAL the total radioactivity and radionuclide Described below are the methods used to measure or approximate composition in effluents.

A) FISSION AND ACTIVATION GASES monitoring (intrinsic germanium Noble gas effluent activity is determined by on-line gamma spectroscopic crystal) of an isokinetic sample stream.

B) IODINES analysis (at least weekly) of charcoal cartridges Iodine effluent activity is determined by gamma spectroscopic sampled from an isokinetic sample stream.

C) PARTICULATES Building vent is determined by Activity released from the main stack and the combined Radwaste/Reactor filters sampled from an isokinetic sample stream gamma spectroscopic analysis (at least weekly) of particulate and composite analysis of the filters for non-gamma emitters.

D) TRITIUM gas proportional counting of monthly samples Tritium effluent activity is measured by liquid scintillation or taken with an air sparging/water trap apparatus.

E) LIQUID EFFLUENTS of a representative sample of each Isotopic contents of liquid effluents are determined by isotopic analysis batch and composite analysis of non-gamma emitters.

F) SOLID EFFLUENTS spectroscopy analyses of a representative Isotopic contents of waste shipments are determined by gamma composite sample analyses conducted off-site sample of each batch. Scaling factors established from primary of non-gamma emitters. For low activity are applied, where appropriate, to find estimated concentration and application of appropriate scaling trash shipments, curie content is estimated by dose rate measurement factors.

ATTACHMENT 1 Summary Data Page 1 of 2 Unit 1 Unit 2 X Reporting Period January - December 2001 Liquid Effluents:

10CFR20.1001-20.2402, Appendix B, Table 2, Column 21 Average MEC - pCi/ml (Qtr. 1) = 8.55E-03 Average MEC - ýtCi/ml (Qtr. 3) = 7.84E-03 Average MEC - pCi/ml (Qtr. 2) = 7.49E-03 Average MEC - jCi/ml (Qtr. 4) = 5.95E-03 Average Energy (Fission and Activation gases - Mev):

Qtr. 1: Ey = 3.22E-01 B = 4.26E-01 Qtr. 2: By = 2.80E-01 = 3.33E-01 Qtr. 3: By = 6.66E-01 Fp = 1.04E+00 Qtr. 4: ty = 7.86E-01 Ep = 8.91E-01 Liquid:

Number of batch releases : 89 Total time period for batch releases (hrs) : 2.85E + 02 Maximum time period for a batch release (hrs) : 3.32E+00 Average time period for a batch release (hrs) 3.20E + 00 Minimum time period for a batch release (hrs) 1.88E-02 Total volume of water used to dilute the liquid effluent during the release lt 2 d 3rd 4 th period (L) 1.66E+08 3.85E+08 7.79E+08 6.58E+08 Total volume of water used to dilute the liquid effluent during reporting lst 2 nd 3 rd 4 1h Period (L) : 1.23E+10 1.37E+ 10 1.46E+10 1.39E+10 Gaseous (Emergency Condenser Vent): "Not Applicable for Unit 2" Number of batch releases : N/A Total time period for batch releases (hrs) : N/A Maximum time period for a batch release (hrs) : N/A Average time period for a batch release (hrs) : N/A Minimum time period for a batch release (hrs) : N/A Gaseous (Primary Containment Purge):

Number of batch releases : 14 Total time period for batch releases (hrs) : 3.57E+02 Maximum time period for a batch release (hrsl : 5.58E+01 Average time period for a batch release (hrs) : 2.55E+01 Minimum time period for a batch release (hrs) 4.50E+00 1 Improved Technical Specifications limit the concentration of radioactive material released in the liquid effluents to unrestricted areas to ten times the concentrations specified in 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2.

Maximum Effluent Concentrations (MEC) numerically equal to ten times the 10CFR20.1001-20.2402 concentrations were adopted to evaluate liquid effluents.

ATTACHMENT 1 Summary Data Page 2 of 2 Unit 1 __ Unit 2 X Reporting Period January - December 2001 Abnormal Releases: There were no abnormal releases during this report period.

A. Liquids:

Number of releases 0 Total activity released N/A Ci B. Gaseous:

Number of releases 0 Total activity released N/A Ci

ATTACHMENT 2 Unit 1 Unit 2 X Reporting Period January - December 2001 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES, ELEVATED AND GROUND LEVEL 1st 2nd 3rd 4th EST.

QUAR--TER QUARTER QUARTER QUARTER TOTAL ERROR, %

A. Fission & Activation gases 1.62E + 01 1.02E+00 1.35E+00 7.67E+00 5.00E+01

1. Total release Ci 9.65E-01 2.03E+00 jCi/se 1.31 E-01 1.72E-01

2. Average release rate C

B. lodines 3.00E+01 Ci 1.47E-05 9.23E-05 1.93E-05 4.53E-04

1. Total Iodine-1 31 5.75E-05 gCi/se 1.87E-06 1.17E-05 2.45E-06

2. Average release rate for period C

1 C. Particulates 3.00E+01 Ci 6.17E-04 3.52E-03 4.91 E-04 1.74E-03

1. Particulates with half-lives >8 2.22E-04 XCi/se 7.85E-05 4.67E-04 6.25E-05 days 2.50E+01 C 1.1 8E-05 2.46E-05 3.76E-05 2.09E-05

2. Average release rate for period

3. Gross alpha radioactivity Ci 1

D. Tritium 5.OOE+01 Ci 4.92E+00 5.49E+ 00 9.OOE+00 9.30E+00

1. Total release 1 .18E+00

,uCi/se 6.26E-01 7.28E-01 1.15E+00

2. Average release rate for period C

E. Percent of Tech. Spec. Limits Fission and Activation Gases 7.28E-04 8.1 OE-04 1.18E-02 2.94E-02 Percent of Quarterly Gamma Air Dose Limit (5 mR) 1.59E-03 4.80E-05 4.97E-05 8.84E-04 Percent of Quarterly Beta Air Dose Limit 3.64E-04 7.69E-04 6.67E-03 2.14E-02 (10 mrad)

Percent of Annual Gamma Air Dose 4.91 E-04 1.29E-03 2.40E-05 4.88E-05 Limit to Date (10 mR)

Percent of Annual Beta Air Dose 1.13E-03 2.82E-05 3.13E-05 4.49E-04 Limit to Date (20 mradl Percent of Whole Body Dose Rate 2.35E-04 6.OOE-06 6.57E-06 9.76E-05 Limit (500 mrem/yr)

Percent of Skin Dose Rate Limit (3000 mrem/yr) 1 Tritium, lodines, and Particulates (with half-lives greater than 8 days) 7.69E-03 4.03E-02 9.18E-03 1.19E-01 Percent of Quarterly Dose Limit (7.5 mrem) 2.42E-02 2.88E-02 1.16E-01 3.87E-03 Percent of Annual Dose Limit (15 mrem) 1.54E-04 8.43E-04 1.84E-04 2.39E-03 Percent of Organ Dose Rate Limit (1500 mrem/yr) ___________ .1___________ - ______

1 Tritium, Iron-55, and Strontium results for the fourth quarter were not received from the off-site vendor at the time of this Release Report.

reo*rt. These values include estimates, and actual numbers will be provided in the next Radioactive Effluent

ATTACHMENT 3 Unit 1 Unit 2 X Reporting Period January - December 2001 GASEOUS EFFLUENTS - ELEVATED RELEASE 3

CONTINUOUS MODE 1st 2"d 3rd 4th Nuclides Released QUARTER QUARTER QUARTER QUARTER

1. Fission Gases 1 Argon-41 Ci 1.79E-02 1.85E-02 1 .16E-02 3.99E-01 Krypton-85 Ci Krypton-85m Ci 6.68E-01 8.68E-01 4.75E-01 1.51E+00 Krypton-87 Ci 1.93E-02 2.18E-01 2.20E-01 Krypton-88 Ci 2.74E-02 7.20E-01 3.08E+00 Xenon-127 Ci Xenon-131 m Ci Xenon-1 33 Ci 8.70E-02 Xenon-133m Ci Xenon-135 Ci 4.57E-02 2.60E-01 5.36E-02 4.51 E-01 Xenon-1 35m Ci 6.87E-02 1.30E-01 4.14E-01 8.31 E-01 Xenon-1 37 Ci 9.99E-02 5.66E-02 3.62E+00 6.11 E+00 Xenon-138 Ci 1.17E-01 2.53E-02 2.16E+00 3.48E+00

2. lodines' Iodine-1 31 Ci 1.47E-05 9.23E-05 1.93E-05 4.53E-04 Iodine-1 33 Ci 8.49E-05 1.93E-05 4.10E-05 4.28E-03 Iodine-1 35 Ci

3. Particulates1.2 Strontium-89 Ci ** ** ** 1.39E-05 Strontium-90 Ci ** ** ** 4.54E-06 Cesium-134 Ci Cesium-1 37 Ci Cobalt-60 Ci 1.42E-05 2.01 E-05 3.67E-05 3.49E-05 Cobalt-58 Ci Manganese-54 Ci 4.49E-06 7.52E-06 2o80E-05 Barium-Lanthanum-1 40 Ci Antimony-125 Ci ** ** ** **

Niobium-95 Ci Cerium-141 Ci ** ** ** **

Cerium-144 Ci ** 77 ** 46 Iron-59 Ci ** *5.1E0 ** **

Cesium-136 Ci Chromium-51 Ci 1.1E0 ** ** **

Zinc-65 Ci ** 5.6Oi1-06 TW Iron-55 Ci ITT 4.15E-05 **4.1i6E-05 Molybdenum-99 Ci 1.71E-06 * **

Silver-i 1Om Ci 2.04E-06 ** ** **

Ci 2

4. Tritium Ci 3.49E+ 00 4.58E+00 7.60E+00 7.90E+00 1 Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk. A lower limit of detection of 1.OOE-04 gCi/ml for required noble gases, 1.OOE-11i.Ci/ml for required particulates, 1.OOE- 12 p.tCi/ml for required lodines, and 1.OOE-06 tLCi/ml for Tritium, as required by the Off-Site Dose Calculation Manual (ODCM),

has been verified.

2 Tritium, Iron-55, and Strontium results for the fourth quarter were not received from the off-site vendor at the time of this report. These values include estimates. Actual values will be included in the next Radioactive Effluent Release Report.

3 Contributions from purges are included.

ATTACHMENT 4 Page 1 of 2 Unit 1 Unit 2 X Reporting Period January - December 2001 GASEOUS EFFLUENTS - GROUND LEVEL RELEASES CONTINUOUS MODE 1 st 2nd 3rd 4 th QUARTER QUARTER QUARTER QUARTER 1

1. Fission Gases Argon-41 Ci Krypton-85 Ci Krypton-85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-127 Ci Xenon-131 m Ci Xenon-1 33 Ci Xenon-1 33m Ci Xenon-1 35 Ci Xenon-135m Ci Xenon-1 37 Ci Xenon- 138 Ci 1

2. lodines Iodine-1 31 Ci Iodine-133 Ci Iodine-135 Ci

3. Particulates"2 Strontium-89 Ci ** ** ** 2.67E-05 Strontium-90 Ci ** ** ** 3.29E-05 Cesium-134 Ci Cesium-1 37 Ci 4.47E-06 ** **

Cobalt-60 Ci 1.08E-04 6.34E-04 1.18E-04 1.10E-04 Cobalt-58 Ci 6.13E-06 7.21E-05 2.36E-05 Manganese-54 Ci 5.92E-05 4.77E-04 2.75E-05 1.33E-04 Barium-Lanthanum-140 Ci Antimony-125 Ci Niobium-95 Ci Cerium- 141 Ci Cerium- 144 Ci Ci ** 1.85E-04 ** 5.28 E-05 Iron-59 Cesium-1 36 Ci ** 4.09E-04 ** 7 T73E-05 Chromium-51 Ci

- 1.08E-04 ** **

Zinc-65 Ci 3.98E-04 1.57E-03 3.15E-04 1.16E-03 Iron-55 Ci Molybdenum-99 Ci Silver- 11Om Ci 1.86E-05 ** ** **

Ci 2

4. Tritium Ci 1.43E + 00 9.16E-01 1.40E+00 1.40E + 00 1 Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk. A lower limit of detection of 1.OOE-04 gCi/ml for required noble gases, 1.OQE-11 u Ci/ml for required particulates, 1.OOE- 12

,Ci/ml for required lodines, and 1.OOE-06 uCi/ml for Tritium, as required by the Off-Site Dose Calculation Manual (ODCM),

has been verified.

2 Tritium, Iron-55 and Strontium 89 and 90 results for the fourth quarter were not received from the off-site vendor at the time of this report. These values include estimates, and actual values will be included in the next Radioactive Effluent Release Report.

ATTACHMENT 4 Page 2 of 2 Unit 1 Unit 2 X Reporting Period January - December 2001 GASEOUS EFFLUENTS - GROUND LEVEL RELEASES BATCH MODE There were no batch releases during the reporting period.

1"t 2nd 3rd 4th QUARTER QUARTER QUARTER QUARTER

1. Fission Gases1 Argon-41 Ci Krypton-85 Ci Krypton-85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-127 Ci Xenon-131m Ci Xenon-133 Ci Xenon-133m Ci Xenon-1 35 Ci Xenon-135m Ci Xenon-1 37 Ci Xenon-138 Ci

2. lodines 1 Iodine-131 Ci Iodine-133 Ci Iodine-135 Ci

3. Particulates 1.2 Strontium-89 Ci Strontium-90 Ci Cesium-134 Ci Cesium-137 Ci Cobalt-60 Ci Cobalt-58 Ci Manganese-54 Ci Barium-Lanthanum-140 Ci Antimony-1 25 Ci Niobium-95 Ci Cerium-141 Ci Cerium-1 44 Ci Iron-59 Ci Cesium-1 36 Ci Chromium-51 Ci Zinc-65 Ci Iron-55 Ci Molybdenum-99 Ci Silver-1 1Om Ci

4. Tritium 2 Ci Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk. A lower limit of detection of 1.iOE-04 uCi/ml for required noble gases, 1.OOE-11 uCi/ml for required particulates, 1.iOE- 12 pCi/ml for required lodines, and 1.OOE-06 pCi/ml for Tritium, as required by the Off-Site Dose Calculation Manual (ODCM),

has been verified.

2 Tritium, Iron-55 and Strontium 89 and 90 results for the fourth quarter were not received from the off-site vendor at the time of this report. These values include estimates, and actual values will be included in the next Radioactive Effluent Release Report.

ATTACHMENT 5 Page 1 of 2 Unit 1 Unit 2 X Reporting Period January - December 2001 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES 1st 2nd 3rd 4th EST. TOTAL QUARTER QUARTER QUARTER QUARTER ERROR, %

1 A. Fission & Activation Products

1. Total release (not including Ci 4.07E-03 3.04E-02 3.48E-02 7.04E-02 5.00E+01 Tritium, gases, alpha)

2. Average diluted concentration /UCi/ml 3.31E-10 2.22E-09 2.38E-09 5.07E-09 during reporting period 1

B. Tritium

1. Total release Ci 1.90E+00 6.81E+00 1.16E+01 1.08E+01 5.00E+01

2. Average diluted concentration p*Ci/ml 1.54E-07 4.97E-07 7.90E-07 7.77E-07 during reporting period 3

C. Dissolved and Entrained Gases

1. Total release Ci 5.OOE+01

2. Average diluted concentration gCi/ml during reporting period 3

D. Gross Alpha Radioactivity

1. Total release Ci *** ** 6.03E-05 5.OOE+01 E. Volumes

1. Prior to dilution Liters 6.19E+05 1.50E+06 3.09E+06 2.56E+06 5.O0E+01

2. Volume of dilution water used Liters 1.66E+08 3.85E+08 7.79E+08 6.58E+08 5.OOE+01 during release period

3. Volume of dilution water available Liters 1.23E+ 10 1.37E+ 10 1.46E+ 10 1.39E+ 10 5.OOE+01 during reporting period:

F. Percent of Technical Specification Limits 2.OOE-02 8.25E-02 1.52E-01 2.19E-01 Percent of Quarterly Whole Body Dose Limit (1.5 mrem) 2.78E-02 1.42E-01 1.60E-01 3.52E-01 Percent of Quarterly Organ Dose Limit (5 mrem) 1.OOE-02 5.15E-02 1.28E-01 2.46E-01 Percent of Annual Whole Body Dose Limit to Date (3 mrem) 1.39E-02 8.46E-02 1.62E-01 3.39E-01 Percent of Annual Organ Dose Limit to Date (10 mrem) 1.81 E-03 6.67E-03 1.01 E-02 1.31E-02 Percent of 1 OCFR20 Concentration Limit 2.4 Percent of Dissolved or Entrained Noble Gas Limit (2.OOE-04 UiCi/ml) 34 SIron-55, Strontium 89 and 90 and Tritium results for the fourth quarter were not received from the off-site vendor at the time of this report. These values include estimates, and actual values will be included in the next Radioactive Effluent Release Report.

2 The percent of 1 OCFR20 concentration limit is based on the average concentration during the quarter.

3 Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk. A lower limit of detection of 5.00E-07 uCi/ml for required gamma emitting nuclides, 1.00E-05 uCi/ml for required dissolved and entrained noble gases and Tritium, 5.00E-08 pCi/ml for Sr-89/90, 1.00E-06,uCi/ml for Fe-55 and 1.OOE-07 uCi/ml for gross alpha radioactivity, as required by the Off-Site Dose Calculation Manual (ODCM), has been verified.

4 Improved Technical Specifications limit the concentration of radioactive material released in the liquid effluents to unrestricted areas to ten times the concentrations specified in 10CFR20.1001-20.2402, Appendix B, Table 2, Column 2. Maximum Effluent Concentrations (MEC) numerically equal to ten times the 10CFR20.1001-20.2402 concentrations were adopted to evaluate liquid effluents.

ATTACHMENT 5 Page 2 of 2 Unit 1 Unit 2 X Reporting Period January - December 2001 LIQUID EFFLUENTS RELEASED 3

BATCH MODE 1st 2nd 3r 4th QUARTER QUARTER QUARTER QUARTER Nuclides Released'12 Ci 1.86E-05 6.47E-04 1.33E-03 4.11E-04 Silver- 11Om Ci ** 7.43E-05 5.46E-05 3.02E-05 Arsenic-76 Ci 2.07-E-05 2.44E-04 6.78E-04 1.37E-04 Gold-199 Barium-140 Ci ** ** ** **

Cerium-141 Ci**-*--* **

Ci ** **

Cerium-144 Ci 2.45-E-05 3.17E-04 2.31E-04 1.27E-03 Cobalt-58 Ci 5.20E-04 3.60E-03 5.75E-03 1.52E-02 Cobalt-60 Ci 1.82E-04 2.46E-03 1.07E-03 3.17E-03 Chromium-51 Cesium-134 Ci ** ** ** **

Cesium-136 Ci *- 2.72E-05 ** **

Ci *I ** ** **

Cesium-137 Ci 6.99E-06 6.22E-04 5.31E-04 1.70E-03 Copper-64 Ci 2.57E-04 2.82E-03 ** 4.93E-03 Iron-55 Ci 1.31 E-04 2.43E-03 2.05E-03 4.38E-03 Iron-59 Iodine-131 Ci **

Ci *T **

Iodine-132 lodine-133 Ci *- *-

Lanthanum-140 Ci *- ** ** **

Ci 2.70-E-03 1 .64E-02 1 .25E-02 3.67E-02 Manganese-54 Manganese-56 Ci ** ** ** **

Molybdenum-99 Ci 1.12E-05 ** **

Sodium-24 Ci *- ** ** **

Ci *7 *7 1.43E-05 Niobium-95 Nickel-65 Ci ** *- *- **

Ci ** ** ** **

Neptunium-239 Ci *' 9.20E-05 3.15E-05 3.81 E-04 Antimony-124 Ci *- ** ** 5.08E-05 Strontium-89 Ci *7 *--** 3.50E-05 Strontium-90 **

Ci *- *- **

Strontium-92 Technecium-99m Ci *- 1.18E-05 ** **

Ci ** ** **

Tellurium-132 Ci IT-T V Tungsten-187 Ci 2.11E-04 6.94E-04 1.70E-03 1.99E-03 Zinc-65 Zinc-69m Ci 1.14E-05 2.65E-05 Ci .- .. ** **

Zirconium-95 Ci *V *- ** **

Zirconium-97 Ci ** ** **

Dissolved or Entrained Gases' **

6.81 E+ 00 1.16E + 01 1.08E+01 Tritium 2 Ci 1.90E+00 1 Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk. A lower dissolved and limit of detection of 5.OOE-07 gCi/ml for required gamma emitting nuclides, 1.OOE-05 ACi/ml for required uCi/ml for gross entrained noble gases and Tritium, 5.OOE-08 ,Ci/ml for Sr-89/90, 1.OOE-06 uCi/ml for Fe-55 and 1.OOE-07 alpha radioactivity, as required by the Off-Site Dose Calculation Manual (ODCM), has been verified.

2 Iron-55, Strontium 89 and 90 and Tritium results for the fourth quarter were not received from the off-site vendor at the time of Effluent Release Report.

this report. These values include estimates, and actual values will be included in the next Radioactive 3 No continuous mode releases occurred during the reporting period.

ATTACHMENT 6 Page 1 of 6 Unit 1 Unit 2 X Reporting Period January - December 2001 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS 1

A. TYPE Volume Activity 3

(M ) (Ci)

Class Class A B C A B C

1. Spent Resins (Dewatered) 8.40E+01 0 0 1.46E + 02 0 0

2. Dry Active Waste 0 0 _0 0 0 0

3. Irradiated Components, Control Rods, 0 3.41E+00 3.25E+00 0 2.17E+01 3.97E+04 etc.

4. Other:

(to Vendor for Processing or Consolidation)

a. Dry Active Waste (Compactible 1.76E+02 0 0 4.06E+00 0 0 and Non-Compactible)

b. Spent Resins (Dewatered) 3.50E+01 0 0 8.32E+01 0 0

c. DAW - Contaminated Equipment 8.34E+01 0 0 4.30E-01 0 0 (Non-compactible) 1 The estimated total error is 5.OOE+01%.

Pans 2 of 6 Unit 1 Unit 2 X Reporting Period January - December 2001 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A.1 TYPE Solidification Container Package Agent

1. Spent Resins (Dewatered) HIC - Poly w/ STP None steel shell HIC - Poly STP None

2. Dry Active Waste (Compactible and Non-Compactible) N/A N/A N/A

3. Irradiated Components, Control Rods, etc. Steel Liner Type B None HIC - Poly Type B None

4. Other: (To Vendor for Processing or Consolidation)

a. Dry Active Waste (Compactible and Non-Compactible) Metal Box STP None HIC STP None

b. Spent Resins (Dewatered) HIC Type A None

c. DAW - Contaminated Equipment (Non-Compactible) Metal Box STP None

ATTACHMENT 6 Paqe 3 of 6 Unit 1 Unit 2 X Reporting Period January - December 2001 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A.2 ESTIMATE OF MAJOR NUCLIDE COMPOSITION (BY TYPE OF WASTE)

1. Spent Resins (Dewatered):

Nuclide Percent (1) Fe-55 6.63E+01 (2) Co-60 1.50E+01 (3) Zn-65 9.29E+00 (4) Mn-54 8.37E+00 (5) Other 1.04E+00

2. Dry Compressible Waste:

Nuclide Percent

3. Irradiated Components, Control Rods, etc.:

Nuclide Percent (1) Co-60 5.87E+01 (2) Fe-55 3.73E+01 (3) Ni-63 3.38E+00 (4) Other 6.20E-01

4. Other: (to Vendor for Processing or Consolidation)

a. Dry Active Waste (Compactible and Non Compactible)

Nuclide Percent (1) Fe-55 4.70E+01 (2) Zn-65 2.56E+01 (3) Co-60 2.13E+01 (4) Mn-54 4.40E + 00 (5) Other 1.70E+00

b. Spent Resins (Dewatered)

Nuclide Percent (1) Fe-55 6.97E+01 (2) Co-60 1.59E+01 (3) Mn-54 7.06E +00 (4) Zn-65 6.05E+00 (5) Other 1.29E+00

c. DAW - Contaminated Equipment (Non Compactible)

Nuclide Percent (1) Co-60 5.47E+01 (2) Fe-55 2.43E+01 (3) Zn-65 1.35E+01 (4) Ni-63 4.16E+00 (5) Mn-54 1.87E+00 (6) Other 1.47E+00

ATTACHMENT 6 Page 4 of 6 Unit 1 Unit 2 X Reporting Period January - December 2001 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A.3. SOLID WASTE DISPOSITION Mode of Transportation Destination Number of Shipments Truck GTS Duratek 9 Oak Ridge, TN Truck ATG Richland Corporation 1 Richland, WA Truck ATG Catalytics, LLC 6 Kingston, TN Truck Barnwell Waste Management 15 Facility Barnwell, SC Truck Chem Nuclear Consolidation Facility 4 Barnwell, SC B. IRRADIATED FUEL SHIPMENTS (DISPOSITION): There were no shipments.

Number of Shipments Mode of Transportation Destination 0 N/A N/A

Page 5 of 6 Unit 1 Unit 2 X Reporting Period January - December 2001 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS C. SOLID WASTE SHIPPED OFF-SITE TO VENDORS FOR PROCESSING AND SUBSEQUENT BURIAL These totals Below is a summary of NMP-2 radwaste buried by vendor facilities during January - December 2001.

and burial were reported separately from "10CFR61 Solid Waste Shipped for Burial" since (a) waste classification and (b) Improved Technical Specification (ITS) Section 5.6.3 requires reporting of was performed by the vendors, 10CFR61) shipped off-site during the reporting period." The "information for each class of solid waste (as defined by data represents the actual shipments made from the off-site vendors of our radwaste (e.g., non-compacted following to burial.

÷*h ,4r* nnn-,mnre*rihl waste, scrap metal, and resins) that was processed and commingled prior C.1. TYPE OF WASTE - Non-compacted trash, dry non compressible waste, scrap metals, and resins processed Burial Volume Activity Est. Total 3

by vendor facilities prior to burial. (m ) (Ci) Error, %

1.78E+01 4.23E-01 5.OOE+01 C.2 ESTIMATE OF MAJOR NUCLIDE COMPOSITION Nuclide Percent (1) Co-60 4.03E+01 (2) Mn-54 1.90E+01 (3) Zn-65 1.61E+01 (4) Fe-55 1.39E+01 (5) Cr-51 7.52E+00 (6) Co-58 1.43E+ 00 (7) Other 1.75E+00 C.3 SOLID WASTE DISPOSITION Number of Shipments Mode of Transportation Destination 43 Truck Envirocare, UT

ATTACHMENT 6 Page 6 of 6 Unit 1 Unit 2 X Reporting Period January - December 2001 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS D. SEWAGE WASTES SHIPPED TO A TREATMENT FACILITY FOR PROCESSING AND BURIAL Below is a summary of the sewage sludge, which was removed from the site sanitary treatment facility and transferred to a municipal sewage treatment facility, for subsequent drying and disposal to a landfill. This is a site release and therefore includes the results from both Unit 1 and Unit 2 activities.

Burial Volume 3 Activity Est. Total D. 1 TYPE OF WASTE - (M ) (Ci) Error, %

Sewage Sludge 2.72E+O01 1.72E-03 5.OOE+01 D. 2 ESTIMATE OF MAJOR NUCLIDE COMPOSITION Nuclide Percent (1) H-3 5.63E +01 (2) Ni-63 1.48E+01 (3) C-14 1.31E+01 (4) Tc-99 7.88E+00 (5) 1-129 7.13E +00 (6) Other 7.90E-01 D. 3 SOLID WASTE DISPOSITION Number of Shipments Mode of Transportation Destination 1 Truck Landfill

ATTACHMENT 7 Page 1 of 2 Unit 1 Unit 2 X Reporting Period January - December 2001

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL (ODCM)

The Unit 2 Off-Site Dose Calculation Manual (ODCM) Revision 22 was implemented in December 2001. Administrative changes were made to reflect a change in site ownership, editorial changes for clarification, and a new control location for milk sampling.

The ODCM changes do not reduce the overall conformance of existing criteria in accordance with Technical Specifications. A copy of the ODCM, Revision 22 is attached and below is a summary of the changes accepted by the Station Operations Review C,,,nrnmit-fp nn rlc~emher 11. 2001.

New/Amended Description of Old Page # New Page # Section # Change Reason for Change In the definition for MEMBER(S) OF THE PUBLIC, replaced "the Niagara Mohawk Power Corporation, the Nine Mile Point Unit 2 co 1 1.0-1 I 1.0-1 1.0 tenants, the New York State Power Authority" Administrative with "the owners and operators of the Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant".

In the definition for SITE BOUNDARY, replaced "the Niagara Mohawk Power Corporation or the I 1.0-2 I 1.0-2 1.0 New York Power Authority" with "the owners Administrative and operators of the Nine Mile Point Station and James A. Fitzpatrick Nuclear Power Plant".

In the definition for UNRESTRICTED AREA, replaced "the Niagara Mohawk Power 1 1.0-2 1 1 .0-2 1 .0 Corporation and the New York State Power Authority" with "the owners and operators of the Administrative Nine Mile Point Nuclear Station and James A.

Fitzpatrick Nuclear Power Plant".

Replaced the words "Niagara Mohawk Power Corporation" and "Power Authority State of New York" with "Nine Mile Point Nuclear Station, LLC*" and "ENTERGY", respectively. Also added note: "Niagara Mohawk Power I 1.0-4 I. 1 .0-4 1 .0 Corporation retains ownership in certain Administrative transmission line and switchyard facilities within the exclusion area boundary. Access and usage are controlled by Nine Mile Point Nuclear Station, LLC by Agreement".

Added words to clarify that even though the Offgas System Noble Gas Activity Monitor is included in Table D 3.3.2-1, "Radioactive IB 3.3-2 IB 3.3-2 lB 3.3-2 Gaseous Effluent Monitoring Instrumentation", it Administrative is not an effluent monitor. Its alert setpoint is based on ITS SR 3.7.4.1 and its trip setpoint is based on 10CFR100.

Added REPORTING REQUIREMENTS Section and D 4.1.2, "Annual Radiological Environmental 14.1-1 14.1-1 and D 4.1.2 and Operating Reports", and Section D 4.1.3, Administrative N/A I 4.1-1a D 4.1.3 "Radiological Effluent Release Report" which were omitted in the ITS implementation.

Added "in accordance with Technical 1111 1111 2.1.1 Specification 5.5.4.g" at the end of the 1' Administrative sentence of Section 2.1.1.

Added "in accordance with Technical 1111 1111 2.1.1 Specification 3.7.4" at the end of the last Administrative sentence of Section 2.1.1.

Replaced "per 10CFR20" with "per Technical 1115 II 15 2.2 Specification 5.5.4.g" at the end of the 2 nd Administrative paragraph of Section 2.2.

Section 4.1, combined the 2 nd and 3,d sentences Editorial II 29 II 29 4.1 in the 1t paragraph for clarification. I

ATTACHMENT 7 Page 2 of 2 Unit 1 Unit 2 X Reporting Period January - December 2001

SUMMARY

OF CHANGES TO THE OFF-SITE DOSE CALCULATION MANUAL (ODCM)

Old Page # New Page # New/Amended Description of Section # Sectin # CangeReason Change for Change Reworded the 2 nd paragraph of Section 4.1 to clarify that the calculated dispersion and II 29 II 29 4.1 deposition parameters are averages, and to refer Administrative to Controls 3.5.1 and 3.5.2 and the Radiological Environmental Monitoring Program.

Table D 2-2 through D 2-5; Added adult, teen, 13 33 through Table D 2-2 child and infant liquid dose factors for Co-57, I through II 36 11 through 135 and Nb-95m, per DER 2000-4230, using the Administrative 36 II 36 Table D 2-5 methodology and equations given in Appendix A to the ODCM.

II 63 I163 Table D 5.1 Map Location 45, Milk Location #7 was deleted. Administrative have been Administrative 1163 1163 Table D 5.1 "*"

deletedandasthe they are no longerfootnote corresponding necessary. Administrative II 63 II 63 Table D 5.1 Map Location 47, Milk Location #65 was deleted. Administrative Map Location 73, Milk Control Location (Woodworth), 13.9 mi @ 2340 SW, was replaced Administrative II 63 II 63 Table D.5.1 by Map Location 77, Milk Location (Summerville), ODCM Table D3.5.1-1 13.9 mi @ 191* SSW, as Mr. Woodworth retired and sold his herd.

Figure 5.1-2 (page 1 of 2) was revised to delete I11107 II1107 Figure 5.1-2 Map Location 45. Administrative Figure 5.1-2 (page 2 of 2) was revised to delete 11108 11108 Figure 5.1-2a Map Locations 47 and 73, and add Map Location Administrative 77.

ATTACHMENT 8 Unit 1 - Unit 2 X Reporting Period January - December 2001

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM (RPCP)

The Unit 2 Radwaste Process Control Program (RPCP) Revision 5 was implemented in February 2001. Administrative changes were made to reflect changes in training requirements and clarification of existing procedure requirements. The RPCP changes do not reduce the overall conformance in accordance with Technical Specifications. A copy of the RPCP, Revision 5 is attached and below is a summary of the changes accepted by the Station Operations Review Committee.

Old Page # New Page # New/Amended Description of Reason for Change Section # Change All All N/A Changed page format, font, and font size. Editorial 6 7 3.2.9 Replaced "bi-annual recertification" with "recertification every two years". Clarification 11 12 3.6.2.e Replaced "Retraining of Radwaste Operator personnel on an annual basis" with "Continuing Administrative Training of Radwaste Operator personnel on a cyclic basis (i.e., every 4 years)".

ATTACHMENT 9 Unit 1 Unit 2 X Reporting Period January - December 2001

SUMMARY

OF INOPERABLE MONITORS There were no inoperable monitors for a period greater than 30 days during the reporting period.

ATTACHMENT 10 Doses to Members of the Public Due To Their Activities Inside the Site Boundary

Page 1 of 4 ATTACHMENT 10 RADIOACTIVE EFFLUENT RELEASE REPORT (2001)

NINE MILE POINT NUCLEAR STATION UNIT 2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY JANUARY - DECEMBER 2001 Doses to members of the public (as defined by the Unit 2 Off-Site Dose Calculation Manual (ODCM)) from the operation of the Nine Mile Point Unit 2 (NMP2) facility as a result of activity inside the site boundary are based on activities at the Energy Center located approximately one quarter mile west of Nine Mile Point Unit 1 (NMP1). This facility was open to the public and offered educational information, summer picnicking activities and fishing; however, since the events of September 11, 2001, access has been restricted. Any possible doses received by a member of the public by utilizing the private road that transverses the east and west site boundaries prior to September 11, 2001 are not considered here since it takes a matter of minutes to travel the distance.

The activity at the Energy Center that is used for the dose analysis is fishing near the shoreline adjacent to the NMP site. Although access to this area has been restricted since September 11, 2001, the dose analysis will assume access granted for all of 2001. Dose pathways considered for this activity include direct radiation, inhalation and external ground (shoreline sediment or soil) doses. Other pathways, such as ingestion pathways, are not considered because they are either not applicable, insignificant, or are considered as part of the evaluation of the total dose to a member of the public located off-site. In addition, only releases from the NMP2 stack and vent were evaluated for the inhalation pathway.

The direct radiation pathway is evaluated in accordance with the methodology found in the Off Site Dose Calculation Manual (ODCM). This pathway considers four components: direct radiation from the generating facilities, direct radiation from any possible overhead plume, direct radiation from ground deposition and direct radiation plume submersion. The direct radiation pathway is evaluated by the use of high sensitivity environmental Thermoluminescent Dosimeters (TLDs). Since any significant fishing activity near the Energy Center occurs between April through December, environmental TLD data for the approximate period of April 1 - December 31, 2001 were considered. Data from environmental TLDs from the approximate area where the fishing occurs were compared to control environmental TLD locations for the same time period.

The average fishing area TLD dose rate was 7.02E-03 mRem per hour for the period. The average control TLD dose rate was 6.02E-03 mRem per hour for the period (approximate second, third and fourth calendar quarters of the year). The average increase in dose as a result of fishing in this area at a conservative frequency of eight hours per week for thirty-nine weeks is

3. 1OE-01 mRem from direct radiation for the period in question. The majority of the dose from this pathway is from the NMP1 facility because of its proximity to the fishing area. A small portion may be due to the NMP2 facility.

Page 2 of 4 ATTACHMENT 10 RADIOACTIVE EFFLUENT RELEASE REPORT (2001)

NINE MILE POINT NUCLEAR STATION UNIT 2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY JANUARY - DECEMBER 2001 The inhalation dose pathway is evaluated by utilizing the inhalation equation in the ODCM, as adapted from Regulatory Guide 1.109. The equation basically gives a total inhalation dose in mRem for the time period in question (April - December). The total dose equals the sum, for all applicable radionuclides, of the NMP2 stack and vent release concentrations, times the average NMP2 stack and vent flow rate, times the applicable five-year average calculated X/Q, times the inhalation dose factors from Regulatory Guide 1.109, Table E-7, times the Regulatory Guide 1.109 annual air intake, times the fractional portion of the year in question. In order to be slightly conservative, no radiological decay is assumed.

The 2001 calculation utilized the following information:

NMP2 Stack:

Unit 2 average stack flowrate = 4.45E+0 1 m3 /sec

X/Q value = 9.60E-07 (annual NWN sector, historical average)

Inhalation dose factor = Table E-7 of Regulatory Guide 1.109

Annual air intake = 8000 m3 per year (adult) e Fractional portion of the year = 0.03 56 (312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months ) 3

H-3 = 1.90E+04 pCi/m 3

Mn-54 = 3.52E-02 pCi/m 3

e Fe-55 = 8.24E-02 pCi/m 3

0 Co-60 = 8.11E-02 pCi/m 3

Zn-65 = 4.99E-03 pCi/m

Sr-89= 1.38E-02 pCi/m3

Sr-90 4.49E-03 pCi/m3 3

1-131 = 5.57E-01 pCi/m

1-133 = 4.29E+00 pCi/m3

Page 3 of 4 ATTACHMENT 10 RADIOACTIVE EFFLUENT RELEASE REPORT (2001)

NINE MILE POINT NUCLEAR STATION UNIT 2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE SITE BOUNDARY JANUARY - DECEMBER 2001 NMP2 Vent:

3

"* Unit 2 average vent flowrate = 9.82E+01 m /sec

"* X/Q value = 2.8E-06 (conservative ground level value)

"* Inhalation dose factor = Table E-7 of Regulatory Guide 1.109

"* Annual Air intake = 8000 m3 per year (adult)

"* Fractional portion of the 3year = 0.0356 (312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months )

"* H-3 = 1.59E+03 pCi/m 3

"* Cr-51 = 2.20E-01 pCi/m 3

"* Mn-54 2.86E-01 pCi/m

"* Fe-55 = 1.34E+00 pCi/m3

"* Fe-59 = 1.07E-01 pCi/m3

"* Co-58 4.31E-02 pCi/m 3 3

"* Co-60 3.83E-01 pCi/m 3

"* Zn-65 = 4.92E-02 pCi/mr 3

"* Sr-89 = 1.17E-02 pCi/m

"* Sr-90 = 1.45E-02 pCi/m3 inside the site The inhalation dose to a member of the public from NMIP2 as a result of activities to the boundary is 8.76E-05 mRem to the lung (maximum organ dose) and 5.71E-05 mRem whole body.

The dose from standing on the shoreline while fishing is based on the methodology in the ODCM, was performed as adapted from Regulatory Guide 1.109. During 2001, it was noted that fishing to be from the shoreline on many occasions although waders were also utilized. In order at all conservative, it is assumed that the maximum exposed individual fished from the shoreline times.

of all plant The ODCM equation gives the total dose to the whole body and skin from the sum related radionuclides detected in shoreline sediment samples. The plant-related radionuclide concentration is adjusted for background sample results, as applicable. The equation, therefore, adjusted for yields the whole body and skin dose by multiplying the radionuclide concentration soil density in any background data (as applicable), times a usage factor, times the sediment or grams per square meter (to a depth of one centimeter), times the applicable shore width factor, which the times the regulatory guide dose factor, times the fractional portion of the year over decay dose is applicable. In order to be conservative and to simplify the equation, no radiological is assumed since the applicable radionuclides are usually long lived.

Page 4 of 4 ATTACHMENT 10 RADIOACTIVE EFFLUENT RELEASE REPORT (2001)

NINE MILE POINT NUCLEAR STATION UNIT 2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY JANUARY - DECEMBER 2001 The calculation utilized the following information:

"* Usage factor = 312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months

"* Density in grams per square meter = 40,000

"* Shore width factor = 0.3

"* Whole body and skin dose factor for each radionuclide = Regulatory Guide 1.109, Table E-6

"* Fractional portion of the year = 1 (used average radionuclide concentration over total time period)

"* Average Cs-137 concentration = 0.22 pCi/g The total whole body and skin dose from standing on the shoreline to fish is 3.44E-03 mRem whole body and 4.02E-03 mRem skin dose for the period.

Doses to members of the public relative to activities inside the site boundary from aquatic pathways other than ground dose from shoreline sediment/soil are not applicable.

In summary, the total dose to a member of the public as a result of activities inside the site boundary from the direct radiation, inhalation and shoreline dose pathways is 3.1 OE-0 I mRem to the whole body and 8.76E-05 mRem to the maximum exposed internal organ (lung). The dose to the skin of an adult is 4.02E-03 mRem. These doses are generally a result of the operation of NMP2. However, a portion of these doses for the direct radiation pathway may be attributable to the NvP 1 facility.

ATTACHMENT 11 Doses to Members of the Public Due To Their Activities Outside the Site Boundary

Page 1 of 3 ATTACHMENT 11 RADIOACTIVE EFFLUENT RELEASE REPORT (2001)

NINE MILE POINT NUCLEAR STATION UNIT 2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY JANUARY - DECEMBER 2001 Radiation doses to the likely most exposed member of the public outside of the site boundary are evaluated relative to 40 CFR 190 requirements. The dose limits of 40 CFR 190 are 25 mRem (whole body or organ) per calendar year and 75 mRem (thyroid) per calendar year. The intent of 40 CFR 190 also requires that the effluents of Nine Mile Point Unit 2 (NMP2), as well as other nearby uranium fuel cycle facilities, be considered. In this case, the effluents of Nine Mile Point Unit 1 (NMP 1), NMP2 and the James A. FitzPatrick (JAF) facilities must be considered.

Doses to the likely most exposed member of the public as a result of effluents from the site can be evaluated by using calculated dose modeling based on the accepted methodologies of the facilities' Off-Site Dose Calculation Manuals (ODCMs) or may, in some cases, be calculated from the analysis results of actual environmental samples. Acceptable methods of calculating doses on from environmental samples are also found in the facilities' ODCMs. These methods are based Regulatory Guide 1.109 methodology.

Dose calculations from actual environmental samples are, at times, difficult to perform for some pathways. Some pathway doses should be estimated using calculational dose modeling. These pathways include noble gas air dose, inhalation dose, etc. Other pathway doses may be calculated directly from environmental sample concentrations using Regulatory Guide 1.109 methodology.

Since the effluents from the generating facilities are low, the resultant gaseous and liquid effluent are doses are anticipated to be low. In view of this, doses can be based on calculated data. Doses not based on actual environmental data for 2001 with the exception of doses from direct radiation, fish consumption and shoreline sediment. In addition, in order to be conservative and for the sake of simplicity, it is assumed in the dose calculations that the likely most exposed member of the public is positioned in the maximum receptor location for each pathway at the same time. This approach is utilized because the doses are very low and the computations are greatly simplified.

The following pathways are considered:

1. The inhalation dose is calculated at the critical residence because of the high occupancy factor.

In order to be conservative, the maximum whole body and organ dose assumes no correction for residing inside a residence.

Page 2 of 3 ATTACHMENT 11 RADIOACTIVE EFFLUENT RELEASE REPORT (2001)

NINE MILE POINT NUCLEAR STATION UNIT 2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY JANUARY - DECEMBER 2001

2. The milk ingestion dose is calculated utilizing the maximum milk cow location. As noted previously, in order to be conservative and for the sake of simplicity, the likely most exposed member of the public is assumed to be at all critical receptors at one time. In this case, the member of the public at the critical residence is assumed to consume milk from the critical milk location.

3. The maximum dose from the milk ingestion pathway as a result of consuming goat's milk is based on the same criteria established for item "2" above (ingestion of cow's milk).

4. The maximum dose associated from consuming meat is based on the critical meat animal.

The likely most exposed member at the critical residence is assumed to consume meat from the critical meat animal location.

5. The maximum site dose associated with the consumption of vegetables is calculated from the critical vegetable garden location. As noted previously, the likely most exposed member of the public is assumed to be located at the critical residence and is assumed to consume vegetables from the critical garden location.

6. The dose, as a result of direct gamma radiation from the site, encompasses doses from direct "shine" from the generating facilities, direct radiation from any overhead gaseous plumes, plume submersion and from ground deposition. This total dose is measured by environmental TLDs. The critical location is based on the closest year-round residence from the generating facilities as well as the closest residence in the critical downwind sector in order to evaluate both direct radiation from the generating facilities and gaseous plumes as determined by the local meteorology. During 2001, the closest residence and the critical downwind residence are at the same location.

The measured average dose for 2001 at the critical residence was 54.8 mRem. The average control dose was 52.8 mRem. The average dose at the critical residence is slightly greater than the average control location dose. The net increase in dose is due to the differences between doses from naturally occurring radionuclides in the soil and rock at the different locations and due to the standard deviation in TLD measurements. There is not a significant difference between the control and critical resident dose and is within expected normal statistical variation.

Page 3 of 3 ATTACHMENT 11 RADIOACTIVE EFFLUENT RELEASE REPORT (2001)

NINE MILE POINT NUCLEAR STATION UNIT 2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY JANUARY - DECEMBER 2001

7. The dose, as a result of fish consumption, is considered as part of the aquatic pathway. The dose for 2001 is calculated from actual results of the analysis of environmental fish samples.

For the sake of being conservative, the average plant-related radionuclide concentrations were utilized from fish samples taken near the site discharge points. No plant related radionuclides were detected in either the control or indicator samples. Therefore, no dose was calculated and was assumed to be zero for this pathway.

8. The shoreline sediment pathway is considered relative to recreational activities. The dose due to recreational activities from shoreline sediment is based on the methodology in the ODCM, as adapted from Regulatory Guide 1.109. The ODCM gives the total dose to the whole body and skin from the sum of plant-related radionuclides detected in actual shoreline sediment samples. The plant-related radionuclide concentration is adjusted for background sample results, as applicable. The total whole body and skin dose from shoreline recreational activities are 7.5 1E-04 mRem whole body and 8.77E-04 mRem skin dose for the period.

In summary, the maximum dose to the likely most exposed member of the public is 3.25E-01 mRem to the Thyroid (maximum organ dose) and 2.45E-0 1 mRem to the whole body. It should be noted that the maximum organ dose and maximum whole body doses are based on the sum of the maximum doses observed for all three facilities regardless of age group. This results in some conservatism. The maximum organ and whole body doses were a result of gaseous effluents.

Doses as a result of liquid effluents are secondary. The total whole body and skin dose from shoreline recreational activities are 7.5 1E-04 mRem whole body and 8.77E-04 mRem skin dose for the period. The direct radiation dose to the critical residence from the generating facilities was insignificant or zero. The dose to an individual as a result of fish consumption was zero. These maximum total doses are a result of operations at the NMP1, NMP2 and the JAF facilities. The maximum organ dose and whole body dose are below the 40 CFR 190 criteria of 25 mRem per calendar year to the maximum exposed organ or the whole body, and below 75 mRem per calendar year to the thyroid.

ATTACHMENT 12 Update of Actual Data for the Third and Fourth Quarter 2000

ATTACHMENT 1 2 Paqe 1 of 1 Unit 1 Unit 2 X Reporting Period July - December 2001 UPDATE OF RELEASE AND DOSE DATA FOR GASEOUS (ELEVATED AND GROUND LEVEL) AND LIQUID EFFLUENTS Update of data using actual results from the off-site vendors for Strontium, Tritium, and Iron-55 for the third and fourth quarter 2000. r T I LIQUID GASEOUS GASEOUS LIQUID LIQUID 4th QUARTER 2000 3rd QUARTER 2000 4th QUARTER 2000 3rd QUARTER 2000 Activity (Ci) Activity (Ci) Activity (Ci)

Nuclide' Activity (Ci)

Sr-89 3.40E-04 Sr-90 4.64E+00 3.74E+ 00 3.08E + 00 8.85E+00 H-3 1.15E-03 2.50E-03 9.29E-05 6.48E-03 Fe-55 I T r LIQUID GASEOUS GASEOUS LIQUID LIQUID 3rd QUARTER 4th QUARTER 3rd QUARTER 4th QUARTER Particulates 6.61 E-04 2.14E-02 5.20E-02 Ci 3.71 E-03

1. Particulates with half-lives >8 days 1.58E-09 3.93E-09 I.Ci/sec 4.38E-04 8.41 E-05

2. Average release rate (gaseous) or diluted concentration (liquid) (gaseous) for reporting period I.Ci/ml (liquid)

Tritium 3.74E+00 3.08E+00 8.85E+00 4.64E +-00

1. Total release Ci 3.51E-07 4.41E-01 3.92E-01 6.51 E-07

2. Average release rate for period p.Ci/sec (gaseous) or diluted (gaseous) concentration (liquids) for the I.Ci/ml reporting period (liquid) 4 + I I LIQUID GASEOUS GASEOUS LIQUID LIQUID Tritium, lodines, and Particulates 3rd QUARTER 4th QUARTER 3rd QUARTER 4th QUARTER (with half-lives greater than 8 days) 2 1.34E-02 1.26E-01 1.98E-01 Percent of Quarterly 2.50E-02

1. (Quarterly) (Quarterly)

(Quarterly (Quarterly Dose Limit (Gaseous - 7.5 mrem, Liquid - 1.5 mrem)2 4.47E-02 3.83E-01 4.78E-01 Percent of Annual 3.79E-02

2. (Annual) (Annual)

(Annual) (Annual)

Dose Limit to Date (Gaseous 15 mrem, Liquid - 3 mrem) 2.70E-04 1.16E-01 3.17E-01 4.66E-04

3. Percent of Organ -Dose Rate (Quarterly) (Quarterly)

Quarterly Quarterly Limit (Gaseous - 1500 2.51E-01 3.89E-01 mrem/yr), Dose Limit (Liquid (Annual) (Annual) 5 mrem Quarter, 10 mrem Annual) 2.35E-02 1.67E-02

4. Percent of 10CFR203 (Quarterly) (Quarterly)

Concentration Limit (Liquid) 1.41 E-06

5. Percent of Dissolved or (Quarterly) (Quarterly)

Entrained Noble Gas (Liquid 2.00E-04 p.Ci/ml) _____________ .1 Concentrations less than the lower limit of detection, as required by the Off-Site Dose Calculation Manual (ODCM) following implementation of Improved Technical Specifications (ITS) are indicated with a double asterisk.

2 The dose is to the whole body for liquid effluents and to the maximally exposed organ for gaseous effluents.

3 The oercent of the 10CFR20 concentration limit is based on the average concentration during the quarter.

ATTACHMENT 13 Summary of Changes to the Environmental Monitoring and Dose Calculation Locations

ATTACHMENT 13 Unit 1 Unit 2 X Reporting Period January - December 2001 LOCATIONS

SUMMARY

OF CHANGES TO THE ENVIRONMENTAL MONITORING AND CALCULATION Changes in Environmental Monitoring Locations by a new control During the report period, the control location for milk, (Woodworth) map location 73, was replaced herd. Sample location location for milk (Summerville), map location 77, as Mr. Woodworth has retired and sold his for changes in selections are based on the annual land use census. Refer to Attachment 7, changes to the ODCM, distance and direction.

New Locations for Dose Calculations on the results of the During the report period, no changes in Dose Calculation Receptor Locations were required based land use census.

ATTACHMENT 14 Off-Site Dose Calculation Manual (ODCM)

NINE MILE POINT NUCLEAR STATION NINE MILE POINT UNIT 2 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

DATE APPROVALS SIGNATURES REVISION 22 Prepared by:

G. R. Stinson Health Physicist Reviewed by: 42. 1/ 0/

T. M. Kurtz ,/

Health Physicist Concurred by: -ýýd Aýýýý V 121/1 01 T. G. Kulczyc1'I Supervisor, Analysis Services Concurred by: /9//2, 0/-

W. R. Yaeger Manager, Engineering Services M. F. Peckham vi~'hiyY-Plalnt General Manager Unit 2

/I. /y&'

B. S. Montgomeryi" Genera Mana Ncer ern V-/SZO

SUMMARY

OFREVISIONS Revision 22 (Effective December 2001 PAGE DATE 1 3.3-13,14 August 2000 1 3.3-6 November 2000 14.0-1 November 2000 II 2-10,26,33-36,66,67,75,80 November 2000 ix, I 1.0-1, 11.0-2, 1 B 3.3-2, 14.1-1 & la, 1111, 1115, 1129,1163, 11107, 11108 December 2001 i Unit 2 Revision 22 December 2001

TABLE OF CONTENTS PAGE List of Tables vii List of Figures ix Introduction x PART I - RADIOLOGICAL EFFLUENT CONTROLS I SECTION 1.0 DEFINITIONS 1 1.0-0 SECTION 2.0 Not Used SECTION 3.0 APPLICABILITY 1 3.0-0 D 3.1 Radioactive Liquid Effluents 1 3.1-1 D 3.1.1 Liquid Effluents Concentration 13.1-1 D 3.1.2 Liquid Effluents Dose 13.1-4 D 3.1.3 Liquid Radwaste Treatment System 1 3.1-7 D 3.2 Radioactive Gaseous Effluents 13.2-1 D 3.2.1 Gaseous Effluents Dose Rate 13.2-1 D 3.2.2 Gaseous Effluents Noble Gas Dose 13-2-4 D 3.2.3 Gaseous Effluents Dose - Iodine-13 1, Iodine-133, Tritium, and Radioactive Material in Particulate Form 1 3.2-7 D 3.2.4 Gaseous Radwaste Treatment System 1 3.2-10 D 3.2.5 Ventilation Exhaust Treatment System 13.2-12 D 3.2.6 Venting or Purging 13.2-14 D 3.3 Instrumentation I 3.3-1 D 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation 1 3.3-1 D 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation 1 3.3-7 D 3.4 Radioactive Effluents Total Dose 13.4-1 D 3.5 Radiological Environmental Monitoring 1 3.5-1 D 3.5.1 Monitoring Program 13.5-1 D 3.5.2 Land Use Census 13.5-13 D 3.5.3 Interlaboratory Comparison Program 1 3.5-16 BASES IB 3.1-0 B 3.1 Radioactive Liquid Effluents IB 3.1-1 B 3.1.1 Liquid Effluents Concentration IB 3.1-1 B 3.1.2 Liquid Effluents Dose IB 3.1-2 B 3.1.3 Liquid Radwaste Treatment System IB 3.1-3 ii Unit 2 Revision 22 December 2001

TABLE OF CONTENTS (Cont)

PAGE Radioactive Gaseous Effluents I B 3.2-1 B 3.2 I B 3.2-1 B 3.2.1 Gaseous Effluents Dose Rate I B 3-2-2 B 3.2.2 Gaseous Effluents Noble Gas Dose B 3.2.3 Gaseous Effluents Dose - Iodine-131, Iodine-133, Tritium, LB 3.2-3 and Radioactive Material in Particulate Form IB 3.2-5 B 3.2.4 Gaseous Radwaste Treatment System LB 3.2-6 B 3.2.5 Ventilation Exhaust Treatment System L B 3.2-7 B 3.2.6 Venting or Purging L B 3.3-1 B3.3 Instrumentation 1 B 3.3-1 B 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation IB 3.3-2 B 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation 1 B 3.4-1 B3.4 Radioactive Effluents Total Dose 1B 3.5-1 B3.5 Radiological Environmental Monitoring IB 3.5-1 B 3.5.1 Monitoring Program LB 3.5-2 B 3.5.2 Land Use Census LB 3.5-3 B 3.5.3 Interlaboratory Comparison Program ADMINISTRATIVE CONTROLS 1 4.0-1 SECTION 4.0 Reporting Requirements 14.1-1 D4.1 1 4.1-1 D 4.1.1 Special Reports D4.2 Major Changes to Liquid, Gaseous and Solid Radwaste 1 4.2-1 Treatment Systems Unit 2 iii Revision 22 December 2001

TABLE OF CONTENTS (Cont)

QIi'PTflATV .gTJRIPCT REF SECTION PAGE Q17rTMV VTIR JrFCT PART II - CALCULATIONAL METHODOLOGIES II1 1.0 LIQUID EFFLUENTS 112 1.1 Liquid Effluent Monitor Alarm Setpoints 112 1.1.1 Basis 3.1.1 112 1.1.2 Setpoint Determination Methodology 3.3.1 112 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint 112 1.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculations II 5 1.1.2.3 Service Water and Cooling Tower Blowdown Effluent Radiation Alarm Setpoint II 6 1.2 Liquid Effluent Concentration Calculation 3.1.1 117 DSR 3.1.1.2 1.3 Liquid Effluent Dose Calculation Methodology 3.1.2 II 8 DSR 3.1.2.1 1.4 Liquid Effluent Sampling Representativeness Table D 3.1.1-1 119 note b 1.5 Liquid Radwaste System Operability 3.1.3 1110 DSR 3.1.3.1 B 3.1.3 GASEOUS EFFLUENTS II 11 2.0 1111 II 11 2.1 Gaseous Effluent Monitor Alarm Setpoints 2.1.1 Basis 3.2.1 2.1.2 Setpoint Determination Methodology Discussion 3.3.2 2.1.2.1 Stack Noble Gas Detector Alarm Setpoint Equation 1112 2.1.2.2 Vent Noble Gas Detector Alarm Setpoint Equation 1113 2.1.2.3 Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation 1114 2.2 Gaseous Effluent Dose Rate Calculation Methodology 3.2.1 1115 II 15 2.2.1 X/Q and W, - Dispersion Parameters for Dose Rate, Table D 3-23 2.2.2 Whole Body Dose Rate Due to Noble Gases DLCO 3.2.1.a 1116 DSR 3.2.1.1 2.2.3 Skin Dose Rate Due to Noble Gases DLCO 3.2.1.a 1117 DSR 3.2.1.1 iv Unit 2 Revision 22 December 2001

TABLE OF CONTENTS (Cont)

T

.SPC'TIflV , TR TFCT RPFSFCTUW PAGF RprTiniv REFR rTTQ7V PAr-F 2.2.4 Organ Dose Rate Due to 1-131, 1-133, Tritium and DLCO 3.2.1 .b Particulates with half-lives greater than 8 days DSR 3.2.1.2 II 18 2.3 Gaseous Effluent Dose Calculation Methodology 3.2.2 1119 3.2.3 3.2.5 2.3.1 W. and -Wv - Dispersion Parameters For Dose, Table D 3-23 1119 2.3.2 Gamma Air Dose Due to Noble Gases 3.2.2 1120 DSR 3.2.2.1 2.3.3 Beta Air Dose Due to Noble Gases 3.3.2 1120 2.3.4 Organ Dose Due to 1-13 1, 1-133, Tritium and Particulates 3.2.3 with half-lives 3.2.5 DSR 3.2.3.1 DSR 3.2.5.1 1120 2.4 1-133 and 1-135 Estimation 1121 2.5 Isokinetic Sampling 1121 2.6 Use of Concurrent Meteorological Data vs. Historical Data 1121 2.7 Gaseous Radwaste Treatment System Operation 3.2.4 1121 2.8 Ventilation Exhaust Treatment System Operation 3.2.5 1122 3.0 URANIUM FUEL CYCLE 3.4 1123 3.1 Evaluation of Doses From Liquid Effluents DSR 3.1.2.1 1124 3.2 Evaluation of Doses From Gaseous Effluents DSR 3.2.2.1 II 26 3.3 Evaluation of Doses From Direct Radiation DSR 3.2.3.1 1126 3.4 Doses to Members of the Public Within the Site Boundary 4.1 1126 4.0 ENVIRONMENTAL MONITORING PROGRAM 3.5 1129 4.1 Sampling Stations 3.5.1 1129 DSR3.5.1.1 4.2 Interlaboratory Comparison Program DSR 3.5.3.2 1129 4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements 1130 V Unit 2 Revision 22 December 2001

TABLE OF CONTENTS (Cont)

SECTION SUBJECT REF SECTION PA[GE Liquid Dose Factor Derivation 1165 Appendix A Plume Shine Dose Factor Derivation 1168 Appendix B Appendix C Dose Parameters for Iodine 131 and 133, Particulates and Tritium 1172 Appendix D Diagrams of Liquid and Gaseous Radwaste Treatment Systems and Monitoring Systems II 82 11105 Appendix E Nine Mile Point On-Site and Off-Site Maps vi Unit 2 Revision 22 December 2001

LIST OF TABLES PART I - RADIOLOGICAL EFFLUENT CONTROLS TABLE NO TITLE PAGE D 3.1.1-1 Radioactive Liquid Waste Sampling and Analysis 1 3.1-2 D 3.2.1-1 Radioactive Gaseous Waste Sampling and Analysis 1 3.2-2 D 3.3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation 1 3.3-6 D 3.3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation 1 3.3-13 D 3.5.1-1 Radiological Environmental Monitoring Program 1 3.5-6 D 3.5.1-2 Reporting Levels for Radioactivity Concentrations in Environmental Samples 13.5-10 D 3.5.1-3 Detection Capabilities for Environmental Sample Analyses 1 3.5-11 vii Unit 2 Revision 22 December 2001

LIST OF TABLES (Cont)

PART II- CALCULA TIONAL METHODOLOGIES TABLE NO TITLE PAGE D2-1 Liquid Effluent Detector Response 1132 D 2-2 thru D 2-5 Aiat Values - Liquid Effluent Dose Factor II 33 D3-1 Offgas Pretreatment Detector Response 1137 D 3-2 Finite Plume - Ground Level Dose II 38 Factors from an Elevated Release D 3-3 Immersion Dose Factors 1139 D 3-4 thru D 3-22 Dose And Dose Rate Factors, Ri 1140 D 3-23 Dispersion Parameters at Controlling 1159 Locations, X/Q, Wv and W. Values D 3-24 Parameters For the Evaluation of Doses to 1160 Real Members of the Public From Gaseous And Liquid Effluents D5.1 Radiological Environmental Monitoring 1161 Program Sampling Locations viii Unit 2 Revision 22 December 2001

LIST OF FIGURES FIGURE NO TITLE PAGE D 1.0-1 Site Area and Land Portion of Exclusion Area Boundaries 11.0-4 D 5.1-1 Nine Mile Point On-Site Map 11106 D 5.1-2 Nine Mile Point Off-Site Map (page 1 of 2) 11107 D 5.1-2 Nine Mile Point Off-Site Map (page 2 of 2) 11108 I ix Unit 2 Revision 22 December 2001

INTRODUCTION The OFFSITE DOSE CALCULATION MANUAL (ODCM) is a supporting document of the Technical Specifications Section 5.5.1. The previous Limiting Conditions for Operation that were contained in the Radiological Effluent Technical Specifications are now transferred to the ODCM as Radiological Effluent Controls. The ODCM contains two parts: Radiological Effluent Controls, Part I; and Calculational Methodologies, Part II. Radiological Effluent Controls, Part 1, includes the following: (1)

The Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification 5.5.1 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 5.6.2 and 5.6.3. Calculational Methodologies, Part II, describes the methodology and parameters to be used in the calculation of liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints and the calculation of offsite doses due to radioactive liquid and gaseous effluents. The ODCM also contains a list and graphical description of the specific sample locations for the radiological environmental monitoring program, and liquid and gaseous radwaste treatment system configurations.

The ODCM follows the methodology and models suggested by NIJREG-0133 and Regulatory Guide 1.109, Revision 1. Simplifying assumptions have been applied in this manual where applicable to provide a more workable document for implementing the Radiological Effluent Control requirements; this simplified approach will result in a more conservative dose evaluation for determining compliance with regulatory requirements.

The ODCM will be maintained for use as a reference and training document of accepted methodologies and calculations. Changes to the calculation methods or parameters will be incorporated into the ODCM to assure that the ODCM represents the present methodology in all applicable areas. Any changes to the ODCM will be implemented in accordance with Section 5.5.1 of the Technical Specifications.

x Unit 2 Revision 22 December 2001

PART I - RADIOLOGICAL EFFLUENT CONTROLS Unit 2 Revision 22 I December 2001

Definitions 1.0 PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 1.0 DEFINITIONS Unit 2 Revision 22 December 2001 11.0-0

Definitions 1.0 1.0 DEFINITIONS

NOTE ---------------------------------------------------

Technical Specifications defined terms and the following additional defined terms appear in capitalized type and are applicable throughout these specifications and bases.

TERM DEFINITION GASEOUS A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system RADWASTE designed and installed to reduce radioactive gaseous effluents by collecting TREATMENT offgases from the main condenser evacuation system and providing for SYSTEM delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

MEMBER(S) MEMBER(S) OF THE PUBLIC shall include all persons who are not OF THE PUBLIC occupationally associated with the Nine Mile Point Nuclear Station and James A. FitzPatrick Nuclear Power Plant. This category does not include employees of owners and operators of the Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant, their contractors or 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 Nine Mile Point Nuclear Station and James A. FitzPatrick Nuclear Power Plant.

MILK SAMPLING A MILK SAMPLING LOCATION is a location where 10 or more head of LOCATION milk animals are available for collection of milk samples.

OFFSITE DOSE The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain CALCULATION the current methodology and parameters used in the calculation of offsite MANUAL doses that result from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the environmental radiological monitoring program.

The ODCM shall also contain: (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Program required by Specification 5.5.1 of Technical Specifications and, (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 5.6.2 and 5.6.3.

(continued)

Unit 2 Revision 22 11.0-1 December 2001

Definitions 1.0 1.0 DEFINITIONS (continued)

TERM DEFINITION PURGE PURGE and PURGING shall be the controlled process of discharging air PURGING or gas from a confinement to maintain temperature, pressure, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

REPORTABLE A REPORTABLE EVENT shall be any of those conditions specified in EVENT 10 CFR 50.73.

SITE BOUNDARY The SITE BOUNDARY shall be that line around the Nine Mile Point Nuclear Station beyond which the land is not owned, leased or otherwise controlled by the owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant. See Figure D 1.0-1.

SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

UNRESTRICTED An UNRESTRICTED AREA shall be any area at or beyond the SITE AREA BOUNDARY, access to which is not controlled by the owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant 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 for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION A VENTILATION EXHAUST TREATMENT SYSTEM shall be any EXHAUST system designed and installed to reduce gaseous radioiodine or radioactive TREATMENT material in particulate form in effluents by passing ventilation or vent SYSTEM exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered safety features (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

(continued)

Unit 2 Revision 22 11.0-2 December 2001

Definitions 1.0 1.0 DEFINITIONS (continued)

TERM DEFINITION VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

Unit 2 Revision 22 11.0-3 December 2001

Definitions NMP2 1.0 LIQUID

,DISCHARGE R I 0 0 N TA AK E Lycoming SITE AREA AND LAND PORTION OF EXCLUSION AREA BOUNDARIES 1A/,4 '12 SCALE -MIL ES Niagara Mohawk Power Corporation retains ownership in certain transmission line and switchyard facilities within the exclusion area boundary. Access and usage are controlled by Nine Mile Point Nuclear Station, LLC by agreement.

Figure D 1.0-1 (Page 1 of 1)

Site Area and Land Portion of Exclusion Area Boundaries Unit 2 Revision 22 1 1.0-4 December 2001

PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 3.0 APPLICABILITY Unit 2 Revision 22 13.0-0 December 2001

Applicability 3.0 3.0 APPLICABILITY The Offsite Dose Calculation Manual (ODCM) Specifications are contained in Section 3.0 of Part I. They contain operational requirements, Surveillance Requirements, and reporting requirements.

Additionally, the Required Actions and associated Completion Times for degraded Conditions are specified. The format is consistent with the Technical Specifications (Appendix A to the NMP2 Operating License).

The rules of usage for the ODCM Specification are the same as those for the Technical Specifications. These rules are found in Technical Specifications Sections 1.2, "Logical Connectors," 1.3, "Completion Times," and 1.4, "Frequency."

The ODCM Specifications are subject to Technical Specifications Section 3.0, "Limiting Condition for Operation (LCO) Applicability and Surveillance Requirement (SR) Applicability,"

with the following exceptions:

1. LCO 3.06, regarding support/supported system ACTIONS is not applicable to ODCM Specifications.

2. LCO 3.0.7, regarding allowances to change specified Technical Specifications is not applicable to ODCM Specifications.

3. Section 3.0 requirements are not required when so stated in notes within individual specifications.

Unit 2 Revision 22 13.0-1 December 2001

Liquid Effluents Concentration D 3.1.1 D 3.1 RADIOACTIVE LIQUID EFFLUENTS D 3.1.1 Liquid Effluents Concentration DLCO 3.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (Figure D 1.0-1) shall be limited to:

a. Ten times the concentration specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases; and

b. 2 x 10-4 ptCi/ml total activity concentration for dissolved or entrained noble gases.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A. 1 Initiate action to restore Immediately radioactive material concentration to within limits.

released in liquid effluents to UNRESTRICTED AREAS exceeds limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.1.1 Perform radioactive liquid waste sampling and activity In accordance with analysis. Table D 3.1.1-1 DSR 3.1.1.2 Verify the results of the DSR 3.1.1.1 analyses to In accordance with assure that the concentrations at the point of release Table D 3.1.1-1 are maintained within the limits of DLCO 3.1.1.

Unit 2 Revision 22 13.1-1 December 2001

Liquid Effluents Concentration D 3.1.1 Table D 3.1.1-1 (Page I of 2)

Radioactive Liquid Waste Sampling and Analysis SAMPLE LOWER SAMPLE LIMITLOWER OF SAMPLE DETECTION SAMPLE SAMPLE ANALYSIS (LLD) (a)

TYPE FREQUENCY FREQUENCY ANALYSIS LIQUID RELEASE TYPE Grab Sample Each Batch (g) Each Batch (g) Principal 5 x 10,7 ý.Ci/ml

1. Batch Waste Release Gamma Emitters Tanks (b) (c)

a. 2LWS-TK4A 1-131 1 x 10-6 t.Ci/ml

b. 2LWS-TK4B

c. 2LWS-TK5A

d. 2LWS-TK5B 31 days Dissolved and 1 x 10.5 .Ci/ml Grab Sample One batch/31 days (g) Entrained Gases (gamma emitters)

Each batch (g) 31 days H-3 1 x 10.1 .Ci/ml Proportional Composite of grab samples Gross Alpha 1 x 10.7 ý.Ci/ml (d)

Proportional Each batch (g) 92 days Sr-89 5 x 10"' ,LCi/ml Composite of grab samples (d)

Sr-90 5 x 10. 4Ci/ml Fe-55 I x 10-6 -Ci/ml Principal 5 x 10-7 RCi/ml

2. Continuous Releases Grab Sample 31 days (e) 31 days (e) Gamma Emitters (c)

a. Service Water Effluent A 1-131 1 x 10-6 4Ci/ml Grab Sample 31 days (e) 31 days (e)

b. Service Water Effluent B

c. Cooling Tower Blowdown 31 days (e) Dissolved and 1 x 10-5 ýLCi/ml Grab Sample 31 days (e)

Entrained Gases (gamma emitters)

Grab Sample 31 days (e) H-3 1 x 10.1 p.Ci/ml 31 days (e)

Grab Sample 31 days (e) 31 days (e) Gross Alpha I x l1-Ol.Ci/ml Grab Sample 92 days (e) 92 days (e) Sr-89 5 x 10a8 RCi/ml Grab Sample 92 days (e) 92 days (e) Sr-90 5 x 108 4sCi/ml 92 days (e) 92 days (e) Fe-5 5 I x 10"6 l.Ci/ml Grab Sample Grab Sample 31 days (f) 31 days (f) Principal 5x 10"7 pCi/ml

3. Continuous Release Gamma Emitters (c)

Auxiliary Boiler Pump Seal and 92 days (f) H-3 I x l0"1 pLCi/ml Sample Cooling Grab Sample 92 days (f)

Discharge (Service Water)

Unit 2 Revision 22 13.1-2 December 2001

Liquid Effluents Concentration D 3.1.1 Table D 3.1.1-1 (Page 2 of 2)

Radioactive Liquid Waste Sampling and Analysis (a) The LLD is defined as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

(4.66) (Sb)

LLD =

(E) (V) (2.22x10') (Y) e-5:t where:

LLD The before-the-fact lower limit of detection (piCi per unit mass or volume),

Sb - The standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E = The counting efficiency (counts per disintegration),

V = The sample size (units of mass or volume),

2.22 x 106 = The number of disintegrations per minute per L+/-Ci, Y The fractional radiochemical yield, when applicable, X - The radioactive decay constant for the particular radionuclide (sec"), and At The elapsed time between the midpoint of sample collection and the time of counting (seconds).

Typical values of E, V, Y, and At should be used in the calculation.

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

and (b) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, then thoroughly mixed by the method described in Part II, Section 1.4 to assure representative sampling.

(c) The principal gamma emitters for which the LLD applies include 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"j.+/-Ci/mi. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be in analyzed and reported in the Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3 in the format outlined RG 1.21, Appendix B, Revision 1,June 1974.

in which (d) A composite sample is one in which the-quantity of liquid sampled is proportional to the quantity of liquid waste discharged and the method ofsampling employed results in a specimen that is representative of the liquids released.

no longer (e) If the alarm setpoint of the effluent monitor is exceeded, the frequency of sampling shall be increased to daily until the condition exists. Frequency of analysis shall be increased to daily for principal gamma emitters and an incident composite for H-3, gross alpha, Sr 89, Sr-90, and Fe-55.

(f) If the alarm setpoint of Service Water Effluent Monitor A and/or B is exceeded, the frequency of sampling shall be increased to daily until the condition no longer exists. Frequency ofanalysis shall be increased to daily for principal gamma emitters and an incident composite for H-3, gross alpha, Sr-89, Sr-90, and Fe-55.

(g) Complete prior to each release.

Unit 2 Revision 22 13.1-3 December 2001

Liquid Effluents Dose D 3.1.2 D 3.1 RADIOACTIVE LIQUID EFFLUENTS D 3.1.2 Liquid Effluents Dose DLCO 3.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials released in liquid effluents from each unit to UNRESTRICTED AREAS (Figure D 1.0-1) shall be limited to:

a. _-1.5 mrem to the whole body and < 5 mrem to any organ during any calendar quarter; and

b. < 3 mrem to the whole body and < 10 mrem to any organ during any calendar year.

APPLICABILITY: At all times.

ACTIONS N O TE S-------------------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose to a A. 1 Prepare and submit to the 30 days MEMBER OF THE NRC, pursuant to D 4.1.1, a PUBLIC from the release Special Report that of radioactive materials in (1) Identifies the cause(s) for liquid effluents to exceeding the limit(s) and UNRESTRICTED AREAS (2) Defines the corrective exceeds limits, actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with DLCO 3.1.2.

(continued)

Unit 2 Revision 22 13.1-4 December 2001

Liquid Effluents Dose D 3.1.2 ACTTONS (continued CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose to a B. 1 Calculate the annual dose to a Immediately MEMBER OF THE MEMBER OF THE PUBLIC PUBLIC from the release which includes contributions of radioactive materials in from direct radiation from the liquid effluents exceeds 2 units (including outside storage times the limits, tanks, etc.).

AND B.2 Verify that the limits of DLCO Immediately 3.4 have not been exceeded.

C. Required Action B.2 and C. 1 Prepare and submit to the 30 days Associated Completion NRC, pursuant to D 4.1.1, a Special Report, as defined in time not met. 10 CFR 20.2203 (a)(4), of Required Action A. 1 shall also include the following:

(1) The corrective action(s) to be taken to prevent recurrence of exceeding the limits of DLCO 3.4 and the schedule for achieving conformance, (2) An analysis that estimates the dose to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s), and (3) Describes the levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations.

Unit 2 Revision 22 1 3.1-5 December 2001

Liquid Effluents Dose D 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.2.1 Determine cumulative dose contributions from liquid 31 days effluents for the current calendar quarter and the current calendar year.

Unit 2 Revision 22 13.1-6 December 2001

Liquid Radwaste Treatment System D 3.1.3 D 3.1 RADIOACTIVE LIQUID EFFLUENTS D 3.1.3 Liquid Radwaste Treatment System DLCO 3.1.3 The liquid radwaste treatment system shall be OPERABLE.

APPLICABILITY: At all times.

ACTIONS

NOTES -----------------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Radioactive liquid waste A. 1 Prepare and submit to the 30 days being discharged without NRC, pursuant to D 4.1.1, a treatment. Special Report that includes:

(1) An explanation of why AND liquid radwaste was being discharged without Projected doses due to the treatment, identification of liquid effluent, from the any inoperable equipment unit, to UNRESTRICTED or subsystems, and the AREAS would exceed reason for the inoperability, 0.06 mrem to the whole (2) Action(s) taken to restore body or 0.2 mrem to any the inoperable equipment to organ in a 31 day period. OPERABLE status, and (3) Summary description of AND action(s) taken to prevent a recurrence.

Any portion of the liquid radwaste treatment system not in operation.

Unit 2 Revision 22 13.1-7 December 2001

Liquid Radwaste Treatment System D 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.3.1 -------------------- NOTE ----------------

Only required to be met when liquid radwaste treatment systems are not being fully utilized.

Project the doses due to liquid effluents from each 31 days unit to UNRESTRICTED AREAS.

Unit 2 Revision 22 1 3.1-8 December 2001

Gaseous Effluents Dose Rate D 3.2.1 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.1 Gaseous Effluents Dose Rate DLCO 3.2.1 The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (Figure D 1.0-1) shall be limited to:

a. For noble gases, _ 500 mrem/yr to the whole body and

< 3000 mrem/yr to the skin and

b. For 1-131, 1-133, H-3 and all radionuclides in particulate form with half-lives > 8 days, _<1500 mrem/yr to any organ.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. The dose rate(s) at or A. 1 Restore the release rate to Immediately beyond the SITE within the limit, BOUNDARY due to radioactive gaseous effluents exceeds limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.1.1 The dose rate from noble gases in gaseous effluents In accordance with shall be determined to be within the limits of DLCO Table D 3.2.1-1 3.2.1.a.

DSR 3.2.1.2 The dose rate from 1-131, 1-133, H-3 and all In accordance with Table D 3.2.1-1 radionuclides in particulate form with half-lives

> 8 days in gaseous effluents shall be determined to be within the limits of DLCO 3.2.1.b.

Unit 2 Revision 22 1 3.2-1 December 2001

Gaseous Effluents Dose Rate D 3.2.1 Table D 3.2.1-1 (Page 1 of 2)

Radioactive Gaseous Waste Sampling and Analysis SAMPLE LOWER LIMIT OF DETECTION GASEOUS RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE (LLD) (a)

TYPE FREQUENCY FR1EQUENCY ANALYSIS TYPE Grab Sample Each Purge (h) Principal 1x 10-4 pCi/ml

1. Containment (b) Gamma Emitters (c)

Each Purge H-3 (oxide) I X 10"6 1 Ci/ml Each Purge Principal 1 x 10.4 i.Ci/ml Gamma Emitters (c) 4 31 days (d) Principal I x 10 ýLCi/ml

2. Main Stack, Grab Sample 31 days (d)

Gamma Emitters Radwaste/Reactor (c)

Building Vent 6

31 days (e) H-3 (oxide) lx I0" jCi/mi Grab Sample 31 days (e)

Charcoal Continuous (t) 7 days (g) 1-131 Ix 10-12 ýiCi/ml Sample Particulate Continuous (f) 7 days (g) Principal 1 x 10"n1 pCi/ml Sample Gamma Emitters (c)

Gross Alpha lx 10-" ICi/mI Composite Continuous (f) 92 days Sr-89 1 x 10"n p.Ci/mi Particulate Sample Sr-90 I x 10"1 iCi/ml See the notes on the next page.

Unit 2 Revision 22 13.2-2 December 2001

Gaseous Effluents Dose Rate D 3.2.1 Table D 3.2.1-1 (Page 2 of 2)

Radioactive Gaseous Waste Sampling and Analysis yield a net count, above system background, (a) The LLD is defined as the smallest concentration of radioactive material in a sample that will blank observation represents a "real" a

that will be detected with 95% probability with only 5% probability of falsely concluding that signal.

For a particular measurement system, which may include radiochemical separation:

LLD (Sb)

-(4.66)

(E) (V) (2.22x,06) MYe -'At where:

LLD = The before-the-fact lower limit of detection (1tCi per unit mass or volume),

blank sample as Sb = The standard deviation ofthe background counting rate or of the counting rate of a appropriate (counts per minute),

E The counting efficiency (counts per disintegration),

V - The sample size (units of mass or volume),

2.22 x 10' The number of disintegrations per minute per 40Ci, Y = The fractional radiochemical yield, when applicable, 1

X = The radioactive decay constant for the particular radionuclide (sec ), and (seconds).

At = The elapsed time between the midpoint of sample collection and the time of counting Typical values of E, V, Y, and At should be used in the calculation.

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

and analysis during actual PURGE is (b) Sample and analysis before PURGE is used to determine permissible PURGE rates. Sample used for offsite dose calculations.

The principal gamma emitters for which the LLD applies include the following radionuclides:1-131, Kr-87, Kr-88, Xe-133, Xe-133m, (c) Cs-134, Cs-137, Ce-141 and Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99,are to be considered. Other gamma peaks that Ce-144 in iodine and particulate releases. This list does not mean that only these nuclides reported in the Radioactive Effluent Release Report are identifiable, together with those ofthe above nuclides, shall also be analyzed and B, Revision 1, June 1974.

pursuant to Technical Specification 5.6.3 in the format outlined in RG 1.21, Appendix sampling and analysis shall also be performed (d) If the main stack or reactor/radwaste building isotopic monitor is not OPERABLE,monitor.

following shutdown, startup, or when there is an alarm on the offgas pretreatment system when fuel is offloaded until stable H-3 (e) H-3 grab samples shall be taken once every 7 days from the reactor/radwaste ventilation release levels can be demonstrated.

for the time period covered by each dose or dose rate (0) The ratio of the sample flow rate to the sampled stream flow rate shall be known calculation made in accordance with DLCO 3.2.1.b and DLCO 3.2.3.

setpoint, the iodine and particulate (g) When the release rate of the main stack or reactor/radwaste building vent exceeds its alarm release rates. The analysis shall be done device shall be removed and analyzed to determine the changes in iodine and particulate collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the release no longer exceeds the alarm setpoint. When samples corresponding LLDs may be increased by a factor of 10.

(h) Complete prior to each release.

Unit 2 Revision 22 I 3.2-3 December 2001

Gaseous Effluents Noble Gas Dose D 3.2.2 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.2 Gaseous Effluents Noble Gas Dose DLCO 3.2.2 The air dose from noble gases released in gaseous effluents from each unit to areas at or beyond the SITE BOUNDARY (Figure D 1.0-1) shall be limited to:

a. During any calendar quarter: < 5 mrad for gamma radiation and

< 10 mrad for beta radiation and

b. During any calendar year: < 10 mrad for gamma radiation and

< 20 mrad for beta radiation.

APPLICABILITY: At all times.

ACTIONS

N O TE S ----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. The air dose at or beyond A. 1 Prepare and submit to the 30 days the SITE BOUNDARY NRC, pursuant to D 4.1.1, a due to noble gases released in gaseous ettluents Special Report that exceeds limits. (1) Identifies the cause(s) for exceeding the limit(s) and (2) Defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with DLCO 3.2.2.

(continued)

Unit 2 Revision 22 December 2001 1 3.2-4

Gaseous Effluents Noble Gas D Dose 3.2.2 ACTIONS (continued)

REQUIRED ACTION COMPLETION CONDITION TIME B.I Calculate the annual dose to a Immediately B. Calculated dose to a MEMBER OF THE PUBLIC MEMBER OF THE which includes contributions PUBLIC from the release from direct radiation from the of radioactive materials in units (including outside storage gaseous effluents due to tanks, etc.).

noble gases exceeds 2 times the limits.

AND Immediately B.2 Verify that the limits of DLCO 3.4 have not been exceeded.

C. 1 Special Report, as defined in 30 days C. Required Action B.2 and 10 CFR 20.2203 (a)(4), of Associated Completion time not met. Required Action A. 1 shall also include the following:

(1) The corrective action(s) to be taken to prevent recurrence of exceeding the limits of DLCO 3.4 and the schedule for achieving conformance, (2) An analysis that estimates the dose to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes.

the release(s), and (3) Describes the levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations.

Unit 2 Revision 22 December 2001 13.2-5

Gaseous Effluents Noble Gas D Dose 3.2.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.2.1 Determine cumulative dose contributions for the 31 days current calendar quarter and current calendar year.

Unit 2 Revision 22 December 2001 1 3.2-6

Gaseous Effluents Dose 131, 1-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.3 Gaseous Effluents Dose 131, 1-133, H-3 and Radioactive Material in Particulate Form DLCO 3.2.3 The dose to a MEMBER OF THE PUBLIC from 1-131, 1-133, H-3, and all radioactive material in particulate form with half-lives > 8 days in gaseous effluents released, from each unit, to areas at or beyond the SITE BOUNDARY (Figure D 1.0-1) shall be limited to:

a. During any calendar quarter: _<7.5 mrem to any organ and

b. During any calendar year: < 15 mrem to any organ.

APPLICABILITY: At all times.

ACTIONS

N OTE S ----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. The dose from 1-131, 1-133, A. 1 Prepare and submit to the NRC, 30 days H-3 and radioactive material pursuant to D 4.1.1, a Special in particulate form with half- Report that lives > 8 days released in (1) Identifies the cause(s) for gaseous effluents at or beyond exceeding the limit(s) and the SITE BOUNDARY (2) Defines the corrective actions exceeds limits, that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with DLCO 3.2.3.

(continued)

Unit 2 Revision 22 December 2001 1 3.2-7

Gaseous Effluents Dose 131, 1-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose to a B. 1 Calculate the annual dose to a Immediately MEMBER OF THE PUBLIC MEMBER OF THE PUBLIC from the release of radioactive which includes contributions from materials in gaseous effluents direct radiation from the units exceeds 2 times the limits. (including outside storage tanks, etc.).

AND B.2 Verify that the limits of DLCO Immediately 3.4 have not been exceeded.

C. Required Action B.2 and C. 1 Special Report, as defined in 10 30 days Associated Completion time CFR 20.2203 (a)(4), of Required not met. Action A. 1 shall also include the following:

(1)The corrective action(s) to be taken to prevent recurrence of exceeding the limits of DLCO 3.4 and the schedule for achieving conformance, (2)An analysis that estimates the dose to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s), and (3)Describes the levels of radiation and concentrations of radioactive material involved and the cause of the exposure levels or concentrations.

Unit 2 Revision 22 December 2001 13.2-8

Gaseous Effluents Dose 131, 1-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.3.1 Determine cumulative dose contributions for the 31 days current calendar quarter and current calendar year for 1-131, 1-133, H-3 and radioactive material in particulate form with half-lives > 8 days.

Unit 2 Revision 22 1 3.2-9 December 2001

Gaseous Radwaste Treatment System D 3.2.4 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.4 Gaseous Radwaste Treatment System DLCO 3.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM shall be in operation.

APPLICABILITY: Whenever the main condenser air ejector system is in operation.

ACTIONS mNOTES -----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. The gaseous radwaste from A. 1 Restore treatment of gaseous 7 days the main condenser air radwaste effluent.

ejector system is being discharged without treatment.

B. Required Action and B.1 Prepare and submit to the 30 days associated Completion NRC, pursuant to D 4.1.1, a Time not met. Special Report that includes the following:

(1) Identification of any inoperable equipment or subsystems and the reason for the inoperability, (2) Action(s) taken to restore the inoperable equipment to OPERABLE status, and (3) Summary description of action(s) taken to prevent a recurrence.

Unit 2 Revision 22 December 2001 1 3.2-10

Gaseous Radwaste Treatment System D 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.4.1 Check the readings of the relevant instruments to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensure that the GASEOUS RADWASTE TREATMENT SYSTEM is functioning.

Unit 2 Revision 22 December 2001 13.2-11

Ventilation Exhaust Treatment System D 3.2.5 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.5 Ventilation Exhaust Treatment System DLCO 3.2.5 The VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE.

APPLICABILITY: At all times.

ACTIONS

NO TE S ----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable. --------------

REQUIRED ACTION COMPLETION CONDITION TIME A.1I Prepare and submit to the 30 days A. The radioactive gaseous waste is being discharged NRC, pursuant to D 4 .1.1, a without treatment. Special Report that includes the following:

AND (1) Identification of any inoperable equipment or Projected doses in 31 days subsystems and the reason for the inoperability, from iodine and particulate releases, from each unit, to (2) Action(s) taken to restore the inoperable equipment to areas at or beyond the SITE BOUNDARY (see OPERABLE status, and Figure D 1.0-1) would (3) Summary description of action(s) taken to prevent a exceed 0.3 mrem to any recurrence.

organ of a MEMBER OF THE PUBLIC.

Unit 2 Revision 22 December 2001 13.2-12

Ventilation Exhaust Treatment System D 3.2.5 SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY I

DSR 3.2.5.1 N O TE

-------- O TE--------------

Only required to be met when the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.

31 days Project the doses from iodine and particulate releases from each unit to areas at or beyond the SITE BOUNDARY.

Unit 2 Revision 22 13.2-13 December 2001

Venting or Purging D 3.2.6 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.6 Venting or Purging DLCO 3.2.6 VENTING or PURGING of the drywell and/or suppression chamber shall be through the standby gas treatment system.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

N OTES ----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable. .

COMPLETION CONDITION REQUIRED ACTION TIME Suspend all VENTING and Immediately A. VENTING or PURGING A. 1 PURGING of the drywell of the dryowell and/or and/or suppression chamber.

SUDDression chamber not through the standby gas treatment system.

Unit 2 Revision 22 December 2001 13.2-14

Venting or Purging D 3.2.6 SURVEILLANCE REQUIREMENTS T

SURVEILLANCE FREQUENCY I

DSR 3.2.6.1 The drywell and/or suppression chamber shall be Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months determined to be aligned for VENTING or PURGING before start of through the standby gas treatment system. VENTING or PURGING AND 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter during VENTING or PURGING Unit 2 Revision 22 13.2-15 December 2001

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 D 3.3 INSTRUMENTATION D 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation DLCO 3.3.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table D 3.3.1-1 shall be OPERABLE with:

a. The minimum OPERABLE channel(s) in service.

b. The alarm/trip setpoints set to ensure that the limits of DLCO 3.1.1 are not exceeded.

APPLICABILITY: According to Table D 3.3.1-1.

ACTIONS

NO TE S ----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

3. Separate condition entry is allowed for each channel.

I. r CONDITION REQUIRED ACTION COMPLETION TIME Immediately A. Liquid effluent monitoring A. 1 Suspend the release of instrumentation channel radioactive liquid effluents alarm/trip setpoint less monitored by the affected conservative than required. channel.

OR A.2 Immediately Declare the channel inoperable.

OR Change the setpoint so it is Immediately A.3 acceptably conservative.

(continued)

Unit 2 Revision 22 December 2001 I 3.3-1

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 ACTIONS (continued)

REQUIRED ACTION COMPLETION CONDITION TIME B. One or more required B. 1 Enter the Condition referenced Immediately channels inoperable. in Table D 3.3.1-1 for the channel.

AND B.2 Restore inoperable channel(s) 30 days to OPERABLE status.

C. 1 Analyze at least 2 independent Prior to initiating a C. As required by Required release Action B.1 and referenced samples in accordance with in Table D 3.3.1-1. Table D 3.1.1-1.

AND C.2 -------- NOTE ---------------

Verification Action will be performed by at least 2 separate technically qualified members of the facility staff.

Prior to initiating a Independently verify the release release rate calculations and discharge line valving.

D. 1 Collect and analyze grab 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. As required by Required Action B. 1 and referenced samples for radioactivity at a AND in Table D 3.3.1-1. limit of detection of at least 5 x 10-7 pCi/ml.

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter (continued)

Unit 2 Revision 22 December 2001 1 3.3-2

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME E. As required by Required E. 1 --------------- NOTE------

Action B. 1 and referenced Pump performance curves in Table D 3.3.1-1. generated in place may be used to estimate flow.

Estimate the flow rate during 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months actual releases.

AND Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter F. 1 Estimate tank liquid level. Immediately F. As required by Required Action B. I and referenced AND in Table D 3.3.1-1.

During liquid additions to the tank G. 1 Explain in the next In accordance with G. Required Action B.2 and Radioactive Radioactive Effluent Release associated Completion Effluent Release Time not met. Report why the inoperability Report was not corrected in a timely manner.

H. 1 Suspend liquid effluent Immediately H. Required Action and releases monitored by the associated Completion inoperable channel(s).

Time for Condition C, D, or E not met.

1.1 Suspend liquid adctitions to Immediately I. Required Action and the tank monitore A. 1h~thA associated Completion Time inoperable channe i(s).

for Condition F not met.

Unit 2 Revision 22 December 2001 1 3.3-3

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 SURVEILLANCE REQUIREMENTS

......... ----------------------------------.N O T E -------------------------------------------------------

Refer to Table D 3.3.1-1 to determine which DSRs apply for each function.

SURVEILLANCE FREQUENCY DSR 3.3.1.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DSR 3.3.1.2 Perform CHANNEL CHECK by verifying indication 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on any of flow during periods of release. day on which continuous, periodic, or batch releases are made DSR 3.3.1.3 Perform SOURCE CHECK. Prior to release DSR 3.3.1.4 Perform SOURCE CHECK. 31 days DSR 3.3.1.5 Perform CHANNEL FUNCTIONAL TEST. The 31 days CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the alarm/trip setpoint; and control room alarm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, or instrument controls not set in operate mode.

DSR 3.3.1.6 Perform CHANNEL FUNCTIONAL TEST. 92 days (continued)

Unit 2 Revision 22 1 3.3-4 December 2001

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY DSR 3.3.1.7 Perform CHANNEL FUNCTIONAL TEST. The 184 days CHANNEL FUNCTIONAL TEST shall also demonstrate control room alarm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, or instrument controls not set in operate mode.

DSR 3.3.1.8 Perform CHANNEL CALIBRATION. The initial 18 months CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST), standards that are traceable to NIST standards, or using actual samples of liquid effluents that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

DSR 3.3.1.9 Perform CHANNEL CALIBRATION. 18 months Unit 2 Revision 22 I 3.3-5 December 2001

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 Table D 3.3.1-1 (page I of 1)

Radioactive Liquid Effluent Monitoring Instrumentation APPLICABILITY REQUIRED CONDITIONS OR OTHER CHANNELS REFERENCED SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTION B. 1 REQUIREMENTS Radioactivity Monitors Providing Alarm and Automatic Termination of Release Liquid Radwaste Effluent (a) 1 C DSR 3.3.1.1 Line DSR 3.3.1.3 DSR 3.3.1.5 DSR 3.3.1.8

2. Radioactivity Monitors Providing Alarm but not Providing Automatic Termination of Release

a. Service Water Effluent (a) 1 D DSR 3.3.1.1 Line A DSR 3.3.1.4 DSR 3.3.1.7 DSR 3.3.1.8

b. Service Water Effluent (a) 1 D DSR3.3.1.1 Line B DSR 3.3.1.4 DSR 3.3.1.7 DSR 3.3.1.8

c. Cooling Tower Blowdown (a) 1 D DSR 3.3.1.1 Line DSR 3.3.1.4 DSR 3.3.1.7 DSR 3.3.1.8

3. Flow Rate Measurement Devices

a. Liquid Radwaste (a) 1 E DSR 3.3.1.2 Effluent Line DSR 3.3.1.6 DSR 3.3.1.9

b. Service Water Effluent (a) 1 E DSR 3.3.1.2 Line A DSR 3.3.1.6 DSR 3.3.1.9

c. Service Water Effluent (a) 1 E DSR 3.3.1.2 Line B DSR 3.3.1.6 DSR 3.3.1.9 Cooling Tower (a) 1 E DSR 3.3.1.2

d. DSR 3.3.1.6 Blowdown Line DSR 3.3.1.9 Tank Level Indicating (b) 1 F DSR 3.3.1.1 4.

Devices (c) DSR 3.3.1.6 DSR 3.3.1.9 (a) During releases via this pathway.

(b) During liquid addition to the associated tank.

holding the tank (c) Tanks included in this DLCO are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of such as contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system, temporary tanks.

Unit 2 Revision 22 1 3.3-6 December 2001

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 D 3.3 INSTRUMENTATION D 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation DLCO 3.3.2 The radioactive gaseous effluent monitoring instrumentation channels shown in Table D 3.3.2-1 shall be OPERABLE with:

a. The minimum OPERABLE channel(s) in service.

b. The alarm/trip setpoints set to ensure that the limits of DLCO 3.2.1 are not exceeded.

APPLICABILITY: According to Table D 3.3.2-1.

ACTIONS

N O TE S ----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

3. Separate condition entry is allowed for each channel.

I CONDITION REQUIRED ACTION COMPLETION TIME A. Gaseous effluent A. 1 Suspend the release of Immediately monitoring instrumentation radioactive gaseous effluents channel alarm/trip setpoint monitored by the affected less conservative than channel.

required.

OR A.2 Declare the channel inoperable. Immediately OR A.3 Change the setpoint so it is Immediately acceptably conservative.

(continued)

Unit 2 Revision 22 December 2001 13.3-7

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more channels B. 1 Enter the Condition referenced Immediately inoperable, in Table D 3.3.2-1 for the channel.

AND B.2 Restore inoperable channel(s) 30 days to OPERABLE status.

C. As required by Required C. 1 Place the inoperable channel 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B. 1 and referenced in the tripped condition.

in Table D 3.3.2-1. OR C.2.1 Take grab samples. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND C.2.2 Analyze samples for gross 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time activity, of sampling completion (continued)

Unit 2 Revision 22 December 2001 13.3-8

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 A CTT(ThTV (r't-,,tniipd'

/ .II , .1 vy.,i l.* .,',*

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D. 1 Estimate the flow rate for the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action B. 1 and referenced inoperable channel(s).

in Table D 3.3.2-1. AND Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter E. As required by Required E. 1 Continuously collect samples 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Action B. 1 and referenced using auxiliary sampling in Table D 3.3.2-1. equipment as required in Table D 3.2.1-1.

F. As required by Required F.1.1 Take grab samples. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B. 1 and referenced AND in Table D 3.3.2-1.

Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND Analyze samples for gross 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time F. 1.2 activity with a radioactivity of sampling limit of detection of at least completion I X10-4 ýtCi/ml.

AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months F.2.1 Restore the inoperable channel(s) to OPERABLE status.

OR 14 days F.2.2 In lieu of another required report, prepare and submit to the NRC, pursuant to D 4.1.1, a special report that:

(1) Identifies the cause(s) of the inoperability.

(2) Outlines the action taken and the schedule for restoring the system to OPERABLE status (continued)

Unit 2 Revision 22 December 2001 13.3-9

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action B.2 and G. 1 Explain in the next In accordance with associated Completion Radioactive Effluent Release Radioactive Time not met. Report why the inoperability Effluent Release was not corrected in a timely Report frequency manner.

H Required Action and H. 1 Suspend gaseous effluent Immediately associated Completion releases monitored by the Time for Condition C, D, E inoperable channel(s).

or F not met.

Unit 2 Revision 22 December 2001 1 3.3-10

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 SURVEILLANCE REQUIREMENTS FREQUENCY SURVEILLANCE Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DSR 3.3.2.1 Perform CHANNEL CHECK. 7 days DSR 3.3.2.2 31 days DSR 3.3.2.3 Perform SOURCE CHECK.

31 days DSR 3.3.2.4 Perform CHANNEL FUNCTIONAL TEST. The CHANNEL FUNCTIONAL TEST shall also demonstrate the automatic isolation capability of this pathway, and that control room alarm annunciation occurs if the instrument indicates measured levels above the alarm/trip setpoint (each channel will be tested independently so as to not initiate isolation during operation); and control room alarm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, and instrument controls not set in operate mode.

DSR 3.3.2.5 Perform CHANNEL FUNCTIONAL TEST. 92 days 92 days DSR 3.3.2.6 Perform CHANNEL FUNCTIONAL TEST. The CHANNEL FUNCTIONAL TEST shall also demonstrate control room alarm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, and instrument controls not set in operate mode.

(continued)

Unit 2 Revision 22 December 2001 13.3-11

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 SURVEILLANCE REQUIREMENTS (continued) I SURVEILLANCE FREQUENCY i

DSR 3.3.2.7 Perform CHANNEL CALIBRATION. The initial 18 months 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 with NIST, or using actual samples of gaseous effluents that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

The CHANNEL CALIBRATION shall also demonstrate that automatic isolation of this pathway occurs when the instrument channels indicate measured levels above the Trip Setpoint.

DSR 3.3.2.8 Perform CHANNEL CALIBRATION. 18 months DSR 3.3.2.9 Perform CHANNEL CALIBRATION. The initial 18 months 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 with NIST, or using actual samples of gaseous effluents that have been analyzed on a system that has been calibrated with NIST traceable sources. These standards shall permit calibrating the system over its intended range of energy and measurement. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration may be used.

Unit 2 Revision 22 1 3.3-12 December 2001

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 Table D 3.3.2-1 (page 1 of 2)

Radioactive Gaseous Effluent Monitoring Instrumentation REQUIRED CONDITIONS APPLICABILITY OR CHANNELS REFERENCED OTHER SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTION B. 1 REQUIREMENTS Offgas System

a. Noble Gas (a) 2 C DSR 3.3.2.1 Activity Monitor DSR 3.3.2.4

- Providing DSR 3.3.2.7 Alarm and Automatic Termination of Release

b. System Flow-Rate (a) I D DSR 3.3.2.1 Measuring Device DSR 3.3.2.5 DSR 3.3.2.8

c. Sample Flow- (a) 2 D DSR 3.3.2.1 Rate Measuring DSR 3.3.2.5 Device DSR 3.3.2.8

2. Radwaste/Reactor Building Vent Effluent System

a. Noble Gas (b) 1 F DSR 3.3.2.1 Activity Monitor DSR 3.3.2.3 (c) DSR 3.3.2.6 DSR 3.3.2.9

b. Iodine Sampler (b) 1 E DSR 3.3.2.2

c. Particulate (b) I E DSR 3.3.2.2 Sampler

d. Flow-Rate (b) 1 D DSR 3.3.2.1 Monitor DSR 3.3.2.5 DSR 3.3.2.8

e. Sample Flow- (b) I D DSR 3.3.2.1 Rate Monitor DSR 3.3.2.5 DSR 3.3.2.8 (continued)

(a) During offgas system operation.

(b) At all times.

(c) Includes high range noble gas monitoring capability.

Unit 2 Revision 22 13.3-13 December 2001

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 Table D 3.3.2-1 (page 2 of 2)

Radioactive Gaseous Effluent Monitoring Instrumentation REQUIRED CONDITIONS APPLICABILITY OR CHANNELS REFERENCED OTHER SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTION B. 1 REQUIREMENTS

3. Main Stack Effluent

a. Noble Gas (b) 1 F DSR 3.3.2.1 Activity Monitor DSR 3.3.2.3 (c) DSR 3.3.2.6 DSR 3.3.2.9

b. Iodine Sampler (b) 1 E DSR 3.3.2.2 C. Particulate (b) 1 E DSR 3.3.2.2 Sampler

d. Flow-Rate (b) 1 D DSR 3.3.2.1 Monitor DSR 3.3.2.5 DSR 3.3.2.8

e. Sample Flow- (b) 1 D DSR 3.3.2.1 Rate Monitor DSR 3.3.2.5 DSR 3.3.2.8 (b) At all times.

(c) Includes high range noble gas monitoring capability.

Unit 2 Revision 22 1 3.3-14 December 2001

Radioactive Effluents Total Dose D 3.4 D 3.4 RADIOACTIVE EFFLUENTS TOTAL DOSE D 3.4 Radioactive Effluents Total Dose DLCO 3.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to < 25 mrem to the whole body or any organ, except the thyroid, which shall be limited to _ 75 mrem.

APPLICABI[LITY: At all times.

ACTIONS

-NOTES ------------------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Estimated dose or dose A. I Verify the condition resulting in Immediately commitment due to direct doses exceeding these limits has radiation and the release of been corrected.

radioactive materials in liquid or gaseous effluents exceeds the limits.

B. Required Action and B. 1 ----- NOTE---

associated Completion Time This is the Special Report required not met. by D 3.1.2, D 3.2.2, or D 3.2.3 supplemented with the following.

Submit a Special Report, 30 days pursuant to D 4.1.1, including a request for a variance in accordance with the provisions of 40 CFR 190. This submission is considered a timely request, and a variance is granted until staff action on the request is complete.

Unit 2 Revision 22 1 3.4-1 December 2001

Radiological Environmental Monitoring Program D 3.5.1 D3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING D 3.5.1 Monitoring Program DLCO 3.5.1 The Radiological Environmental Monitoring Program shall be conducted as specified in Table D 3.5.1-1.

APPLICABILITY: At all times.

ACTIONS

NOTES -----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Radiological Environmental A. 1 Prepare and submit to the NRC In accordance with Monitoring Program not in the Annual Radiological the Annual conducted as specified in Environmental Operating Report, Radiological Table D 3.5.1-1. a description of the reasons for Environmental not conducting the program as Operating Report required and the plans for frequency preventing a recurrence.

B. Level of radioactivity in an B. ---- NOTES ------------

environmental sampling 1. Only applicable if the medium at a specified location radioactivity/radionuclides are exceeds the reporting levels of the result of plant effluents.

Table D 3.5.1-2 when 2. For radionuclides other than averaged over any calendar those in Table D 3.5.1-2, this quarter. report shall indicate the methodology and parameters OR used to estimate the potential annual dose to a MEMBER OF THE PUBLIC.

(continued)

Unit 2 Revision 22 13.5-1 December 2001

Radiological Environmental Monitoring Program D 3.5.1 A rTIflN (cAntinh1Pd CONDITION REQUIRED ACTION COMPLETION TIME More than one of the Prepare and submit to the NRC, 30 days radionuclides in Table pursuant to D 4.1.1, a Special D 3.5.1-2 are detected in the Report that environmental sampling (1) Identifies the cause(s) for medium and exceeding the limit(s) and (2) Defines the corrective actions Concentration 1 + to be taken to reduce reporting level 1 radioactive effluents so that the potential annual dose to a concentration 2 + . 1.0. MEMBER OF THE PUBLIC reporting level 2 is less than the calendar year limits of D 3.1.2, D 3.2.2, or OR D 3.2.3.

Radionuclides other than OR those in Table D 3.5.1-2 are detected in an environmental B.2 -------- NOTES------

sampling medium at a 1.Only applicable if the specified location which are radioactivity/radionuclides are the result of plant effluents not the result of plant effluents.

and the potential annual dose 2.For radionuclides other than to a MEMBER OF THE those in Table D 3.5.1-2, this PUBLIC from all report shall indicate the radionuclides is > the methodology and parameters calendar year limits of used to estimate the potential D 3.1.2, D 3.2.2 or D 3.2.3. annual dose to a MEMBER OF THE PUBLIC.

In accordance with Report and describe the condition the Annual in the Annual Radiological Radiological Environmental Operating Report. Environmental Operating Report frequency (continued)

Unit 2 Revision 22 December 2001 1 3.5-2

Radiological Environmental Monitoring Program D 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Milk or fresh leafy C. 1 Identify specific locations for 30 days vegetation samples obtaining replacement unavailable from one or samples and add them to the more of the sample Radiological Environmental locations required by Table Monitoring Program.

D 3.5.1-1.

AND C.2 Delete the specific locations 30 days from which samples were unavailable from the Radiological Environmental Monitoring Program.

AND C.3 Pursuant to Technical In accordance with Specification 5.6.3, submit in the Radioactive the next Radioactive Effluent Effluent Release Release Report Report documentation for a change in the ODCM reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location(s) for obtaining samples.

4 -I-D. Environmental samples D. I Ensure all efforts are made to Prior to the end of required in Table D 3.5.1-1 complete corrective action(s). the next sampling are unobtainable due to period sampling equipment AND malfunctions.

D.2 Report all deviations from the In accordance with sampling schedule in the the Annual Annual Radiological Radiological Environmental Operating Environmental Report. Operating Report

__________________________________ .1______________________________________ +/-

(continued)

Unit 2 Revision 22 1 3.5-3 December 2001

Radiological Environmental Monitoring Program D 3.5.1 A C'TTCThJ (rntrnid Af"rT NTQ CONDITION REQUIRED ACTION COMPLETION TIME Samples required by Table E. 1 Choose suitable alternative 30 days E.

D 3.5.1-1 not obtained in media and locations for the the media of choice, at the pathway in question.

most desired location, or at the most desired time. AND E.2 Make appropriate 30 days substitutions in the Radiological Environmental Monitoring Program.

AND E.3 Submit in the next In accordance with Radioactive Effluent Release the Radioactive Report documentation for a Effluent Release change in the ODCM Report reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location(s) for obtaining samples.

Unit 2 Revision 22 December 2001 1 3.5-4

Radiological Environmental Monitoring Program D 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.1.1 Collect and analyze radiological environmental In accordance with monitoring samples pursuant to the requirements of Table D 3.5.1-1 Table D 3.5.1-1 and the detection capabilities required by Table D 3.5.1-3.

Unit 2 Revision 22 December 2001 13.5-5

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 1 of 4)

Radiological Environmental Monitoring Program EXPOSURE NUMBER OF SAMPLING AND PATHWAY SAMPLES COLLECTION TYPE AND FREQUENCY AND/OR STATIONS SAMPLE FREQUENCY OF ANALYSIS SAMPLE LOCATIONS (a)

1. Direct 32 routine (1) An inner ring of stations, Once per 3 months Gamma dose: once per 3 Radiation monitoring one in each meteorological months stations (b) sector in the general area of the SITE BOUNDARY (2) An outer ring of stations, one in each land base meteorological sector in the 4 to 5 mile (c) range from the site (3) The balance of the stations should be placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations (d)

2. Airborne 5 locations (1) 3 samples from offsite Continuous sampler Radioiodine canister: Analyze Radioiodine locations close to the site operation with weekly for 1-131 and boundary (within 1 mile) in sample collection Particulates different sectors (e) weekly or more Particulate sampler:

1sample from the vicinity frequently if required (1) Analyze for gross beta (2) by dust loading radioactivity >_24 hours of an established year-roun (e)(f).

(e) comunit following filter change round community (3) 1 sample from a control (2) Perform on analysis gamma isotopic each sample location, at least 10 miles (g) in which groas beta activity i s s th distant and in a least prevalent wind direction (d) activity is> 10 times the previous yearly mean of control samples.

(3) Gamma isotopic analysis of composite sample (g)

(by location) once per 3 months

3. Waterbome

a. Surface 1 sample Upstream (d) (h) Composite sample (1) Gamma isotopic analysis over a one month of each sample (g) once period (i) per month 1 sample Site's downstream cooling water (2) H-3 analysis of each intake (h) composite sample and once per 3 months

b. Ground As required From one or two sources if likely Grab sample once per (1) Gamma isotopic analysis to be affected (j) 3 months of each sample (g) once per 3 months (2) H-3 analysis ofeach sample once per 3 months (continued)

Unit 2 Revision 22 1 3.5-6 December 2001

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 2 of 4)

Radiological Environmental Monitoring Program EXPOSURE PATHWAY SAMPLING AND NUMBER OF SAMPLE COLLECTION TYPE AND FREQUENCY AND/OR SAMPLES LOCATIONS (a) FREQUENCY OF ANALYSIS SAMPLE

3. Waterborne (continued) 1 sample of each One to three of the nearest water When 1-13 1 analysis (1) 1-131 analysis on each

c. Drinking supplies that could be affected by is performed, a composite sample when its discharge (k) composite sample the dose calculated for over a two week the consumption of the period (i); otherwise, water is greater than 1 a composite sample mrem/yr (I) monthly (2) Gross beta and gamma isotopic analyses of each composite sample (g) monthly (3) H-3 analysis of each composite sample once per 3 months 1 sample From a downstream area with Twice per year Gamma isotopic analysis of

d. Sediment from existing or potential recreational each sample (g)

Shoreline value

4. Ingestion (1) 3 samples from In 3 locations within 3.5 miles Twice per month, (1) Gamma isotopic (g) and

a. Milk MILK (e) April through 1-131 analysis of each SAMPLING December (m) sample twice per month LOCATIONS April through December (2) Gamma isotopic (g) and (2) If there are none, In each of 3 areas 3.5-5.0 miles 1-131 analysis of each then I sample distant (e) sample once per month from MILK January through March SAMPLING if required LOCATIONS (3) 1 sample from a At a control location 9-20 miles MILK distant and in a least prevalent SAMPLING wind direction (d)

LOCATION (1) 1 sample each of In the vicinity of a plant Twice per year Gamma isotopic analysis of

b. Fish 2 commercially discharge area each sample (g) on edible or recreationally portions twice per year important species (n)

(2) 1 sample of the In areas not influenced by station same species discharge (d)

(continued)

Unit 2 Revision 22 December 2001 13.5-7

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 3 of 4)

Radiological Environmental Monitoring Program EXPOSURE PATHWAY SAMPLING AND TYPE AND FREQUENCY AND/OR NUMBER OF SAMPLE COLLECTION OF ANALYSIS SAMPLE SAMPLES LOCATIONS (a) FREQUENCY

4. Ingestion (continued)

c. Food (1) 1 sample of Any area that is irrigated by At time of harvest (p) Gamma isotopic (g) and 1-131 Products each principal water in which liquid plant analysis of each sample of class of food wastes have been discharged (o) edible portions products (2) Samples of 3 Grown nearest to each of 2 Once per year during different kinds different offsite locations (e) the harvest season of broad leaf vegetation (such as vegetables)

(3) 1 sample of Grown at least 9.3 miles distant Once per year during each of the in a least prevalent wind the harvest season similar broad direction leaf vegetation.

Unit 2 Revision 22 1 3.5-8 December 2001

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 3 of 4)

Radiological Environmental Monitoring Program (a) Specific parameters of distance and direction sector from the centerline of one reactor, and additional descriptions where pertinent, shall be provided for each and every sample location in Table D 3.5.1-1. Refer to NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable because of such circumstances as hazardous conditions, seasonal unavailability (which includes theft and uncooperative residents), or malfunction of automatic sampling equipment.

(b) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to integrating dosimeters. Each of the 32 routine monitoring stations shall be equipped with 2 or more dosimeters or with 1 instrument for measuring and recording dose rate continuously. For the purpose of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor, 2 or more phosphors in a packet are considered as 2 or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation.

(c) At this distance, 8 windrose sectors (W, WNW, NW, NNW, N, NNE, NE, and ENE) are over Lake Ontario.

(d) The purpose of these samples is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites, which provide valid background data, may be substituted.

(e) Having the highest calculated annual site average ground-level D/Q based on all site licensed reactors.

(f) Airborne particulate sample filters shall be analyzed for gross beta activity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay.

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

(h) The upstream sample shall be taken at a distance beyond significant influence of the discharge. The downstream sample shall be taken in an area beyond but near the mixing zone.

(i) In this program, representative composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.

(j) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.

(k) Drinking water samples shall be taken only when drinking water is a dose pathway.

(1) Analysis for 1-131 may be accomplished by Ge-Li analysis provided that the lower limit of detection (LLD) for 1-131 in water samples found on Table D 3.5.1-2 can be met. Doses shall be calculated for the maximum organ and age group.

(in) Samples will be collected January through March if 1-131 is detected in November and December of the preceding year.

(n) In the event 2 commercially or recreationally important species are not available, after 3 attempts of collection, then 2 samples of one species or other species not necessarily commercially or recreationally important may be utilized.

(o) Applicable only to major irrigation projects within 9 miles of the site in the general downcurrent direction.

(p) If harvest occurs more than once/year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be taken monthly. Attention should be paid to including samples of tuberous and root food products.

Unit 2 Revision 22 1 3.5-9 December 2001

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-2 (page 1 of 1)

Reporting Levels for Radioactivity in Environmental Samples AIRBORNE FOOD PARTIUCLATE 3OR FISH MILK PRODUCTS RADIONUCLIDE WATER (pCi/L) GASES (pCi/m ) (pCi/kg. wet) (pCi/L) (pCi/kg, wet)

ANALYSIS H-3 20,000 (a)

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-95 400 Nb-95 400 1-131 2 (b) 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-140 200 300 La-140 200 300 may be used.

(a) For drinking water samples. This is a 40 CFR 141 value. If no drinking water pathway exists, a value of 30,000 pCi/L (b) If no drinking water pathway exists, a value of 20 pCi/L may be used.

Unit 2 Revision 22 December 2001 1 3.5-10

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-3 (page 1-of2)

Detection Capabilities for Environmental Sample Analysis (a)(b)

LOWER LIMIT OF DETECTION (LLD)()'

AIRBORNE PARTIUCLATE OR FOOD 3 PRODUCTS SEDIMENT RADIONUCLIDE WATER GASES (pCi/m ) FISH MILK (pCi/L) (pCi/kg, wet) (pCi/L) (pCi/kg, wet) (pCi/kg, dry)

ANALYSIS Gross Beta 4 0.01 H-3 2,000 (

Mn-54 15 130 30 260 Fe-59 15 130 Co-58 15 130 Co-60 30 260 Zn-65 Zr-95 15 Nb-95 15 1-131 1W 0.07 60 15

^ 60 Cs-134 15 0. 05 130 150 150 18 80 180 Cs-137 18 0. 06 Ba-140 15 15 15 15 La-140 See the notes on the next page Unit 2 Revision 22 December 2001 13.5-11

Radiological Environmental Monitoring Program D 3.5.1 Table 3.5.1-3 (page 2 of 2)

Detection Capabilities for Environmental Sample Analysis (a)()

(a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report.

(b) Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in ANSI N-545, Section 4.3 1975. Allowable exceptions to ANSI N-545, Section 4.3 are contained in the ODCM.

(c) The LLD is defined as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 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) MYe-t where:

LLD = The before-the-fact lower limit of detection (pCi per unit mass or volume),

Sb = The standard deviation ofthe background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),

E = The counting efficiency (counts per disintegration),

V = The sample size (units of mass or volume),

2.22 = The number of disintegrations per minute per pCi, Y = The fractional radiochemical yield, when applicable,

-1 The radioactive decay constant for the particular radionuclide (sec'), and At = The elapsed time between environmental collection or end of the sample collection period, and the time of counting (seconds).

Typical values of E, V, Y, and At should be used in the calculation.

It should be recognized that the LLD is defined as a before-the-fact limit representing the capability of a measurement system and not as an 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, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

(d) If no drinking water pathway exists, a value of 3,000 pCi/L may be used.

(e) If no drinking water pathway exists, a value of 15 pCi/L may be used.

Unit 2 Revision 22 1 3.5-12 December 2001

Land Use Census D 3.5.2 D3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING D 3.5.2 Land Use Census DLCO 3.5.2 A land use census shall:

a. Be conducted,

b. Identify within a distance of 5 miles the location in each of the 16 meteorological sectors of the nearest milk animal and the nearest residence, and the nearest garden (broad leaf vegetation sampling controlled by Table D 3.5.1-1, part 5.c may be performed in lieu of the garden census) of > 500 ft2 producing broad leaf vegetation, and

c. For elevated releases, identify within a distance of 3 miles the locations in each of the 16 meteorological sectors of all milk animals and all gardens (broad leaf vegetation sampling controlled by Table D 3.5.1-1, 5.c may be performed in lieu of the garden census) > 500 ft2 part producing broad leaf vegetation.

APPLICABILITY: At all times.

ACTIONS

-.........................................................N O T E S -----------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Land use census identifies A. 1 Identify the new location(s) in In accordance with location(s) that yields a the next Radioactive Effluent the Radioactive calculated dose, dose Release Report. Effluent Release commitment, or D/Q value Report

> than the values currently being calculated in DSR 3.2.3.1.

(continued)

Unit 2 Revision 22 I 3.5-13 December 2001

Land Use DCensus 3.5.2 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. 1 Add the new location(s) to the 30 days B. Land use census identifies location(s) that yields a Radiological Environmental Monitoring Program.

calculated dose, dose commitment, or D/Q value (via the same exposure AND pathway) 50% > than at a After October 31 of B.2 Delete the sampling location from which the year in which location(s), excluding the samples are currently being the land use census control station location, obtained in accordance was conducted having the lowest calculated with Table D 3.5.1-1.

dose, dose commitment(s) or D/Q value, via the same exposure pathway, from the Radiological Environmental Monitoring Program.

AND In accordance with B.3 Submit in the next Radioactive Effluent Release Report the Radioactive Effluent Release documentation for a change in Report the ODCM including revised figure(s) and table(s) for the ODCM reflecting the new location(s) with information supporting the change in sampling locations.

Unit 2 Revision 22 December 2001 1 3.5-14

Land Use DCensus 3.5.2 SURVEILLANCE REQUIREMENTS FREQUENCY SURVEILLANCE 366 days DSR 3.5.2.1 Conduct the land use census during the growing season using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities.

Report the results of the land use census in the Annual In accordance with DSR 3.5.2.2 Radiological Environmental Operating Report. the Annual Radiological Environmental Operating Report Unit 2 Revision 22 December 2001 13.5-15

Interlaboratory Comparison Program D 3.5.3 D 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING D 3.5.3 Interlaboratory Comparison Program DLCO 3.5.3 The Interlaboratory Comparison Program shall be described in the ODCM.

AND Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the NRC, that correspond to samples required by Table D 3.5.1-1.

Participation in this program shall include media for which environmental samples are routinely collected and for which intercomparison samples are available.

APPLICABILITY: At all times.

ACTIONS

N O TES OTES------------------------------ -----------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Analyses not performed as A. 1 Report the corrective actions In accordance with required. taken to prevent a recurrence the Annual to the NRC in the Annual Radiological Radiological Environmental Environmental Operating Report. Operating Report Unit 2 Revision 22 December 2001 1 3.5-16

Interlaboratory Comparison Program D 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.3.1 Report a summary of the results obtained as part of the In accordance with Interlaboratory Comparison Program in the Annual the Annual Radiological Environmental Operating Report. Radiological Environmental Operating Report Unit 2 Revision 22 13.5-17 December 2001

PART I - RADIOLOGICAL EFFLUENT CONTROLS BASES Unit 2 Revision 22 1B 3.1-0 December 2001

Liquid Effluents Concentration B 3.1.1 B3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.1 Liquid Effluents Concentration BASES This is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than ten times the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of Appendix I to 10 CFR 50, to a MEMBER OF TIE PUBLIC and (2) the levels required by 10 CFR 20.1301(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its effluent 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.

This applies to the release of radioactive materials in liquid effluents from all units at the site.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in L. A. Currie, "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

Unit 2 Revision 22 IB 3.1-1 December 2001

Liquid Effluents Dose B 3.1.2 B 3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.2 Liquid Effluents Dose BASES This is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I to 10 CFR 50. This 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 IV.A.of Appendix I to assure that the releases of radioactive materials in liquid effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the potable drinking water that are in excess of the requirements of 40 CFR 141.

For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside storage tanks, etc., are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBERS OF THE PUBLIC from other uranium 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. The dose calculation methodology and parameters 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, so that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified for calculating the doses that result from actual release rates of radioactive material in liquid effluents are consistent with the methodology provided in RG 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 R.G. 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. This applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

Unit 2 Revision 22 I B 3.1-2 December 2001

Liquid Radwaste Treatment System B 3.1.3 B3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.3 Liquid Radwaste Treatment System BASES The installed liquid radwaste treatment system shall be considered OPERABLE by meeting DLCO 3.1.1 and DLCO 3.1.2. The OPERABILITY of the liquid radwaste treatment system ensures that this system will be available for use whenever liquid effluents require treatment before release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable. This implements the requirements of 10 CFR 50.36a, GDC 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 limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I to 10 CFR 50 for liquid effluents. This applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

Unit 2 Revision 22 IB 3.1-3 December 2001

Gaseous Effluents Dose Rate B 3.2.1 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.1 Gaseous Effluents Dose Rate BASES This is provided to ensure that the dose rate at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR 20 to UNRESTRICTED AREAS.

The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table 2, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR 20 or as governed by 10 CFR 20.1302(c). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in Part II. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC 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 mremryear to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrem/year. This applies to the release of radioactive materials in gaseous effluents from all units at the site.

The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in L. A. Currie, "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environments Measurements,"

NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

Unit 2 Revision 22 I B 3.2-1 December 2001

Gaseous Effluents Noble Gas Dose B 3.2.2 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.2 Gaseous Effluents Noble Gas Dose BASES This is provided to implement the requirements of Section II.B, III.A, and IV.A of Appendix I to 10 CFR 50. The DLCO implements the guides set forth in Section II.B of Appendix I. The REQUIRED ACTIONS 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 material in gaseous effluents to UNRESTRICTED AREAS 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 guidelines of Appendix I be shown by calculational procedures based on models and data so that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside storage tanks, etc., are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium 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. The dose calculation methodology and parameters for calculating the doses from the actual release rates of radioactive noble in gaseous effluents are consistent with the methodology provided in RG 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 RG 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 or beyond the SITE BOUNDARY are based upon real-time meteorological conditions or the historical average atmospheric conditions. This applies to the release of radioactive material in gaseous effluents from each unit at the site.

Unit 2 Revision 22 I B 3.2-2 December 2001

Gaseous Effluents Dose' Iodine- 13 1, Iodine- 133, Tritium, and Radioactive Material In Particulate Form B 3.2.3 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.3 Gaseous Effluents Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material In Particulate Form BASES This is provided to implement the requirements of Sections II.C, III.A, and IV.A of Appendix I to 10 CFR 50. The DLCO implements the guides set forth in Section.II.C of Appendix I. The REQUIRED ACTIONS 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 to UNRESTRICTED AREAS will be kept as low as is reasonably the achievable. The calculational methods specified in the Surveillance Requirements implement guides of Appendix I be requirements in Section III.A of Appendix I that conformance with the shown by calculational procedures based on models and data, so that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside storage tanks, etc., are kept small. The Special dose to a Report will describe a course of action that should result in the limitation of the annual MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium 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.

release The calculational methodology and parameters for calculating the doses from the actual rates of the subject materials are consistent with the methodology provided in RG 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose and of Evaluating Compliance with 10 CFR Part 50, Appendix I, "Revision 1, October 1977, RG 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric material conditions. The release rate DLCO for iodine-13 1, iodine-133, tritium, and radioactive in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide were pathways to man, in the areas at or beyond the SITE BOUNDARY. The pathways that inhalation of airborne examined in the development of these calculations were: (1) individual radioactive material, (2) deposition of radioactive material onto green leafy vegetation Unit 2 Revision 22 I B 3.2-3 December 2001

Gaseous Effluents Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material In Particulate Form B 3.2.3 B 3.2.3 Gaseous Effluents Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material In Particulate Form (continued) with subsequent consumption by man, (3) deposition onto grassy areas where milk-producing animals and meat-producing animals graze (human consumption of the milk and meat is assumed),

and (4) deposition on the ground with subsequent exposure to man. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

Unit 2 Revision 22 IB 3.2-4 December 2001

Gaseous Radwaste Treatment System B 3.2.4 B3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.4 Gaseous Radwaste Treatment System BASES The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM ensures that the system. will be available for use whenever gaseous effluents require treatment before release to the environment. The requirement that the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable. This implements the requirements of 10 CFR 50.36a, GDC 60 of Appendix A to 10 CFR 50, and the design objectives given in Section II.D of Appendix I to 10 CFR 50. Limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I to 10 CFR 50, for gaseous effluents. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportional among the units sharing that system.

Unit 2 Revision 22 I B 3.2-5 December 2001

"VentilationExhaust Treatment System B 3.2.5 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.5 Ventilation Exhaust Treatment System BASES The OPERABILITY of the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the system will be available for use whenever gaseous effluents require treatment before release to the environment. The requirement that the appropriate portions of this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable. This implements the requirements of 10 CFR 50.36a, GDC 60 of Appendix A to 10 CFR 50, and the design objectives given in Section II.D of Appendix I to 10 CFR 50. Limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I to 10 CFR 50, for gaseous effluents. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportional among the units sharing that system.

The appropriate components, which affect iodine or particulate release, to be OPERABLE are:

1) HEPA Filter - Radwaste Decon Area

2) HEPA Filter - Radwaste Equipment Area

3) HEPA Filter - Radwaste General Area Whenever one of these filters is not OPERABLE, iodine and particulate dose projections will be made for 31-day intervals starting with filter inoperability, and continuing as long as the filter remains inoperable, in accordance with DSR 3.2.5.1.

Unit 2 Revision 22 I B 3.2-6 December 2001

Venting or Purging B 3.2.6 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.6 Venting or Purging BASES chamber This provides reasonable assurance that releases from drywell and/or suppression unrestricted areas.

purging operations will not exceed the annual dose limits of 10 CFR 20 for Unit 2 Revision 22 1B 3.2-7 December 2001

Radioactive Liquid Effluent Monitoring Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation BASES 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 Part II to ensure that the alarm/trip will occur before exceeding ten times the limits of 10 CFR 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of GDC 60, 63, and 64 of Appendix A to 10 CFR 50. The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of radioactive materials to UNRESTRICTED AREAS.

Tanks included are those outdoor tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system, such as temporary tanks.

Unit 2 Revision 22 December 2001 I B 3.3-1

Radioactive Gaseous Effluent Monitoring Instrumentation B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation BASES 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 Part II to ensure that the alarm/trip will occur before exceeding the limits of 10 CFR 20. Although the Offgas System Noble Gas Activity Monitor is listed in Table D 3.3.2-1, "Radioactive Gaseous Effluent Monitoring Instrumentation", these monitors are actually located upstream of the Main Stack noble gas activity monitor and are not effluent monitors. They were included in Table D 3.3.2-1 in accordance with NUREG-0473. As such, Offgas System Noble Gas Activity Monitor alarm and trip setpoints are not based on 10CFR20. The offgas system noble gas monitor alert setpoint is set at 1.5 times nominal full power background to assure compliance with ITS SR 3.7.4.1 which requires offgas sampling be performed within four hours of a 50% increase in offgas monitoring readings, and to support MSLRM trip removal. The offgas system noble gas monitor trip setpoint is based on the 10CFR100 limits for the limiting design basis gaseous waste system accident which is the offgas system rupture. The range of the noble gas channels of the main stack and radwaste/reactor building vent effluent monitors is sufficiently large to envelope both normal and accident levels of noble gas activity. The capabilities of these instruments are consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident,"

December 1980 and NUREG-0737, "Clarification of the TMI Action Plan Requirements,"

November 1980. This instrumentation also includes provisions for monitoring and controlling the concentrations of potentially explosive gas mixtures in the offgas system. The OPERABILITY and use of this instrumentation is consistent with the requirements of GDC 60, 63, and 64 of Appendix A to 10 CFR 50.

Unit 2 Revision 22 I B 3.3-2 December 2001

Radioactive Effluents Total Dose B 3.4 B 3.4 RADIOACTIVE EFFLUENTS TOTAL DOSE BASES This is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46 FR 18525. This requires the preparation and submittal of a Special Report whenever the calculated doses from releases of radioactivity and from radiation from uranium fuel cycle sources exceed 25 mrem to the whole body or any organ, except the thyroid (which shall be limited to less than or equal to 75 mrem). If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR 20, as addressed in 3.1.1 and 3.2.1. An individual is not considered a IMEMBER OF THE PUBLIC during any period in which the individual is engaged in carrying out any operation that is part of the nuclear fuel cycle.

Unit 2 Revision 22 IB 3.4-1 December 2001

Monitoring Program B 3.5.1 B 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 3.5.1 Monitoring Program BASES The Radiological Environmental Monitoring Program provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR 50 and 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 the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. Program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table D 3.5.1-3 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as a before-the-fact limit representing the capability of a measurement system and not as an after-the-fact limit for a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in L. A. Currie, "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984),

and in the HASL Procedures Manual, HASL-300 (revised annually).

Unit 2 Revision 22 I B 3.5-1 December 2001

Land Use Census B 3.5.2 B 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 3.5.2 Land Use Census BASES This is provided to ensure that changes in the use of areas at or beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program are made if required by the results of this census. The best information, such as from a door-to-door survey, from an aerial survey, or from consulting with local agricultural authorities, shall be used.

This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR 50.

Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in RG 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf 2

vegetation (i.e., similar to lettuce and cabbage) and (2) the vegetation yield was 2 kg/m .

A MILK SAMPLING LOCATION, as defined in Section 1.0, requires that at least 10 milking cows are present at a designated milk sample location. It has been found from past experience, and as a result of conferring with local farmers, that a minimum of 10 milking cows is necessary to guarantee an adequate supply of milk twice a month for analytical purposes. Locations with fewer than 10 milking cows are usually utilized for breeding purposes, eliminating a stable supply of milk for samples as a result of suckling calves and periods when the adult animals are dry.

Elevated releases are defined in RG 1.111, Revision 1, July 1977.

Unit 2 Revision 22 I B 3.5-2 December 2001

Interlaboratory Comparison Program B 3.5.3 B 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 3.5.3 Interlaboratory Comparison Program BASES The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR 50.

Unit 2 Revision 22 IB 3.5-3 December 2001

PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 4.0 ADMINISTRATIVE CONTROLS Unit 2 Revision 22 14.0-0 December 2001

Administrative Controls 4.0 4.0 ADMINISTRATIVE CONTROLS 5.5.4, "Radioactive The ODCM Specifications are subject to Technical Specifications Section Operating Effluent Controls Program," Section 5.6.2, "Annual Radiological Environmental 5.5.1, "Offsite Dose Report," Section 5.6.3, "Radioactive Effluent Release Report," and Section Calculation Manual."

Unit 2 Revision 22 14.0-1 December 2001

Special Reports D 4.1.1 D 4.1.2 D 4.1.3 D 4.1 REPORTING REQUIREMENTS D 4.1.1 Special Reports Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.

D 4.1.2 Annual Radiological Environmental Operating Reports In addition to the requirements of Technical Specification 5.6.2 the report shall also include the following:

A summary description of the Radiological Environmental Monitoring Program; at least two legible maps, one shall cover stations near the SITE BOUNDARY and the second shall include the more distant stations, covering all sample locations keyed to a table giving distances and directions from the centerline of one reactor; the results of license participation in the Interlaboratory Comparison Program, required by Control D 3.5.3; discussion of all deviations from the Sampling Schedule of Table D 3.5.1-1; and discussion of all analysis in which the LLD required by Table D 3.5.1-3 was not achievable.

D 4.1.3 Radioactive Effluent Release Report The Radiological Effluent Release Report described in Technical Specification section 5.6.3 shall include:

"* An annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distribution of wind speed, wind direction, and atmospheric stability. In lieu of submission with the Radiological Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

"* An assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during the previous year.

(Continued)

Unit 2 Revision 22 14.1-1 December 2001

Special Reports D 4.1.3 D 4.1.3 Radioactive Effluent Release Report (continued)

"* As assessment of radiation doses from the radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC from their activities inside the SITE BOUNDARY during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location shall be included in these reports. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in Part II.

"* As assessment of doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR 190, "Environmental Radiation Protection Standards for Nuclear Power Operation."

Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Part II.

"* A list of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

"* Any changes made during the reporting period to the PROCESS CONTROL PROGRAM and to the OFFSITE DOSE CALCULATION MANUAL (ODCM).

"* Any major changes to liquid, gaseous, or solid radwaste treatment systems pursuant to D 4.2.

"* A listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Control D 3.5.2.

"* An explanation of why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Controls D 3.3.1 and D 3.3.2.

"* Description of events leading to liquid holdup tanks exceeding the limits of TRM 3.7.7.

Unit 2 Revision 22 14.1 -la December 2001

Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment System D4.2 D 4.2 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEM

NOTE ---------------------------------------------------

Licensees may choose to submit this information as part of the annual FSAR update.

Licensee-initiated major changes to the radwaste treatment systems (liquid, gaseous, and solid):

a. Shall be reported to the Commission in the Radioactive Effluent Release report for the period in which the evaluation was reviewed by the SORC. The discussion of each change shall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59.

2. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

4. An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

5. An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto;

6. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period that precedes the time when the change is to be made;

7. An estimate of the exposure to plant operating personnel as a result of the change; and (Continued)

Unit 2 Revision 22 14.2-1 December 2001

Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment System D4.2 D 4.2 MAJOR CHANGES TO LIQUID, GASEOUS, AND SOLID RADWASTE TREATMENT SYSTEM (continued)

8. Documentation of the fact that the change was reviewed and found acceptable by the SORC.

b. Shall become effective upon review and acceptance by the SORC.

Unit 2 Revision 22 December 2001 1 4.2-2

PART II - CALCULATIONAL METHODOLOGIES 1.0 LIQUID EFFLUENTS Service Water A and B, Cooling Tower Blowdown and the Liquid Radioactive Waste Discharges comprise the Radioactive Liquid Effluents at Unit 2. Presently there are no temporary outdoor tanks containing radioactive water capable of affecting the nearest known or future water supply in an unrestricted area. NUREG 0133 and Regulatory Guide 1. 109, Rev. 1 were followed in the development of this section.

1.1 Liquid Effluent Monitor Alarm Setpoints 1.1.1 Basis The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) 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 nobles gases, the concentration shall be limited to 2E-04 uCi/ml total activity.

1.1.2 Setpoint Determination Methodology 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint The Liquid Radioactive Waste System Tanks are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. At the end of the discharge tunnel in Lake Ontario, a diffuser structure has been installed. Its purpose is to maintain surface water temperatures low enough to meet thermal pollution limits. However, it also assists in the near field dilution of any activity released. Service Water and the Cooling Tower Blowdown are also pumped to the discharge tunnel and will provide dilution. If the Service Water or the Cooling Tower Blowdown is found to be contaminated, then its activity will be accounted for when calculating the permissible radwaste effluent flow for a Liquid Radwaste discharge. The Liquid Radwaste System Monitor provides alarm and automatic termination of release if radiation levels above its alarm setpoint are detected.

The radiation detector is a sodium iodide crystal. It is a scintillation device. The crystal is sensitive to gamma and beta radiation. However, because of the metal walls of the sample chamber and the absorption characteristics of water, the monitor is not particularly sensitive to beta radiation. Actual detector response Yj (CGi/CFi), cpm, has been evaluated by placing a sample of typical radioactive waste into the monitor and recording the gross count rate, cpm. A calibration ratio was developed by dividing the noted detector response, Yj (CGj/CFj) cpm, by total concentration of activity Yj (CGD), uci/cc. The quantification of the gamma activity was completed with gamma spectrometry equipment whose calibration is traceable to NIST. This calibration ratio verified the manufacturer's prototype calibration, and any subsequent transfer calibrations performed. The current calibration factor (expressed as the reciprocal conversion factor, uCi/ml/cpm), will be used for subsequent setpoint calculations in the determination of detector response:

Ei(CGi/CFj) = Yi(CGi)/CF Unit 2 Revision 22 II 2 December 2001

Where the factors are as defined above.

For the calculation of RDF = E MEC fraction = Yi (Ci /MECj) the contribution from non gamma emitting nuclides except tritium will be initially estimated based on the expected ratios to quantified nuclides as listed in the FSAR Table 11.2.5. Fe-55, Sr-89 and Sr-90 are 2.5, 0.25 and 0.02 times the concentration of Co-60. The contribution will be estimated using the results from the latest analysis of composite samples, when available.

Tritium concentration is assumed to equal the latest concentration detected in the monthly tritium analysis (performed offsite) of liquid radioactive waste tanks discharged.

Nominal flow rates of the Liquid Radioactive Waste System Tanks discharged is

< 165 gpm while dilution flow from the Service Water Pumps, and Cooling Tower Blowdown cumulatively is typically over 10,200 gpm. Because of the large amount of dilution the alarm setpoint could be substantially greater than that which would correspond to the concentration actually in the tank. Potentially a discharge could continue even if the distribution of nuclides in the tank were substantially different from the grab sample obtained prior to discharge which was used to establish the detector alarm point. To avoid this possibility of "Non representative Sampling" resulting in erroneous assumptions about the discharge of a tank, the tank is recirculated for a minimum of 2.5 tank volumes prior to sampling.

This monitor's setpoint takes into account the dilution of Radwaste Effluents provided by the Service Water and Cooling Tower Blowdown flows. Detector response for the nuclides to be discharged (cpm) is multiplied by the Actual Dilution Factor (dilution flow/waste stream flow) and divided by the Required Dilution Factor (total fraction of the effluent concentration in the waste stream). A safety factor is used to ensure that the limit is never exceeded. Service Water and Cooling Tower Blowdown are normally non-radioactive. If they are found to be contaminated prior to a Liquid Radwaste discharge then an alternative equation is used to take into account the contamination. If they become contaminated during a Radwaste discharge, then the discharge will be immediately terminated and the situation fully assessed.

Normal Radwaste Effluent Alarm Setpoint Calculation:

Alarm Setpoint < 0.8

TDF/PEF

TGC/CF

1/RDF + Background.

Where:

Alarm Setpoint = The Radiation Detector Alarm Setpoint, cpm 0.8 = Safety Factor, unitless TDF = .Nonradioactive dilution flow rate, gpm. Service Water Flow (ranges from 30,000 to 58,000 gpm) +

Blowdown flow (typically 10,200 gpm) - Tempering Unit 2 Revision 22 II 3 December 2001

Ci Concentration of isotope i in Radwaste tank prior to dilution, tCi/ml (gamma + non-gamma emitters)

CFj - Detector response for isotope i, net 1,Ci/ml/cpm See Table D 2-1 for a list of nominal values PEF = The permissible Radwaste Effluent Flow rate, gpm, 165 gpm is the maximum value used in this equation MECi = Maximum Effluent Concentration, ten times the limiting effluent concentration for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, 1tCi/ml Background = Detector response when sample chamber is filled with nonradioactive water, cpm CF - Monitor Conversion Factor, ýtCi/ml/cpm, determined at each calibration of the effluent monitor CG, - Concentration of gamma emitting nuclide in Radwaste tank prior to dilution, *Ci/ml TGC = YCGi = Summation of all gamma emitting nuclides (which monitor will respond to)

Y (CGi/CFi) - The total detector response when exposed to the concentration of nuclides in the Radwaste tank, cpm RDF = Yj (Ci/MEC1 ) - The total fraction of ten times the 10 CFR 20, Appendix B, Table 2, Column 2 limit that is in the Radwaste tank, unitless. This is also known as the Required Dilution Factor (RDF), and includes non-gamma emitters TGC/CF An approximation to Y' (CGi/CFi) using CF determined at each calibration of the effluent monitor TDF/PEF = An approximation to (TDF + PEF)/PEF, the Actual Dilution Factor in effect during a discharge.

Tempering A diversion of some fraction of discharge flow to the intake canal for the purpose of temperature control, gpm.

Permissible effluent flow, PEF, shall be calculated to determine that the maximum effluent concentration will not be exceeded in the discharge canal.

PEF = TDF (RDF) 1.5 If Actual Dilution Factor is set equal to the Required Dilution Factor, then the alarm points required by the above equations correspond to a concentration of 80% of the Radwaste Tank concentration. No discharge could occur, since the monitor would be in alarm as soon as the discharge commenced. To avoid this situation, maximum allowable radwaste discharge flow is calculated using a multiple (usually 1.5 to 2) of the Required Dilution Factor, resulting in discharge canal concentration of 2/3 to 1/2 of the maximum effluent concentration prior to alarm and termination of release. In Unit 2 Revision 22 II 4 December 2001

performing the alarm calculation, the smaller of 165 gpm (the maximum possible flow) and PEF will be used.

To ensure the alarm setpoint is not exceeded, an alert alarm is provided. The alert alarm will be set in accordance with the equation above using a safety factor of 0.5 (or lower) instead of 0.8.

1.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculation:

The allowable discharge flow rate for a Radwaste tank, when one of the normal dilution streams (Service Water A, Service Water B, or Cooling Tower Blowdown) is contaminated, will be calculated by an iterative process. Using Radwaste tank concentrations with a total liquid effluent flow rate, the resulting fraction of the maximum effluent concentration in the discharge canal will be calculated.

FMEC = + MECJ)]

+[F.,IY(Fs)Y(C.

Then the permissible radwaste effluent flow rate is given by:

PEF = Total Radwaste Effluent Flow FMEC The corresponding Alarm Setpoint will then be calculated using the following equation, with PEF limited as above.

TGC/CF Alarm Setpoint < 0.8 + Background FMEC Where:

Alarm Setpoint = The Radiation Detector Alarm Setpoint, cpm 0.8 = Safety Factor, Unitless Fs = An Effluent flow rate for stream s, gpm Ci = Concentration of isotope i in Radwaste tank prior to dilution, *Ci/ml cis Concentration of isotope i in Effluent stream s including the Radwaste Effluent tank undiluted, ýtCi/ml CF - Average detector response for all isotopes in the waste stream, net ýLCi/ml/cpm MECi Maximum Effluent Concentration, ten times the effluent concentration limit for isotope i from 10CFR20 Appendix B, Table 2, Column 2, ýCi/ml PEF = The permissible Radwaste Effluent Flow rate, gpm Background = Detector response when sample chamber is filled with nonradioactive water, cpm Unit 2 Revision 22 II 5 December 2001

TGC/CF = The total detector response when exposed to the Yi (CGi/CF) concentration of nuclides in the Radwaste tank, cpm F.Ci ] = The total activity of nuclide i in all Effluent streams, tCi-gpm/ml x [F.] = The total Liquid Effluent Flow rate, gpm (Service Water & CT Blowdown & Radwaste) 1.1.2.3 Service Water and Cooling Tower Blowdown Effluent Alarm Setpoint These monitor setpoints do not take any credit for dilution of each respective effluent stream. Detector response for the distribution of nuclides potentially discharged is divided by the total MEC fraction of the radionuclides potentially in the respective stream. A safety factor is used to ensure that the limit is never exceeded.

Service Water and Cooling Tower Blowdown are normally non-radioactive. If they are found to be contaminated by statistically significant increase in detector response then grab samples will be obtained and analysis meeting the LLD requirements of Table D 3. 1. 1-1 completed so that an estimate of offsite dose can be made and the situation fully assessed.

Service Water A and B and the Cooling Tower Blowdown are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. Normal flow rates for each Service Water Pump is 10,000 gpm while that for the Cooling Tower Blowdown may be as much as 10,200 gpm. Credit is not taken for any dilution of these individual effluent streams.

The radiation detector is a sodium iodide crystal. It is a scintillation device. The crystal is sensitive to gamma and beta radiation. However, because of the metal walls in its sample chamber and the absorption characteristics of water, the monitor is not particularly sensitive to beta radiation.

Detector response i (ci/cFr) has been evaluated by placing a diluted sample of Reactor Coolant (after a two hour decay) in a representative monitor and noting its gross count rate. Reactor Coolant was chosen because it represents the most likely contaminant of Station Waters.

A two hour decay was chosen by judgement of the staff of Nine Mile Point. Reactor Coolant with no decay contains a considerable amount of very energetic nuclides which would bias the detector response term high. However assuming a longer than 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months decay is not realistic as the most likely release mechanism is a leak through the Residual Heat Removal Heat Exchangers which would contain Reactor Coolant during shutdowns.

Unit 2 Revision 22 II 6 December 2001

Service Water and Cooling Tower Blowdown Alarm Setpoint Equation:

Alarm Setpoint < 0.8 1/CF Ei Ci/[Yi(Ci/MECi)] + Background.

Where:

Alarm Setpoint = The Radiation Detector Alarm Setpoint, cpm 0.8 - Safety Factor, unitless Ci = Concentration of isotope i in potential contaminated stream,

ýCi/ml CFj = Detector response for isotope i, net ýtCi/ml/cpm See Table 2-1 for a list of nominal values MECi - Maximum Effluent Concentration, ten times the effluent concentration limit for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, tCi/ml Background = *Detector response when sample chamber is filled with nonradioactive water, cpm YZ(Ci/CFi) = The total detector response when exposed to the concentration of nuclides in the potential contaminant, cpm Yi (Ci/MECi) The total fraction of ten times the 10CFR20, Appendix B, Table 2, Column 2 limit that is in the potential contaminated stream, unitless.

(1/CF) iCi An approximation to Y (ci/CFi), determined at each calibration of the effluent monitor CF = Monitor Conversion Factor, tCi/ml/cpm 1.2 Liquid Effluent Concentration Calculation This calculation documents compliance with Section D 3.1.1 of Part 1:

As required by Technical Specification 5.5.4, "Radioactive Effluent Controls Program," the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure D 1.0-1) 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 be limited to 2E-04 microcurie/ml total activity.

The concentration of radioactivity from Liquid Radwaste, Service Water A and B and the Cooling Tower Blowdown are included in the calculation. The calculation is performed for a specific period of time. No credit is taken for averaging. The limiting concentration is calculated as follows:

FMEC Es [ F.Fs/Y(Fs) Yi (Cis-.MECi) ]

Where: FMEC - The Fraction of Maximum Effluent Concentration, the ratio at the point of discharge Unit 2 Revision 22 II 7 December 2001

of the actual concentration to ten times the limiting concentration of 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases, unitless Cis = The concentration of nuclide i in a particular effluent stream s, ýCi/ml

- The flow rate of a particular effluent stream s, gpm MEC, = Maximum Effluent Concentration, ten times the limiting Effluent Concentration of a specific nuclide i from 10CFR20, Appendix B, Table 2, Column 2 (for noble gases, the concentration shall be limited to 2E-4 microcurie/ml), tCi/ml Yj (CF./MEC)) = The Maximum Effluent Concentration fraction of stream s prior to dilution by other streams

= The total flow rate of all effluent streams s, gpm A value of less than one for the MEC fraction is required for compliance.

1.3 Liquid Effluent Dose Calculation Methodology The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure D 1.0-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the whole 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 whole body and to less than or equal to 10 mrem to any organ.

Doses due to Liquid Effluents are calculated monthly for the fish and drinking water ingestion pathways and the external sediment exposure pathways from all detected nuclides in liquid effluents released to the unrestricted areas using the following expression from NUREG 0133, Section 4.3.

Dt = Yi [Ai, EL (ATLCiLFL) I Where:

Dt = The cumulative dose commitment to the total body or any organ, t from the liquid effluents for the total time period YL (ATL), mrerm ATL = The length of the L th time period over which CiL and FL are averaged for all liquid releases, hours CiL The average concentration of radionuclide, i, in undiluted liquid effluents during time period ATL from any liquid release, ýtCi/ml Unit 2 Revision 22 II 8 December 2001

A =, The site related ingestion dose commitment factor for the maximum individual to the total body or any or gan t for each identified principal gamma or beta emitter, mrem/hr per pCi/ml. Table D 2-2.

FL The near field average dilution factor for Cil during any liquid effluent release. Defined as the ratio of the maximum undiluted liquid waste flow during release to the product of the average flow from the site discharge structure to unrestricted receiving waters times 5.9. (5.9 is the site specific applicable factor for the mixing effect of the discharge structure.)

See the Nine Mile Point Unit 2 Environmental Report - Operating License Stage, Table 5.4-2 footnote 1.

These factors can be related to batch release parameters as follows:

FL = PEF / (TDF x 5.9) (Terms defined in Section 1.1.2.1 and above)

ATLFL = [PEF (gpm) x ATL (min) x 1.67E-2 (hr/min)] / [TDF (gpm) x 5.9]

= [TV x 2.83E-3 (hours)] / TDF For each batch, PEF (gpm) x ATL (min) = Tank Volume. For each batch, a dose calculation common constant (ATLFL) is calculated to be used with the concentration of each nuclide and dose factor, A1 , to calculate the dose to a receptor. Normally, the highest dose factor for any age group (adult, teen, child, infant) will be used for calculation, but specific age-group calculations to demonstrate compliance may be performed if required.

1.4 Liquid Effluent Sampling Representativeness There are four tanks in the radwaste system designed to be discharged to the discharge canal. These tanks are labeled 4A, 4B, 5A, and 5B.

Liquid Radwaste Tank 5A and 5B at Nine Mile Point Unit 2 contain a sparger spray ring which assists the mixing of the tank contents while it is being recirculated prior to sampling. This sparger effectively mixes the tank four times faster than simple recirculation.

Liquid Radwaste Tank 4A and 4B contain a mixing ring but no sparger. No credit is taken for the mixing effects of the ring. Normal recirculation flow is 150 gpm for tank 5A and 5B, 110 gpm for tank 4A and 4B while each tank contains up to 25,000 gallons although the entire contents are not discharged. To assure that the tanks are adequately mixed prior to sampling, it is a plant requirement that the tank be recirculated for the time required to pass 2.5 times the volume of the tank:

Recirculation Time = 2.5T/RM Unit 2 Revision 22 119 December 2001

Where:

Recirculation Time = Is the minimum time to recirculate the Tank, min 2.5 = Is the plant requirement, unitless T = Is the tank volume, gal R = Is the recirculation flow rate, gpm.

M = Is the factor that takes into account the mixing of the sparger, unitless, four for tank 5A and B, one for tank 4A and B.

Additionally, the Alert Alarm setpoint of the Liquid Radwaste Effluent monitor is set at approximately 60 % of the High alarm setpoint. This alarm will give indication of incomplete mixing with adequate margin before exceeding ten times the effluent concentration.

Service Water A and B and the Cooling Tower Blowdown are sampled from the radiation monitor on each respective stream. These monitors continuously withdraw a sample and pump it back to the effluent stream. The length of tubing between the continuously flowing sample and the sample spigot contains less than 200 ml which is adequately purged by requiring a purge of at least 1 liter when grabbing a sample.

1.5 Liquid Radwaste System Operability The Liquid Radwaste Treatment System shall be OPERABLE and used when projected doses due to liquid radwaste effluents would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period. Cumulative doses will be determined at least once per 31 days (as indicated in Section 1.3) and doses will also be projected if the radwaste treatment systems are not being fully utilized.

The system collection tanks are processed as follows:

1) Low Conductivity (Waste Collector): Radwaste Filter and Radwaste Demineralizer or the Thermex System.

2) High Conductivity (Floor Drains): Regenerant Evaporator or the Thermex System.

3) Regenerant Waste: If resin regeneration is used at NMP-2; the waste will be processed through the regenerant evaporator or Thermex System.

The dose projection indicated above will be performed in accordance with the methodology of Section 1.3.

Unit 2 Revision 22 11 10 December 2001

2.0 GASEOUS EFFLUENTS The gaseous effluent release points are the stack and the combined Radwaste/Reactor Building vent. The stack effluent point includes Turbine Building ventilation, main condenser offgas (after charcoal bed holdup), and Standby Gas Treatment System exhaust. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.

2.1 Gaseous Effluent Monitor Alarm Setpoints 2.1.1 Basis The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited to the following in accordance with Technical Specification 5.5.4. g:

a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mrem/yr to the skin, and

b. For iodine-131, for iodine-133, for tritium, and for all radionuclides with half lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.

The radioactivity rate of noble gases measured downstream of the recombiner shall be limited to less than or equal to 350,000 microcuries/second during offgas system operation in accordance with Technical Specification 3.7.4.

2.1.2 Setpoint Determination Methodology Discussion Nine Mile Point Unit 1 and the James A FitzPatrick nuclear plants occupy the same site as Nine Mile Point Unit 2. Because of the independence of these plants' safety systems, control rooms and operating staffs it is assumed that simultaneous accidents are not likely to occur at the different units. However, there are two release points at Unit 2. It is assumed that if an accident were to occur at Unit 2 that both release points could be involved.

The alarm setpoint for Gaseous Effluent Noble Gas Monitors are based on a dose rate limit of 500 mRem/yr to the Whole Body. Since there are two release points at Unit 2, the dose rate limit of 500 mRem/yr is divided equally for each release point, but may be apportioned otherwise, if required. These monitors are sensitive to only noble gases. Because of this it is considered impractical to base their alarm setpoints on organ dose rates due to iodines or particulates. Additionally skin dose rate is never significantly greater than the whole body dose rate. Thus the factor R which is the basis for the alarm setpoint calculation is nominally taken as equal to 250 mRem/yr.

If there are significant releases from any gaseous release point on the site (> 25 mRem/yr) for an extended period of time then the setpoint will be recalculated with an appropriately smaller value for R.

The high alarm setpoint for the Offgas Noble Gas monitor is based on a limit of 350,000 uCi/sec. This is the release rate for which a FSAR accident analysis was Unit 2 Revision 22 I 11 December 2001

completed. At this rate the Offgas System charcoal beds will not contain enough activity so that their failure and subsequent release of activity will present a significant offsite dose assuming accident meteorology.

Initially, in accordance with Part I, Section D 3.3.2, the Germanium multichannel analysis systems of the stack and vent will be calibrated with gas standards (traceable to NIST) in accordance with DSR 3.3.2.9. Subsequent calibrations may be performed with gas standards, or with related solid sources. The quarterly Channel Functional Test will include operability of the 30cc chamber and the dilution stages to confirm monitor high range capability. (Appendix D, Gaseous Effluent Monitoring System).

2.1.2.1 Stack Noble Gas Detector Alarm Setpoint Equation:

The stack at Nine Mile Point Unit 2 receives the Offgas after charcoal bed delay, Turbine Building Ventilation and the Standby Gas Treatment system exhaust. The Standby Gas Treatment System Exhausts the primary containment during normal shutdowns and maintains a negative pressure on the Reactor Building to maintain secondary containment integrity. The Standby Gas Treatment will isolate on high radiation detected (by the SGTS monitor) during primary containment purges.

The stack noble gas detector is made of germanium. It is sensitive to only gamma radiation. However, because it is a computer based multichannel analysis system it is able to accurately quantify the activity released in terms of uCi of specific nuclides.

Only pure alpha and beta emitters are not detectable, of which there are no common noble gases. A distribution of Noble Gases corresponding to offgas is chosen for the nominal alarm setpoint calculation. Offgas is chosen because it represents the most significant contaminant of gaseous activity in the plant. The release rate Qj, corresponds to offgas concentration expected with the plant design limit for fuel failure. The alarm setpoint may be recalculated if a significant release is encountered.

In that case the actual distribution of noble gases will be used in the calculation.

The following calculation will be used for the initial Alarm Setpoint.

0.8R Y.i(Qi)

Alarm Setpoint, .tCi/sec < Yi(QiVi) 0.8 = Safety Factor, unitless R = Allocation Factor. Normally, 250 mrem/yr; the value must be 500 mrem/yr or less depending upon the dose rate from other release points within the site such that the total dose rate corresponds to

< 500 mrem/yr QI = The release rate of nuclide i, ýCi/sec Vi = The constant for each identified noble gas nuclide accounting for the whole body dose from the elevated finite plume listed on Table D 3-2, mrem/yr per gCi/sec Unit 2 Revision 22 1112 December 2001

i (Qi) = The total release rate of noble gas nuclides in the stack effluent,

ýItCi/sec Ei (QiVi) - The total of the product of each isotope release rate times its respective whole body plume constant, mrem/yr, PCi/sec The alert alarm is normally set at less than 10% of the high alarm.

2.1.2.2 Vent Noble Gas Detector Alarm Setpoint Equation:

The vent contains the Reactor Building ventilation above and below the refuel floor and the Radwaste Building ventilation effluents. The Reactor Building Ventilation will isolate when radiation monitors detect high levels of radiation (these are separate monitors, not otherwise discussed in the ODCM). Nominal flow rate for the vent is 2.37E5 CFM.

This detector is made of germanium. It is sensitive to only gamma radiation.

However, because it is a computer based multichannel analysis system it is able to accurately quantify the activity released in terms of *Ci of specific nuclides. Only pure alpha and beta emitters are not detectable, of which there are no common noble gases. A distribution of Noble Gases corresponding to that expected with the design limit for fuel failure offgas is chosen for the nominal alarm setpoint calculation.

Offgas is chosen because it represents the most significant contaminant of gaseous activity in the plant. The alarm setpoint may be recalculated if a significant release is encountered. In that case the actual distribution of noble gases will be used in the calculation.

0.8R Yj (Qi)

Alarm Setpoint, uCi/sec < (X/Q) v Yi (QiKi)

Where:

0.8 = Safety Factor, unitless R = Allocation Factor. Normally, 250 mrem/yr; the value must be 500 mrem/yr or less depending upon the dose rate from other release points within the site such that the total rate corresponds to < 500 mrem/yr QI = The release rate of nuclide i, p Ci/sec (X/Q), = The highest annual average atmospheric dispersion coefficient at the site boundary as listed in the Final Environmental Statement, NUREG 1085, Table D-2, 2.OE-6 sec/m 3 Ki The constant for each identified noble gas nuclide accounting for the whole body dose from the semi-infinite cloud, listed on Table D 3-3, mrem/yr per tCi/m3 Unit 2 Revision 22 1113 December 2001

zi (Qi) - The total release rate of noble gas nuclides in the vent effluent, uCi/sec Yi (QiKj) The total of the product of the each isotope release rate times its respective whole body immersion constant, mrem/yr per sec/m3 The alert alarm is normally set at less than 10% of the high alarm.

2.1.2.3 Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation:

The Offgas system has a radiation detector downstream of the recombiners and before the charcoal decay beds. The offgas, after decay, is exhausted to the main stack. The system will automatically isolate if its pretreatment radiation monitor detects levels of radiation above the high alarm setpoint.

The Radiation Detector contains a plastic scintillator disc. It is a beta scintillation detector. Detector response Yj (Ci/CFi) has been evaluated from isotopic analysis of offgas analyzed on a multichannel analyzer, traceable to NIST. A distribution of offgas corresponding to that expected with the design limit for fuel failure was used to establish the initial setpoint. However, the alarm setpoint may be recalculated using an updated nuclide distribution based on actual plant process conditions. The monitor nominal response values will be confirmed during periodic calibration using a Transfer Standard source traceable to the primary calibration performed by the vendor.

Particulates and Iodines are not included in this calculation because this is a noble gas monitor.

To provide an alarm in the event of failure of the offgas system flow instrumentation, the low flow alarm setpoint will be set at or above 10 scfm, (well below normal system flow) and the high flow alarm setpoint will be set at or below 110 scfm, which is well above expected steady-state flow rates with a tight condenser.

To provide an alarm for changing conditions, the alert alarm will normally be set at 1.5 times nominal full power background to ensure that the Specific Activity Action required by ITS SR 3.7.4.1, are implemented in a timely fashion.

(3.50E+05) (2.12 E-03) Y,*(C_/CF_ )

Alarm Setpoint, cpm < 0.8 F Y (Ci) + Background Where:

Alarm Setpoint = The alarm setpoint for the offgas pretreatment Noble Gas Detector, cpm 0.8 - Safety Factor, unitless Unit 2 Revision 22 1114 December 2001

350,000 The Technical Specification Limit for Offgas Pretreatment,

ýtCi/sec 2.12E-03 = Unit conversion Factor, 60 sec/min / 28317 cc/CF Ci = The concentration of nuclide, i, in the Offgas, pCi/cc CFi = The Detector response to nuclide i, ptCi/cc/cpm; See Table D 3-1 for a list of nominal values F = The Offgas System Flow rate, CFM Background = The detector response to non-fission gases and general area dose rates, cpm Y+/- (cJ/CFi) The summation of the nuclide concentration divided by the corresponding detector response, net cpm Y'i (Ci) The summation of the concentration of nuclides in offgas, 1ICi/cc 2.2 Gaseous Effluents Dose Rate Calculation Dose rates will be calculated monthly at a minimum to demonstrate that the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the dose rate limits specified in 10CFR20. These limits are as follows:

The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) shall be limited per Technical Specification 5.5.4.g to the following:

a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mrem/yr to the skin, and

b. For iodine-131, iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500.mrem/yr to any organ:

2.2.1 X/Q and W, - Dispersion Parameters for Dose Rate, Table D 3-23 The dispersion parameters for the whole body and skin dose rate calculation correspond to the highest annual average dispersion parameters at or beyond the unrestricted area boundary. This is at the east site boundary. These values were obtained from the Nine Mile Point Unit 2 Final Environmental Statement, NUREG 1085 Table D-2 for the vent and stack. These were calculated using the methodology of Regulatory Guide 1.111, Rev. 1. The stack was modeled as an elevated release point because its height is more than 2.5 times any adjacent building height. The vent was modeled as a ground level release because even though it is higher than any adjacent building it is not more than 2.5 times the height.

The NRC Final Environmental Statement values for the site boundary X/Q and D/Q terms were selected for use in calculating Effluent Monitor Alarm Points and compliance with Site Boundary Dose Rate specifications because they are conservative Unit 2 Revision 22 11 15 December 2001

when compared with the corresponding Nine Mile Point Environmental Report values.

In addition, the stack "intermittent release" X/Q was selected in lieu of the "continuous" value, since it is slightly larger, and also would allow not making a distinction between long term and short term releases.

The dispersion parameters for the organ dose calculations were obtained from the Environmental Report, Figures 7B-4 (stack) and 7B-8 (vent) by locating values corresponding to currently existing (1985) pathways. It should be noted that the most conservative pathways do not all exist at the same location. It is conservative to assume that a single individual would actually be at each of the receptor locations.

2.2.2 Whole Body Dose Rate Due to Noble Gases The ground level gamma radiation dose from a noble gas stack release (elevated),

referred to as plume shine, is calculated using the dose factors from Appendix B of this document. The ground level gamma radiation dose from a noble gas vent release accounts for the exposure from immersion in the semi-infinite cloud. The dispersion of the cloud from the point of release to the receptor at the east site boundary is of factored into the plume shine dose factors for stack releases and through the use X/Q in the equation for the immersion ground level dose rates for vent releases. The release rate is averaged over the period of concern. The factors are discussed in Appendix B.

Whole body dose rate (DR)y due to noble gases:

(DR) y = 3.17E-08 Yj [ViQis + Ki (X/Q)vQiv]

Where:

DRy Whole body dose rate (mrem/sec)

V1 The constant accounting for the gamma whole body dose rate from the finite plume from the elevated stack releases for each identified noble gas nuclide, i. Listed on Table D 3-2, mrem/yr per ,Ci/sec Ki The constant accounting for the gamma whole body dose rate from immersion in the semi-infinite cloud for each identified noble gas 3

nuclide, i. Listed in Table D 3-3, mrem/yr per uCi/m (From Reg.

Guide 1.109)

X/Q, The relative plume concentration at or beyond the X/Q, land sector site boundary. Average meteorological data is used.

Elevated X/Q values are used for the stack releases (s =stack);

ground X/Q values are used for the vent releases (v-=vent). Listed on Table D 3-23 (s) or QisQ=v The release rate of each noble gas nuclide i, from the stack vent (v). Averaged over the time period of concern. (tCi/sec)

Unit 2 Revision 22 1116 December 2001

3.17E-08 = Conversion Factor; the inverse of the number of seconds in one year. (yr/sec) 2.2.3 Skin Dose Rate Due to Noble Gases There are two types of radiation from noble gas releases that contribute to the skin dose rate: beta and gamma.

For stack releases this calculation takes into account the dose from beta radiation in a semi infinite cloud by using an immersion dose factor. Additionally, the dispersion of the released activity from the stack to the receptor is taken into account by use of the factor (X/Q). The gamma radiation dose from the elevated stack release is taken into account by the dose factors in Appendix B.

For vent releases the calculations also take into account the dose from the beta (p) and gamma (y) radiation of the semi infinite cloud by using an immersion dose factor.

Dispersion is taken into account by use of the factor (X/Q).

The release rate is averaged over the period of concern.

Skin dose rate (DR)0, 0 due to noble gases:

(DR) yp = 3.17E-8 iH[ (Li(X/Q)s+1.11Bi)Qis+(Li+l.11Mi) (X/Q) Qiv]

Where:

(DR) , = Skin dose rate (mrem/sec)

Li The constant to account for the gamma and beta skin dose rates for each noble gas nuclide, i, from immersion in the semi-infinite cloud, mrem/yr per 1 /Ci/m 3 , listed on Table D 3-3 (from R.G. 1.109)

Mi The constant to account for the air gamma dose rate for each noble gas nuclide, i, from immersion in the semi-infinite cloud, mrad/yr per pCi/m 3, listed on Table D 3-3 (from R.G. 1.109) 1.11 = Unit conversion constant, mrem/mrad

.7 = Structural shielding factor, unitless B, = The constant accounting for the air gamma dose rate from exposure to the overhead plume of elevated releases of each identified noble gas nuclide, i. Listed on Table D 3-2, mrad/yr per gCi/sec.

Unit 2 Revision 22 1117 December 2001

(X/Q), The relative plume concentration at or beyond the land Elevated (X/Q)X sector site boundary. Average meteorological data is used.

X/Q values are used for the stack releases (s = stack); ground X/Q values are used for the vent releases (v =vent).

3.17E-8 = Conversion Factor; the inverse of the number of seconds in a year; (yr/sec)

Qiv=s The release rate of each noble gas nuclide i, from the stack(s) or vent (v) averaged over the time period of concern, gtCi/sec.

greater 2.2.4 Organ Dose Rate Due to 1-131, 1-133, Tritium, and Particulates with Half-lives than 8 days.

C. The The organ dose rate is calculated using the dose factors (Ri) from Appendix inhalation and factor Ri takes into account the dose rate received from the ground plane, from the ingestion pathways. W, and Wv take into account the atmospheric dispersion respective release point to the location of the most conservative receptor for each of the pathways. The release rate is averaged over the period of concern.

in Organ dose rates (DR)at due to iodine-131, iodine-133, tritium and all radionuclides particulate form with half-lives greater than 8 days:

(DR) at = 3.17E-8 Y[jEiRijat [W3Qis + WvQivl I Where:

(DR)at = Organ dose rate (mrem/sec)

Rijat - The factor that takes into account the dose from nuclide i through pathway j to an age group a, and individual organ t. Units for 3

inhalation pathway, mrem/yr per PCi/m . Units for ground and ingestion pathways, m-mrem/yr per uCi/sec. See Tables D 3-4 through D 3-22).

3 W3 , WV Dispersion parameter either X/Q (sec/m ) or D/Q (1/m2) depending on pathway and receptor location. Average meteorological data is used (Table D 3-23). Elevated W, values are used for stack releases (s=stack); ground Wv values are used for vent releases (v=vent).

QiS, Q" The release rates for nuclide i, from the stack (s) and vent (v) respectively, ý,Ci/sec.

dose rate When the release rate exceeds 0.75 uCi/sec from the stack or vent, the The use of the 0.75 assessment shall, also, include JAF and NMP1 dose contributions.

because it is based on the dose conversion kICi/sec release rate threshold is conservative Unit 2 Revision 22 1118 December 2001

factor (Ri) for the Sr-90 child bone which is significantly higher than the dose factors for the other isotopes present in the stack or vent release.

2.3 Gaseous Effluent Dose Calculation Methodology Doses will be calculated monthly at a minimum to demonstrate that doses resulting from the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the limits specified in 10 CFR 50. These limits are as follows:

The air dose from noble gases released in gaseous effluents, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) 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.

The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radioactive material in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) 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.

The VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE and appropriate portions of this system shall be used to reduce releases of radioactivity when the projected doses in 31 days from iodine and particulate releases, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure D 1.0-1) would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

2.3.1 W, and W, - Dispersion Parameters for Dose, Table D 3-23 The dispersion parameters for dose calculations were obtained chiefly from the Nine Mile Point Unit 2 Environmental Report Appendix 7B. These were calculated using the methodology of Regulatory Guide 1.111 and NUREG 0324. The stack was modeled as an elevated release point because height is more than 2.5 times the height of any adjacent building. The vent was modeled as a combined elevated/ground level release because the vent's height is not more than 2.5 times the height of any adjacent building.

Average meteorology over the appropriate time period was used. Dispersion parameters not available from the ER were obtained from C.T. Main Data report dated November, 1985, or the FES.

Unit 2 Revision 22 1119 December 2001

2.3.2 Gamma Air Dose Due to Noble Gases Gamma air dose from the stack or vent noble gas releases is calculated monthly. The gamma air dose equation is similar to the gamma dose rate equation except the receptor is air instead of the whole body or skin of whole body. Therefore, the stack noble gas releases use the finite plume air dose factors, and the vent noble gas releases use semi infinite cloud immersion dose factors. The factor X/Q takes into account the dispersion of vent releases to the most conservative location. The release activity is totaled over the period of concern. The finite plume factor is discussed in Appendix B.

Gamma air dose due to noble gases:

DY = 3.17E-8 Yi [Mi (X/Q) v Qiv + Bi Qis] x t O- The gamma air dose for the period of concern, mrad t = The duration of the dose period of concern, sec Where all other parameters have been previously defined.

2.3.3 Beta Air Dose Due to Noble Gases The beta air dose from the stack or vent noble gas releases is calculated using the semi infinite cloud immersion dose factor in beta radiation. The factor X/Q takes into account the dispersion of releases to the most conservative location.

Beta air dose due to noble gases:

DP - 3.17E-8 ZiNi[ (X/Q)v Qiv + (X/Q), Qi,] X t D

= Beta air dose (mrad) for the period of concern N = The constant accounting for the beta air dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed on Table D 3-3, mrad/yr per uCi/m 3 . (From Reg. Guide 1.109).

t The duration of the dose period of concern, sec Where all other parameters have been previously defined.

2.3.4 Organ Dose Due to 1-131, 1-133, Tritium and Particulates with half-lives greater than 8 days.

The organ dose is based on the same equation as the dose rate equation except the dose is compared to the 10CFR50 dose limits. The factor R, takes into account the dose received from the ground plane, inhalation, food (cow milk, cow meat and vegetation) pathways. Ws and W, take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. The release is totaled over the period of concern. The R, factors are discussed in Appendix C.

Unit 2 Revision 22 1120 December 2001

Organ dose Dat due to iodine-131, iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days.

Dat 3.17E-8 *j [ Yi Rijat [Ws Qis + Wv Qiv]] x t Where:

Dat = Dose to the critical organ t, for age group a, mrem t = The duration of the dose period of concern, sec Where all other parameters have been previously defined in Section 2.2.4.

2.4 1-133 and 1-135 Estimation Stack and vent effluent iodine cartridges are analyzed to a sensitivity of at least 1E-12 uCi/cc. If detected in excess of the LLD, the 1-131 and 1-133 analysis results will be a

reported directly from each cartridge analyzed. Periodically, (usually quarterly but on monthly frequency if effluent iodines are routinely detected) a short-duration (12 to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) effluent sample is collected and analyzed to establish an 1-135/1-131 ratio and an 1-133/1-131 ratio, if each activity exceeds LLD. The short-duration ratio is used to confirm the routinely measured 1-133 values. The short-duration 1-135/1-131 ratio (if determined) is used with the 1-131 release to estimate the 1-135 release. The short duration 1-133/1-131 ratio may be used with the 1-131 release to estimate the 1-133 release if the directly measured 1-133 release appears non-conservative.

2.5 Isokinetic Sampling Sampling systems for the stack and vent effluent releases are designed to maintain isokinetic sample flow at normal ventilation flow rates. During periods of reduced ventilation flow, sample flow may be maintained at a minimum flow rate (above the calculated isokinetic rate) in order to minimize sample line losses due to particulate deposition at low velocity.

2.6 Use of Concurrent Meteorological Data vs. Historical Data set It is the intent to use dispersion parameters based on historical meteorological data to to alarm points and to determine or predict dose and dose rates in the environment due gaseous effluents. If effluent levels approach limiting values, meteorological conditions concurrent with the time of release may be used to determine gaseous pathway doses.

2.7 Gaseous Radwaste Treatment System Operation Part I, Section D 3.2.4 requires the GASEOUS RADWASTE TREATMENT SYSTEM The to be in operation whenever the main condenser air ejector system is in operation.

facilitate system may be operated for short periods with the charcoal beds bypassed to Unit 2 Revision 22 1121 December 2001

transients. The components of the system which normally should operate to treat offgas are the Preheater, Recombiner, Condenser, Dryer, Charcoal Adsorbers, HEPA Filter, and Vacuum Pump. (See Appendix D, Offgas System).

2.8 Ventilation Exhaust Treatment System Operation Part I, Section D 3.2.5 requires the VENTILATION EXHAUST TREATMENT SYSTEM to be OPERABLE when projected doses in 31 days due to iodine and particulate releases would exceed 0.3 mrem to any organ of a member of the public.

The appropriate components, which affect iodine or particulate release, to be OPERABLE are:

1) HEPA Filter - Radwaste Decon Area

2) HEPA Filter - Radwaste Equipment Area

3) HEPA Filter - Radwaste General Area Whenever one of these filters is not OPERABLE, iodine and particulate dose projections will be made for 31-day intervals starting with filter inoperability, and continuing as long as the filter remains inoperable, in accordance with DSR 3.2.5.1. Predicted release rates will be used, along with the methodology of Section 2.3.4. (See Appendix D, Gaseous Radiation Monitoring.)

Unit 2 Revision 22 1122 December 2001

3.0 URANIUM FUEL CYCLE The "Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b) as follows:

"Uranium fuel cycle means the operations of milling of uranium ore chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the reuse of recovered non-uranium special nuclear and by-product materials from the cycle."

Sections D 3.1.2, D 3.2.2, and D 3.2.3 of Part I requires that when the calculated doses associated with the effluent releases exceed twice the applicable quarter or annual limits, the licensee shall evaluate the calendar year doses and, if required, submit a Special Report to the NRC and limit subsequent releases such that the dose commitment to a real individual from all uranium fuel cycle sources is limited to 25 mrem to the total body or any organ (except the thyroid, which is limited to 75 mrem). This report is to demonstrate that radiation exposures to all real individuals from all uranium fuel cycle sources (including all liquid and gaseous effluent pathways and direct radiation) are less than the limits in 40 CFR Part 190. If releases that result in doses exceeding the 40 CFR 190 limits have occurred, then a variance from the NRC to permit such releases will be requested and if possible, action will be taken to reduce subsequent releases.

The report to the NRC shall contain:

1) Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site, that contribute to the annual dose of the maximum exposed member of the public.

2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from all existing pathways and sources of radioactive effluents and direct radiation.

The total body and organ doses resulting from radioactive material in liquid effluents from Nine Mile Point Unit 2 will be summed with the doses resulting from the releases of noble gases, radioiodines, and particulates. The direct dose components will also be determined by either calculation or actual measurement. Actual measurements will utilize environmental TLD dosimetry. Calculated measurements will utilize engineering calculations to determine a projected direct dose component. In the event calculations are used, the methodology will be detailed as required by Technical Specification 5.6.3.

The doses from Nine Mile Point Unit 2 will be added to the doses to the maximum exposed individual that are contributed from other uranium fuel cycle operations within 5 miles of the site.

Unit 2 Revision 22 1123 December 2001

For the purpose of calculating doses, the results of the Environmental Monitoring Program may be included to provide more refined estimates of doses to a real maximum exposed individual. Estimated doses, as calculated from station effluents, may be replaced by doses calculated from actual environmental sample results.

3.1 Evaluation of Doses From Liquid Effluents For the evaluation of doses to real members of the public from liquid effluents, the fish consumption and shoreline sediment ground dose will be considered. Since the doses from other aquatic pathways are insignificant, fish consumption and shoreline sediment are the only two pathways that will be considered. The dose associated with fish consumption may be calculated using effluent data and Regulatory Guide 1.109 methodology or by calculating a dose to man based on actual fish sample analysis data.

Because of the nature of the receptor location and the extensive fishing in the area, the critical individual may be a teenager or an adult. The dose associated with shoreline sediment is based on the assumption that the shoreline would be utilized as a recreational area. This dose may be derived from liquid effluent data and Regulatory Guide 1.109 methodology or from actual shoreline sediment sample analysis data.

Equations used to evaluate fish and shoreline sediment samples are based on Regulatory Guide 1.109 methodology. Because of the sample medium type and the half-lives of the radionuclides historically observed, the decay corrected portions of the equations are deleted. This does not reduce the conservatism of the calculated doses but increases the simplicity from an evaluation point of view. Table D 3-24 presents the parameters used for calculating doses from liquid effluents.

The dose from fish sample media is calculated as:

Rapj = Yi [Cif (U) (Daipj) f] (IE+3)

Where:

RP=j The total annual dose to organ j, of an individual of age group a, from nuclide i, via fish pathway p, in mrem per year; ex. if calculating to the adult whole body, then RPpj = Rb and Daipj = DjwB Cif= The concentration of radionuclide i in fish samples in pCi/gram U = The consumption rate of fish 1E+3 = Grams per kilogram (Daipj) The ingestion dose factor for age group a, nuclide i, fish pathway p, and organ j, (Reg. Guide 1.109, Table E- 11) (mrem/pCi). ex. when calculating to the adult whole body Dipj = DjwB f The fractional portion of the year over which the dose is applicable Unit 2 Revision 22 II 24 December 2001

The dose from shoreline sediment sample media is calculated as:

Rapj = Zi

Ci (u) (4E+4) (0.3) (Daipj) f]

Where:

RaPj The total annual dose to organ j, of an individual of age group a, from nuclide i, via the sediment pathway p, in mrem per year; ex. if calculating to the adult whole body, then Rpj = RwB and Daipj = DjwB Cis The concentration of radionuclide i in shoreline sediment in pCi/gram U - The usage factor, (hr/yr) (Reg. Guide 1. 109) 4E+4 - The product of the assumed density of shoreline sediment (40 kilogram per square meter to a depth of 2.5 cm) times the number of grams per kilogram 0.3 = The shore width factor for a lake Daipj The dose factor for age group a, nuclide i, sediment pathway s, and 2

organ j. (Reg. Guide 1.109, Table E-6) (mrem/hr per pCi/mi ); ex.

when calculating to the adult whole body Da 5 pi = DjwB f = The fractional portion of the year over which the dose is applicable NOTE: Because of the nature of the receptor location and the extensive fishing in the area, the critical individual may be a teenager or an adult.

Unit 2 Revision 22 1125 December 2001

3.2 Evaluation of Doses From Gaseous Effluents For the evaluation of doses to real members of the public from gaseous effluents, the pathways contained in section 2 of the calculational methodologies section will be considered and include ground deposition, inhalation, cows milk, goats milk, meat, and food products (vegetation). However, any updated field data may be utilized that concerns locations of real individuals, real time meteorological data, location of critical receptors, etc. Data from the most recent census and sample location surveys should be utilized. Doses may also be calculated from actual environmental sample media, as available. Environmental sample media data such as TLD, air sample, milk sample and vegetable (food crop) sample data may be utilized in lieu of effluent calculational data.

Doses to members of the public from the pathways considered in section 2 as a result of gaseous effluents will be calculated using the methodology of Regulatory Guide 1.109 or the methodology of the ODCM, as applicable. Doses calculated from environmental sample media will be based on methodologies found in Regulatory Guide 1.109.

3.3 Evaluation of Doses From Direct Radiation The dose contribution as a result of direct radiation shall be considered when evaluating whether the dose limitations of 40 CFR 190 have been exceeded. Direct radiation doses as a result of the reactor, turbine and radwaste buildings and outside radioactive storage tanks (as applicable) may be evaluated by engineering calculations or by evaluating environmental TLD results at critical receptor locations, site boundary or other special interest locations. For the evaluation of direct radiation doses utilizing environmental TLDs, the critical receptor in question, such as the critical residence, etc., will be compared to the control locations.

The comparison involves the difference in environmental TLD results between the receptor location and the average control location result.

3.4 Doses to Members of the Public Within the Site Boundary The Radioactive Effluent Release Report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary as defined by Figure D 1.0-1. A member of the public, would be represented by an individual who visits the sites' Energy Center for the purpose of observing the educational displays or for picnicking and associated activities.

Fishing is a major recreational activity in the area and on the Site as a result of the salmon and trout populations in Lake Ontario. Fishermen have been observed fishing at the shoreline near the Energy Center from April through December in all weather conditions. Thus, fishing is the major activity performed by members of the public within the site boundary. Based on the nature of the fishermen and undocumented observations, it is conservatively assumed that the maximum exposed individual spends an average of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week fishing from the shoreline at a location between the Unit 2 Revision 22 II 26 December 2001

Energy Center and the Unit 1 facility. This estimate is considered conservative but not necessarily excessive and accounts for occasions where individuals may fish more on weekends or on a few days in March of the year.

The pathways considered for the evaluation include the inhalation pathway with the resultant lung dose, the ground dose pathway with the resultant whole body and skin dose and the direct radiation dose pathway with the associated total body dose. The direct radiation dose pathway, in actuality, includes several pathways. These include:

the direct radiation gamma dose to an individual from an overhead plume, a gamma submersion plume dose, possible direct radiation dose from the facility and a ground plane dose (deposition). Because the location is in close proximity to the site, any beta plume submersion dose is felt to be insignificant.

Other pathways, such as the ingestion pathway, are not applicable. In addition, pathways associated with water related recreational activities, other than fishing, are not applicable here. These include swimming, boating and wading which are prohibited at the facility.

The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium and particulates) in the effluent for the appropriate time period.

The radionuclide concentrations are then multiplied by the appropriate X/Q value, inhalation dose factor, air intake rate, and the fractional portion of the year in question.

Thus, the inhalation pathway is evaluated using the following equation adapted from Regulatory Guide 1.109. Table D 3-24 presents the reference for the parameters used in the following equation.

NOTE: The following equation is adapted from equations C-3 and C-4 of 3 Regulatory Guide 1.109. Since many of the factors are in units of pCi/m ,

m3/sec., etc., and since the radionuclide decay expressions have been deleted because of the short distance to the receptor location, the equation presented here is not identical to the Regulatory Guide equations.

Dja = i [ (Ci) F (X/Q) (DFA) ija (BR) at]

Where:

Dia The maximum dose from all nuclides to the organ j and age group (a) in mrem/yr; ex. if calculating to the adult lung, then Dja = DL and DFA.j. = DFAiL C, The average concentration in the stack or vent release of nuclide i for the period in pCi/m 3 .

3 F = Unit 2 average stack or vent flowrate in m /sec.

Unit 2 Revision 22 1127 December 2001

X/Q The plume dispersion parameter for a location approximately 0.50 miles west of NMP-2 (The plume dispersion parameters are 9.6E-07 (stack) and 2.8E-06 (vent) and were obtained from the C.T. Main five year average annual X/Q tables.

The vent X/Q (ground level) is ten times the listed 0.50 mile X/Q because the vent is approximately 0.3 miles from the receptor location. The stack (elevated) X/Q is conservative when based on 0.50 miles because of the close proximity of the stack and the receptor location.

(DFA)ija the dose factor for nuclide i, organ j, and age group a in mrem per pCi (Reg. Guide 1.109, Table E-7); ex. if calculating to the adult lung the DFAija = DFAiL 3

(BR)a annual air intake for individuals in age group a in M per year (obtained from Table E-5 of Regulatory Guide 1.109).

t = fractional portion of the year for which radionuclide i was detected and for which a dose is to be calculated (in years).

The ground dose pathway (deposition) will be evaluated by obtaining at least one soil or shoreline sediment sample in the area where fishing occurs. The dose will then be calculated using the sample results, the time period in question, and the methodology based on Regulatory Guide 1.109 as presented in Section 3.1. The resultant dose may be adjusted for a background dose by subtracting the applicable off-site control soil or shoreline sediment sample radionuclide activities. In the event it is noted that fishing is not performed from the shoreline but is instead performed in the water (i.e., the use of waders), then the ground dose pathway (deposition) will not be evaluated.

The direct radiation gamma dose pathway includes any gamma doses from an overhead plume, submersion in the plume, possible radiation from the facility and ground plane dose (deposition). This general pathway will be evaluated by average environmental TLD readings. At least two environmental TLDs will be used at one location in the approximate area where fishing occurs. The TLDs will be placed in the field on approximately the beginning of each calendar quarter and removed approximately at the end of each calendar quarter (quarter 2, 3, and 4).

The average TLD readings will be adjusted by the average control TLD readings. This is accomplished by subtracting the average quarterly control TLD value from the average fishing location TLD value. The applicable quarterly control TLD values will be used after adjusting for the appropriate time period (as applicable). In the event of loss or theft of the TLDs, results from a TLD or TLDs in a nearby area may be utilized.

Unit 2 Revision 22 II 28 December 2001

4.0 ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Stations The current sampling locations are specified in Table D 5-1 and Figures D 5.1-1 and D 5.1-2. The meteorological tower location is shown on Figure D 5.1-1 and is located where TLD location #17 is identified. The Environmental Monitoring Program is a joint effort between the owners and operators of the Nine Mile Point Units 1 and 2 and the James A. FitzPatrick Nuclear Power Plants. Sampling locations are chosen on the basis of historical average dispersion or deposition parameters from both units. The environmental sampling location coordinates shown on Table D 5-1 are based on the NMP-2 reactor centerline.

The average dispersion and deposition parameters for the three units have been calculated for a 5 year period, 1978 through 1982. Average dispersion or deposition parameters for the site are calculated using the 1978 through 1982 data and are used to compare the results of the annual land use census. If it is determined that sample locations required by Control D 3.5.1 are unavailable or new locations are identified that yield a significantly higher (i.e., 50%) calculated D/Q value, actions will be taken as required by Controls D 3.5.1 and D 3.5.2 and the Radiological Environmental Monitoring Program updated accordingly.

4.2 Interlaboratory Comparison Program Analyses shall be performed on samples containing known quantities of radioactive materials that are supplied as part of a Commission approved or sponsored Interlaboratory Comparison Program, such as the EPA Crosscheck Program.

Participation shall be only for those media, e.g., air, milk, water, etc., that are included in the Nine Mile Point Environmental Monitoring Program and for which cross check samples are available. An attempt will be made to obtain a QC sample to program sample ratio of 5 % or better. The Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission staff may evaluate the results.

Specific sample media for which EPA Cross Check Program samples are available include the following:

"* gross beta in air particulate filters

"* gamma emitters in air particulate filters

"* gamma emitters in milk

gamma emitters in water

tritium in water

1-131 in water Unit 2 Revision 22 1129 December 2001

4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements required by the Technical Specifications are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use. In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD.

Furthermore, the above capabilities may be determined by the vendor who supplies the TLDs. Required detection capabilities are as follows.

4.3.1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less than 7.5 %. A total of at least 5 TLDs shall be evaluated.

4.3.2 Reproducibility shall be determined by giving TLDs repeated exposures equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The average of the relative standard deviations of the responses shall be less than 3.0%. A total of at least 4 TLDs shall be evaluated.

4.3.3 Dependence of exposure interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constant. This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures. For these tests, the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85. At least 6 TLDs shall be evaluated.

4.3.4 Energy dependence shall be evaluated by the response of TLDs to photons for several energies between approximately 30 keV and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25 % for photons with energies greater than 80 keV and shall not be enhanced by more than a factor of two for photons with energies less than 80 keV. A total of at least 8 TLDs shall be evaluated.

4.3.5 The directional dependence of the TLD response shall be determined by comparing the response of the TLD exposed in the routine orientation with respect to the calibration source with the response obtained for different orientations. To accomplish this, the TLD shall be rotated through at least two perpendicular planes. The response averaged over all directions shall not differ from the response obtained in the standard calibration position by more than 10%. A total of at least 4 TLDs shall be evaluated.

4.3.6 Light dependence shall be determined by placing TLDs in the field for a period equal to the field cycle under the four conditions found in ANSI N545, section 4.3.6. The results obtained for the unwrapped TLDs shall not differ from those obtained for the TLDs wrapped in aluminum foil by more than 10%. A total of at least 4 TLDs shall be evaluated for each of the four conditions.

Unit 2 Revision 22 1130 December 2001

4.3.7 Moisture dependence shall be determined by placing TLDs (that is, the phosphors packaged for field use) for a period equal to the field cycle in an area where the exposure rate is known to be constant. The TLDs shall be exposed under two conditions: (1) packaged in a thin, sealed plastic bag, and (2) packaged in a thin, sealed plastic bag with sufficient water to yield observable moisture throughout the field cycle.

The TLD or phosphor, as appropriate, shall be dried before readout. The response of the TLD exposed in the plastic bag containing water shall not differ from that exposed in the regular plastic bag by more than 10%. A total of at least 4 TLDs shall be evaluated for each condition.

4.3.8 Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is less than 10 uR/hr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3).

The average exposure inferred from the responses of the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least 3 TLDs shall be evaluated.

Unit 2 Revision 22 1131 December 2001

TABLE D 2-1 LIQUID EFFLUENT DETECTORS RESPONSES*

NUCLIDE (CPM/ tCi/mi X 108)

Sr 89 0.78E-04 Sr 91 1.22 Sr 92 0.817 Y91 2.47 Y 92 0.205 Zr 95 0.835 Nb 95 0.85 Mo 99 0.232 Tc 99m 0.232 Te 132 1.12 Ba 140 0.499 Ce 144 0.103 Br 84 1.12 1131 1.01 1 132 2.63 1 133 0.967 1134 2.32 1135 1.17 Cs 134 1.97 Cs 136 2.89 Cs 137 0.732 Cs 138 1.45 Mn 54 0.842 Mn 56 1.2 Fe 59 0.863 Co 58 1.14 Co 60 1.65

Values from SWEC purchase specification NMP2-P281F.

Unit 2 Revision 22 II 32 December 2001

TABLE D 2-2 Aw+/- VALUES - LIQUID' ADULT mrem - ml hr - uCi T BODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG NUCLIDE 3.67E-1 3.67E-1 3.67E-1 3.67E-1 3.67E-1 H3 3.67E-1 --

3.13E2 1.18E-2 1.18E-2 2.86E-1 7.56E-1 1.66 Cr 51 1.26 -

2.33E2 -- 2.73 6.89 --

Cu 64 1.28 1.34E4 3.98 4.38E3 1.31E3 3.98 3.98 Mn 54 8.38E2 2.62E2 6.62E2 4.57E2 ..-- 2.55E2 Fe 55 1.07E2 9.28E2 8.06E3 1.03E3 2.42E3 7.53E-1 7.53E-1 6.76E2 Fe 59 -- -

5.43E1 5.36E2 -- 2.11El --

Co 57 1.81E3 1.07 9.04E1 1.07 1.07 1.07 Co 58 2.01E2 6.36E2 4.93E3 6.47E1 3.24E2 6.47E1 6.47E1 6.47E1 Co 60 2.21 3.32E4 4.63E4 2.31E4 7.35E4 4.92E4 2.21 Zn 65 6.38E2 3.57E3 2.22E4 6.18E-5 6.18E-5 6.18E-5 6.18E-5 Sr 89 Sr 90 1.36E5 1.60E4 5.55E5 ..

Sr 92 1.44E-2 6.61 3.34E-1 .....

7.59E-1 2.83E2 9.77E-1 7.88E-1 8.39E-1 6.99E-1 6.99E-1 Zr 95 Mn 56 3.07E-2 5.52 -- 1.73E-1 2.20E-1 ....

1.60E1 1.95E2 1.97E-3 8.42E1 1.91E2 1.97E-3 1.97E-3 Mo 99 1.34E2 1.34E2 1.34E2 1.34E2 1.34E2 1.34E2 1.34E2 Na 24 1.16E2 5.36E1 1.42E2 2.03E2 3.48E2 6.65E4 2.77E-2 1131 1132 4.34E-3 2.33E-3 4.64E-3 1.24E-2 1.98E-2 4.34E-1 -

1133 1.22E1 3.59E1 2.30E1 3.99E1 6.97E1 5.87E3 -

1135 1.32E0 3.79E0 1.28E0 3.36E0 5.39E0 2.22E2 -

Ni 65 1.14E-2 6.35E-1 1.93E-1 2.50E-2 .....

1.24E4 2.98E5 7.08E5 2.29E5 2.04E1 7.61E4 Cs 134 5.79E5 1.33E4 2.96E4 1.17E5 6.5 1E4 3.28E-1 8.92E3 Cs 136 8.42E4 3.42E5 1.01E4 3.82E5 5.22E5 1.77E5 3.1OEl 5.89E4 Cs 137 1.37E1 4.30E2 2.09E2 3.04E-1 1.3 1E-1 4.17E-2 1.92E-1 Ba 140 3.79E-2 8.81E1 6.93E-2 5.83E-2 4.60E-2 3.53E-2 3.53E-2 Ce 141 Nb 95m 1.51E1 1.44E6 3.53E1 2.74E1 2.70E1 ..

1.31E2 1.48E6 4.38E2 2.44E2 2.41E2 3.56E-1 3.56E-1 Nb 95 1.62E-2 3.72E3 1.03E-1 5.36E-2 2.83E-3 2.83E-3 2.83E-3 La 140 6.15E2 2.02 9.66E-1 6.57E-1 2.06E-1 2.06E-1 Ce 144 3.03E-1 2.05E-2 9.54E-01 5.71E-4 1.61E-3 2.45E-2 -- 7.90E-4 Tc 99m 1.8E-3 4.47E2 2.28E-2 2.78E-3 7.40E-3 5.95E-4 5.95E-4 Np 239 5.97E4 1.95E3 1.26E3 1.22E4 1.39E3 2.66E-3 Te 132 1.18E3 3.39E2 5.44E-3 1.10E-3 1.66E-3 7.11E-6 7.11E-6 Zr 97 5.08E-4 4.31E1 4.04E4 1.48E2 1.23E2 4.43E-5 4.43E-5 4.43E-5 W 187 1.09E1 3.94E2 1.14E1 1.13El 1.22E1 1.04E1 1.04E1 Ag 110m 3.36E2 1.07E3 4.33E1 4.31E1 4.31E1 5.12E1 Sb 124 4.72E1 3.60E4 2.46E2 5.90E2 3.57E2 6.90E-2 6.90E-2 Zn 69m 5.40E1 7.33E2 1.26E-1 4.67 1.79E1 1.26E-1 1.26E-1 Au 199 3.95 1.24E4 1.60E-1 6.19 1.16E1 1.60E-1 1.60E-1 As 76 5.94 Regulatory position C, Section 1.

Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Unit 2 Revision 22 II 33 December 2001

TABLE D 2-3 Aj VALUES - LIQUID' TEEN mrem - ml hr -uCi__ _ _ _ _ _ _ _

GI-TRACT BONE LIVER ml uCi TEEN

--2-3 mrem D hr TABLE KIDNEY THYROID LUNG NUCLIDE T BODY AiM VALUES

-LIQUIDl 2.73E-1 2.73E-1 2.73E-1 2.73E-1 2.73E-1 H3 2.73E-1 3.47E-1 7.79E-1 1.90 1.35 2.16E2 6.56E-2 6.56E-2 Cr51 2.87 7.27 Cu 64 1.35 2.23E2 2.22E1 2.22E1 2.22E1 4.32E3 1.31E3 Mn 54 8.75E2 8.84E3 3.11E2 2.13E2 6.93E2 4.91E2 Fe 55 1.15E2 4.20 4.20 7.84E2 9.59E2 5.85E3 1.06E3 2.48E3 Fe 59 2.19E1 Co 57 1.44E2 4.08E2 5.98 5.98 1.23E3 5.98 9.47E1 5.98 Co 58 2.10E2 3.61E2 3.6 1E2 3.61E2 9.44E2 3.73E3 3.61E2 6.20E2 Co 60 7.28E4 4.66E4 1.24E1 1.24E1 3.40E4 3.08E4 2.10E4 Zn 65 3.45E-4 3.45E-4 3.45E-4 3.45E-4 6.92E2 2.88E3 2.42E4 Sr 89 1.14E5 1.30E4 4.62E5 Sr 90 1.54E-2 9.19E1 3.61E-1 3.90 Sr 92 3.99 4.03 3.90 3.96 2.10E2 4.19 Zr 95 1.81E-1 2.29E-1 Mn 56 3.22E-2 1.19El 1.1 OE-2 1.10E-2 1.60E2 1.1OE-2 8.95E1 2.05E2 Mo 99 1.71E1 1.38E2 1.38E2 1.38E2 1.38E2 1.38E2 1.38E2 1.38E2 Na 24 2.12E2 3.66E2 6.19E4 1.55E-1 1.14E2 4.21E1 1.52E2 1131 1.27E-2 2.OOE-2 4.29E-1 4.56E-3 5.54E-3 4.86E-3 1132 4.19EI 7.35E1 5.85E3 1.02E-4 1.28E1 3.17E1 2.47E1 1133 3.46E0 5.47E0 2.23E2 1.76E0 3.84E0 1.34E0 1135 2.66E-2 1.21E-2 1.44 2.08E-1 Ni 65 7.18E5 2.28E5 1.14E2 8.72E4 3.33E5 9.05E3 3.05E5 1.01E4 Cs 134 1.17E5 6.38E4 1.83 7.87E4 9.44E3 2.98E4 7.21E4 Cs 136 5.44E5 1.85E5 1.73E2 1.90E5 7.91E3 4.09E5 4.15E-1 Cs 137 5.03E-1 3.25E-1 2.33E-1 1.44E1 3.40E2 2.21E2 Ba 140 2.21E-1 2.08E-1 1.97E-1 1.97E-1 2.00E-1 6.85E1 2.33E-1 Ce 141 2.99E1 2.96E1 1.69E1 1.14E6 3.87E1 1.99 Nb 95m 2.47E2 2.39E2 1.99 1.17E2 1.05E6 4.43E2 Nb 95 6.82E-2 1.58E-2 1.58E-2 1.58E-2 2.97E-2 3.01E3 1.22E-1 1.15 La 140 1.94 1.62 1.15 1.25 4.83E2 3.07 Ce 144 1.63E-3 2.43E-2 9.04E-4 2.11E-2 1.07 5.84E-4 3.32E-3 Tc 99m 5.67E-3 1.07E-2 3.32E-3 4.63E-3 3.78E2 2.82E-2 1.48E-2 Np 239 1.30E3 1.25E4 1.37E3 1.23E3 4.13E4 2.06E3 3.97E-5 Te 132 1.19E-3 1.78E-3 3.97E-5 5.68E-4 3.11E2 5.84E-3 2.47E-4 Zr 97 1.30E2 2.47E-4 2.47E-4 4.55E1 3.52E4 1.59E2 5.79E1 W 187 5.88E1 5.97E1 5.79E1 5.85E1 3.17E2 5.89E1 Ag 110m 2.41E2 2.41E2 2.41E2 2.50E2 2.45E2 4.5 3E2 2.5 1E2 3.85E-1 Sb 124 6.24E2 3.79E2 3.85E-1 5.76E1 3.43E4 2.65E2 7.04E-1 Zn 69m 5.60 2.01E1 7.04E-1 4.85 5.78E2 7.04E-1 Au 199 7.40 1.33E1 8.92E-1 8.92E-1 7.18 1.06E4 8.92E-1 As 76 position C, Section 1.

'Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory Unit 2 Revision 22 December 2001 II 34

TABLE D 2-4 Aj.+/- VALUES - LIQUID1 CHILD mrem - ml hr - uCi NUCLIDE T BODY GI-TRACT BONE LIVER ml uCi 4

Dhr-TABLE mrem CHILD KIDNEY THYROID LUNG Ai.,

VALUES

-LIQUID1 H3 3.34E-1 3.34E-1 3.34E-1 3.34E-1 3.34E-1 3.34E-1 Cr51 1.39 7.29E1 1.37E-2 1.37E-2 2.22E-1 7.76E-1 1.41 Cu 64 1.60 1.25E2 2.65 6.41 Mn 54 9.02E2 2.83E3 4.65 3.37E3 9.49E2 4.65 4.65 1.50E2 8.99E1 9.15E2 4.85E2 2.74E2 Fe 55 Fe 59 1.04E3 2.18E3 1.29E3 2.09E3 8.78E-1 8.78E-1 6.08E2 Co 57 6.24E1 1.62E2 2.00E1 Co 58 2.21E2 4.20E2 1.25 7.30E1 1.25 1.25 1.25 7.03E2 1.25E3 7.55E1 2.88E2 7.55E1 7.55E1 7.55E1 Co 60 Zn 65 3.56E4 1.01FA 2.15E4 5.73E4 3.6 1E4 2.58 2.58 Sr 89 9.13E2 1.24E3 3.20E4 Sr 90 1.06E5 5.62E3 4.17E5 Sr 92 1.85E-2 8.73 4.61E-1 Zr 95 8.95E-1 9.36E1 1.22 9.04E-1 9.43E-1 8.15E-1 8.15E-1 Mn 56 3.73E-2 2.39E1 1.65E-1 2.00E-1 Mo 99 2.22E1 7.42E1 2.30E-3 8.98E1 1.92E2 2.30E-3 2.30E-3 Na 24 1.51E2 1.51E2 1.51E2 1.51E2 1.51E2 1.5 1E2 1.51E2 1131 1.14E2 1.80El 2.00E2 2.01E2 3.31E2 6.66E4 3.23E-2 1132 5.08E-3 1.30E-2 6.01E-3 1. 10E-2 1.69E-2 5.13E-1 1133 1.51E1 1.60El 3.22E1 3.98E1 6.64E1 7.40E3 1135 1.53E0 2.30E0 1.68E0 3.02E0 4.63E0 2.67E2 Ni 65 1.46E-2 3.07 2.66E-1 2.5 1E-2 Cs 134 1.27E5 3.28E3 3.68E5 6.04E5 1.87E5 2.38E1 6.72E4 Cs 136 6.26E4 3.40E3 3.52E4 9.67E4 5.15E4 3.82E-1 7.68E3 Cs 137 7.28E4 3.12E3 5.15E5 4.93E5 1.61E5 3.62E1 5.78E4 Ba 140 1.87E1 1.62E2 3.19E2 3.28E-1 1.40E-1 4.87E-2 2.15E-1 Ce 141 4.61E-2 4.14E1 1.08E-1 7.43E-2 5.57E-2 4.12E-2 4.12E-2 Nb 95m 2.14E1 5.28E5 4.99E1 2.92E1 2.68E1 1.45E2 3.75E5 5.21E2 2.03E2 1.91E2 4.16E-1 4.16E-1 Nb 95 La 140 1.93E-2 1.33E3 1.39E-1 5.09E-2 3.30E-3 3.30E-3 3.30E-3 4.31E-1 2.92E2 3.81 1.36 8.61E-1 2.40E-1 2.40E-1 Ce 144 2.29E-2 7.87E-1 7.05E-4 1.38E-3 2.01E-2 7.02E-4 Tc 99m 2.40E-3 1.79E2 3.44E-2 3.12E-3 7.70E-3 6.94E-4 6.94E-4 Np 239 Te 132 1.38E3 1.15E4 2.57E3 1.14E3 1.06E4 1.66E3 3. 1OE-3 Zr 97 6.99E-4 1.77E2 8.11E-3 1.18E-3 1.69E-3 8.29E-6 8.29E-6 5.37E1 1.68E4 2.02E2 1.202 5.16E-5 5.16E-5 5.16E-5 W 187 1.29E1 1.24E2 1.35E1 1.30E1 1.39E1 1.21E1 1.21E1 Ag 110m 5.69E1 1.68E2 6.92E1 5.06E1 5.03E1 5.04E1 6.08E1 Sb 124 6.80E1 1.87E4 3.37E2 5.75E2 3.34E2 8.05E-2 8.05E-2 Zn 69m 5.58 2.75E2 1.47E-1 5.02 1.80E1 1.47E-1 1.47E-1 Au 199 8.31 5.47E3 1.86E-1 6.58 1.15E1 1.86E-1 1.86E-1 As 76

'Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

Unit 2 Revision 22 II 35 December 2001

TABLE D 2-5 1

A1w VALUES - LIQUID INFANT mrem - ml hr - uCi GI-TRACT BONE LIVER KIDNEY THYROID LUNG NUCLIDE T BODY 1.87E-1 -- 1.87E-1 1.87E-1 1.87E-1 1.87E-1 H3 1.87E-1

.... 1.17E-3 5.36E-3 1.04E-2 Cr 51 8.21E-3 2.39E-1 Cu 64 1.96E-2 8.70E-1 -- 4.24E-2 7.17E-2 ....

-- 1.20El 2.67 ..-

Mn 54 2.73 4.42 6.91E-1 8.42 5.44 .

-- 2.66 Fe 55 1.45 1.52E1 1.82E1 3.18E1 .... 9.41 Fe 59 1.25E1 2.37E0 -- 6.95E1 ..

Co 57 1.13E0 Co 58 5.36 5.36 -- 2.15 ......

Co 60 1.55EI 1.56E1 -- 6.55 .....

Zn 65 1.76E1 3.22E1 1.11El 3.81El 1.85E1 ....

Sr 89 4.27E1 3.06E1 1.49E3 .....

Sr 90 2.86E3 1.40E2 1.12E4 .....

Sr 92 1.56E-5 4.54E-3 4.21E-4 .....

Zr 95 2.12E-2 1.49E1 1.23E-1 2.99E-2 3.23E-2 ....

Mn 56 1.81E-6 9.56E-4 -- 1.05E-5 9.05E-6 ....

Mo 99 2.65 4.48 -- 1.36E1 2.03E1 ....

9.61E-1 9.61E-1 9.61E-1 9.61E-1 9.61E-1 9.61E-1 9.61E-1 Na 24 1131 9.78 7.94E-1 1.89E1 2.22E1 2.60E1 7.31E3 -

3.43E-6 7.80E-6 4.75E-6 9.63E-6 1.07E-5 4.52E-4 -

1132 8.26E-1 4.77E-1 1.94 2.82 3.31 5.13E2 -

1133 1135 2.38E2 2.36E2 3.29E2 6.54E2 7.28E2 5.86E0 -

Ni 65 2.96E-6 4.96E-4 5.75E-5 6.5 1E-6 --...

1.16 2.28E2 4.26E2 1.10E2 -- 4.50E1 Cs 134 4.30E1 1.14 2.56E1 7.53E1 3.00El -- 6.13 Cs 136 2.81E1 1.16 3.17E2 3.71E2 9.95E1 -- 4.03E1 Cs 137 2.63E1 2.33E1 9.48E1 9.48E-2 2.25E-2 -- 5.82E-2 Ba 140 4.88 Ce 141 3.31E-3 1.45E1 4.61E-2 2.81E-2 8.67E-3 ....

Nb 95m 1.02E3 1.20E1 2.39E3 1.73E3 1.10E3 ....

Nb 95 5.87E-3 8.57 2.47E-2 1.02E-2 7.28E-3 ....

La 140 6.52E-4 2.98E1 6.43E-3 2.53E-3 ......

Ce 144 1.01E-1 1.03E2 1.80 7.37E-1 2.98E-1 ....

3.17E-4 7.14E-3 1.19E-5 2.46E-5 2.64E-4 -- 1.28E-5 Tc 99m 2.08E-4 1.06E1 4.12E-3 3.68E-4 7.34E-4 --..

Np 239 Te 132 4.08 1.62E1 8.83 4.37 2.74E1 6.46 -

1.38E-4 1.92E1 1.76E-3 3.02E-4 3.04E-4 --..

Zr 97 W 187 4.13E-2 7.02 1.72E-1 1.19E-1 .....

6.02E-1 4.39E-1 6.28E-1 ..-

Ag 110m 2.91E-1 2.28E1 3.93E1 1.27E1 1.87E-1 -- 3.38E-2 7.98 Sb 124 3.95 1.24E-1 2.52E-1 1.02E-1 ..-

Zn 69m 2.30E-2 3.50 2.23E-1 5.38 -- 2.48E-1 6.26E-1 ....

Au 199 ..-

2.85E1 -- 8.46E-2 1.03E-1 As 76 8.67E-2 position C, Section 1.

'Calculated in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory Unit 2 Revision 22 II 36 December 2001

TABLE D 3-1 OFFGAS PRETREATMENT*

DETECTOR RESPONSE NUCLIDE NET CPM/,.Ci/cc Kr 83m Kr 85 4. 28E+03 Kr 85m 3. 85E+03 Kr 87 6.68E+03 Kr88 3.97E+03 Kr 89 6.48E+03 Xe 131m Xe 133 1.69E+03 Xe 133m Xe 135 4.91E+03 Xe 135m Xe 137 6.89E+03 Xe 138 5.51E+03

Values from calculation H21C-070 Unit 2 Revision 22 December 2001 II 37

TABLE D 3-2' 1

PLUME SHINE PARAMETERS NUCLIDE B. mrad/yr V_uCi/sec mrem/yr uCi/sec Kr 83m 9.01E-7 Kr 85 6.92E-7 5.09E-4 4.91 E-4 Kr 85m 2.57E-3 Kr 87 2.72E-3 7.23E-3 7.04E-3 Kr 88 1.13E-2 Kr 89 1.15E-2

6. 57E-3 4.49E-3 Kr 90 Xe 131m 7.76E-6 7.46E-5 6.42E-5 Xe 133 3.95E-5 4.79E-5 Xe 133m 7.44E-4 Xe 135 7.82E-4 1.45E-3 1.37E-3 Xe 135m 5.98E-4 Xe 137 6.25E-4 4 .46E-3 4.26E-3 Xe 138 1.31E-3 Xe-127 1.96E-3 5.OOE-3 4.79E-3 Ar 41 Bi and Vi are calculated for critical site boundary location; 1.6km in the easterly direction. See Appendix B. Those values that show a dotted line were negligible because of high energy absorption coefficients.

Unit 2 Revision 22 II 38 December 2001

TABLE D 3-3 IMMERSION DOSE FACTORS1 Ki(X-Bodv) Li (0-Skin) 2 Mi (y-Air) Ni (.-Air)

Nuclide 7.56E-02 1.93E1 2.88E2 Kr 83m

1. 17E3 1.46E3 1.23E3 1.97E3 Kr 85m 1.61E1 1.34E3 1.72E1 1.95E3 Kr 85 5.92E3 9.73E3 6.17E3 1.03E4 Kr 87 1.47E4 2.37E3 1.52E4 2.93E3 Kr 88 1.66E4 1.01E4 1.73E4 1.06E4 Kr 89 1.56E4 7.29E3 1.63E4 7.83E3 Kr 90 9.15E1 4.76E2 1.56E2 1.1 1E3 Xe 131m 2.51 E2 9.94E2 3.27E2 1.48E3 Xe 133m 2.94E2 3.06E2 3.53E2 1.05E3 Xe 133 3.12E3 7.11E2 3.36E3 7.39E2 Xe 135m 1.81E3 1.86E3 1.92E3 2.46E3 Xe 135 1.42E3 1.22E4 1.51E3 1.27E4 Xe 137 8.83E3 4.13E3 9.2 1E3 4.75E3 Xe 138 8.84E3 2.69E3 9.30E3 3.28E3 Ar 41

'From, Table B- 1.Regulatory Guide 1.109 Rev. 1 2

mrem/yr per uCi/m3 .

3mrad/yr per uCi/m 3 .

Unit 2 Revision 22 II 39 December 2001

TABLE D 3-4 DOSE AND DOSE RATE R, VALUES - INHALATION - INFANT' mrem/yr uCi/m3 nA* t t*r*O T~T'D rr D nlnX7 "I'H YKI)IU

~ 7T~ Y P,,JJ31* P.* I ir Ir TX'T1I-

Uz*t f(ZT..J I U JL--*L.4*LI*

NUCLIDE DUINL1r 11 A~J 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 5.3 1E3 H 3* 5.31 E3 5.3 1E3 5.3 1E3 2.65E4 5.31E3 5.3 1E3 3.57E2 C 14* 5.75E1 1.32E1 1.28E4 8.95E1 7.06E3 Cr 51 4.98E3 1.00E6 2.53E4 4.98E3 1.09E3 Mn 54 8.69E4 1.97E4 1. 17E4 3.33E3 2.48E4 Fe 55 1.02E6 1.36E4 2.35E4 9.48E3 1.11E4 Fe 59 7.77E5 1.22E3 1.82E3 3.19E4 Co 58 4.51 E6 8.02E3 1.18E4 5.14E4 Co 60 3.25E4 6.47E5 1.93E4 6.26E4 3.11E4 6.40E4 Zn 65 2.03E6 3.98E5 1. 14E4 1.31E5 Sr 89 1. 12E7 4.09E7 2.59E6 2.17E4 Sr 90 3.11 E4 1.75E6

1. 15E5 2.79E4 2.03E4 1.27E4 Zr 95 4.72E3 4.79E5 1.57E4 6.43E3 3.78E3 4.87E4 Nb 95 2.65E2 1.35E5 1.65E2 3.23E1 1.06E3 Mo 99 1.48E7 5.18E4 3.79E4 4.44E4 1.96E4 2.16E3 1-131 3.56E6 2.24E4 1.32E4 1.92E4 5.60E3 1.33E3 1133 1.90E5 7.97E4 3.96E5 7.03E5 7.45E4 1.33E3 Cs 134 1.72E5 7.13E4 5.49E5 6.12E5 4.55E4 3.84E4 Cs 137 1.34E1 1.60E6 5.60E4 5.60E1 2.90E3 8.48E4 Ba 140 1.68E5 5.05E2 2.00E2 5.15El 2.16E4 La 140 5.25E3 5.17E5 2.77E4 1.67E4 1.99E3 1.48E5 Ce 141 5.38E5 9.84E6 3.19E6 1.21E6 1.76E5 3.12E4 Ce 144 3.15E3 3.22E5 7.94E3 8.13E3 5.00E2 3.30E4 Nd 147 1.09E4 3.67E6 9.99E3 7.22E3 5.00E3 Ag 110m 3

mrem/yr per 4Ci/m Section 5.3. 1, except C 14 values in

'This and following Ri Tables Calculated in accordance with NUREG 0133, accordance with Regulatory Guide 1.109 Equation C-8.

Unit 2 Revision 22 II 40 December 2001

TABLE D 3-5 DOSE AND DOSE RATE Ri VALUES - INHALATION - CHILD mrem/yr uCi/m 3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H3* 1.12E3 1. 12E3 1. 12E3 1. 12E3 1.12E3 1. 12E3 C 14* 3.59E4 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 Cr 51 1.54E2 8.55E1 2.43E1 1.70E4 1.08E3 Mn 54 4.29E4 9.5 1E3 1.00E4 1.58E6 2.29E4 Fe 55 4.74E4 2.52E4 7.77E3 1.1 1E5 2.87E3 Fe 59 2.07E4 3.34E4 1.67E4 1.27E6 7.07E4 Co 58 1.77E3 3.16E3 1.1 1E6 3.44E4 Co 60 1.31E4 2.26E4 7.07E6 9.62E4 Zn 65 4.26E4 1.13E5 7.03E4 7.14E4 9.95E5 1.63E4 Sr 89 5.99E5 1.72E4 2.16E6 1.67E5 Sr 90 1.01E8 6.44E6 1.48E7 3.43E5 Zr 95 1.90E5 4.18E4 3.70E4 5.96E4 2.23E6 6.11E4 Nb 95 2.35E4 9.18E3 6.55E3 8.62E3 6.14E5 3.70E4 Mo 99 1.72E2 4.26E1 3.92E2 1.35E5 1.27E5 I 131 4.81 E4 4.81E4 2.73E4 1.62E7 7.88E4 2.84E3 1133 1.66E4 2.03E4 7.70E3 3.85E6 3.38E4 5.48E3 Cs 134 6.51 E5 1.01E6 2.25E5 3.30E5 1.21E5 3.85E3 Cs 137 9.07E5 8.25E5 1.28E5 2.82E5 1.04E5 3.62E3 Ba 140 7.40E4 6.48E1 4.33E3 2.11El 1.74E6 1.02E5 La 140 6.44E2 2.25E2 7.55E1 1.83E5 2.26E5 Ce 141 3.92E4 1.95E4 2.90E3 8.55E3 5.44E5 5.66E4 Ce 144 6.77E6 2.12E6 3.61 E5 1. 17E6 1.20E7 3.89E5 Nd 147 1.08E4 8.73E3 6.8 1E2 4.8 1E3 3.28E5 8.21E4 Ag 110m 1.69E4 1. 14E4 9.14E3 2.12E4 5.48E6 1.00E5

mrem/yr per ýCi/m3 Unit 2 Revision 22 1141 December 2001

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