Text

Radioactive Effluent Release Report, January - December 2012
	ML13130A006
Person / Time
Site:	Nine Mile Point
Issue date:	05/01/2013
From:	Dosa J; EDF Group, Constellation Energy Nuclear Group, Nine Mile Point
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML13130A006 (240)
	v • d • e

P.O. Box 63 CENGS.

a joint venture of Lycoming, NY 13093 Constelation eDF NINE MILE POINT NUCLEAR STATION May 1, 2013 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

Nine Mile Point Nuclear Station Unit No. 2; Docket No. 50-410 Radioactive Effluent Release Report, January - December 2012 In accordance with 10 CFR 50.36a and the Nine Mile Point Unit 2 (NMP2) Technical Specifications, enclosed is the Radioactive Effluent Release Report for the period January through December 2012.

Included in this report is a summary of gaseous and liquid effluents and solid waste released from the station during the reporting period (Attachments 1 - 6), a summary of revisions to the Offsite Dose Calculation Manual (ODCM) 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 any station liquid and/or gaseous effluent monitoring instrumentation that was non-functional for greater than 30 days (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. Attachment 12 provides a summary of the tritium results from the groundwater protection program. Attachment 13 is a copy of Revision 35 of the ODCM.

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 ODCM. During the reporting period from January through December 2012, NMP2 did not exceed any 10 CFR 20, 10 CFR 50, Technical Specification, or ODCM limits for gaseous or liquid effluents.

Should you have questions regarding the information in this submittal, please contact me at (315) 349-5219.

Very truly yours, John/.D/osa JJD/KES Director Licensing A 0

Document Control Desk May 1, 2013 Page 2

Enclosure:

Nine Mile Point Nuclear Station, Unit 2, Radioactive Effluent Release Report, January -

December 2012 cc: Regional Administrator, Region I, NRC Project Manager, NRC Resident Inspector, NRC B. Dionne, NRC

ENCLOSURE NINE MILE POINT NUCLEAR STATION, UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT January - December 2012 Nine Mile Point Nuclear Station, LLC May 1, 2013

1 -4 NINE MILE POINT NUCLEAR STATION - UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT January- December 2012 CENGSM a joint venture of 0 Conistellation e Energy 0,',,eDF NINE MILE POINT NUCLEAR STATION

Page 1 of 2 NINE MILE POINT NUCLEAR STATION - UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER 2012 SUPPLEMENTAL INFORMATION Facility: Nine Mile Point Unit 2 Licensee: Nine Mile Point Nuclear Station, LLC

1. TECHNICAL SPECIFICATION/ODCM LIMITS A) FISSION AND ACTIVATION GASES

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

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

B&C) TRITIUM, IODINES AND PARTICULATES, HALF LIVES > 8 DAYS

1. The dose rate limit of Iodine-131, Iodine-133, Tritium and all radionuclides in particulate form with half-lives greater than eight days, released in gaseous effluents from the site to areas at or beyond the site boundary shall be less than or equal to 1500 mrem/year to any organ.

2. The dose to a member of the public from Iodine-131, Iodine-133, Tritium and all radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released from Nine Mile Point Unit 2 to areas at or beyond the site boundary shall be limited during any calendar quarter to less than or equal to 7.5 mrem to any organ, and during any calendar year to less than or equal to 15 mrem to any organ.

D) LIQUID EFFLUENTS

1. Improved Technical Specifications (ITS) 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 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-04 microcuries/ml total activity.

2. The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released from Nine Mile Point Unit 2 to unrestricted areas shall be limited 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 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.

Page 2 of 2

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

A) FISSION AND ACTIVATION GASES Noble gas effluent activity is determined by an on-line scintillation detector (calibrated against gamma isotopic analysis of a 4.0L Marinelli grab sample) of an isokinetic sample stream.

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

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

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

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

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

G) C-14 The production of C-14 and the effluent dose consequences are estimates based on EPRI methodology provided in EPRI Report 1021106, Estimation of Carbon-14 in Nuclear Power PlantGaseous Effluents, December 2010 and NUREG-00 16, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents for Boiling Water Reactors (BWR-GALE Code).

ATTACHMENT 1

SUMMARY

DATA Page 1of 2 Unit I Unit 2 X Reporting Period January - December 2012 Liquid Effluents:

ODCM Required Maximum Effluent Concentration (MEC) = 10 x 10CFR20.1001 - 20.2402, Appendix B, Table 2, Column 2 Average MEC - pCi/ml (Qtr. 1)= NO RELEASES Average MEC - pCi/ml (Qtr. 3) = NO RELEASES Average MEC - pCi/mi (Qtr. 2) - I N RLESES Average MEC - pCi/mi (Qtr. 4) = NORLEA~SES Average Energy (Fission and Activation gases - MEV):

Qrtr. 1: Ey = 8.29E-01 E13 = 2.72E-01 Qrtr. 2: Ey = 6.89E-01 EP = 2.56E-01 Qrtr. 3: Ey = 3.44E-01 EP = 3.13E-01 Qrtr. 4: Ey = 9.90E-01 EP3 = 3.10E-01 Liquid:

Number of Batch Releases 0 Total Time Period for Batch Releases (hrs) 0.0 Maximum Time Period for a Batch Release (hrs) 0.00 Average Time Period for a Batch Release (hrs) 0.00 Minimum Time Period for a Batch Release 0.00 Total volume of water used to dilute the liquid I 1st 2nd 3rd 4th during the release period (L) N/A N/A N/A N/A lTotal volume of water available to dilute the liquid I 1st 2nd 3rd 4th effluent during the report period (L) 1"255+10 1"29E+10 1"30E+10 1'16E+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 N/A Gaseous (Primary Containment Purge)

Number of Batch Releases 20 Total Time Period for Batch Releases (hrs) 285.7 Maximum Time Period for a Batch Release (hrs) 60.6 Average Time Period for a Batch Release (hrs) 14.3 Minimum Time Period for a Batch Release (hrs) 0.3

ATTACHMENT 1

SUMMARY

DATA Page 2 of 2 Unit I Unit 2 X Reporting Period January - December 2012 Abnormal Releases:

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 Page 1 of 1 Unit I Unit 2 X Reporting Period January - December 2012 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES, ELEVATED AND GROUND LEVEL 1st 2nd 3rd 4th Est. Total Quarter Quarter Quarter Quarter Error %

A. Fission & Activation Gases

1. Total Release Ci 6.86E+01 8.60E+00 1.13E+01 1.57E+01 5.OOE+01

2. Average Release Rate pCi/sec 8.73E+00 1.09E+00 1.42E+00 1.98E+00 B. Iodines

1. Total Iodine - 131 Ci 6.14E-04 7.40E-04 2.76E-04 1.77E-04 3.OOE+01

2. Average Release Rate for Period pCi/sec 7.90E-05 9.19E-05 3.51E-05 2.11E-05 C. Particulates

1. Particulates with Half-lives>8days Ci 4.11E-04 2.08E-03 4.03E-04 2.26E-03 3.OOE+01

2. Average Release Rate for Period pCi/sec 5.26E-05 2.47E-04 5.13E-05 2.69E-04

3. Gross Alpha Radioactivity Ci 0.00E+00 0.OOE+00 0.OOE+00 0.OOE+00 2.50E+01 D. Tritium

1. Total Release Ci 2.34E+01 5.55E+00 2.26E+01 2.25E+01 5.OOE+01

2. Average Release Rate for Period pCi/sec 3.01 E+00 6.65E-01 2.88E+00 2.68E+00 E. Percent of Tech. Spec. Limits Fission and Activation Gases Percent of Quarterly Gamma Air Dose 1.31 E-01 1.35E-02 8.58E-03 3.62E-02 Limit (5 mR)

Percent of Quarterly Beta Air Dose Limit 2.17E-03 2.45E-04 3.93E-04 5.39E-04 (10 mrad)

Percent of Annual Gamma Air Dose 6.55E-02 7.22E-02 7.64E-02 9.45E-02 Limit to Date (10 mR)

Percent of Annual Beta Air Dose Limit to 1.09E-03 1.21 E-03 1.41 E-03 1.68E-03 Date (20 mrad)

Percent of Whole Body Dose Rate Limit 5.1OE-03 5.23E-04 3.26E-04 1.39E-03 (500 mrem/yr)

Percent of Skin Dose Rate Limit (3000 9.95E-04 1.03E-04 6.70E-05 2.72E-04 mrem/yr)

Tritium, lodines, and Particulates (with half-lives greater than 8 days)

Percent of Quarterly Dose Limit (7.5 1.68E-01 1.94E-01 7.59E-02 6.37E-02 mrem)

Percent of Annual Dose Limit to Date 8.44E-02 1.82E-01 2.20E-01 2.53E-01 (15 mrem)

Percent of Organ Dose Limit (1500 3.37E-03 3.90E-03 1.51E-03 1.28E-03 mrem/yr

ATTACHMENT 3 Page 1 of 1 Unit I Unit 2 X Reporting Period January - December 2012 GASEOUS EFFLUENTS - ELEVATED RELEASE Continuous Mode (2)

Nuclides Released 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Fission Gases (1)

Argon-41 Ci Krypton-85 Ci Krypton-85m Ci 2.50E+01 2.85E+00 6.31E+00 8.29E+00 Krypton-87 Ci ....

1.20E-01 **

Krypton-88 Ci 2.67E+01 2.30E+00 2.55E-01 7.31E+00 Xenon-127 Ci Xenon-131m Ci Xenon-133 Ci 1.68E+01 7.41 E-01 1.42E-01 1.33E-01 Xenon-133m Ci Xenon-135 Ci ** 1.16E+00 2.84E+00 **

Xenon-135m Ci ** 1.55E+00 4.39E-01 Xenon-137 Ci Xenon-138 Ci ....

1.18E+00 **

lodines (1)

Iodine-131 Ci 5.61 E-04 6.04E-04 2.76E-04 1.69E-04 Iodine-133 Ci 8.36E-03 3.04E-03 2.01 E-03 2.01 E-03 Iodine-135 Ci Particulates (1)

Chromium-51 Ci Manganese-54 Ci .... _ 1.97E-05 2.48E-04 Iron-55 Ci 8.37E-05 5.98E-05 6.97E-05 2.78E-04 Iron-59 Ci Cobalt-58 Ci ** 1.40E-05 1.17E-04 Cobalt-60 Ci 1.38E-05 7.95E-06 1.07E-04 8.27E-04 Neodymium-147 Ci Zinc-65 Ci ** 8.16E-05 4.76E-04 Strontium-89 Ci Strontium-90 Ci Niobium-95 Ci Zirconium-95 Ci Molybdenum-99 Ci ** ** 7.30E-06 8.59E-05 Ruthenium-103 Ci Cesium-134 Ci Cesium-136 Ci Cesium-1 37 Ci Barium-140 Ci Lanthanum-140 Ci Cerium-141 Ci Cerium-144 Ci Tritium (1) Ci I1.51E+01 i1.22E+00 i1.33E+01 i1.62E+01 (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 ot 1.OOE-04 pCi/ml for required noble gases, 1.OOE-11 pCi/ml for required particulates and gross alpha, 1.OOE-12 pCi/ml for required lodines, 1.OOE-11 pCi/ml for Sr-89/90 and 1.OOE-06 pCi/ml for Tritium, as required by the ODCM, has been verified.

(2) Contributions from purges are included. There were no other batch releases during the reporting period.

ATTACHMENT 4 Page 1 of 1 Unit I Unit 2 X Reporting Period January - December 2012 GASEOUS EFFLUENTS - GROUND LEVEL RELEASES Continuous Mode (2)

Nuclides Released 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Fission Gases (1)

Argon-41 Ci Krypton-85 Ci Krypton-85m Ci Krypton-87 Ci Krypton-88 Ci Xenon-127 Ci Xenon-1 31 m Ci Xenon-1 33 Ci Xenon-133m Ci 9.41E-02 .....

Xenon-135 Ci Xenon-135m Ci Xenon-137 Ci Xenon-138 Ci

- I ** ** **

lodines (1)

Iodine-131 Ci 5.29E-05 1.36E-04 ** 8.25E-06 Iodine-133 Ci 7.70E-04 1.75E-04 ** 9.01 E-05 Iodine-135 Ci Particulates (1)

Chromium-51 Ci ** 1.46E-04 ....

Manganese-54 Ci ** 9.OOE-05 5.07E-06 1.39E-05 Iron-55 Ci 2.76E-04 1.23E-03 6.20E-05 1.51E-04 Iron-59 Ci ** 1.42E-05 ....

Cobalt-58 Ci ** 9.15E-06 ....

Cobalt-60 Ci 3.82E-05 4.43E-04 4.44E-05 1.31E-04 Neodymium-147 Ci Zinc-65 Ci 6.22E-05 ** 1.85E-05 Strontium-89 Ci Strontium-90 Ci Niobium-95 Ci Zirconium-95 Ci Molybdenum-99 Ci Ruthenium-103 Ci ** 1.80E-05 ....

Cesium-1 34 Ci Cesium-1 36 Ci Cesium-1 37 Ci Barium-140 Ci Lanthanum-140 Ci Cerium-141 Ci Cerium-144 Ci Tritium (1) Ci 8.29E+00 4.32E+00 9.27E+00 6127E+00-1 (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 pCi/ml for required noble gases, 1.OOE-1i1 pCi/ml for required particulates and gross alpha, 1.OOE-12 pCi/ml for required iodines, 1.OOE

-11 pCi/ml for Sr-89/90 and 1.OE-06 pCi/mI for Tritium, as required by the ODCM, has been verified.

(2) There were no batch releases from this path during the reporting period.

ATTACHMENT 5 Page 1 of 2 Unit I Unit 2 X Reporting Period January - December 2012 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES (1) 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Est. Total Error, %

A. *i-ssion & Activation P-'roaucts

1. Total Release (not including Tritium, gases, alpha)

2. Average diluted concentration during Ci No Releases 1No Releases No Releases No Releases 5.OOE+01 pCi/ml No Releases No Releases No Releases No Releases reporting period B. Tritium 1.Total release Ci No Releases No Releases No Releases No Releases 5.OOE+01

2. Average diluted concentration during the pCi/ml No Releases No Releases No Releases No Releases reporting period C. Dissolved and Entrained Gases

1. Total release Ci No Releases No Releases No Releases No Releases 5.OOE+01

2. Average diluted concentration during the reporting period pCi/ml No Releases No Releases No Releases No Releases D. Gross Alpha Radioactivity

1. Total release 5.OOE+01 Ci No Releases I No Releases I No Releases I No Releases]

E. Volumes

1. Prior to Dilution Liters No Releases No Releases No Releases No Releases 5.OOE+01

2. Volume of dilution water used during release period Liters No Releases No Releases No Releases No Releases 5.OOE+01

3. Volume of dilution water available during Liters 1.25E+10 1.29E+10 1.30E+10 1.16E+10 5.OOE+01 reporting period F. Percent of Tech. Spec. Limits Percent of Quarterly Whole Body Dose 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Limit (1.5 mrem)

Percent of Annual Whole Body Dose Limit 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 to Date (3 mrem)

Percent of Quarterly Organ Dose Limit (5 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 mrem)

Percent of Annual Organ Dose Limit to 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 Date (10 mrem)

Percent of 10CFR20 Concentration Limit 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 (2), (3)

Percent of Dissolved or Entrained Noble 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 Gas Limit (2.OOE-04 pCi/ml)

(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 5.OOE-07 pCi/ml for required gamma emitting nuclides, 1.OOE-05 pCi/ml for required dissolved and entrained noble gases and tritium, 5.OOE-08 pCi/ml for Sr-89/90, 1.OOE-06 pCi/ml for 1-131 and Fe-55, and 1.OOE-07 pCi/ml for gross alpha radioactivity, as required by the Off-Site Dose Calculation Manual (ODCM), has been verified.

(2) The percent of 10CFR20 concentration limit is based on the average concentration during the quarter.

(3) 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 I Unit 2 X Reporting Period January - December 2012 LIQUID EFFLUENTS RELEASED Batch Mode (1),(2)

Nuclides Released 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Nuclides Released Strontium-89 Ci No Releases No Releases No Releases No Releases Strontium-90 Ci No Releases No Releases No Releases No Releases Cesium-134 Ci No Releases No Releases No Releases No Releases Cesium-137 Ci No Releases No Releases No Releases No Releases Iodine-131 Ci No Releases No Releases No Releases No Releases Cobalt-58 Ci No Releases No Releases No Releases No Releases Cobalt-60 Ci No Releases No Releases No Releases No Releases Iron-59 Ci No Releases No Releases No Releases No Releases Zinc-65 Ci No Releases No Releases No Releases No Releases Manganese-54 Ci No Releases No Releases No Releases No Releases Chromium-51 Ci No Releases No Releases No Releases No Releases Zirconium-95 Ci No Releases No Releases No Releases No Releases Niobium-95 Ci No Releases No Releases No Releases No Releases Molybdenum-99 Ci No Releases No Releases No Releases No Releases Technetium-99m Ci No Releases No Releases No Releases No Releases Barium-140 Ci No Releases No Releases No Releases No Releases Lanthanum-140 Ci No Releases No Releases No Releases No Releases Cerium-141 Ci No Releases No Releases No Releases No Releases Tungsten-187 Ci No Releases No Releases No Releases No Releases Arsenic-76 Ci No Releases No Releases No Releases No Releases Iodine-133 Ci No Releases No Releases No Releases No Releases Iron-55 Ci No Releases No Releases No Releases No Releases Neptunium-239 Ci No Releases No Releases No Releases No Releases Silver-1 1Om Ci No Releases No Releases No Releases No Releases Gold-1 99 Ci No Releases No Releases No Releases No Releases Cerium-144 Ci No Releases No Releases No Releases No Releases Cesium-136 Ci No Releases No Releases No Releases No Releases Copper-64 Ci No Releases No Releases No Releases No Releases Dissolved or Entrained Gases Ci No Releases No Releases No Releases No Releases Tritium Ci No Releases No Releases No Releases No Releases (1) No continuous mode release occurred during the report period as indicated by effluent sampling.

(2) 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.OOE-07 pCi/ml for required gamma emitting nuclides, 1.OOE-05 pCi/ml for required dissolved and entrained noble gases and tritium, 5.OOE-08 pCi/ml for Sr-89/90, 1.OOE-06 pCi/mI for 1-131 and Fe-55, and 1.OOE-07 pCi/ml for gross alpha radioactivity, as identified in the ODCM, has been verified.

ATTACHMENT 6 Page 1 of 4 Unit I Unit 2 X Reporting Period January - December 2012 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS Al. TYPE Volume Activity (1) 3 (M ) (Ci)

Class Class A B C A B C a.1 Spent Resins 5.63E+01 0.OOE+00 O.OOE+00 3.47E+02 0.OOE+00 0.OOE+00 (Dewatered) a.2 Filter Sludge 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 O.OOE+00 O.OOE+00 a.3 Concentrated Waste 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OQE+00 Totals 5.63E+01 0.OOE+00 0.OOE+00 3.47E+02 O.OOE+00 0.OOE+00 b.1 Dry, compressible waste 9.71E+02 0.OOE+00 0.OOE+00 1.38E+00 0.OOE+00 0.OOE+00 b.2 Dry, non-compressible O.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 0.OOE+O0 waste (contaminated equipment)

Totals 9.71E+02 0.OOE+00 O.OOE+00 1.38E+00 O.OOE+00 0.OOE+00

c. Irradiated Components, 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 Control Rods

d. Other (to vendor for processing) d.1 Carbon, High Rad Trash, 1.97E+02 0.OOE+00 0.OOE+00 9.06E+00 O.OOE+00 O.OOE+00 EPU Waste (1) The estimated total error is 5.O0E+01%.

IATTACHMENT 6 Page 2 of 4 Unit 1 Unit 2 X Reporting Period January - December 2012 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS Al. TYPE Container Package Solidification Agent a.1 Spent Resin (Dewatered) Poly Liner General Design None a.2 Filter Sludge N/A N/A N/A b.1 Dry Compressible Waste Seavan General Design None b.2 Dry, Non-Compressible N/A N/A N/A Waste

c. Irradiated Components, N/A N/A N/A Control Rods

d. Other (to vendor for processing) d .1 Ca rb oEPU n , H igWaste h R a d T ra s h , Poly P l Liner, Seavan, i e ,S FlexibleGeneral Design None a a ,F ei l

__________________________ j Packaging System

ATTACHMENT 6 Page 3 of 4 Unit 1 Unit 2 X Reporting Period January - December 2012 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A2. ESTIMATE OF MAJOR NUCLIDE COMPOSITION (BY TYPE OF WASTE)

a. Spent Resins, Filter Sludges, Concentrated Waste Nuclide Percent Fe-55 45.7 Co-60 36.5 Zn-65 10.9 Mn-54 4.4 C-14, Cr-51, Fe-59, Co-58, Ni-63, Sr-90, Nb-95, Tc-99, Sb-124, 2.5 Cs-134, Cs-137, Ce-144

b. Dry, compressible waste, dry, non-compressible waste (contaminated equipment)

Nuclide Percent Fe-55 69.3 Co-60 24.5 Mn-54 5.1 Fe-59, Co-58, Ni-63, Zn-65, Sr-90, Sb-125, Cs-134, Cs-137, Ce-144 1.1

c. Irradiated Components, Control Rods: There were no shipments.

Nuclide Percent N/A N/A

d. Other (To Vendor for Processing)

1. CarbonHigh Rad Trash, EPU Waste Nuclide Percent Fe-55 50.5 Co-60 43.5 Mn-54 4.2 H-3, C-14, Fe-59, Ni-63,Zn-65, Sr-89, Sr-90, Sb-125, Cs-134, 1.8 Cs-137, Ce-144

ATTACHMENT 6 Page 4 of 4 Unit I Unit 2 X Reporting Period January - December 2012 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A3. SOLID WASTE DISPOSITION Number of Shipments Mode of Transportation Destination 15 Hittman Transport Duratek Services, Inc 1 Hittman Transport Studsvik Processing Facility Memphis,LLC 1 Landstar Studsvik Processing Facility Memphis,LLC 27 R&R Trucking/AATCO Studsvik Processing Facility Memphis,LLC 13 Hittman Transport Studsvik Processing Facility,LLC B. IRRADIATED FUEL SHIPMENTS (Disposition): There were no shipments.

Number of Shipments Mode of Transportation Destination 0 N/A N/A D. SEWAGE WASTES SHIPPED TO A TREATMENT FACILITY FOR PROCESSING AND BURIAL There were no shipments of sewage sludge with detectable quantities of plant-related nuclides from NMP to the treatment facility during the reporting period.

ATTACHMENT 7 Page 1 of 1 Unit I Unit 2 X Reporting Period January - December 2012

SUMMARY

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

The Unit 2 Off-Site Dose Calculation Manual (ODCM) was revised once during the reporting period. Revision 35 included identification of the new Independent Spent Fuel Storage Installation (ISFSI) as a potential source of radiation. In addition, dose conversion factors for Ruthenium 103 (Ru-103), which was identified during the Unit 2 refuel outage. These changes do not affect the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50 Appendix I, and does not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. A copy of Revision 35 is attached (Attachment 13) and a summary of the changes is provided below.

REVISION 35 New/Amended Section Page # # Description of Change Reason for Change 13.1-5 D 3.1.2 I 3.2-5 D 3.2.2 I 3.2-8 D 3.2.3 A new Independent Spent Fuel Storage Installation has I B 3.1-2 B 3.1.2 been built on site and has become operational. It has I B 3.2-2 B 3.2.2 been identified in the ODCM as a potential source ol I B 3.2-3 B 3.2.3 radiation to be considered in dose calculations.

II 27 3.3 & 3.4 1129 3.4 II 41- II 59 Tables D 3-4 thru Added dose conversion Ru-103 was identified in a gaseous effluent sample. This D 3-22 factors for Ru-103 is the first time ever identified. Dose conversion factors for Ru-1 03 are added.

ATTACHMENT 8 Page 1of 1 Unit I Unit 2 X Reporting Period January - December 2012

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM (PCP) I There were no changes to the NMP2 Process Control Program (PCP) during the reporting period.

ATTACHMENT 9 Page 1of 1 Unit I Unit 2 X Reporting Period January - December 2012

SUMMARY

OF NON-FUNCTIONAL MONITORS Monitor Dates Monitor was Cause and Corrective Actions Non-Functional 2LWS-CAB206, January 1, 2012 to This monitor exceeded 30 days non-functional twice during 2012.

2LWS-FT330 & September 7, 2012 No liquid waste discharges were performed during these times, 2LWS-FT331, and and therefore, this monitor was not made operable. The Liquid Waste November 4, 2012 discharge manual isolation valves, 2LWS-V420 and 2LWS-V422, Discharge Monitor to December 31, are locked closed during inoperable periods, therefore, no 2012 inadvertent discharge can occur. Reference ESL 2010-0211 and ESL-2011-0243.

2OFG-CAB13A and June 17, 2012 to These monitors were originally removed from service to perform 2OFG-CAB13B July 18, 2012 condenser in-leakage tesing. During the inoperable time, the Offgas Monitors monthly funtional surveillance N2-RSP-RMS-M104 could not be kept current. After the condenser in-leakage testing and repair was completed, the surveillance was performed, and a low detector response was identified. Trouble shooting lead to a detector replacement, and the monitor was returned to service on July 18, 2012. Reference ESL-2012-0146.

2RMS-PNL180A, August 21, 2012 to The total effluent flow transmitter was degraded, and could not be Vent Wide Range October 3, 2012 calibrated. The lead time on the replacement transmitter caused Gaseous Monitoring the 30 day DLCO to be exceeded. The transmitter was replaced System (WRGMS) and successfully calibrated. The system was returned to service on October 3, 2012. Reference ESL-2012-0179.

1~

ATTACHMENT 10 Page 1of 5 Unit 1 Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY Introduction An assessment of the radiation dose potentially received by a Member of the Public due to their activities inside the site boundary from Nine Mile Point Unit 2 (NMP2) liquid and gaseous effluents has been conducted for the period January through December 2012.

This assessment considers the maximum exposed individual and the various exposure pathways resulting from liquid and gaseous effluents to identify the maximum dose received by a Member of the Public during their activities within the site boundary.

Prior to September 11, 2001, the public had access to the Energy Information Center for purposes of observing the educational displays or for picnicking and associated activities. Fishing also occurred near the shoreline adjacent to the Nine Mile Point (NMP) site. Fishing near the shoreline adjacent to the NMP site was the onsite activity that resulted in the potential maximum dose received by a Member of the Public. Following September 11, 2001 public access to the Energy Information Center has been restricted and fishing by Members of the Public at locations on site is also prohibited. Although fishing was not conducted during 2012 the annual dose to a hypothetical fisherman was still evaluated to provide continuity of data for the location.

Dose Pathways Dose pathways considered for this evaluation included 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 Radwaste/Reactor Building Vent were evaluated for the inhalation pathway. Dose due to aquatic pathways such as liquid effluents is not applicable since swimming is prohibited at the NMP site.

Dose to a hypothetical fisherman is received through the following pathways while standing on the shoreline fishing:

External ground pathway; this dose is received from plant related radionuclides detected in the shoreline sediment.

Inhalation pathway; this dose is received through inhalation of gaseous effluents released from the NMP2 Stack and Radwaste/Reactor Building Vent.

" Direct radiation pathway; dose resulting from the operation of Nine Mile Point Unit I (NMP1), NMP2 and the James A. Fitzpatrick Nuclear Power Plant (JAFNPP) Facilities.

Methodologies for Determininp_ Dose for Applicable Pathways External Ground (Shoreline Sediment) Pathway Dose from the external ground (shoreline sediment) is based on the methodology in the NMP2 Offsite Dose Calculation Manual (ODCM) as adapted from Regulatory Guide 1.109. For this evaluation it is assumed that the hypothetical maximum exposed individual fished from the shoreline at all times.

ATTACHMENT 10 Page 2 of 5 Unit 1 Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY The total dose received by the whole body and skin of the maximum exposed individual during 2012 was calculated using the following input parameters:

Usage Factor = 312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months (fishing 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week, 39 weeks per year)

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 = I (used average radionuclide concentration over total time period)

Average Cs-137 concentration = 1.59E-01 pCi/g The total whole body and skin doses received by a hypothetical maximum exposed fisherman from the external ground pathway is presented in Table 1, Exposure Pathway Annual Dose.

Inhalation Pathway The inhalation dose pathway is evaluated by utilizing the inhalation equation in the NMP2 ODCM, as adapted from Regulatory Guide 1.109. The total whole body dose and organ dose received by the hypothetical maximum exposed fisherman during 2012 calculated using the following input parameters for gaseous effluents released from both the NMP2 Stack and Radwaste/Reactor Building Vent for the time period exposure is received:

NMP2 Stack:

Variable Fisherman 1 X/Q (s/m3) 9.6 E-07 Inhalation dose factor Table E-7, Regulatory Guide 1.109 Annual air intake (m 3/year) (adult) 8000 Fractional portion of the year 0.0356 H-3 (pCi/sec) 1.27 E+06 C-14 (pCi/sec) 2 6.40 E+05 Mn-54 (pCi/sec) 1.15 E+0I Fe-55 (pCi/sec) 1.68 E+01 Fe-59 (pCI/sec) 5.79 E-01 Co-58 (pCI/sec) 4.82 E+00 Co-60 (pCI/sec) 3.89 E+01 Mo-99 (pCl/sec) 4.21 E+00 1-131 (pCi/sec) 4.33 E+01 1-133 (pCi/sec) 2.91 E+02

ATTACHMENT 10 Page 3 of 5 Unit 1 _ Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY NMP2 Radwaste/ReactorBuilding Vent:

Variable Fisherman 3 2.8 E-06 X/Q (s/m )

Inhalation dose factor Table E-7, Regulatory Guide 1.109 Annual air intake (m3/year) (adult) 8000 Fractional portion of the year 0.0356 H-3 (pCi/sec) 8.16 E+05 Cr-51 (pCi/sec) 6.68 E+00 Mn-54 (pCi/sec) 4.99 E+00 Fe-55 (pCi/sec) 6.39 E+01 Fe-59 (pCi/sec) 6.50 E-01 Co-58 (pCi/sec) 4.19 E-01 Co-60 (pCi/sec) 2.59 E+01 Zn-65 (pCi/sec) 3.69 E+00 Ru-103 (pCi/sec) 8.24 E-01 1-131 (pCi/sec) 6.41 E+00 1-133 (pCi/sec) 8.02 E+00 1 The maximum exposed fisherman is assumed to be present on site during the period of April through December at a rate of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week for 39 weeks per year equivalent to 312 hours0.00361 days 0.0867 hours 5.15873e-4 weeks 1.18716e-4 months for the year (fractional portion of the year = 0.0356). Therefore, the Average Stack and Radwaste/Reactor Building Vent flow rates and radionuclide concentrations used to determine the dose are represented by second, third and fourth quarter gaseous effluent flow and concentration values.

2. C-14 release rate determined from NUREG-0016, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents for Boiling Water Reactors (BWR-GALE Code),"

and EPRI Technical Report 1021106, "Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents."

The total whole body dose and maximum organ dose received by the hypothetical maximum exposed fisherman is presented in Table 1, Exposure Pathway Annual Dose.

ATTACHMENT 10 Page 4 of 5 Unit 1 _ Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY Direct Radiation PathwayT The direct radiation pathway is evaluated in accordance with the methodology found in the NMP2 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 from plume submersion.

The direct radiation pathway is evaluated by the use of high sensitivity environmental Thermoluminescent Dosimeters (TLDs). Since fishing activities occur between April 1 and December 31, TLD data for the second, third, and fourth quarters of 2012 from TLDs placed in the general area where fishing once occurred were used to determine an average dose to the hypothetical maximum exposed fisherman from direct radiation. The following is a summary of the average dose rate and assumed time spent on site used to determine the total dose received:

Variable Fisherman Average Dose Rate (mRem/hr) 1.69 E-03 Exposure time (hours) 312 Total Doses received by the hypothetical maximum exposed fisherman from direct radiation is presented in Table 1, Exposure Pathway Annual Dose.

Dose Received Bv A Hvoothetical Maximum Exoosed Member of the Public Inside the Site Boundary During 2012 The following is a summary of the dose received by a hypothetical maximum exposed fisherman from liquid and gaseous effluents released from NMP2 during 2012:

TABLE 1 Exposure Pathway Annual Dose Fisherman Exposure Pathway Dose Type (mrem)

External Ground Whole Body 2.51 E-03 Skin of Whole Body 2.92 E-03 Inhalation Whole Body 2.33 E-04 Maximum Organ Bone: 3.98 E-04 Thyroid 2.81 E-04 Direct Radiation Whole Body 0.53

ATTACHMENT 10 Page 5 of 5 Unit 1 Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY Based on these values, the total annual dose received by a hypothetical maximum exposed Member of the Public inside the site boundary is as follows:

TABLE 2 Annual Dose Summary Fisherman Total Annual Dose for 2012 (mrem)

Total Whole Body 5.29 E-0I Skin of Whole Body 2.92 E-03 Maximum Organ Bone: 3.98 E-04 Thyroid 2.81E-04

ATTACHMENT 11 Page 1 of 5 Unit 1 __ Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Introduction An assessment of radiation doses potentially received by the likely most exposed Member of the Public located beyond the site boundary was conducted for the period January through December 2012 for comparison against the 40 CFR 190 annual dose limits.

The intent of 40 CFR 190 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 I (NMP1), NMP2 and the James A. FitzPatrick Nuclear Power Plant (JAFNPP) facilities must be considered.

40 CFR 190 requires the annual radiation dose received by Members of the Public in the general environment, as a result of plant operations, be limited to:

< 25 mRem whole body

< 25 mRem any organ (except thyroid)

< 75 mRem thyroid This evaluation compares doses resulting from liquid and gaseous effluents and direct radiation originating from the site as a result of the operation of the NMP1, NMP2 and JAFNPP nuclear facilities.

Dose Pathways Dose pathways considered for this evaluation included doses resulting from liquid effluents, gaseous effluents and direct radiation from all nuclear operating facilities located on the Nine Mile Point Site.

Dose to the likely most exposed Member of the Public, outside the site boundary, is received through the following pathways:

" Fish consumption pathway; this dose is received from plant radionuclides that have concentrated in fish that is consumed by a Member of the Public.

Vegetation consumption pathway; this dose is received from plant radionuclides that have concentrated in vegetation that is consumed by a Member of the Public.

Shoreline Sediment; this dose is received as a result of an individual's exposure to plant radionuclides deposited in the shoreline sediment, which is used as a recreational area.

" Deposition, Inhalation and Ingestion pathways resulting from gaseous effluents; this dose is received through exposure to gaseous effluents released from NMP1, NMP2 and JAFNPP operating facilities.

Direct Radiation pathway; radiation dose resulting from the operation ofNMPI, NMP2 and JAFNPP facilities.

Methodologies for Determining Dose for Applicable Pathways Fish Consumption Dose received as a result of fish consumption is based on the methodology specified in the NMP2 Off-Site Dose Calculation Manual (ODCM) as adapted from Regulatory Guide 1.109. The dose for 2012 is calculated from actual analysis results of environmental fish samples taken near the site discharge points. For this evaluation it is assumed that the most likely exposed Member of the Public consumes fish taken near the site discharge points.

No radionuclides were detected in fish samples collected and analyzed during 2012; therefore, no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2012.

ATTACHMENT 11 Page 2 of 5 Unit 1 _ Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Veeetation Consumption Dose received as a result of vegetation consumption is based on the methodology specified in the NMP2 ODCM as adapted from Regulatory Guide 1.109. The dose for 2012 is calculated from actual analysis results of environnmental vegetation samples taken near the likely most exposed Member of the Public.

No radionuclides were detected in vegetation samples collected and analyzed during 2012; therefore, no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2012.

For estimating C-14, dose received as a result of vegetation consumption is based on the methodology specified in the NMP2 ODCM as adapted from Regulatory Guide 1.109. The estimated concentration of C-14 in vegetation is based on the estimated concentration of C-14 in the plant gaseous effluents.

Shoreline Sediment Dose received from shoreline sediment is based on the methodology in the NMP2 ODCM as adapted from Regulatory Guide 1.109. For this evaluation it is assumed that the most likely exposed Member of the Public spends 67 hour7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months s/year along the shoreline for recreational purposes.

No radionuclides were detected in shoreline sediment samples collected and analyzed during 2012; therefore, no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2012.

Dose Pathways Resulting From Gaseous Effluents Dose received by the likely most exposed Member of the Public due to gaseous effluents is calculated in accordance with the methodology provided in the NMP1 ODCM, NMP2 ODCM, and the JAFNPP ODCM. These calculations consider deposition, inhalation and ingestion pathways. The total sum of doses resulting from gaseous effluents from NMP1, NMP2 and JAFNPP during 2012 provides a total dose to the whole body and maximum organ dose for this pathway.

Carbon-14 Dose Pathways Resulting from Gaseous Effluents The carbon-14 (C-14) effluent source terms are used to estimate radiological doses from C-14 in site gaseous waste effluents. These estimates were generated in order to meet the NRC requirement to incorporate C-14 in nuclear power plant 2012 Annual Radiological Effluent Release Reports (ARERRs). The C-14 production and effluent source term estimates were based on EPRI methodology provided in EPRI Report 1021106, Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents, December 2010. The following methodology was used to estimate C-14 gaseous release activity and dose components for the 2012 ARERR.

ATTACHMENT 11 Page 3 of 5 Unit 1 Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY EPRI Methodology for Estimating C-14 Production Rates in Boiling Water Reactors (BWRs):

For BWRs, EPRI Report 1021106 summarized the distribution of C-14 in release pathways as follows: gaseous 95% to 99%, liquid <0.5% and solid 1% to 5%. The report also states that -95% of C-14 in BWR gaseous waste effluents exists in the carbon dioxide form, which contributes to population dose via photosynthesis uptake in the food consumption cycle.

For NMP 1 and NMP2, C-14 gaseous dose calculations in the site ARERR are made using the following assumptions for each unit: (1) continuous release of the estimated C-14 generated during power operation based on the number of Effective Full Power Days (EFPDs) for the period, (2) maximum C-14 activity from literature values cited in EPRI Report 1021106, and (3) typical fraction as carbon dioxide for gaseous releases from literature values also cited in EPRI Report 1021106.

Equation 1 estimates the maximum annual production of C-I14, PRmtx, for each BWR unit.

PRux = 5.1 # MWT / 1000 [Eq 1]

Where:

5.1 = BWR Normalized Production (Ci/GWt-yr)

MWT MegaWatts Thermal (MWt) 1000 Conversion Factor (MWt to GWt)

Equation 2 estimates the C-14 activity released, A C-!4, into the gaseous pathway during the time period for each BWR unit.

A C-14 PRALJ4v # 0.99

EFPD / 365, Ci (for time period) [Eq 2]

Where:

PR AV = maximum annual production rate of C-14 0.99 = fraction C-14 in BWR gaseous pathway releases (maximum literature value in EPRI Report 1021106; also Table 1)

EFPD number of effective full power days for the unit during the time period; e.g., quarterly or yearly (Table 1) 365 number of days in a typical year

.4 ATTACHMENT 11 Page 4 of 5 Unit 1 Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Equation 3 estimates the C- 14 activity released in carbon dioxide form, A c-14, C02, into the gaseous pathway during the time period for each BWR unit.

A C-14, C02 PRm4

0.99

0.95 # EFPD / 365, Ci (for time period) [Eq 3]

Where:

PRAL -- maximum annual production rate of C-14 0.99 = fraction of C-14 in BWR gaseous pathway releases (maximum literature value in EPRI Report 1021106; also Table 1) 0.95 = fraction of C-14 as carbon dioxide in BWR gaseous pathway releases (typical literature value in EPRI Report 1021106, also Table 1)

EFPD number of effective full power days for the unit during the time period, e.g. quarterly or yearly (Table 1) 365 number of days in a typical average year For each BWR unit, the 2012 estimated C-14 activity releases (total and carbon dioxide chemical form) are summarized in Table 1.

Table 1 2012 BWR Estimated C-14 Gaseous Releases BWR Gaseous Release CO 2 Form Release EFPD Operation Max.

Prod.Annual Rate 2012 Total (Eq 2)Release 2012 CO Release (Eq2 3)

(Eq 1) (Eq 2) (Eq 3)

Fraction(a) Fraction(b) 328 EFPD NMP! 0.99 0.95 (89.9%) 9.44 Ci/yr 8.39 Ci 7.97 Ci 295 EFPD NMP2 0.99 0.95 (80.8%)

(80.8%) 20.34 Ci/yr(C) 16.27 Ci 15.5 Ci JAFNPP 0.99 0.95 303 EFPD 10.84 Ci/yr 8.89 Ci 8.45 Ci (83.0%)

(a) Maximum literature values from EPRI Report 1021106.

(b) Typical value from EPRI Report 1021106.

(c) NMP2 Reactor Power Rating increased to 3988 Megawatts thermal.

As long as the core designs and power ratings are not significantly changed, the maximum annual production rates and annual total and carbon dioxide activity releases in Table I should be acceptable for use in estimating C-14 gaseous release activity and dose components for the ARERR.

ATTACHMENT 11 Page 5 of 5 Unit 1 Unit 2 X Reporting Period: January - December 2012 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Direct Radiation Pathway 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 2012, the closest residence and the critical downwind residence are at the same location.

Table 2 Dose Potentially Received by the Likely Most Exposed Member of the Public Outside the Site Boundary During 2012 Exposure Pathway Dose Type Dose (mrem)

Fish Consumption Total Whole Body No Dose Total Maximum Organ No Dose Shoreline Sediment Total Whole Body No Dose Total Skin of Whole Body No Dose Gaseous Effluents Total Whole Body 6.45 E-04 (excluding C- 14) Thyroid 9.48 E-04 Maximum Organ Lung: 7.19 E -04 Gaseous Effluent Total Whole Body 4.08 E-02 (C- 14) Maximum Organ Bone: 2.04 E-01 Direct Radiation Total Whole Body 3.9 Based on these values the maximum total annual dose potentially received by the likely most exposed Member of the Public during 2012 is as follows:

Total Whole Body: 4.0 E+00 mrem

Total Thyroid: 9.48 E-04 mrem

Maximum Organ: Bone: 2.04 E-01 mrem 40 CFR 190 Evaluation The maximum total doses presented in this attachment are the result of operations at the NMPI, NMP2 and JAFNPP facilities. The maximum organ dose (Bone: 0.204 mrem), maximum thyroid dose (0.0009 mrem) and the maximum whole body dose (4.0 mrem) 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.

Page 1 of 3 ATTACHMENT 12 GROUNDWATER MONITORING DATA Reporting Period January - December 2012 CONCENTRATION OF TRITIUM IN GROUNDWATER SAMPLES (pCi/I)

MW-9(2 ) MW- 1O(

2 Date GMX-MWl MW-B119' MW-1 MW-5 MW-6 MW-7 MW-8 ) MW-ll 2/7/2012 <386 <386 <386 2/8/2012 <386 <386 <386 <386 2/9/2012 <386 5/1/2012 <406 <392 <406 <392 5/2/2012 <392 <406 <406 <406 7/30/2012 <393 <373 <372 <372 <372 <372 <372 7/31/2012 <372 <372 <372 8/19/2012 <480 <487 8/20/2012 <463 <486 <482 <482 8/21/2012 <481 <479 <480 <483 8/27/2012 <499 <499 8/28/2012 <499 <499 <442 <442 <442 <499 <499 <499 9/4/2012 <442 <442 <442 9/10/2012 <430 <430 <367 9/17/2012 <402 <402 <402 9/24/2012 <375 <375 <375 10/2/2012 <415 10/4/2012 <415 <415 10/15/2012 <387 <387 10/16/2012 <363 <387 <387 <387 <363 10/17/2012 <387 10/18/2012 <387 <387 10/24/2012 <363 10/25/2012 <363 <363 11/1/2012 <316 11/12/2012 <345 <345 <345 11/13/2012 12/5/2012 <314 <314 12/6/2012 <314 Notes:

Control Location (1) New well locations (2) Sentinel well location

Page 2 of 3 ATTACHMENT 12 GROUNDWATER MONITORING DATA Reporting Period January - December 2012 CONCENTRATION OF TRITIUM IN GROUNDWATER SAMPLES (pCi/I) 1 1 1 1 1 2 Date MW-12 MW-13 MW-14*( ) MW-15(1) MW-16( ) MW-17 1) MW-18 ) MW-19(1) MW-20( ) MW-21 (' )

2/7/2012 <386 <386 5/1/2012 <406 <406 5/2/2012 <406 7/30/2012 <373 <373 1 7/31/2012 <373 8/19/2012 <480 8/20/2012 <478 <481 8/27/2012 <499 8/28/2012 <499 <499 9/10/2012 <367 9/17/2012 <402 9/24/2012 <375 <375 <375 10/2/2012 <415 <415 <415 10/4/2012 <415 <415 <415 <415 10/15/2012 <363 <387 10/16/2012 <363 <387 10/17/2012 <387 <387 <387 <387 10/18/2012 <363 <363 10/24/2012 <363 <363 <363 <363 <363 <363 10/26/2012 <363 11/12/2012 <345 11/13/2012 <345 <345 <345 12/5/2012 1 _ 1 <314 <314 <416 1 NMP2 Groundwater Depression Cone (pCi/I) 3 _)

1/3/2012 <415 8/27/2012 <419 10/11/2012 <345 4/2/2012 <400 8/30/2012 399 +/- 115 10/15/2012 <345 6/4/2012 402 +/- 117 9/4/2012 <499 10/29/2012 <386 6/11/2012 <370 9/7/2012 <430 11/5/2012 467 +/-116 7/2/2012 <386 9/10/2012 <430 11/12/2012 <416 7/16/2012 <298 9/13/2012 <440 11/19/2012 <416 7/23/2012 <298 9/17/2012 <440 11/26/2012 908 +/-208 7/30/2012 <393 9/21/2012 <440 12/3/2012 <416 8/7/2012 <298 9/27/2012 <379 12/31/2012 <376 8/13/2012 584 +/- 114 10/1/2012 <379 8/20/2012 <389 10/4/2012 <328 8/23/2012 <389 1 10/8/2012 <328 Notes:

Control Location (1) New well locations (2) Sentinel well location (3) Samples collected from storm drain system, which include precipitation

U ATTACHMENT 12 Page 3 of 3 FIGURE 1 NMPNS ON-SITE GROUNDWATER MONITORING WELLS AND UNIT 2 STORM DRAIN OUTFALL I

- Monitoring Well Location

/ - Storm Drain Outfall 001 NW 303.75 SSW 168.75 I I I MW-B119 IGround Water Monitoring well is located SCALE (YARDS) south of Lake Road at a distance and direction of 0 100 200 300 1,355 yrds @ 195 degrees, FGMX-MWlI (SSW). Based on NMP2 Rx Building.

4 1,

ATTACHMENT 13 Page 1 of 207 Unit 1 Unit 2 X Reporting Period January - December 2012 OFF-SITE DOSE CAL1CIUILATION MANIuAL (ODCMI Revision 35

O Constellation Energy NINE MILE POINT NUCLEAR STATION NINE MILE POINT UNIT 2 OFF-SITE DOSE CALCULATION MANUAL (ODCNF REVISION 35 APPROVALS SIGNATURES DATE Prepared by: ~4 It.-

G. I. Stinson Principle Chemist Reviewed by: 6lr~oIia C. L. Wid4-Poindexter Supervisor - Chemistry Support M. R. Faivus GeneAl SuperviW Chemistry M. A. Philippon 61pZ///

Plant General Manager

SUMMARY

OFREVISIONS Revision 35 (Effective June 2012)

PAGE DATE 13.3-13,14 August 2000 1 3.3-6 November 2000 1 4.0-1 November 2000 II 2-10,26,33-36,66,67,75,80 November 2000 ix, 1 1.0-1, 11.0-2, 1 B 3.3-2, 14.1-1 & Ia, 11 11, 11 15,I1 29, 1163, 11 107,11 108 December 2001 1 3.3-9 December 2002 1 3.3-10 March 2003 13.3-7, 13.3-12, and 13.3-13 January 2004 II 63,1164, and 11107 December 2005 II 3 and I 4 May 2006 iv, 11.0-1, 13.1-7, 1 3.2-3, 13.2-10, 13.2-12,13.3-1, 13.3-2,13.3-3, 1 3.3-7,1 3.3-8, 13.3-9, 1 3.3-10, 1 B 3.1-3, I B 3.2-5, 1B 3.2-6, 1 B 3.3-1, 1 B 3.3-2,14.1-la, II 10, 11 13, 1120, and 1123 September 2006 I1 12, 1115, 11 16 September 2007 11 16 September 2007 11.0-3, 11.0-4, 1 3.2-2, 1 3.2-3, 14.2-1, 1 4.2-2, 11 9, 1113, 1 14, II 16, II 20,11 58, 1162-65, II 90, If 104,11105, 11 107, 11108, and 11109 December 2008 1 3.5-6, If 63, 11 64, I165, 11 108 October 2009 13.3-13 February 2011 13.3-9, 13.3-10, II 63 July 2011 1 3.1-5, 1 3.2-5, 1 3.2-8, I B 3.1-2, 1 B 3.2-2, 1 B 3.2-3, 1127, 1129, II 41 thru II 59 June 2012 i Unit 2 Revision 35 June 2012

TABLE OF CONTENTS PAGE List of Tables vii List of Figures ix Introduction x PART I - RADIOLOGICAL EFFLUENT CONTROLS I SECTION 1.0 DEFINITIONS I !.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 I 3.2-1 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- 131, 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 1 3.2-12 D 3.2.6 Venting or Purging 13.2-14 D 3.3 Instrumentation 1 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 I 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 R 3 1.3 Liquid Radwaste Treatment System IB 3.1-3 ii Unit 2 Revision 35 June 2012

TABLE OF CONTENTS (Con!)

PAGE B 3.2 Radioactive Gaseous Effluents I B 3.2-1 B 3.2.1 Gaseous Effluents Dose Rate I B 3.2-1 B 3.2.2 Gaseous Effluents Noble Gas Dose I B 3-2-2 B 3.2.3 Gaseous Effluents Dose - Iodine-13 1, Iodine- 133, Tritium, and Radioactive Material in Particulate Form IB 3.2-3 B 3.2.4 Gaseous Radwaste Treatment System 1B 3.2-5 B 3.2.5 Ventilation Exhaust Treatment System IB 3.2-6 B 3.2.6 Venting or Purging 1B 3.2-7 B 3.3 Instrumentation I B 3.3-1 B 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation 1 B 3.3-1 B 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation 1 B 3.3-2 B 3.4 Radioactive Effluents Total Dose 1B 3.4-1 B 3.5 Radiological Enviroinental Monitoring IB 3.5-1 B 3.5.1 Monitoring Program IB 3.5-1 B 3.5.2 Land Use Census IB 3.5-2 B 3.5.3 Interlaboratory Comparison Program IB 3.5-3 SECTION 4.0 ADMINISTRATIVE CONTROLS 14.0-1 D4.1 Reporting Requirements 14.1-1 D 4.1.1 Special Reports 14.1-1 D 4.2 Major Changes to Liquid, Gaseous and Solid Radwaste Treatment Systems 1 4.2-1 iii Unit 2 Revision 35 June 2012

TABLE OF CONTENTS (Cont)

.;FCT(ON NIJR.!ECT RFFSRCTION REFSECTION PAGP PAGE SECTION SUBJECT PART II- CALCULATIONAL METHODOLOGIES II 1 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 11.2.1 Liquid Radwaste Effluent Radiation Alarm Selpoint 112 1.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculations 115 1.1.2.3 Service Water and Cooling Tower Blowdown Effluent Radiation Alarm Setpoint 116 1.2 Liquid Effluent Concentration Calculation 3.1.1 117 DSR 3.1.1.2 1.3 Liquid Effluent Dose Calculation Methodology 3.1.2 118 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 FUNCTIONALITY 3.1.3 II .10 DSR 3.1.3..1 B 3.1.3 2.0 GASEOUS EFFLUENTS 1112 2.1 Gaseous Effluent Monitor Alarm Setpoints 1112 2.1.1 Basis 3.2.1 1112 2.1.2 Setpoint Determination Methodology Discussion 3.3.2 1112 2.1.2,1 Stack Noble Gas Detector Alarm Setpoint Equation I 13 2.1.2.2 Vent Noble Gas Detector Alarm Setpoint Equation 1114 2.1.2,3 Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation 1115 2.2 Gaseous Effluent Dose Rate Calculation Methodology 3.2.1 1116 2.2.1 X/Q and W, - Dispersion Parameters for Dose Rate, Table D 3-23 1116 2.2.2 Whole Body Dose Rate Due to Noble Gases DLCO 3.2,1 .a 1117 DSR 3.2.1.1 2.2.3 Skin Dose Rate Due to Noble Gases DLCO 3.2. L.a 1118 DSR 3.2.1.1 iv Unit 2 Revision 35 June 2012

TABLE OF CONTENTS (Cont)

SECTION SUBJECT REF SECTION PA GE 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 11 19 2-3 Gaseous Effluent Dose Calculation Methodology 3.2.2 II 20 3.2.3 3.2.5 2.3.1 W, and W, - Dispersion Parameters For Dose, Table D 3-23 1120 2.3.2 Gamma Air Dose Due to Noble Gases 3.2.2 1121 DSR 3.2.2.1 2.3.3 Beta Air Dose Due to Noble Gases 3.3.2 1121 2.3.4 Organ Dose Due to 1-131, 1-133, Tritium and Particulates 3.2.3 with Half-Lives Greater than 8 Days 3.2.5 DSR 3.2.3.1 DSR 3.2.5.1 1121 2.4 1-133 and 1-135 Estimation II 22 2.5 Isokinetic Sampling 11 22 2.6 Use of Concurrent Meteorological Data vs. Historical Data 11 22 2.7 Gaseous Radwaste Treatment System Operation 3.2.4 11 22 2.8 Ventilation Exhaust Treatment System Operation 3.2.5 11 23 3.0 URANIUM FUEL CYCLE 3.4 II 24 3.1 Evaluation of Doses From Liquid Effluents DSR 3.1.2.1 11 25 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 II27 3.4 Doses to Members of the Public Within the Site Boundary 4.1 II 27 4.0 ENVIRONMENTAL MONITORING PROGRAM 3.5 II 30 4.1 Sampling Stations 3.5.1 11 30 DSR 3.5.1.1 4.2 Interlaboratory Comparison Program DSR 3.5.3.2 II 30 4.3 Capabilitics for Thermolumincsccnt Dosimcters Uscd for Environmental Measurements II 31 V Unit 2 Revision 35 June 2012

TABLE OF CONTENTS (Cont)

SECTION SUBJECT REF SECTION PAGE Appendix A Liquid .Dose Factor Derivation 1166 Appendix B Plume Shine Dose Factor Derivation 1169 Appendix C Dose Parameters for Iodine 131 and 133, Particulates and Tritium 11 73 Appendix D Diagrams of Liquid and Gaseous Radwaste Treatment Systems and Monitoring Systems 11 83 Appendix E Nine Mile Point On-Site and Off-Site Maps 11106 vi Unit 2 Revision 35 June 2012

LIST OF TABLES PART !- RADIOLOGICAL EFFLUENT CONTROLS TABLE NO TITLE PA GE D 3.1.1-1 Radioactive Liquid Waste Sampling and Analysis 13.1-2 D 3.2.1-1 Radioactive Gaseous Waste Sampling and Analysis .13.2-2 D 3.3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation 13.3-6 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 1 3.5-10 D 3.5.1-3 Detection Capabilities for Environmental Sample Analyses 1 3.5-11 vii Unit 2 Revision 35 June 2012

LIST OF TABLES (Cont)

PART II - CAL CULA TIONAL METHODOLOGIES TABLE NO TITLE PAGE D2-1 Liquid Effluent Detector Response 11 33 D 2-2 thru D 2-5 Ajat Values - Liquid Effluent Dose Factor II 34 D3-1 Offgas Pretreatment Detector Response 11 38 D 3-2 Finite Plume - Ground Level Dose 11 39 Factors from an Elevated Release D 3-3 Immersion Dose Factors 11 40 D 3-4 thru D 3-22 Dose And Dose Rate Factors, Ri II 41 D 3-23 Dispcrsion Parameters at Controlling .11 60 Locations, X/Q, W, and W, Values D 3-24 Parameters for the Evaluation of Doses to 11 61 Real Members of the Public from Gaseous and Liquid Effluents D5.1 Radiological Environmental Monitoring II 62 Program Sampling Locations viii Unit 2 Revision 35 June 2012

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 11107 D 5.1-2 Nine Mile Point Off-Site Map (page 1 of 2) 11 108 D 5.1-2a Nine Mile Point Off-Site Map (page 2 of 2) 11109 ix Unit 2 Revision 35 June 2012

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 I1. 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 11, 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 NUR(EG-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 35 June 2012

PART I - RADIOLOGICAL EFFLUENT CONTROLS Unit 2 Revision 35 I June 2012

Definitions 1.0 PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 1.0 DEFINITIONS Unit 2 Revision 35 11.0-0 June 2012

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 FUNCTIONAL FUNCTIONALITY is an attribute of Structures, Systems, or Components (FUNCTIONALITY) (SSCs) that is not controlled by Technical Specifications. An SSC shall be functional or have functionality when it is capable of performing its specified function as set forth in the Current Licensing Basis (CLB).

FUNCTIONALITY does not apply to specified safety functions, but does apply to the ability of non-Technical Specifications SSCs to perform specified support functions.

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 fox 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.

(continued)

Unit 2 Revision 35 11.0-1 June 2012

Definitions 1.0 1.0 DEFINITIONS (continued)

TERM DEFINITION 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.

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.

(continued)

Unit 2 Revision 35 11.0-2 June 2012

Definitions 1.0 1.0 DEFINITIONS (continued)

TERM DEFINITION 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 andlor IEPA 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.

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 35 I 10-3 June 2012

Definitions 1.0 FIGURE D 1.0-1 NMP U.ii,2

- JAF Liquid Disrharge Lake Ontario Site Area and Land Liquid DisIah-r.

Portion of Exclusion Area Boundaries

.1 Note: National Grid retains ownership in certain transmission I I I I I I I I I I I line and switchyard facilities within the exclusion area boundary.

0 1 Access and usage are controlled by Nine Mile Point Nuclear Scale (Miles)

Station, LLC by agreement.

Unit 2 Revision 35 11.0-4 June 2012

PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 3.0 APPLICABILITY Unit 2 Revision 35 1 3.0-0 June 2012

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 3.0.6, 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 35 13.0-1 June 2012

Liquid Effluents Concentration D13.1.1 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 dissolvedor entrained noble gases; and

b. 2 x 10-4 jCi/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 In accordance with activity 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. .L1-are maintained within the limits of DLCO 3.1.1.

Unit 2 Revision 35 13.1-1 Junc 2012

Liquid Effluents Concentration D 3.1.1 Table D 3.1.1-1 (P1age I of 2)

Radioactive Liquid Waste Sampling and Analysis SAMPLE LOWER LIMIT OF SAMPLE TYPE SAMPLE ANALYSIS SAMPLE DETECI'ION LIQUID RELEASE TYPE FREQUENCY FREQUENCY ANALYSIS (LLD) (a)

1. Batch Waste Release Grab Sample Each Batch (g) Each Batch (g) Principal 5 x 10-' pCilml Tanks (b) Gamma Emitters (c)

a. 2LWS-TK4A

b. 2LWS-TK4B 1-131 I x I0.

wt.Ci/ml

c. 2LWS-TKSA

d. 2LWS-TK5B Giab Saniple Ot: btcdt/31 31 days Dissolved and I x I0- pCi/mI days (g) Entrained Gases (gamma emitters)

Propoltional Each batch (g) 31 days 11-3 I x 10" pCi/lril Composite of 7

grab samples Gross Alpha I x 10 pCi/ml (d)

Proportional Each batch (g) 92 days Sr-89 3.%104 pLCi/nl Composite of grab samples (d)

Sr-90 5 x 10 tCi/ml Fe-55 1 x 10.6 ACi/ml 7

2. Con tinuous Releases Grab Sample 31 days (e) 31 days (e) Principal 5x 10. ACi/ml Gamma

a. Service Water Emitters (c)

Effluent A

b. Service Water Grab Sample 31 days (c) 31 days (e) 1-131 I x 10t) ACi/ml Effluent B

c. Cooling rower Blowdý.,'n Grab Sample 31 days (e) 31 days (e) Dissolved and I x 10-' p0Ci/ml Entrained Gases (gamma emitters)

Grab Sample 31 days (e) 31 days (e) 11-3 I x 10-, pCi/ml Grab Sample 31 days (e) 31 days (e) Gross Alpha I x 10-7 PCi/ml Grab Sample 92 days (c) 92 days (e) Sr-89 5 x I0"8 11Ci/m1 Grab Sample 92 days (e) 92 days (e) Sr-90 5 x 108 utCi/mi Grab Sample 92 days (e) 92 days (e) Fe-55 Ix 10"6 pCi/ml

3. Continuous Release Grab Sample 31 days (1) 31 days (f) Principal 5 x 10"IpCi/mIl Gamma Auxiliary Boiler Emitters (c)

Pump Seal and Sample Cooling Grab Sample 92 days (f) 92 days (f) H-3 I x 10 iCihnl Discharge (Service Water)

Unit .2 Revision 35 1 3.1-2 June 2012

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 concelration 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 repre~sents a "real" signal.

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

lLD (4 . 66 )(Sb)

(E)(V) (2.22x 106) (Y) e ý-'

where:

LLD = The before-the-fact lower limit of detection (ptCi 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 10" = The number of disintegrations per minute per .Ci, Y The fractional radiochemical yield, when applicable,

?. = The radioactive decay constant for the particular radionuclide (see"), 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-(b) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated.

and then thoroughly mixed by the meltod described in Part 1l, Section 14 to assure representative sampling.

(c) The principal gamma emitte.5 for which the LLD applies include the following radionuclidcs: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-14 1. Ce-144 shall also be measured,but with an LLD of 5 x 10'gsCi/ml. 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 nucl ides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3 in the fomat outlined in RG 1.21, Appendix B, Revision 1, June 1974.

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

(e) If the alarm setpouit of'the effluent monitor is exceeded, the frequency of sampl ing shall be increased to daily until the condition no longer exists. Fiequency ofraalysis shall be iacica.cd tu daily foi pi incipal gajiuna cinittem$ and an iicident complosite 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 of analysis shall beincreased to daily for principal gamma emitters and an incident composite for 11-3, gross alpha, Sj-89, Si-90, and Fe-55.

(g) Complete prior to each release.

Unit 2 Revision 35 13.1-3 June 201.2

Liquid Effluents Dose D 3.12 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 <5 10 mrem to any organ during any calendar year.

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. 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)

UNRESTRICTED AREAS and exceeds limits. (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 3.1.2.

(continued)

Unit 2 Revision 35 1 3.1-4 June 2012

liquid Effluents Dose D 3.1.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose to a 1B.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 times the limits. storage tanks, the on-site ISFSI, etc.). I AND Immediately B.2 Verify that the limits of DLCO 3.4 have not been exceeded.

C. Required Action B.2 and C. I Prepare and submit to the 30 days Associated Completion NRC, pursuant to D 4.1. 1, a time not met. Special Report, as defined in 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 35 1 3.1-5 Junc 2012

Liquid Effluents Dose

.D 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE I 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 35 1 3.1-6 June 2012

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 FUNCTIONAL.

APPLICABILITY: At all times.

ACTIONS

NO TES ---

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

COMI CONDITION REQUIRED ACTION COMtPLETION T"IME 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 nonfunctional unit, to UNRESTRICTED equipment or subsystems, AREAS would exceed and the reason for the 0.06 rnrem to the whole nonfunctionality, body or 0.2 mrem to any (2) Action(s) taken to restore organ in a 31 day period. the nonfunctional equipment to AND FUNCTIONAL status, and (3) Suiunary description of Any portion of the liquid action(s) taken to prevent a radwaste treatment system recurrence.

not in operation.

Unit 2 Revision 35 I 3.1-7 June 2012

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 filly utilized.

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

Unit 2 Revision 35 1 3.1-8 June 2012

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 bcyond the SITE BOUNDARN (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. I 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 rddionuclides in particulate forin 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 35 1 3.2-1 June 2012

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 SAMPLE SAMPLE ANALYSIS SAMPLE DETECTION GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ANALYSIS (I.I.D) (a)

TYPE

1. Containment (b) Grab Sample Each Purge (h) Principal Gamma Emitiers I X I1 4pCi/mIn (c)

Each Purge 11-3 (oxide) I x I.Y.iCi!ml Each Purge Principal 1 . IlpCi/ml 0-Gamma Emitters (c) 2 Main Stack, Grab Sample 31 days (d) 31 days (d) Principal I x 10'.Cil Radwaste/Reactor Gamma Emitlers Building Vent (c)

Grab Sample 31 days (e) 31 days (e) 11-3 (oxide) Ix 10plaCi/m!

Charcoal Continuous (t) 7 days (g) 1-131 Ix 10- p.Ciiml Sample Particulate Continuous (1) 7 days (g) Principal I x 10-" pCiiml Sample Gamma Emitters (e) 1 Gross Alpha lx rti/mI Composite Continuous (f) 92 days Sr-89 1 x 10"1 pCi/mt Particulate Sample Sr-90 I x 10" pCi/mI See the notes on the next page.

Unit 2 Revision 35 I 3.2-2 June 2012

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

Radioactive Gaseous 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, thal 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 radiochernical separation:

LLD = (4.06)(Sb)

(E) (V)(2.22x 106) (Y)e 4-'

where:

L..LD The before-the-fact lower limit of detection (jaCi per unit mass or volume),

St = 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 snmple size (units of MaS3or volume),

2 22 x 10' = The number of disintegrations per minute per ýtCi.

Y The fractional radiochemical yield, when applicable, X = The radioactive decay constant for the particular radionuclide (see-), 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 LID is defined as a befoTre-the-fact limit representing the capability of a measurement system and not as an after-the-fact limit for a particular measutrement.

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

(c) The principal gamma emitters for which the LLD applies include the following radionuclides: Kr-87, Kr.-88, Xe-133. Xe-133m.

Xe-135. and Xe-138 in noble gas releases and Mn-54, Fe-59, C-SR, Co-60, 7n-65, Mo-99, 1-131, Cs-134. C%,-117, Ce-141 and Cc- 144 in iodine and particulate releases. 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 analyzed and reported in the Radioactive Elfluent Release Report pursuant to Technical Specification 5.6.3 in the format outlined in RO 1.21, Appendix B, Revision I, June 1974.

(d) If the main stack or reactor/radwast.e building radiation monitor isrnot Fl FNCTIONAI., qampling ,nd analysis shall als.o he performed following shutdown, startup, or when there is an alarm on the offgas pretreatment monitor.

(c) H-3 grab samples shall be taken once every 7 days front the reaaor/radwaste ventilatiot system when fuel is omlrsoded until stable 11-3 release levels can be demonstrated.

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

(g) When the release rate of the main stack or reactor/radwaste building vent exceeds itsalarm setpoint, the iodine and particulate device shall be removed and analyzed to determine the changes in iodine and particulate relea-se rates. The analysis shall he dorne once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the release no longe1r exceeds the alarm setpoint. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analy7Zd, the corresponding LLDs may be increased by a factor of] 0.

(h) Complete prior to each release.

Unit 2 Revision 35 13.2-3 June 2012

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

-..--------------.------------------ NOTES ------ --------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION T REQUIRED ACTION COMPLETION

_TIME A. The air dose at or beyond A. I 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 effluents 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 35 1 3.2-4 June 2012

Gaseous Effluents Noble Gas Dose D 3.2.2 ACTIONS (continued)_

CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose to a B.I 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 gaseous effluents due to units (including outside noble gases exceeds 2 times the limits.

storage tanks, the on-site ISFSI, etc.). I AND Immediately B.2 Verify that the limits of DLCO 3.4 have .not been exceeded.

C. Required Action B.2 and C. 1 Special Report, as defined in 30 days Associated Completion 10 CFR 20.2203 (a)(4), of 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 35 1 3.2-5 June 2012

Gaseous Effluents Noble Gas Dose D 3.2.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.2.1 Determine cumulative dose contributions for the 3 1 days current calendar quarter and current calendar year.

Unit 2 Revision 35 13.2-6 June 2012

Gaseous Effluents Dose - I-131, l-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 D 3.2 RADIOACTIVE GASEOUS EFFLUENTS D 3.2.3 Gaseous Fffluents Dose 131, 1-133, H-3 and Radioactive Material in Particulate Form DICO 3.2-3 The dose to a MEMBER OF THE P1IJRITC 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

...----------------------........-------------------- NOTES --------------------------------------------. -

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.] Prepare and submit to the NRC, 30 days E-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 exceeding the limit(s) and beyond the SITE (2) Defines the corrective actions BOUNDARY 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 35 1 3.2-7 June 2012

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 which includes contributions radioactive materials in from direct radiation from the gaseous effluents exceeds 2 units (including outside storage times the limits. tanks, the on-site ISFSI, etc.). I 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 (aX4), of Required not met. Action A.1 .shall also include the following:

(I)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 35 1 3.2-8 June 2012

Gaseous Effluents Dose - I- 131, l- 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 1-13 1, 1-133, H-3 and radioactive material in particulate form with half-lives > 8 days.

Unit 2 Revision 35 1 3.2-9 June 2012

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

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

LCO 3.0.3 is not applicable.

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

air ejector system is being discharged without treatment.

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

(1) Identification of any nonfunctional equipment or subsystems and the reason ior the nonfunctionality, (2) Action(s) taken to restore the nonfunctional equipment to FUNCTIONAL status, and (3) Summary description of action(s) taken to prevent a recurrence.

Unit 2 Revision 35 13.2-10 June 2012

Gaseous Radwaste Treatment System D)3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE JFREQUENCY 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 35 13.2-11 June 2012

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 FUNCTIONAL.

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. The radioactive gaseous A. I Prepare and submit to the 30 days 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 nonfunctional equipment or Projected doses in 31 days subsystems and the reason from iodine and particulate for the nonfunctionality, releases, from each unit, to (2) Action(s) taken to restore areas at or beyond the SITE the nonfunctional BOUNDARY (see Figure L) equipment to 1.0-1) would exceed 0.3 FUNCTIONAL status, and mrem to any organ of a (3) Summary description of MEMBER OFTHE action(s) taken to prevent a PUBLIC. recurrence.

Unit 2 Revision 35 1 3.2-12 June 2012

Ventilation Exhaust Treatment System D 3.2.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.5.1 --------------------- NOTE ----------------

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 35 I 3.2-13 June 2012

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

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

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. VENTING or PURGING A. 1 Suspend all VENTING and Immediately of the drywell and/or PURGING of the drywell suppression chamber not and/or suppression chamber.

through the standby gas treatment system.

Unit 2 Revision 35 I 3.2-14 June 2012

Venting or Purging D 3.2.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.6.1 The drywelI 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 before start of PURGING 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 35 1 32-15 June 2012

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 FUNCTIONAL with:

a. The minimum FUNCTIONAL 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

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

1. LCO 3.0.3 is not applicable.

2. Separate condition entry is allowed for each channel.

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

OR A.2 Declare the channel Immediately nonfunctional.

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

(continued)

Unit 2 Revision 35 1 3.3-1 June 2012

Radioactive l[iquid Effluent Monitoring Instrumentation D 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

~ .3_______ S.--

B. One or more required B. 1 Enter the Condition referenced Immediately channels nonfunctional. in Table D 3.3.1-1 for the channel.

AND B.2 Restore nonfunctional 30 days channel(s) to FUNCTIONAL status.

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

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

+

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

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

__........____ -. ~

... ._____________ ~ .

(continued)

Unit 2 Revision 35 1 3.3-2 June 2012

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 ACTIONS (continued)

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

Action B. I 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. As required by Required F.1 Estimate tank liquid level. Immediately Action B. I and refcrcnccd in 'Fable D 3.3.1-1. AND During liquid additions to the tank

+ 4 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 Effluent Release nonfunctionality was not Report corrected in a timely manner.

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

or E not met.

I. Required Action and 1.1 Suspend liquid additions to Immediately associated Completion Time the tank monitored by the for Condition F not met. nonfunctional channel(s).

Unit 2 Revision 35 13.3-3 June 2012

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 SURVEILLANCE REQUIREMENTS

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

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 35 1 3.3-4 June 2012

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 35 1 3.3-5 June 2012

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

Radioactive L.iquid Effluent Monitoring Instrrumentation APPLICABILITY REQUIRED CONDITIONS OR OTHER CHANNELS REFERENCED SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTION B. I REQUIREMENTS Radioactivity Monitors Providing Alarm and Automatic Termination of Release Liquid Radwasie Effluent (a) I C DSR 3.3. I.1 Line DSR 3.3.1.3 DSR 3.3.1.5 DSR 3.3, 1.g

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

a. Service Water EUfluent (a) 1 D DSR 3.3.1. I Line A DSR 3,3.1,4 DSV 3.3.1.7 DSR 3.3.1.8

b. Service Water Eltluent (a) 1 D DSR 3.3. 1.1 Line B DSR 3,3.1.4 DSR 3.3.1.7 DSR 3.3.1.8

c. Cooling lower (a) I D DSR 33. I.1 Blowdown Line DSR 33.1.4 DSR 3.3.1.7 DSR 3.3.1.8

3. Flow Rate Measurement Devices a, Liquid Radwaste (a) I E DSR 3.3.1.2 Ettluent Line DSR 3.3.1.6 DSR 3.3.1.9

b. Service Water Effluent (a) I E DSR 3.3.1.2 Line A DSR 3.3.1.6 DSR 3.3.1.9 C. Service Water Effluent (a) I E DSR 3.3.1.2 Line B DSR 3.3.1.6 DSR 3.3.1.9 d, Cooling Tower (a) I E DSR 3.3.1.2 Blowdown Line DSR 3.3.1.6 DSR 3.3.1.9

4. Tank Level Indicating (b) 1 F DSR 3.3.1.1 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.

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

Unit 2 Revision 35 1 3.3-6 June 2012

Radioactive Gaseous Effluent Monitoring Instrumentation

) 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 FUNCTIONAL with:

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

b. The alarm/trip setpoints of Offgas Noble Gas Activity Monitor set to ensure that the limit of Technical Specification LCO 3.7.4 is not exceeded.

c. The alarm/trip setpoints of Radwaste/Reactor Building Vent Effluent Noble Gas Activity Monitor and Main Stack Effluent Noble Gas Activity Monitor set to ensure that the limits of DLCO 3.2.1 are not exceeded.

APPLICABILITY: According to Table D 3.3.2-1.

ACTIONS

.NOTES -----------------------------------------------

I. LCO 3.0.3 is not applicable.

2. Separate condition entry is allowed for each channel.

CONDITION REQUIRED ACTION J COMPLETION TIME A. Gaseous effluent A.I 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 Immediately nonfunctional.

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

(continued)

Unit 2 Revision 35 I 3.3-7 June 2012

Radioactive Gaseous Effluent Monitoring Instrumentation

.D 3.3.2 ACTIONS (continued) ..... ....

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

AND B.2 Restore nonfunctional 30 days channel(s) to FUNCTIONAL status.

i i C. As required by Required C.I Place the nonfunctional 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B. 1 and referenced channel in the tripped in Table D 3.3.2-1. condition.

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 35 1 3.3-8 June 2012

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (continued)... ..

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D. I 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 nonfunctional 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 Establish continuous 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Action B. 1 and referenced sampling using auxiliary in Table D 3.3.2-1. sampling equipment as required in Table D 3.2.1-1.

F. As required by Required F.1.1 Take grah samples. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B. 1 and referenced in Table D 3.3.2-1. AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND F. 1.2 Analyze samples for gross 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from time activity with a radioactivity of sampling limnit of detection of at least completion I x 104 "Ci/ml.

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

OR F.2.2 Through a CR, deterrr 14 days (1) The cause(s) of t nonfunctionality.

(2) The actions to be and the schedule restoring the systi FUNCTIONAL s (contmnued)

Unit 2 Revision 35 1 3.3-9 June 2012

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (_ontinu.ed.)_.---

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

H. Required Action and H.1.1 Initiate action to implement 1 Hour associated Completion appropriate compensatory Time for Condition C, D, E actions.

or F.1 not met.

AND H. 1.2 Obtain Shift Manager 12 Hours approval of the compensatory actions and the plan for exiting this Condition H.

Unit 2 Revision 35 1 3.3-10 June 2012

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 31.3.2.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DSR 3.3.2.2 Perform CHANNEL CHECK. 7 days DSR 3.3.2.3 Perform SOURCE CHECK. 31 days DSR 3.3.2.4 Perform CHANNEL FUNCTIONAL TEST. The 31 days 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 alarn/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 DSR 3.3.2.6 Perform CHANNEL FUNCTIONAL TEST. The 92 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 downsuale failure, and instrument controls not set in operate mode.

(continued)

Unit 2 Revision 35 13.3-11 June 2012

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

SI JRVFILLANCF FRFQI JFNCY DSR 3.3.2.7 Perform CHANNEL CALIBRATION. The initial 24 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 trom 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.

DSR 3.3.2.10 Perform CHANNEL CALIBRATION. 24 months Unit 2 Revision 35 1 3.3-12 June 2012

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

Radioactive Gaseous Effluent. Monitoring ln.stnmentation REQUIRED CONDITIONS APPLICAI3L1.,ITY OR CI ANNELS REFERENCED OTHER SPECIFIED PER FROM REQUIRED SURVEILLANCE INSTRUMENT CONDITIONS INSTRUMENT ACTION B. I REQUIREMENTS Offgas Systen

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

- Providing DSR 3.3.2,7 Alari-i aid Automatic Termination of Release

b. System Flow- (a) DSR 3.3:2.1 Rate Measuring DSR 3.3.2.5 Device DSR 3.3.2.10

2. Radwaste/Reaetor Building Vent Effluent System

a. Noble Gas (b) F DSR 3.3.2.1 Activity Monitor DSR 3.3.2.3 (c) DSR 3.3.2.6 DSR 33.2.9

b. Iodine Sampler (b) I E DSR 3.3.2.2

c. Particulate (b) I E DSR 3.3.2.2 Sailplel

d. Flow-Rate (b) D DSR 3.3.2.1 Monitor DSR 3.3.2.5 DSR 3,3.2.8 C. Sample Flow- (b) D DSR 3.3.2.1 Rate Monitor DSR 3.3.2.5 DSR 3,3.2.8 (continued)

(a) Du.iikg offgas system uperatitn.

(b) At all times.

(c) Includes high range noble gas monitoring capability.

Unit 2 Revision 35 1 3.3-13 June 2012

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. I REQUIREMENTS

3. Main Stack Effluent (b)

a. Noble Gas F DSR 3.3.2.1 Activiy M.onitor DSR 3.3.2.3 (c) DSR 3.3.2.6 DSR 3.3.2.9 (b)

b. Iodine Sampler E DSR 3.3.2.2 (b)

C. Particulate DSR 3.3.2.2 Sampler (b) D

d. Flow-Rate DSR 3.312.1 Monitor DSR 3.3.2.5

[) DSR 3.3.2.8 (b)

e. Sample Flow- 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 35 1 3.3-14 June 2012

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 TI IE 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.

APPLICABILITY: At all times.

ACTIONS

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPI.ETION 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. --------- 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 35 1 3.4-1 June 2012

Radiological Environmental Monitoring Prooram D S.5.1 D 3.5 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.I Prepare and submit to the NRC In.accordance with Monitoring Program not in the Annual Radiological the Annual conducted as specified in Environmental Operating Radiological Table D 3.5.1-1. Report, a description of the Environmental reasons for not conducting the Operating Report program as required and the frequency plans for preventing a recurrence.

B. Level of radioactivity in an B.1 --------- NOTES-------

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

levels of Table D 3-.5.1-2 2. For radionuclides other than when averaged over any those in Table.D 3.5. 1-2, this calendar quarter. report shall indicate tire methodology and parameters OR used to estimate the potential annual dose to a MEMBER OF THE PUBLIC.

(continued)

Unit 2 Revision 35 13.5-1 June 2012

Radiological Environmental Monitoring Program

.D 1 .5.1 ACTIONS (continued)

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 I + to be taken to reduce reporting level I 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 ofD 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 lucation which are radioactivitv/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 35 1 3.5-2 June 2012

Radiological Environmental Monitoring Program D 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Milk or fresh leafy C.I 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 3,51-I.

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.

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

D.2 Report all deviations frorn the In accordance with sampling schedule in the i the Annual Annual Radiological Radiological Environmental OperatinE Environmental Report. Operating Report (continued)

Unit 2 Revision 35 13.5-3 June 2012

Radiological Environmental Monitoring Program D 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Samples required by Table E. I Choose suitable alternative 30 days 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 35 1 3.5-4 June 2012

Radiological Environmental Monitoring Program D31.5.1 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 35 1 3.5-5 June 2012

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

Radiological Environmental Monitoring Program EXPOSURE NUMBER OF SAMPLING AND PATHWAY SAMPLE COLLECTION TYPE AND FREQUENCY ANDIOR 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 per3 Radiation monitoring one ineach months stations (b) meteorological 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. Airbome 5 locations (1) 3 samples from offsitc Continuous samplci Radioiodiinc canistco.

Radioiodine locations close to the site operation with Analyze weekly for 1-131 and boundary (within I mile) sample collection Particulates in different sectors (e) weekly or more Particulate sampler:

(2) 1 sample from the vicinity frequently if (1) Analyze for gross beta of an established year- required by dust radioactivity >Ž 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months round community (e) loading following filter change (0.

racontol (2) Perform gamma isotopic (3) 1 sample atfrom location, leasta t0miles mil analysis (g) on each in which grosssample beta distant and in aleast activity is> 10 times the prevalent wind direction previous yearly mean of (d) control samples.

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

(by location) once per 3 months

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

b. Ground As required From one or two sources if Grab sample once (1) Gamma isotopic likely to be affected (0) per 3 months analysis of each sample (g) once per 3 months (2) H-3 analysis of each sample once per 3 months (continued)

Unit 2 Revision 35 1 3.5-6 June 2012

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

Radiological Environmental Monitoring Program EXPOSURE PATHWAY SAMPLING AND AND/OR NUMBER OF SAMPLE COLLECTION TYPE AND FREQUENCY SAMP1LE SAMPLES LOCATIONS (a) FREQUENCY OF ANALYSIS

3. Walerbome (continued)

c. Drinking I sample of each One to three ofthe nearest water When 1-131 analysis (I) 1-131 analysis on each supplies that could be affected is performed, a composite sample by its discharge (k) composite sample when the dose over a two week calculated for the period (i)- otherwise, consumption of the a composite sample water is greater than I monthly mrem/yr (I)

(2) Gro*s beta and gamma isotopic analyses of each composite sample (g) monthly (3) H-3 analysis of each composite sample once per 3 months

d. Sediment I sample From a downstream area with Twice per year Gamma isotopic analysis of from existing or potential recreational each samplc (g)

Shoreline value

4. Ingestion

a. Milk (1) 3 samples from In 3 locations within 3.5 miles Twice per month, (1) Gamma isotopic (g) and MILK (e) April through 1-131 analysis of each SAMPLING December (m) sample twice per month LOCATIONS April through Dccenabci (2) If there are In each of 3 areas 3.5-5.0 miles (2) Gamma isotopic (g) and none, then I distant (e) 1-131 analysis of each sample from sample once per month 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 teat prevalcjkt SAMPLING wind direction (d)

LOCATION

b. Fish (1) 1 sample each In the vicinity of a plant Twice per year Gamma isotopic analysis of of 2 discharge area each sample (g) on edible commercially portions twice per year or recreationally important species (n)

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

(continued)

Unit 2 Revision 35 I 3.5-7 June 2012

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-1 (page 3 ot4)

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

4. Ing¢stiun (continued)

c. Food (1) 1 sample of Any area that is irrigated by At time of harvest Gamma isotopic (g) and 1-131 Products each principal waler in wvhich liqnid plant (p) 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 dvring each of the in a least prevalent wind the harvest season similar broad diiection leaf vegetation.

Unit 2 Revision 35 13.5-4 June 2012

Radiological Environmental Monitoring Program D 35.1 Table D 3.5.1-1 (page 4 of 4)

Radiological Environmental Monitoring Program (a) Specific parameters of distance and direction sector from the centerline of one reactor, md additional descriptions where pertinent, shall be provided for each and every sample location inTable D 3.5.1 -1. Refer to NUREG-Ol33, -'Preparation of Radiolugitwal Efflucut'Tuchnical 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 amhl71rdnls conditions.seasonal unavailability (which includes theft and uncooperative residents), or malfsrction of automatic sampling equipment.

(b) One uVInule ihrstruunicts, stich as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place ol, or in addition to integrating dosimeters. Each of the 32 routine monitoring stations shall be equipped with 2 or more dosimeters or with I instrument for measuring and recording dosne rate confinurinsly For the. purpose of this table, a thermoluminescicnt dosimeter (ILD) 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 DIQ based on all site licmscd reactors.

(I) Airborne particulate sample filters shall be analy~zd 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.

(g) Gamma isotopic analysis means the identification and quantification of gamma--emitting radionuclides that may be attributable to 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 aea bLyond but near the mixing 4one.

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

(j) Groundwatce sainples .hall be takei when this source is tapped for drinking or irrigaticon purposes in areas where the hydraulic gradient or recharge properties are suitable for contanrination.

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

(I) Analysis for 1-131 may be accomplished by Ge-Li analysis providedthat the lower limit ofdetection (I..,D) 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.

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

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

(o) Applicable only to major irrigation projects within 9 miles of the site in the general dow ncurrent direction (p) tfbhrvest occurs more thadi umrcc/yea,:Sarrptiug shail bepiefumned duiinu eacti discrete harvest. Ifharvest occurscontinuously, sampling shall be taken monthly. Attention should be paid to including samples of tuberous and root food products.

Unit 2 Revision 35 13.5-9 June 2012

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 RADIONUCLIDE PARTIUCLATE OR FISH MILK PRODUCTS ANALYSIS WATER (pCiIL) GASES (pCi/mr) (pCi/kg. wet) (pCi/L) (pCi/kg, wet) 11-3 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 0OO Us-I134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba- 140 200 300 La- 140 200 300 (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 may be used.

(b) If no drinking water pathway exists, a value of 20 pCi/L may be used.

Unit 2 Revision 35 13.5-10 June 2012

Radiological Environmental Monitoring Program D 3.5.1 Table D 3.5.1-3 (page 1 of 2)

Detection Capabilities for Environmental Sample Analysis LOWER LIMIT OF DETECTION (LLD1,"

AIRBORNE PARTIUCLATE OR FOOD RADIONUCLIDE WATER GASES (pCi/m*) FISH MILK PRODUCTS SEDIMENT ANALYSIS (pCi/L) (pCi/kg. wet) (pCI/I..) (pLVkg, wet) (pCi/kg, dry)

Gross Beta 4 0.01 H--3 2,000 Id, Mln-54 15 130 Fe-59 30 260 Co-S8 15 130 Co-60 is 130 Zn-65 30 260 ZT-95 15 Nb-95 15 Mc 1-131 0.07 60 Cs-134 is 00.05 130 15 60 Iso Cs-137 18 00.06 ISO 18 80 180 Ba-140 15 15 La-140 15 15 See the notes on the next page Unit 2 Revision 35 13.5-11 June 2012

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

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

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

(b) Required detection capabilities tbr thermoluminescentt dosimeters used for environmertal 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 LELD is defined as the smallest concertration 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 conclud ing that a blank observation represents a "real" signral.

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

LLD (4.66)(S 6 )

(E) (V) (2.22) (Y) e-where:

LLD The before-the-fact lower limit of detection (pCi 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 The number of disintegrations per minute per pCi.

Y The fractional radiochemical yield, when applicable, X The radioative decay constant for the particular radionuclide (seec'), and At The elapsed tine between environmental collection or end ofthe 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 aich a manner that the stated LLDs will be achaeved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of itrerfering 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, avalue 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 35 13.5-12 June 2012

Land Use Census D 3.5.2

3.5 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 gampling controlled by Table D 3,,5.1-1, part 5.c may be performed in lieu of the garden census) > 500 ft producing broad leaf vegetation.

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. 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 35 13.5-13 June 2012

Land Use Census D 3.5.2 ACTIONS (continued)

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

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

AND B.3 Submit in. the next In accordance with Radioactive Effluent Release the Radioactive Report documentation for a Effluent Release change in the ODCM Report 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 35 1 3.5-14 June 2012

Land Use Census D 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.2.1 Conduct the land use census during the growing 366 days 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.

DSR 3.5.2.2 Report the results of the land use census in the Annual In accordance with Radiological Environmental Operating Report. the Annual Radiological Environmental Operating Report Unit 2 Revision 35 1 3.5-15 June 2012

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

1. LCO 3.0.3 is not applicable.

2. ,CO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Analyses not performed as A. I 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 35 13.5-16 June 2012

Interlaboratory Comparison Pruoram D .3.5.3 SURVEIILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.3.1 Report a summary of the results obtained as part of In accordance with the Interlaboratory Comparison Program in the the Annual Annual Radiological Environmental Operating Radiological Report. Environmental Operating Report Unit 2 Revision 35 1 35-17 June 2012

PART I - RADIOLOGICAL EFFLUENT CONTROLS BASES Unit 2 Revision 35 I B 3.1-0 June 2012

Liquid Effluents Concentration B 3.1.1 B 3.1 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 THE PUBLIC and (2) the levels required by 10 CFR 20.1301(c) 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, I-ASL-300 (revised annually).

Unit 2 Revision 35 IB 3.1-1 June 2012

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 IL.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 inmplement 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 1, and if direct radiation doses from the units including the on-site Independent Spent Fuel Storage Installation (ISFSI), and the 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 ard 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 35 I B 3.1-2 June 2012

Liquid Radwaste Treatment System B 3.1.3 B33.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.3 Liquid Radwaste Treatment System BASES The installed liquid radwaste treatment system shall be considered FUNCTIONAL by meeting DLCO 3.1.1 and DLCO 3.1.2. The FUNCTIONALITY 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 requircmcnts 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 35 I B 3.1-3 June 2012

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, Colunm 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 Il. 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 mrem/year 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 35 I B 3.2-1 June 2012

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 ll.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 1, and if direct radiation doses from the units including the on-site ISFSI, and the 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 gases 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 I," 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 35 I B 3.2-2 June 2012

Gaseous Effluents Dose- Iodine-131, 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, lI.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 achievable. The calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, 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 the on-site ISFSI, and the 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 comrilitrnent 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 calculational methodology and. parameters for calculating the doses from the actual release 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 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. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate DLCO for iodine-131, iodine-133, tritium, and radioactive material in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas at or beyond the SITE BOUNDARY. The pathways that were examined in the development of these calculations were: (1) individual inhalation of airborne radioactive material, (2) deposition of radioactive material onto green leafy vegetation with subsequent consumption by man, (3)

Unit 2 Revision 35 I B 3.2-3 June 2012

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) 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 35 I B 3.2-4 June 2012

Gaseous Radwaste Treatment System B 3.2.4 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.4 Gaseous Radwaste Treatment System BASES The FUNCTIONALITY 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 IL.B and I.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 35 IB 3.2-5 June 2012

Ventilation Exhaust Treatment System B 3.2.5 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.5 Ventilation Exhaust Treatment System BASES The FUNCTIONALITY 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 thifs 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, awd the design objectives given in Section I.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 I1,B and [l.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 FUNCTIONAL 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 FUNCTIONAL, iodine and particulate dose projections will be made for 3.1-day intervals starting with filter nonfunctionality, and continuing as long as the filter remains nonfunctional, in accordance with DSR 3.2.5.1.

Unit 2 Revision 35 1 B 3.2-6 June 2012

Venting or Purging B 3.2.6 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.6 Venting or Purging BASES This provides reasonable assurance that releases from drywell and/or suppression chamber purging operations will not exceed the annual dose limits of 10 CFR 20 for unrestricted areas.

Unit 2 Revision 35 I B 3.2-7 June 2012

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 alarmltrip will occur before exceeding ten times the limits of 10 CFR 20. The FUNCTIONALITY 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 35 I B 3.3-1 June 2012

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 seLpoints for these instruments shall be calculated and adiusted 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 I OCFR20. 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 offgus 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 10CFR 100 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 FUNCTIONALITY 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 35 I B 3.3-2 June 2012

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 itoern). If the dose to any MEMBER OF TIE 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 MEMBER 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 35 I B 3.4-1 June 2012

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 Environmeental 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 35 1 B 3.5-1 June 2012

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 mininiurn required to produce die 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 vegetation (i.e., similar to lettuce and cabbage) and (2) the vegetation yield was 2 kg/m2.

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 35 1B 3.5-2 June 2012

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 35 I B 3.5-3 June 2012

PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 4.0 ADMINISTRATIVE CONTROLS Unit 2 Revision 35 14-0-0 June 2012

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

Unit 2 Revision 35 14.0-1 June 2012

Special Reports D4.1.1 D4.1.2

.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 4.1.2Annual 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 ofjoint 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 forom the unit during the previous year.

(Continued)

Unit 2 Revision 35 14.1-1 June 2012

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 nonfunctionality 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 35 1 4.1-la June 2012

Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment System D 4.2 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 PORC. 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 supplcmcntal 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 amcndmcnts 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 35 14.2-1 June 2012

Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment System D 4.2 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 PORC.

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

Unit 2 Revision 35 14.2-2 June 2012

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. I 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-I) 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 uCi/mI 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 Fj (c*,/cFj), 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, Ej (CGjCFd) cpm, by total concentration of activity Yi (cG), 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:

Yi(CGi/CFi) = Yi(CGC)/CF Where the factors are as defined above.

Unit 2 Revision 35 112 June 2012

The calculations of the required dilution factors (RDF) are performed as follows:

RDF,= I MEC gamma .fraction = Z(CG,/MECi) i i RDFTOTAL = MEC total fraction = Z (Ci/MECG) i I RDF, is used to calculate the liquid radwaste effluent radiation monitor setpoint. This monitor is a gamma detector and has little or no response to non gamma emitters. Use of RDF, rather than RDFTOTAL, to determine the monitor setpoint prevents the condition where a tank with gamma concentrations near their LLD cannot be discharged due to spurious alarms received because the setpoint is close to the monitor background.

RDFTOTAL is used to determine the minimum dilution factor required to discharge the tank contents based on all activity, both gamma and non gamma, in the tank. This ensures that the concentrations of all radioactive materials released in liquid effluents will meet DLCO

3. 1.1. Non gamma emitting nuclide activity, except tritium was 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, respectively, the concentration of Co-60.

Currently, non gamma activity except tritium is estimated using the results from the latest analysis of composite samples.

Tritium concentration is assumed to equal the latest concentration detected in the monthly tritium analysis 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 nucl ides 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.

Unit 2 Revision 35 113 June 2012

Normal Radwaste Effluent Alarm Setpoint Calculation:

Alarm Setpoint < 0.8

TDF/PEF

TGC/CF

I/RDF7 + 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 PEF = The permissible Radwaste Effluent Flow rate, gpm, 165 gpm is the maximum value used in this equation TDF/PEF An approximation to (TDF + PEF)/PEF, the Actual Dilution Factor in effect during a discharge.

TGC = ZCG = Summation of all gamma emitting nuclides (which monitor will respond to)

CGi - Concentration of gamma emitting nuclide in Radwaste tank prior to dilution, [tCi/ml CF Monitor Conversion Factor, ýtCi/mllcpm, determined at each calibration of the effluent monitor TGC/CF = An approximation to Ei (CGi/CFi) using CF determined at each calibration of the effluent monitor RDFy = Y-i (CG 1/MECj)= The total fraction often 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 Gamma (RDFy)-

Background = Detector response when sample chamber is filled with nonradioactive water, cpm Ci - Concentration of isotope i in Radwaste tank prior to dilution, RCi/ml (gamma + non-gamma emitters)

CFi - Detector response for isotope i, net pzCi/ml/cpm See Table D 2-1 for a list of nominal values MECj = Maximum Effluent Concentration, ten times the limiting effluent concentration for isotope i from 10 CFR 20 Appendix B, Table 2, Column 2, p.Ci/ml Y (CGI/CFi) = The total detector response when exposed to the concentration of nuclides in the Radwaste tank, cpm RDFTotai=Zi (ci/MECO) = The total fraction often 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-Total and includes both the gamma and non-gamma emitters.

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

Unit 2 Revision 35 114 June 2012

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

PEF = TDF (RDFIoial) 1.5 If Actual Dilution Factor is set equal to the Required Dilution Factor, then the alarm points required by thc 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 radwastc 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. If no gamma emitters are detected in the Radwaste Tank samples, then the radiation monitor setpoint will be based on assuming gamma activity at the LLD of the most limiting nuclide from recent discharges. In 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 he calculated.

FMEC = F - MECI)]

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 Unit 2 Revision 35 115 June 2012

Where:

Alarm Setpoint = The Radiation Detector Alarm Setpoint, cpm 0.8 - Safety Factor, Unitless F, = An Effluent flow rate for stream s, gpm Cj = Concentration of isotope i in Radwaste tank prior to dilution, pCi/ml Cis Concentration of isotope i in Effluent stream s including the Radwaste Effluent tank undiluted, pCi/ml CF Average detector response for all isotopes in the waste stream, net pCi/ml/cpm MEFCi - Maximum Effluent Concentration, ten times the effluent concentration limit for isotope i from 10CFR20 Appendix B, Table 2, Column 2, /Ci/ml PEF = The permissible Radwaste Effluent Flow rate, gpm Background = Detector response when sample chamber is filled with nonradioactive water, cpm TGC/CF - The total detector response when exposed to the E. (CGi/CF) concentration of nuclides in the Radwaste tank, cpm Es. [F.Ci] = The total activity of nuclide i in all Effluent streams, ptCi-gpm/ml Eý; [Fj] = 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 tunncl 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.

Unit 2 Revision 35 116 June 2012

Detector response y (c.i/cr-j) 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.

Service Water and Cooling Tower Blowdown Alarm Setpoint Equation:

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

Where:

Alarm Setpoint = The Radiation Detector Alarm Setpoint, cpm 0.8 = Safety Factor, unitless Ct - Concentration of isotope i in potential contaminated stream, j.Ci/mI CF, = Detector response for isotope i, net ptCi/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 Yi (CI/CFi) The total detector response when exposed to the concentration of nuclides in the potential contaminant, cpm Y. (C1/MEC) The total fraction of ten times the IOCFR20, Appendix B, Table 2, Column 2 limit that is in the potential contaminated stream, unitless.

(U/CF)Ei*C An approximation to .i (C1/CFv), determined at each calibration of the effluent monitor CF Monitor Conversion Factor, pCi/ml/cpm 1.2 Liquid Effluent Concentration Calculation This calculation documents compliance with Section D 3.1.1 of Part I:

Unit 2 Revision 35 11 7 June 2012

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 cntraincd 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 E. [ Fs/I& (F,) J~ (Cjs;-MECj)]I Where: FMEC = The Fraction of Maximum Effluent Concentration, the ratio at the point of discharge 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, pCi/ml F, = The flow rate of a particular effluent stream s, gpm MECi = Maximum Effluent Concentration, ten times the limiting Effluent Concentration of a specific nuclide i from IOCFR20, Appendix B, Table 2, Column 2 (for noble gases, the concentration shall be limited to 2E-4 microcurie/ml), ltCi/ml YX(C4o/MEcQ) = The Maximum Effluent Concentration fraction of stream s prior to dilution by other streams V (Fr) = The total tlow 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.

Unit 2 Revision 35 118 June 2012

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 = Y.[Ait YrIATLCiLFL)]

Where:

D, = The cumulative dose commitment to the total body or any organ, t from the liquid effluents for the total time period ET, (ATL) , mrern 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 AT,. from any liquid release, liCi/ml Ail= The site related ingestion dose commitment factor for the maximum individual to the total body or any organ t for each identified principal gamma or beta emitter, mrem/hr per tCi/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 I Jnit 2 Environmental Report - Operating License Stage, Table 5.4-2 footnote 1.

These factors can he 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/miin)] / [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, Ai, 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.

When actual results for the non-gamma emitter, (Gross Alpha, H-3, Sr-89, Sr-90 and Fe-55), concentrations are not available, dose contributions are calculated using concentration estimates. The doses are revised when actual results are obtained.

1.4 Liquid Effluent Sampling Representativeness There are four tanks in the radwaste system designed to be discharged to the discharge Unit 2 Revision 35 119 June 2012

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 spargcr cffectively 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 as;sure 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/R!vl 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 FUNCTIONALITY The Liquid Radwaste Treatment System shall be FUNCTIONAL 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 Unit 2 Revision 35 if 10 June 2012

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 wvill bc processed through the regenerant evaporator or Thermex System.

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

Unit 2 Revision 35 i1 11 June 2012

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. I 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 mremryr to the whole body and less than or equal to 3000 mremryr to the skin, and

b. For iodine-131, for iodine-] 33, for tritium, and for all radionuclides with half-lives greater than 8. days: Less than or equal to 1500 mrernryr to any organ.

The radioactivity rate of noble gases measured at the recombiner effluent shall be less than or equal to 350,000 microcuries/second after decay of 30 minutes during offgas system operation in accordance with Technical Specification 3.7.4.

2.1.2 Setpoint Determination Methodology Discussion Nine Mile Point Unit I 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 mRern/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/scc after 30-minute delay. This is the release rate for which a FSAR accident analysis was completed. At this rate the Offgas System charcoal beds will not contain Unit 2 Revision 35 11 12 June 2012

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 1, Section D 3.3.2, the Radiation Monitoring System on 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.

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 a beta-gamma sensitive detector. It is able to accurately quantify the activity released in terms of uCi of Xe-133 equivalent activity. 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.BR

ý iR (0ij Alarm Setpoint, pCi/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 Q = The release rate of nuclide i, jtCi/sec V= 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, rnrem/yr per utCi/sec XU. (Qi) The total release rate of noble gas nuclides in the stack effluent, pCi/sec i (Qivi) The total of the product of each isotope release rate times its respective whole body plume constant, mrem/yr The alert alarm is normally set at less than 10% of the high alarm.

Unit 2 Revision 35 11 13 June 2012

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 arc separate monitors, not otherwise discussed in the ODCM). Nominal flow rate for the vent is 2.37E5 CFM.

This detector is a beta-gamma sensitive detector. It is able to accurately quantify the activity released in terms of uCi of Xe-133 equivalent activity. 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.

O.8R .i(Qi)

Alarm Setpoint, uCi/sec < (X/Q)v'i(QjKi)

Where:

0.g = 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, ItCi/sec (XJQ)ý = 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/in 3 Ki The constant for each identified noble gas nuclide accounting for the whole body dose from the3 semi-infinite cloud, listed on Table D 3-3, mrem/yr per pCi/m F.i (Qi) The total release rate of noble gas nuclides in the vent effluent, uCi/sec Yj (QOK}) The total of the product of the each isotope release rate times its3 respective whole body immersion constant, mrcm/yr per sec/m The alert alarm is normally set at less than 10% of the high alarm.

Unit 2 Revision 35 11 14 June 2012

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. Calculation H21C-070, Offgas Radiation Monitor (2OFG-RE13A and 13B)

Alert and Alarm Setpoint Determination, assumes a distribution of offgas corresponding to that expected with the design limit for fuel failure to establish the setpoint. Calculation Table 5 tabulates Alarm/Trip setpoints as a function of Offgas flow due to decay from the Recombiner discharge to the monitor/sampling location. The monitor design response values are confirmed during periodic calibration using a Transfer Standard source traceable to the primary calibration performed by the vendor.

Particulates and lodines 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 scfin, (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 less than or equal to 1.5 timies nominal full power background (average, +3 standard deviations) to ensure that the Specific Activity Action required by ITS SR 3.7.4.1, are implemented in a timely fashion.

(ACT) (2.12 E-03) i,(Cj/CFJ) + Background Alarm Setpoint, cpm < 0.8 F i (CW)

Where:

Alarm Setpoint The alarm setpoint for the offgas pretreatment Noble Gas Detector, cpm 0.8 = Safety Factor, unitless ACT = The Technical Specification Limit for Offgas Pretreatment, tpCi/sec, equivalent to 350,000 jiCi/sec after 30 minutes decay.

2.12E-03 - Unit conversion Factor, 60 sec/min / 28317 cc/CF Ci - The concentration of nuclide, i, in the Offgas, RCi/cc CFj - The Detector response to nuclide i, ýiCi/cc/cpm; See Table D 3-I for a list of nominal values F = The Offgas System Flow rate, CFM Unit 2 Revision 35 11 15 June 2012

Background The detector response to something other than noble gases that will be released after 30 minutes decay. Includes purge background, response to activation gases, and response to fission gases with half'life less than 4 minutes, cpm F (ci/CF) The summation of the nuclide concentration divided by the corresponding detector response, net cpm

- (Cc) = The summation of the concentration of nuclides in offgas, jtCi/cc NOTE: Calculation H21C-070 indicates Activity at the Monitor will range from 9.64 E+05 jtCi/sec to 1.22 E +06 l.Ci/sec. A conservative Offgas Activity Release Rate of 900,000 jiCi/sec applies to this location at all flow rates.

2.2 Gaseous Effluents Dose Rate Calculation Methodology 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:

When actual results for the gross alpha, particulate, iodine, H-3, Sr-89, Sr-90 and Fe-55, concentrations are not available, dose contributions are calculated using concentration estimates. The doses are revised when actual results are obtained.

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 disper.ion 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 arc conservative Unit 2 Revision 35 11 16 June 2012

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 factored into the plume shine dose factors for stack releases and through the use of 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 Ej [ViQj. + Ki (X/Q),,Qi.,]

Where:

DRy = Whole body dose rate (mrem/sec)

Vi = 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 jiCi/sec Ki The constant accounting for the gamma whole body dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed in Table D 3-3, mrem/yr per uCi/m 3 (From Reg.

Guide 1.109)

X/Qv 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 (sec/mr3)

Qis,Qi, The release rate of each noble gas nuclide i, from the stack (s) or vent (v). Averaged over the time period of concern. (jtCi/sec)

Unit 2 Revision 35 It 17 June 2012

3.177E-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 (P3) and gamma (y) radiation of the semi infinite cloud by using an immcrsion 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)y- due to noble gases:

(DR)yp = 3.17E-8 *i[(Li(X/Q)s+1.11Bi)Qis+(Li+I.llMi) (X/Q)IQ.i0 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 p.Ci/m 3 , listed on Table D 3-3 (from R.G. 1.109)

Mj 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 tCi/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 Bi 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 piCi/sec.

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(X/Q)S The relative plume concentration at or beyond the land (X/Q)X 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). (sec/re')

3.17E-8 = Conversion Factor; the inverse of the number of seconds in a year; (yr/sec)

Qi",Qis The release rate of each noble gas nuclide i, from the stack(s) or vent (v) averaged over the time period of concern, piCi/sec.

2.2.4 Organ Dose Rate Due to 1-131, 1-133, Tritium, and Particulates with Half-lives greater than 8 days.

The organ dose rate is calculated using the dose factors (Ri) from Appendix C. The factor Ri takes into account the dose rate received from the ground plane, inhalation and ingestion pathways. W, 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 rate is averaged over the period of concern.

Organ dose rates (DR)at due to iodine-131, iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days:

(DR)a t = 3.17E-8 Ej[EiRijat [W.Q 1 s + WQiv] 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 inhalation pathway, mrem/yr per pCi/m-3 . Units for ground and ingestion pathways, m2 -mrem/yr per uCi/sec. (See Tables D 3-4 through D 3-22).

W",W, Dispersion parameter either X/Q (sec/mr3) or D/Q (I/m 2) 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 W, values are used for vent releases (v=vent).

Qi., QIV The release rates for nuclide i, from the stack (s) and vent (v) respectively, p.Ci/sec.

When the release rate exceeds 0.75 VCi/sec from the stack or vent, the dose rate assessment shall, also, include JAF and NMPI dose contributions. The use of the 0.75 piCi/sec release rate threshold is conservative because it is based on the dose conversion factor (Ri) for the Sr-90 child bone which is significantly higher than the dose factors for Unit 2 Revision 35 11 19 June 2012

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 FUNCTIONAL 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.

When actual results for the gross alpha, particulate, iodine, H-3, Sr-89, Sr-90 and Fe-55, concentrations are not available, dose contributions are calculated using concentration estimates. The doses are revised when actual results are obtained.

2.3.1 WV and W.K - 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.

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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[M+/-(X/Q) Qi, + Bi Qis] x t Dy The gamma air dose for the period of concern, mrad t The duration of the dose period of concern, scc 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:

D = 3.17E-8 EtiNi[(X/Q)v Qiv + (X/Q), Qss] x t Dp = Beta air dose (mrad) for the period of concern Ni = 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 IOCFR50 dose limits. The factor R1 takes into account the dose received from the ground plane, inhalation, food (cow milk, cow meat and vegetation) pathways. W, 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 Ri factors are discussed in Appendix C.

Organ dose D, due to iodine-131, iodine-133, tritium and radionuclides in particulate form Unit 2 Revision 35 1121 June 2012

with half-lives greater than 8 days.

Da, = 3.17E-8 ji [ Yi Riiat [WS Qi. + 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 IE-12 uCi/cc. If detected in excess of the LLD, the 1-131 and 1-133 analysis results will be reported directly from each cartridge analyzed. Periodically, (usually quarterly but on a 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 l-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 I- 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 1-13-3/I-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 It is the intent to use dispersion parameters based on historical mctcorological data to set alarm points and to determine or predict dose and dose rates in the environment due to gaseous effluents. If effluent levels approach limiting values, meteorological conditions concurrent with the. time of release may bc used to detcrminc gaseous pathway doses.

2.7 Gaseous Radwaste Treatment System Operation Part I, Section D 3.2.4 requires the GASEOUS RADWASTE TREATMENT SYSTEM to be in operation whenever the main condenser air ejector system is in operation. The system may be operated for short periods with the charcoal beds bypassed to facilitate 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).

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2.8 Ventilation Exhaust Treatment System Operation Part 1, Section D 3.2.5 requires the VENTILATION EXHAUST TREATMENT SYSTEM to bc FUNCTIONAL 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 FUNCTIONAL are:

1) HEPA Filter - Radwaste Decon Area

2) HEPA Filter - Radwaste Equipment Area

3) HEPA Filter - Radwaste General Arca Whenever one of these filters is not FUNCTIONAL, iodine and particulate dose projections will be made for 31-day intervals starting with filter nonfunctionality, and continuing as long as the filter remains nonfunctional, 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.)

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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 variai.uce 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.

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 Unit 2 Revision 35 11 24 June 2012

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 hal f-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 = Ej [Cif (U) (Daipj) f] (1E+3)

Where:

Rapj 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 Rapj = R*,b and Daipj = DiWB Cif The concentration of radionuclide i in fish samples in pCi/gram U The consumption rate of fish IE+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-1 I) (mrem/pCi). ex. when calculating to the adult whole body Daipj = DiwE, f The fractional portion of the year over which the dose is applicable The dose from shoreline sediment sample media is calculated as:

Unit 2 Revision 35 1125 June 2012

Rapj ý Ei [Cis (U) (4E+4) (0.3) (Daipj) f3 Where:

Rapi 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 Raj = RwB and Dipj = DjWB Ci = 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 "aThedose factor for age group a, nuclide i. sediment pathway s, and organ

j. (Reg. Guide 1.109, Table E-6) (mrem/tr per pCi/rm2); ex. when calculating to the adult whole body Daipj = DiwB 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.

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.

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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, the Independent Spent Fuel Storage Installation (ISFSI), and outside radioactive storage tanks (as applicable) may be evaluated by engineering calculations or by evaluating environmental TLD results at critical receptor I

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 die 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 Energy Center and the Unit I 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 (including the ISFSI) 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.

Unit 2 Revision 35 1127 June 2012

The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium and particulates) in the cffiucnt 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 Regulatory Guide 1.109. Since many of the factors are in units of pCi/m 3, m 3/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.

D5, Ej (((Cj) F (X!Q) (DFA) ij,(BR).tl Where:

Dja = 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 DFAtja DFAiL Ci The average concentration in the stack or vent release of nuclide i for the period in pCi/mn3 .

F Unit 2 average stack or vent flowrate in m 3/see.

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)ij, 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 DFAIj, = DFAiL (BR)a annual air intake for individuals in age group a in M3 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).

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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. Theresultant 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 cvaluated.

The direct radiation gamma dose pathway includes any gamma doses from an overhead plume, submersion in the plume, possible radiation from the facility (including the ISFSI) and ground plane dose (deposition). This general pathway will be evaluated by average I

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.

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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 Coiminssion 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 35 1130 June 2012

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 ILD 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.

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 Unit 2 Revision 35 11 31 June 2012

to be constant. The TLDs shall be exposed under two conditions: (1) packaged in a thin, scaled 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 cxposcd 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 he evaluated.

Unit 2 Revision 35 11 32 June 2012

TABLE D 2-1 LIQUID EFFLUENT DETECTORS RESPONSES*

NUCLIDE (CPMI/ICi/ml 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 Cc 144 0.103 Br 84 1.12 1131 1.01 1 132 2.63 1133 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 35 11 33 June 2012

TABLE D 2-2 Ajý, VAIUES - I.,IQIID' ADULT mrem - ml hr- uCi NUCLIDE T BODY Cl-TRACT BONE LIVER KIDNEY THYROID LUNG q 3 3.67E-1 3.67E-1 -- 3.67E- I 3.67E-1 3.67E-1 3,67E-1 Cr51 1.26 3.13E2 1.18E-2 1.18E-2 2.86E-I 7.56E-1 1.66 Cu 64 1.28 2.332E -- 2.73 6.89 -- --

Mn 54 8.38E2 1.3424 3.98 4.38E3 1.31 E3 3.98 3.98 Fe 55 1.07E2 2.62E2 6.62E2 4.57E2 - -- 2.55E2 Fe 59 9.2812 8.06E3 1.03E3 2.42E3 7.53E-I 7.53E-1 6.76E2 Co 57 5.43E1 5.36E2 -- 2.111 -- -- --

Co 58 2:01E2 1.81E3 1.07 9.04E 1 1.07 1.07 1.07 Co 60 6.36E2 4.93E3 6.47E] 3.24E2 6.47E1 6.47E1 6.471]

Zn 65 3.32E4 4,63E4 2.312A 7.35E4 4.92E4 2.21 2.21 Sr 89 6.38E2 3.57E3 2.22E4 6.18E-5 6.18E-5 6.18E-5 6.18E-5 Sr 90 1.36E5 1.60E4 5.55E5 - .......

Sr 92 1.44E-2 6.61 3.34E-1 ........

Zr 95 7.59E-1 2.83E2 9.77E-1 7.88E-I 8.39E-1 6.99E-I 6.99E-1 Mn 56 3.07E-2 5.52 -- 1.73E-1 2.20E-1 ....

Mo 99 1.60EI 1.95E2 1.97E-3 8.42E! 1.91E2 1.97E-3 1.97E-3 Na 24 1.34E2 1.34E2 1.34E2 1.34 F2 1.34122 1.34E2 1.34F2 1131 1.16E2 5.36EI 1.42E2 2.03 E2 3.4822 6.65 E4 2.77E-2 1 132 4.34E-3 2.33E-3 4.64E-3 1.24E-2 1.98E-2 4.34E-i --

1 133 1.22E 1 3.59EI 2.30E1 3.99E1 6.97EI 5.87E3 1 135 1.32ED 3.79E0 1.2 NEU 3.36E0 5.39E0 2.22E2 Ni 65 1.14E-2 6.35E-1 1.93E-1 2.50E-2 ..

Cs 134 5.79E5 1.24E4 2.98E5 7.08E5 2.29E5 2.04E1 7.61E4 Cs 136 9.42E4 1.33E4 2.96E4 1-17E5 6.51 F4 3.29P.-I 8-92F3 Cs 137 3.42E5 1.01E4 3.82E5 5.22E5 1.77E5 3.10EI 5.89E4 Ba 140 1.37E1 4.30E2 2.09E2 3.04E-I 1.31E-1 4.17E-2 1.92E-1 Ce 141 3.79E-2 8.81EI 6.93E-2 5.83E-2 4.60E-2 3.53E-2 3.53E-2 Nb95m 1.51E] 1.44E6 3.53E1 2.74E1 2.70E1 --...

Nb95 1.31E2 1.48E6 4.38E2 2.44E2 2.41E2 3.56E-1 3.56E-I La 140 1.62E-2 3.72E3 1.03E-I 5.36E-2 2.83E-3 2.83E-3 2.83E-3 Ce 144 3.03E-1 6.15E2 2.02 9.66E- I 6.57E- I 2,06E-I 2.06E-1 Tc 99m 2.05E-2 9.54E-01 5.7 1E-4 1.61 E-3 2.45E-2 -- 7.90E-4 Np 239 1.8E-3 4.47E2 2.28E-2 2.78E-3 7.40F-3 5.95F.-4 5.95F-4 Te 132 .1.18E3 5.97E4 1.95E3 1.26E3 1.22E4 1.39E3 2.66E-3 Zr 97 5.08E-4 3.39E2 5.44E-3 1.10E-3 1.662-3 7.11E-6 7.11E-6 W 187 4.31E I 4.04E4 1.48E2 1.23E2 4.43E-5 4.43E-5 4.43E-5 Ag IOta 1.09EI 3.94E2 1.14E1 1.13EI 1.22E] 1.04E1 1.04EI Sb 124 4.72E1 3.36E2 1.07E3 4.33E1 4.31I1 4.31E1 5.12El Zn 69m 5.40E1 3.60E4 2.46E2 5.90E2 3.57E2 6.90E-2 6.90E-2 Au 199 3.95 7.33E2 1.26E-1 4.67 1.79E! 1.26E-1 1.26E-1 As 76 5.94 1.24E4 1.60E-I 6.19 1.16E] 1.60E-1 1.60E-I Calculated in accordance with NUREG 0133, Scction 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

Unit 2 Revision 35 1I 34 June 2012

TABLE D 2-3 A;., VALUES - LIQUID i TEEN mrcm - ml hr-uCi NUCLIDE T BODY Gl-TRACT BONE LIVER I.,IQlIIDI KIDNEY THYROID LUNG Illl - uCi 2-3

--Dml H3 2.73E-1 2.73E- I mrcm TABLE A:,I hr TEEN VALiIES 2.73E-1 2.73E-1 2.7313-1 2.73E-I Cr 51 1.35 2.16E2 6.5613-2 6.56E-2 3.47E-1 7.79E-1 1.90 Cu 64 1.35 2.23132 2.87 7.27 Mn 54 8.75E2 8.84E3 2.22E1 4.32E3 1.3 1E33 2.221E I 2.22E1 Fe 55 1.15 E2 2.131,2 6.93 E2 4.91 E2 3.11 E2 Fe 59 9.59E2 5.85E3 1.06E3 2.48E3 4.20 4.20 7.84E2 Co 57 1.44.E2 4.08E2 2.19131 Co 58 2.10112 1.23E3 5.98 9.47E1 5.98 5.98 5.98 Co 60 9.44E2 3.73E3 3.611E2 6.20E2 3.611E2 3.61 E2 3.611E2 Zn 65 3.40E4 3.08E4 2.103E4 7.28154 4.66E4 1.24E1 1.24E11 Sr 89 6.92E2 2.88E3 2.421E4 3.45E-4 3.45E-4 3.45E-4 3.45E-4 Sr 90 1.14.E5 1.30E4 4.62E5 Sr 92 1.54E-2 9.19E1 3.61 E-I Zr 95 3.96 2.10E2 4.19 3.99 4.03 3.90 3.90 Mn 56 3.22E-2 1.19El 1.81E-1 2.29E-1 Mo 99 1.71E] 1.60E2 1.101E-2 8.95E1 2.05E2 1.101E-2 1.10E-2 Na 24 1.38E2 1.38E2 1.38E2 1.38E2 1.38E2 1.38E2 1.38132 1 131 1. 14E2 4.21E1 1,52E2 2.12E2 3.66E2 6.19E4 1.55E-1 1 132 4.56E-3 5.5413-3 4.86E-3 1.27E-2 2.0013-2 4.29E- 1 1 133 1.28E I 3.17E1 2.47EI 4.19ElI 7.35E1 5.85E3 1.02E-4 1 135 1.76EO 3.84 E0 1.34E1U 3.46E0 5-47E) 2.23E2 Ni 65 1.21E-2 1.44 2.08E-1 2.66E-2 Cs 134 3.33E5 9.05E3 3.05E5 7.1 8E5 2.28E5 1.14E2 8.72E4 Cs 136 7.87E4 9.44E33 2.98E4 1.177E5 6.38E4 1.83 1.0 1E4 Cs 137 1.90E5 7.9) E3 4.09E5 5.44E5 1.85E5 1.7362 7.21E4 Ba 140 1.44E I 3.40E2 2.2 11E2 5.03E-1 3.25E-1 2.3313-1 4.15E--1 Ce 141 2.0013-I 6.85EI 2.33E-1 2.21E-I 2.08E-1 1.97E- I 1.97E-1 Nb 95m 1.69131 1.14E6 3.87E1 2.99E11 2.96E1I Nb 95 1.177E2 1.05E6 4.43E2 2.47E2 2.39E2 1.99 1.99 La 140 2.9713-2 3.01 E3 1.22E-1 6.82E-2 1.58E-2 1.58E-2 1.58E-2 Ce 144 1.25 4.83E2 3.07 1.94 1.62 1.15 1.15 Tc 99m 2.1 IE-2 1.07 5.84E-4 1.63E-3 2.43E-2 9.04E-4 Np 239 4.63E-3 3.78E2 2.82E-2 5.67E-3 1.07E-2 3.32E,-3 3.32F-3 Te 132 1.23E3 4.13134 2.06133 1.30E3 1.25E4 1.37E3 1.48E-2 Zr 97 5.68E-4 3.111E2 5.84E-3 1.19E-3 1.78.13-3 3.97E-5 3.9713-5 W 187 4.55E1 3.52E4 1.59E2 1.30E2 2.47E-4 2.47E-4 2.47E-4 Ag 110m 5.85E1I 3.17E2 5.89E1 5.88E1 5.97151 5.79E 1 5.79E]

Sb 124 2.45E2 4.53132 2.51 E2 2.4 1E2 2.41 E2 2.41 E2 2.50E2 Zn 69m 5.76E 1 3.43E4 2.65E2 6.24E2 3.79E2 3.85E-1 3.85E-1 Au 199 4.85 5.78E2 7.04E-1 5.60 2.011El 7.04E- 1 7.04E-)

As 76 7.18 1.06E4 8.92E-! 7.40 1.33E1 8.92E- 1 8.92E- 1

'Calculated in accordance with NIJREG 0133, Section 4.3. ; and Regulatory Guide 1.109, Regulatory position C, Section 1.

Unit 2 Revision 35 II 35 June 2012

TABLE D 2-4 A;., VALUES - LIQUID' CHILD mrem - ml hr - uCi NUCLIDE T BODY 0I-TRACT BONE LIVER KIDNEY THYROID LUNG H3 3.34E-1 3.34E-1 3.34E-1 3.34E- 1 3.34E-1 3.34E-1 Cr 51 1.39 7.29E1 1.37E-2 1.37E-2 2.22E-I 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 Fe 55 1.50E2 8.99E1 9.15E2 4.85E2 2.74E2 Fe 59 1.04E3 2.18E3 1.29E3 2.09E3 8.78E-1 8.78E-1 6.08E2 Co 57 6.24E1 1.62E2 2.00El Co 58 2.21 E2 4.20E2 1.25 7.30EI 1.25 1.25 !1.25 Co 60 7.03E2 1.25E3 7.55E1 2.88E2 7.55EI 7.55E1 7.55Et Zn 65 3.56E4 1.01E4 2.15E4 5.73E4 3.61E4 2.58 2.58 Sr 89 9.13E2 1.24E3 3.20E4 Sr 90 1.06.F5 5.6263 4.17E5 Sr 92 1.85E-2 8.73 4.6 1E-I Zr 95 8.95E-1 9.36E1 1.22 9.04 E- I 9.43E-1 8.15E-1 8.15E-1 Mn 56 3.73E-2 2.39E! 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 1.5 I.E2 1.5 1 E2 1.5 1E2 1.51E2 1.51E2 1.5 1FE2 1.5182 1131 1.1482 i.80E] 2.00E2 2.01E2 3.31 E2 6.66E4 3.23E-2 1132 5.08E-3 1.30E-2 6.0 1E-3 1.10E-2 1.69E-2 5.13E-I 1133 1.51E] 1.60E) 3.22EI 3.98E I 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 2.51E-2 Cs 134 1.27E5 3.28E3 3.68E5 6.04E5 1.87E5 2.38E1 6.72E4 Cs 136 6.26E4 3.40E3 3.52F4 9.67F4 5.15E4 3.82E-1 7.68E3 Cs 137 7.28E4 3.12E3 5.15E5 4.93E5 1.61E5 3.62 E1 5.78E4 Ba 140 1.87E1 1.62E2 3.19E2 3.28E-1 1.40E- I 4.87E-2 2.15E-I Ce 141 4.6 1E-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 Nb 95 1.45E2 3.75E5 5.21E2 2.03E2 1.91 E2 4.16E-I 4.16E-1 La 140 1.93E-2 1.33E3 1.39E-1 5.09E-2 3.30E-3 3.30E-3 3.30E-3 Ce 144 4.31E-1 2.92E2 3.81 1.36 8.61E-1 2.40E-1 2.40E- I Tc 99m 2.29E-2 7.87E-1 7.05E-4 1.38E-3 2.0 1E-2 7.02E-4 Np 239 2.40E-3 1.79F2 3.44F-2 3.128-3 7.70E-3 6.94.F-4 6.948-4 Te 132 1.38E3 1.15E4 2.57E3 1.148E3 1.06E4 1.6613 3. 10E-3 Zr 97 6.998-4 1.77E2 8.11 E-3 1.18E-3 1.69E-3 8.29E-6 8.29E-6 W 187 5.37E1 1.68E4 2.02E2 1.20E2 5.16E-5 5.161-5 5.16E-5 Ag I 10om 1.29E I 1.24M2 1.35E1 1.30EI 1.39E1 1.21EI 1.21EI Sb 124 5.69E1 1.68E2 6.92E1 5.06EI 5.03EI 5.04E1 6.08E1 Zn 69m 6.80E 1 1.87E4 3.37E2 5.75E2 3.34E2 8.05E-2 8.05E-2 Au 199 5.58 2.7512 1.47E-1 5.02 1.80E] 1.47E- 1 1.47E-I As 76 8.31 5.4713 1.86E-1 6.58 1.15E1 1.86E-1 1.86E-1

'Calculaled in accordance with NUREG 0133, Section 4.3.1; and Regulatory Guide 1.109, Regulatory position C, Section 1.

Unit 2 Revision 35 1136 June 2012

TABLE D 2-5 A,., VALULES - ...IQUID' INFANT mrem - ml hr - uCi NUCLIDE T BODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H3 1.87E-1 1.871-1 1.87E-1. 1.87E-1 1.87E-1 1.87E-1 Cr 51 8.21 E-3 2.39E-I 1.17E-3 5.36E-3 1.04E-2 Cu 64 1.96E-2 8.70E- 1 4.24E-2 7.17E-2 Mn 54 2.73 4.42 1.20E 1 2.67 Fe 55 1.45 6.91E-1 8.42 5.44 2.66 Fe 59 1.25E1 1.52EI 1.82EI 3.18E1 9.41 Co 57 1.13E0 2.37E0 6.95E]

Co 58 5.36 5.36 2.15 Co 60 1.55E1 1.56E£ 6.55 Zn 65 1.76EI 3.22E] 1.1 IEI 3.81E1I 1.85EI Sr 89 4.27E 1 3.06E1 1.49E3 Sr 90 2 .86E3 J.40E2 1.12E4 Sr 92 1.56E-5 4.54E-3 4.2 1E-4 Zr 95 2.12E-2 1.49E 1 1.23E-I 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 1.36E] 2.03E1 Na 24 9.61E-1 9.61E-1 9.61E-I 9.61E-I 9.6 1E-I 9.61E-1 9.61 E-I 1 131 9.78 7.94E- I 1.89EI 2.22El 2.60E I 7.31 E3 1 132 3.43E-6 7.80E-6 4.75E-6 9.63E-6 1.0713-5 4.52E-4 1 133 8.26E- I 4.77E-1 1.94 2.82 3.31 5.13E2 1 135 2.38E2 2.36E2 3.29E2 6.54E2 7.28E2 5.86E0 Ni 65 2.96E-6 4.96E-4 5.75E-5 6.51E-6 Cs 134 4.30E] 1.16 2.28E2 4.26E2 I. 10E2 4.50E1 Cs 136 2.81EI 1.14 2.56EI 7.53F.I 3.0013I 6.13 Cs 137 2.63E] 1.16 3.17E2 3.71 E2 9.95EI 4.03E1 Ba 140 4.88 2.33EI 9.48E1 9.48E-2 2.25E-2 5.82E-2 Ce 141 3.31E-3 1.45E1 4.61 E-2 2.81 E-2 8. 67E-3 Nb 95m 1.02E3 1.20E] 2.39E3 1.73E3 1.1 0E3 Nb 95 5.87E-3 857 2.47E-2 1.02E-2 7.28E-3 La 140 6.52E-4 2.98E] 6A3E-3 2.53E-3 Ce 144 1.01E-I 1,03E2 1.80 7.37E-I 2.98E- I Tc 99rm 3.17E-4 7.14E-3 1. 19E-5 2.46E-5 2.64E-4 1.28E-5 Np 239 2.08E-4 1.06E I 4.12F.-3 3.69F-4 7.34R-4 Te 132 4.08 1.62EI 8.83 4.37 2.74E1 6.46 Zr 97 1.38E-4 1.92E] 1.76E-3 3.02E-4 3.04E-4 W 187 4.13E-2 7.02 1.72E-1 1.19E-I Ag I 1Dn 2.91 E-1 2.28E] 6.02£-1 4.39E-1 6.28E- I Sb 124 3.95 3.93E1 i.27E1 1.87E-1 3.38E-2 7.98 Zn 69m 2.30E-2 3.50 1.24E-I 2.52E-1 1.02E-I Au 199 2.23E-1 5.38 2.48E-£ 6.26E-1 As 76 8.67E-2 2.85E! 8.46E-2 1.03E-1 t

Calcu lated in accordance with NUREG 0133, Section 4.3.1 and Regulatory Guide 1.109, Regulatory position C,Section I.

Unit 2 Revision 35 11 37 June 2012

TABLE D 3-1 OFFGAS PRETREATMENT*

DETECTOR RESPONSE NUCLIDE NET CPMItCi/cc Kr 83m Kr85 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 5.5 1EE+03

Values from calculation H2 IC-070 IJnit 2 Revision 35 II 38 June 2012

TABLE D 3-2 PLUME SHINE PARAMETERS' NUCLIDE B&Mrad/yr VX mrem/vr uCilsec uCi/sec Kr 83m 9.01 E-7 Kr 85 6.92E-7 Kr 85m 5.09E-4 4.91E-4 Kr 87 2.72E-3 2.57E-3 Kr88 7.23E-3 7.04E-3 Kr 89 1.15E-2 1.13E-2 Kr 90 6.57E-3 4.49E-3 Xe 131m 7.76E-6 Xe 133 7.46E-5 6.42E-5 Xe 133m 4.79E-5 3.95E-5 Xe 135 7.82E-4 7.44E-4 Xe 135m Xe 137 1.45E-3 1.37E-3 Xe 138 6.25E-4 5.98E-4 4.46E-3 4.26E-3 Xe-127 1.96E-3 1.31E-3 Ar 41 5.OOE-3 4.79E-3 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 35 11 39 June 2012

TABLE D 3-3 IMMERSION DOSE FACTORS1 2

Nuclide K_ (I-Body) 2 L(-Skin) ý(y-Air) 3 Ni (P -Air) 3 Kr 83m 7.56E-02 1.93E1 2.88E2 Kr 85m 1.17E3 1.46E3 1.23E3 1.97E3 Kr85 1.61El 1.34E3 1.72E1 1.95E3 Kr 87 5.92E3 9.73E3 6.17E3 1.03E4 Kr88 1.47E4 2.37E3 1.52E4 2.93E3 Kr89 1.66E4 1.01E4 1.73E4 1.06E4 Kr 90 1.56E4 7.29E3 1.63E4 7.83E3 Xe 131m 9.15E1 4.76E2 1.56E2 1.11E3 Xe 133m 2.5 1E2 9.94E2 3.27E2 1.48E3 Xe 133 2.94E2 3.06E2 3.53E2 1.05E3 Xe 135m 3.12E3 7.11E2 3.36E3 7.39E2 Xe 135 1.81E3 1.86E3 1.92E3 2.46E3 Xe 137 1.42E3 1.22E4 1.51E3 1.27E4 Xe 138 8.83E3 4.13E3 9.21 E3 4.75E3 Ar 41 8.84E3 2.69E3 9.30E3 3.28E3

'From, 2

Table B- 1.Regulatory Guide 1.109 Rev. 1 mrem/yr per uCi/m 3.

3 mrad/yr per uCi/m 3 .

Unit 2 Revision 35 II 40 June 2012

TABLE D 3-4 DOSE AND DOSE RATE Ri VALUES - INHALATION - INFANT' mrem/Vr uCi/m3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H3 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 C 14 2.65E4 5.311E3 5.3 1E3 5.3 11E3 5.3 1 E3 5.3 1E3 5.31 E3 Cr51 8.95E1 5.75E1 1.32E1 1.28E4 3.57E2 Mn 54 2.53 E4 4.98E3 4.98E3 1.00E6 7.06E3 Fe 55 1.97E4 1.17E4 3.33E3 8.69E4 1.09E3 Fe 59 1.36E4 2.35E4 9.48E3 1.02E6 2.48E4 Co 5 1-22E3 1.92E31 7.77E5 I.-I IE4 Co 60 8.02E3 1.188E4 4.511E6 3.19E4 Zn 65 1.93E4 6.26E4 3.1 IE4 3.25E4 6.47E5 5.14E4 Sr 89 3.98E5 1.14E4 2.031E6 6.40E4 Sr 90 4.09E7 2.59E6 I.1 2E7 1.3 IE5 Zr 95 1.15E5 2.79E4 2.03 E4 3.111E4 1.75E6 2.17E4 Nb95 1.57E4 6.43E3 3.78E3 4.72E3 4.79E5 1.27E4 Mo 99 1.65E2 3.23E1 2.65E2 1.35E5 4.87E4 Ru 103 2.02E3 6.79E2 4.24E3 5.52E5 1 .61E4 I Ag I orn 9,99F3 7.22-E3 5.00E3 1.09E4 3.67.E6 3.30E4 1-131 3.79E4 4.44E4 1.96E4 1.48E7 5.18.E4 1.06E3 I 133 t.32E4 1.92E-4 5.6013 3.56E6 2.24E4 2.16E3 Cs 134 3.96E5 7.03E5 7.45E4 1.90E5 7.97E4 1.33E3 Cs 137 5.49E5 6.12E5 4.55E4 1.72E5 7.13E4 1-33E3 Ba 140 5.60E4 5.60E1 2.90E3 1.34E1 1.60E6 3.84E4 La 140 5.05E2 2.00E2 5.15E1] 1.68E5 8.48E4 Ce 141 2.77E4 1.67E4 1.99E3 5.25E3 5.17E5 2.16E4 Ce 144 3.19E6 1.2 1 E6 1.76E5 5.38E5 9.84E6 1.48E5 Nd 147 7.94E3 8.13E3 5.00E2 3.15E3 3.22E5 3.12E4

'This and following R, Tables Calculated in accordance with NUREG 0133, Section 5.3.1, except C 14 values in accordance with Regulatory Guide 1.109 Equation C-8.

Unit 2 Revision 35 II 41 June 2012

TABLE D 3-5 DOSE AND DOSE RATE Ri VALUES - INHALATION - CHILD uCim 3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H3 1.12E3 1.12E3 1.1203 1.12 3 1.12E3 1.12E3 C 14 3.59E4 6,733E3 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 Cr51 1.54E2 8&55E1 2.43EI 1.70E4 1.081E3 Mn 54 4.29E4 9.51 E3 1.00E4 1.58E6 2.29E4 Fe 55 4.74 E4 2.52E4 7.77E3 1.11 E5 2.87E3 Fe 59 2.07E4 3.34E4 1.67E4 1.27E6 7.07E4 Co 58 1.77E3 3.16E3 1.11 E6 3.44E4 Co 60 1.3 11E4 2.26E4 7.07E6 9.62E4 Zn 65 4.26E4 1.13E5 7.031E4 7.14E4 9.95E5 1.63E4 Sr 89 5.99E5 1.72E4 2.16E6 1.67E5 Sr 90 1.01 E8 6.44E6 1.48E-7 3.43E5 Zr 95 1.90E5 4. 1 E4 3.70E4 5.96E4 2.23E6 6.11 E4 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 Ru 103 Ag 11Om 2.79E3 1.69E4 1,14 E4 1.07E3 9.14E3 7.03E3 2.12E4 6.62E5 5.48E6 4.48E4 1.00E5 I

1 131 4.81EF4 4.811E4 2.731A 1.62E7 7.88E4 2.84E3 1 133 1.66E4 2.03E4 7.70E3 3.85E6 3.38E4 5.48E3 Cs 134 6.51E5 1.0 1E6 2.25E5 3.30E5 1.211E5 3.85E3 Cs 137 9.07E5 8.25E5 1.28E5 2.82E5 1.04E5 3.62E3 Ba 140 7.40E4 6.48EI 4.33E3 2.111El 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 2.12F6 3.611E5 1.17E6 1.20E7 3.89E5 Nd 147 1.08E4 8.73E3 6.8 11E2 4.81 E3 3.28E5 8.21 E4 Unit 2 Revision 35 11 42 June 2012

TABLE D 3-6 DOSE AND DOSE RATE Ri VALUES - INHALATION - TEEN mrem/vr uCi/m3 NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI 113 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 C 14 2.60E4 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 Cr51 1.35E2 7.50E1 3.07E1 2.10E4 3.00E3 Mn 54 5,11E4 8.40E3 1.27E4 1.98E6 6.68E4 Fe 55 3.34E4 2.38E4 5.5403 1.24E5 6.3913 Fe 59 1.59E4 3.70F4 1.43F.A t.53E6 1.78E5 Co 58 2.07E3 2,78E3 1.34E6 9.52E4 Co 60 1.51E4 1.98E4 9.72E6 2.59E5 Zn 65 3.86E4 1.34E5 6.24E4 8.64E4 1.24E6 4.66F4 Sr 89 4.34E5 1.25E4 2.42E6 3.71E5 Sr 90 1.08E8 6.68E6 1.65E7 7.65E5 Zr 95 1.46E5 4.58E4 3.15E4 6.74E4 2.69E6 1.49E5 Nb 95 1.86E4 1.03E4 5.66E3 1.00FA 7.51 E5 9.68E4 Mo 99 1.69E2 3.22E] 4.11E2 1.54E5 2.69E5 Ru 103 2.10E3 .96F.2 7,43E.3 7.83E5 1.09E5 I Ag I10m 1.38E4 1.31E4 7,99E3 2.5024 6.75E6 2.73E5 1131 3.54E4 4.91E4 2.64E4 1.46E7 8.40E4 6.49E3 1133 1.22E4 2.05E4 6.22E3 2.92E6 3.59E4 1.03E4 Cs 134 5.02E5 I. 13E6 5.49E5 3.75E5 1.46E5 9.76E3 Cs 137 6.70E5 8.48E5 3.11 E5 3.04E5 1.2 1E5 8.48E3 Ba 140 5.47E4 6.70EI 3.52E3 2.28E1 2.03E6 2.29E5 La 140 4-79E2 2.36E2 6.26E1 2.14E5 4.87E5 Ce 141 2.84E4 1.90E4 2.17E3 8.88E3 6.14E5 1.26E5 Ce 144 4.. 8(E6 2.02E6 2.62E5 1.21 E6 1.34E7 8.64E5 Nd 147 7.86E3 8.56E3 5.13E2 5.02E3 3.72E5 1.82E5 Unit 2 Revision 35 II 43 June 2012

TABLE D 3-7 DOSE AND DOSE RATE Ri VALUES - INHALATION - ADULT mrcm/yr 3

uCi/m NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI RH3 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 C 14 1.82P4 3.41E3 3.41 E3 3.41E3 3.41133 3.41E3 3.41 E3 Cr 51 1.00E2 5.95EI 2.28EI 1.44L4 3.32E3 Mn 54 3.9614 6.30E3 9.84E3 1.40E6 7.74E4 Fe 55 2,46E4 1.70E4 3.94E3 7.21E4 6.03 E3 Fe 59 L.18E4 2.78E4 1.06E4 1.02E6 1.88E5 Co 58 1.5813 2.07E3 9.28E5 1.06E5 Co 60 1.15E4 1.48E4 5.97E6 2.85E5 Zn 65 3.24E4 1.03E5 4.66E4 6.90E4 8.64E5 5.34E4 Sr 89 3.04E5 8.72E3 1.40,6 3.50E5 Sr 90 9.92E7 6. 10E6 9.60E6 7.22E5 Zr 95 1.07E5 3.4E4 2.33E4 5.42E4 1.77E6 1.50E5 Nb 95 1.41E4 7.82E3 4.21 E3 7.74E3 5.05E5 1.04E5 Mo99 1.21 E2 2.30E1 2.91E2 9.12E4 2.48E5 Ru 103 1.53E3 6.58E2 5.83E3 5.05E5 1.10E5 I Ag I lore 1.38E4 1.31E4 7.99E3 2.50E4 6.75E6 2.73E5 1131 2.52E4 3.58E4 2.05E4 1.19E7 6.13E4 6.28E3 1133 8.64E3 1.48E4 4.52E3 2.15E6 2.58E4 8.88E3 Cs 134 3.73E5 8.48E5 7.28E5 2.87F5 9.76E4 1.04E4 Cs 137 4.78E5 6.21 E5 4.28E5 2.22E5 7.52E4 8.40E3 Ba 140 3.90E4 4.90EI 2.57E3 1.67EI 1.27E6 2.18E5 La 140 3.44E2 1.74E2 4.58E1 1.36E5 4.58E5 Ce 141 1.99E4 1.35E4 1.53E3 6.2603 3.62E5 1.20E5 Ce 144 3.43 E6 1.43.6 1.84E5 8.48E5 7.78E6 g.16E5 Nd 147 5.27E3 6.10E3 3.65E2 3.56E3 2.212E5 1.73E5 Ag I1hOm 1.38E'1 1.31E4 7.99E3 2.50E4 6.75E6 2.73E5 Unit 2 Revision 35

!I 44 June 2012

TABLE D 3-8 DOSE AND DOSE RATE Rj VALUES - GROUND PLANE ALL AGE GROUPS m -mrem/yr uCi/sec NUCLIDE TOTAL BODY SKIN FI 3 C 14 Cr 51 4.65E6 5.50E6 Mn 54 1.40E9 1.64E9 Fe 55 Fe 59 2.73E8 320E8 Co 58 3.80E8 4.45E8 Co 60 2.15EI0 2.53E10 Zn 65 7.46E8 8.57E8 Sr 89 2.16E4 2.51 E4 Sr 90 ..

Zr 95 2.45E8 2.g5E8 Nb 95 1.36E8 1.61 E8 Mo 99 3.99E6 4.63E6 Ru 103 1.08E8 1.26E8 Ag 110m 3.44E9 4.01E9 1 131 1.72E7 2.09E7 1 133 2.39E6 2.91E6 Cs 134 6.83E9 7.97E9 Cs 137 1.03EI0 1,20El0 Ba 140 2.05E7 2.35E7 La 140 1.92E7 2.18E7 Ce 141 1.37E7 1.54E7 Ce 144 6.96E7 8.07E?

Nd 147 8.46E6 1.01 E7 Unit 2 Revision 35 1145 June 2012

TABLE D 3-9 DOSE AND DOSE RATE Rj VALUES - COW MILK - INFANT m 2 -mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KII )NEY LUNG GI-LLI H 3" 2.38E3 2.3 8E3 2.38E3 2.38E3 2.38E3 2.38E3 C 14 3.23E6 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 Cr 51 8.35E4 5.45E4 I. 19FA 1 .06E5 2.43 E6 Mn 54 2.5 1E7 5.68E6 5.56E6 9.21 E6 Fe 55 9.43E7 5.44E7 1.45E7 2.66E7 6.91 E6 Fe 59 1.22E8 2.13E8 8.3 8E7 6.29E7 1.02E8 Co 58 1.39E7 3.46E7 3.46E7 Co 60 5.90E7 1.39E8 1-40E8 Zn 65 3.53E9 1.21E10 5.5 8E9 5.87E9 1.02E10 Sr 89 6.93E9 1.99E8 1.42E8 Sr 90 8.19E10 2.09E10 1.02E9 Zr 95 3.85E3 9.39E2 6.66E2 1.0O1E3 4.68E5 Nb 95 4.2 1E5 1.64E5 1.17E5 1.54E5 3.03E8 Mo 99 1.04E8 2.03E7 1.55E& 3.43E7 Ru 103 4.65E3 1.55E3 9.6713 5.65E4 I Ag Ir0m 2.46E8 1.79E8 1.19E8 2.56118 9.29E9 1131 6.81E 8.02E8 3.53F8 2.64 E.11 9.37E8 2.86E7 1133 8.52E6 1.24E7 3.63E6 2.26E9 1.46E7 2.10E6 Cs 134 2.41E10 4.49E 10 4.54E9 1.16E10 4.74E9 1.22E8 Cs 137 3.47E10 4.06E10 2.88E9 1.09E 10 4.41E9 1.27E8 Ba 140 1.2.1ES 1.21 E5 6.22E6 2.87E4 7.42E4 2.97E7 La 140 2.03E I 7.99 2.06 9.39E4 Cc 141 2.28E4 1.39E4 1.64E3 4.28E3 7.18E6 Ce 144 1.49E6 6.10E5 8.34F4 2.46FE5 8.5467 Nd 147 4.43E2 4.55E2 2.79E1 1.76E2 2.89E5 "mrem/yr per pCi//m3.

Unit 2 Revision 35 11 46 June 2012

TABLE D 3-10 DOSE AND DOSE RATE Ri VALUES - COW MILK - CHILD m2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H3 1.57E3 1.57E3 1.57E3 1.57E3 1.57E3 1.57E3 C 14' 1.65E6 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 Cr 51 5.27E4 2.93E4 7.99E3 5.34E4 2.80E6 Mn 54 1.35E7 3.59E6 3.78E6 1.13E7 Fe 55 6.97E7 3.07E7 1.15E7 2.09E7 6.85E6 Fe 59 6.52E7 1.06E8 5.26E7 3.06E7 1.10E8 Co 58 6.94E6 2.13E7 4.05E7 Co 60 2.89E7 8.52E7 1.60E&

Zn 65 2.63E9 7.00E9 4.35E9 4.41E9 1.23E9 Sr 89 3.64E9 1.04E8 1.41E8 Sr 90 7.53E10 1.91E10 1.01E9 Zr 95 2.17E3 4.77E2 4.25E2 6.83E2 4.98E5 Nb 95 1.86E5 1.03E4 5.69E4 1.00E5 4.42E8 Mo 99 4.07E7 1.0187 8.69E7 3.37E7 Ru 103 Ag 110m 2.29E3 1.33E8 8.97E7 8.82E2 7.17E7 5.78E3 1.67E&

5.93E4 1.07E10 I

1131 3.26E8 3.28E8 1.86E8 1.08E8I 5.39Eg 2.92E7 1133 4.04E6 4.99E6 1,89E6 9.27E8 8.32E6 2.01E6 Cs 134 1.50E10 2.45E10 5.18E9 7.61E9 2.73E9 1.32E8 Cs 137 2.17E10 2.08E10 3.07E9 6.78E9 2.44E9 1.30E8 Ba 140 5.87E7 5.1484 3.43E6 1.67E4 3.07E4 2.97E7 La 140 9.70 3.39 1.14 9.45E4 Ce 141 1.15E4 5.73E3 8.5 1E2 2.5 1E3 7.15E6 Ce 144 1.04E6 3.26E5 5.55E4 1.80E5 8.49E7 Nd 147 2.24E2 1.81E2 1.40E1 9.94E1 2.87E5 mrem!yr per g.Ci/m 3.

Unit 2 Revision 35 1147 June 2012

TABLE D 3-11 DOSE ANT) DOSE RATE P, VALUES - COW MILK - TEEN m2-mrem/yr uCilsec

- TEEN NUCLIDE BONE LIVER T. BODY THYROID 3-11RATE KIDNEY MILK LUNG GI-LLI TABLE DOSE D DOSE

.AND

- 9.94132 uCi/sec 11 3 9.94E2 1*COW VALUES 9.94E2 mZ-mremlyr 9.94132 9.94E2 9.94E2 C 14' 6.70,5 1.34E5 1.34E5 1.34E5 1.34E5 1.35E5 1.34E5 Cr51 2.58E4 1.44E4 5.66E3 3.69E4 4.34E6 Mn 54 9.01 E6 1.79E6 2.69E6 1.85E7 Fe 55 2.78E7 1.97E7 4.59E6 1.25E7 8.52E6 Fe 59 2.81 E7 6.57E7 2.54E7 2.07E7 1.55E8 Co 58 4.55E6 1.05E7 6.27E7 Co 60 1.86E7 4.19E7 2.42Eg Zn 65 1.34E9 4.65E9 2.17E9 2.97E9 1.97E9 Sr 89 1.47E9 4.21 E7 1.75E8 Sr 90 4.45E10 1.10E10 1.25E9 Zr 95 9.34E2 2.95E2 2.03E2 4.33E2 6.803E5 Nb 95 1.86E5 1.03E5 5.69E4 1.00E5 4.42E8 Mo 99 2.24E7 4.27E6 5.12E7 4.01E7 Ru 103 9.70E2 4.15E2 3.42E3 8. 1OE4 I Ag I lOr 6.13E7 5.80E7 3.53E7 1.11 E8 1.63E 10 1 131 1.34E8 1.88E8 1.01 E8 5.49E10 3.2413E 3.72E7 1133 1.66E6 2.82E6 8.59E5 3.93E8 4.94E6 2.13E6 Cs 134 6.49E9 1.53E10 7.08E9 4.85E9 1.85E9 1.90E8 Cs 137 9.02E9 1.20EI0 4.18E9 4.08E9 1.59E9 1.71)E Ba 140 2.43E7 2.98E4 1.57E6 i.0LE4 2.00E4 3.75E7 La 140 4.05 1.99 5.30E-I 1.14E5 Ce 141 4.67E3 3.12E3 3.58E2 1.47E3 8.911E6 Ce 144 4.22E5 1.74F5 2.27F4 1.41)45 1.06g.9 Nd 147 9.12E1 9.91E1E 5.94E0 5.82EI 3.58E5 mrern/yr per p~Ci/m 3 Unit 2 Revision 35 11 48 June 2012

TABLE D 3-12 DOSE AND DOSE RATE Rj VALUES - COW MILK - ADULT m2 -mrem/yr uCi/sec NUCLIDE DONE LIVER T. BODY THYROID K1DNEY LUNG GI-LLI H3 7.63E2 7.63E2 7.63E2 7.6 3E2 7.63E2 7.63E2 C 14 3.63E5 7.26E4 7.26E4 7.26E4 7.2:6E4 7.26E4 7.26E4 Cr 51 1.48E4 8.85E3 3.2:6E3 1.96E4 3.72E6 Mn 54 5.411E6 1.03E6 -- 1.6 51E6 1.66E7 Fe 55 1.57E7 1.08E7 2.52E6 ..-- 6.04136 6.211E6 Fe 59 1.61 E7 3.79E7 1.45E7 .... 1.06E7 1.26E8 Co 58 2.70E6 6.05E6 5.47E7 Co 60 1.10E7 2.42E7 2.06E&

Zn 65 8.71E8 2.77E9 1.25E9 1.85E9 1.75E9 Sr 89 7.99E8 2.29E7 1.28E8 Sr 90 3.15E10 7.74E9 9.11 ES Zr 95 5.34E2 1.711E2 1.16E2 2.69E2 5.43E5 Nb95 1.09E5 6.07E4 3,27E4 6.00E4 3.69E8 Mo 99 1.24E7 2,36E6 2.81 E7 2.87E7 Ru 103 5.45E2 2.35E2 2.08E3 6.34E4 I Ag I 10m 3.71 E7 3.43E7 2.04E7 6.74E7 1.40E 10 1131 7.41E7 1.06FE8 6.08E7 3.47FE10 1.82F& 2.80E7 1133 9.09E5 1.58E6 4,82E5 2.32E8 2.76E6 1.42E6 Cs 134 3.74E9 8.89.E9 7.27E9 2.88E9 9.55E8 1.56E8 Cs 137 4.97E9 6.80E9 4.46E9 2.31E9 7.68E8 1.32ES Ba 140 1.35E7 1.69E4 8.83E5 5.75E3 9.69E3 2,77E7 La 140 2,26 1.14 3.01E-1 8.35E4 Ce 141 2.54E3 1.72E3 1.95E2 7.99E2 6.58E6 Ce 144 2.29E5 9.58E4 1.23E4 5.68E4 7.74E7 Nd 147 4.74E1 5.48E] 3.28E0 3.20E] 2.63E5

~mrem/yr per pCi/rn3 I Jnit 2 Revision 35 II 49 June 2012

TABLE D 3-1.3 DOSE AND DOSE RATE R, VALUES - GOAT MILK - INFANT m2 -mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI 1-3" 6.33E3 6.33E3 6.33E3 6.33E3 6.33E3 6.33E3 3.23E6 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 Cr51 I.00E4 6.56E3 1.43E3 1.28E4 2.93E5 Mn 54 3.01 E6 6.82E5 6.67E5 1.11 E6 Fe 55 1.10E6 7.08E5 1.89E5 3.46E5 8.98E4 Fe 59 1.59E6 2.78E6 1.09E6 8.21E5 1.33E6 Co 58 1.67E6 4.16E6 4.16E6 Co 60 7.08E6 1.67E7 1,68E7 Zn 65 4.24E8 1.45E9 6.70E8 7.04E8 1.23E9 Sr 89 1.48E10 4.24E8 3.04E8 Sr 90 1.72E 11 4.38E10 2.15E9 Zr 95 4.66E2 1.13.E2 8.04E1 1.22E2 5.65E4 Nb 95 9.42E4 3.88E4 2.24E4 2.78E4 3.27E7 Mo 99 1.27E7 2.47E6 1.89E7 4.17E6 Ru 103 5.57E2 1.86E2 1. 16E3 6.78E3 I Ag I lOre 2.95E7 2.15E7 1.43E7 3.07E7 1.11E9 1131 g. ! 7E8 9.63E8 4.23E& 3.16El1 1.12E9 3.44E7 1.133 1.02E7 1.49E7 4.36E6 2.71E9 1.75E7 2.52E6 Cs 134 7.23E10 1.35E11 1.36EI0 3.47E10 1.42E 10 3.66E8 Cs 137 1.04E 11 1.22E11 8.63E9 3.27E10 1.32E 10 3.81E8 Ba 140 1.45E7 1.45E4 7.48E5 3.44E3 8.91E3 3.56E6 La H40 2.430 9.59E-1 2.47E-1 1.13E4 Ce 141 2.74E3 1.67E3 1.96E2 5.14E2 8.62E5 Ce 144 1.79E5 7.32E4 1.00E4 2.96E4 1.03E7 Nd 147 5.32E! 5.47E 1 3.35E0 2.11El 3.46E4 mrem/yr per pCil/m3 .

IJnit 2 Revision 35 II 50 June 2012

TABLE D 3-14 DOSE AND DOSE RATE Ri VALUES - GOAT MILK - CHILD m2m 2rrem/lyr uCi/sec NUCLIDE - CHILD BONE LIVER T. BODY THYROID 3-14RATE KIDNEY MILK LUNG GI-LLI AND GOATD DOSE

- 4.17E3 TABLE DOSE uCi/sec W3F 4.17E3 VALUES R;

4.17E3 m_2-mrem/yr 4.17E3 4.17E3 4-17E3 C 14* 1.65E6 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 Cr 51 6.34E3 3.52E3 9.62E2 6.43E3 3.36E5 Mn 54 1.62E6 4.3 1E5 4.54E5 1.36E6 Fe 55 9.06E5 4.8 1E5 1.49E5 2.72E5 8.91E4 Fe 59 8.52E5 1.38E6 6.86E5 3.99E5 1.43E6 Co 58 8.35E5 2.56E6 4.87E6 Co 60 3.47E6 1.02E7 1.92E7 Zn 65 3.15E8 8.40E8 5.23E8 5.29E8 1.48E8 Sr 89 7.77E9 2.22E8 3.01E8 Sr 90 1.58E I1I 4.01E10 2.13E9 Zr 95 2.62E2 5.76E1 5.13EI 8.25E1 6.01E4 Nb 95 5.05E4 1.96E4 1.40E4 1.85E4 3.63E7 Mo 99 4.95E6 1.22E6 1.06E7 4.09E6 Ru 103 2.75E2 1.06E2 6.93E2 7.12E3 I Ag 1Orm 1.60E7 1.08E7 8.60E6 2.00E7 1.28E9 1 131 3.91EF 3.94E8 2.24E8 1.30E 1I 6.46E8 3.50E7 1133 4.84E6 5.99E6 2.27E6 1.11E9 9.98E6 2-41 E6 Cs 134 4.49E 10 7.37E10 1.55E10 2.28E10 8. 19E9 3.97E8 Cs 137 6.52E10 6.24E 10 9.21 E9 2.03EI0 7.32E9 3.916E8 Ba 140 7.05E6 6.18E3 4. 12E5 2.01E3 3.68E3 3.57E6 La 140 1.16 4.07E-1 1.37E-1 1.13E4 Cc 141 1.38E3 6.88E2 1.02E2 3.02E2 8.59E5 Ce 144 1.25E5 3.916E4 6.66E3 2.16E4 1.02E7 Nd 147 2.68E 1 2.17E1 1.68E0 1.19EI 3.44E4 Imrern/yr per pCi/m3.

I Jnit 2 Revision 35 1 51 June 2012

TABLE D 3-15 DOSE AND DOSE RATIE Rj VALUES - GOAT MILK - TEEN ml-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H33 2.64E3 2.6403 2.64E3 2.64E3 2.64E3 2.64E3 C 14" 6.70E5 1.34E5 1.34E5 1.34E5 1.34E5 1.35E5 1.34E5 Cr 51 3.11 E3 1.73E3 6.82E2 4.44E3 5.23E5 Mn 54 1.08E6 2.15E5 3.23E5 2.22E6 Fe 55 3.61 E5 2.56E5 5.97F4 1.62E5 1.11 E5 Fe 59 3.67E5 8.57E5 3.3 1E5 2.70E5 2.03E6 Co 58 5.46E5 1.26E6 7.53E6 Co 60 2.23 E6 5.03 E6 2.91 E7 Zn 65 1.61ES 5.58E8 2.60E8 3.57E8 2.36E8 Sr 89 3.14E9 8.99E7 3.74E8 Sr 90 9.36E10 2.31E10 2.63E9 Zr 95 1.13 E2 3.56E1 2.45E1 5.23E1 8.22E4 Nb 95 2.23E4 1.24E4 6.82E3 1.20E4 5.30E7 Mo 99 2.72E6 5.19E5 6.23F6 4.87E6 Ru 103 1.16E2 4.98E] 4.103E2 9.72E3 I Ag 110m 7.36E6 6.96E6 4.24E6 1.33E7 1.96E9 I 131 i.61 E8 2.26E8 1.21 E8 6.59E10 3.99E8 4.47E7 1 133 1.99E6 3.38E36 1.03E6 4.72E8 5.93E6 2,56E,6 Cs 134 1.95E 10 4.58E10 2.13E10 1.46E110 5.56E9 5.70E8 Cs 137 2.71E10 3.60E110 1.25E10 1.23E10 4.76E9 5.12E8 Ba 140 2.92E6 3.58E3 1.88E!5 1.21E3 2.41 E3 4.50E6 La 140 4.86E-1 2.39E- I 6.36E-2 1.37E4 Ce 141 5.60E2 3.74E2 4.30E1 1-76E2 1.07E6 Ce 144 5.06E4 2.09E4 2.72E3 1.25E4 1.27E7 Nd 147 1.09E 1 1.19E I 7.13 E-I 6.99E0 4.29E4 mrem/yr per pCi/m3 .

Unit 2 Revision 35

[1 52 June 2012

TABLE D 3-16 DOSE AND DOSE RATE Ri VALUES - GOAT MILK - ADULT mz-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H3 -- 2.03E3 2.03E3 2.03E3 2.03E3 2.03E3 2.03E3 C 14" 3.63E5 7-26E4 7.26.E4 7.26F4 7.26E4 7.26F4 7.26E4 Cr 51 .... 1.78E3 1.06E3 3.92E2 2.36E3 4.48L5 Mn 54 -- 6.50E5 1.24E5 - 1.93E5 -- 1.99E6 Fe 55 2.04E5 1.41 E5 3.28E4 .... 7.85E4 8.07E4 Fe 59 2.10E5 4.95E5 1.90E5 .... 1.38E5 1.65E6 Co 58 -- 3.25E5 7.27E5 ...... 6.58E6 Co 60 -- 1.32E6 2.91E6 ...... 2.48E7 Zn 65 1.05E8 3.33E8 1.51E8 -- 2.23E8 -- 2.10E8 Sr 89 1.70E9 -- 4.89E7 ...... 2.73E8 Sr 90 6.62E 10 -- 1.63E 10 ...... 1.91 E9 Zr 95 6.45E1 2.07E 1 1.40E2 -- 3.25Ei -- 6.56E4 Nb 95 1.3 1E4 7.29E3 3.92E3 -- 7.21E3 -- 4.42E7 Mo 99 -- 1.51 E6 2.87E5 -- 3.41 E6 -- 3.49E6 Ru 103 6.55E 1 -- 2.82E2 -- 2.50E2 -- 7.64E3 Ag I10m 4.45E6 4.12E6 2.45E6 -- 8.09E6 -- 1,68E9 1 131 8.89E7 1.27E8 7.29E7 4.17E10 2.18E8 -- 3.36E7 1 133 1.09E6 1.90E6 5.79E5 2.79E8 3.31 E6 -- 1712E6 Cs 134 1.12E10 2.67E10 2.18E10 -- 8.63E9 2.86E9 4.6728 Cs 137 1.49E10 2.04E10 1.34E10 -- 6.93E9 2.30E9 3.95E8 Ba 140 1.62E6 2.03E3 1.06E5 -- 6.91E2 1.16E3 3.33E6 La 140 2.71E-1 1.36E-1 3.61E-2 ...--. 1.00E4 Cc 141 3.06E2 2.07E2 2.34E1 -- 9.60E] -- 7.90E5 Cc 144 2.75E4 1.15FA 1.48M3 -- 6.92F3 -- 9.30E6 Nd 147 5.69E0 6.57E0 3.93E-1 -- 3.84E0 -- 3.15E4 "mrem/yr per pCi/lm3 .

Unit 2 Revision 35 11 53 June 2012

TABLE D 3-17 DOSE AND DOSE RATE Ri VALUES - COW MEAT - CHILD m 2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BOD Y THYROID KIDNEY LUNG G1-LLI 113: 2.34E2 2.34E2 2.34E-2 2.34132 2.34E2 2.34E2 C 14 5.29E5 1.06E5 1.06E5 1.0 6E5 1.06E5 1.06E5 1.06E5 Cr51 4.55E3 2.52E3 6.90E2 4.6 1E3 2.411E5 Mn 54 5.15 E6 1.37E6 1.44 E6 4.32E6 Fe 55 2.89E8 1.53E8 4.74E7 8.66E7 2.84E7 Fe 59 2.04E8 3.30E8 1.65E8 9.58E7 3.44E8 Co 58 9.41 E6 2.88E7 5.49E7 Co 60 4.641E7 1.37E8 2.57E8 Zn 65 2.38E8 6.35E8 3.95E8 4.00E8 1.12E8 Sr 89 2.65E8 7.57E6 1.03E7 Sr 90 7.011E9 1.78E9 9.44E7 Zr 95 1.5 1E6 3.32E5 2.95E5 4.75E5 3.46E8 Nb 95 4.10E6 1.59E6 1.14E6 1.50E6 2.95E9 Mo 99 5.42E4 1.34134 1.116E5 4.48E4 Ru 103 8.30E7 3.19E7 2.09E8 2.15139 I Ag I 10m 5.62E6 3.79E6 3.03 E6 7.05E6 4.52E8 1131 4.15E6 4.18 E6 2.37E6 1.38E9 6.86E6 3.72E-5 1133 9.38E-2 1.16E-1 4.39E-2 2.15E11 1.93E-1 4.67E-2 Cs 134 6.09E8 1003E9 2.11 E8 3.1l0E8 I.11 E8 5.39E6 Cs 137 8.99E8 8.60E8 1.27E8 2.80E8 1.01E8 5.39E6 Ba 140 2.20E7 1.93E4 1.28E6 6.27E3 1.155E4 1.1 1E7 La 140 2.80E-2 9.78E-3 3.30E-3 2.73E2 Ce 141 1.17E4 5.82E3 8.64E2 2.55E3 7.26E6 Ce 144 1_48F6 4-65E5 7.911F4 2.57F5 1-21ER Nd 147 5193E3 4.80E3 3.72E2 2.64E3 7.611E6 "mremiyrper pCiJm3 .

Unit 2 Revision 35 11 54 June 2012

TABLE D 3-18 DOSE AND DOSE RATE Ri VALUES - COW MEAT - TEEN m -_mrcm/vr uCilsec - TEEN 3-18RATE DMEAT DOSE I

II - COW AND NUCLIDE BONE LIVER T. BODY uCi/sec R* VALUES DOSE TABLE mZ-mrcm/vr THYROID KIDNEY LUNG GI-LL!

H 3 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 C 14" 2,81E5 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 Cr 51 2.93E3 1.62E3 6.39E2 4.16E3 4.90E5 Mn 54 4.50E6 8.93E5 1.34E6 9.24E6 Fe 55 1.50E8 1.07E8 2.49E7 6.77E7 4.62E7 Fe 59 I ,ISF.8 -. 69E8 1.04E8 9.47E7 6.36E8 Co 58 8.05E6 1.86E7 1.11E8 Co 60 3.90E7 8.80E7 5.09E8 Zn 65 1.59E8 5.52E8 2.57E8 3.53Eg 2.34E8 Sr 89 1.40E8 4.0 1E6 1.67E7 Sr 90 5.42E9 1.34E9 1.52E8 Zr 95 8.50E5 2.68E5 1.84E5 3.94E5 6.19E8 Nb 95 2.37E6 1.32E6 7.24E5 1.28E6 5.63E9 Mo 99 3.90E4 7.43E3 8.92E4 6.98E4 Ru 103 4.59F7 1.96E7 1.62E9 3.84E9 I Ag I 10m 3.39E6 3.20E6 1.95E7 6.13E6 9.01 E8 1131 2.24E6 3.13E6 1.68E6 9. 15E8 5.40E6 6.20E5 1133 5.05E-2 8.57E-2 2.6 1E-2 1.20EI 1.50E-1 6,48E-2 Cs 134 3.46E8 8.13E8 3.77E8 2.58E2 9.87E7 1.01E7 Cs 137 4.88E8 6.49E8 2.26E8 2.2 128 9.58E7 9.24E6 Ba 140 1.19E7 1.46E4 7.68E5 4.95E3 9.812E3 1.84E7 La 140 1.53E-2 7.5 1E-3 2.00E-3 4.31E2 Ce 141 6.19E3 4.14E3 4.75E2 1.95E3 1.18E7 Ce 144 7.87E5 3.26E5 4.23E4 1.94E5 1.98E&

Nd 147 3.16E3 3.44E3 2.06E2 2.02E3 1.24E7

'mremnyr per pCi/m3 .

Unit 2 Revision 35 I1 55 June 2012

TABLE D 3-19 DOSE AND DOSE RATE Ri VALUES - COW MEAT - ADULT M2-mrem/yr uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI 11 3 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 C 14" 3.33 E5 6.66E4 6.66E4 6.66E4 6.66E4 6.66E4 6.66E4 Cr 51 3.65E3 2.18E3 8.03E2 4.84E3 9.17E5 Mn 54 5.90E6 1.13E6 1.76E6 1.81E7 Fe 55 1.85E8 1.28E8 2.98E7 7.14E7 7.34E7 Fe 59 1.44E8 3.39E8 13 0FR 9.4627 1.13F9 Co 58 1.04E7 2.34E7 2.12E8 Co 60 5.03E7 1.1 IE8 9.45E8 Zn 65 2.26E8 7.19E8 3.25E& 4.81 Eg 4.53E8 S1 89 1.66E8 4.76E6 2.66E7 Sr 90 8.38E9 2.06E9 2.42E8 Zr 95 1.06E6 3.40E5 2.30E5 5.34E5 1.08E9 Nb 95 3.04E6 1.69E6 9.08E5 1.67E6 1.03E10 Mo 99 4.71 E4 8.97E3 1.07E5 1.09E5 Ru 103 Ag I lOre 5.64E7 2.43F7 2.15F9 6.58.F9 I 4.48E6 4.14E6 2.46E6 8. 13E6 1.69E9 1131 2.69E6 3.85E6 2.21 E6 1.26E9 6.6 1E6 1.02E6 1133 6.04E-2 1.05E-1 3.20E-2 1.54E1 1.83E-I 9-44E-2 Cs 134 4.35E8 1.03E9 8.45E8 3.35E8 1.1 IE8 1,ý81E7 Cs 137 5.88E8 8.04E8 5.26E8 2.73Eg 9.07E7 1.56E7 Ba 140 1.4E7 1.81E4 9.44E5 6.15E3 1.04E4 2.97E7 La 140 1.86E-2 9.37E-3 2.48E-3 6.88E2 Ce 141 7.38E3 4.99E3 5.66E2 2.32E3 1.91E7 Ce 144 9.33 E5 3-90E5 5.01 E4 2.31E5 3.16E2 Nd 147 3.59E3 4.15E3 2.48E2 2.42E3 1.99E7 rmrem/yr per pCi/m 3 .

Unit 2 Revision 35 1 56 June 2012

TABLE D 3-20 DOSE AND DOSE RATE R, VALUES - VEGETATION - CHILD uCi/seic

- CHILD D 3-20RATE DOSE i i i i VEGETATION

- AND NUCLIDE BONE LIVER T.DOSE Ri VALUES BODY uCi/sec TABLE THYROID mZ-mremlyr KIDNEY LUNG GI-LLI H-3' 4.01E3 4.01 E3 4.01 E3 4.01 E3 4.01 E3 4.01E3 C 14' 3.50E6 7.01 E5 7.01E5 7.01 E5 7.01E5 7.01 E5 7.01E5 Cr51 1.17E5 6.49E4 1-77E4 1.18E5 6.20E6 Mn 54 6.65E8 1.77E8 1-86E8 5.58E8 Fe 55 7.63E8 4.05E8 1.25E8 2.29E8 7.50E7 Fe 59 3.97E8 6.42E8 3.20E8 6.69E8 Co 58 6.45E7 1.97E8 3.76E8 Co 60 3.78E8 1.12E9 2.10E9 Zn 65 8.12E8 2.16E9 1.35E9 1.36E9 3.80E8 Sr 89 3.59E 10 1.03E9 1.39E9 Sr 90 1.24E 12 3.15EII 1.67EI0 Zr 95 3.96E6 8.50E5 7.56E5 1.22E6 8.86E8 Nb 95 1.02E6 3.99E5 2.85E5 3.75E5 7.37E8 Mo 99 7.70E6 1.91 E6 1.65E7 6.37E6 Ru 103 1.53E7 5.90E6 3.R6F7 3.97F8 I Ag I r0m 3.21E7 2.17E7 1.73E7 4.04E7 2.58E9 1 131 7.16E7 7.20E7 4.09E7 2.38EIO 1.18E8 6.41 E6 1133 1,69E6 2.09E6 7.92E5 .3.89E8 3.49E6 8.44E5 Cs 134 1.60E 10 2.63E10 5.5 5E9 8.15E9 2.93E9 1.42E8 Cs 137 2.39E10 2.29E10 3.38E9 7.46E9 2,68E9 1.43E8 Ba 140 2.77E8 2.43E5 1.62E7 7.90E4 1.45E5 1.40E8 La 140 3.25E3 1.13E3 3.83E2 3.16E7 Ce 141 6.56E5 3.27E5 4.85E4 1.43E5 4.08E8 Cc 144 1.27E8 3.98E7 6.78E6 2.2117 1.04E 10 Nd 147 7.23E4 5.86E4 4.54E3 3.22E4 9.28E7 rmrern/yr per iCilms.

Unit 2 Revision 35 1157 June 2012

TABLE D 3-21 DOSE AND DOSE RATE Ri VALUES - VEGETATION - TEEN In 2_mrem/vr uCi/sec il DOSE - TEEN D 3-21RATE VEGETATION

-AND NUCLIDE BONE LIVER T.DOSE Ri VALUESBODY uCi/sec TABLE THYROID mZ-mrcm/vr KIDNEY LUNG GI-LLI H- 3 2.5903 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3 C 14" 1.45E6 2.911E5 2.911E5 2.91 E5 2.91E5 2.91 E5 2.911E5 Cr 51 6.16E4 3.42E4 1.35E4 8.79E,4 1.03E7 Mn 54 4.54E8 9.013E7 1.36E8 9.32E8 Fe 55 3.10E8 2.20E8 5.13E7 1.40E8 9.53E7 Fe 59 1.79E8 4.18E8 1.61E8 1.32E8 9.89F8 Co 58 4.37E7 1.01 E8 6.02 E8 Co 60 2.49E8 5.60E8 3.24E9 Zn 65 4.24E8 1.47E9 6.86E8 9.4 1E8 6.23E8 Sr 89 1.51E10 4.33E8 1.80E9 Sr 90 7.51E11 1.85E1 II 2.11E 10 Zr 95 1.72E6 *5.44E5 3.74E5 7.99E5 1.26E9 Nb 95 4.80E5 2.66E5 1.46E5 2.58E5 1.14E9 Mo 99 5.64E6 1.08E6 1.29E7 1.0 11E7 Ru 103 6.82E6 2.92E6 2.40E7 5.70F8 I Ag 1O1m 1.511E7 1.43E7 8.72E6 2.74E7 4.03E9 1.131 3.85E7 5.39E7 2.89E7 1.57E10 9.28E7 1.07E7 1133 9.29E5 1.58E6 4.80E5 2.20E8 2.76E6 1.193E6 Cs 134 7.10E9 1.67E10 7.75E9 5-311E9 2.03E9 2.08E8 Cs 137 1.01E10 1.35E10 4.69E9 4.59E9 1.78E9 1.92E8 Ba 140 1.38E8 1.69E5 8.9 1E6 5.74E-4 1.14E5 2.13E8 La 140 1.81E3 8.88E2 2.36E2 5.103E7 Ce 141 2.83E5 1.89E5 2.17E4 8.89E4 5.40E8 Ce 144 5.27E7 2.1 RF.7 2.-83F6 1.30E7 1.33E10 Nd 147 3.66E4 3.98E4 2.38E3 2.34E4 1.44E8

.mremlyr per pCi/m 3 Unit 2 Revision 35 11 58 June 2012

TABLE D 3-22 DOSE AND DOSE RAT.E

.R, VALUES - VEGETATION - ADULT m 2-mrem/vr uCi/sec

- ADULT NUCLIDE BONE LIVER T. BODY RATE THYROID 3-22 KIDNEY LUNG GI-LLI i

AND D DOSE

-VEGETATION DOSE TABLE uCi/sec 113* 2.26E3 2.26E3 VALUES m2-mrem/vr

.Ri 2.26E3 2.26E3 2.26E3 2.26E3 C 14*

8.97.E5 1.79E5 1.79E5 1.79E5 1.79E5 !.79E5 1.79E5 Cr51 4.164E4 2.77FA 1.02E4 6.15E4 1.17E7 Mn 54 3.13E8 5.97E7 9.3 1E7 9.58E8 Fe 55 2.00E8 1.38E8 3.22E7 7.69E7 7.91E7 Fe 59 1.26E8 2.96E8 1.13Eg 8.27E7 1.02E9 Co 58 3.08E7 6.90E7 6.24E8 Co 60 1.67E8 3.69E8 3.14E9 Zn 65 3.17E8 1.01E9 4.56E8 6.75E8 6.36E8 Sr 89 9.96E9 2.6E8 1.60E9 Sr 90 6.05E II 1.48E 11 1.75E10 Zr 95 1.1816 3.77E5 2.55E5 5.92E5 1.20E9 Nb 95 3.55E5 1.98E5 1.06E5 1.95E5 1.20E9 Mo 99 6.14E6 1.17E6 1.39E7 i .42E7 Ru 103 4.77E6 2.06F6 1.82F7 5.57.F8 I Ag I lore 1.05E7 9.75E6 5.79E6 1.92E7 3.98E9 1131 4.041E7 5.78E7 3.3 1E7 1.90E10 9.91E7 1.53E7 1133 1.00E6 1.74E6 5.30E5 2.56E8 3.03E6 1.56E6 Cs 134 4.67E9 1.11E10 9.08E9 3.59E9 1.19E9 1-94E8 Cs 137 6.36E9 8.70E9 5.70E9 2.95E9 9.812E8 1.68E8 Ba 140 1.29E8 1.61 E5 8.42E6 5.49E4 9.25E4 2.65E8 La 140 1.98E3 9.97E2 2.63E2 7.32E7 Ce 141 1.97E5 1.33E5 1.5 1E4 6.19E4 5.09E8 Ce 144 3.29F7 1.38E7 1.77E6 8. 16E6 1.11E10 Nd 147 3.36E4 3.88E4 2.32E3 2.27E4 1.86E8 mrem/yr per pCi/m3

. Unit 2 Revision 35 11 59 June 2012

TABLE D 3-23 DISPERSION PARAMETERS AT CONTROLLING LOCATIONS' X/Q,W, and W, VALUES VENT DIRECTION X/O (see/M3) D/o (M-2 Site Boundary' E 1,600 2.00 E-6 2.10E-9 Inhalation and Ground E (104-) 1,800 1.42E-7 2.90E-9 Plane Cow Milk ESE (130-) 4,300 4.11 E-8 4 .7313- 10 3

Goat Milk SE (140') 4,800 3.56E-08 5.32E-10 Meat Animal E (114) 2,600 1.17E-7 1 .86E-9 Vegetation E (96) 2,900 1.04E-7 1.50E-9 STACK Site 2 Boundary E 1,600 4.50E-8 6.OOE-9 Inhalation and Ground E (1090) 1,700 9.48E-9 1.34E-9 Plane Cow Milk ESE (135) 4,200 1.05E-8 3.64E- 10 Goat Milk3 SE (140*) 4,800 2.90E-08 5.71E-10 Meat Animal E (114) 2,500 1.13E-8 1.15E-9 Vegetation E (96) 2,800 1.38E-8 9.42E-10 NOTE: Inhalation and Ground Plane are annual average values. Others are grazing season only.

X/Q and D/Q values from NMP-2 ER-OLS.

2 X/Q and D/Q from NMP-2 FES, NUREG-1085, May 1985, Table D-2.

3 X/Q and D/Q from C.T. Main Data Report dated November 1985.

Unit 2 Revision 35 11 60 June 2012

TABLE D 3-24 PARAMETERS FOR THE EVALUATION OF DOSES TO REAL MEMBERS OF THE PUBLIC FROM GASEOUS AND LIQUID EFFLUENTS Pathwal Parameter Value Reference Fish U (kg/yr) - adult 21 Reg. Guide 1.109 Table E-5 Fish D~pj (mrem/pCi) Each Radionuclide Reg. Guide 1.109 Table E-I I Shoreline U (hr/yr)

- adult 67 Reg. Guide 1.109

- teen 67 Assumed to be Same as Adult Shoreline Drmhpj Each Radionuclide Reg. Guide 1.109 pCUr 2 (mrem/hir perpi/) ) Table E-6 Inhalation DFAija Each Radionuclide Reg. Guide 1.109 Table E-7 Unit 2 Revision 35 1161 June 2012

TABLE D 5.1 NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collection Site Tv'l orfSaimole Location (Env. Program No, Location Radioiodine and 1 Nine Mile Point Road North 1.8 mi @7,92*F Particulates (air) (R-1)

Radioiodine and 2 County Route 29 & Lake Road Particulates (air) (R-2) 1.1mitq 1346ESr Radioiodine and 3 County Route 29 Particulates (air) (R-3)

Radiuiodiuc and 4 Village of Lycoming, NY 1.8 mi Ca145' SE Particulates (air) (R-4)

Radioiodine and 5 Montario Point Road 16.2 mi@ 42'NE Particulateis (air) (R-5)

Direct Radiation (TLD) 6 North Shoreline Area 0.1 mi@ 354'N

{75)

Direct Radiation (TLD) 7 North Shoreline Area 0.1rni C@27' NINE (76)

Direct Radiation (TLD) 8 North Shoreline Area 0.1- iiC@)37' NE

<77)

Direct Radiation (I .1)) 9 North Shoreline Area (23)

Direct Radiation (TLD) 10 JAF East Boundary *.0 mi @ 86' E (78)

Direct Radiation (TLD) 11 Route 29 (79)

Direct Radiation (TLD) 12 Route 29 .5 i Ca:3136- SE (80)

Direct Radiation (FLD) 13 Miner Road .7 mi r~ 160* SSE (81)

Direct Radiation (TLD) 14 Miner Road .6tni@180.

(82)

Direct Radiation (TLD) 15 Lakeview Road .2 mi qý,,203- SW (83)

Direct Radiation (TLD) 16 Lakeview Road .1 mi nd225'SW (84)

Direct Radiation (TLD) 17 Site Meteorological Tower 0'.7 mi (-- 244' WSW (7)

Direct Radiation (TLD) 18 Energy Information Center 0 .5 mi Cy266" W (18)

Direct Radiation (TLD) 19 North Shoreline 0'.2 tni @ 290* WNW (85)

Map = SeeFigures .D 5.1-1 and D 5.1-2.

Unit 2 Revision 35 11 62 June 2012

TABLE D 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING I OCATIONS I

Map Collection Site Tv¥1 of Sample Location 'Env. Program No.) Iocatlan Direct Radiation (TLD) 20 North Shoreline 0.1 mi. 310'NW (86)

Direct Radiation (TLD) 21 North Shoreline 0.1 mi @332'NNW (87)

Direct Radiation (TLD) 22 Hickory Grove 4.5 mi (O97"1 E (88)

Dijmti Radiation (TLD) 23 Leavilt Road 4 3 mi @ 112 ESE (89)

Direct Radiation (TLD) 24 Route 104 4.2 mi @ 135* SE (90)

Direct Radiation (TLD) 25 Route 5]A 4.9 mi @4157' SSE (91)

Direct Radiation (TLD) 26 Maiden Lane Road 4.5 mi@ 183'S (92)

Direct Radiation (TLD) 27 County Route 53 4.4 mi @ 206' SSW (93)

Direct Radiatinn (TI .D) 28 County Route I 44 mi @ 224' SW (94)

Direct Radiation (TLD) 29 Lake Shoreline 3.7 mi @ 239' WSW (95)

Direct Radiation (TLD) 30 Phoenix, NY Control 19,7 mi @ 168' SSE (49)

Direct Radiation (TLD) 31 S, W- Oswego, Control 12.5 mi @ 227'SW (14)

Direct Radiation (TLD) 32 Scriba, NY 3.7 mi r@ 199" SSW (96)

Direct Radiation (TLD) 33 Novelis, Route IA 3 ,.0 mi @ 222' SW I (58)

Direct Radiation (TLD) 1.-8mi (a, 145' SE 34 Lycoming, NY (97)

Direct Radiation (TLD) 35 New Haven, NY 5.2 mi @ 124' SE (56)

Direct Radiation (TLD) 36 W. Boundary. Bible Camp 0.9 mi @ 239' WSW (15)

Direct Radiation (TLD) 37 Lake Road .2 mi@ 103- ESE (98)

Map See Figures D5. 1-1 and D.5.1 Unit 2 Revision 35 II 63 June 2012

TABLE D 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS Map Collection Site Tvyp of Sample LoUalion (Env. Prolram No.) LUcathon Surface Water 38 OSS Inlet Canal 7.6 mi @ 236" SW (NA)

Surface Water 39 JAFNPP Inlet Canal 0.5 mi @, 71" ENE (NA)

Shoreline Sediment 40 Sunset Bay Shorcline 1.2 mi @ 84" E (NA)

Fish 41 NMP Site Discharge Area 0.3 mi g315'NW

.(NA)

(and/or)

Fish 42 NMP Sile Discharge Area 0.6 mi to 5_' NE (NA)

Fish 43 Oswego Harbor Area 5.9 mi @ 237* WSW (NA)

Milk 64 Milk Location 455 8.8 ti @ 97' E Milk (CR) 77 Milk Location 16.0 mi @ 190 S (Summerville)

Food Product 48 Produce Location #6** 1.9 mi@ 143 SE (Hergenstock) (NA)

Food Product 49 Produce Location #I .. 1.6 mi @.84' E (Culeton) (NA)

Food Product 50 Produce Location #24* 1.9 mi@ 101' E (Vitullo) (NA)

Food Product 51 Produce Location 05* 1.5 mi 0a I6"ESE (C.S. Parkhurst) (NA)

Food Product 52 Produce Location #3** 1.5 mi @q4" E (C. Narewski) (NA)

Map See Figures 1 51-1 and D 5.1-2.

Food Product Samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculaled site average DiQ.

(NA) Not applicable.

CR Control Result (location).

Unit 2 Revision 35 1164 June 2012

TABLE D 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS Map Collection Site Type of Sample Location (Env. Program No.) Location Food Product 53 Produce Location #4** 1.7 mi @ 126"SE (P. Parkhurst) (NA)

Food Product (CR) 54 Produce Location #7** 15.1 mi @ 222' SW (Mc Millen) (NA)

Food Product (CR) 55 Produce Location #8"* 12.5 mi @ 227* SW (Denman) (NA)

Food Product 56 Produce Location #9** 1.6 mi@ 717]S (O'Connor) (NA)

Food Product 57 Produce Location # 10* 2.3 mi @ 124' SE (C. Lawton) (NA)

Food Product 58 Produce Location if1 1* 2.Oini @ 112' ESE (C. R. Parkhurst) (NA)

Food Product 59 Produce Location # i2* 2.0 mi @ 110 ESE (Barton) (NA)

Food Product (CR) 60 Produce Location #113** 15.4 mi @ 222 SW (Flack) (NA)

Food Product 61 Produce Location #14"* 1.9 mi @ 97* E (Koeneke) (NA)

Food Product 62 Produce Location # l5** 1.6 mi @ 139'SE (Whaley) (NA)

Food Product 63 Produce Location # 16* 1.2 mi @ 209 SSW (Murray) (NA)

Food Product 67 Produce Location #17** 1.7 mi @ 98' E (Battles) (NA)

Food Product 68 Product Location # 1l8* 1.5 mi @ 840 E (Kronenbitter)

Food Product 69 Product Location #11 9** 1.4 mi @ 132* SE (O'Connor)

Map = See Figures D 5.1- I and D 5.1-2.

= Food Product Samples need notnecessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q.

(NA) - Not applicable.

CR = Control Result (location).

Unit 2 Revision 35 1165 June 2012

APPENDIX A LIQUID DOSE FACTOR DERIVATION Unit 2 Revision 35 II 66 June 2012

Appendix A Liquid Effluent Dose Factor Derivation, Aiat Ai,, (mremihr per uCi/ml) which embodies the dose conversion factors, pathway transfer factors (e.g., bioaccumulation factors), pathway usage factors, and dilution factors for the points of pathway origin takes into account the dose from ingestion of fish and drinking water and the sediment. The total body and organ dose conversion factors for each radionuclide will be used from Table E-Il of Regulatory Guide 1. 109. To expedite time, the dose is calculated for a maximum individual instead of each age group. The maximum individual dose factor is a composite of the highest dose factor Aia, of each nuclide i age group a, and organ t, hence Ai,,. It should be noted that the fish ingestion pathway is the most significant pathway for dose from liquid effluents. The water consumption pathway is included for consistency with NUREG 0133.

The equation for calculating dose contributions given in section 1.3 requires the use of the composite dose factor Ait for each nuclide, i. The dose factor equation for a fresh water site is:

i. [A'

0. eK +U B~e-tfD 4 69.3U, P ek"1. (1- CA b DFSl D, f Ia t+ A1 Where:

Aiat Is the dose factor for nuclide i, age group a, total body or organ t, for all appropriate pathways, (mrem/hr per uCi/ml)

K0 Is the unit conversion factor, 1.14E5=IE6pCi/uCi x 1E3 ml/liter 8760 hr/yr U= Water consumption (liters/yr); from Table E-5 of Reg. Guide 1.109 U- = Fish consumption (kg/yr); from Table E-5 of Reg. Guide 1.109 U, = Sediment Shoreline Usage (hr/yr); from Table E-5 of Reg. Guide 1.109 BFi Bioaccumulation factor for nuclide, i, in fish, (pCi/kg per pCi/liter),

from Table A-1 of Reg. Guide 1.109 DFLiat = Dose conversion factor for age, nuclide, i, group a, total body or organ t, (mrem/pCi); from Table E-1I1 of Reg. Guide 1.109 DFSi = Dose conversion factor for nuclide i and total body, from standing on contaminated ground (mrem/hr per pCi/m ); from Table E-6 of Reg.

Guide 1.109 Unit 2 Revision 35 1167 June 2012

Appendix A (Cont'd)

D, Dilution factor from the near field area within one-quarter milc of the release point to the potable water intake for the adult water consumption. This is the Metropolitan Water Board, Onondaga County intake structure located west of the City of Oswego. (Unitless)

D, Dilution factor from the near field area within one quarter mile of the release point to the shoreline deposit (taken at the same point where we take environmental samples 1.5 miles; unitless) 69.3 conversion factor -693 x 100, 100 = K, (litersikg-hr)*40 kg/m 2*24 hr/day/.693 in liters/m 2 -d, and K, = transfer coefficient from water to sediment in liters/kg per hour.

tP%, tPf, Average transit time required for each nuclide to reach the tPS point of exposure for internal dose, it is the totaltime elapsed from release of the nuclides to either ingestion for water (w) and fish (f) or shoreline deposit (s), (hr) tb Length of time the sediment is exposed to the contaminated water, nominally 15 yrs (approximate midpoint of facility operating life),

(hrs).

i= decay constant for nuelide i (hr")

W= Shore width factor (unitless) from Table A-2 of Reg. Guide 1.109 Example Calculation For 1-131 Thyroid Dose Factor for an Adult from a Radwaste liquid effluents release:

2 (DFS)i = 2.80E-9 mrem/hr per pCi/mr (DFL)iat = 1.95E-3 mrem/pCi tPW = 40 hrs. (w = water)

BFi = 15 pCi/kg per pCi/liter tPr = 24 hrs. (f = fish)

Uf = 21 kg/yr tb = 1.314E5 hr (5.48E3 days)

Dw = 62 unitless U% = 730 liters/yr D, = 17.8 unitless K, = 1.14E5 (pCi/uCi)(mlikg)

U, = 12 hr/yr (hr/yr)

W = 0.3 )Li = 3.61E-3hr' tPS = 7.3 hrs (s=Shoreline Sediment)

These values will yield an Aiat Factor of 6.65E4 mrem-ml per uCi-hr as listed in Table D 2-2. It should be noted that only a limited number of nuclides are listed on Tables D 2-2 to D 2-5. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in a revision to the ODCM.

In addition, not all dose factors are used for the dose calculations. A maximum individual is used, which is a composite of the maximum dose factor of each age group for each organ as reflected in the applicable chemistry procedures.

Unit 2 Revision 35 11 68 June 2012

APPENDIX B PLUME SHINE DOSE FACTOR DERIVATION Unit 2 Revision 35 II 69 June 2012

Appendix B For elevated releases the plume shine dose factors for gamma air (Bi) and whole body (Vi), are calculated using the finite plume model with an elevation above ground equal to the stack height.

To calculate the plume shine factor for gamma whole body doses, the gamma air dose factor is adjusted for the attenuation of tissue, and the ratio of mass absorption coefficicnts between tissue and air. The equations are as follows:

Gamma Air Bi = K aE I, Where: K' conversion tactor (see s RO V, below for actual value).

pa mass absorption coefficient (cm:/g; air for Bi, tissue for Vi)

E = Energy of gamma ray per disintegration (Mev)

VS average wind speed for each stability class (s), ni/s R downwind distance (site boundary, m) 0 = sector width (radians) s subscript for stability class Is I function = I, + k12 for each stability class. (unitless, see Regulatory Guide 1.109) k* Fraction of the attenuated energy that is actually absorbed in air (see Regulatory Guide 1.109, see below for equation)

Whole Body

- [Lat.d Vi = 1. 11SFBie Where: td tissue depth (g/cm2)

SF = shielding factor from structures (unitless) 1.11 = Ratio of mass absorption coefficients between tissue and air.

Where all other parameters are defined above.

Unit 2 Revision 35 1170 June 2012

Appendix B (Cont'd) 1K = conversion factor 3.7ElOdis 1.6E-6crg Ci-sec Mev .46 12 93 g 100 erg m3 g-rad 2k = *Al-_a.

Where: t= mass attenuation coefficient (cm 2/g; air for Bi, tissue for V1 )

Ila = defined above There are seven stability classes, A thru G. The percentage of the year that each stability class is taken from the U-2 FSAR. From this data, a plume shine dose factor is calculated for each stability class and each nuclide, multiplied by its respective fraction and then summed.

The wind speeds corresponding to each stability class are, also, taken from the Unit 2 FSAR. To confirm the accuracy of these values, an average of the 12 month wind speeds for 1985, 1986, 1987 and 1988 was compared to the average of the FSAR values. The average wind speed of the actual data is equal to 6.78 m/s, which compared favorably to the FSAR average wind speed equal to 6.77 m/s.

The average gamma energies were calculated using a weighted average of all gamma energies emitted from the nuclide. These energies were taken from the handbook "Radioactive Decay Data Tables", David C. Kocher.

The mass absorption (Ri,) and attenuation (p) coefficients were calculated by multiplying the mass absorption (p-t/p) and mass attenuation p[t/p) coefficients given in the Radiation Health Handbook by the air density equal to 1.293 E-3 g/cc or the tissue density of I glcc whcre applicable. The tissue depth is 5g/cm 2 for the whole body.

The downwind distance is the site boundary.

SAMPLE CALCULATION Ex. Kr-89 F STABILITY CLASS ONLY - Gamma Air

-DATA E 2.22MeV k = = .872 K .46

=a 2.943 E-3m- Ia V= 5.55 m/Sec p 5.5064E-3m"' R = 1600m S= 39

= 1 9m vertical plume spread taken from "Introduction to Nuclear Engineering", John R. LaMarsh Unit 2 Revision 35 11 71 June 2012

Appendix B (Cont'd)

-1 Function uoz = .11 I1 .3 12 =.4 I + kI 2 - .3-+ (.871) (.4) =.65 dis.

B = 0.46 ICi-sec) (Mev/ergsl (2.943E-3m-') (2.22Mev) (.65) 2 (iTI/ (W/mr ) (ergs) (5.55 m/s) (.39) (1600m)

(g - rad) 3.18(-7) rad/s (3600 s/hr) (24 h/d) (365 d/v) (lE3rmrad/rad)

Ci/s (1IE6uCi)

Ci

= 1.00(-2) mrad/yr uCi/sec (.0253 cm3 /g) (bg/cm2)

Vi = 1.11 (.7) L(E-2)mrad/yrl [e I 6.85(-3) mrad/yr ILCi /.sec Note: The above calculation is for the F stability class only- For Table D 3-2 and procedure values, a weighted fraction of each stability class was used to determine the Bi and Vi values.

Unit 2 Revision 35 II 72 June 2012

APPENDIX C DOSE PARAMETERS FOR IODINE 131 and 133, PARTICULATES AND TRITIUM Unit 2 Revision 35 II 73 June 2012

Appendix C DOSE PARAMETERS FOR IODINE - 131 AND - 133, PARTICULATES AND TRITIUM This appendix contains the methodology which was used to calculate the organ dose factors for 1-131, 1-133, particulates, and tritium. The dose factor, Ri, was calculated using the methodology outlined in NUREG-0133. The radioiodine and particulate DLCO 3.2.1 is applicable to the location in the unrestricted area where the combination of existing pathways and receptor age groups indicates the maximum potential exposure occurs, i.e., the critical receptor. Washout was calculated and determined to be negligible. Ri values have been calculated for the adult, teen, child and infant age groups for all pathways. However, for dose compliance calculations, a maximum individual is assumed that is a composite of highest dose factor of each age group for each organ and pathway. The methodology used to calculate these values follows:

C. 1 Inhalation Pathway Ri(I) -K'(BR)a,(DFA)ija, where:

Ri(I) - dose factor for each identified radionuclide i of the organ of interest (units = mrem/yr per uCi/m 3);

K? = a constant of unit conversion, I E6 pCi/gCi (BR)a = Breathing rate of the receptor of age group a, (units = m 3/yr);

(DFA)ija The inhalation dose factor for nuclide i, organ j and age group a, and organ j (units = mrem!pCi).

The breathing rates (BR)a fbr the various age groups, as given in Table E-5 of Regulatory Guide 1.109 Revision ], are tabulated below.

Age Group (a) Breathing Rate (m3/vr)

Infant 1400 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DFA)ija for the various age groups are given in Tables E-7 through E-I 0 of Regulatory Guide 1.109 Revision I.

Unit 2 Revision 35 1174 June 2012

Appendix C (Cont'd)

C.2 Ground Plane Pathway R (G) = K'K' ' (SF) (DFG) i (1-e Where:

Ri(G) Dose factor for the ground plane pathway for each identified 2

radionuclide i for the organ of interest (units =mM -rnrem/yr per uCi/sec)

K' = A constant of unit conversion, 1E6 pCi/uCi K" A constant of unit conversion, 8760 hr/year

- The radiological decay constant for radionuclidc i, (units = scc-1) t The exposure time, sec, 4.73E8 sec (15 years)

(DFG)i = The ground plane dose conversion factor for radionuclide i; (units =

mrermIr per pCi/mr2 )

SF = The shielding factor (dimensionless)

A shielding factor of 0.7 is discussed in Table E-15 of Regulatory Guide 1.109 Revision 1.

A tabulation of DFGi values is presented in Table E-6 of Regulatory Guide 1.109 Revision 1.

Unit 2 Revision 35 I175 June 2012

Appendix C (Cont'd)

C.3 Grass-(Cow or Goat)-Milk Pathway K' U F "+/-

R.(C) UQ apFtm(r)(DFL),.. ff, e(1-(;,J++ ) e" Where:

RX(C) Dose factor for the cow milk or goat milk pathway, for each identified radionuclide i for the organ of interest, (units = m 2 -mrem/yr per uCi/sec)

Kt = A constant of unit conversion, 1E6 pCiipCi Q= The cow's or goat's feed consumption rate, (units = kg/day-wet weight)

Uap *Thereceptor's milk consumption rate for age group a, (units = liters/yr) 2 Yp The agricultural productivity by unit area of pasture fccd grass, (units = kg/mn )

YS= The agricultural productivity by unit area of stored feed, (units = kg/mr)

FM = The stable element transfer coefficients, (units = pCi/liter per pCi/day) r Fraction of deposited activity retained on cow's feed grass (DFL)iaj The ingestion dose factor for nuclide i, age group a, and total body or organ j (units = mrem/pC.i)

- The radiological decay constant for radionuclide i, (units=sec)

-W The decay constant for removal of activity on leaf and plant surfaces by weathering equal to 5.73E-7 sec' (corresponding to a 14 day half-life) t= The transport time from pasture to cow or goat, to milk, to receptor, (units = sec) ti, The transport time from pasture, to harvest, to cow or goat, to milk, to receptor (units

= sec) fp - Fraction of the year that the cow or goat is on pasture (dimensionless)

S= Fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless)

Unit 2 Revision 35 1176 June 2012

Appendix C (Cont'd)

Milk cattle and goats are considered to be fed from two potential sources, pasture grass and stored feeds. Following the development in Regulatory Guide 1. 109 Revision 1, the value of f, is considered unity in lieu of site specific information. The value of fp is 0.5 based on 6 month grazing period. This value for fp was obtained from the environmental group.

Table C-I contains the appropriate values and their source in Regulatory Guide 1. 109 Revision 1.

The concentration of tritium in milk is based on the airborne concentration rather than the deposition. Therefore, the RT(C) is based on X/Q:

R-1 (C) = K'K'" FmQrQUap(DFL)iat 0.75(0.5/H)

Where:

R.1.(C) = Dose factor for the cow or goat milk pathway for tritium for the organ of interest, (units = mrem/yr per pCi/mn 3)

= A constant of unit conversion, I E3 g/kg 3

H - Absolute humidity of the atmosphere, (units =g/hi )

0.75 - The fraction of total feed that is water 0.5 - The ratio of the specific activity of the feed grass water to the atmospheric water Other values are given previously. A site specific value of H. equal to 6.14 g/m 3 is used.

This value was obtained from the environmental group using actual site data.

Unit 2 Revision 35 1177 June 2012

Appendix C (Cont'd)

CA Grass-Cow-Meat Pathway R,= () (A,,+'p + (1- *ff,)e -

0 , I.

[_

[LLf, K'QrUpFf(r)(DFL)i Ri(M) = Dose factor for the meat ingestion pathway for radi.onuclide i for any organ of interest, (units = m 2-mrem/yr per ltCi/sec)

Ff = The stable element transfer coefficients, (units = pCi/kg per pCi/day)

U,, = The receptor's meat consumption rate for age group a, (units - kg/year) th = The transport time from harvest, to cow, to receptor, (units = sec) tr= The transport time from pasture, to cow, to receptor, (units = see)

All other terms remain the same as defined for the milk.pathway. Table C-2 contains the values which were used in calculating Ri(M).

The concentration of tritium in meat is based on airborne concentration rather than deposition.

Therefore, the RT(M) is based on X/Q.

RT(M) = K'K"'.FfQfUap(DFL)iat [0.75(0.5/H)]

Where:

R-r(M) = Dose factor for the meat ingestion pathway for tritium for any organ of interest, (units = mrem/yr per pCi/m 3)

All other terms are defined above.

C.5 Vegetation Pathway The integrated concentration in vegetation consumed by man follows the expression developed for milk. Man is considered to consume two types of vegetation (fresh and stored) that differ only in the time period between harvest and consumption, therefore:

- xit -;i* th Ri(V) K' Y.(oLi r

+ Xw)

(DFL) aj I UL'*Fze + osaFge ]

Unit 2 Revision 35 II 78 June 2012

Appendix C (Cont'd)

Where:

Ri(V) Dose factor for vcgctablc pathway for radionuclide i for the organ of interest, (units = m2-mrem/yr per pCi/sec)

K' = A constant of unit conversion, 1E6 pCi/LCi U La = The consumption rate of fresh leafy vegetation by the receptor in age group a, (units - kg/yr) uSa = The consumption rate of stored vegetation by the receptor in age group a (units = kg/yr)

FL = The fraction of the annual intake of fresh leafy vegetation grown locally Fý The fraction of the annual intake of stored vegetation grown locally tL Thie average time between harvest of leafy vegetation and its consumption, (units =sec) th The average time between harvest of stored vegetation and its consumption, (units = see)

Y, The vegetation areal P density, (units = kg/mr)

All other factors have been defined previously.

Table C-3 presents the appropriate parameter values and their source in Regulatory Guide 1.109 Revision 1.

In lieu of site-specific data, values for FL and F. of, 1.0 and 0.76, respectively, were used in the calculation. These values were obtained from Table E- 15 of Regulatory Guide 1.109 Revision 1.

The concentration of tritium in vegetation is based on the airborne concentration rather than the deposition. Therefore, the Rr(V) is based on X/Q:

R-1 (V) = K'K." [UIa fL + Uqa fg](DFL)iaj 0.75(0.5/H)

Where:

RT(V) dose factor for the vegetable pathway for tritium for any organ of interest, (units = mrem/yr per PCi/m 3).

All other terms are defined in preceeding sections.

Unit 2 Revision 35 1.179 June 2012

TABLE C-I Parameters for Grass - (Cow or Goat) - Milk Pathways Reference Parameter Value (Reg. Guide 1.109 Rev. 1)

Qf (kg/day) 50 (cow) Table E-3 6 (goat) Table E-3 r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-1i5 (DFL)ij, (mrem/pCi) Each radionuclide Tables E- I1to E-14 F, (pCi/liter per pCi/day) Each stable element Table E-1 (cow)

Table E-2 (goat)

Ys (kg/mr2) 2.0 Table E-15 Y. (kg/m 2 ) 0.7 Table E- 15 th (seconds) 7.78 x 106 (90 days) Table E-15 tf (seconds) 1.73 x 10" (2 days) Table E-15 Uap (liters/yr) 330 infant Table E-5 330 child Table E-5 400 teen Table E-5 310 adult Table E-5 Unit 2 Revision 35 11 80 June 2012

TABLE C-2 Parameters for the Grass-Cow-Meat Pathway Reference Parameter Value (Reg. Guide 1.109 Rev. 1) r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 Ff (pCi/kg per pCi/day) Each stable element Table E- I UWp (kg/yr) 0 infant Table E-5 41 child Table E-5 65 teen Table E-5 110 adult Table E-5 (DFL)Ija (.mrem/pCi) Each radionuclide Tables E-11 to E-14 Yp (kg/m 2) 0.7 Table E-15 Y5 (kg/mr2) 2.0 Table E-15 th (seconds) 7.78E6 (90 days) Table E- 15 tf (seconds) 1.73E6 (20 days) Table E-15 Qf (kg/day) 50 Table E-3 Unit 2 Revision 35 II 81 June 2012

TABLE C-3 Parameters for the Vegetable Pathway Reference Value (Rev- Guide 1.109 Rev. 1)

Parameter r (dimensionless) 1.0 (radioiodines) Table E-I 0.2 (particulates) Table E-I Each radionuclide Tables E-I I to E-14 (DFL)ija (mrem/pCi) 0 Table E-5 UL) (kg/yr) - infant

- child 26 Table E-5

- teen 42 Table E-5 64 Table E-5

- adult 0 Table E-5 US% (kg/yr) - infant

- child 520 Table E-5

- teen 630 Table E-5

- adult 520 Table E-5 (seconds) 8.64E4 (1 day) Table E-15 tL 5.1 8E6 (60 days) Table E- 15 th (seconds) 2.0 Table E- 15 Y, (kg/mr2)

Unit 2 Revision 35 11 82 June 2012

APPENDIX D DIAGRAMS OF LIQUID AND GASEOUS TREATMENT SYSTEMS AND MONITORING SYSTEMS Unit 2 Revision 35 I183 June 2012

Liquid Radwaste Treatment System Diagrams Unit 2 Revision 35 II 84 Jume 2012

TYPICAL OF A SPENT FUIL PL OOLING SYSTEM RNVASTE.

i3ERINERAL IZER '

REACTOR WATER, CLEAMUP SYSTEM REGENERANT .

EVAPORATOR*

CONDENSATE, DEHINERALIZERS REACTOR SUILDIKC E*UIPMENT DRAINS PHASE SEPARATOR RECOVERY, SAMPLE TAWKS' TURBINE RL.,O EQUIPMENT DRAINS RESIDUAL HEATI REMOVAL SYSTEM RECOVERY SAMPLE SYSTEM RADWASTE, AOV3iB FIL.TERS'i FLOOR ORA IN FILTER RADTASRE AOV38'7 FILTERS I WASTE COLLECTION Unit 2 Revision 34 II 85 July 2011

ltwvu WASMiCOM FLCC WA WAM1 CLECTM eMMsaWl WASTE COLLECTOR TREATMENT SYSTEM1 Unit 2 Revision 34 1 86 July 2011

TYPICAL OF 2 WWII

&"UM -- *.ab" f I~b Z 1 IKEN AOV279 EVAP RECOVERY SAMPLE SYSTEM and WASTE DISCHARGE SAMPLE SYSTEM Unit 2 Revision 34 II 87 July 2011

RECEN AUX.

EVAP CST RX $PENI BOILER RW DIT5 BLOE. THERtEX BLDG RESIN BLDG BLOC COOLER GRAINS SYSTEM DRAINS TANKS. S"RP DRAINS FLOOR DRAIN COLLECTION SYSTEM Unit 2 Revision 34 II 88 July 2011

WASTE COLLECTOR SURCE TANK INSTRUMFNT FLOOR DRAIN AIR COLLECTOR SURGE TANK FLOOR DRAIN COLLECTOR TANKS REGEN. WASTE 1 ANK S WASTE COLLECTOR SLUDGE TANK TANKS ACV21 4 REGEN. WASTE TANK 1AOV127 WASTE DISCH.

SAMPLE TANK 1A0Y12 3 FLOOR DRAIN COLLECTOR 1ANK AOV 126 WASTE COLLECTOR.

TANK FLOOR DRAIN FILTER SYSTEM Unit 2 Revision 34 11 89 July 2011

CONDENSATE FILTRATION (IRON PREFILTER)

ALTERNATE TYPICAL OF 2 THERMEX -------------------------------------------

SYSTEM COND DEMIN RW FLOOR AND EQUIPMENT DRAINS

'T 3 RW FILTER AC BACK WASH PUMPS T-I I

I J

)

OTHER REGENERANT I

) WASTE TANK 0

OTHER REGEN WASTE PUMP REGENERANT WASTE TANK PUMP P3A/B SYSTEM REGENERANT WASTE SYSTEM Unit 2 Revision 34 II 90 July 2011

4 501 - COLL. TANKS SOTs EFLOOR DRAIN AgV 32 AtS COLLi T LWAASTE DISCH AOVI47 SAMPLE 1ANKS REGENERANT EVAPORATOR SYSTEM Unit 2 Revision 34 1191 July2011

Gaseous Treatment System Diagrams Unit 2 Revision 34 11 92 July 2011

-VQ--- to MA SOy, 11t ."o TOWora" I m~rmw

Title:

CONDENSER AIR REMOVAL SYSTEM Unit 2 Revision 34 II 93 July 2011

10aft'S OFFGAS RECOMBINERS Unit 2 Revision 34 11 94 July 2011

To 04%"oft

, .. .I_

,ou,,- -__..- "- - - "- -" J Li I-OCAT AMD EQUZVvh E WqSm* TMZS 'T;.........

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OWACYG' OAkaPa Title- STANDBY GAS TREATMENT SYSTEM Unit 2 Revision 34 I197 July 2011

Liquid Radiation Monitoring Diagrams Unit 2 Revision 34 11 98 July 2011

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NIMUR4' MIOHAWK PO~Wf COI0*C5'AflON NINE MILE POINT-UNrf 2 PtS~AL&AF'~TY ANIA,'rSI RUf*V Unit 2 Revision 34 II 99 July 2011

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Gaseous Effluent Monitoring System Diagrams Unit 2 Revision 34 11 103 July 2011

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Unit 2 Revision 34 11 104 July 2011

Deleted Unit 2 Revision 34 11105 July 2011

APPENDIX E NINE MILE POINT ON-SITE AND OFF-SITE MAPS Unit 2 Rcvision 34 11106 July 2011

FIGURE D 5.1-1 NINE MILE POINT ON-SITE MAJ SENVIRONMENTAL SAMPLE .1 LOCATIONS Scale(Mles) 1 Unit 2 Revision 34 11 107 July 2011

FIGURE D 5.1-2 (1 of 2)

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ML13130A006

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