Text

Radioactive Effluent Release Report, January - December 2008
	ML091280433
Person / Time
Site:	Nine Mile Point
Issue date:	05/01/2009
From:	Syrell T; Constellation Energy Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML091280433 (149)
	v • d • e

Constellation Energy- P.O. Box 63NY 13093 Lycoming, Nine Mile Point Nuclear Station May 1, 2009 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 2008 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 2008.

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 the inoperability of any station liquid and/or gaseous effluent monitoring instrumentation 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 is a copy of Revision 31 of the ODCM. 3 is a copy of Revision 07 of the PCP. Attachment 14 includes updated 2007 Radioactive Effluent Release Report Attachments 2 and 3 that correct a units error.

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 2008, 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.

Ve truly yours, Terry SyrellV~

Director Licensing TFS/KES

Enclosure:

Radioactive Effluent Release Report, January - December 2008 ---

Document Control Desk May 1, 2009 Page 2 cc: S. J. Collins, NRC Region I Administrator R.V. Guzman, NRC Project Manager J. Furia, NRC Senior NRC Resident Inspector

ENCLOSURE NINE MILE POINT NUCLEAR STATION, UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT January - December 2008 Nine Mile Point Nuclear Station, LLC May 1, 2009

NINE MILE POINT NUCLEAR STATION - UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT January- December 2008 Nine Mile Point Nuclear Station Constellation Energy-

Page 1 of 2 NINE MILE POINT NUCLEAR STATION - UNIT 2 RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER 2008 SUPPLEMENTAL INFORMATION Facility: Nine Mile Point Unit #2 Licensee: Nine Mile Point Nuclear Station, LLC

l. 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) limits 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 on-line gamma spectroscopic monitoring (intrinsic germanium crystal) 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.

ATTACHMENT 1

SUMMARY

DATA Page 1 of 2 Unit 1 Unit 2 X Reporting Period January - December 2008 Liquid Effluents:

ODCM Required MEC = 10 x 10CFR20.1001 - 20.2402, Appendix B, Table 2, Column 2 Average MEC - pCi/ml (Qtr. 1) = I NO RELEASES Average MEC - pCi/m, (Qtr. ) = i REEASES]

Average MEC - pCi/ml (Qtr. 2) = 1.59E-02 Average MEC - pCi/ml (Qtr. 4) = 1.O0E-02 Average Energy (Fission and Activation gases - MEV):

Qrtr. 1: Ey = 9.34E-01 = 3.06E-01 Qrtr. 2: Ey = 9.21 E-01 I3 = 3.67E-01 Qrtr. 3: Ey = 8.92E-01 I3 = 2.91E-01 Qrtr. 4: Ey = 7.80E-01 1I3 = 7.49E-01 Liquid:

Number of Batch Releases 16 Total Time Period for Batch Releases (hrs) 3.91 E+01 Maximum Time Period for a Batch Release (hrs) 3.38E+00 Average Time Period for a Batch Release (hrs) 2.44E+00 Minimum Time Period for a Batch Release (hrs) 1.66E-02 Total volume of water used to dilute the liquid ist 2nd 3rd 4th during the release period (L) INo Releasesl 1.26E+08 INo Releasesl 1.36E+07 I Total volume of water available to dilute the liquid I 1st 2nd 3rd 4th effluent during the report period (L)I 1.17E+10 1.15E+10 1.36E+10 1-32E+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 14 Total Time Period for Batch Releases (hrs) 3.40E+02 Maximum Time Period for a Batch Release (hrs) 6.67E+01 Average Time Period for a Batch Release (hrs) 2.43E+01 Minimum Time Period for a Batch Release (hrs) 1.58E+00

ATTACHMENT 1

SUMMARY

DATA Page 2 of 2 Unit I Unit 2 X Reporting Period January - December 2008 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 2008 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 1.71E+02 2.71E+01 3.83E+01 1.23E+02 5.OOE+01

2. Average Release Rate pCi/sec 2.18E+01 3.44E+00 4.81E+00 1.55E+01 B. lodines

1. Total Iodine- 131 Ci 1.57E-03 2.12E-03 3.45E-04 1.56E-04 3.O0E+01

2. Average Release Rate for Period pCi/sec 2.OOE-04 2.70E-04 4.08E-05 2.15E-05 C. Particulates

1. Particulates with half-lives>8days Ci 3.80E-04 1.61 E-03 9.21E-05 2.69E-04 3.OOE+01

2. Average Release Rate for Period pCi/sec 7.16E-05 2.46E-04 1.50E-05 3.78E-05

3. Gross alpha radioactivity Ci 0.OOE+00 0.OOE+00 0.OOE+00 7.07E-08 2.50E+01 D. Tritium

1. Total release Ci 3.96E+01 1.50E+01 2.11E+01 1.57E+01 5.OOE+01

2. Average Release Rate for Period pCi/sec 5.05E+00 1.91E+00 2.49E+00 2.17E+00 E. Percent of Tech. Spec. Limits Fission and Activation Gases Percent of Quarterly Gamma Air Dose 3.78E-01 5.80E-02 8.42E-02 2.24E-01 Limit (5 mR)

Percent of Quarterly Beta Air Dose Limit 7.80E-03 1.10E-03 3.08E-03 1.21 E-02 (10 mrad)

Percent of Annual Gamma Air Dose 1.89E-01 2.18E-01 2.61E-01 3.73E-01 Limit to Date (10 mR)

Percent of Annual Beta Air Dose Limit to 3.90E-03 4.46E-03 6.OOE-03 1.21 E-02 Date (20 mrad)

Percent of Whole Body Dose Rate Limit 1.47E-02 2.26E-03 3.24E-03 8.59E-03 (500 mrem/yr)

Percent of Skin Dose Rate Limit (3000 2.89E-03 4.42E-04 6.48E-04 1.79E-03 mrem/yr)

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

Percent of Quarterly Dose Limit (7.5 4.05E-01 5.05E-01 9.12E-02 4.61E-02 mrem)

Percent of Annual Dose Limit to Date 2.04E-01 4.58E-01 5.04E-01 5.27E-01 (15 mtrem)

Percent of Organ Dose Limit (1500 8.15E-03 1.02E-02 1.82E-03 9.17E-04 mrem/yr

ATTACHMENT 3 Page 1 of 1 Unit I Unit 2 X Reporting Period January - December 2008 GASEOUS EFFLUENTS - ELEVATED RELEASE Continuous Mode (2)

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

Argon-41 Ci 1.35E-01 .....

Krypton-85 Ci Krypton-85m Ci 5.20E+01 1.16E+01 1.80E+01 1.82E+01 Krypton-87 Ci 4.13E+00 6.11E-01 **_ 2.14E+00 Krypton-88 Ci 7.36E+01 8.88E+00 1.56E+01 2.55E+01 Xenon-127 Ci Xenon-131m Ci Xenon-133 Ci 3.28E+01 7.06E-01 2.07E+00 5.12E+00 Xenon-133m Ci 8.78E-03 .... *. **

Xenon-1 35 Ci 1.61E+00 5.46E-01 1.30E-01 2.86E+00 Xenon-1 35m Ci 3.36E+00 6.06E-01 **_ 6.63E+00 Xenon-1 37 Ci 4.76E-02 .... 3.44E+01 Xenon-1 38 Ci 2.54E-01 4.09E+00 ** 2.62E+01 lodines (1)

Iodine-131 Ci 1.45E-03 1.91 E-03 6.60E-05 7.35E-05 Iodine-133 Ci 1.12E-02 9.84E-05 4.OOE-04 5.64E-04 Iodine-135 Ci Particulates (1)

Chromium-51 Ci Manganese-54 Ci ** 1.89E-05 ....

Iron-55 Ci 7.69E-05 2.04E-04 ** 6.50E-05 Iron-59 Ci Cobalt-58 Ci Cobalt-60 Ci Neodymium-i147 Ci Zinc-65 Ci Strontium-89 Ci ** 2.37E-05 ** 1.22E-05 Stronium-90 Ci Niobium-95 Ci Zirconium-95 Ci 1.30E-05 ......

Molybdenum-99 Ci Cesium-134 Ci Cesium-136 Ci Cesium-137 Ci ......

2.20E-06 Barium-140 Ci Lanthanum-140 Ci Cerium-141 Ci Cerium-144 Ci Tritium (1) Ci i1.25E+01 i 8.77E+00 1.49E+01 i1.02E+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 of 1.00E-04 pCi/ml for required noble gases, 1.OOE-11 pCi/ml for required particulates and gross alpha, 1.00E-12 pCi/ml-for required lodines, 1.OOE-11 pCi/ml for Sr-89/90 and 1.00E-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 2008 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-131m Ci Xenon-133 Ci 6.25E-01 ** 7.17E-01 1.1OE+00 Xenon-133m Ci Xenon-135 Ci 4.03E-01 **_ 5.52E-01 7.63E-01 Xenon-135m Ci 2.12E+00 ** 1.16E+00 Xenon-137 Ci Xenon-138 Ci lodines (1)

Iodine-131 Ci 1.21 E-04 2.08E-04 2.79E-04 8.22E-05 Iodine-133 Ci 8.18E-04 6.79E-04 7.08E-04 1.81E-03 Iodine-135 Ci

- I ** I ** **

Particulates (1)

Chromium-51 Ci Manganese-54 Ci 4.48E-05 1.95E-04 ..

Iron-55 Ci 4.17E-05 1.04E-03 7.28E-05 1.90E-04 Iron-59 Ci ** 3.41 E-05 ....

Cobalt-58 Ci Cobalt-60 Ci 2.36E-05 9.33E-05 1.93E-05 **

Neodymium-147 Ci Zinc-65 Ci Stronium-89 Ci 1.80E-04 .....

Stronium-90 Ci Niobium-95 Ci Zirconium-95 Ci Ci ** 1.08E-05 9.69E-06 **

Molybdenum-99 Cesium-134 Ci Cesium-136 Ci Cesium-137 Ci Barium-140 Ci Lanthanum-140 Ci Cerium-141 Ci Cerium-144 Ci Tritium Ci 2.71E+01 6.26E+00 6.15E+00 5.55E+00 (1) Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk. A lower limit of detection of 1.00E-04 pCi/mi for required noble gases, 1.00E-11 pCi/ml for required particulates and gross alpha, I.00E-12 pCi/ml for required lodines, 1.00E-11 pCi/ml for Sr-89/90 and 1.OOE-06 pCi/ml 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 2008 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES (1) 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Est. Total Error, %

A. Fission & Activation Products

1. Total Release (not including Tritium, No Releases 3.07E-03 No Releases **

Ci 5.OOE+01 gases, alpha)

2. Average diluted concentration during pCi/ml No Releases 2.68E-10 No Releases **

reporting period B. Tritium 1 .Total release Ci No Releases 7.26E+00 No Releases 1.54E-04 5.OOE+01

2. Average diluted concentration during the No Releases 6.33E-07 No Releases 11.16E-11 pCi/ml' reporting period C. Dissolved and Entrained Gases

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

2. Average diluted concentration during the pCi/ml No Releases ** No Releases **

reporting period D. Gross Alpha Radioactivity

1. Total release Ci [No Releases [ ** ]No Releases J ** 5.OOE+01 E. Volumes

1. Prior to Dilution Liters No Releases 9.77E+05 No Releases 1.51 E+04 5.OOE+01

2. Volume of dilution water used during Liters No Releases 1.26E+08 No Releases 1.36E+07 5.OQE+01 release period

3. Volume of dilution water available during Liters 1.17E+10 1.15E+10 1.36E+10 1.32E+10 5.OOE+01 reporting period F. Percent of Tech. Spec. Limits Percent of Quarterly Whole Body Dose No Releases 1.69E-02 No Releases 1.06E-07 Limit (1.5 mrem)

Percent of Annual Whole Body Dose Limit No Releases 8.46E-03 No Releases 8.46E-03 to Date (3 mrem)

Percent of Quarterly Organ Dose Limit (5 No Releases 2.72E-02 No Releases 3.17E-08 mrem)

Percent of Annual Organ Dose Limit to 1.36E-02 No Releases 1.36E-02 No Releases Date (10 mrem)

Percent of 10CFR20 Concentration Limit No Releases 6.76E-03 No Releases 1.16E-07 (2), (3)

Percent of Dissolved or Entrained Noble No Releases No Releases ** **

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), have 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 2008 LIQUID EFFLUENTS RELEASED Batch Mode (1),(2)

Nuclides Released 1st Quarter 2nd Quarter 3rd Quarter 4th Quarter Nuclides Released Strontium-89 Ci Strontium-90 Ci Cesium-134 Ci Cesium-137 Ci Iodine-131 Ci Cobalt-58 Ci ** 2.OOE-05***

Cobalt-60 Ci .30E-03 1* **

Iron-59 Ci .iiE-04*

1*

Zinc-65 Ci Manganese-54 Ci ** 1.31E-03

Chromium-51 Ci 2.99E-04*

Zirconium-95 Ci Niobium-95 Ci Molybdenum-99 Ci Technetium-99m Ci Barium-140 Ci Lanthanum-140 Ci Cerium-141 Ci ** 3.29E-05 **

Tungsten-187 Ci Arsenic-76 Ci Iodine-133 Ci Iron-55 Ci Neptunium-239 Ci Silver-11 Om Ci Gold-1 99 Ci Cerium-144 Ci Cesium-136 Ci Copper-64 Ci Dissolved or Entrained Gases Ci Tritium (3),(4) Ci ** 7.26E+00 * . ý54E-04 (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.O0E-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/mI for gross alpha radioactivity, as identified in the ODCM, has been verified.

(3) 2nd Quarter tritium activity includes the 4/2/08 contribution of 1.71 E-02 Ci from one post-analysis discharge via the service water discharge pathway for Closed Cooling Primary cross-tie to Spent Fuel Pool Cooling heat exchanger.

(4) 4th Quarter tritium activity reflects the contribution from four tanks of water collected from the Circulating Water Pump bay pits and discharged via the service water discharge pathway following analysis.

ATTACHMENT 6 Page 1 of 4 Unit I Unit 2 X Reporting Period January - December 2008 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS Al. TYPE Volume Activity (1)

(m3) (Ci)

Class Class A B C A B C a.1 Spent Resins 5.63E+01 O.OOE+00 0.OOE+00 2.57E+02 O.OOE+00 O.OOE+00 (Dewatered) a.2 Filter Sludge O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 a.3 Concentrated Waste 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Totals 5.63E+01 O.OOE+00 0.OOE+00 2.57E+02 O.OOE+00 0.OOE+00 b.1 Dry, compressible waste 5.89E+02 O.OOE+00 O.OOE+00 1.72E+00 O.OOE+00 O.OOE+00 b.2 Dry, non-compressible 0.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 waste (contaminated equipment)

Totals 5.89E+02 0.OOE+00 0.OOE+00 1.72E+/--0O 0.OOE+00 0.OOE+00

c. Irradiated Components, 0.OOE+00 0.OOE+00 1.52E-01 0.OOE+00 0.0OE+00 1.89E+04 Control Rods

d. Other (to vendor for processing) d.1 DryActiveWaste 5.37E+01 O.OOE+00 0.OOE+00 2.11E+01 0.OOE+00 0.00E+00 (compactible) High Rad Trash (1) The estimated total error is 5OOE+01%

ATTACHMENT 6 Page 2 of 4 Unit 1 Unit 2 X Reporting Period January - December 2008 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 a.3 Concentrated Waste N/A N/A N/A b.1 Dry, Compressible waste Metal Box General Design None b.2 Dry, non-compressible waste (Contaminated N/A N/A N/A Equipment)

c. Irradiated Components, Steel Liner Type B None Control Rods

d. Other (to vendor for processing) d.1 Dry Active Waste (Compactible) (High Rad Steel Liner / Metal Box General Design None Trash)

ATTACHMENT 6 Page 3 of 4 Unit I Unit 2 X Reporting Period January - December 2008 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 75.4 Co-60 12.9 Mn-54 7.1 Zn-65 3.4 H-3, Cr-51, Fe-59, Co-58, Ni-63, Nb-95, Se-124, Cs-134, Cs-137, 1.2 La-140

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

Nuclide Percent Fe-55 80.6 Co-60 15.1 Mn-54 3.4 Fe-59, Ni-63, Zn-65, Sb-125, Cs-137, Ce-144 0.9

c. Irradiated Components, Control Rods Nuclide Percent Co-60 58.9 Fe-55 36.1 Ni-63 4.4 H-3, C-14, Cr-51, Mn-54, Fe-59, Co-58, Ni-59, Zn-65, Sr-90, Zr-95, 0.6 Nb-94, Mo-93, Tc-99, Ag-110m, 1-129, Cs-134, Cs-137

d. Other (To Vendor for Processing)

1. Dry Active Waste (Compactible) (High Rad Trash)

Nuclide Percent Fe-55 65.7 Co-60 29.0 Mn-54 3.0 H-3, C-14, Fe-59, Co-58, Ni-59, Ni-63, Zn-65, Sr-90, Tc-99, Ag-11im, Sb-125, Cs-134, Cs-137, Ce-144, Pu-238, Pu-239, Pu-241, Am-241, 2.3 Cm-242, Cm-243

ATTACHMENT 6 Page 4 of 4 Unit I Unit 2 X Reporting Period January - December 2008 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A3. SOLID WASTE DISPOSITION Number of Shipments Mode of Transportation Destination 1 Hittman Transport Barnwell Disposal Facility 15 Hittman Transport Duratek Services, Inc.

3 Hittman Transport Studsvik Processing Facility - Memphis 1 Studsvik Logistics Studsvik Processing Facility - Memphis 13 Hittman Transport Studsvik Processing Facility - Erwin 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 Sewage wastes shipped to a treatment facility for processing and burial, if any, are reported in the Nine Mile Point Nuclear Station's Unit No. 1 Radioactive Effluent Release Report.

ATTACHMENT 7 Page 1 of 1 Unit I Unit 2 X Reporting Period January - December 2008

SUMMARY

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

The Unit 2 Off-Site Dose Calculation Manual (ODCM) was revised during the reporting period to incorporate plant modifications (addition of iron prefilters, replacement of GEMS stack and vent effluent monitors), include more accurate environmental monitoring program sample locations based on GPS data, add clarifications and enhancements, and correct typos.

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 do not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. A copy of the ODCM, Revision 31 is attached and a summary of the changes presented to and approved by the Plant Operations Review Committee on November 20, 2008 is provided below. The summary also includes the justification for the change.

REVISION 31 Page # New/Amended Description of Change Reason For Change I 1.0-3 1.0 Format alignment error Correct typo I 1.0-4 Figure D 1.0-1 Electonically generated figure Enhancement I 3.2-2 Table D 3.2.1-1 Reformatted for clarity Enhancement 13.2-3 Table D 3.2.1-1 .Footnote (d); replaced "isotopic" with Plant modification; replacement of stack "radiation" and vent monitors 14.2-1 & I Section D 4.2 Replaced "SORC" with "PORC" Title change 4.4-2 119 Section 1.3Added text to use estimates until Clarification actual results are obtained.

1113 & Sections 2.1.2, Revised text to describe new stack Plant modification; replacement of stack 1114 2.1.2.1 & 2.1.2.2 and vent replacement monitors and vent monitors 1116 Section 2.2 Added text to use estimates until Clarification actual results are obtained.

II 20 Section 2.3 Added text to use estimates until aculrslsaeotie.Clarification actual results are obtained.

II 58 Table D 3-21 Nd-147, Vegetation, Teen, Total Correct typo Body dose factor II 62 to 65 Table D 5.1 Updated location distances and Enhancement azimuths using GPS data II 90 Regenerant Waste Added input from Iron Prefilter Plant modification

_System Figure 11104 Gaseous Radiation Deleted the word "isotopic" Plant modification; replacement of stack

_ Monitoring Figure and vent monitors 11'105 Block Diagram Deleted figure Plant modification; replacement of stack ITypical Gaseous and vent monitors 11107, II Figures D 5.1-1, & D Electonically generated figure Enhancement 109 & 5.1-2 109

ATTACHMENT 8 Page 1 of 3 Unit I Unit 2 X Reporting Period January - December 2008

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)

The Unit 2 Radwaste Process Control Program (RPCP) Revision 07 was implemented in June 2008. The Revision reflects the implementation of a condensate "prefilter" filtration system upstream of the deep bed condensate demineralizers and an associated change in the use, operation and naming of the Evaporator Bottoms Tank and Pump. The affect on the RPCP is the addition of a new generation source (the Iron Prefilter System) for both Liquid and Solid waste. The RPCP changes do not reduce the overall conformance of a solidified waste product to existing criteria for solid waste. A copy of the RPCP, Revision 07 is attached and a summary of the changes presented to and approved by the Plant Operations Review Committee on June 18, 2008 is provided below. The summary also includes the justification for the change.

REVISION 07 Page # New/Amended Description of Change Reason For Change Section #

Changes text from 'The Plant Manager" Change to reflect the current organizational titles.

Page 1 Section 2.1 to "The Plant General Manager".

The text at the end of the sentence that Eliminates guidance that is otherwise controlled under separate had indicated "in accordance with the programs.

Page 1 Section 2.1.2 Quality Assurance program" is deleted.

Changes text from 'The Managers of Grammatical correction.

Page 1 Section 2.2 Operations" to 'The Manager of Operations".

Changed the specified equipment name The modification for implemenetation of the condensate "prefilter" from "Evaporator Bottoms Tank" to filtration system changed the use, operation and specified name of Page 2 Section 3.1.2 "Condensate Filtration System Phase the Evaporator Bottoms Tank and Pump.

Separator".

Replaced information regarding the The modification for implementation of the condensate "prefilter" Evaporator Bottoms Tank with new. filtration system changed the use, operation and specified name of information identifying that the the Evaporator Bottoms Tank and Pump. The tank is now, Page 2 Section 3.1.2 (a) Condensate Filtration System Phase designated as the Condensate Filtration System Phase Separator Separator tank may be decanted to the Tank and remains 2LWS-TK10.

Floor Drain Collector System.

Identifies that the contents of the The Evaporator Bottoms Tank (2LWS-TK10) has been re-named as Condensate Filtration System Phase the Condensate Filtration System Phase Separator Tank (2LWS-Separator Tank may be recirculated or TK1 0) and is now being used to process the insoluble waste Page 2 Section 3.1.2 (b) transferred to a liner in the Radwaste generated by the new condensate "prefilter" system.

Truck Bay for further offsite processing.

Replaced Evaporator Bottoms Tank The Evaporator Bottoms Tank (2LWS-TK10) has been re-named as (TK10) with Condensate Prefilter/Phase the Condensate Filtration System Phase Separator Tank (2LWS-Page 9 Section 3.4 (a) Separator Tank (TK1 0) designation for TK1 0) and is now being used to process the insoluble waste isolation when preparing to process generated by the new condensate "prefilter" system.

waste.

Replaced Evaporator Bottoms Tank The ability to process insoluble impurities removed from the (TK10) with Condensate Prefilter/Phase condensate process stream by the prefilters requires a collection Separator Tank (TK1 0) designation for tank for settling out of the solids. The pre-existing Concentrated Page 9 Section 3.4 (b) recirculation to ensure homogenous Waste Tank now serves this function and was effectively integrated mixture for sampling, into the overall design as an efficiency in the design process.

ATTACHMENT 8 Page 2 of 3 Unit I Unit 2 X Reporting Period January - December 2008

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM (PCP) (continued)

The Unit 2 Radwaste Process Control Program (RPCP) Revision 07 was implemented in June 2008. The Revision reflects the implementation of a condensate "prefilter" filtration system upstream of the deep bed condensate demineralizers and an associated change in the use, operation and naming of the Evaporator Bottoms Tank and Pump. The affect on the RPCP is the addition of a new generation source (the Iron Prefilter System) for both Liquid and Solid waste. The RPCP changes do not reduce the overall conformance of a solidified waste product to existing criteria for solid waste. A copy of the RPCP, Revision 07 is attached and a summary of the changes presented to and approved by the Plant Operations Review Committee on June 18, 2008 is provided below. The summary also includes the justification for the change.

iped to a treatment facility for processing and burial, if any, are reported in the Nine Mile Point Nuclear Station's Unit No. 1 Radioactive Effit Page # New/Amended Description of Change Reason For Change Replaced the position title from "The Change to reflect the current organizational titles.

Manager, Nuclear QA Operations" to Page 10 NOTE for Section 3.6 "The Director of Quality and Performance Assessment".

The step text ", under the cognizance of Eliminates guidance that is otherwise controlled under separate Page 10 Section 3.6.1 (b) the SRAB," is deleted, programs.

Change from "Meet the applicable Updated to reflect the current Quality Assurance Topical Report requirements of QATR-1, Quality section reference and administrative procedure references for Assurance Program Topical Report for control of training records.

Nine Mile Point Nuclear Station Operations, Section 17.0, Quality Assurance Records, NIP-TQS-01, Qualification and Certification, and NIP-RMG-01, Identification, Maintenance, Page 11 Section 3.6.2 (b.3) Storage and Transfer of Nuclear Division Records" to "Meet the applicable requirements of the Quality Assurance Topical Report Section B.15 Records, CNG-TR-1.01-1000 Conduct of Training and CNG-PR-3.01-1000 Records Management" for the administration and control of training records.

Change from "Station Operations Updated to reflect the current title of the Plant Operations Review Review Committee" to "Plant Committee.

Operations Review Committee."

Page 11 Section 3.6.4 (b)

Change from "The Supervisor of Updated to reflect the current organizational title and Radwaste" to 'The Supervisor responsibilities.

Radioactive Materials Processing."

Change the reference from "QATR-1, Updated reference to reflect the current Quality Assurance Topical Quality Assurance Program Topical Report.

Report for Nine Mile Point Nuclear Page 13 Section 5.1.1 Station Operations, Section 17.0, Quality Assurance Records", to "Quality Assurance Topical Report Section B.15 Records."

.ATTACHMENT 8 Page 3 of 3 Unit I ' Unit 2 X 'Reporting Period January - December 2008

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM (PCP) (continued)

The Unit 2 Radwaste Process Control Program (RPCP) Revision 07, was implemented in June 2008. The Revision reflects the implementation of a condensate "prefilter" filtration system upstream of the deep bed condensate demineralizers and an associated change in the use, operation and naming of the Evaporator Bottoms Tank and Pump. The affect on the RPCP is the addition of a new generation source (the Iron Prefilter System) for both Liquid and Solid waste. The RPCP changes do not reduce the overall conformance of a solidified waste product to existing criteria for solid waste. A copy of the RPCP, Revision 07 is attached and a summary of the changes presented to and approved by the Plant Operations Review Committee on June 18, 2608 is provided below. The summary also includes the justification for the change.

REVISION 07 Page # New/Amended Description of Change Reason For Change New supplemental reference added for The new supplemental reference identifies the design change under DCP N2-05-064, Condensate Filtration which modification to repurpose and rename the Evaporator Page 14 Section 5.3.11 Radwaste Processing. Bottoms Tank and Pump was made.

Edited the contained program reference Updated to reflect the current procedure series title designation.

from "Quality Assurance Procedures Page 15 Attachment 1 (QAPs)" to Quality Assurance Procedures (CNG-QL series)."

Inserted a new Section 3.0 (withsub- The modification for implementation of the condensate filtration sections 3.1, 3.2 and 3.3) for the solid "prefilter" system creates a new generation source for both Liquid waste source from the condensate and Solid waste. The filtration system removes iron particulates filtration system. Renumber the from the BWR condensate process stream. The associated filters remainder of Attachment 2 as are periodically backwashed. The bachwash is collected in a, Page 16 Attachment 2, appropriate, receiving tank and the iron ladened waste is pumped to a' phase Section 3.0 separator tank where flocculent is added to assist in separation.

The resultant sludge is pumped to a liner for offsite processing.

The Condensate Prefilter elements are treated as solid Radwaste at the end of there useful life. Elements are shipped offsite for vendor processing.

The Waste Evaporator and Regenerant The Waste Evaporator and Regenerant Evaporator are not being Evaporator solid waste sources utilized to concentrate solids from the waste stream and future use identifed as Section 10.0 and 11.0, is not anticipated.

Page17hrespectively in Attachment 2 of the and Attachment 2 previous Revision 6 have been deleted.

Page 18 Renumber the remainder of Attachment 2 as appropriate.

ATTACHMENT 9 Page 1of 1 Unit 1 Unit 2 X Reporting Period January - December 2008

SUMMARY

OF INOPERABLE MONITORS Monitor Dates of Inoperability Cause and Corrective Actions Liquid Radwaste May 27, 2008 @ The monthly functional test frequency was exceeded. The Effluent Line 15:30 and continuing monitor is not required to be operable if no discharge is in Radioactivity Monitor, progress. Discharge isolation manual valves 2LWS-V420 and 2LWS-CA8206 2LWS-V422 are locked closed, therefore no inadvertent discharge can occur. The monitor remains inoperable until a liquid waste discharge is needed.

Stack Gaseous The Stack GEMS became inoperable due to an unidentified Effluent Monitoring failure. Prior to the identification of the failure, the system was System (GEMS), removed from service for replacement with a new design.

2RMS-CAB170 and Required compensatory actions remained in effect for the 2RMS-RAK170. balance of 2008. The new system was made functional on a) Noble Gas Activity a) April 16, 2008 February 3, 2009.

Monitor through December 31, 2008.

b) Iodine Sampler b, c, d, e) April 21, c) Particulate 2008 through Sampler December 31, 2008 d) Flow-Rate Monitor e) Sample Flow Rate Monitor Radwaste/Reactor The Vent GEMS became inoperable due to an unidentified Building Vent Effluent failure. Prior to the identification of the failure, the system was Monitoring System removed from service for replacement with a new design.

(GEMS), Required compensatory actions remained in effect for the 2RMS-CAB180 and balance of 2008. The new system was made functional on 2RMS-RAK1 80. February 3, 2009.

a) Noble Gas Activity a) April 16, 2008 Monitor through December

31,2008 b) Iodine Sampler b, c, d, e) April 21, c) Particulate 2008 through Sampler December 31, 2008 d) Flow-Rate Monitor e) Sample Flow Rate Monitor

Attachment 10 Page I of 3 Unit 1 Unit 2 X Reporting Period January - December 2008 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 2008.

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 NMP. 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 2008 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 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 Nine Mile Point 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 NMP2 Stack and Vent.

" Direct radiation pathway; dose resulting from the operation of NMP2, Nine Mile Point Unit 1 (NMP1) and the James A.

Fitzpatrick (JAF) Facilities.

Methodologies for Determining Dose for Applicable Pathways External Ground (Shoreline Sediment) pathway Dose from the external ground (shoreline sediment) is based on the methodology in the Unit 2 Offsite Dose Calculation Manual (NMP2 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.

The total dose received by the whole body and skin of the maximum exposed individual during 2008 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 = 1 (used average radionuclide concentration over total time period)

Average Cs-137 concentration = 1.57 E-01 pdi/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.

Attachment 10 Page 2 of 3 Unit 1 Unit 2 X Reporting Period January - December 2008 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY 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 2008 is calculated using the following input parameters for gaseous effluents released from both the NMP2 Stack and Vent for the time period exposure is received:

NMP 2 Stack:

Variable Fisherman

3 X/Q (s/m ) 9.60E-07 Inhalation dose factor Table E-7 Regulatory Guide 1.109 3

Annual air intake m /year) (adult) 8000 Fractional portion of the year (hours) 0.0356 H-3 (pCi/sec) 1.43 E+06 Mn-54 (pCi/sec) 8.02 E-01 Fe-55 (pCi/sec) 1.14 E+01 Sr-89 (pCi/sec) 1.18 E+00 Cs-137 (pCi/sec) 9.34 E-02 1-131 (pCi/sec) 8.70 E+01 1-133 (pCi/sec) 4.50 E+01 NMP2 Vent:

Variable Fisherman

3 X/Q (s/m ) 2.80E-06 Inhalation dose factor Table E-7 Regulatory Guide 1.109 Annual air intake (m3/year) (adult) 8000 Fractional portion of the year (hours) 0.0356 H-3 (pCi/sec) 7.62 E+05 Mn-54 (pCi/sec) 8.27 E+00 Fe-55 (pCi/sec) 5.53 E+01 Fe-59 (pCi/sec) .1.45 E+00 Co-60 (pCi/sec) 4.78 E+00 Mo-99 (p~i/sec) 8.70 E-01 1-131 (pCi/sec) 2.42 E+01 1-133 (pCi/sec) 1.35 E+02 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 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.

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 3 of 3 Unit 1 Unit 2 X Reporting Period January - December 2008 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY Direct Radiation Pathway 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

- December 31, TLD data for the second, third, and fourth quarters of 2008 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) 1.53 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 By A Hypothetical Maximum Exposed Member Of The Public Inside the Site Boundary During 2008 The following is a summary of the dose received by a hypothetical maximum exposed fisherman from Liquid and Gaseous effluents released from NMP2 during 2008:

Table I Exposure Pathway Annual Dose Exposure Pathway Dose Type Fisherman (mRem)

External Ground Whole Body 2.47 E-03 Skin of Whole Body 2.88 E-03 Inhalation Whole Body 1.58 E-04 Maximum Organ Thyroid: 2.55,E-04 Direct Radiation Whole Body 0.48 Based on these values the total annual dose received by a hypothetical maximum exposed member of the public isas follows:

Table 2 Annual Dose Summary Total Annual Dose for 2008 Fisherman (mRem)

Total Whole Body 4.83 E-01 Skin of Whole Body 2.88 E--03 Maximum Organ Thyroid: 2.55 E-04

Attachment 11 Page 1 of 2 Unit 1 - Unit 2 X Reporting Period January - December 2008 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 2008 for comparison against the 40CFR190 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 NMP2, Nine Mile Point Unit 1 (NMP1) and the James A. FitzPatrick (JAF) facilities must be considered.

40CFR190 requires the annual radiation dose received by members of the public in the general environment, as a result of plant operations, be limited to:

S< 25 mRem wholebody S< 25 mRem any organ (except thyroid)

S< 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 NMP2, NMP1 and JAF 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 most likely 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.

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 JAF operating facilities.

Direct Radiation pathway; radiation dose resulting from the operation ofNMP1, NMP2 and JAF facilities.

Methodologies for Determining Dose for Applicable Pathways Fish Consumotion Dose received as a result of fish consumption is based on the methodology specified in the NMP2 Off-site Dose Calculation Manual (NMP2 ODCM) as adapted from Regulatory Guide 1.109. The dose for 2008 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 2008; therefore no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2008.

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.

Attachment 11 Page 2 of 2 Unit 1 Unit 2 X Reporting Period January - December 2008 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Shoreline Sediment continued:

No radionuclides were detected in shoreline sediment samples collected and analyzed during 2008;' therefore no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2008.

Dose Pathways Resultina 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 NMP2 ODCM, NMPI Offsite Dose Calculation Manual, and the JAF Offsite Dose Calculation Manual. These calculations consider deposition, inhalation and ingestion pathways. The total sum of doses resulting from gaseous effluents from NMPI, NMP2 and JAF during 2008 provide a total dose to the whole body and maximum organ dose for this pathway.

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 2008, the closest residence and the critical downwind residence are at the same location.

Dose Potentially Received by the Likely Most Exposed Member of the Public Outside the Site Boundary Durinca 2008 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 8.18 E-03 Total Maximum Organ Thyroid: 1.08 E-01 Direct Radiation Total Whole Body 0.48 Based on these values the maximum total annual dose potentially received by the most likely exposed member of the public during 2008 is as follows:

" Total Whole Body: 4.92 E-01 mRem

Total Skin of Whole Body: 4.90E-03 mRem

" Maximum Organ: Thyroid: 1.08 E-01 mRem 40CFR1 90 Evaluation The maximum total doses presented in this attachment are the result of operations at the NMP1, NMP2 and the JAF facilities.

The maximum organ dose (Thyroid: 0.108 mRem) and the maximum whole body dose (0.492 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.

ATTACHMENT 12 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

Constellation Energy, Nine Mile Point Nuclear Station NINE MILE POINT NUCLEAR STATION NINE MILE POINT UNIT 2 OFF-SITE DOSE CALCULATION MANUAL (ODCM)

REVISION 31 APPROVALS SIGNATURES DATE Prepared by: & f7I 8 G. R. Stinson Principle Chemist Reviewed by: 168 C. L. Widay /

Supervisor - Chemistry Support Concurred by:

M. R. Faivus General Supervisor Chemistry

SUMMARY

OFREVISIONS Revision 31 (Effective December 2008)

PAGE DATE 1 3.3-13,14 August 2000 I 3.3-6 November 2000 14.0-1 November 2000 II 2-10,26,33-36,66,67,75,80 November 2000 ix,I 1.0-1,I 1.0-2, 1 B 3.3-2,14.1-1 & la, II 11, 1115,1129,1163, 11107, 11108 December 2001 1 3.3-9 December 2002 1 3.3-10 March 2003 I 3.3-7, i 3.3-12, and I 3.3-13 January 2004 II 63, I1 64, and 11107 December 2005 II 3 and II 4 May 2006 iv, 11.0-1, 1 3.1-7, 1 3.2-3, 1 3.2-10, 1 3.2-12, 1 3.3-1, 1 3.3-2, 1 3.3-3, 13.3-7,13.3-8,1 3.3-9,1 3.3-10, 1 B 3.1-3, 1 B 3.2-5, 1 B 3.2-6, 1 B 3.3-1, I B 3.3-2,1 4.1-la, 1110, 1113,1120, and 1123 September 2006 11 12,1115,1116 September 2007 1116 September 2007 11.0-3,11.0-4,13.2-2,13.2-3,14.2-1,14.2-2,119,1113,1114,1116, II 20, 11 58, 1162-65, II 90,11104, 11105, 11107, 11 108, and 11109 December 2008 i Unit 2 Revision 31 December 2008

TABLE OFCONTENTS PAGE List of Tables vii List of Figures ix Introduction x PART I - RADIOLOGICAL EFFLUENT CONTROLS I SECTION 1.0 DEFINITIONS I 1.0-0 SECTION 2.0 Not Used SECTION 3.0 APPLICABILITY 1 3.0-0 D 3.1 Radioactive Liquid Effluents 13.1-1 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 1 3.2-1 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 I 3.3-1 D 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation I 3.3-1 D 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation I 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 1 3.5-13 D 3.5.3 Interlaboratory Comparison Program 1 3.5-16 BASES I B 3.1-0 B 3.1 Radioactive Liquid Effluents I B 3.1-1 B 3.1.1 Liquid Effluents Concentration I B 3.1-1 B 3.1.2 Liquid Effluents Dose I B 3.1-2 B 3.1.3 Liquid Radwaste Treatment System I B 3.1-3 ii Unit 2 Revision 31 December 2008

TABLE OF CONTENTS (Cont)

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- 131, Iodine-i 33, Tritium, and Radioactive Material in Particulate Form 1B 3.2-3 B 3.2.4 Gaseous Radwaste Treatment System IB 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 1 B 3.3-1 B 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation I B 3.3-1 B 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation I B 3.3-2 B 3.4 Radioactive Effluents Total Dose I B 3.4-1 B 3.5 Radiological Environmental Monitoring 1B 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 1B 3.5-3 SECTION 4.0 ADMINISTRATIVE CONTROLS 1 4.0-1 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 31 December 2008

TABLE OF CONTENTS (Cont)

SECTION SUBJECT REF SECTION PA GE PART II - CALCULATIONAL METHODOLOGIES II1 1.0 LIQUID EFFLUENTS 112 1.1 Liquid Effluent Monitor Alarm Setpoints 112 1.1.1 Basis 3.1.1 112 1.1.2 Setpoint Determination Methodology 3.3.1 112 1.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint 112 1.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculations II 5 1.1.2.3 Service Water and Cooling Tower Blowdown Effluent Radiation Alarm Setpoint 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 II 8 DSR 3.1.2.1 1.4 Liquid Effluent Sampling Representativeness Table D 3.1.1 -1 119 note b 1.5 Liquid Radwaste System FUNCTIONALITY 3.1.3 1110 DSR 3.1.3.1 B 3.1.3 2.0 GASEOUS EFFLUENTS 11 12 2.1 Gaseous Effluent Monitor Alarm Setpoints 11 12 2.1.1 Basis 3.2.1 1112 2.1.2 Setpoint Determination Methodology Discussion 3.3.2 11 12 2.1.2.1 Stack Noble Gas Detector Alarm Setpoint Equation 11 13 2.1.2.2 Vent Noble Gas Detector Alarm Setpoint Equation 11 14 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 DSR3.2.1.1 2.2.3 Skin Dose Rate Due to Noble Gases DLCO 3.2.1.a 11 18 DSR 3.2.1.1 iv Unit 2 Revision 31 December 2008

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 1119 2.3 Gaseous Effluent Dose Calculation Methodology 3.2.2 II 20 3.2.3 3.2.5 2.3.1 W, and Wv - Dispersion Parameters For Dose, Table D 3-23 II 20 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 II 22 2.6 Use of Concurrent Meteorological Data vs. Historical Data I 22 2.7 Gaseous Radwaste Treatment System Operation 3.2.4 II 22 2.8 Ventilation Exhaust Treatment System Operation 3.2.5 II 23 3.0 URANIUM FUEL CYCLE 3.4 II 24 3.1 Evaluation of Doses From Liquid Effluents DSR 3.1.2.1 II 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 II 27 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 II 30 DSR 3.5.1.1 4.2 Interlaboratory Comparison Program DSR 3.5.3.2 II 30 4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements 1131 v Unit 2 Revision 31 December 2008

TABLE OF CONTENTS (Cont)

SECTION SUBJECT REF SECTION PAGE Appendix A Liquid Dose Factor Derivation II 66 Appendix B Plume Shine Dose Factor Derivation II 69 Appendix C Dose Parameters for Iodine 131 and 133, Particulates and Tritium 1173 Appendix D Diagrams of Liquid and Gaseous Radwaste Treatment Systems and Monitoring Systems II 83 Appendix E Nine Mile Point On-Site and Off-Site Maps 11 106 vi Unit 2 Revision 31 December 2008

LIST OF TABLES PART I - RADIOLOGICAL EFFLUENT CONTROLS TABLE NO TITLE PAGE D 3.1.1-1 Radioactive Liquid Waste Sampling and Analysis I 3.1-2 D 3.2.1-1 Radioactive Gaseous Waste Sampling and Analysis I 3.2-2 D 3.3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation I 3.3-6 D 3.3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation 1 3.3-13 D 3.5.1-1 Radiological Environmental Monitoring Program I 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 31 December 2008

LIST OF TABLES (Cont)

PART H - CALCULA TIONAL METHODOLOGIES TABLE NO TITLE PAGE D2-1 Liquid Effluent Detector Response 1133 D 2-2 thru D 2-5 Aiat Values - Liquid Effluent Dose Factor II 34 D 3-1 Offgas Pretreatment Detector Response II 38 D 3-2 Finite Plume - Ground Level Dose II 39 Factors from an Elevated Release D 3-3 Immersion Dose Factors II 40 D 3-4 thru D 3-22 Dose And Dose Rate Factors, Ri 1141 D 3-23 Dispersion Parameters at Controlling II 60 Locations, X/Q, Wv and Ws Values D 3-24 Parameters For the Evaluation of Doses to 1161 Real Members of the Public From Gaseous And Liquid Effluents D5.1 Radiological Environmental Monitoring II 62 Program Sampling Locations viii Unit 2 Revision 31 December 2008

LIST OF FIGURES FIGURENO TITLE PAGE D 1.0-1 Site Area and Land Portion of Exclusion Area Boundaries I 1.0-4 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) 11108 D 5.1-2a Nine Mile Point Off-Site Map (page 2 of 2) 11109 ix Unit 2 Revision 31 December 2008

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

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

The ODCM follows the methodology and models suggested by NUREG-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 31 December 2008

PART I - RADIOLOGICAL EFFLUENT CONTROLS Unit 2 Revision 31 I December 2008

Definitions 1.0 PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 1.0 DEFINITIONS Unit 2 Revision 31 11.0-0 December 2008

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 for recreational, occupational, or other purposes not associated with Nine Mile Point Nuclear Station and James A. FitzPatrick Nuclear Power Plant.

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

(continued)

Unit 2 Revision 31 11.0-1 December 2008

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 31 11.0-2 December 2008

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 and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered safety features (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

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 31 11.0-3 December 2008

Definitions 1.0 FIGURE D1.0-1

JAF Liquld Discharge Lake Ontario NMP Uri[t2, Site Area and Land L-iqu[d Diischarge Portion of Exclusion Area Boundaries

.1 Note: National Grid retains ownership incertain transmission F ] 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 31 11.0-4 December 2008

PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 3.0 APPLICABILITY Unit 2 Revision 31 I 3.0-0 December 2008

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 31 1 3.0-1 December 2008

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

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

b. 2 x 10-4 RtCi/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 3.1.1-1 are maintained within the limits of DLCO 3.1.1.

Unit 2 Revision 31 I 3.1-1 December 2008

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

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

Batch Waste Release Grab Sample Each Batch (g) Each Batch (g) Principal 5 x 107jICi/ml Tanks (b) Gamma Emitters (c)

a. 2LWS-TK4A

b. 2LWS-TK4B 1-131 l x 10-6 lCi/ml

c. 2LWS-TK5A

d. 2LWS-TK5B Grab Sample One batch/31 31 days Dissolved and I x 10.' pýCi/ml days (g) Entrained Gases (gamma emitters)

Proportional Each batch (g) 31 days H-3 Ix 10-5 ptCi/ml Composite of grab samples Gross Alpha Ix 10-7 pCi/ml (d)

Proportional Each batch (g) 92 days Sr-89 5 x 10-' PCi/mI Composite of grab samples (d)

Sr-90 5 x 10-' p-Ci/ml Fe-55 Ix 10.6 PCi/mI

2. Continuous Releases Grab Sample 31 days (e) 31 days (e) Principal 5 x 10-7 pCi/ml Gamma

a. Service Water Emitters (c)

Effluent A

b. Service Water Grab Sample 31 days (e) 31 days (e) 1-131 1 x 10-6 pCi/ml Effluent B

c. Cooling Tower Blowdown Grab Sample 31 days (e) 31 days (e) Dissolved and I x 10-5 gCi/mI Entrained Gases (gamma emitters)

Grab Sample 31 days (e) 31 days (e) H-3 1 x 10-' tCi/ml Grab Sample 31 days (e) 31 days (e) Gross Alpha 1 x 10-7 ptCi/ml Grab Sample 92 days (e) 92 days (e) Sr-89 5 x 10-8 IlCi/ml Grab Sample 92 days (e) 92 days (e) Sr-90 5 x 10-8 IlCi/ml Grab Sample 92 days (e) 92 days (e) Fe-55 I x 10-6 p-Ci/ml

3. Continuous Release Grab Sample 31 days (f) 31 days (f) Principal 5 x 10-7 p-Ci/ml Gamma Auxiliary Boiler Emitters (c)

Pump Seal and Sample Cooling Grab Sample 92 days (f) 92 days (1) H-3 1 x 10.5 pCi/ml Discharge (Service Water)

Unit 2 Revision 31 I 3.1-2 December 2008

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

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

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

LLD (4 .6 6 )(Sb)

(E) (V) (2.22x 106) (Y) e -Et where:

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

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

E The counting efficiency (counts per disintegration),

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

2.22 x 106 = The number of disintegrations per minute per ptCi, Y The fractional radiochemical yield, when applicable,

)L The radioactive decay constant for the particular radionuclide (sec'), and At = The elapsed time between the midpoint of sample collection and the time of counting (secads).

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 beforethe-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 bdsolated, and then thoroughly mixed by the method described in Part II, Section 1.4 to assure representative sampling (c) The principal gamma emitters for which the LLD applies include the following radionuclides: Mn54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 106 .pCi/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 nulides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to Technical Specification 5.6.31i the format 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 tothe quantity of liquid waste discharged and in which the method of sampling employed results in a speimen that is representative of the liquids released.

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

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

(g) Complete prior to each release.

Unit 2 Revision 31 1 3.1-3 December 2008

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

a.

1.5 mrem to the whole body and < 5 mrem to any organ during any calendar quarter; and

b.

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

APPLICABILITY: At all times.

ACTIONS

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 31 I 3.1-4 December 2008

Liquid Effluents Dose D 3.1.2 ACTIONS (continued)

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

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

C. Required Action B.2 and C. 1 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 31 I 3.1-5 December 2008

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

Unit 2 Revision 31 I 3.1-6 December 2008

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

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

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

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

(1) An explanation of why AND liquid radwaste was being discharged without Projected doses due to the treatment, identification of liquid effluent, from the any nonfunctional unit, to UNRESTRICTED equipment or subsystems, AREAS would exceed and the reason for the 0.06 mrem 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) Summary description of Any portion of the liquid action(s) taken to prevent a radwaste treatment system recurrence.

not in operation.

Unit 2 Revision 31 I 3.1-7 December 2008

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

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

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

Unit 2 Revision 31 I 3.1-8 December 2008

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

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

< 3000 mrem/yr to the skin and

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

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. The dose rate(s) at or A. 1 Restore the release rate to Immediately beyond the SITE within the limit.

BOUNDARY due to radioactive gaseous effluents exceeds limits.

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

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

> 8 days in gaseous effluents shall be determined to be within the limits of DLCO 3.2.l.b.

Unit 2 Revision 31 I 3.2-1 December 2008

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 (LLD) (a)

TYPE

1. Containment (b) Grab Sample Each Purge (h) Principal Gamma Emitters I x 104 OPCi/ml (c)

Each Purge H-3 (oxide) I x 10-6 PCi/ml Each Purge Principal 1 x 10"4 ýCi/mI Gamma Emitters (c)

2. Main Stack, Grab Sample 31 days (d) 31 days (d) Principal Ix 104LOCi/ml Radwaste/Reactor Gamma Emitters Building Vent (c)

Grab Sample 31 days (e) 31 days (e) H-3 (oxide) lx 106jsCi/ml Charcoal Continuous (f) 7 days (g) 1-131 Ix 10"12 p.Ci/ml Sample Particulate Continuous (f) 7 days (g) Principal 1 x 101 p.Ci/mI Sample Gamma Emitters (c)

Gross Alpha Ix 10"- p-Ci/ml Composite Continuous (f) 92 days Sr-89 I x 10-" pCi/ml Particulate Sample Sr-90 I x 10" p.Ci/ml See the notes on the next page.

Unit 2 Revision 31 I 3.2-2 December 2008

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, abovcsystem background, that will be detected with 95% probability with only 5% probability of falsely concludig that a blank observation represents a "real" signal.

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

LLD = (4.66) (Sb )

(E)(V) (2.22x 106 ) (Y) e-?,At where:

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

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

E The counting efficiency (counts per disintegration),

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

2.22 x 106 = The number of disintegrations per minute per pCi, Y = The fractional radiochemical yield, when applicable,

= The radioactive decay constant for the particular radionuclide (se&l), and At = The elapsed time between the midpoint of sample colledion 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 beforethe-fact limit representing the capability of a measurementsystem and not as an after-the-fact limit for a particular measurement.

(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 theLLD applies include the following radionuclides: Kr87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, 1-131, Cs-134, Cs-137, Ce-141 and Ce-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 beanalyzed and reported in the Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3 in the format outlined in RG 1.21, Appendix B, Revision 1, June 1974.

(d) If the main stack or reactor/radwaste buildingradiation monitor is not FUNCTIONAL, sampling and analysis shall also be performed following shutdown, startup, or whoa there is an alarm on the offgas pretreatment monitor.

(e) H-3 grab samples shall be taken once every 7 days from the reactor/radwaste ventilation system when fuel is offloaded until sthle H-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 went exceeds its alarm setpoint, the iodine and particulate device shall be removed and analyzed to determine the changes in iodine and particulate release rates. The analysis shall bdone once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the release no longer exceeds the alarmsetpoint. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10.

(h) Complete prior to each release.

Unit 2 Revision 31 I 3.2-3 December 2008

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 REQUIRED ACTION COMPLETION TIME A. The air dose at or beyond A.1 Prepare and submit to the 30 days the SITE BOUNDARY NRC, pursuant to D 4.1 1, a due to noble gases released in gaseous 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 31 I 3.2-4 December 2008

Gaseous Effluents Noble Gas Dose D 3.2.2 ACTIONS (continued)

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

times the limits.

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 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 31 1 3.2-5 December 2008

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

Unit 2 Revision 31 1 3.2-6 December 2008

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

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

b. During any calendar year:

15 mrem to any organ.

APPLICABILITY: At all times.

ACTIONS 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 1-13 1, 1-133, A.1 Prepare and submit to the NRC, 30 days H-3 and radioactive material pursuant to D 4.1.1, a Special in particulate form with half- Report that lives > 8 days released in (1) Identifies the cause(s) for gaseous effluents at or 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 31 I 3.2-7 December 2008

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, etc.).

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

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

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

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

Unit 2 Revision 31 1 3.2-8 December 2008

Gaseous Effluents Dose 13 1, 1-133, H-3 and Radioactive Material in Particulate Form D 3.2.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.3.1 Determine cumulative dose contributions for the 31 days current calendar quarter and current calendar year for 1-131, 1-133, H-3 and radioactive material in particulate form with half-lives > 8 days.

Unit 2 Revision 31 1 3.2-9 December 2008

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. 1 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. 1 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 for 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 31 13.2-10 December 2008

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

Unit 2 Revision 31 13.2-11 December 2008

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. 1 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 D equipment to 1.0-1) would exceed 0.3 FUNCTIONAL status, and mrem to any organ of a (3) Summary description of MEMBER OF THE action(s) taken to prevent a PUBLIC. recurrence.

Unit 2 Revision 31 1 3.2-12 December 2008

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 31 1 3.2-13 December 2008

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 31 1 3.2-14 December 2008

Venting or Purging D 3.2.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i

DSR 3.2.6.1 The drywell and/or suppression chamber shall be Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months determined to be aligned for VENTING or PURGING before start of through the standby gas treatment system. VENTING or PURGING AND 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter during VENTING or PURGING Unit 2 Revision 31 1 3:2-15 December 2008

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. 1 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 31 I 3.3-1 December 2008

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME 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. 1 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. 1 Collect and analyze grab 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B. 1 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 itCi/ml.

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

Unit 2 Revision 31 I 3.3-2 December 2008

Radioactive Liquid Effluent Monitoring Instrumentation D 3.3.1 ACTIONS (continued)

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

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

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

AND Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter F. As required by Required F. 1 Estimate tank liquid level. Immediately Action B. 1 and referenced in Table D 3.3.1-1. AND During liquid additions to the tank 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 31 I 3.3-3 December 2008

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 31 1 3.3-4 December 2008

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 31 13.3-5 December 2008

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

Radioactive Liquid Effluent Monitoring Instrumentation 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 Radwaste Effluent (a) I C DSR 3.3.1.1 Line DSR 3.3.1.3 DSR 3.3.1.5 DSR 3.3.1.8

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

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

b. Service Water Effluent (a) I D DSR 3.3.1.1 Line B DSR 3.31.4 DSR 3.31.7 DSR 3.3.1.8

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

3. Flow Rate Measurement Devices

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

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

c. Service Water Effluent (a) I E DSR 3.3.1.2 Line B DSR 3.3.1.6 DSR 3.3.1.9

d. Cooling Tower (a) 1 E DSR 3.3.1.2 Blowdown Line DSR 3.3.1.6 DSR 3.3.1.9

4. Tank Level Indicating (b) I 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 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 31 I 3.3-6 December 2008

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

1. LCO 3.0.3 is not applicable.

2. Separate condition entry is allowed for each channel.

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

required.

OR A.2 Declare the channel Immediately nonfunctional.

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

(continued)

Unit 2 Revision 31 1 3.3-7 December 2008

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (continued)

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

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

C. As required by Required C. 1 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 31 I 3.3-8 December 2008

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. As required by Required D.1 Estimate the flow rate for the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action B. 1 and referenced 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 Continuously collect samples 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Action B. 1 and referenced using auxiliary sampling in Table D 3.3.2-1. equipment as required in Table D 3.2.1-1.

F. As required by Required F.1.1 Take grab samples. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action B. 1 and referenced 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 limit of detection of at least completion 1 x 10-4 tCi/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, determine: 14 days (1) The cause(s) of the nonfunctional.

(2) The actions to be taken and the schedule for restoring the system to FUNCTIONAL status.

(continued)

Unit 2 Revision 31 1 3.3-9 December 2008

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 ACTIONS (continued)

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

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

or F. I not met.

Unit 2 Revision 31 13.3-10 December 2008

Radioactive Gaseous Effluent Monitoring Instrumentation D 3.3.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.2.1 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 alarm/trip setpoint (each channel will be tested independently so as to not initiate isolation during operation); and control room alarm annunciation occurs for instrument indication levels measured above the alarm setpoint, circuit failure, instrument indicating a downscale failure, and instrument controls not set in operate mode.

DSR 3.3.2.5 Perform CHANNEL FUNCTIONAL TEST. 92 days 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 downscale failure, and instrument controls not set in operate mode.

(continued)

Unit 2 Revision 31 13.3-11 December 2008

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

SURVEILLANCE FREQUENCY i

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 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.

DSR 3.3.2.10 Perform CHANNEL CALIBRATION. 24 months Unit 2 Revision 31 I 3.3-12 December 2008

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

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

1. Offgas System

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

- Providing DSR 3.3.2.7 Alarm and Automatic Termination of Release

b. System Flow- (a) I D DSR 3.3.2.1 Rate Measuring DSR 3.3.2.5 Device DSR 3.3.2.10 (a) 2 D DSR 3.3.2.1

c. Sample Flow- DSR 3.3.2.5 Rate Measuring DSR 3.3.2.10 Device

2. Radwaste/Reactor Building Vent Effluent System

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

b. Iodine Sampler (b) I E DSR 3.3.2.2

c. Particulate (b) I E DSR 3.3.2.2 Sampler

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

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

(a) During offgas system operation.

(b) At all times.

(c) Includes high range noble gas monitoring capability.

Unit 2 Revision 31 13.3-13 December 2008

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

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

b. Iodine Sampler (b) E DSR 3.3.2.2

c. Particulate (b) E DSR 3.3.2.2 Sampler

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

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

(c) Includes high range noble gas monitoring capability.

Unit 2 Revision 31 1 3.3-14 December 2008

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

APPLICABILITY: At all times.

ACTIONS

NOTE I'% ------------------------------------------------------

1. LCO 3.0.3 is not applicable.

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Estimated dose or dose A. 1 Verify the condition resulting in Immediately commitment due to direct doses exceeding these limits has radiation and the release of been corrected.

radioactive materials in liquid or gaseous effluents exceeds the limits.

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

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

30 days Submit a Special Report, 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 31 I 3.4-1 December 2008

Radiological Environmental Monitoring Program D 3.5.1 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.1 Prepare and submit to the NRC In accordance with Monitoring Program not in the Annual Radiological the Annual conducted as specified in Environmental Operating 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. -------- NOTES-------

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

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

(continued)

Unit 2 Revision 31 I 3.5-1 December 2008

Radiological Environmental Monitoring Program D 3.5.1 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 of D 3.1.2, D 3.2.2, or OR D 3.2.3.

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

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

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

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

Unit 2 Revision 31 1 3.5-2 December 2008

Radiological Environmental Monitoring Program D 3.5.1 ACTIONS (continued)

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

D 3.5.1-1.

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

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

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

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

Unit 2 Revision 31 1 3.5-3 December 2008

Radiological Environmental Monitoring Program D 3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION ITIME E. Samples required by Table E.1 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 31 I 3.5-4 December 2008

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

Unit 2 Revision 31 I 3.5-5 December 2008

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

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

1. Direct 32 routine (1) An inner ring of stations, Once per 3 months Gamma dose: once per 3 Radiation monitoring one in each 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. Airborne 5 locations (1) 3 samples from offsite Continuous sampler Radioiodine canister:

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 r 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months round community (e) loading (1).

following filter change (2) Perform gamma isotopic (3) 1 sample from a control analysis on each sample location, at least 10 miles (g) in which gross beta distant and in a least 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. Waterborne

a. Surface 1 sample Upstream (d) (h) Composite sample (1) 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 (3) Gamma isotopic

b. Ground As required From one or two sources if Grab sample once analysis of each sample likely to be affected (j) per 3 months (g) once per 3 months (4) H-3 analysis of each sample once per 3 months (continued)

Unit 2 Revision 31 1 3.5-6 December 2008

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

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

3. Waterborne (continued)

c. Drinking 1 sample of each One to three of the nearest When 1-131 analysis (1) 1-131 analysis on each water supplies that could be is performed, a composite sample affected 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 (1)

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

d. Sediment 1 sample From a downstream area with Twice per year Gamma isotopic analysis of from existing or potential recreational each sample (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 December (2) If there are In each of 3 areas 3.5-5.0 miles (2) Gamma isotopic (g) and none, then 1 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 least prevalent 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 31 1 3.5-7 December 2008

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

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

4. Ingestion (continued)

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

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

Unit 2 Revision 31 1 3.5-8 December 2008

Radiological Environmental Monitoring Program D 3.5.1 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, and additional descriptions where pertinent, shall be provided for each and every sample location in Table D 3.5.11. Refer to NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainablebecause of such circumstances as hazardous conditions, seasonal unavailability (which includes theft and uncooperative residents), or malfunction of automatic sampling equipment.

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

(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 groundlevel D/Q based on all site licensed reactors.

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

(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 area beyond but near the mixing zone.

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

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

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

(1) Analysis for 1-131 may be accomplished by Ge-Li analysis provided that the lower limit of detection (LLD) for 1131 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 1131 is detected in November and December of the preceding year.

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

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

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

Unit 2 Revision 31 I 3.5-9 December 2008

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 (pCi/L) GASES (pCi/m3 ) (pCi/kg, wet) (pCi/L) (pCi/kg, wet)

H-3 20,000 (a)

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-95 400 Nb-95 400 1-131 2 (b) 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-140 200 300 La-140 200 300 (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 31 I 3.5-10 December 2008

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

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

LOWER LIMIT OF DETECTION (LLDIc)

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

Gross Beta 4 0.01 H-3 2,000 (d)

Mn-54 15 130 Fe-59 30 260 Co-58 15 130 Co-60 15 130 Zn-65 30 260 Zr-95 15 Nb-95 15 1-131 1 (0) 0.07 I 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-140 15 15 La-140 15 15 See the notes on the next page Unit 2 Revision 31 13.5-11 December 2008

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

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

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

(b) Required detection capabilities for thermoluminescent dosimetersused for environmental measurements are given in ANSI N-545, Section 4.3 1975. Allowable exceptions to ANSI N545, Section 4.3 are contained in the ODCM.

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

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

4 66 LLD ( . )(Sb)

(E) (V) (2.22) (Y) e- xA 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 radioactive decay constant for the particular radionuclide (sec'), and At = The elapsed time between environmental collection or end of the sample collection period, and the time of counting (seconds).

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

It should be recognized that the LLD is defined as a beforethe-fact limit representing the capability of a measurement system and not as an after-the-fact limit for a particular measurement. Analyses shall be performed in such a manner that he stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

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

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

Unit 2 Revision 31 1 3.5-12 December 2008

Land Use Census D 3.5.2 D 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 sampling controlled by Table D 3.5.1-1, part 5.c may be performed in lieu of the garden census) > 500 ft2 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 31 I 3.5-13 December 2008

Land Use Census D 3.5.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Land use census identifies B.1 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 same 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 commitment(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 31 1 3.5-14 December 2008

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 31 13.5-15 December 2008

Interlaboratory Comparison Program D 3.5.3 D3.5 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

1. LCO 3.0.3 is not applicable. NOTES ----------------------------

2. LCO 3.0.4 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Analyses not performed as A. 1 Report the corrective actions In accordance with required. taken to prevent a recurrence the Annual to the NRC in the Annual Radiological Radiological Environmental Environmental Operating Report. Operating Report Unit 2 Revision 31 I 3.5-16 December 2008

Interlaboratory Comparison Program D 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.3.1 Report a summary of the results obtained as part of the In accordance with Interlaboratory Comparison Program in the Annual the Annual Radiological Environmental Operating Report. Radiological Environmental Operating Report Unit 2 Revision 31 I 3.5-17 December 2008

PART I - RADIOLOGICAL EFFLUENT CONTROLS BASES Unit 2 Revision 31 I B 3.1-0 December 2008

Liquid Effluents Concentration B 3.1.1 BA3.1 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(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its effluent concentration in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

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

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

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

Unit 2 Revision 31 IB 3.1-1 December 2008

Liquid Effluents Dose B 3.1.2 B 3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.2 Liquid Effluents Dose BASES This is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I to 10 CFR 50. This implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in liquid effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable.

Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the potable drinking water that are in excess of the requirements of 40 CFR 141. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside storage tanks, etc., are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBERS OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. The dose calculation methodology and parameters implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by Calculational procedures based on models and data, so that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

The equations specified for calculating the doses that result from actual release rates of radioactive material in liquid effluents are consistent with the methodology provided in RG 1.109, "Calculation of Annual Doses To Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and R.G. 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977. This applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

Unit 2 Revision 31 I B 3.1-2 December 2008

Liquid Radwaste Treatment System B 3.1.3 B3.1 RADIOACTIVE LIQUID EFFLUENTS B 3.1.3 Liquid Radwaste Treatment System BASES The installed liquid radwaste treatment system shall be considered 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 requirements of 10 CFR 50.36a, GDC 60 of Appendix A to 10 CFR 50 and the design objective given in Section II.D of Appendix I to 10 CFR 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I to 10 CFR 50 for liquid effluents. This applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared radwaste treatment systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

Unit 2 Revision 31 I B 3.1-3 December 2008

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 toensure that the dose rate at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR 20 to UNRESTRICTED AREAS.

The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table 2, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR 20 or as governed by 10 CFR 20.1302(c). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in Part II. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the whole body or to less than or equal to 3000 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 31 I B 3.2-1 December 2008

Gaseous Effluents Noble Gas Dose B 3.2.2 B 3.2 RADIOACTIVE GASEOUS EFFLUENTS B 3.2.2 Gaseous Effluents Noble Gas Dose BASES This is provided to implement the requirements of Section II.B, III.A, and IV.A of Appendix I to 10 CFR 50. The DLCO implements the guides set forth in Section II.B of Appendix I. The REQUIRED ACTIONS provide the required operating flexibility and, at the same time, implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept as low as is reasonably achievable. The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guidelines of Appendix I be shown by calculational procedures based on models and data so that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units including outside storage tanks, etc., are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 5 miles must be considered. The dose calculation methodology and parameters for calculating the doses from the actual release rates of radioactive noble in gaseous effluents are consistent with the methodology provided in RG 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977, and RG 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1," July 1977. The ODCM equations provided for determining the air doses at or beyond the SITE BOUNDARY are based upon real-time meteorological conditions or the historical average atmospheric conditions.

This applies to the release of radioactive material in gaseous effluents from each unit at the site.

Unit 2 Revision 31 I B 3.2-2 December 2008

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, III.A, and IV.A of Appendix I to 10 CFR 50. The DLCO implements the guides set forth in Section II.C of Appendix I. The REQUIRED ACTIONS provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept as low as is reasonably 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 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 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 Unit 2 Revision 31 I B 3.2-3 December 2008

Gaseous Effluents Dose -Iodine-131, Iodine-133, Tritium, and Radioactive Material In Particulate Form B 3.2.3 B 3.2.3 Gaseous Effluents Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material In Particulate Form (continued) with subsequent consumption by man, (3) deposition onto grassy areas where milk-producing animals and meat-producing animals graze (human consumption of the milk and meat is assumed), and (4) deposition on the ground with subsequent exposure to man. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

Unit 2 Revision 31 I B 3.2-4 December 2008

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 II.B and II.C of Appendix I to 10 CFR 50, for gaseous effluents. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportional among the units sharing that system.

Unit 2 Revision 31 I B 3.2-5 December 2008

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 this system be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable. This implements the requirements of 10 CFR 50.36a, GDC 60 of Appendix A to 10 CFR 50, and the design objectives given in Section II.D of Appendix I to 10 CFR 50. Limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I to 10 CFR 50, for gaseous effluents. This applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportional among the units sharing that system.

The appropriate components, which affect iodine or particulate release, to be 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 31-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 31 I B 3.2-6 December 2008

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 31 I B 3.2-7 December 2008

Radioactive Liquid Effluent Monitoring Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Radioactive Liquid Effluent Monitoring Instrumentation BASES The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Part II to ensure that the alarm/trip will occur before exceeding ten times the limits of 10 CFR 20. The 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 31 I B 3.3-1 December 2008

Radioactive Gaseous Effluent Monitoring Instrumentation B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Radioactive Gaseous Effluent Monitoring Instrumentation BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm/trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Part II to ensure that the alarm/trip will occur before exceeding the limits of 10 CFR 20. Although the Offgas System Noble Gas Activity Monitor is listed in Table D 3.3.2-1, "Radioactive Gaseous Effluent Monitoring Instrumentation", these monitors are actually located upstream of the Main Stack noble gas activity monitor and are not effluent monitors. They were included in Table D 3.3.2-1 in accordance with NUREG-0473. As such, Offgas System Noble Gas Activity Monitor alarm and trip setpoints are not based on I 0CFR20. The offgas system noble gas monitor alert setpoint is set at 1.5 times nominal full power background to assure compliance with ITS SR 3.7.4.1 which requires offgas sampling be performed within four hours of a 50% increase in offgas monitoring readings, and to support MSLRM trip removal. The offgas system noble gas monitor trip setpoint is based on the 10CFRI 00 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 31 I B 3.3-2 December 2008

Radioactive Effluents Total Dose B 3.4 B 3.4 RADIOACTIVE EFFLUENTS TOTAL DOSE BASES This is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46 FR 18525. This requires the preparation and submittal of a Special Report whenever the calculated doses from releases of radioactivity and from radiation from uranium fuel cycle sources exceed 25 mrem to the whole body or any organ, except the thyroid (which shall be limited to less than or equal to 75 mrem). If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR 20, as addressed in 3.1.1 and 3.2.1. An individual is not considered a 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 31 I B 3.4-1 December 2008

Monitoring Program B 3.5.1 B 3.5 RADIOLOGICAL ENVIRONMENTAL MONITORING B 3.5.1 Monitoring Program BASES The Radiological Environmental Monitoring Program provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR 50 and thereby supplements the Radiological Effluent Monitoring Program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. Program changes may be initiated based on operational experience.

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

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

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

Unit 2 Revision 31 I B 3.5-1 December 2008

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

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

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

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 31 I B 3.5-2 December 2008

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 31 IB 3.5-3 December 2008

PART I - RADIOLOGICAL EFFLUENT CONTROLS SECTION 4.0 ADMINISTRATIVE CONTROLS Unit 2 Revision 31 1 4.0-0 December 2008

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, "Annual Radiological Environmental Operating Report," Section 5.6.3, "Radioactive Effluent Release Report," and Section 5.5.1, "Offsite Dose Calculation Manual."

Unit 2 Revision 31 1 4.0-1 December 2008

Special Reports D4.1.1 D 4.1.2 D 4.1.3 D 4.1 REPORTING REQUIREMENTS D 4.1.1 Special Reports Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.

D 4.1.2Annual 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 of joint frequency distribution of wind speed, wind direction, and atmospheric stability.

In lieu of submission with the Radiological Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

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

(Continued)

Unit 2 Revision 31 1 4.1-1 December 2008

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 31 I 4.1-la December 2008

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 supplemental information;

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

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

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

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

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

Unit 2 Revision 31 1 4.2-1 December 2008

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 31 1 4.2-2 December 2008

ATTACHMENT 13 RADWASTE PROCESS CONTROL PROGRAM (RPCP)

NINE MILE::POINT NUCLEAR STATION UNIT 2 UNIT 2 RADWASTE PROCESS CONTROL PROGRAM REVISION 07 TECHNICAL SPECIFICATION REQUIRED Approved by:

S. L..Belcher Date THIS IS AFULL REVISION Effective Date: 6/'42n6

TABLE OF CONTENTS.

SECTION PAGE 1.0 PURPOSE .......................... . . ...................

2.0. RESPONSIBILITIES. .. ........................ .................... 1 3.0 PRO G RA M.. . ................................. ........... ........ ........ . ........ ....... 1 3 .1 S yste m De scriptio n ....................................................................................... ............................... 1 3.2 Radioactive Waste Dewatering System ..................................... 3 3.3 Disposition of other Radioactive Material .............................. 8.......

8 3 .4 n............

Sa mp ling ..... ........ e. .................... .. .......................... .... ........... ................... 99....

3.5 Waste Classification..... ........................ ..........oq-9 3.6 Adm inistrative Controls ............................................................................................................... 10 4.0 DEFINITIONS........ ..................... ..................... 12 5.0D :REFER ENC ES ........ ..... ..... ... . ....

.... :-...... . ........... .... ........ .... .............. :;.13 ATTACHMENT 1: UNIT 2 RADWASTE PROCESS-CONTROL PROGRAMREFERENCE AND IMPLEMENTING PROCEDURES......... ...................... 15 ATTACHMENT 2: SOLID WASTE SOURCES..; ... .. ....... ....... ..... ........................... 16 Page i Rev 07

1.0 PURPOSE 1.1 To describe the methods for processing, packaging and transporatio'n of low-level radioactive waste and 'provide assurance of complete stabilization of various radioactive "wet wastes" in.accordancewith applicableMNRCOregulations and guidelines..

1.2 To satisfy the Nuclear Regulatory Comnmission's Low-Level Waste-and LUanium Recovery Projects Branch (WMLU) requirement and establish process parameters within which the vendor supplied Rapid Dewatering System must be operated to meet current disposal criteria at low-level waste, disposal facilities..

NOTE: Conformance with WMLU requirements provides assurancethat the requirements identified.in 10CFR61., Subpart D, Technical Requirements for Land.Disposal Facilities, and Final Waste Classification are satisfied.

2.0 RESPONSIBILITIES 2.1 The Plant General Manager is responsible for:

2.1.1 Ensuring the Unit 2 Radwaste Process Control Program provides forzthe health and safety of the general public as it applies to* Radwaste Management.

2.1.2 Reviewing and approving changes.to the: Unit 2 Radwaste Process Control Program.

2.2 The Manager of Operations is responsible for the content and maintehance of.this

.procedure.

2.3 The Supervisor Radioactive: Materials Processing, is responsible for overall implementation ofthe Radwaste Process Control Program.

2.4 Operators are responsible toprocess and package wastes in accordance with-applicable Waste Handling Procedures.

3.0 PROGRAM 3.1 System Description 3.1.1 General a.. The. Solid Waste Management System (SWMS) .isimplemented by the procedures identified in the Unit 2 Radwaste Process Control Program Implementing Procedures (Attachment 1). SWMS describes the collection Volume reduction, dewatering, or solidification and packaging of wet and dry types of radioactive Waste in preparation fo- shipment off-site for further processing or disposal at a licensed processing or burial site. The processing and storage methods used for interim storage are consistent with the present.waste form stability requirements; Page 1 Rev 07

3.1.1 (Cont)

b. Types of solid waste sources are identified in Solid Waste Sources.

(Attachment2)..

c. TheSolid Waste Management System accommodates dry solid trash which is either compacted With a trash compactor (when physically

.possible) or-sent~offsite for separation and processing.

NOTE: When required, Unit 2 will use the services of a vendor to solidify, dewater, separate, recover, or incinerate waste,

d. Bead resins, powdered resins and charcoal are dewatered using a Dewatering System in-vendor certified polyethylene containers or'High Integrity Containers (HIC) ifgoing for, burial.

e. Bead resins .powdered resins, and charcoal, are sent..to a vendor for volume reduction using vendor approved containers.

f. Concentrated wastes~are processed offsite by an Approved vendor.

3.1.2 Condensate Filtration System Phase Separator

a. The Condensate Filtration.System Phase Separator tank may be decanted to the. Floor Drain Collector System.

b. Contents of the tank may be recirculated, or transferred to a liner in the.

Radwaste Truck bay forfurther offsite processing.

3.1.3 Waste-Sludge Tank

a. The waste sludge:tank is supplied with waste from the following sources:.

1. Radwaste filters

2. The Thermex System

3. Spent Resin Tank

b. The waste sludge, tank has the:ability for decantation. A decant pump takes a suction off the sludge tank and discharges to the spent resin tank.

The tankcan also be gravity decanted manually to the floor drains.

Page 2 Rev 07

3.1.3 (Cont)

c. Contents of the wastesiudge tank are transferred by one of two redundant waste sludge. pumps, to the Radwaste Truck bay for dewatering by an approved vendor using dewatering System or cement solidification.

3.1.4 Ventilation System The Radwaste Building Ventilation System (HVW) provides filtered, conditioned outside air to various areas of the.Radwaste Building and exhausts the air to the atmosphere'through the Reactor Building Ventilation. The HVW system maintains the building at a pressure, below atmospheric to help. prevent any unmonitored air leakage to the environment.

3.1.5 Liners

a. Dewatering System is compatiblewith vendor supplied dewatering waste, containers.

b. These containers and their dewatering internals are designed to ensure uniform dewatering of waste slurries., They are fabricated and inspected

.inaccordance with a vendor approved Quality Assurance Program -and

.are compatible with the waste they are designed to contain.

c. Selection of liner typeAwiHl be determined by waste classification requirements; 3.1.6 Crane

a. Liner movements normally are completed using a. radio controlled remote operated crane. Linear movements. may be completed using the pendent controller.

b.. When liners stored in the Radwaste Building storage area are to be shipped, they are loaded using the-crane for transportation to a processing or burial facility.

3.2 Radioactive Waste Dewaterincg System 3.2.1 DewaterinQ System.

a.. The dewatering system is a self-contained, free-standing portableisystem for dewatering radioactive spent resins and filter sludge's in a variety of liners to meet current disposal criteria at low-level waste disposal facilities.

The system is comprised of:

Page 3 Rev 07

3.2.i.a (Cont)

1. A plant connection skid

2. A container fillhead, complete with:interconnecting hoses and cables

3. Adewatering skid

4. Acontrol panel

5. Awaste container.

b. The radioactive waste slurry istransferred by waste transfer pumps to the Dewatering System..

c. The water removed from the radioactive waste is pumped from,.the waste liner by a dewatering pump through a media-specific filtering device and returned to. a plant floor drain.

d. Fill operation. is controlled remotely and viewed with a video monitor on the control panel. A remote level-control system detects and monitors waste level in the1liner and provides overfill protection. An independent level control system in the fillhead provides redundant overfill protection.

e. Upon completion-of dewatering, warm air between 180-195 deg. F is recirculated through the liner and moisture separator until water content of the waste, is within the low-level burial site Acceptance Criteria.

NOTE: The limiting factor on air temperature recirculated through the liner is based on.maximum allowable temperature of aHIC. The maximum measured acceptable temperature is 200 deg. F.

f. The type of' media which: can be:.dewatered by the Dewatering System is divided.into two. categories:

1. Granular media which includes bead resin, charcoal, and zeolites

2. Filterprecoat media which includes ecodex, powdex, ecosorb, eacocoat, and diatomateous earth,

g. All discharge air is passed through HEPA filtration units contained Within the. Dewatering Skid before passing to permanent plant ventilation.

Page 4. Rev 07

3.2.2 Acceptance Criteria Acceptance Criteria for process completion is established by a minimumn dewatering time.and a maximum water collection rate. The resultant. waste form meets the requirements of 10 CFR 6&1Licensing Requirements for Land*Disposal of Radioactive Waste"' and NRC Branch TechinicalPosition on, WasteForm (May, 1983.Rev 0).

a. Bead Resin Type Liners

1. The dewatering pump. has run for one houir or until unable to maintain 1.6" vacuum,after the final waste transfer.

2, The Dewatering System has been runfora minimum of four hours.

.3, Thermoisture separator sight glass level does not increase more than 1/2'inch during a thirty minute period;

b. Precoat Media Type Liners

1. The dewatering pump has run, for one hour or until unable to maintain 22.' vacuum after !fina!iwaste transfer.

2. The Dewatering System has been run for a minimum Of eleven hours.

3. The moisture separator sight glass level does. notOncrease more than 1 inch during a thirtyminute period.

3.2.3 Plant ConnectionlSkid The plant connection skid consists of the following:

a-.. Aremotely operated waste inlet control.valve to.regulate influent to the liner. This valve is interlocked to close on High Level, High High: Level (mechanical float inside fillhead), and decreasing: air pressure or loss of electrical power.

b. Adiaphragm pump with connections to the fillhead for gross initial dewatering.

c. Manifold for air and water supplies to control valves and to flush components.

Page.5 Rev: 07

3.2:4 Fillhead

a. Camera and light provides remote visual observation of the container level during the resin transfer and dewatering.

-b. Connections on the underside of the.fillhead connect to break away fittings to facilitate:remote removal from the container for ALARA.

c, The external connections on the fillhead are camlock, with the, exception of the waste inlet.

d. A.float switch inside the fillhead provides redundant liner level detection.The float switch provides automatic closure.of the waste isolation valve on high high level.

ý32.5 Dewatering Skid The Dewatering. Skid consists of avacuum pump, moisture separator; air conditioning unit, and piping interface to the plant connection stand. Pressures and temperatures are monitored at various points on this component to safeguard mechanical operations. AHEPA filter is installed downstream of the safety relief and manual bypass valves.

3.2.6 Control Panel Acontrol panel containing electrical and pneumatic controls to allow remote operation of all components and monitoring of individual parameters. Avideo monitor of the liner is.provided aswell as~temperature and pressure indications of primary components. Audible. and visual alarms to indicate, off-rnormal conditions are also found on the control panel.

3.2.7 Waste Containers

a. Waste Containers used for dewatering satisfy: stability requirements.

1. Polyethylene container may be used as the disposal package for NRC Class "A"waste.

2. Polyethylene container may also be used for NRC Class "B"and "C"waste, but enhanced structural stability is required for burial at the Barnwell site.

NOTE: The enhanced structural stability required for 10CFR61.56 and.the State of South Carolinas accomplished by the use of DHEC approved concrete overpack structures at the Barnwell burial site.

b. Each Waste Container is accompanied by a certificate of compliance.

Page 6 Rev 07

312.7 (Cont)

C. Dewaiering procedures will be based on an NRC approved vendor process control program or Topical Repor and are part of Dewatering System Procedures.

d. Nopolyethylenecontainer is stored indirect sunlight for a period greater than one year.

e. Waste containers used to transport concentrated waste are Compatible with the type of waste they are designed to contain.

f. Containers are protected from overfiUl by a level detection system. FAVA is the manufacturer's designation for a leveldetection system which is installed in the liner with a remote readout. display on the control panel.

Four probes inserted at different levels in the linerwork on the conduction principlejto determine the level of waste in the:container.

32.18 Operators Operators shall ensure proper equipment is available before beginning radwaste:

processing. Operators may process wastes when the following equipment is operable:

a. Closed circuit television system stations

b. Radwaste Building Ventilation

c. Radwaste:Building.Floor Drain System

d. Radwaste Building CNS System

e. Service Air System

3.2.9 Vendor Operators All operations of the Dewatering System shall be performed by techniciansloperators that have successfully. completed the Vendor training program, The technician/operator shall have practical experience and certification on the Dewatering System. The technicianloperator is subjected to recertification; every two years.

3.2.10, Quality Assurance Vendor approved Quality Assurance Program, and Vendor QA Procedures, shall be employed to control.the design, fabrication, inspection, testing, operation, and record keeping for the Dewatering System.

Page .7 Rev 07

3.2.11 Records The Vendor maintains records of the design, fabrication and testing of each Dewaterinpg System. The setup and operation of the systerh is maintained in accordance with vendor supplied procedures.

3.3 Disposition of other Radioactive Material 3.3.1 Contaminated Fluids

a. Contaminated fluids are stored in containers at designated areas within the plant.

b. A vendor with an approved, process control program acceptable at the selected processing or burial site is used to dispose of contaminated fluids..

c. A vendor may also be usedlto incinerate oils.

3.3.2 Temporary Radwaste Processing

a. Vendors are NRC:approved and have demonstrated:a comm'itment to 10CFR61., Subpart D,Technical.Requirements.for Land. Disposal.Facilities and Final Waste Classification and Waste Form. Technical Position Papers stability requirementsl

b. Vendors have completed Class B and'Cwaste. testing or have.provideda schedule of completion.

c. Vendors have approved procedures to process Class A waste (Dewatering, Evaporation, ahd Solidification):

d. Vendor procedures are Reviewed and approvedin accordance with NIP-PRO-03, Preparation and Review of Technical Procedures.

3.3.3 Dry Active Waste (DAW)

a. The proper and safe steps are performed to collect and prepare low specific activity (LSA) DAW in accordance with N2-WHP-:1 2, Solid Dry Waste Collection and Compaction and N2-WHP-4, Waste Transfer Procedure.

b., DAW is examined for liquids or items that would compromise the irtegrity of the package or violate the burial site license andlor criteria before compacting. These liquids or items are removed or separated.

Page 8 Rev 07

.3.3.3- (COnt)

c. DAW is shipped in containers meeting the: transport requirements of 49CFR 173.427, Transport Requirements. for Low Spenific Activity (LSA)

Radioactive Materials;, -and any additional >vendor requirements, if specified.

dI Waste precluded from: disposal in LSA boxes or drumsdue toradiation limits is:disposed of in liners in accordance with WHPs, Waste Handling Procedures.

3.4 Sampling

a. The Condensate Prefilter/ Phase Separator Tank (TKIfO), the Waste Sludge. Tank (TK8), and the Spent Resin Tank (TK7) are isolated from further inputwhen preparing to process waste and a batch number is.assigned.

b. The Condensate Prefilter/Phase Separator Tank (TK10) and the Spent!Resin Tank (TK7) are recirculated to ensure a homogeneous mixture..

C.. The Waste Sludge Tank (TK8). is agitated to ensure a homogeneous mixture.

d. Asample is obtained from the tank(s) to be processed in accordance with N2-WHP-4, Waste Transfer Procedure.

e. The sample from the tank(s) to be processed is analyzed by Chemistry-and the sample data sheet form in N2-CSP-WSS-@406, Dewatered Waste:Surveillance at Unit 2, is completed.

3.5 Waste Classification

a. The Unit 2 Radwaste Process Control Program, procedure assures that wastes determined acceptable fornear surface disposal are. properly classified.

b. Waste classification is performed:consistent with the guidance provided in the Branch Technical Position pertaining to Waste Classification and is based upon the concentration of certain radionuclide's in the waste form ýas given in 100FR61.55, Waste Classification, and 10CFR61..56, Waste Characteristics.

NOTE: The methods used ýand the frequency for.determining the radionuclide concentration of the final waste form are conducted in accordance with N2-CSP-WSS-@406, Dewatered Waste Surveillance at Unit 2.

c. Classification of waste is performed in accordance with S-WHP-03, Classification and Shipment of Radioactive Material, using the RADMAN computer code or S-WHP-04, Classification and Shipment of Radioactive Material, using the RAMSHP computer program.

Page 9 Rev 07

3.6 Administrative Controls NOTE: The Director of Quality and Performance Assessment ha-s the authority to stop work when significant conditions adverse to quality exist and require corrective action.

3.6.1 Quality Assurance (QA).procedures and the Nuclear QA. Program require:

a. Ongoing review, monitoring, and audit functions.

b. Performance of audits of the Process Control Program and implementing procedures [or processing and: packaging of radioactive waste at least once every 24 months.

c. Compliance with the waste classification and characterization.

requirements of 10CFRI61.55, Waste Classification and 10CFR61.56, Waste Characteristics,

d. Quality Assurance Inspectors performing radwaste inspections have documented training in Department of3Transportation and NRC radwaste

regulatory requirfements.

e. Quality Assurance review of vendor programs to ensure compliance with i0CFR1I, Packaging.and Transportation of Radioactive.Materials, Quality Assurance requirements.

3.6.2 Training:Procedures and Training Programs require:

a. Operator qualification by completion of the. Operations Unit 2 Plant Training Program including:

1. On-the-job training in conjunction with classroom instruction to ensure each operator demonstrates an acceptable revel of skill and familiarity associated with.radwaste controls and operational procedures.

2. Continuing Training in accordance with approved training procedures.

b. Training records to:

1. Be maintained for audit. and inspection purposes:

2. Be considered permanent records,.

Page 10 Rev 07

3.6.2.b (Cont)

3. Meet the applicable requirements of the Quality Assurance Topical Report Section B.15 Records' CNG-TR-1.01M1000 Conduct of Training and CNG-PR-ý3.01 -1000 Records

.Manlagement.

3.6.3 DocumentationControl and Record Retention

a. Station management shall evaluate QA program audits of waste classification records to satisfy .tl-e requirements of 10CFR20.2006.d, Transfer for Disposaland Manifests.

b. Personnel shall process changes to operating procedures in accordance with NIP-PRO series.

c. Site Records Management shall maintain waste management records in accordance with the appropriate administrative procedures.

3.6.4 Licensee-initiated changes to the Unit 2 RadwasteProcess Control Program:

a. Aresubmitted to the Commission in the Radioactive Effluent.Release Report;forthe period in which the change(s) was made, and contain the.

'information required by.USAR Section 11.4.7, Process Control Program.

b. Become effective upon"1review and acceptance:by the Plant Operations Review .Committee.:

3.6.5 The Supervisor Radioactive Materials Processing shall ensure:

a. Shi pping manifests are completed and tiracked to satisfy-the requirements of 10CFR20, Transfer for Disposal and Manifests, in accordance with Waste Handling Procedures.

b. Temporary storage of solid radioactive material awaiting shipment.in an area other than a designated area is done in accordance with :GAP-INV-02, COntrol of Material Storage Areas.

3;6.61 Solid Radioactive Wastes Specification

a. Technical Requirements Manual (TRM) Specification 3.11.1 contains the requirement to solidify or dewater wastes to meet shipping, transportation.

and disposal site requiiements. Required actions and Icompletion times associated with failure to meet the requirements are also contained in TRM 3.11.1.

b. TRM 3.11.1 is a part of the Process Control Program and is therefore subject to the'same controls and change processes as the POCP.

Page 11 Rev 07

4.0 DEFINITIONS 4.1 Class "A" Waste Waste usually segregated from other waste classes atithe disposal site. The physical form and characteristics shall meet the minimum requirements of 10CFR61.56; Waste Characteristics.

4.2 Class "B" Waste, Waste meeting more rigorous waste form requirements to risure stability after disposal.

Class B waste form shall meet both.the minimum~and stability requirements of I OCFR6I,56, Waste Characteristics.

4.3 Class "C" Waste Waste meeting Class B standards and requiring additional measures at the disposal acility to prevent inadvertent intrusion.

4.4 Homogeneous Of the same kind :or nature; essrentiallyalike. Most waste streams are considered homogeneous for purposes of waste classification.

4.5 Batch An isolated quantityof feedwaste to be processed having' essentially constant physical and chemical characteristics.

4.6 Dewatered.Waste Refers to waste that has been processed by means other than solidification;,

encapsulation, or absorption to meet the free standing liquid requirements of 10 CFR 61.56 (a)(3) and .(b)(2).

4.7 ConcentratedWaste Liquid waste that has a high level of dissolved and/or particulate solid content.

4.8 Dried Waste Solid waste that has been processed by evaporation to dryness.

Page 12 Rev 07

5.0 REFERENCES

5.1 Licensee Documentation 5.1,1 QualityAssurance Topical Report, Sectibn B.15 Records 5 .1.2 Unit 2 Technical Specifications Section 5.6.3, Radioactive Effluent Release Report 5.1.3 Nuclear Quality Assuiance Program 5.1.4 Unit 2 Updated Safety Ana!ysis Report Section 11.4.7, Process Control Program 5A..5 Unit 2Technical Requirements Manual Specification 3.11.1, Solid Radioactive wastes 5.2 Standards, Regqulations, and Codes 5:.2.1 ANSI/ANS 55.1, 1979, American Nat onal. Standard for So idRadioactiVe Was te Processing System for Light Water Cooled Reactor Plants 5.2.2 10CFR20.2006.d, Transferfor Disposal and ManifeSt 5.2.3 10CFR20 App G, Requiremehts for Transfers of Low Level Radioactive Waste intended for Disposal at Licensed Land Disposal. Facilit ies and Manifests, 5.2.4 10CFR61 ,.,Subpar.t. D,Technical, Requirementsfor Land. Disposal Facilities. and Final Waste Classification and Waste. Form Technical Position Papers 5.2.5. !IOCFR61 55, Waste: Classification 5.2.6 10CFR61 :56, Waste Characteristics 5.2.7 10CFR71, Packaging and Transportation of..Radioactive.Material 5.2.8. 49CFR173.1.b, Transportation 5.2.9 49CFR173.427;Transport Requi-erments for Low Specific Activity (LSA)

Radioactive Materials 5.2.10 NU REG-01 23, Standard Radiological Effluent Technical Specifications for Boiling Water Reactors 5.2.11 NUREG-0800,

a. Section 11.2, Standard Review Plan for Liquid Waste Management Systems

b. Section 11.4, Standard Review Plan for Solid Waste Management Systems Page 13 .Rev 07

5.2.12 Resource Conservation and Recovery Act (RCRA) of 1976 (Ref. Corporate&Guide.

to Hazardous Waste Disposal and. Spill.Reporting) 52.13 Regulatory Guide 1.143, Rev. 0, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Ins tal ed in Light Water Cooled Nuclear Power Plants 5.3 Supplemental.References 5.3.1 South Carolina Department of Health and Environmental Control, Radioactive Material License 097, as amended 5.3;2 State of Washington Radioactive Material License No. WN-1019-2, as amended 513:3 NRC Special Nuclear Material License No. .12-13536-02, as amended, for Banwell, SC.

5.3A4 NRC Special Nuclear Material License No. 16-19204-01 as:amended,.for Richland, WA 563.5 Nuclear Regulatory Commission Branc.h.Technical Position on Waste Classification and Waste Form, May 1983 5.3.6 CNSI ProprietaryTopical Report No. RDS-25506-01-NP-A, .Rev. 1- March 1988.

Appendix A, B1 C, D and Material Safety-Data Sheets 5.3.7 SE 92-049, Interim On-Site Storage of Low Level Radioactive Waste (LLRW).in!

the Radwaste Solidification and.Storage Building (RSSB).at Unit 1.

5.3.8 SE 92-061, Upgradeý Radwaste 245' Elevation Storage, at Unit 2.

5.3.9 N2-WHP-25, Thermex Operating Procedure 5.3.10 Safety Evaluation 94-074, Installation of the'Thermex System 5.3.11 DCP N2-05-064, Condensate Filtration Radwaste Processing Page 14 Rev 07

ATTACHMENT 1; UNIT 2 RADWASTE PROCESS CONTROL PROGRAM REFERENCE AND IMPLEMENTING PROCEDURES Waste Handling Procedures (WHPs)

Radiation Protection Procedures.(S-RPIPs)

Chemistry Procedures (CSPs)

Quality.Assurance Procedures:(CNG-QL Series)

Operating Procedures (OPs)

Generation. Administrative Procedures (GAP/APs)

Nuclear Division Interfacing Procedures (N Ps).

Page 15 Rev 07

ATTACHMENT 2: SOLID WASTE SOURCES (Sheet 1 of 3) 1.0 RADWASTE FILTERS 1..1 Mechanical radwaste filters filter particulates (backwash: material) from the waste collector subsystem.

1.2 When a filter reaches apre-determihed differential pressure, the filter media is backwashed into the backwash tank, which is then pumped to the spent resin tank, Regen Waste Tanks, or to a liner in the Radwaste Truckbay.

2.0 RADWASTE DEMINERALIZERS 2.1 The radwaste demineralizers are loaded with' an ionic exchange media for processing water from.

the wastfecollector tanks.

2.2 When determined, the resin canno longer be used, the depleted resin is:pumped to the spent resin tank..

3.0 CONDENSATE FILTRATION SYSTEM 3.1 The Condensate filtration system removes iron particulates.,to extend Condensate demineralizer bed lifeý 3.2 The filters are periodically backwashed and the iron laden waste is sent tola backwash receiving tank. The.waste is then pumped~to aphase separator tank where a flocculent is.added turning:the iron into sludge. This.sludge is then .pumpedto.a liner for offsite processing.

3.3 The Condensate Prefilter Elements are treated as solid Radwaste at the end of their useful life, Filter Elements are shipped offsite for vendor processing, 4.0 CONDENSATE DEMINERALIZER 4.1 The condensate demineralizers remove soluble and insoluble impurities from the condensate water to maintain reactor feedwater purity, 4.2 After it is determined these resins can no longer be used, the depleted resins are pumped to the Radwaste Derhineralizer or Spent, Resin Tank.

Page 16 Rev 07

ATTACHMENT 2: SOLID WASTE SOURCES-(Cant)

(Sheet 2 of:3) 5.0 THERMEX SYSTEM 5.1 Concentrated waste will be pumped to aRegen Waste Tahkfor further concentration by an evaporator, stored in a temporary liner for:offsite processii0g, orpumped to the Spent Resin Tank.

5.2 Exhausted resin and charcoal are sluiced to the Waste Sludge Tank. This Waste may be transferted to the Spent Resin Tank, mixed to a homogenous. mixture, and then transferred to a:

liner in: the .truckbay for dewatering, or transferred to a liner in the Radwaste Truckbay.

.5.3 Exhausted reverse osmosis membranesaand filters will be processed as DAW.!

6.0 SPENT FUEL POOL PHASESEPARATOR This tank receives the exhausted powdered filter media-.(resins) from the Spent Fuel Pool Cleanup System which is subsequently pumped directly to a liner in the Radwaste Truckbayforprocessing.

7.0 RWCU PHASE SEPARATOR These.separator tanks receive exhausted.powdered filter~media (resins) from the water:cleanup' system which is subsequently pumped directly, to a liner in the Radwaste Truckbay for processing..

8.0 CONTAMINATED FLUIDS Fluids rlfom sourceswithin Unit 2 that become contaminatedis either stored in containers (to beý solidified by a vendor with an approved procedure) or shipped off-site, for incineration.

9.0 COMPACTIBLE SOLIDS Compactable low level trash is either processed and compacted in a hydraulically operated box compactor, or shipped off-site for vendor separation and processing.

10.0 FILTERS AND. MISCELLANEOUS ITEMS.

Solid items with high dose. rates are handled on a case-b'-case basis, being disposed of by methods acceptable to the burial site or shipped off-site for vendor recovery or disposal.

Page 17 Rev 07

ATTACHMENT 2: SOLID WASTE SOURCES (Cant)

(Sheet 3 of3) 11.0 SPENT RESIN TANK 11.1 Exhausted resin from the condensate dernineralizerand the Radwaste demineralizer' are sluiced to the Spent Resin Tank. Exhausted resin from the RWCUI phase separator(s), the Spent Fuel Pool phase separator and the Radwaste Filter Backwash Tanks may be sluiced to the Spent Resin Tank.

1ii2 Bleed water from the Thermex System may be pumped to the Spent Resin: Tank.

11.3 The waste from the Spent Resin Tank is,pumped::to.the Waste Sludge Tank for processing by the..

Dewatering.System in the Radwaste Truckbay.

Page 18 Rev 07.

ATTACHMENT 14 Unit 1 Unit 2 X Reporting Period January - December 2008 ATTACHMENT 14 Includes Attachment 2 and Attachment 3, Revision 1 updates to the Nine Mile Point Nuclear Station Unit 2 2007 Radioactive Effluent Release Report that correct a units error.

ATTACHMENT 14 Unit I Unit 2 X Reporting Period January - December 2007 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES, ELEVATED AND GROUND LEVEL lst 2nd 3rd 4th Est. Total Quarter Quarter Quarter Quarter Error, %

A. Fission & Activation Gases

1. Total Release Ci 2.26E+00 4.38E+00 2.OOE+02 2.06E+02 ý5.OOE+01

2. Average Release Rate pCi/sec 2.90E-01 5.58E-01 2.52E+01 2.58E+01 B. lodines

1. Total Iodine- 131 Ci 4.68E-06 I 3.83E-06 5.76E-04 1.1OE-03 3.OOE-'01

2. Average Release Rate for Period pCi/sec 5.96E-07 4.87E-07 6.79E-05 1.51E-04 C. Particulates 5.27E+00

1. Particulates with half-lives>8 Ci 2.42E-04 4.33E-04 2.33E-04 1.30E-03 3.00E+401I 6.71E 01

2. Average Release Rate for Period pCi/sec 3.08E-05 5.51E-05 2.75E-05 1.79E-04

3. Gross alpha radioactivity Ci 3.98E-08 ....... 2.50E+01 D. Tritium

1. Total release Ci 3.32E+01 i 2.62E+01 4.01E+01 3789E+01 5.OOE+01

2. Average Release Rate for Period pCi/sec 4.23E+00 3.33E+00 4.73E+00 5.36E+00 E. Percent of Tech. Spec. Limits Fission and Activation Gases Percent of Quarterly Gamma Air Dose 1.18E-02 3.92E-01 4.26E-01 6.50E-03 Limit (5 mR)

Percent of Quarterly Beta Air Dose Limit 1.13E-02 9.65E-05 1.62E-04 7.40E-03 (10 mrad)

Percent of Annual Gamma Air Dose 3.25E-03 9.13E-03 2.07E-01 4.21E-01 Limit to Date (10 mR)

Percent, of Annual Beta Air Dose Limit to 4.83E-05 1.30E-04 3.83E-03 9.50E-03 Date (20 mrad)

Percent of Whole Body Dose Rate Limit 1.51 E-02 1.64E-02 2.55E-04 4.57E-04 (500 mrem/yr)

Percent of Skin Dose Rate Limit 8.92E-05 2.96E-03 3.22E-03 4.97E-05 (3000 mrem/yr)

Tritium, lodines, and Particulates (with Percent of Quarterly Dose Limit 1.33E-02 7.95E-03 6.95E-03 1.03E-02 (7.5 mrem)

Percent of Annual Dose Limit to Date 1.27E-02 1.94E-02 401E-03 7.51E-03 (15 mrem)

Percent of Organ Dose Limit 6.38E-03 1.82E-04 1.56E-04 2.83E-03 (1500 mrem/yr)

ATTACHMENT 14 Unit 1 Unit 2 x Reporting Period January - December 2007 GASEOUS EFFLUENTS - ELEVATED RELEASE Continuous Mode (2)

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

Argon-41 2.48E-01 5.16E-01 1.49E-01 1.70E-01 Krypton-85 9.44E-03 ** 1.17E-02 **

Krypton785m 6.56E-01 1.78E+00 3.70E+01 5.29E+01 6.98E-03 ** 1.41E+01 1.14E+01 Krypton-87 Krypton-88 1.13E+00 2.09E+00 7.71 E+01 7.86E+01 Xenon-127 Xenon-131m

- 7.12E+01 4.22E+01 Xenon-1 33 Xenon-1 33m Xenon-1 35 2.41 E-02 ** 1.03E-01 5.33E+00 Xenon-135m 4.23E-02 .... 3.07E+00 Xenon-137 6.18E-02 .... 3.99E+00 Xenon-138 8.46E-02 ** ** 4.43E+00 lodines (1)

Iodine-131 1.98E-06 2.88E-06 3.56E-04 7.65E-04

.... 1.40E-03 6.59E-03 Iodine-133 Iodine-135 Particulates (1)

Chromium-51 Manganese-54

- ** 1.32E-05 3.18E-04 Iron-55 Iron-59 Cobalt-58 Cobalt-60 Neodymium-147 5.27-E+OO 5.27E-06 ** 7.77E-06 ** I Zirconium-95 Zinc-65

.... 3.59E-05 1.10E-04 Strontium-89 Stronium-90 Niobium-95 Molybdenum-99 Cesium-134 Cesium-136 Cesium-137 Barium-140 Lanthanum-140 Cerium-141 Cerium-144 Tritium (1) Ci 2.09E+O1 1.69E+01 2.35E+01 I'.63E+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 of 1.OOE-04 pCi/ml for required noble gases, 1.00E-11 pCi/mI for required particulates and gross alpha, 1.OOE-12 pCi/ml for required lodines, 1.0OE-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.

ML091280433

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5.0 REFERENCES

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