Text

January - December 2003 Radioactive Effluent Release Report
	ML041320372
Person / Time
Site:	Nine Mile Point
Issue date:	05/01/2004
From:	William Holston; Constellation Energy Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NMP1L 1816
	Download: ML041320372 (226)
	v • d • e

P.O. Box 63 Lycoming, NY 13093 Constellation Energy Nine Mile Point Nuclear Station May 1, 2004 NMP1L 1816 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Nine Mile Point Unit 1 Docket No. 50-220 NPF-63 January - December 2003 Radioactive Effluent Release Report Gentlemen:

In conformance with the Nine Mile Point Unit 1 (NMP1) Technical Specifications, enclosed is the Radioactive Effluent Release Report for the reporting period January through December 2003.

Included in this report is a summary of gaseous, liquid, and solid effluents released from the stations during the reporting period (Attachments 1 - 6), a summary of revisions to the Offsite Dose Calculation Manual and the Radwaste Process Control Program during the reporting period (Attachments 7 and 8), and an explanation as to the cause and corrective actions regarding the inoperability of any station liquid and/or gaseous effluent monitoring instrumentation (Attachment 9). Attachments 10 and 11 provide a summary and assessment of radiation doses to members of the public within and outside the site boundary, respectively, from liquid and gaseous effluents as well as direct radiation in accordance with 40 CFR190.

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 NMPI Offsite Dose Calculation Manual. Distribution is in accordance with 10CFR50.4(b)(1) and the Technical Specifications. 2 is a copy of Revision 7 of the Radwaste Process Control Program (RPCP) 3 is a copy of Revision 24 of the Offsite Dose Calculation Manual.

During the reporting period from January through December 2003, NMP1 did not exceed any 10 CFR 20, 10 CFR 50, Technical Specification, or Offsite Dose Calculation Manual limits for gaseous or liquid effluents.

Page 2 NMP1L 1816 If you have any questions concerning the attached report, please contact Mr. Anthony Salvagno, (315) 349-1456, Reliability Engineering.

Very truly yours, William C. Holston Manger Engineering Services WCH/CWP/jm Enclosure cc: Mr. H.J. Miller, NRC Regional Administrator, Region I Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. P. S. Tam, Senior Project Manager, NRR (2 copies)

NINE MILE POINT NUCLEAR STATION - UNIT I ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January - December 2003 Constellation Energy Nine Mile Point Nuclear Station

Page 1 of 2 NINE MILE POINT NUCLEAR STATION - UNIT 1 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY - DECEMBER 2003 SUPPLEMENTAL INFORMATION Facility: Nine Mile Point Unit #1 Licensee: Nine Mile Point Nuclear Station, LLC

1. TECHNICAL SPECIFICATION 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 due to noble gases released in gaseous effluents from Nine Mile Point Unit 1 to areas beyond the site boundary shall be limited during any calendar quarter to less than or equal to 5 milliroentgen 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 milliroentgen 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 1 to areas 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. The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to ten times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-04 microcuries/mi 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 1 to unrestricted areas shall be limited during any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and during any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems 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 gross activity monitoring (calibrated against gamma isotopic analysis of a 4.OL Marinelli grab sample) of an isokinetic stack sample stream.

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

C) PARTICULATES Activity released from the main stack 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. Tritium effluent activity is measured during purge and weekly when fuel is offloaded until stable tritium release rates are demonstrated.

E) EMERGENCY CONDENSER VENT EFFLUENTS The effluent curie quantities are estimated based on the isotopic distribution in the Condensate Storage Tank water and the Emergency Condenser shell water. Actual isotopic concentrations are found via gamma spectroscopy. Initial release rates of Sr-89, Sr-90 and Fe-55 are estimated by applying scaling factors to release rates of gamma emitters and actual release rates are determined from offsite analysis results. The activity of fission and activation gases released due to tube leaks is based on reactor steam leak rates using offgas isotopic analyses.

F) 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. Tritium concentration is estimated to be the same as the most recent analysis of the Condensate Storage Tank water. Initial release rates of Sr-89, Sr-90, and Fe-55 are estimated by applying scaling factors to release rates of gamma emitters and actual release rates are determined from post offsite analysis results.

G) SOLID EFFLUENTS Isotopic contents of waste shipments are determined by gamma spectroscopy analysis 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 X Unit 2 Reporting Period January - December 2003 Uquld Effluents:

ODCM Required MEC - 10 x 10CFR20, Appendix B, Table 2, Column 2 There were no discharges of Liquid Radwaste requiring use of MEC to determine allowable release rate.

MECs for discharges from Emergency Condenser Vents are as follows:

Average MEC - pCi/ml (Qtr. 1) = N/A Average MEC - pCi/ml (Qtr. 3) = 9.92E-03 Average MEC - RCi/ml (Qtr. 2) = 9.94E-03 Average MEC - pCi/ml (Qtr. 4) - NIA Average Energy (Fission and Activation gases - Mev):

Qtr. 1 : Ey - 2.47E-01 = 3.17E-01 Qtr. 2: fy - 2.47E-01 fl 3.17E-01 Qtr. 3: y - 2.34E-01 = 3.06E-01 Qtr. 4: By - 2.47E-01 = 3.17E-01 Uqutd: Radwaste EC Vent Number of batch releases : 0 2 Total time period for batch releases (hrs) : N/A 3.21E+01 Maximum time period for a batch release (hrs) : N/A 3.18E+01 Average time period for a batch release {hrs) : N/A 1.61E+01 Minimum time period for a batch release (hrs) : N/A 3.17E-01 Total volume of water used to dilute the liquid effluent during release 1n 2-d 3 rd 4 th period (L)

Radwaste : N/A N/A N/A N/A EC Vent : N/A 2.04E+05 5.11E+05 N/A Total volume of water available to dilute the liquid effluent during report 1" 2nd 3r 4th period (L)

Radwaste : 1.09E1+11 1.04E + 11 1.35E + 11 1.35E +11 EC Vent . N/A 3.55E+07 3.02E+07 N/A Gaseous - (There were two releases from the operation of the Emergency Condenser Vent):

Number of batch releases : 2 Total time period for batch releases (hrs) : 3.21E+01 Maximum time period for a batch release (hrs) : 3.18E+01 Average time period for a batch release (hrs) : 1.61E+01 Minimum time period for a batch release (hrs) : 3.17E-01 Gaseous IPrImary Containment Purge):

Number of batch releases I Total time period for batch releases (hrs) : 1.21E+02 Maximum time period for a batch release (hrs) : 1.21E+02 Average time period for a batch release (hrs) : 1.21E+02 Minimum time period for a batch release (hrs) : 1.21E+02

ATTACHMENT 1 Summary Data Page 2 of 2 Unit 1 X Unit 2 _ Reporting Period January - December 2003 Abnormal Releases:

A. Uqulds:

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 1 X Unit 2 Reporting Period January - December 2003 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 I I I I I

1. Total release Ci 1 .44E-01 4.30E-03 2.15E-01 3.19E-05 5.OOE+01

2. Average release rate pCi/sec 1 .85E-02 5.51E-04 2.71 E-02 4.01 E-06 B. lodines

1. Total lodine-1 31 Ci 1.71E-04 3.97E-05 1.09E-04 9.30E-05 3.OOE+01

2. Average release rate for period ,uCi/sec 2.23E-05 4.99E-06 1.39E-05 1.19E.05 C. Particulates

1. Particulates with half-lives >8 days Ci 3.0312-03 7.33E-04 1.36E-03 3.91 E-04 3.OOE+01

2. Average release rate for period uCilsec 3.96E-04 9.23E-05 1.74E-04 5.O1E-05

3. Gross alpha radioactivity Ci 4.53E-05 5.97E-05 5.30E-05 4.73E-05 2.50E+01 D. Tritium

1. Total release Ci 6.69E+O0 5.31E+00 1.65E+02 7.50E+O0 5.00E+011

2. Average release rate for period ,Mci/sec S.73E-01 6.68E-01 2.11E+01 9.60E-01 E. Percent of Tech. Spec. Limits Fission and Activation Gases Percent of Quarterly Gamma Air Dose Limit (5 mR) 6.58E.04 1.98E-05 9.66E-04 2.58E-07 Percent of Quarterly Beta Air Dose Umit (10 mrad) 2.71 E.04 8.06E-06 4.27E-04 1.65E-07 Percent of Annual Gamma Air Dose Limit to Date 110 mR) 3.29E-04 3.39E.04 8.22E-04 8.22E-04 Percent of Annual Beta Air Dose Limit to Date (20 mrad) 1.36E-04 1.40E-04 3.53E-04 3.53E-04 Percent of Whole Body Dose Rate Limit (500 mrem/yr) 1.78E-05 5.29E-07 2.53E-05 6.74E-09 Percent of Skin Dose Rate Limit (3000 mrem/yr) 6.22E-06 ¶1.86E-07 9.05E-06 2.97E-09 Tritium, lodines, and Particulates (with half-lives greater than 8 daysl Percent of Quarterly Dose Limit (7.5 mrem) 1.31 E-01 3.64E-02 2.25E-01 6.16E-02 Percent of Annual Dose Limit to Date 115 mrem) 6.60E-02 8.44E-02 1 .98E-01 2.23E-01 Percent of Organ Dose Rate Limit (1500 mrem/yr) 2.70E-03 7.24E-04 4.53E-03 1.04E-03

ATTACHMENT 3 Page 1 of 1 Unit 1 X Unit 2 Reporting Period January - December 2003 GASEOUS EFFLUENTS - ELEVATED RELEASE CONTINUOUS MODE 2 1st 2nd 3rd 4th Nuclides Released QUARTER QUARTER QUARTER QUARTER

1. Fission Gases' Argon-41 Ci *- .*

- *v Krypton-85 Ci *-

.4 Krypton-85m Ci *-

.-4.

Krypton-87 Ci *-

44 *-

Krypton-88 Ci *T Xenon-127 Ci 4-Ci *- 44 7*-

Xenon-1 31 m .4

T Xenon-1 33 Ci 44 Xenon-1 33m Ci *-

Xenon-1 35 Ci I .44E-Oi 4.30E-03 1. 6R-0 1 Ci *- *0 *-

Xenon-1 35m 4.

Xenon-1 37 Ci .4 44 77 Xenon-1 38 Ci 44 44

2. Iodines1 Iodine-131 Ci 1.71E-04 3.97E-05 1.09E-04 9.30E-05 Iodine-133 Ci 1.32E-04 lodine-1 35 Ci

3. Particulatesi Strontium-89 Ci 1.62E-04 ..

Strontium-90 Ci Cesium-1 34 Ci Cesium-1 37 Ci 3.63E-06 2.1 3E-05 Cobalt-60 Ci 1.44E-03 4.30E-04 8.1SE-04 3.43E-04 Cobalt-58 Ci 7.78E-05 1.06E-05 1.98E-05 Manganese-54 Ci 4.25E-04 9.13E-05 2.75E-04 2.72E-05 Barium-Lanthanum-140 Ci

7-6 *- *-

Antimony-1 25 Ci *W *r- IrT*

Niobium-95 Ci Cerium-141 Ci Cerium-144 Ci Iron-59 Ci 3.65E-05 .. .

Cesium-136 Ci Chromium-51 Cl 2.40E-04 4.30E-05 3.30E-05 .

Zinc-65 Cl Iron-55 Ci 6.64E-04 1.55E-04 2.11E-O4 Molybdenum-99 Ci Neodymium-147 Ci

4. Tritium Ci 5.43E + 00 3.40E+0o 6.79E+00 6.18E+00 Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk. A lower limit of detection of 1.OOE-04 pCi/ml for required noble gases, 1.OOE-1 1 ttCi/ml for required particulates, 1.OOE-1 2 pCi/ml for required lodines, and 1.OOE-06 pCi/ml for Tritium, as required by 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 2 Unit 1 X Unit 2 _ Reporting Period January - December 2003 GASEOUS EFFLUENTS - GROUND LEVEL RELEASES Ground level releases are determined in accordance with the Off-Site Dose Calculation Manual and Chemistry procedures.

CONTINUOUS MODE 1St 2nd 3rd 4th QUARTER QUARTER QUARTER QUARTER T r

1. Fission Gases' Argon-41 Ci *-

F*

Krypton-85 Ci *-

Ci 7* *V*

Krypton-85m ** ;*4 Krypton-87 Ci W**

Ci *-i 44 Krypton-88 ;**

Xenon-127 Ci r*_

Xenon-131 m Ci Xenon-133 Ci 5.603E-05 77 04 Xenon-133m Ci Xenon-135 Ci 2.33E-05 2.270E--05 3.1i9E-05 Xenon-1 35m Ci 4.

Xenon-1 37 Ci ;**

4* 4;*-

Xenon-1 38 Ci

2. lodines' lodine-1 31 Ci 44 ** *4 *4, lodine-1 33 Ci lodine-1 35 Ci 44 *- *- 4*

3. Particulates' Strontium-89 Ci 4* *4 *- *4 4* *- 44 *-

Strontium-90 Ci Cesium-1 34 Ci 44 _ *4

_ 4*_* *-_

Cesium-1 37 Ci 44 *4 4* *4 Cobalt-60 Ci Cobalt-58 Ci T.* *4 4*

44 *4 *4 *4 Manganese-54 Ci

4 *4 *- *4 Barium-Lanthanum-140 Ci

4 *4 *- *4 Antimony-125 Ci Niobium-95 Ci

- *- 4* *-

Cerium-141 Ci

4 44 4* *4 Cerium-144 Ci 44 4* *4 44 Iron-59 Ci 44 4* iF*4 *4 Cesium-1 36 Ci

- 4*F *- *4 Chromium-51 Ci 44 *4 *4 4*

Zinc-65 Ci 44 *4i *4 *4 Iron-55 Ci

-4* *4 *-

Molybdenum-99 Ci 44 4* 44 *4 Neodymium-1i47 Ci

4. Tritium Ci 1.26E+O0 1.14E+00 1.66E+00 1.32E+00 I Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk.

ATTACHMENT 4 Pace 2 of 2 Unit 1 X Unit 2_ Reporting Period January - December 2003 GASEOUS EFFLUENTS - GROUND LEVEL RELEASES Ground level releases are determined in accordance with the Off-Site Dose Calculation Manual and Chemistry procedures.

BATCH MODE 1st 2nd 3rd

4th QUARTER QUARTER QUARTER QUARTER

1. Fission Gases' Argon-41 Ci No Releases ** No Releases Krypton-85 Ci No Releases W** No Releases Krypton-85m Ci No Releases No Releases Krypton-87 Ci No Releases V**

No Releases Krypton-88 Ci No Releases No Releases Xenon-127 Ci No Releases i** No Releases Xenon-131 m Ci No Releases No Releases Xenon-133 Ci No Releases 7*6 ** No Releases Xenon-133m Ci No Releases No Releases Xenon-135 Ci No Releases No Releases Xenon-135m Ci No Releases No Releases Xenon-137 Ci No Releases No Releases Xenon-138 Ci No Releases No Releases

2. lodines' lodine-131 Ci No Releases No Releases lodine-1 33 Ci No Releases 7 No Releases Iodine-135 Ci No Releases 7 No Releases

3. Particulatesi Strontium-89 Ci No Releases No Releases Strontium-90 C No Releases 77 No Releases Cesium-1i34 Ci No Releases W No Releases Cesium-1i37 Ci No Releases 2.04E-07 No Releases Cobalt-60 Ci No Releases 2.18E-08 4.32E-06 No Releases Cobalt-58 Ci No Releases 2.86E-09 3.90E-07 No Releases Manganese-54 Ci No Releases 8.30E-09 2.27E-06 No Releases Barium-Lanthanum-140 Ci No Releases No Releases Antimony-1 25 Ci No Releases No Releases Niobium-95 Ci No Releases 7 No Releases Cerium-141 Ci No Releases 77 No Releases Cerium-144 Ci No Releases 7T 77 No Releases Iron-59 Ci No Releases 77 77 No Releases Cesium-1 36 Ci No Releases 7 W No Releases Chromium-51 Ci No Releases 2.23E-0s 77 No Releases Zinc-65 Ci No Releases 7*_ No Releases Iron-55 Ci No Releases *_ 7 No Releases Molybdenum-99 Ci No Releases No Releases Neodymium-1i47 Ci No Releases No Releases

4. Tritium Ci No Releases 7.70E-01 1.57E+02 No Releases

Refects initiation of Emergency Condensers in August 2003 during the Northeast blackout.

1 Concentrations less than the lower limit of detection of the counting system used are indicated with a double asterisk.

ATTACHMENT 5 Page 1 of 2 Unit I X Unit 2 _ Reporting Period January - December 2003 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES 1st 2nd 3rd 4th EST.

QUARTER QUARTER' QUARTER' QUARTER TOTAL ERROR. %

A. Fission & Activation Products

1. Total release (not including Tritium, gases, alpha) Ci No Releases 6.09E.07 2.04E-06 No Releases 5.OOE+01

2. Average diluted concentration during reporting period MCi/ml No Releases 1.72E-1 I 6.75E-1 1 No Releases B. Tritium

1. Total release Ci No Releases 1.56E-02 4.47E-02 No Releases 5.OOE+01

2. Average diluted concentration during reporting period ,pCi/ml No Releases 4.39E-07 1.48E-06 No Releases C. Dissolved and Entrained Gases

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

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

1. Total release Ci No Releases No Releases 5.OOE+01 E. Volumes2

1. Prior to dilution Liters No Releases 1.89E + 03 3.78E + 03 No Releases 5.OOE+01

2. Volume of dilution water used during release period Liters No Releases 2.04E+05 5.11E+05 No Releases 5.OOE +0 1

3. Volume of dilution water available during reporting period Liters No Releases 3.55E+07 3.02E+07 No Releases 6.OOE+01 F. Percent of Technical Specification Limits Percent of Quarterly Whole Body Dose Limit (1.5 mreml No Releases 7.74E-03 6.12E-02 No Releases Percent of Quarterly Organ Dose Limit (5 mrem) No Releases 4.64E-03 2.62E-02 No Releases Percent of Annual Whole Body Dose Limit to Date (3 mrem) No Releases 3.86E-03 3.43E-02 No Releases Percent of Annual Organ Dose Limit to Date (10 mrem) No Releases 2.32E-03 1.44E-02 No Releases Percent of 10CFR20 Concentration Limit No Releases 4.43E-03 1.50E 02 No Releases Percent of Dissolved or Entrained Noble Gas Limit (2.OOE-04 jsCi/ml) No Releases I No Radwaste Batch releases. Liquid Batch Releases associated with Emergency Condensesr operation are assumed to be discharged via storm drain in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2 Dilution water volumes based on SPDES report storm drain flow estimates for EC vent releases only.

ATTACHMENT 5 Page 2 of 2 Unit 1 X Unit 2 Reporting Period January - December 2003 LIQUID EFFLUENTS RELEASED BATCH MODE 12 Nuclides Released Ist 2nd 3rd 4"'

NcdeReesdQUARTER QUARTER QUARTER QUARTER Strontium-89 Ci No Releases No Releases Strontium-90 Ci No Releases No Releases W*

Cesium-134 Ci No Releases . 77 No Releases Cesium-1 37 Ci No Releases . 5.79E-08 No Releases lodine-1 31 Ci No Releases . No Releases Cobalt-58 Ci No Releases .. 1.11E-07 No Releases Cobalt-60 Ci No Releases 4.471E-07 1.23E-06 No Releases Iron-59 Ci No Releases *. No Releases Zinc-65 Ci No Releases 77 .. No Releases Manganese-54 Ci No Releases 11.68E-07 6.43E-07 No Releases Chromium-51 Ci No Releases .. .. No Releases Zirconium-Niobium-95 Ci No Releases *0 No Releases Molybdenum-99 Ci No Releases **

No Releases Technetium-99m Ci No Releases .. ;7 No Releases Barium-Lanthanum-140 Ci No Releases ..

No Releases Cerium-141 Ci No Releases .*0* No Releases Tungsten-1 87 Ci No Releases .. *0 No Releases Iodine-1 33 Ci No Releases .. 0* No Releases Iron-55 Ci No Releases *. No Releases Neptunium-239 Ci No Releases * . No Releases Iodine-135 Ci No Releases No Releases Dissolved or Entrained Gases Ci No Releases O.OOE+00 O.OOE+00 No Releases Tritium Ci No Releases 1.56E-02 4.47E-02 No Releases No continuous mode release occurred during the report period as indicated by effluent sampling. No Radwaste Batch releases. Liquid Batch Releases associated with Emergency Condenser operation in April and August 2003 are reflected.

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 ACilml for required gamma emitting nuclides, 1.OOE-05 pCi/ml for required dissolved and entrained noble gases and tritium, 5.00E-08 jxCi/ml for Sr 89/90, 11.OOE-06 jCi/ml for I-1 31 and Fe-55, and 1.OOE-07 PiCi/ml for gross alpha radioactivity, as identified in the Off-Site Dose Calculation Manual, has been verified.

ATTACHMENT 6 Page 1 of 4 Unit 1 X Unit 2 Reporting Period January - December 2003 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A.1 TYPE Volume Activity' 3

(M ) (Ci)

Class Class

a. Spent Resins (Dewatered)

A B C A B C 2.35E+01 0 0 1.53E+02 0 0

b. Dry Compressible 6.16E+02 0 0 B.55E-01 0 0 W aste _ _ _ _ _ _ _ _ _ _ _ _ _

c. Irradiated Components, 0 0 0 Control Rods

d. Other:

(to vendor for processing)

1. Sewage Sediment 5.44E+01 0 0 2.93E-03 0 0

' The estimated total error is 5.OOE+01%.

ATTACHMENT 6 Page 2 of 4 Unit 1 X Unit 2 Reporting Period January - December 2003 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A.1 TYPE Solidification Container Package Agent

a. Spent Resins (Dewatered) Poly Lner STP I Type A None Type B

b. Dry Compressible Waste [ Metal Box STP J None

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

d. Other: (To Vendor for Processing)

1. Sewage Sediment Metal Box STP None

ATTACHMENT 6 Paae 3 of 4 Unit 1 X Unit 2 Reporting Period January - December 2003 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A.2 ESTIMATE OF MAJOR NUCLIDE COMPOSITION (BY TYPE OF WASTE)

a. Spent Resins (Dewatered)

Nuclide Percent (1) Mn-54 3.30E+01 (2) Fe-55 3.10E+01 (3) Co-60 3.03E+01 (4) Co-58 3.1OE+00 (5) Cs-1 37 1.30E+OO (6) Other 1.30E+00

b. Dry Compressible Waste Nuclide Percent (1) Fe-55 6.91E+01 (2) Co-60 2.07E+01 (3) Mn-54 5.20E+00 (4) Cs-1 37 3.40E+OO (5) Other 1.60E+OO

c. Irradiated Components, Control Rods Nuclide Percent

d. Other: (to Vendor for Processing)

1. Sewage Sediment Nuclide Percent (1) Ce-144 6.56E+01 (2) Co-60 1.48E+01 (3) Cs-137 1.23E+01 (4) Mn-54 7.20E +00 (5) Other 1.00E-O1

ATTACHMENT 6 Page 4 of 4 Unit 1 X Unit 2 _ Reporting Period January - December 2003 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A.3. SOLID WASTE DISPOSITION:

Number of Shipments Mode of Transportation Destination 11 Hittman Transport -Truck Duratek Services. Inc.

1 TAG Transport - Truck Duratek Services. Inc.

5 Hlttman Transport - Truck Studsvik Processing Facility, LLC B. IRRADIATED FUEL SHIPMENTS (DISPOSITION): There were no shipments.

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

ATTACHMENT 7 Page 1 of 1 Unit 1 X Unit 2 _ Reporting Period January - December 2003

SUMMARY

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

The Unit 1 Off-Site Dose Calculati9on Manual (ODCM) was revised during the reporting period to correct references to the Technical Specification Administrative Controls Sections in support of TS Amendment 181 implementation. This revision also clarified wording In the 6th paragraph of ODCM Part II, Section 2.1.1. 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 24 is attached and a summary of the changes presented to and approved by the Stations Operations Review Committee on July New/Amended Page # Description of Change Reason for Change Section #

ix, I 1.0-1, Introduction, Part I Revised to correct references To support Technical Specification Amendment 181.

13.1-22,1 1.0, Table D 4.6.15-2 to the Technical 3.1-38,1 Notes, Table D Specification Administrative 3.1-39, I B 4.6.20-1 Notes, B Controls Sections.

3.1-1, I 3/4 .6.14, and 6.0; 6.0-1 and I Part II 1.3.2.1, 2.2, 6.0-3; Part 2.2.1, 2.2.3.1, and Il 11 11, II 2.2.3.2.

17,1118, II 24, and 11 25.

11 12 2.1.1 Insert the words steady This change clarifies the intention of the wording state' between 'until' and that indicates the alarm setpoints are not to be

.power' in the 6th changed during outages, AND during startup from paragraph of ODCM Part II, outages to steady state operation. This changing Section 2.1.1. effluent condition during startup would require constant changes to the effluent setpoints until steady state conditions exist.

ATTACHMENT 8 Page 1 of 2 Unit 1 X Unit 2 Reporting Period January - December 2003

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)

The Unit 1 Radwaste Process Control Program (PCP) Revision 7 was implemented in December 2003. Revisions were made to reflect the alignment of responsibilities with the current management structure, reflect a recent amendment for Section 6 of the Technical Specifications, and to enhance existing information and make grammatical corrections. The PCP changes do not reduce the overall conformance of a solidified waste product to existing criteria for solid waste. A copy of the PCP, Revision 7 is attached and below is a summary of the changes accepted by the Station Onerations Review Committee.

Old Page # New Page # NewlAmended Change Reason for Change

__ _ _Section #

Changed Radiation Protection Alignment of responsibilities N/A Page 1 Step 2.2 r to Operations Manager. to support the Optimize Manaer o Opratons anaer. the Organization' initiative Added newresponsibilities.

step to include Enhancement to explicitly N/A Page 1 Step 2.4 Operator's state responsibility Changed Manager of Chemistry Alignment of responsibilities N/A Page 4 Step 4.3.2 to General Supervisor Chemistry to support the Optimize toSuervsorthe Gnerl Organization' initiative Changed 'by applicable Unit 1 Technical Specification NIA Page 5 Step 4.4.2.a Tech Spec' to 'by the UFSAR Amendment Section B.2.2.16'.

Old Step 6.1.1.a changed from P6.5.2.11 Technical Review and Technical Specification N/APage 7 Step 6.1.1 control' to 6.6.3 Radioactive Amendment 181 Effluent Release Report'.

N/A Page 7 Step 6.1.1 Deleted Step 6.1.1.b. Amendment 181 Changed from 'Radioactive Effluent TechnicalTehiaSpcfato N/A Page 7 Step 6.1.2 Specifications, Amendment No. Tenilect181 66" to 'Offsite Dose Calculation Manual' N/A Page 11 Section 2.1 h an ionic'. Grammatical correction N/A Page 11 Section 2.2 Changed NO to lower case no. Grammatical correction N/A Page 11 Section 3.2 Changed NO to lower case no. Grammatical correction Split into substeps a and b and Change to reflect actual N/A Page 11 Section 4.2 Added or transferred to a liner practice and provide in the truck bay. efficiency Grammatical correction changed c . and drained in the thickener N/A Page 12 Section 8.0 tanks to, and drained to the thickener tanks.

ATTACHMENT 8 Pace 2 of 2 Unit 1 X Unit 2 Reporting Period January - December 2003

SUMMARY

OF CHANGES TO THE PROCESS CONTROL PROGRAM 1PCPl Old Page # New Page 1 NewlAmended Change Reason for Change Section #

Added cleanup demineralizer to Not included previously and N/A Page 12 Section 9.0 exhaust resins that are added to should have the spent resin tank.

Removed for incineration to processing and/or disposal. This N/A Page 12 Section 10.0 includes incineration or other processing methods that maybe employed by the vendor.

Deleted Step 11.2 that further Eliminate unnecessary text.

N/A Page 13 Section 11.0 described low level compactible The description is covered solids. under the previous step.

Changed 'evaporator' to Reflects actual plant N/A Page 13 Section 13.0 "concentrator' in the nomenclature for the description. system Changed 'evaporator' to Reflects actual plant NIA Page 13 Section 13.1 concentrator' in the nomenclature for the description. system Changed evaporator to concentrator' In the description Reflects actual plant N/A Page 13 Section 13.2 and changed evaporator bottoms nomenclature for the to #13 Concentrated Waste system Tank.

ATTACHMENT 9 Page 1 of 1 Unit 1 X Unit 2 _ Reporting Period January - December 2003

SUMMARY

OF INOPERABLE MONITORS Off-gas Radiation Monitor RAM-RN12A was inoperable from 7/14/03 at 15:15 until 8/28/03 at 16:55. The monitor was declared inoperable due to erratic readings. The problem was traced to the main power supply for the monitor. The power supply was replaced, and the monitor was calibrated and returned to service. During this period, redundant monitor RAM-RN12B was operable. Therefore, the minimum number of channels required by ODCM D 3.6.14.b was always maintained.

Service Water radiation monitor RAM-72-406 was Inoperable from 10/8/03 at 02:39 until 11/8/03 at 11:20. The inoperability was caused by a failure of the automatic sample stream switch. Blocking valve BV-72-398 would not close.

The switch could not be repaired, and is obsolete. A rebuilt switch was obtained and installed. The monitor was then calibrated and returned to service. There is no redundant monitor, therefore, the minimum number of channels required by ODCM D 3.6.14.a was not met during this period.

ATTACHMENT 10 Page 1 of 3 Unit 1 X Unit 2 Reporting Period January -December 2003 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 1 (NMP1) liquid and gaseous effluents has been conducted for the period January through December 2003.

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 2003 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 NMP1 stack and emergency condenser 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 NMP1 Stack and Emergency Condenser Vent.

Direct radiation pathway; dose resulting from the operation of NMP1, Nine Mile Point Unit 2 (NMP2) 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 NMP1 Offsite Dose Calculation Manual (NMP1 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 2003 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-1 37 concentration = 0.22 pCi/g The total whole body and skin doses received by a hypothetical maximum exposed fisherman from the external ground pathway is presented in Table 1, Exposure Pathway Dose.

ATTACHMENT 10 Pace 2 of 3 Unit 1 X Unit 2 _ Reporting Period January - December 2003 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 NMP1 ODCM, as adapted from Regulatory Guide 1.109. The total whole body dose and organ dose received by the hypothetical maximum exposed fisherman during 2003 is calculated using the following input parameters for gaseous effluents released from both the NMP1 Stack and Emergency Condenser Vent for the time period exposure is received:

NMP 1 Stack:

Variable Fisherman

Average Stack flow rate (m3 /sec) 1.043E+02 XIQ Is/m3 ) 8.9E-06 Inhalation dose factor Table E-7 Regulatory Guide 1.109 3

Annual air intake m lyear) (adult) 8000 Fractional portion of the year (hours) 0.0356 H-3 (pCi/M 3 ) 6.573E+03 3

Cr-51 (pCi/M ) 3.154E-02 3

Mn-54 (pCi/M ) 1.607E-01 3

Fe-55 (pCi/M ) 1.508E-01 Co-58 (pCi/M 3 ) 1.245E-02 3

Co-60 (pCiIm ) 6.427E-01 3

Cs-1 37 (pCiIM ) 9.764E-03 3

1-131 (pCiIm ) 9.668E-02 1-133 (pCi/M 3 ) 5.346E-02 NMP1 Emer 3ency Condenser Vent:

Variable Fisherman

Average Vent flow rate (m3/sec) 5.825E-04 3

X/Q (s/m 1 6.63E-06 Inhalation dose factor Table E-7 Regulatory Guide 1.109 Annual air intake m3/year) (adult) 8000 Fractional portion of the year 0.0356 H-3 IpCI/m31 8.29E+09 Cr-61 _I__I_

_ _ 3.01E+01 Mn-54 pCIM3__ 6.73E+01 Co-58 _pCim__ 1.35E+01 Co-60 (pCi/M 3 ) 1.36E+02 3

Cs-137 IpCi/m ) 5.04E+00 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 Dose.

ATTACHMENT 10 Page 3 of 3 Unit 1 X Unit 2 Reporting Period January - December 2003 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 NMP1 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 2003 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 (mRemlhr) 1.01 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 Dose Dose Received By A Hypothetical Maximum Exposed Member Of The Public Inside the Site Boundary During 2003 The following is a summary of the dose received by a hypothetical maximum exposed fisherman from Liquid and Gaseous effluents released from NMP1 during 2003:

TABLE I Exposure Pathway Annual Dose Exposure Pathway Dose Type Fisherman (mReml External Ground Whole Body 3.44E-03 Skin of Whole Body 4.01 E-03 Inhalation Whole Body 1.72E-03 Maximum Organ Lung: 1.85E-03 Direct Radiation Whole Body 0.314 Based on these values the total annual dose received by a hypothetical maximum exposed member of the public inside the site boundary is as follows:

TABLE 2 Annual Dose Summary Total Annual Dose for 2003 Fisherman Total Whole Body (mRem) 0.319 Skin of Whole Body (mReml 4.01 E-03 Maximum Organ (mRem) Lung: 1.85E-03

ATTACHMENT 11 Page 1 of 2 Unit 1 X Unit 2 Reporting Period January - December 2003 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE 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 2003 for comparison against the 40CFR1 90 annual dose limits.

The intent of 40 CFR 190 requires that the effluents of Nine Mile Point Unit 1 (NMP1), as well as other nearby uranium fuel cycle facilities, be considered. In this case, the effluents of NMP1, Nine Mile Point Unit 2 (NMP2) 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:

< 25 mRem wholebody

< 25 mRem any organ (except thyroid)

< 75 mRem thyroid This evaluation compares doses resulting from Liquid and Gaseous effluents and direct radiation originating from the site as a result of the operation of the NMP1, NMP2 and 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 of NMP1, NMP2 and JAF facilities.

Methodologies for Determining Dose for Applicable Pathways Fish Consumption Dose received as a result of fish consumption is based on the methodology specified in the NMP1 Off-site Dose Calculation Manual (NMP1 ODCM) as adapted from Regulatory Guide 1.109. The dose for 2003 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 2003; therefore no dose was received by the whole body and organs of the likely most exposed Member of the Public during 2003.

ATTACHMENT 11 Page 2 of 2 Unit 1 X Unit 2 Reporting Period January - December 2003 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES OUTSIDE THE SITE BOUNDARY Shoreline Sediment Dose received from shoreline sediment is based on the methodology in the NMP1 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.

The total dose received by the whole body and skin of the maximum exposed individual during 2003 is calculated using the following input parameters:

Usage Factor - 67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months 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

Average Cs = 137 Concentration = 0.048 pCilg Dose Pathways Resulting From Gaseous Effluents Dose received by the likely most exposed member of the public due to gaseous effluents is calculated In accordance with the methodology provided in the NMP1 ODCM, NMP2 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 NMP1, NMP2 and JAF during 2003 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 2003, 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 During 2003 1 Exposure Pathway [ Dose Type - Dose (mRem) l Fish Consumption Total Whole Body No Dose Total Maximum Organ No Dose Shoreline Sediment Total Whole Body 11.62E-04

_ Total Skin of Whole Body 1.89E-04 Gaseous Effluents Total Whole Body 2.65E-02 Total Maximum Organ Thyroid: 4.21 E-02 Direct Radiation Total Whole Body 1.9 Based on these values the maximum total annual dose potentially received by the most likely exposed member of the public during 2003 is as follows:

Total Whole Body 1.9 mRem

Total Skin of Whole Body: 1.89E-04 mRem

Maximum Organ: Thyroid: 4.21E-02 mRem 40CFR190 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: 0.042 mRem) and the maximum whole body dose (1.9 mReml 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 vI Energy Group Con trO11edj Documen-t Cp Nine Mile Point Fo ,Ltest Iormation Nuclear Station I Check CDS NINE MILE POINT NUCLEAR STATION NINE MILE POINT UNIT 1 OFF-SITE DOSE CALCULATION hlAN (ODCM)

DATE APPROVALS SIGNATURES REVISION 24 Prepared by:

T. M. Kurtz"f Health Physicist Checked by: Jt~7 G. R. Stinson 7/9/03 Health Physicist 7,>O9/°03 Reviewed by:

T. G. Kulczyckf Principal Engineer - Reliability Engineering L. A. Hopkins F.- is b Plant Gener anager B.S. Montgomery 7/3g/oo3 Mana~7 Engine&mng ServiceW

SUMMARY

OF REVISIONS Revision 24 (Effective 7/15/2003)

PAGE DATE 1, 2,5,6,8,9, 11-13/15-18, February 1987 21, 24, 25, 36-44, 47-49, 52-81,86-116 3, 4, 7, 10, 14, 19, 20, 22, 23, 26-35 December 1987 45,46,50,51,82-85 January 1988

29 May 1988 (Reissue)

64,77,78 May 27, 1988 (Reissue) 1, 19, 21, 22A, 22B, 124, 25, 26, 112 February 1990 i, ii, iii, 12-16,18,28-40,4S47 52,55,59-89,92,93,97-129 June 1990 91-93, 95 June 1992 3, 4, 21, 92, 95a-c February 1993 10, 16-20 March 1993 5,13, 18, 20, 25-30, 65,79 June 1993 66,69 December 1993 16,69 June 1994 10, 12 February 1995 10,18,67, 69 December 1995 5, D-1 June 1996 5, D-1 June 1997 5, D-1 April 1999 D-1 December 1999 iv, 3,6,8,9, 11, 13, 14,27,29,65,66,69, 69a December 2001 Added Part I & Revised Part I - I 2-16,1 20-23, II 25, II 26, II 29, I 30 November 2002 iv, v, viu, viii, I 1.0-1 and 2, I 3.1-1,7 to 9, 11, 14, 18 to 24, 26 and 27, I B 3.1-1, 3 to 7, I 6.0-2, 4, and 5, I 2, II 3, I 4, II 6, I 9 to 11, 1 13 to 22, 1 42, Figure D-8, Deleted Figures D-7, D-9, D-10 November 2002 Unit 1 ODCM Revision 24 July 2003

SUMMARY

OF REVISIONS (continued)

Revision 24 (Effective 7115/2003)

PAGE DATE x, I 1.0-1, I 3.1-22, I 3.1-38 and 39, I B 3.1-1, I 6.1-0 and 3,1 11, 12,17,18,24 and 25 July 2003 I Unit 1 ODCM Revision 24 ii July 2003

ODCM - NINE MILE POINT UNIT 1 TABLE OF CONTENTS PAGE List of 'V..-'-

I Uit. .................................................................................................................................................... viii List 01. ures ....

INTRODUCTION ..................................... ........................................................................................................

x PART I - Radiological Effluent Controls SECTION 1.0: Defmitions.. ................................................................................................. I 1.0-0 SECTION 2.0: Not Used SECTIONS 3.0/4.0: Applicability ................................................................................................... 13.0-0 D 3/4.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION ............ .................... I 3.1-1 D 3/4.6.14.a. Liquid Effluent ..................................................... I 3.1-1 D 3/4.6.14.b Gaseous Process and Effluent ..................................................... 13.1-7 D 3/4.6.15 RADIOACTIVE EFFLUENTS ........................................ 13.1-13 D 3/4.6.15.a.(1) Liquid Concentration ........................................ 13.1-13 D 314.6.15Sa(2) Liquid Dose. ........................................ 3.1-14 D 3/4.6.15.b.(1) Gaseous Dose Rate ........................................ I 3.1.18 D 3/4.6.15.b.(2) Gaseous Air Dose ......................................... 13.1-19 D 3/4.6.15.b.(3) Gaseous Tritium, lodines and Particulates ......................................... 13.1-20 D 314.6.15.d Uranium Fuel Cycle ......................................... I 3.1-23 D 3/4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS ......................... I 3.1-26 D 3/4.6.16.a Liquid ......................................... I 3.1-26 D 3/4.6.16.b Gaseous ......................................... I 3.1-26 D 3/4.6.17 Not Used D 3/4.6.18 MARK I CONTAINMENT ........................................ 13.1-28 D 3/4.6.19 LIQUID WASTE HOLDUP TANKS ........................................ I 3.1-29 D 3/4.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM... I 3.1-30 D 3/4.6.21 INTERLABORATORY COMPARISON PROGRAM . ...................... I 3.1-40 D 3/4.6.22 LAND USE CENSUS .......................................... 3.1-41 Unit 1 ODCM Revision 24 iii July 2003

ODCM - NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Cont)

PAGE R ACICQ AJX J.!... 1 TV I A.P 1 E 1J -W1 B 3/4.6.14 BASES FOR RADIOACTIVE EFFLUENT INSTRUMENTATION ............... I B 3.1-1 B 3/4.6.15 BASES FOR RADIOACTIVE EFFLUENTS . ..................................

I B 3.1-1 Liquiid Concentration .................................................... I B 3.1-1 Liquid Dose .................................................... I B 3.1-2 Gaseous Dose Rate .................................................... I B 3.1-3 Dose-Noble Gases .................................................... I B 3.14 Dose-Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form .................................................... I B 3.1-5 Total Dose-Uranium Fuel Cycle ............................................. I B 3.1-6 B 3/4.6.16 BASES FOR RADIOACTIVE EFFLUENT TREATMENT SYSTEMS ........... I B 3.1-7 Liquid Radwaste Treatment System ................................................... I B 3.1-7 Gaseous Effluent Treatment Systems ................................................... I B 3.1-7 B 3/4.6.18 BASES FOR MARK I CONTAINMENT ................................................... I B 3.1-8 B 3/4.6.19 BASES FOR LIQUID WASTE HOLDUP TANKS .............................................. I B 3.1-8 B 3/4.6.20 BASES FOR RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ................................................... I B 3.1-9 B 3/4.6.21 BASES FOR INTERLABORATORY COMPARISON PROGRAM ............... I B 3.1-10 B 3/4.6.22 BASES FOR LAND USE CENSUS ................................................... I B 3.1-11 SECTION 5.0 Not Used SECTION 6.0 ADMINISTRATIVE CONTROLS, Iv.*.

...........

v.e*.*.**...................... ................. .16.0-1 Reporting Requirements .............. .......................................................................... .I6.0-2 S~npni

- it. JC; Pes Rnrt4

xspV . ...................................................

T Kn-A I v.v-s Unit 1 ODCM Revision 24 iv July 2003

ODCM - NINE MILE POINT UNIT 1 TABLE OF CONTENTS (Cont)

PAGE

'a if .-

PART II - Calculational . -

IMMOUoao0gSl 1 -

ui...

-I--

I 1.0 LIQUID EFFLUENTS . ................................................................................................ .112 1.1 Setpoint Determinations s.............................................................................................I .112 1.1.1 Basis. ................................................................................................ .112 1.1.2 Service Water System Elffluent Line Alarm Setpoint............................................. .112 1.13 Liquid Radwaste Effluei it Line Alarm Setpoint........ ................................................. ,113 1.1.4 Discussion. ................................................................................................ .,u5 1.1.4.1 Control of Liquid Effluent Batch Discharges .. ................................... I 5 1.1.4.2 Simultaneous Discharges of Radioactive Liquids .................................. I 15 1.1.4.3 Sample Representativeness .................................. 15 1.1.4.4 Liquid Radwaste System Operation .................................. I1 6 1.1.4.5 Service Water System Contamination ................................... I 7 1.2 Liquid Effluent Concentration Calculation .................................. II 7 1.3 Dose Determinations .................................. II 8 1.3.1 Maximum Dose Equivalent Pathway .................................. II 8 1.3.2 Dose Projections - Determination of Need to Operate the Liquid Radwaste Treatment System ..................... H 11 2.0 GASEOUS EFFLUENTS .................... 1 12 2.1 Setpoint Determinations ...................................... I 12 2.1.1 Basis ....................................... 12 2.1.2 Stack Monitor Setpoints ...................................... I 12 2.1.3 Recombiner Discharge (Off Gas) Monitor Setpoints ..................................... II 14 2.1.4 Emergency Condenser Vent Monitor Setpoint . . .1......................

II i Unit 1 ODCM Revision 24 v July 2003

ODCM - NINE MILE POINT UNIT I TABLE OF CONTENTS (Cont)

PAGE 2.1.5 Discussion ............................................. . II 15 2.1.5.1 Stack Effluent Monitoring System Description . ...................................I 15 2.1.5.2 Stack Sample Flow Path - RAGEMS Auxiliary Sample Point . . 1 15 2.1.53 Stack Sample Flow Path - OGESMS ................... .......................... II 16 2.1.5.4 Sample Frequency/Sample Analysis ............................................... II 16 2.1.5.5 I-133 and I-135 Estimates ............................................. I 16 2.1.5.6 Gaseous Radwaste Treatment System Operation ............................................. II 17 2.2 Dose and Dose Rate Determinations .............................................. II 17 2.2.1 Dose Rate ............................................. II 18 2.2.1.1 Noble Gases ............................................. I 19 2.2.1.2 Tritium, Iodines and Particulates ............................................... II 20 2.2.2 Dose .............................................. H 22 2.2.2.1 Noble Gas Air Dose ............................................... I 22 2.2.2.2 Tritium, Iodines and Particulates ............................................. II 23 2.2.2.3 Accumulating Doses ............................................. II 24 2.2.3 Dose Projections - Determination of Need to Operate Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System ............................ 1 24 2.3 Critical Receptors ............................................. II 25 2.4 Refinement of Offsite Doses Resulting From Emergency Condenser Vent Releases .. I 26 Unit 1 ODCM Revision 24 vi July 2003

ODCM- NINE MILE POINT UNIT I TABLE OF CONTENTS (Cont)

PAGE 3.0 40 CFR 190 REQUIREMENTS. ,................................ II 27 .

3.1 Evaluation of Doses From Liquid Effluents......................... ,................................ II 28 .

3.2 Evaluation of Doses From Gaseous Effluents ...................... ,................................ II 29 .

3.3 Evaluation of Doses From Direct Radiation ........................ nT2fn

.V .

3A Doses to Members of the Public Within the Site Boundary 30 .

4.0 ENVIRONMENTAL MONITORING PROGRAM................................................................... 1 33 4.1 Sampling Stations .I 33 4.2 Interlaboratory Comparison Program .1 33 4.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements .11 34 Appendix A Liquid Dose Factor Derivation (A)...........................................................................H175 Appendix B Plume Shine Dose Factor Derivation (B. and V) ........................................................II78 Appendix C Organ Dose Parameters for Iodine - 131 & 133, Particulates and Tritium (R.)............................................ n 0O1 Ou 0& ....

Appendix D Diagrams of Radioactive Liquid and Gaseous Effluent Treatment Systems and Monitoring Systems................. TVall

... 7A .

Unit 1 ODCM Revision 24 vii July 2003

ODCM - NINE MILE POINT UNIT 1 LIST OF TABLES PART I - Radiological Effluent Controls PAGE D 3.6.14-1 Radioactive Liquid Effluent Monitoring Instrumentation ........................ I 3.1-3 D 4.6.14-1 Radioactive Liquid Effluent Monitoring Instrumentation - SR ............... I 3.1-5 D 3.6.14-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation............................................................................................. 13.1-8 D 4.6.14-2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation - SR................................................ I 3.1-11 D 4.6.15-1 Radioactive Liquid Waste Sampling and Analysis Program - SR .......... I 3.1-15 D 4.6.15-2 Radioactive Gaseous Waste Sampling and Analysis Program - SR ....... I 3.1-21 D 3.6.20-1 Operational Radiological Environmental Monitoring Program ............. I 3.1-33 D 4.6.20-1 Detection Capabilities for Environmental Sample Analysis Lower Limit of Detection LLD - SR............................................... I 3.1-37 D 6.9.3-1 Reporting Level for Radioactivity Concentration in Environmental Samples ............................................... I 6.0-5 PART II - Calculational Methodologies Table 1-1 Average Energy Per DisintegrationI................................................................ 136 Tables 2-1 Aht Values for the NMP-1 Facility . ....... .......................... 1 37 to 2-8 Table 3-1 Critical Receptor Dispersion Parammeters for Ground Level and Elevated Releases.................... 45 .

Table 3-2 Gamma Air and Whole Body Plume Shine Dose Factors for Noble Gases (Bi and VO ...................... ,........................................... 1146 Table 3-3 Immersion Dose Factors for Noble Gases ............ ,........................................... 1147 Tables 3-4 to 3-22 Dose and Dose Rate Factors (R.) .......................... .......................................... .1148 Table 3-23 Parameters for the Evaluation of Doses to Real Members of the Public from Gaseous and Liquid Effluents .................................................................... ......................................... .II67 Table 5.1 Nine Mile Point Nuclear Station Radiological Environmental Monitoring Program Sampling Tldrfinnc 1 68 w ......................................................

Unit 1 ODCM Revision 24 viii July 2003

ODCM - NINE MILE POINT UNIT 1 LIST OF FIGURES PAGE Figure 5.1-1 Nine Mile Point On-Site Map ,...................................................................... *....1172 Figure 5.1-2 Nine Mile Point Offsite Map.. F....................................................................... .....1173 Figure 5.13-1 Site Boundaries....................... .................................................................. .....II 74 Figure D-0 Piping Instrument and Equipment Symbols .

" uA .

Figure D-1 Radioactive Waste Disposal........................... T% .1 V-l ................................

- A Figure D-2 Stm Packing, Exhauster, and Recombiner... v-] .

Figure D-3 Reactor Building Vent System...................... ..................... D-3 .

Figure D-4 Waste Disposal Building Vent System.......... 1 U-.

A Figure D-5 NMP-1 Stack.................................................. 1Sw.V .

Figure D-6 Offgas Building Vent System......................... ..................................................... ..D-6 Figure D-7 This Page/Figure Deleted Figure D-8 Stack Sample and Sample Return................. f-U

"-a .

Figures D-9, D-10 These Pages/Figures Deleted Figure D-11 OGESMS Schematic...................................... D-11 .

Unit 1 ODCM Revision 24 ix July 2003

INTRODUCTION The Offsite Dose Calculation Manual (ODCM) provides the methodology to be used for demonstrating compliance with 10 CFR 20, 10 CFR 50, and 40 CFR 190. The contents of the ODCM are based on Draft NUREG-0472, Revision 3, "Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors," September 1982; Draft NUREG-0473, Revision 2, "Radiological Effluent Technical Specifications for BWR's", July 1979; NUREG 0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants,"

October 1978; the several Regulatory Guides referenced in these documents; and, communication with the NRC staff.

Should it be necessary to revise the ODCM, these revisions will be made in accordance with Technical Specifications.

The Offsite Dose Calculation Manual (ODCM) is a supporting document of the Technical Specifications Section 6.5.1, "Offsite Dose Calculation Manual." 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 I, includes the following: (1) The Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specifications 6.5.3, "Radioactive Effluent Controls Program" and 6.5.1, "Offsite Dose Calculation Manual", respectively, 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 6.6.2, "Annual Radiological Environmental Operating Report" and 6.6.3 "Radioactive Effluent Release Report". Calculational Methodologies, Part II, describes 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.

Unit 1 ODCM Revision 24 x July 2003

PART I - RADIOLOGICAL EFFLUENT CONTROLS Unit 1 ODCM Revision 24 I July 2003

PART I - RADIOLOGICAL EFFLUENT CONTROLS Section 1.0 Definitions Unit 1 ODCM Revision 24 I 1.0-0 July 2003

1.0 DEFINITIONS DEFINITIONS 1.0 NOTE:

Technical Specifications defined terms and the following additional defined terms are applicable throughout these controls and bases.

Gaseous Radwaste Treatment System A gaseous radwaste treatment system is any system designed and installed to reduce radioactive gaseous effluents by collecting main condenser offgas and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

Member(s) of the Public Member(s) of the public shall include persons who are not occupationally associated with the Nine Mile Point Nuclear Station. This category does not include employees of owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant, their contractors or vendors who are occupationally associated with Nine Mile Point Unit 1. 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 Unit 1.

Milk SamDling Location A milk sampling location is that location where 10 or more head of milk animals are available for the collection of milk samples.

Offsite Dose Calculation Manual (ODCM)

The Offsite Dose Calculational Manual shall contain the current methodology and parameters used in the calculation of offsite doses due to 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 the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specifications 6.6.2, "Annual Radiological Environmental Operating Reporr and 6.6.3, "Radioactive Effluent Release Report, and Controls D 6.9.1.d and D 6.9.1.e.

Unit 1 ODCM Revision 24 I 1.0-1 July 2003

Purae - Purging DEFINITIONS 1.0 Purge or purging is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement. The purge is completed when the oxygen concentration exceeds 19.5 percent.

Site Boundary The site boundary shall be that line around the Nine Mile Point Nuclear Station beyond which the land is neither owned, leased, nor otherwise controlled by the owners and operators of Nine Mile Point Nuclear Station and James A. Fitzpatrick Nuclear Power Plant.

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 Area The unrestricted area shall be any area at or beyond the site 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. That area outside the restricted area (10 CFR 20.1003) but within the site boundary will be controlled by the owner as required.

Unit 1 ODCM Revision 24 I 1.0-2 July 2003

Ventilation Exhaust Treatment System DEFINTIONS 1.0 A ventilation exhaust treatment system is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing lodines 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 Feature (ESF) atmospheric cleanup systems are not considered to be ventilation exhaust treatment system components.

Ventina Venting is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, 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 1 ODCM Revision 24 I 1.0-3 July 2003

PART I - RADIOLOGICAL EFFLUENT CONTROLS Sections 3.0/4.0 Applicability Unit 1 ODCM Revision 24 I3.0-0 July 2003

3.0 CONTROLS APPLICABILITY 3.0/4.0 The Offsite Dose Calculation Manual (ODCM) Part 1,Radiological Effluent Controls, is subject to Technical Specifications Section 3.0 requirements, as applicable.

4.0 SURVEILLANCE REQUIREMENTS The ODCM Part 1,Radiological Effluent Controls, is subject to Technical Specifications Section 4.0 requirements, as applicable.

Unit 1 ODCM Revision 24 13.0-1 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 CONTROLS 9.

SURVEILLANCE REQUIREMENT DLCO 3.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION DSR 4.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION ApDlicability: Aodlicability:

Applies to the operability of plant instrumentation that Applies to the surveillance of instrumentation that monitors monitors plant effluents. plant effluents.

Obiective: Obiective:

To assure the operability of instrumentation to monitor To verify operation of monitoring instrumentation.

the release of radioactive plant effluents.

Sfecification: Specification:

a. Liquid Effluent a. Liquid Effluent The radioactive liquid effluent monitoring instru- Each radioactive liquid effluent monitoring mentation channels shown in Table D 3.6.14-1 instrumentation channel shall be demonstrated operable shall be operable with their alarm setpoints set to by performance of the sensor check, source check, ensure that the limits of Control instrument channel calibration and channel test DLCO 3.6.15.a.1 are not exceeded. The alarm operations at the frequencies shown in Table D 4.6.14-1.

setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in Part II.

With a radioactive liquid effluent monitoring Records - Auditable records shall be maintained, in instrumentation channel alarm setpoint less accordance with procedures in Part II, of all radioactive conservative than a value which will ensure that the liquid effluent monitoring instrumentation alarm limits of DLCO 3.6.15.a.1 are met, immediately setpoints. Setpoints and setpoint calculations shall be suspend the release of radioactive liquid effluents available for review to ensure that the limits of Control monitored by the affected channel, or declare the DLCO 3.6.15.a.1 are met.

channel inoperable, or change the setpoint so it is acceptably conservative.

Unit 1 ODCM Revision 24 I3.1-1 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 CONTROLS 9.

SURVEILLANCE REQUIREMENT With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels operable, take the action shown in Table D 3.6.14-1. Restore the instruments to OPERABLE status within 30 days, or outline in the next Radioactive Effluent Release Report the cause of the inoperability and how the instruments were or will be restored to operable status.

Unit 1 ODCM Revision 24 I3.1-2 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 TABLE D 3.6.14-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Instrument Minimum Channels Operable ADplIcablity

1. Gross Radioactivity Monitors(a)

A. Liquid Radwaste Effluent Line 1(C) At all times(b)

B. Service Water System Effluent Line 1 (d At all times0 )

2. Flow Rate Measurement Devices A. Uquid Radwaste Effluent Une 1 (e) At all times B. Discharge Canal ** **

3. Tank Level Indicating Devices(g)

A. Outside Liquid Radwaste Storage Tanks 1M' At all times

- Pumps curves or rated capacity will be utilized to estimate flow.

Unit 1 ODCM Revision 24 13.1-3 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 NOTES FOR TABLE D 3.6.14-1 (a) Provide alarm, but do not provide automatic termination of release.

(b) An operator shall be present in the Radwaste Control Room at all times during a release.

(c) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases may continue provided that prior to initiating a release:

1. At least two independent samples are analyzed in accordance with Specification DSR 4.6.15.a, and

2. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and discharge line valving.

Otherwise suspend release of radioactive effluents via this pathway.

(d) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for gamma radioactivity at a lower limit of detection of at least 5x10-7 microcurie/ml.

(e) During discharge, with the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases.

(f) With the number of channels operable less than required by the minimum channels operable requirement, liquid additions to this tank may continue provided the tank liquid level is estimated during liquid additions to the tank.

(g) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents.

(h) Deleted.

(i) Monitoring will be conducted continuously by alternately sampling the reactor building and turbine building service water return lines for approximately 15-minute intervals.

Unit 1 ODCM Revision 24 I 3.1-4 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 TABLE D 4.6.14-1 RADIOA LCTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirement Instrument Sensor Check Source Checks Channel Test Channel Calibration

1. Gross Beta or Gamma Radioactivity Monitors

a. Liquid Radwaste Effluent Line Once/day* Once/discharge* Once/3 months(a)* Once/year(b)*

b. Service Water Effluent Line Once/day Once/month Once/3 months(a) Once/year"')

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line Once/day(c) None None Once/year

b. Discharge Canal(d None None None Once/year

3. Tank Level Indicating Devices(e)

a. Outside Uquid Radwaste Storage Tanks Once/day** None Once/3 months Once/1 8 months

Required prior to removal of blank flange in discharge line and until blank flange is replaced.

- During liquid addition to the tank.

Unit 1 ODCM Revision 24 I3.1-5 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - LIQUID D 3/4.6.14 NOTES FOR TABLE D 4.6.14-1 (a) The channel test shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:

1. Instrumentation indicates measured levels above the alarm setpoint.

2. Instrument indicates a downscale failure.

3. Instrument controls not set in operate mode.

(b) The channel calibration shall be performed using one or more reference standards certified by the National Institute of Standards and Technology (NIST), or using standards that are traceable to the NIST or using actual samples of liquid waste 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.

(c) Sensor check shall consist of verifying indication of flow during periods of release. Sensor check shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic or batch releases are made.

(d) Pump performance curves or rated data may be used to estimate flow.

(e) Tanks included in this specification are those outdoor tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents.

(f) Source check may consist of an installed check source, response to an external source, or (for liquid radwaste monitors) verification within 30 minutes of commencing discharge of monitor response to effluent.

Unit 1 ODCM Revision 24 I 3.1-6 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 CONTROLS 4.

SURVEILLANCE REQUIREMENT

b. Gaseous Process and Effluent b. Gaseous Process and Effluent The radioactive gaseous process and effluent Each radioactive gaseous process and monitoring instrumentation channels shown in Table effluent monitoring instrumentation channel D 3.6.14-2 shall be operable with their alarm setpoints shall be demonstrated operable by set to ensure that the limits of Control DLCO performance of the sensor check, source 3.6.15.b.1 are not exceeded. The alarm setpoints of check, instrument channel calibration and these channels shall be determined and adjusted in instrument channel test operations at the accordance with the methodology and parameters in frequencies shown in Table D 4.6.14-2.

Part II.

Auditable records shall be maintained of With a radioactive gaseous process and effluent the calculations made, in accordance with monitoring instrumentation channel alarm setpoint procedures in Part II, of radioactive less conservative than required by the above gaseous process and effluent monitoring specification, immediately suspend the release of instrumentation alarm setpoints. Setpoints radioactive gaseous effluents monitored by the and setpoint calculations shall be available affected channel, or declare the channel inoperable, for review to ensure that the limits of or change the setpoint so it is acceptably Control DLCO 3.6.15.b.1 are met.

conservative.

With less than the minimum number of radioactive gaseous process and effluent monitoring instrumentation channels operable, take the action shown in Table D 3.6.14-2. Restore the instruments to OPERABLE status within 30 days or outline In the next Radioactive Effluent Release Report the cause of the inoperability and how the instruments were or will be restored to operable status.

Unit 1 ODCM Revision 24 I3.1-7 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 TABLE D 3.6.14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Minimum Instrument Channels Operable Applicability Action Stack Effluent Monitoring

a. Noble Gas Activity Monitor 1 * (a)

b. Iodine Sampler Cartridge 1 * (b)

c. Particulate Sampler Filter 1 * (b)

d. Sampler Flow Rate Measuring 1 * (c)

Device

e. Stack Gas Flow Rate 1 * (c), (d)

Measuring Device

2. Deleted

At all times.

- Note Deleted.

Unit 1 ODCM Revision 24 13.1-8 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 TABLE D 3.6.14-2 (cont'd)

RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Minimum Instrument Channels Operable Applicability Action

3. Condenser Air Ejector Radioactivity Monitor (Recombiner Discharge or Air Ejector Discharge)

a. Noble Gas Activity Monitor 1 (g)

b. Offgas System Flow Rate 1 (c)

Measuring Device

c. Sampler Flow Rate Measuring 1 (c)

Device

4. Emergency Condenser System

a. Noble Gas Activity Monitor 1 per vent (h)

During operation of the main condenser air ejector During power operating conditions and whenever the reactor coolant temperature is greater than 2121F except for hydrostatic testing with the reactor not critical.

Unit 1 ODCM Revision 24 I 3.1-9 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 NOTES FOR TABLE D 3.6.14-2 (a) With the number of channels operable less than required by the minimum channels operable requirement, effluent releases via this pathway may continue provided grab samples are taken once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(b) With the number of channels operable less than required by the minimum channels operable requirements, effluent releases via this pathway may continue provided that samples are continuously collected with auxiliary sampling equipment starting within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of discovery in accordance with the requirements of Table D 4.6.15-2.

(c) With the number of channels operable less than required by the minimum channels operable requirements, effluent releases via this pathway may continue provided the flow rate is estimated once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

(d) Stack gas flow rate may be estimated by exhaust fan operating configuration.

(e) Deleted (1) Deleted (g) With the number of channels operable less than required by the minimum channels operable requirement, gases from the main condenser offgas treatment system may be released provided:

1. Offgas grab samples are collected and analyzed once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. The stack monitor is operable.

3. Otherwise, be in at least hot shutdown within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(h) With the number of channels operable less than required by the minimum channels operable requirements, steam release via this pathway may commence or continue provided vent pipe radiation dose rates are monitored once per four hours.

Unit 1 ODCM Revision 24 1 3.1-10 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 TABLE D 4.6.14-2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION Surveillance Requirements Instrument Sensor Check Source Check Channel Test Channel Calibration

1. Stack Effluent Monitoring System

a. Noble Gas Activity Monitor Once/daya) Once/month Once/3 months(e) Once/year(b)

b. Iodine Sampler Cartridge None None None None

c. Particulate Sampler Filter None None None None

d. Sampler Flow Rate Measuring Device Once/day(a) None None Once/year

e. Stack Gas Flow Rate Measuring Device Once/day None None Once/year

2. Deleted

3. Condenser Air Ejector Radioactivity Monitor (Recombiner Discharge or Air Elector Discharge)

a. Noble Gas Activity Monitor Once/daVO Once/month Once/operating Once/year°b cycle(c)

b. Offgas System Flow Rate Once/day( None None Once/year Measuring Device Once/day' None None Once/year

c. Sampler Flow Rate Measuring Device

4. Emergency Condenser System Once/dayh) Once/month Once/3 months(9) Once/operating cyce(b)

a. Noble Gas Activity Monitor Unit 1 ODCM Revision 24 13.1-11 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION - GASEOUS D 3/4.6.14 NOTES FOR TABLE D 4.6.14-2 (a) At all times.

(b) The 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 the NIST or using actual samples of gaseous effluent 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.

(c) The channel function test shall demonstrate that control room alarm annunciation occurs if either of the following conditions exist:

1) Instrument indicates measured levels above the Hi or Hi Hi alarm setpoint.

2) Instrument Indicates a downscale failure.

The channel function test shall also demonstrate that automatic isolation of this pathway occurs if either of the following conditions exist:

1) Instruments indicate two channels above Hi Hi alarm setpoint.

2) Instruments indicate one channel above Hi Hi alarm setpoint and one channel downscale.

(d) Deleted (e) Deleted (f) During operation of the main condenser air ejector.

(g) The channel test shall produce upscale and downscale annunciation.

(h) During power operating conditions and whenever the reactor coolant temperature is greater than 2120 F except for hydrostatic testing with the reactor not critical.

Unit 1 ODCM Revision 24 I3.1-12 July 2003

RADIOACTIVE EFFLUENTS - LIQUID CONCENTRATION D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.15 RADIOACTIVE EFFLUENTS DSR 4.6.15 RADIOACTIVE EFFLUENTS ADDlicabilit : ApDlicabilitv:

Applies to the radioactive effluents from the station. Applies to the periodic test and recording requirements of the station process effluents.

Oboective: Obiective:

To assure that radioactive material is not released to the To ascertain that radioactive effluents from the station are environment in any uncontrolled manner and is within the limits within the allowable values of 10CFR20, Appendix B and of 10CFR20 and 10CFR50 Appendix I. 10CFR50, Appendix I.

ftecication:

Specification:

a. Liquid

a. Liquid (1) Concentration (1) Concentration The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to ten times the concentrations specified In Radioactive liquid wastes shall be sampled and 10CFR Part 20, Appendix B, Table 2, Column 2 for analyzed according to the sampling and analysis radionuclides other than dissolved or entrained program of Table D 4.6.15-1.

noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 The results of the radioactivity analyses shall be microcuries/ml total activity. used in accordance with the methodology and parameters in Part II to assure that the Should the concentration of radioactive material concentrations at the point of release are released in liquid effluents to unrestricted areas maintained within the limits of Control DLCO exceed the above limits, restore the concentration 3.6.15.a.(1).

to within the above limits immediately.

Unit 1 ODCM Revision 24 I 3.1-13 July 2003

RADIOACTIVE EFFLUENTS - LIQUID DOSE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT (2) Dose (2) Dose The dose or dose commitment to a member of the Cumulative dose contributions from liquid public from radioactive materials in liquid effluents effluents for the current calendar quarter and the released, from each reactor unit, to unrestricted current calendar year shall be determined in areas (see Figures 5.1-1) shall be limited: accordance with the methodology and parameters in Part II monthly.

(a) During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and (b) During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ.

With the calculated dose from the release of radioactive materials In liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit(s) and 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 the above limits.

Unit 1 ODCM Revision 24 I3.1-14 July 2003

RADIOACTIVE EFFLUENTS - LIQUID D 3/4.6.15 TABLE D 4.6.15-1 RADIOACTIVE UQUID WASTE SAMPLING AND ANALYSIS PROGRAM Surveillance Requirement Minimum Sampling Lower Limit(") of Detection Liquid Release Type Frequency Analysis Frequency Type of Activity Analysis (LLD) (pCI/ml)

A. Batch Waste°b Tanks *

Each Batch Each Batch PRncipal Gamma(c) 5 x 10'7 Emitters 1-131 1 x 10.6 Each Batch(d) Each Batch(d) Dissolved and Entrained 1 x 10.6 Gases (Gamma Emitters)

Monthly H-3 1X Each Batch Composite1 Gross Alpha 1 X107

Quarterly Sr-89, Sr-90 5 x 1098 Each Batch Composite(e)

Fe-55 1lx 1046 B. Service Water Once/month(O Once/monthm0 Principal Gamma(c) Emitters 5x 10-7 System Effluent 1-131 1x 1046 Dissolved and Entrained 1 x 1045 Gases H-3 I x 105 Gross Alpha 1x 10i7 Once/quarter(0 Once/quarterO9 Sr-89, Sr-90 5 x 10.6 Fe-55 1 x 104

Completed prior to each release.

Unit 1 ODCM Revision 24 I 3.1-15 July 2003

RADIOACTIVE EFFLUENTS - LIQUID D 3/4.6.15 NOTES FOR TABLE D 4.6.15-1 (a) The LLD is defined, for purposes of these specifications, 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 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real signal. For a particular measurement system which may include radiochemical separation:

LLD = 4.66 St 6

E*V.2.22 x 10%Yeexp (-XAt)

Where:

LLD is the Oa priori lower limit of detection as defined above, as microcuries per unit mass or volume, Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield, when applicable, X is the radioactive decay constant for the particular radionuclide, and At for plant effluents Is the elapsed time between the midpoint of sample collection and time of counting.

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

Itshould 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 for a particular measurement.

Unit 1 ODCM Revision 24 I 3.1-16 July 2003

RADIOACTIVE EFFLUENTS - LIQUID D 3/4.6.15 NOTES FOR TABLE D 4.6.15-1 (b) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be Isolated and then thoroughly mixed to assure representative sampling.

(c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-1 34, Cs-1 37, Ce-141 and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report.

(d) If more than one batch is released in a calendar month, only one batch need be sampled and analyzed during that month.

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

(f) Ifthe alarm setpoint of the service water effluent monitor, as determined by the method presented in Part II, is exceeded, the frequency of sampling shall be increased to daily until the condition no longer exists. Frequency of analysis shall be increased to daily for principal gamma emitters (including dissolved and entrained gases) and an incident composite for H-3, gross alpha, Sr-89, Sr-90 and Fe-55.

Unit 1 ODCM Revision 24 I 3.1-17 July 2003

RADIOACTIVE EFFLUENTS - GASEOUS DOSE RATE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT

b. Gaseous b. Gaseous (1) Dose Rate (1) Dose Rate The dose rate due to radioactive materials The dose rate due to noble gases in gaseous released In gaseous effluents from the site to effluents shall be determined to be within the limits of areas at or beyond the site boundary shall be Control DLCO 3.6.15 in accordance with the limited to the following: methodology and parameters in Part II.

(a) For noble gases: Less than or equal The dose rate due to iodine-1 31, iodine-1 33, tuitium to 500 mrems/year to the whole body and all radionuclides in particulate form with half lives and less than or equal to 3000 greater than 8 days in gaseous effluents shall be mrems/year to the skin, and determined to be within the limits of Control DLCO 3.6.15 in accordance with methodology and (b) For iodine-1 31, iodine-1 33, tritium and parameters in Part II by obtaining representative all radionuclides in particulate form samples and performing analyses in accordance with with half lives greater than 8 days: the sampling and analysis program specified in Table Less than or equal to 1500 D 4.6.15-2.

mrems/year to any organ.

With the dose rate(s) exceeding the above limits, without delay restore the release rate to within the above limits(s).

Unit 1 ODCM Revision 24 13.1-18 July 2003

RADIOACTIVE EFFLUENTS - GASEOUS DOSE D 3/4.6.15 CONTROLS 9 SURVEILLANCE REQUIREMENT (2) Air Dose (2) Air Dose The air dose due to noble gases released in gaseous Cumulative dose contributions for the current effluents, from each reactor unit, to areas beyond the calendar quarter and current calendar year for site boundary shall be limited to the following: noble gases shall be determined monthly in accordance with the methodology and (a) During any calendar quarter Less than or parameters in Part II.

equal to 5 milllroentgen for gamma radiation and less than or equal to 10 mrads for beta radiation and, (b) During any calendar year Less than or equal to 10 milliroentgen for gamma radiation and less than or equal to 20 mrads for beta radiation.

With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that identifies the cause(s) for exceeding the limit(s) and 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 the above limits.

Unit 1 ODCM Revision 24 I 3.1-19 July 2003

RADIOACTIVE EFFLUENTS - GASEOUS DOSE D 3/4.6.15 CONTROLS 4 SURVEILLANCE REQUIREMENT (3) Tritium, lodines and Particulates (3) Tritium, lodines and Particulates The dose to a member of the public from iodine-1 31, Cumulative dose contributions for the current iodine-1 33, tritium and all radionuclides in particulate calendar quarter and current calendar year for form with half lives greater than 8 days in gaseous iodine-131, iodine-133, tritium and radionuclides effluents released, from each reactor unit, to areas in particulate form with half lives greater than 8 beyond the site boundary shall be limited to the days shall be determined monthly in accordance following: with the methodology and parameters in Part II.

(a) During any calendar quarter Less than or equal to 7.5 mrems to any organ and, (b) During any calendar year Less than or equal to 15 mrems to any organ.

With the calculated dose from the release of iodine-131, iodine-133, tritium and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that Identifies the cause(s) for exceeding the limit and 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 the above limits.

Unit 1 ODCM Revision 24 13.1-20 July 2003

RADIOACTIVE EFFLUENTS - GASEOUS D 3/4.6.15 TABLE D 4.6.15-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Surveillance Requirements MInimum Lower Llmitao) of Detection Gaseous Release Type Sampling Frequency Analysis Frequency Type of Activity Analysis (LLD) (pClImI)

U - U A. Containment Purge(b) Each Purge Prior to each release Principal Gamma Emitters(c) 1 x 10i4 Grab Sample Each Purge Principal Gamma Emitters(c) 1 X 10-4 H-3 1 X 10 4 B. Stack Once/Month(d) Once/Month(d) Principal Gamma Emitters(c) 1 X104 Once/Month(h) Once/Month H-3 1 x 10 4 C. Stack Confinuous(e) Once/Week(O 1-131 1 x 10.12 Charcoal Sample Continuous(e) OnceANeek(O Principal Gamma Emitters(c) 1 x 10-11 Particulate Sample Continuous(e) Once/Month Composite Gross alpha, Sr-89, Sr-90 1 x 10.1 Particulate Sample Continuous(e) Noble Gas Monitor Noble Gases, Gross Gamma 1 x I0 5()

or Principal Gamma Emitters(c)

Unit 1 ODCM Revision 24 I 3.1-21 July 2003

RADIOACTIVE EFFLUENTS - GASEOUS D 3/4.6.15 NOTES FOR TABLE D 4.6.15-2 (a) The LLD is defined in notation (a)of Table D 4.6.15-1.

(b) Purge is defined in Section 1.0.

(c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-1 33, Xe-1 35 and Xe-1 38 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-1 34, Cs-1 37, Ce-1 41,1-131 and Ce-1 44 for particulate emissions.

This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Radioactive Effluent Release Report pursuant to Technical Specification 6.6.3, Radioactive Effluent Release Report, and Control D 6.9.1.

(d) Sampling and analysis shall also be performed following shutdown, startup or an increase on the recombiner discharge monitor of greater than 50 percent, factoring out increases due to changes in thermal power level or dilution flow; or when the stack release rate is in excess of 1000 .Ci/second and steady-state gaseous release rate increases by 50 percent.

(e) The sample flow rate and the stack flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls DLCO 3.6.15.b.(l).(b) and DLCO 3.6.15.b.(3).

(f) When the release rate is in excess of 1000 ,.Ci/sec and steady state gaseous release rate increases by 50 percent, the iodine and particulate collection device shall be removed and analyzed to determine the changes in iodine-131 and particulate release rate. The analysis shall be done daily following each change until it is shown that a pattern exists which can be used to predict the release rate; after which it may revert to weekly sampling frequency. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor of 10.

(g) When the continuous Noble Gas Monitor is inoperable the LLD for noble gas gamma analysis shall be 1 x 10'4 J.CIcc.

(h) Tritium grab samples shall be taken weekly from the station ventilation exhaust (stack) when fuel Is offloaded until stable tritium release levels can be demonstrated.

Unit 1 ODCM Revision 24 I 3.1-22 July 2003

RADIOACTIVE EFFLUENTS - MAIN CONDENSER, URANIUM FUEL CYCLE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT

c. Deleted c. Main Condenser The radioactivity rate of noble gases at the recombiner discharge shall be continuously monitored in accordance with Table D 3.6.14-2.

d. Uranium Fuel Cycle d. Uranium Fuel Cycle The annual (calendar year) dose or dose Cumulative dose contributions from liquid and commitment to any member of the public due gaseous effluents shall be determined in to releases of radioactivity and to radiation accordance with Controls from uranium fuel cycle sources shall be DSR 4.6.15.a.(2), DSR 4.6.15.b.(2) and limited to less than or equal to 25 mrems to DSR 4.6.15.b.(3) and in accordance with the the whole body or any organ, except the methodology and parameters in Part II.

thyroid, which shall be limited to less than or equal to 75 mrems.

Unit 1 ODCM Revision 24 13.1-23 July 2003

RADIOACTIVE EFFLUENTS - URANIUM FUEL CYCLE D 3/4.6.15 CONTROLS U.

SURVEILLANCE REQUIREMENT With the calculated doses from the release of radioactive Cumulative dose contributions from direct radiation from the materials in liquid or gaseous effluents exceeding twice the reactor units and from radwaste storage tanks shall be limits of Controls DLCO 3.6.15.a(2), DLCO 3.6.15.b(2) and determined in accordance with the methodology and parameters DLCO 3.6.15.b(3), calculations shall be made including direct in Part II. This requirement is applicable only under conditions radiation contributions from the reactor units and from outside set forth in Control DLCO 3.6.15.d.

storage tanks to determine whether the above listed 40CFR1 90 limits have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.2203(a)(4), shall include an analysis that estimates the radiation exposure (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) covered by this report.

Unit 1 ODCM Revision 24 I 3.1-24 July 2003

RADIOACTIVE EFFLUENTS - URANIUM FUEL CYLE D 3/4.6.15 CONTROLS SURVEILLANCE REQUIREMENT 9

It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting In violation of 40CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40CFR 190. Submittal of the report is considered a timely request and a variance is granted until staff action on the request is complete.

Unit 1 ODCM Revision 24 I 3.1-25 July 2003

RADIOACTIVE EFFLUENT TREATMENT SYSTEMS D 3/4.6.16 CONTROLS SURVEILLANCE REQUIREMENT 9

DLCO 3.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS DSR 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Aimlicabilitv Applicability:

Applies to the operating status of the liquid and gaseous Applies to the surveillance requirements for the liquid and effluent treatment systems. gaseous effluent treatment systems.

Obiective: Obiective:

To assure operability of the liquid and gaseous effluent To verify operability of the liquid and gaseous effluent treatment system. treatment system.

Soecification: Specification:

a. Uquid a. Liquid The liquid radwaste treatment system shall be used to Doses due to liquid releases to unrestricted areas reduce the radioactive materials in liquid wastes prior shall be projected prior to the release of each batch of to their discharge when the projected dose due to the liquid radioactive waste in accordance with the liquid effluent, from each unit, to the Unrestricted methodology and parameters in Part II.

Areas would exceed 0.06 mrem to the total body or 0.2 mrem to any organ for any batch. b. Gaseous

b. Gaseous (1) Doses due to gaseous releases to areas at or beyond the site boundary shall be calculated in (1) The Gaseous Radwaste Treatment System accordance with the methodology and shall be operable. The Gaseous Radwaste parameters in Part II.

Treatment System shall be used to reduce radioactive materials in gaseous waste prior to their discharge as necessary to meet the requirements of Control DLCO 3.6.15.

Unit 1 ODCM Revision 24 I 3.1-26 July 2003

RADIOACTIVE EFFLUENT TREATMENT SYSTEMS D 3/4.6.16 CONTROLS With gaseous radwaste from the main condenser air ejector system being discharged without treatment for more than 7 days, prepare and submit to the Commission within 30 days, pursuant to Control D 6.9.3, a Special Report that identifies the inoperable equipment and the reason for its inoperability, actions taken to restore the inoperable equipment to OPERABLE status, and a summary description of those actions taken to prevent a recurrence.

(2) The Ventilation Exhaust Treatment System shall be operable (2) and appropriate portions of this system shall be used to reduce releases of radioactivity when the projected doses in NOTE:

31 days due to gaseous effluent releases, from each unit, to Only required to be met when the Ventilation areas beyond the site boundary would exceed 0.3 mrem to Exhaust Treatment System is not being fully utilized.

any organ of a member of the public.

With radioactive gaseous waste being discharged without Project the doses from the iodine and particulate treatment and in excess of the above limit, complete a DER releases from each unit to areas beyond the Site evaluation of the degraded condition within 30 days that Boundary at least every 31 days.

identifies the inoperable equipment, the reason for the inoperability, and plans and schedule to restore the equipment to operable status.

Unit 1 ODCM Revision 24 I 3.1-27 July 2003

MARK I CONTAINMENT D 3/4.6.18 CONTROLS .5 SURVEILLANCE REQUIREMENT DLCO 3.6.18 MARK I CONTAINMENT DSR 4.6.18 MARK I CONTAINMENT A22licability: ApDlicabilitv:

Applies to the venting/purging of the Mark I Applies to the surveillance requirement for venting Containment. and purging of the Mark I Containment when required to be vented/purged through the Emergency Ventilation System.

Oblective: Obiective:

To assure that the Mark I Containment is vented/purged so To verify that the Mark I Containment is vented through the that the limits of Controls DLCO 3.6.15.b(1) and Emergency Ventilation System when required.

DLCO 3.6.15.b(3) are met.

Specification: Specification:

The Mark I Containment drywell shall be vented/ purged The containment drywell shall be determined to be through the Emergency Ventilation System unless Controls aligned for venting/purging through the Emergency DLCO 3.6.15.b.(1) and DICO 3.6.15.b.(3) can be met Ventilation System within four hours prior to start of without use of the Emergency Ventilation System. and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during venting/purging of the drywell.

If these requirements are not satisfied, suspend all venting/purging of the drywell.

Unit 1 ODCM Revision 24 I3.1-28 July 2003

LIQUID WASTE HOLDUP TANKS D 3/4.6.19 CONTROLS SURVEILLANCE REQUIREMENT DLCO 3.6.19 LIQUID WASTE HOLDUP TANKS* DSR 4.6.19 LIQUID WASTE HOLDUP TANKS Applicability: ADplicabilitv:

Applies to the quantity of radioactive material Applies to the surveillance requirements for outdoor that may be stored in an outdoor liquid waste liquid waste holdup tanks.

holdup tank.

Obeective: Objective:

To assure that the quantity of radioactive material stored in To verify the quantity of radioactive material stored in an outdoor holdup tanks does not exceed a specified level. outdoor liquid waste holdup tank.

Specification: Sfecification:

The quantity of radioactive material contained in an outdoor The quantity of radioactive material contained in each of the liquid waste tank shall be limited to less than or equal to 10 tanks listed in Control DLCO 3.6.19 shall be determined to cures, excluding tritium and dissolved or entrained noble be within the limit of Control DLCO 3.6.19 by analyzing a gases. representative sample of the tank's contents at least weekly With the quantity of radioactive material in any such tank when radioactive materials are being added to the tank.

exceeding the above limit, immediately suspend all additions of radioactive material to the tank. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit and describe the events leading to this condition in the next Radioactive Effluent Release Report.

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

Unit 1 ODCM Revision 24 I 3.1-29 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 CONTROLS SURVEILLANCE REQUIREMENT

.9 DLCO 3.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING DSR 4.6.20 RADIOLOGICAL ENVIRONMENTAL PROGRAM MONITORING PROGRAM Apolicabilitv. Applicabilitv Applies to radiological samples of station Applies to the periodic sampling and monitoring environs. requirements of the radiological environmental monitoring program.

Oblective: Objective:

To evaluate the effects of station operations and radioactive To ascertain what effect station operations and radioactive effluent releases on the environs and to verify the effluent releases have had upon the environment.

effectiveness of the controls on radioactive material sources.

Specification: Specification:

The radiological environmental monitoring program shall be The radiological environmental monitoring samples shall be conducted as specified in Table D 3.6.20-1. collected pursuant to Table D 3.6.20-1 from the specific locations given in the table and figure(s) In Part II and shall With the radiological environmental monitoring program not be analyzed pursuant to the requirements of Table D 3.6.20-being conducted as specified InTable D 3.6.20-1, prepare 1 and the detection capabilities required by Table D 4.6.20-and submit to the Commission, in the Annual Radiological 1.

Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

Deviations are permitted from the required sample schedule if samples are unobtainable due to hazardous conditions, seasonal unavailability, theft, uncooperative residents or to malfunction of automatic sampling equipment. In the event of the latter, every effort shall be made to complete corrective action prior to the end of the next sampling period.

Unit 1 ODCM Revision 24 I 3.1-30 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 CONTROLS SURVEILLANCE REQUIREMENT With the level of radioactivity (as the result of plant effluents), in an environmental sampling medium exceeding the reporting levels of Table D 6.9.3-1 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Special Report pursuant to Control D 6.9.3. The Special Report shall identify the cause(s) for exceeding the limit(s) and define the corrective action(s) to be taken to reduce radioactive effluents so that the potential annual dose to a member of the public is less than the calendar year limits of Controls DLCO 3.6.15.a.(2),

DLCO 3.6.15.b.(2) and DLCO 3.6.15.b.(3). When more than one of the radionuclides in Table D 6.9.3-1 are detected Inthe sampling medium, this report shall be submitted if:

concentration (1) + concentration (2) + ........

limit level (1) limit level (2)

1 .0 When radionuclides other than those in Table D 6.9.3-1 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Controls DLCO 3.6.15.a.(2), DLCO 3.6.15.b.(2) and DLCO 3.6.15.b.(3).

Unit 1 ODCM Revision 24 I 3.1-31 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 CONTROLS SURVEILLANCE REQUIREMENT This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

With milk or fruit and/or vegetables no longer available at one or more of the sample locations specified in Table D 3.6.20-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Identify the cause of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for Part II reflecting the new location(s).

Unit 1 ODCM Revision 24 I 3.1-32 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.20-1 OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Number of Samples(a) and Locations Sampling and Collection Type of Analysis and and/or Sample Frequency (a) Frequency Radioiodine & Samples from 5 locations: Continuous sampler operation Radioiodine Canisters analyze Particulates with sample collection weekly or oncetweek for 1-131.

1) 3 Samples from off-site locations in different as required by dust loading, sectors of the highest calculated site average D/Q whichever is more frequent (based on all site licensed reactors)

Particulate Samplers

2) 1 sample from the vicinity of an established year Gross beta radioactivity round community having the highest calculated following filter change, (b) site average D/Q (based on all site licensed composite (by location) for reactors) gamma isotopic analysis(C) once per 3 months, (as a minimum)

3) 1 sample from a control location 10-17 miles distant and In a least prevalent wind direction(d)

Direct Radiation(e) 32 stations with two or more dosimeters to be placed as Once per 3 months Gamma dose once per 3 follows: an inner ring of stations in the general area of months the site boundary and an outer ring in the 4 to 5 mile range from the site with a station in each land based sector.* The balance of the stations should be placed in special interest areas such as population centers, nearby residences, schools and in 2 or 3 areas to serve as control stations.

At this distance, 8 wind rose sectors are over Lake Ontario.

Unit 1 ODCM Revision 24 I3.1-33 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.20-1 (Cont)

OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Collection Type of Analysis and Number of Samples(") and Locations Frecuencv (a) and/or Samole Frequencv WATERBORNE Surface° 1) 1 sample upstream Composite sample over 1 month Gamma isotopic analysis(c) period(g) once/month. Composite for once per 3 months tritium analysis.

2) 1 sample from the site's downstream cooling water intake Sediment from Shoreline 1 sample from a downstream area with existing or Twice per year Gamma isotopic analysis(C) potential recreational value INGESTION Milk 1) Samples from milk sampling locations in 3 Twice per month, April-December Gamma isotopic(c) and 1-131 locations within 3.5 miles distance having the (samples will be collected in analysis twice per month when highest calculated site average D/Q. If there are January-March if 1-131 is detected animals are on pasture (April-none, then 1 sample from milking animals in in November and December of December); once/month at each of 3 areas 3.5-5.0 miles distant having the the preceding year) other times (January-March) if highest calculated site average D/Q (based on required all site licensed reactors)

2) 1 sample from a milk sampling location at a control location (9-20 miles distant and In a least prevalent wind direction)(d)

Unit 1 ODCM Revision 24 I 3.1-34 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D 3.6.20-1 (Cont)

OPERATIONAL RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Exposure Pathway Sampling and Collection Type of Analysis and Number of Samples(s) and Locations and/or Sample Freouency (a) Freauenev Fish 1) 1 sample each of two commercially or Twice per year Gamma isotopic analysis(C) on recreationally important species in the vicinity of a edible portions twice per year plant discharge area.Y'

2) 1 sample each of the same species from an area at least 5 miles distant from the site.(d Food Products 1) Samples of three different kinds of broad leaf Once per year during harvest Gamma isotopic(c) analysis of vegetation (such as vegetables) grown nearest to season edible portions (isotopic to each of two different off-site locations of highest include 1-131 or a separate I-calculated site average D/Q (based on all 131 analysis may be licensed site reactors). performed) once during the harvest season

2) Once sample of each of the similar broad leaf vegetation grown at least 9.3-20 miles distant In a least prevalent wind direction.

Unit 1 ODCM Revision 24 13.1-35 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 314.6.20 NOTES FOR TABLE D 3.6.20-1 (a) It Is recognized that, at times, it may not be possible or practical to obtain samples of the media of choice at the most desired location or time. In these Instances, suitable alternative media and locations may be chosen for the particular pathway in question and may be substituted. Actual locations (distance and directions) from the site shall be provided in the Annual Radiological Environmental Operating Report. Highest D/Q locations are based on historical meteorological data for all site licensed reactors.

(b) Particulate sample filters should be analyzed for gross beta 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. If the gross beta activity in air is greater than 10 times a historical yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

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

(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, such as historical control locations which provide valid background data may be substituted.

(e) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purpose of this table, a thermoluminescent dosimeter may be considered to be one phosphor and two or more phosphors in a packet may be considered as two or more dosimeters. Film badges shall not be used for measuring direct radiation.

(f) The 'upstream sample' should be taken at a distance beyond significant influence of the discharge. The 'downstream samplen should be taken in an area beyond but near the mixing zone, if possible.

(g) Composite samples should be collected with equipment (or equivalent) which is capable of collecting an aliquot at time intervals which are very short (e.g. hourly) relative to the compositing period (e.g. monthly) in order to assure obtaining a representative sample.

(h) In the event commercial or recreational important species are not available as a result of three attempts, then other species may be utilized as available.

Unit 1 ODCM Revision 24 I 3.1-36 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 TABLE D4.6.20-1 DETECTION CAPABILIES FOR ENVIRONMENTAL SAMPLE ANALYSIS"0)

LOWER LIMIT OF DETECTION LLD(c)

Surveillance Requirement Water(c) Airborne Particulate or Fish Milk Food Products Sediment Analysis (pClf) Gases (pCI/m 3) (pCi/kg, wet) (pCIf) (pCI/kg, wet) (pCI/kg, dry) gross beta 4 0.01 H-3 2000*

Mn-54 15 130 Fe-59 30 260 Co-58, Co-60 15 130 Zn-65 30 260 Zr-95, Nb-95 15 1-131 1* 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba/La-140 15 15

If no drinking water pathway exists, a value of 3000 pCIliter may be used.

- If no drinking water pathway exists, a value of 15 pCI/liter may be used.

Unit 1 ODCM Revision 24 I 3.1-37 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 NOTES FOR TABLE D 4.6.20-1 (a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported In the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.6.2, "Annual Radiological Environmental Operating Reporr, and Control D 6.9.1.d.

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

(c) The LLD is defined, for purposes of these specifications, 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 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a *reald signal.

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

LLD = 4.66 Sh E*Ve2.22eYeexp (-MAt)

Where:

LLD Is the 'a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V Is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, where applicable, Xis the radioactive decay constant for the particular radionuclide, and At for environmental samples is the elapsed time between sample collection, or end of the sample collection period and time of counting.

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

Unit 1 ODCM Revision 24 13.1-38 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM D 3/4.6.20 NOTES FOR TABLE D 4.6.20-1 It should be recognized that the ILD is defined as a before the fact limit representing the capability of a measurement system and not as an after the fact limit for the particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.

Occasionally, background fluctuations, unavoidable small sample sizes, the presence of interfering nudides or other uncontrollable circumstances may render these IIDs unachievable. In such cases, the contributing factors shall be Identified and described in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.6.2, "Annual Radiological Environmental Operating Report", and Control D 6.9.1 .d.

I Unit 1 ODCM Revision 24 13.1-39 July 2003

INTERLABORATORY COMPARISON PROGRAM D 3/4.6.21 CONTROLS SURVEILLANCE REQUIREMENT

- q DLCO 3.6.21 INTERLABORATORY COMPARISON PROGRAM DSR 4.6.21 INTERLABORATORY COMPARISON PROGRAM Applicability: ADmlicabilitv:

Applies to participation In an interlaboratory comparison Applies to testing the validity of measurements on program on environmental sample analysis. environmental samples.

Oblective:

Obiective:

To ensure the accuracy of measurements of radioactive To verify the accuracy of measurements on radioactive material in environmental samples.

material in environmental samples.

Specification: Specification:

Analyses shall be performed on radioactive materials The Interlaboratory Comparison Program shall be described supplied as part of an Interlaboratory Comparison Program in Part II. A summary of the results obtained as part of the which has been approved by the Commission. Participation above required Interlaboratory Comparison Program shall be in this program shall include media for which environmental included in the Annual Radiological Environmental Operating samples are routinely collected and for which Report. Participants in the EPA Cross Check Program may Intercomparison samples are available. provide the EPA program code designation in lieu of providing results.

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

Unit 1 ODCM Revision 24 13.1-40 July 2003

LAND USE CENSUS D 3/4.6.22 CONTROLS SURVEILLANCE REQUIREMENT h

DLCO 3.6.22 LAND USE CENSUS DSR 4.6.22 LAND USE CENSUS Applicability: Agplicability.

Applies to the performance of a land use census in the Applies to assuring that current land use Is known.

vicinity of the Nine Mile Point Nuclear Facility.

Oblective:

Obiective:

To determine the utilization of land within a distance of three To verify the appropriateness of the environmental miles from the Facility.

surveillance program.

Specification: Specification:

A land use census shall be conducted and shall identify The land use census shall be conducted during the growing within a distance of three miles the location in each of the 16 season at least once per 12 months using that information meteorological sectors the nearest residence and within a that will provide the best results, such as conducting a door-distance of three miles the location In each of the 16 to-door survey, aerial survey or consulting local agriculture meteorological sectors of all milk animals. In lieu of a authorities. The results of the land use census shall be garden census, specifications for vegetation sampling in included in the Annual Radiological Environmental Operating Table D 3.6.20-1 shall be followed, including analysis of Report.

appropriate controls.

With a land use census Identifying a milk animal location(s) that represents a calculated D/Q value greater than the D/Q value currently being used in Control DSR 4.6.15.b.(3), identify the new location(s) in the next Radioactive Effluent Release ReporL Unit 1 ODCM Revision 24 I 3.1-41 July 2003

LAND USE CENSUS D 3/4.6.22 CONTROLS SURVEILLANCE REQUIREMENT 9

If the D/Q value at a new milk sampling location is significantly greater (50%) than the D/Q value at an existing milk sampling location, add the new location to the radiological environmental monitoring program within 30 days. The sampling location(s) excluding the control station location, having the lowest calculated D/Q may be deleted from this monitoring program after October 31 of the year In which this land use census was conducted. Pursuant to Control D 6.9.1 .e identify the new location(s) in the next Radioactive Effluent Release Report and also Include in the report a revised figure(s) and table for Part II reflecting the new location(s).

Unit 1 ODCM Revision 24 13.1-42 July 2003

PART I - RADIOLOGICAL EFFLUENT CONTROLS Bases Unit 1 ODCM Revision 24 I B 3.1-0 July 2003

RADIOACTIVE EFFLUENT INSTRUMENTATION, RADIOACTIVE EFFLUENTS - LIQUID CONCENTRATION B 314.6.14, B 3/4.6.15 BASES FOR DLCO 3.6.14 and DSR 4.6.14 RADIOACTIVE EFFLUENT INSTRUMENTATION The radioactive liquid and gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid and gaseous effluents during actual or potential releases of liquid and 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 prior to exceeding the limits as described in Technical Specification 6.5.3, Radioactive Effluent Controls Program." Historically, the maximum allowable deviation of

+50% of setpoint for High Radiation-Offgas Line (Reference FSAR, Appendix D) and +100% and -50% of setpoint for High Radiation-Emergency Cooling System Vent had negligible effect on the initiations of these systems. The operability and use of this Instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10CFR Part 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.

BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Liquid Concentration This control 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 I OCFR Part 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 Il.A design objectives of Appendix I, 10CFR Part 50, to a member of the public and (2) the limits of 10 CFR 20.1301 (e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-1 35 is the controlling radioisotope and its effluent concentration in air (submersion) was converted to an equivalent concentration in water using the methods described In Intemational Commission on Radiological Protection (ICRP) Publication 2.

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 HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A.,

Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry,o Anal. Chem. 40. 586-93 (1968), and Hartwell, J.

K., "Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

Unit 1 ODCM Revision 24 I B 3.1-1 July 2003

RADIOACTIVE EFFLUENTS - LIQUID DOSE B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Liquid Dose This control is provided to implement the requirements of Section I.A, III.A and IV.A of Appendix I, 10CFR Part 50. The controls expressed as quarter and annual limits are set at those values found in Section II.A. of Appendix I, in accordance with Section IV.A. 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 material in liquid effluents to unrestricted areas will be kept was low as is reasonably achievable. There are no drinking water supplies that can be potentially affected by plant operations. The dose calculation methodology and parameters in Part II implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculation procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The equations specified in Part II for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, WEstimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix l," April 1977.

Unit 1 ODCM Revision 24 I B 3.1-2 July 2003

RADIOACTIVE EFFLUENTS - GASEOUS DOSE RATE B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Gaseous Dose Rate This control is provided to ensure that the dose at any time at or beyond the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10CFR Part 20 to unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10CFR Part 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 10CFR Part 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. 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 mrems/year to the total body or to less than or equal to 3000 mrems/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 mrems/year.

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 HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A.,

'Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry,m Anal. Chem. 40, 586-93 (1968), and Hartwell, J.

K., wDetection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

Unit 1 ODCM Revision 24 I B 3.1-3 July 2003

RADIOACTIVE EFFLUENTS - DOSE - NOBLE GASES B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Dose - Noble Gases This control is provided to implement the requirements of Sections Il.B, IlI.A and IV.A of Appendix I, 10CFR Part 50. The controls expressed as quarter and annual limits are set at those values found in Section ll.B of Appendix I in accordance with the guidance of Section IV.A. 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 material in gaseous effluents to unrestricted areas will be kept 'as low as is reasonably achievable." The Surveillance Requirement implements the requirements in Section III.A of Appendix I that conform with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in Part II for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, 'Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I, "Revision 1, October 1977 and Regulatory Guide 1.1 11, 'Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Ught-Water Cooled Reactors,' Revision 1, July 1977.

The Offsite Dose Calculation Manual Part II equations provided to determine the air doses beyond the site boundary are based upon the historical average atmospheric conditions.

Unit 1 ODCM Revision 24 I B 3.1-4 July 2003

RADIOACTIVE EFFLUENTS - DOSE - IODINE -131, IODINE -133, TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Dose - lodine-1 31. Iodine-I 33. Tritium and Radionuclides in Particulate Form This control is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10CFR Part 50. The controls expressed as quarter and annual limits are set at those values found in Section I.C of Appendix I in accordance with the guidance of Section IV.A. The action statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to unrestricted areas will be kept 'as low as is reasonably achievable." The Part II calculational methods specified in the Surveillance Requirement implements the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a member of the public through appropriate pathways is unlikely to be substantially underestimated. The Part II calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR Part 50, Appendix I,' Revision 1, October 1977 and Regulatory Guide 1.1 11, 'Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for iodine-1 31, iodine-1 33, tritium and radionuclides in particulate form with half lives greater than 8 days are dependent upon the existing radionuclide pathways to man, in the areas beyond the site boundary. The pathways that were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man and 4) deposition on the ground with subsequent exposure of man.

Unit 1 ODCM Revision 24 I B 3.1-5 July 2003

RADIOACTIVE EFFLUENTS - TOTAL DOSE -

URANIUM FUEL CYCLE B 3/4.6.15 BASES FOR DLCO 3.6.15 AND DSR 4.6.15 RADIOACTIVE EFFLUENTS Total Dose - Uranium Fuel Cycle This control is provided to meet the dose limitations of 40CFR Part 190 that have been incorporated into 10CFR Part 20 by 46FR 18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a member of the public will exceed the dose limits of 40CFR Part 190 if the ind-ivdual reactors remain within twice the dose design objectives of Appendix I and if direct radiation doses from the reactor units and outside storage tanks 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 40CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to a member of the public from other uranium fuel cycle sources is negligible, with the exception that dose contribution from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any member of the public is estimated to exceed the requirements of 40CFR Part 190, the Special Report with a request for variance (provided the release conditions resulting in violation of 40CFR Part 190 have not already been corrected), in accordance with the provisions of 40CFR Part 190.11 and 10 CFR Part 20.2203(a)(4) is considered to be a timely request and fulfills the requirements of 40CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10CFR Part 20, as addressed in Controls DLCO 3.6.15.a.(1) and DLCO 3.6.15.b.(1). An individual is not considered a member of the public during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

Unit 1 ODCM Revision 24 I B 3.1-6 July 2003

RADIOACTIVE EFFLUENT TREATMENT SYSTEMS -

LIQUID AND GASEOUS B 3/4.6.16 BASES FOR DLCO 3.6.16 AND DSR 4.6.16 RADIOACTIVE EFFLUENT TREATMENT SYSTEMS Liquid Radwaste Treatment System The requirement that the appropriate portions of this system be used provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 and the design objective given in Section lI.D of Appendix Ito 10 CFR Part 50. Projected doses are calculated on a batch rather than every 31 days due to the low frequency of releases.

Gaseous Effluent Treatment Systems The operability of the Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment System ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that appropriate portions of these systems 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 control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50 and the design objectives given in Section ll.D of Appendix I to 10 CFR Part 50. The control governing the use of appropriate portions of the Gaseous Radwaste Treatment System is based on time without treatment rather than dose, due to the wide variability in effluent with changing power conditions. Since the capability exists to operate within specification without use of the Gaseous Radwaste Treatment System, it is conceivable that due to unforeseen circumstances, limited operation without the system may be made sometime during the life of the plant. The control governing the use of appropriate portions of the Ventilation Exhaust Treatment System was specified as a suitable fraction of the dose design objectives set forth in II.C of Appendix I, 10CFR Part 50, for gaseous effluents.

Unit 1 ODCM Revision 24 I B 3.1-7 July 2003

MARK I CONTAINMENT, LIQUID HOLDUP TANKS B 3/4.6.18, B 3/4.6.19 BASES FOR DLCO 3.6.18 AND DSR 4.6.18 MARK I CONTAINMENT This control provides reasonable assurance that releases from drywell purging operations will not exceed the annual dose limits of 10CFR Part 20 for unrestricted areas.

BASES FOR DLCO 3.6.19 AND DSR 4.6.19 LIQUID HOLDUP TANKS This control applies to any outdoor tank that is not surrounded by liners, dikes or walls capable of holding the tank contents and that does not have tank overflows and surrounding areas drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than ten times the concentrations of 10CFR Part 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

Unit 1 ODCM Revision 24 I B 3.1-8 July 2003

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM B 3/4.6.20 BASES FOR DLCO 3.6.20 AND DSR 4.6.20 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring program required by this control provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of members of the public resulting from the station operation. This monitoring program ImplementsSection IV.B.2 of Appendix I to 10CFR Part 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. The initially specified monitoring program will be effective for at least the first three years of commercial operation. Following this period, 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 4.6.20-1 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 HASL Procedures Manual, HASL-300 (revised annually), Currie, L.A.,

'Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry,o Anal. Chem 40, 586-93 (1968) and Hartwell, J.K.,

'Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

Unit 1 ODCM Revision 24 I B 3.1-9 July 2003

INTERLABORATORY COMPARISON PROGRAM B 3/4.6.21 BASES FOR DLCO 3.6.21 AND DSR 4.6.21 INTERLABORATORY COMPARISON PROGRAM 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 material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring for the purposes of Section IV.B.2 of Appendix I to 10CFR Part 50.

Unit 1 ODCM Revision 24 I B 3.1-10 July 2003

LAND USE CENSUS B 3/4.6.22 BASES FOR DLCO 3.6.22 AND DSR 4.6.22 LAND USE CENSUS This control is provided to ensure that changes in the use of areas at and 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 survey information such as from a door-to-door survey(s), 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 10CFR Part 50.

In lieu of a garden census, the significance of the exposure via the garden pathway can be evaluated by the sampling of vegetation as specified in Table D 3.6.20-1.

A milk sampling location, as defined in Section 1, 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 per month for analytical purposes. Locations with less than 10 milking cows are usually utilized for breeding purposes which eliminates a stable supply of milk for samples as a result of suckling calves and periods when the adult animals are dry.

Unit 1 ODCM Revision 24 I B 3.1-11 July 2003

PART I - RADIOLOGICAL EFFLUENT CONTROLS Section 6.0 Administrative Controls Unit 1 ODCM Revision 24 I6.0-0 July 2003

Administrative Controls 6.0 6.0 ADMINISTRATIVE CONTROLS The ODCM Specifications are subject to Technical Specification Section 6.6.2, "Annual Radiological Environmental Operating Report," Section 6.6.3, "Radioactive Effluent Release Report," Section 6.5.1, "Offsite Dose Calculation Manual (ODCM)," and Section 6.5.3, "Radioactive Effluent Controls Program."

Unit 1 ODCM Revision 24 I6.0-1 July 2003

REPORTING REQUIREMENTS D 6.9.1.d D.6.9.1 .e D 6.9 Reporting Reauirements D 6.9.1.d Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report shall include a comparison with operational controls as appropriate, and with environmental surveillance reports from the previous 5 years, and an assessment of the observed impacts of the plant operation on the environment. The report shall also include the results of land use censuses required by Control DLCO 3.6.22.

The report shall also include the following: a summary description of the radiological environmental monitoring program; at least two legible maps** covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Control DLCO 3.6.21; discussion of all deviations from the sampling schedule of Table D 3.6.20-1; and discussion of all analyses in which the LLD required in Table D 4.6.20-1 was not achievable.

- One map shall cover stations near the site boundary; a second shall include the more distant stations.

D 6.9.1.e Radioactive Effluent Release Report The Radioactive Effluent Release Report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste releases from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants", Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report 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 distributions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary (Part II Figure 5.1.3-1) during the reporting 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.

The Radioactive Effluent Release Report shall also include an assessment of radiation 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 Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the doses from liquid and gaseous effluents are given in Part II.

In lieu of submission with the Radioactive 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.

Unit 1 ODCM Revision 24 I 6.0-2 July 2003

REPORTING REQUIREMENTS D 6.9.1.e The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period.

a. Container volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and,

f. Solidification agent or absorbent (e.g., cement)

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the Process Control Program (PCP) and to the Offsite Dose Calculation Manual (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Control DLCO 3.6.22.

Changes to the Process Control Program (PCP) shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:

a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;

b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

c. Documentation of the fact that the change has been reviewed and found acceptable.

Changes to the Offsite Dose Calculation Manual (ODCM) shall be in accordance with Technical Specification 6.5.1, "Offsite Dose Calculation Manual (ODCM)".

Unit 1 ODCM Revision 24 I 6.0-3 July 2003

SPECIAL REPORTS D 6.9.3 D 6.9.3Special Reports Special reports shall be submitted in accordance with 10 CFR 50.4 to the Regional Office within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a.

b.

C.

d. Not applicable to RETS e.

f.

g. 4-

h. Calculate Dosiefrom Liquid Effluent In Excess of Limits, Control DLCO 3.6.1 5.a(2) (30 days from the end of the affected calendar quarter).

i. Calculate Air C)ose from Noble Gases Effluent in Excess of Limits, Control DLCO 3.6.15.b(2) (30 days from the end of the affected calendar quarter).

j. Calculate Dosiefrom 1-131, H-3 and Radioactive Particulates with halt lives greater than eight days in Excess of Limits, Control DLCO 3.6.15.b(3)(b) (30 days from the end of the affected calendar quarter).

k. Caculated Doc-es from Uranium Fuel Cycle Source in Excess of Limits, Control DLCO 3.6.1 5.d (30 days from the end of the affected calendar year)

I. Inoperable Gaseous Radwaste Treatment System, Control DLCO 3.6.16.b (30 days from the end of the affected calendar year).

m. Environmental Radiolocical Reports. With the level of radioactivity (as the result of plant effluents) in an environmental sampling media exceeding the reporting level of Table D 6.9.3-1, when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the (Commission within thirty (30) days from the end of the calendar quarter a special report identifying the cause(s) for exceeding the limits, and define the corrective action to be taken.

Unit 1 ODCM Revision 24 I 6.0-4 July 2003

SPECIAL REPORTS D 6.9.3 Table D 6.9.3-1 REPORTING LEVEL FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS Water Airborne Particulate or Fish Milk Food Products Analysis (pClf) Gases (pCIm 3) (pCikg, wet) (pCfl) (pClIkg, wet)

H-3 20,000*

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, Nb-95 400 1-131 2** 0.9 3 100 Cs-134 30 10.0 1,000 60 1,000 Cs-137 50 20.0 2,000 70 2,000 Ba/La-140 200 300

For drinking water samples. This Is a 40 CFR 141 value. If no drinking water pathway exists, a value of 30,000 pCi/liter may be used.

If no drinking water pathway exists, a value of 20 pCi/liter may be used.

Unit 1 ODCM Revision 24 I 6.0-5 July 2003

PART II- CALCULATIONAL METHODOLOGIES Unit 1 ODCM Revision 24 1Il July 2003

1.0 LIQUID EFFLUENTS 1.1 Setpoint Determinations 1.1.1 Basis Monitor setpoints will be established such that the concentration of radionuclides in the liquid effluent releases in the discharge canal shall be limited to ten times the concentrations specified in 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2E-04 gCi/ml total activity. Setpoints for the Service Water System Effluent Line will be calculated quarterly based on the radionuclides identified during the previous year's releases from the liquid radwaste system or the isotopes identified in the most recent radwaste release or other identified probable source. Setpoints for the Liquid Radwaste Effluent Line will be based on the radionuclides identified in each batch of liquid waste prior to its release.

After release, the Liquid Radwaste monitor setpoint may remain as set, or revert back to a setpoint based on a previous Radioactive Effluent Release Report, or install blank flange in the discharge line and declare inoperable in accordance with the ODCM Part L Since the Service Water System effluent monitor and Liquid Radwaste effluent monitor can only detect gamma radiation, the alarm setpoints are calculated by using the concentration of gamma emitting isotopes only (or the corresponding Maximum Effluent Concentration (MEG) values for the same isotopes, whichever are higher) in the X i(gCi/ml)iy expression (Section 1.1.2, 1.1.3).

The Required Dilution Factor (RDF) is calculated using concentrations of all isotopes present (or the corresponding MEC values for the same isotopes, whichever are higher) including tritium and other non-gamma emitters to ensure that all radionuclides in the discharge canal do not exceed Technical Specifications Radioactive Effluent Controls Program limits.

1.1.2 Service Water System Effluent Line Alarm Setpoint The detailed methods for establishing setpoints for the Service Water System Effluent Line Monitor shall be contained in the Nine Mile Point Station Procedures. These methods shall be in accordance with the following:

The General Setpoint Equation is Setpoint < (Conservative Factor) (Concentration)(ADF)(CF)

RDF From the above General Setpoint Equation the Hi and Alert alarms are calculated as follows:

Setpoint (Hi alarm)< 0.9 by (jCi / ml) 1 7(CF)TEDF / FSw +background LI [(joci / ml) ff / MEC If]

Setpoint (Alert alarm) <0.7 2 g(pCi / ml)(CF)TEDF /FW + background of foI

[, / ml) IT / MEC I I (pCi/ml)iy= concentration of gamma emitting isotope i in the sample, or the corresponding MEC of gamma emitting isotope i (MEC)I, whichever is higher (units = FtCi/ml).

Unit 1 ODCM Revision 24 1 2 July 2003

1.1.2 Service Water System Effluent Line Alarm Setpoint (Cont'd)

(JLCi/ml)iT = concentration of any radioactive isotope i in the sample including tritium and other non-gamma emitters or corresponding MEC of isotope i, MECQ, whichever is higher (units = pCi/ml).

TF = Tempering Fraction TDF = Total Dilution Flow (units = gallons/minute).

TEDF = Total Effective Dilution Flow = TDF (1-TF) (units = gallons/minute)

Ft. = Service Water Flow (units = gallons/minute).

CF = Monitor calibration factor (units = net cpm/gCi/ml).

MECQ = Maximum Effluent Concentration, ten times the Effluent Concentration for radionuclide i as specified in 10 CFR 20, Appendix B, Table 2, Column 2 (units RCi/ml).

Sample = Those nuclides present in the previous batch release from the liquid radwaste effluent system or those nuclides present in the last Radioactive Effluent Release Report (units

= JACi/ml) or those nuclides present in the service water system.**

(MEC)~, = same as MEQ but for gamma emitting nuclides only.

0.9 and 0.7= factors of conservatism to account for inaccuracies.

RDF = Required Dilution Factor, i [(pCi/ml) IT/MEC4I. If MEC values are used in the (gCiml)i,, they must also be used in calculating RDF (numerator). RDF= FMEC (See Section It-1.2).

ADF = Actual Dilution Factor, TEDF/FSw

- For periods with known reactor water to Reactor Building Closed Loop Cooling (RBCLC) system leakage, RBCLC concentration may be prudently substituted for the above.

1.1.3 Liquid Radwaste Effluent Line Alarm Setpoint The detailed methods for establishing setpoints for the Liquid Radwaste Effluent Line Monitor shall be contained in the Nine Mile Point Station Procedures. These methods shall be in accordance with the following:

The General Setpoint Equation in Section 11-1. 1.2 is used to develop the Hi-Hi and Hi alarm setpoints below:

Setpoint (Hi-Hi alarm) < 0.9 (pCi Ad / ml)1 7(CF)TEDF / FM + background Unit 1 ODCM Revision 24 113 July 2003

1.1.3 Liquid Radwaste Effluent Line Alarm Setpoint (Cont'd)

Ei (lpCi / ml) 17 (CF)TEDF / F, Setpoint (Hi alarm) < 0.7 zj(u mYC T r+ background Eif (4 uCi / ml)ff / MEC jI (PCi/ml)iy = concentration of gamma emitting isotope i in the sample or the corresponding MEC of gamma emitting isotope i, (MEC)i whichever is higher.

(PCi/ml),T = concentration of any radioactive isotope i in the sample including tritium and other non-gamma emitters or the corresponding MEC of isotope i, MEQ, whichever is higher. (units = pCi/ml).

TF = Tempering Fraction TDF = Total Dilution Flow (units = gallons/minute).

TEDF = Total Effective Dilution Flow = TDF (1-TF) (units = gallons/minute)

F. = Radwaste Effluent Flow (units = gallons/minute).

CF = Monitor calibration factor (units = net cps/pCi/ril).

MEQ. = Maximum Effluent Concentration, ten times the Effluent Concentration for radionuclide i as specified in 10 CFR 20, Appendix B, Table 2, Column 2, for those nuclides detected by spectral analysis of the contents of the radwaste tanks to be released. (units = PCi /ml)

(MEC)y = same as MEQ but for gamma emitting nuclide only.

0.9 and 0.7 = factors of conservatism to account for inaccuracies.

RDF = Required Dilution Factor, z i [(pCi/ml) IT/MEC].

If MEC values are used in the (gCi/ml)iy, they must also be used in calculating RDF (numerator).

ADF = Actual Dilution Factor = TEDF/F,.

Notes: (a) If TEDF/F. = E v[(Ci/ml) rf/MEQ] (if ADF = RDF) the discharge could not be made, since the monitor would be continuously in alarn. To avoid this situation, F.t will be reduced (normally by a factor of 2) to allow setting the alarm point at a concentration higher than tank concentration.

This will also result in a discharge canal concentration at approximately 50%

Maximum Effluent Concentration.

(b) TF is tempering fraction (i.e., diversion of some fraction of discharge flow to the intake canal for the purpose of temperature control).

Unit 1 ODCM Revision 24 14 July 2003

1.1.4 Discussion 1.1A.1 Control of Liquid Effluent Batch Discharges At Nine Mile Point Unit 1 Liquid Radwaste Effluents are released only on a batch mode.

To prevent the inadvertent release of any liquid radwaste effluents, radwaste discharge is mechanically isolated (blank flange installed or discharge valve chain-locked closed) following the completion of a batch release or series of batch releases.

This mechanical isolation remains in place and will only be removed prior to the next series of liquid radwaste discharges after all analyses required in station procedures and Table D 4.6.15-lA of Part I are performed and monitor setpoints have been properly adjusted.

1.1.4.2 Simultaneous Discharges of Radioactive Liquids If during the discharge of any liquid radwaste batch, there is an indication that the service water canal has become contaminated (through a service water monitor alarm or through a grab sample analysis in the event that the service water monitor is inoperable) the discharge shall be terminated immediately. The liquid radwaste discharge shall not be continued until the cause of the service water alarm (or high grab sample analysis result) has been determined and the appropriate corrective measures taken to ensure ten times the effluent concentrations specified in 10CFR20, Appendix B, Table 2, Column 2 (Section D 3.6.15.a(1) of Part I) are not exceeded. In accordance with Liquid Waste procedures, controls are in place to preclude a simultaneous release of liquid radwaste batch tanks. In addition, an independent verification of the discharge valve line-up is performed prior to discharge to ensure that simultaneous discharges are prevented.

1.1.4.3 Sampling Representativeness This section covers Part I Table D 4.6.15-1 Note b concerning thoroughly mixing of each batch of liquid radwaste prior to sampling.

Liquid Radwaste Tanks scheduled for discharge at Nine Mile Point Unit 1 are isolated (i.e.

inlet valves marked up) and at least two tank volumes of entrained fluids are recirculated prior to sampling. Minimum recirculation time is calculated as follows:

Minimum Recirculation Time = 2.0(T/R)

Where:

2.0 = Plant established mixing factor, unitless T = Tank volume, gal R = Recirculation flow rate, gpm Additionally, the Hi Alarm setpoint of the Liquid Radwaste Effluent Radiation Monitor is set at a value corresponding to not more than 70% of its calculated response to the grab sample or corresponding MEC values. Thus, this radiation monitor will alarm if the grab sample, or corresponding MEC value, is significantly lower in activity than any part of the tank contents being discharged.

Unit 1 ODCM Revision 24 I 5 July 2003

1.1.4.4 Liquid Radwaste System Operation Part I Section DLCO 3.6.16.a requires that the liquid radwaste system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge, as necessary, to meet the concentration and dose requirements of Section DLCO 3.6.15.

Utilization of the radwaste system will be based on the capability of the indicated components of each process system to process contents of the respective low conductivity and high conductivity collection tanks:

1) Low Conductivity (Equipment Drains): Radwaste Filter and Radwaste Demin.

(See Fig. D-1) or modular waste water technology ("THERMEX")

2) High Conductivity (Floor Drains): Waste Evaporator (See Fig. D-1) or modular waste water technology ("THERMEX") directly to the Waste Collector Tank or the Waste Sample Tanks.

Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined as described in Section II-1.3 of this manual prior to the release of each batch of liquid waste. This same dose projection of Section II-1.3 will also be performed in the event that untreated liquid waste is discharged, to ensure that the dose limits of Part I DLCO 3.6.15.a(2) are not exceeded. (Thereby implementing the requirements of 10CFR50.36a, General Design Criteria 60 of Appendix A and the Design Objective given in Section II-D of Appendix I to 10 CFR50).

For the purpose of dose projection, the following assumptions shall be made with regard to concentrations of non-garmna emitting radionuclides subsequently analyzed off-site:

a) [H-3]

H-3 Concentration found recent condensate storage tank analysis b) [Sr-89]

4 x Cs-137 Concentration c) [Sr-90]

0.5 xaCs-137 Concentration d) [Fe-551 S 1 x Co-60 Concentration Assumed Scaling Factors used in b, c, and d above represent conservative estimates derived from analysis of historical data from process waste streams. Following receipt of off-site H-3, Sr-89, Sr-90 and Fe-55 analysis information, dose estimates shall be revised using actual radionuclide concentrations and actual tank volumes discharged.

Unit 1 ODCM Revision 24 II 6 July 2003

1A.5 Service Water System Contamination Service water is normally non-radioactive. If contamination is suspected, as indicated by a significant increase in service water effluent monitor response, grab samples will be obtained from the service water discharge lines and a gamma isotopic analysis meeting the LLD requirements of Part I Table D 4.6.15-1 completed. If it is determined that an inadvertent radioactive discharge is occurring from the service water system, then:

a) A 10CFR 50.59 review shall be performed (ref. I&E Bulletin 80-10),

b) daily service water effluent samples shall be taken and analyzed for principal gamma emitters until the release is terminated, c) an incident composite shall be prepared for H-3, gross alpha, Sr-89, Sr-90 and Fe-55 analyses and, d) dose projections shall be performed in accordance with Section II-1.3 of this manual (using estimated concentrations for H-3, Sr-89, Sr-90 and Fe-55 to be conservatively determined by supervision at the time of the incident).

Additionally, service water effluent monitor setpoints may be recalculated using the actual distribution of isotopes found from sample analysis.

When contamination is indicated by quantitative non-gamma emitter results, sample and analyze gamma and non-gamma emitters weekly.

1.2 Liquid Effluent Concentration Calculation This calculation documents compliance with Part I Section DLCO 3.6.15.a (1).

The concentration of radioactive material released in liquid effluents to unrestricted areas (see Figure 5.1.3-1) shall be limited to ten times the effluent concentrations specified in IOCFR20, 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 2 E-4 microcurie/milliliter (PCi/ml) total activity at the point of discharge. For dissolved and entrained noble gases, this limit may also be satisfied by using 2E-4 LCi/ml as the MEC for each noble gas.

The concentration of radioactivity from Liquid Radwaste batch releases and, if applicable, Service Water System and emergency condenser start-up vent discharges are included in the calculation. The calculation is performed for a specific period of time. No credit taken for averaging. The limiting concentration is calculated as follows:

FMEC =E.F) / (M A F.) I Unit I ODCM Revision 24 1 7 July 2003

1.2 Liquid Effluent Concentration Calculation (Cont'd)

Where:

FMEC = The fraction of Maximum Effluent Concentration, the ratio at the point of discharge of the actual concentration to ten times the Effluent Concentration of 10 CFR 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases.

For noble gases, the concentration shall be limited to 2 E4 microcurie/ml total activity.

Cih=(pCi/ml)i, = The concentration of nuclide i in particular effluent stream s, ACi/ml.

F. = The flow rate of a particular effluent stream s, gpm.

MECj = Maximum Effluent Concentration, ten times the Effluent Concentration of a specific nuclide i from 10CFR20, Appendix B, Table 2, Column 2 (noble gas limit is 2E4 pCi/ml).

27 (C.F.) = The total activity rate of nuclide i, in all effluent streams s.

(F.) = The total flow rate of all effluent streams s, gpm (including those streams which do not contain radioactivity).

A value of less than one for FMEC is considered acceptable for compliance with Part I Section DLCO 3.6.15.a.(1).

1.3 Dose Determinations 1.3.1 Maximum Dose Equivalent Pathway A dose assessment report was prepared for the Nine Mile Point Unit 1 facility by Charles T.

Main, Inc., of Boston, MA. This report presented the calculated dose equivalent rates to individuals as well as the population within a 50-mile radius of the facility based on the radionuclides released in liquid and gaseous effluents during the time periods of 1 July 1980 through 31 December 1980 and from January 1981 through 31 December 1981. The radwaste liquid releases are based on a canal discharge rate of 590 ft3 lsec which affects near field and far field dilution; therefore, this report is specific to this situation. Utilizing the effluent data contained in the Semi-Annual Radioactive Effluent Release Reports as source terms, dose equivalent rates were determined using the environmental pathway models specified in Regulatory Guides 1.109 and 1.111 as incorporated in the NRC computer codes LADTAP for liquid pathways, and XOQDOQ and GASPAR for gaseous effluent pathways. Dose equivalent rates were calculated for the total body as well as seven organs and/or tissues for the adult, teen, child, and infant age groups. From the standpoint of liquid effluents, the pathways evaluated included fish and drinking water ingestion, and external exposure to water and sediment.

Unit 1 ODCM Revision 24 1 8 July 2003

1.3.1 Maximum Dose Equivalent Pathway (Cont'd)

The majority of the dose for a radwaste liquid batch release was received via the fish pathway. However, to comply with Part I Specifications for dose projections, the drinking water and sediment pathways are included. Therefore, all doses due to liquid effluents are calculated monthly for the fish and drinking water ingestion pathways and the sediment external pathway from all detected nuclides in liquid effluents released to the unrestricted areas to each organ. The dose projection for liquid batch releases will also include discharges from the emergency condenser vent as applicable, for all pathways. Each age group dose factor, Aiat, is given in Tables 2-1 to 2-8. To expedite time, the dose is calculated for a maximum individual instead of each age group. This maximum individual will be a composite of the highest dose factor of each age group for each organ, hence As,.

The following expression from NUREG 0133, Section 4.3 is used to calculate dose:

D, = E i [Ai 1 L(ATLCiLFL)1 Where:

D, = The cumulative dose commitment to the total body or any organ, from the liquid effluents for the total time period (ATL), mrem.

ATL = The length of the L th time period over which CuL and FL are averaged for all liquid releases, hours.

CQL = The average concentration of radionuclide, i, in undiluted liquid effluents during time period ATL from any liquid release, ACi/ml.

Ait = The site related ingestion dose commitment factor to the total body or any organ t for each identified principal gamma or beta emitter for a maximum individual, mrem/hr per ACi/mI.

FL = The near field average dilution factor for CL during any liquid effluent release.

Defined as the ratio of the maximum undiluted liquid waste flow during release to the average flow from the site discharge structure to unrestricted receiving waters, unitless.

Aat values for radwaste liquid batch releases at a discharge rate of 295 ft3 /sec (one circulating water pump in operation) are presented in tables 2-1 to 2-4. Ajg values for an emergency condenser vent release are presented in tables 2-5 to 2-8. The emergency condenser vent releases are assumed to travel to the perimeter drain system and released ft3 from the discharge structure at a rate of .33 Osec. See Appendix A for the dose factor A.-t derivation. To expedite time the dose is calculated to a maximum individual. This maximum individual is a composite of the highest dose factor Aht of each age group a for each organ t and each nuclide i. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in a revision to the ODCM.

All doses calculated in this manner for each batch of liquid effluent will be summed for comparison with quarterly and annual limits, added to the doses accumulated from other releases in the quarter and year of interest. In all cases, the following relationships will hold:

Unit I ODCM Revision 24 119 July 2003

1.3.1 Maximum Dose Equivalent Pathway (Cont'd)

For a calendar quarter:

Dt < 1.5 mrem total body Dt < 5 mrem for any organ For the calendar year:

Dt < 3.0 mrem total body Dt < 10 mrem for any organ Where:

Dt = total dose received to the total body or any organ due to liquid effluent releases.

If these limits are exceeded, a special report will be submitted to the NRC identifying the cause and proposed corrective actions. In addition, if these limits are exceeded by a factor of two, calculations shall be made to determine if the dose limits contained in 40 CFR 190 have been exceeded. Dose limits, as contained in 40 CFR 190 are total body and organ doses of 25 nirem per year and a thyroid dose of 75 mnrem per year.

These calculations will include doses as a result of liquid and gaseous pathways as well as doses from direct radiation. The liquid pathway analysis will only include the fish and sediment pathways since the drinking water pathway is insignificant. This pathway is only included in the station's effluent dose projections to comply with Part I Specifications.

Liquid, gaseous and direct radiation pathway doses will consider the James A. FitzPatrick and Nine Mile Point Unit 2 facilities as well as Nine Mile Point Unit 1 Nuclear Station.

In the event the calculations demonstrate that the 40 CFR 190 dose limits, as defined above, have been exceeded, then a report shall be prepared and submitted to the Commission within 30 days as specified in Part I Section DLCO 3.6.15.d.

Section 3.0 of the ODCM contains more information concerning calculations for an evaluation of whether 40 CFR 190 limits have been exceeded.

Unit 1 ODCM Revision 24 1 10 July 2003

1.3.2 Dose Projections - Determinations of the Need to Operate the Liquid Radwaste Treatment System 1.3.2.1 Requirements DLCO 3.6.16.a requires that the liquid radwaste system be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent would exceed 0.06 mrem to the total body or 0.2 mrem to any organ for the batch.

This Control implements Technical Specification 6.5.3.f that requires the Radioactive Effluent Controls Program to include limitations on the functional capability and use of the liquid effluent treatment system to ensure the appropriate portion of this system is used to reduce releases of radioactivity. This is required when the projected doses would exceed 0.06 mrem to the total body and 0.2 inrem to any organ. Since releases are performed much less frequently than once per month, doses are to be projected prior to each release and the above limits will be applied on a batch basis.

1.3.2.2 Methodology The dose projection for each batch is calculated in the same manner as cumulative dose calculations for the current calendar quarter and current calendar year. See It-1. 1.4.4 and II-1.3.1. If the calculated dose is greater than 0.06 mrem to the total body or 0.2 mrem to any organ, the appropriate subsystems of the liquid radwaste system shall be used to reduce the radioactivity levels of the batch prior to release.

1.3.2.3 Continuous Liquid Release Dose Projections Each month that a continuous liquid release is in progress, or is anticipated, the expected dose to man can be accounted for or projected. Since a continuous release does not result from not operating a portion of the Liquid Radwaste System, projections are not required to determine or evaluate Radwaste System Operability. Dose projections may be relevant to planning repairs, and in reporting intended actions. See 11-1.1.4.5.

Unit 1 ODCM Revision 24 I111 July 2003

2.0 GASEOUS EFFLUENTS 2.1 Setpoint Determinations 2.1.1 Basis Stack gas monitor setpoints will be established such that the instantaneous release rate of radioactive materials in gaseous effluents does not exceed the 10 CFR 20 limits for annual release rate. The setpoints will be activated if the instantaneous dose rate at or beyond the (land) site boundary would exceed 500 mrem/yr to the whole body or 3000 mrem/yr to the skin from the continuous release of radioactive noble gas in the gaseous effluent.

The offgas (condenser air ejector activity) monitor setpoints provide assurance that the total body exposure to an individual at the exclusion area boundary does not exceed a small fraction of the dose guidelines of 10 CFR 100.

Emergency condenser vent monitor setpoints will be established such that the release rate for radioactive materials in gaseous effluents do not exceed the Technical Specification dose rate limits. Monitor setpoints for emergency condenser vent monitors are conservatively fixed at 5 mr/hr for reasons described in Section 11-2.1.4 and therefore do not require periodic recalculations.

Monitor setpoints from continuous release points will be determined once per quarter under normal release rate conditions and will be based on the isotopic composition of the actual release in progress, or an offgas isotopic distribution or a more conservative default composition specified in the pertinent procedure. If the calculated setpoint is higher than the existing setpoint, it is not mandatory that the setpoint be changed.

Under abnormal site release rate conditions, monitor alarm setpoints from continuous release points will be recalculated and, if necessary, reset at more frequent intervals as deemed necessary by Chemistry Supervision. In particular, contributions from both JAF and NMP-2 and the Emergency Condenser Vents shall be assessed.

During outages and until steady state power operation is again realized, the last operating stack and off gas monitor alarm setpoints shall be used.

Since monitors respond to noble gases only, monitor alarm points are set to alarm prior to exceeding the corresponding whole body dose rates.

The skin dose rate limit is not used in setpoint calculations because it is never limiting.

2.1.2 Stack Monitor Setpoints The detailed methods for establishing setpoints shall be contained in the station procedures.

These methods shall apply the following general criteria:

(1) Rationale for Stack monitor settings is based on the general equation:

release rate. actual = release rate. max. allowable corresp. dose rate, actual corresp. dose rate, max. allowable El_Q_ (Q)max Q, (VI +(SF) K, (X /Q),) 500mremlyr Unit 1 ODCM Revision 24 1 12 July 2003

2.1.2 Stack Monitor Setpoints (Cont'd)

Where:

Q = release rate for each isotope i, ttCi/sec.

Vi = gamma whole body dose factor in units of mremlyr per PCi/sec. (See Table 3-2).

(Q) = instantaneous release rate limit PCi/sec.

SF, Ki, X/Q = See Section II-2.2.1.1.

(2) To ensure that Part I dose rate limits are not exceeded, the Hi Hi alarms on the stack monitors shall be set lower than or equal to (0.9) (Q),,,. HI alarms shall be set lower than or equal to (0.5)

(Q)

M .

(3) Based on the above conservatism, the dose contribution from JAF and NMP-2 can usually be ignored. During Emergency Classifications at JAF or NMP-2 due to airborne effluent, or after emergency condenser vent releases of significant proportions, the 500 mremfyr value may be reduced accordingly.

(4) To convert monitor gross count rates to pCi/sec release rates, the following general formula shall be applied:

(Cm-B) K, = Q = tPCi/sec, release rate Where:

Cm = monitor gross count rate in cps or cpm B = monitor background count rate K, = stack monitor efficiency factor with units of PCi/sec-cps or PCi/sec-cpm (5) Monitor K, factors shall be determined using the general formula:

K, = E IQ/(Cm-B)

Where:

Qj = individual radionuclide stack effluent release rate as determined by isotopic analysis.

K, factors more conservative than those calculated by the above methodology may be assumed.

Alternatively, when stack release rates are near the lower limit of detection, the following general formula may be used to calculate K,:

1/K. E (_ F.

F, kYkEk) (3.7E4 dis) f f sec - pCi Where:

f = stack flow in cc/sec.

E = efficiency in units of cpm-cc/PCi or cps-cc/PCi (cpm = counts per minute; cps = counts per second).

Ek = cpm-cc/bps or cps-cctrps.

From energy calibration curve produced during NIST traceable primary gas calibration or transfer source calibration (bps = beta per second;Yps = gammas per second).

Unit 1 ODCM Revision 24 1 13 July 2003

2.1.2 Stack Monitor Setpoints (Cont'd)

Yk = bid (betas/disintegration) or Ad (gammas/disintegration).

Fi = Activity fraction of nuclide i in the mixture.

i = nuclide counter.

k = discrete energy beta or gamma emitter per nuclide counter.

s = seconds.

This monitor calibration method assumes a noble gas distribution typical of a recoil release mechanism. To ensure that the calculated efficiency is conservative, beta or gamma emissions whose energy is above the range of calibration of the detector are not included in the calculation.

2.1.3 Recombiner Discharge (Off Gas) Monitor Setpoints (1) The Hi-Hi alarm points shall activate with recombiner discharge rates equal to or less than 500,000 PCi/sec. This alarm point may be set equal to or less than 1 Ci/sec for a period of time not to exceed 60 days provided the offgas treatment system is in operation.

According to Part I, Note (c) to Table D 4.6.14-2, the channel functional test of the condenser air ejector radioactivity monitor shall demonstrate that automatic isolation of this pathway occurs if either of the following conditions exist:

i) Instruments indicate two channels above the Hi-Hi alarm setpoint, ii) Instruments indicate one channel above Hi-Hi alarm setpoint and one channel downscale.

This automatic isolation function is tested once per operating cycle in accordance with station procedures.

(2) The Hi alarm points shall be set to activate at equal to or less than five (5) times normal full power background.

If the monitor alarms at this setpoint, the offgas will be immediately sampled and analyzed, followed by an analysis of reactor coolant sample.

(3) To convert monitor mR/hr readings to PCifsec, the formula below shall be applied:

(R)(KR) = QR PCi/sec recombiner discharge release rate Where:

R = mR/hr monitor indicator.

KR = efficiency factor in units of PCi/sec/mR/hr determined prior to setting monitor alarm points.

(4) Monitor KR factors shall be determined using the general formula:

KR= £iQ/R Where:

Q = individual radionuclide recombiner discharge release rate as determined by isotopic analysis and flow rate monitor.

KR factors more conservative than those calculated by the above methodology may be assumed.

Unit 1 ODCM Revision 24 1 14 July 2003

2.1.3 Recombiner Discharge (Off Gas) Monitor Setpoints (Cont'd)

(5) The setpoints chosen provide assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a very small fraction of the limits of 10CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment (thereby implementing the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50). Additionally, these setpoints serve to limit buildup of fission product activity within the station systems which would result if high fuel leakage were to be permitted over extended periods.

2.1.4 Emergency Condenser Vent Monitor Setpoint The monitor setpoint was established by calculation ("Emergency Condenser Vent Monitor Alarm Setpoint", January 13, 1986, NMPC File Code #16199). Assuming a hypothetical case with (1) reactor water iodine concentrations higher than the Technical Specification Limit, (2) reactor water noble gas concentrations higher than would be expected at Technical Specification iodine levels, and (3) leakage of reactor steam into the emergency condenser shell at 300% of rated flow (or 1.3 E6 lbslhr), the calculation predicts an emergency condenser vent monitor response of 20 mR/hr. Such a release would result in less than 10 CFR 20 dose rate values at the site boundary and beyond for typical emergency condenser cooldown periods.

Since a 20 mR/hr monitor response can, in theory, be achievable only when reactor water iodines are higher than permitted by Technical Specifications, a conservative monitor setpoint of 5 mr/hr has been adopted.

2.1.5 Discussion 2.1.5.1 Stack Effluent Monitoring System Description The NMP-1 Stack Effluent Monitoring System consisted of two subsystems; the Radioactive Gaseous Effluent Monitoring System (RAGEMS) and the Offgas Effluent Stack Monitoring System (OGESMS). The OGESMS shall be used to monitor station noble gas effluents and collect particulates and iodine samples in compliance with Part I requirements.

The RAGEMS was designed to be promptly activated from the Main Control Room for use in high range monitoring during accident situations in compliance with NUREG 0737 criteria. In accordance with a letter dated September 11, 2002 from the NRC to NMPNS, LLC, "Nine Mile Point Nuclear Station Unit 1- Use of the Offgas Effluent Stack Monitoring System to Meet Regulatory Guide 1.97, Revision 2 and NUREG-0737,"

OGESMS meets the objective and purpose of NUREG-0737 and RG 1.97. The sample line to RAGEMS will now be used as an additional auxiliary sample point.

2.1.5.2 Stack Sample Flow Path - RAGEMS Auxiliary Sample Point The effluent sample is obtained inside the stack at elevation 530' using an isokinetic probe with four orifices. The sample line then bends radially out and back into the stack; descends down the stack and out of the stack at approximately elevation 257'; runs horizontally (enclosed in heat tracing) some 270' along the off gas tunnel; and enters Turbine Building 250' and Offgas Building 247'.

Unit 1 ODCM Revision 24 1 15 July 2003

2.1.53 Stack Sample Flow Path - OGESMS The OGESMS sample is obtained from the same stack sample probe as the RAGEMS Auxiliary Sample Point. From the exit of the stack at elevation 257, the sample line runs east approximately 20' and then vertically approximately 8' to the OGESMS skid. In the OGESMS, sample flows thru a particulate/iodine cartridge housing and four noble gas scintillation detectors (i.e., 07 and 08 low range beta detectors and RN-03A and RN-03B high range gamma detectors). From OGESMS, the stack sample flows back into the stack at approximately elevation 257'.

All OGESMS detector outputs are monitored and recorded remotely in the Main Control Room. Alarming capabilities are provided to alert Operators of high release rate conditions prior to exceeding Part I Control DLCO 3.6.15.b (1)(a) whole body dose rate limits.

Stack particulate and iodine samples are retrieved manually from the OGESMS and analyzed in the laboratory using gamma spectroscopy at frequencies and LLDs specified in Part I Table D 4.6.15-2.

2.1.5.4 Sampling Frequency/Sample Analysis Radioactive gaseous wastes shall be sampled and analyzed in accordance with the sampling and analysis program specified in Part I Table D 4.6.15-2. Noble gas sample and analysis frequencies are increased during elevated release rate conditions. Noble gas sample and analysis are also performed following startup, shutdown and in conjunction with each drywell purge. Particulate samples are saved and analyzed for principal gamma emitters, gross alpha, Fe-55, Sr-89, Sr-90 at monthly intervals minimally, and in response to an increase in noble gas release rate. The latter three analyses are performed off-site from a composite sample.

Consistent with Part I Table D 4.6.15-2, stack effluent tritium is sampled monthly, during each drywell purge, and weekly when fuel is off loaded until stable release rates are demonstrated. Samples are analyzed off-site.

Line loss correction factors are applied to all particulate and iodine results. Correction factors of 2.0 and 1.5 are used for data obtained from RAGEMS Auxiliary Sample Point and OGESMS respectively. These correction factors are based on empirical data from sampling conducted at NMP-1 in 1985 (memo from J. Blasiak to RAGEMS File, 1/6/86, "Stack Sample Representativeness Study: RAGEMS versus In-Stack Auxiliary Probe Samples").

2.1.5.5 1-133 and 1-135 Estimates Monthly, the stack effluent shall be sampled for iodines over a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period and the 1-13541-131 and the I-133/I-131 ratios calculated. These ratios shall be used to calculate I-133, 1-135 release for longer acquisition samples collected during the month.

Unit 1 ODCM Revision 24 1 16 July 2003

2.1.5.5 1-133 and 1-135 Estimates (Cont'd)

Additionally, the 1-135/1-131 and I-133/1-131 ratios should also be determined after a significant change in the ratio is suspected (eg, plant status changes from prolonged shutdown to power operation or fuel damage has occurred). I-135 will be included in the Radioactive Effluent Release Report in accordance with Regulatory Guide 1.21 but it will not be included when totaling dose rate or dose.

2.1.5.6 Gaseous Radwaste Treatment System Operation Part I Control DLCO 3.6.16.b requires that the gaseous radwaste treatment system shall be operable and shall be used to reduce radioactive materials in gaseous waste prior to their discharge as necessary to meet the requirements of Part I Control DLCO 3.6.15.b.

To ensure Part I Control DLCO 3.6.15.b limits are not exceeded, and to confirm proper radwaste treatment system operation as applicable, cumulative dose contributions for the current calendar quarter and current calendar year shall be determined monthly in accordance with section 2.2 of this manual. Initial dose calculations shall incorporate the following assumptions with regard to release rates of non-gamma emitting radionuclides subsequently analyzed off-site:

a) H-3 release rate

4 PCi/sec b) Sr-89 release rate

4 x Cs-137 release rate c) Sr-90 release rate

0.5 x Cs-137 release rate d) Fe-55 release rate

1 x Co-60 release rate Assumed release rates represent conservative estimates derived from analysis of historical data from effluent releases and process waste streams (See NMP 34023, C. Ware to J.

Blasiak, April 29, 1988, "Dose Estimates for Beta-Emitting Isotopes"). Following receipt of off-site H-3, Sr-89, Sr-90, Fe-55 analysis information, dose estimates shall be revised using actual radionuclide concentrations.

2.2 Dose and Dose Rate Determinations In accordance with Technical Specifications 6.5.3, "Radioactive Effluent Controls Program, and ODCM Part I Controls DSR 4.6.15.b.(l), DSR 4.6.15.b.(2), and DSR 4.6.15.b.(3) dose and dose rate determinations will be made monthly to determine:

(1) Whole body dose rates and gamma air doses at the maximum X/Q land sector site boundary interface.

(2) Skin dose rates and beta air doses at the maximum X/Q land sector site boundary interface.

(3) The critical organ dose and dose rate at a critical receptor location beyond the site boundary.

Average meteorological data (ie, maximum five year annual average X/Q and D/Q values in the case of elevated releases or 1985 annual average X/Q and DIQ values, in the case of ground level releases) shall be utilized for dose and dose rate calculations. Where average meteorological data is assumed, dose and dose rates due to noble gases at locations beyond the site boundary will be lower than equivalent site boundary dose and dose rates.

Therefore, under these conditions, calculations of noble gas dose and dose rates beyond the maximum X/Q land sector site boundary locations can be neglected.

Unit 1 ODCM Revision 24 11 17 July 2003

2.2 Dose and Dose Rate Determinations (Cont'd)

The frequency of dose rate calculations will be upgraded when elevated release rate conditions specified in subsequent sections II-2.2.1.1 and 11-2.2.1.2 are realized.

In accordance with Technical Specification 6.5.3.g, noble gas dose rate to the whole body and skin will be calculated at the site boundary. In accordance with Technical Specification 6.5.3.h, gamma and beta air doses may be calculated at a point beyond the site boundary.

To demonstrate compliance with Technical Specification 6.5.3, "Radioactive Effluent Controls Program", critical organ doses and dose rates may be conservatively calculated by assuming the existence of a maximum individual. This individual is a composite of the highest dose factor of each age group, for each organ and total body, and each nuclide. It is assumed that all pathways are applicable and the highest X/Q and/or D/Q value for actual pathways as noted in Table 3-1 are in effect. The maximum individual's dose is equal to the same dose that person would receive if they were simultaneously subjected to the highest pathway dose at each critical receptor identified for each pathway. The pathways include grass-(cow and goat)-milk, grass-cow-meat, vegetation, ground plane and inhalation. To comply with Part I requirements the maximum individual dose rate will be calculated at this hypothetical critical residence.

If dose or dose rates calculated, using the assumptions noted above, reach Part I limits, actual pathways will be evaluated, and doseldose rates may be calculated at separate critical receptor locations and compared with applicable limits.

Emergency condenser vent release contributions to the monthly dose and dose rate determinations will be considered only when the emergency condenser return isolation valves have been opened for reactor cooldown, if Emergency Condenser tube leaks develop with or without the system's return isolation valve opened, or if significant activity is detected in the Emergency Condenser Shell.

Without tube leakage, dose contributions from emergency condenser vent releases are to be determined based on condensate storage tank and emergency condenser shell isotopic distributions.

When releases from the emergency condenser have occurred, dose rate and dose determinations shall be performed using methodology in 11-2.2.1 and 1-2.2.2. Furthermore, environmental sampling may also be initiated to refine any actual contribution to doses. See Section E-2.4.

2.2.1 Dose Rate Dose rates will be calculated monthly, at a minimum, or when the Hi-Hi stack monitor alarm setpoint is reached, to demonstrate that dose rates resulting from the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the limits specified in Technical Specifications Section 6.5.3, "Radioactive Effluent Controls Program".

These limits are:

Noble Gases Whole Body Dose Rate: 500 mrem/yr Skin Dose Rate: 3000 mrem/yr Tritium. lodines and Particulates Organ Dose Rate: 1500 mrem/yr Unit 1 ODCM Revision 24 1 18 July 2003

2.2.1.1 Noble Gases The following noble gas dose rate equation includes the contribution from the stack (s) elevated release and the emergency condenser vent (v) ground level release when applicable (See section EI-2.2).

For whole body dose rates (mremlsec):

DRy (mremlsec) = 3.17E-8 F I[(VI + (SF) K3 (X/Q).) Q. + (SF)K1 (X/Q)XQV]

For skin dose rates (inren-/sec):

DRyw(mrem/sec)= 3.17E-8 i [(L-(X/Q). + 1.1 (SF)(B1 + Mi(X/Q),))Qi, +

(L + 1.1 I(SF)Mi)(UQ)vQi,]

Where:

DRY = whole body gamma dose rate (mrem/sec).

DRAY = skin dose rate from gamma and beta radiation (mremlsec).

VI = the constant accounting for the gamma whole body dose rate from stack radiation for an elevated finite plume releases for each identified noble gas nuclide, i. Listed on Table 3-2 in mrem/yr per PCi/sec.

K1 = the constant accounting for the gamma whole body dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed in Table 3-3 in mrem/yr per PCCi/m 3 (from Reg.

Guide 1.109)

QaQiv = the release rate of isotope i from the stack(s) or emergency condenser vent(v); (pCi/sec)

SF = structural shielding factor.

X/Q = the relative plume concentration (in units of sec/n 3 ) at the land sector site boundary or beyond. Average meteorological data (Table 3-1) is used. "Elevated" X/Q values are used for stack releases (s = stack);

"Ground" X/Q values are used for Emergency Condenser Vent releases (v = vent).

14 = the constant accounting for the beta skin dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed in Table 3-3 in mremtyr per PCilm3 (from Reg. Guide 1.109)

B1 = the constant accounting for the air gamma radiation from the elevated Finite plume resulting from stack releases for each identified noble gas nuclide, i. Listed in Table 3-2 in mrad/yr per PCi/sec.

Unit 1 ODCM Revision 24 H 19 July 2003

2.2.1.1 Noble Gases (Cont'd)

M; = the constant accounting for the gamma air dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed in Table 3-3 in mrad/yr per pCi/r 3 (from Reg. Guide 1.109)

See Appendix B for derivation of Bi and Vs.

To ensure that the site noble gas dose rate limits are not exceeded, the following procedural actions are taken if the offsite dose rates from Unit 1 exceed 10% of the limits:

1) Notify Unit 1 SSS (Station Shift Supervisor) and Unit 1 Supervisor Chemistry.

2) Notify Unit 2 SSS and Unit 2 Supervisor Chemistry and request the Unit 2 contribution to offsite dose rate.

3) Notify SSS of the James A. Fitzpatrick Nuclear Plant and request the Fitzpatrick contribution to offsite dose rate.

4) Increase the frequency of performing noble gas dose and dose rate calculations, if necessary, to ensure Site (Nine Mile Point Units 1 and 2 and Fitzpatrick) limits are not exceeded.

Additionally, alarm setpoints are set at 90% of the dose rate limit to ensure that site limits are not exceeded. This alarm setpoint is adjusted if the noble gas dose rate from Unit 1 is greater than 10% of the limit.

2.2.1.2 Tritium, lodines and Particulates To ensure that the 1500 mrern/year site dose rate limit is not exceeded, offsite dose rates for tritium, iodine and particulates with half lives greater than 8 days shall be calculated monthly and when release rates (Q) exceed 0.34 PCi/sec using the following equation.

Dgk (mremlsec) = 3.17E-8 j[ i Rijk [W. Q. + W, QWIJ Where:

DA, = Total dose rate to each organ k of an individual in age group a (mremn/sec).

W1 = dispersion parameter either X/Q (sec/mi3 ) or D/Q (1/mn2 ) depending on pathway and receptor location assumed. Average meteorological data is used (Table 3-1). "Elevated" Wj values are used for stack releases (s = stack); "Ground" Wj values are used for Emergency Condenser Vent releases (v = vent).

Qi = the release rate of isotope i, from the stack (s) or vent(v); (ILCi/sec).

Unit 1 ODCM Revision 24 1 20 July 2003

2.2.1.2 Tritium, Iodines and Particulates (Cont'd)

Rijak: = the dose factor for each isotope i, pathway j, age group a, and organ k (Table 3-4, through 3-22; m2 -mremlyr per PCi/sec for all pathways except inhalation, mrem/yr per PCi/m3 . The R values contained in Tables 3-4 through 3-22 were calculated using the methodology defined in NUREG-0133 and parameters from Regulatory Guide 1.109, Revision 1; as presented in Appendix C.

3.17E-8 = the inverse of the number of seconds in a year.

The use of the 0.34 PCi/sec release rate threshold to perform Unit 1 dose rate calculations is justified as follows:

(a) The 1500 rnreml/yr organ dose rate limit corresponds to a minimum release rate limit of 0.34 PCi/sec calculated using the equation:

1500 = (Q, PCi/sec) x (RijWj).

Where:

1500 = site boundary dose rate limit (mnrem/year).

(Rij~j), = the maximum curie-to-dose conversion factor equal to 4.34E3 mrem-sec/iCi-yr for Sr-90, child bone for the vegetation pathway at the critical residence receptor location beyond the site boundary for an elevated release.

(b) The use of 0.34 PCi/sec release rate threshold and the 4.34E3 mrem-sec/pCi-yr curie-to-dose conversion factor is conservative since curie-to-dose conversion factors for other isotopes likely to be present are significantly lower.

If the organ dose rate exceeds 5% of the annual limit, the following procedural actions will be taken:

1) Notify Unit 1 SSS (Station Shift Supervisor) and Unit 1 Supervisor Chemistry.

2) Notify Unit 2 SSS and Unit 2 Supervisor Chemistry and request the Unit 2 contribution to offsite dose rate.

3) Notify SSS of James A. Fitzpatrick Nuclear Plant and request JAF's contribution to offsite dose rate.

4) Increase the frequency of performing dose and dose rate calculations if necessary to ensure site (Nine Mile Point Units 1 and 2 and Fitzpatrick) limits are not exceeded.

Unit 1 ODCM Revision 24 1 21 July 2003

2.2.2 Dose Calculations will be performed monthly at a minimum, to demonstrate that doses resulting from the release of noble gases, tritium, iodines, and particulates with half lives greater than 8 days are within the limits specified in 10 CFR 50, Appendix I.

These limits are:

Noble Gases 5 mR gamma/calendar quarter 10 mrad beta/calendar quarter 10 mR gamma/calendar year 20 mrad beta/calendar year Tritium. Iodines and Particulates 7.5 mrem to any organ/calendar quarter 15 mrem to any organ/calendar year 2.2.2.1 Noble Gas Air Dose The following Noble Gas air dose equation includes contributions from the stack (s) elevated release and the emergency condenser vent (v) ground level release when applicable (see section II-2.2):

For gamma radiation' (nirad):

Dy (mrad) = 3.17E-8 F I[(B1 + Mi(X/Q).) Q1 + M,(X/Q)v Qj, ] t For beta radiation (mrad):

Dp (mrad) = 3.17E-8 F iNi[(X/Q) Qi, + (X/Q), Qj, ] t Where:

Dy = gamma air dose (mrad).

Do = beta air dose (mrad).

Note that the units for the gamma air dose are in mrad compared to the units for the limits are in mR. The NRC recognizes that 1 mR=l mrad, for gamma radiation.

Bi = the constant accounting for the air gamma radiation from the elevated finite plume resulting from stack releases for each identified noble gas nuclide, i. Listed in Table 3-2 in mrad/yr per PCi/sec.

N, = the constant accounting for the air beta dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i.

Listed on Table 3-3 in mrad/yr per IPCi/m 3 (from Reg. Guide 1.109).

QiI Qi, = the release rate of isotope i, from the stack (s) or vent (v);

(ACi/sec).

Unit 1 ODCM Revision 24 1 22 July 2003

2.2.2.1 Noble Gas Air Dose (Cont'd) 3.17E-8 = the inverse of the number of seconds in a year.

M; the constant accounting for the air gamma dose from immersion in the semi-infinite cloud for each identified noble gas nuclide, i.

Listed on Table 3-3 in mrad/yr per pCi/m3 (from Reg. Guide 1.109).

t = total time during release period, sec.

All other parameters are as defined in section II-2.2.1.1.

2.2.2.2 Tritium, Iodines and Particulates To ensure that the 15 mremlyr facility dose limit is not exceeded, offsite doses for tritium, iodines, and particulates with half lives greater than 8 days shall be calculated monthly using the following equation:

D~k (mrem) = 3.17E-8 j[Z i Rqk [W. Q, + W, Qi-] ] t Where:

Dk total dose to each organ k of an individual in age group a(inrem).

W- dispersion parameter either X/Q (sec/m 3 ) or D/Q (/rm 2 )

depending on pathway and receptor location assumed. Average meteorological data is used (Table 3-1). "Elevated" Wj values are used for stack releases (s = stack); "Ground" Wj values are used for Emergency Condenser Vent releases (v = vent).

Qi-QiI= the release rate of isotope i from stack(s) or vent (v); (gCi/sec).

Rij the dose factor for each isotope i, pathway j, age group a, and organ k (Tables 34, through 3-22; n?-mrem/yr per pCi/sec). R values contained in Tables 3-4 through 3-22 were calculated using the methodology defined in NUREG-0133 and parameters from Regulatory Guide 1.109, Revision 1; as presented in Appendix C.

3.17E-8 = the inverse of the number of seconds in a year.

t = total time during the release period, sec.

Unit 1 ODCM Revision 24 II 23 July 2003

2.2.2.3 Accumulating Doses Doses will be calculated monthly, at a minimum, for gamma air, beta air, and the critical organ for each age group. Dose estimates will, also, be calculated monthly prior to receipt of any offsite analysis data i.e., strontium, tritium, and iron-55. Results will be summed for each calendar quarter and year.

The critical doses are based on the following:

- noble gas plume air dose

- direct radiation from ground plane deposition

- inhalation dose

- cow milk ingestion dose

- goat milk ingestion dose

- cow meat ingestion dose

- vegetation (food crops) ingestion dose The quarterly and annual results shall be compared to the limits listed in paragraph 11-2.2.2. If the limits are exceeded, special reports, as required by Part I Section D 6.9.3 shall be submitted.

2.2.3 Dose Projections - Determination of Need to Operate Gaseous Radwaste Treatment System and Ventilation Exhaust Treatment System 2.2.3.1 Requirement DLCO 3.6.16.b requires that the Gaseous Radwaste Treatment System be used to reduce the radioactive materials in gaseous waste prior to their discharge as necessary to meet the requirements of DLCO 3.6.15. DLCO 3.6.16.b(2) requires that the Ventilation Exhaust Treatment System be used to reduce releases of radioactivity when the projected doses in 31 days would exceed 0.3mrem to any organ. These Controls implement Technical Specification 6.5.3.f that requires the Radioactive Effluent Control Program to include limitations on the functional capability and use of the gaseous effluent treatment systems (Gaseous Radwaste Treatment System AND Ventilation Exhaust Treatment System) to ensure the appropriate portions of these systems are used to reduce releases of radioactivity. The Gaseous Radwaste Treatment System is expected to be in service. For the Ventilation Exhaust Treatment System, use is required when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10CFR50, Appendix I, i.e., 3 mrem to any organ. When Treatment systems are not in use, doses are to be projected every 31 days.

The appropriate components, which affect iodine or particular release, to be in use are:

Rad Waste Building FLT-204-24 FLT-204-25 FLT-204-69 FLT-204-70 RSSB FLT-204-147 Unit I ODCM Revision 24 H 24 July 2003

2.2.3.2 Methodolgy Due to system design and operating procedures the charcoal beds are always operated when the offgas system is in operation. Therefore, dose projection is not relevant to determining need to operate.

If the Gaseous Radwaste Treatment System becomes inoperable for more than seven days a Special Report to the NRC is required. This report will include appropriate dose assessments (cumulative and projected).

If Ventilation Exhaust Treatment System components become inoperable which prevent building effluent from being filtered, dose projections will be performed monthly using the methodology of Section II-2.2.2.2. Assumptions for released activity will be added to historical routine stack emissions for calculating dose during the anticipated period of component unavailability. The calculated projected doses for iodine and particulates will be compared to the DLCO 3.6.15.b limits and Technical Specifications Section 6.5.3.f limit, 0.3 mrem to any organ.

2.3 Critical Receptors In accordance with the provisions of 10 CFR 20 and 10 CFR 50, Appendix I, the critical receptors have been identified and are contained in Table 3-1.

For elevated noble gas releases the critical receptor is the site boundary.

When 1985 average annual XIQ values are used for ground level noble gas releases, the critical receptor is the maximum X/Q land sector site boundary interface.

For tritium, iodines, and particulates with half lives greater than eight days, the critical pathways are grass-(cow and goat)-milk, grass-cow meat, vegetation, inhalation and direct radiation (ground plane) as a result of ground deposition.

The grass-(cow and goat)-milk, and grass-cow-meat pathways will be based on the greatest D/Q location. This location has been determined in conjunction with the land use census (Part I Control DLCO 3.6.22) and is subject to change. The vegetation (food crop) pathway is based on the greatest D/Q garden location from which samples are taken. This location may also be modified as a result of vegetation sampling surveys.

The inhalation and ground plane dose pathways will be calculated at the critical residence.

Because Part I states to calculate "at the site boundary or beyond", the doses and/or dose rates must be calculated for a maximum individual who is exposed to applicable pathways at the critical residence. The maximum individual is a composite of the highest dose factor of each age group, for each organ and total body, and each nuclide.

Unit 1 ODCM Revision 24 1 25 July 2003

2.4 Refinement of Offsite Doses Resulting from Emergency Condenser Vent Releases The doses resulting from the operation of the emergency condensers and calculated in accordance with E-2.2.2 may be refined using data from actual environmental samples.

Ground deposition samples will be obtained from an area or areas of maximum projected deposition. These areas are anticipated to be at or near the site boundary and near projected plume centerline. Using the methodology found in Regulatory Guide 1.109, the dose will be calculated to the maximum exposed individual. This dose will then be compared to the dose calculated in accordance with II-2.2.2. The comparison will result in an adjustment factor of less than or greater than one which will be used to adjust the other doses from other pathways. Other environmental samples may also be collected and the resultant calculated doses to the maximum exposed individual compared to the dose calculated per 11-2.2.2. Other environmental sample media may include milk, vegetation (such as garden broadleaf vegetables), etc. The adjustment factors from these pathways may be applied to the doses calculated per II-2.2.2 on a pathway by pathway basis or several pathway adjustment factors may be averaged and used to adjust calculated doses.

Doses calculated from actual environmental sample media will be based on the methodology presented in Regulatory Guide 1.109. The regulatory guide equations may be slightly modified to account for short intervals of time (less than one year) or modified for simplicity purposes by deleting decay factors. Deletion of decay factors would yield more conservative results.

Unit 1 ODCM Revision 24 1 26 July 2003

3.0 40 CFR 190 REQUIREMENTS The "Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b) as follows:

"Uranium fuel cycle means the operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the reuse of recovered non-uranium special nuclear and by-product materials from the cycle."

Control DLCO 3.6.15.d of Part I requires that when the calculated doses associated with the effluent releases exceed twice the applicable quarter or annual limits, the licensee shall evaluate the calendar year doses and, if required, submit a Special Report to the NRC and limit subsequent releases such that the dose commitment to a real individual from all uranium fuel cycle sources is limited to 25 inrem to the total body or any organ (except the thyroid, which is limited to 75 mrem). This report is to demonstrate that radiation exposures to all real individuals from all uranium fuel cycle sources (including all liquid and gaseous effluent pathways and direct radiation) are less than the limits in 40 CFR Part 190. If releases that result in doses exceeding the 40 CFR 190 limits have occurred, then a variance from the NRC to permit such releases will be requested and if possible, action will be taken to reduce subsequent releases.

The report to the NRC shall contain:

1) Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site that contribute to the annual dose of the maximum exposed member of the public.

2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from existing pathways and sources of radioactive effluents and direct radiation.

The total body and organ doses resulting from radioactive material in liquid effluents from Nine Mile Point Unit 1 will be summed with the maximum doses resulting from the releases of noble gases, radioiodines, and particulates for the other calendar quarters (as applicable) and from the calendar quarter in which twice the limit was exceeded. The direct dose components will be determined by either calculation or actual measurement.

Actual measurements will utilize environmental TLD dosimetry. Calculated measurements will utilize engineering calculations to determine a projected direct dose component. In the event calculations are used, the methodology will be detailed as required in Part I Section D 6.9.1.e.

Unit 1 ODCM Revision 24 11 27 July 2003

3.0 40 CFR 190 REQUIREMENTS (Cont'd)

The doses from Nine Mile Point Unit 1 will be added to the doses to the maximum exposed individual that are contributed from other uranium fuel cycle operations within 5 miles of the site. Other uranium fuel cycle facilities within 5 miles of the Site include Nine Mile Point Nuclear Station Unit 2 and the James A. Fitzpatrick Nuclear Power Plant. Doses from other facilities will be calculated in accordance with each facilities' ODCM.

For the purpose of calculating doses, the results of the Radiological Environmental Monitoring Program may be included for providing more refined estimates of doses to a real maximum exposed individual. Estimated doses, as calculated from station effluents, may be replaced by doses calculated from actual environmental sample results. Reports will include all significant details of the dose determination if radiological sampling and analyses are used to determine if the dose limits of 40CFR190 are exceeded.

3.1 Evaluation of Doses From Liquid Effluents For the evaluation of doses to real members of the public from liquid effluents, the fish consumption and shoreline sediment ground dose will be considered. Since the doses from other aquatic pathways are insignificant, fish consumption and shoreline sediment are the only two pathways that will be considered. The dose associated with fish consumption may be calculated using effluent data and Regulatory Guide 1.109 methodology or by calculating a dose to man based on actual fish sample analysis data. Because of the nature of the receptor location and the extensive fishing in the area, the critical individual may be a teenager or an adult. The dose associated with shoreline sediment is based on the assumption that the shoreline would be utilized as a recreational area. This dose may be derived from liquid effluent data and Regulatory Guide 1.109 methodology or from actual shoreline sediment sample analysis data.

Equations used to evaluate doses from actual fish and shoreline sediment samples are based on Regulatory Guide 1.109 methodology. Because of the sample medium type and the half-lives of the radionuclides historically observed, the decay corrected portions of the equations are deleted. This does not reduce the conservatism of the calculated doses but increases the simplicity from an evaluation point of view. Table 3-23 presents the parameters used for calculating doses from liquid effluents.

The dose from fish sample media is calculated as:

Rapj = i [Cif (U)(DiW) ft (1E+3)

Where:

Rmj = The total annual dose to organ j, of an individual of age group a, from nuclide i, via fish pathway p, in mrem per year.

Cif = The concentration of radionuclide i in fish samples in pCi/gram.

U = The consumption rate of fish in kg/yr.

1E+3 = Grams per kilogram.

Unit 1 ODCM Revision 24 11 28 July 2003

3.1 Evaluation of Doses From Liquid Effluents (Cont'd)

(Dwpj) = The ingestion dose factor for age group a, nuclide i, fish pathway p, and organ j, (Reg. Guide 1.109, Table E-11) (inremlpCi).

f = The fractional portion of the year over which the dose is applicable.

The dose from shoreline sediment sample media is calculated as:

,pj = [C, (UJ)(4E+4)(0.3)(Dgj) fi Where:

R~pj = The total annual dose to organ j, of an individual of age group a, from nuclide i, via the sediment pathway p, in mrem per year.

C. = The concentration of radionuclide i in shoreline sediment in pCi/gram.

U = The usage factor, (hr/yr) (Reg. Guide 1.109).

4E+4 = The product of the assumed density of shoreline sediment (40 kilogram per square meter to a depth of 2.5 cm) times the number of grams per kilogram.

0.3 = The shore width factor for a lake.

D,,pj = The dose factor for age group a, nuclide i, sediment pathway s, and organ j. (Reg. Guide 1.109, Table E-6)(mrem/hr per pCi/mi).

f = The fractional portion of the year over which the dose is applicable.

3.2 Evaluation of Doses From Gaseous Effluents For the evaluation of doses to real members of the public from gaseous effluents, the pathways contained in section II-2.2.2.3 of the ODCM will be considered. These include the deposition, inhalation cows milk, goats milk, meat, and food products (vegetation). However, any updated field data may be utilized that concerns locations of real individuals, real time meteorological data, location of critical receptors, etc. Data from the most recent census and sample location surveys should be utilized. Doses may also be calculated from actual environmental sample media, as available. Environmental sample media data such as TLD, air sample, milk sample and vegetable (food crop) sample data may be utilized in lieu of effluent calculational data.

Doses to member of the public from the pathways contained in ODCM section 11-2.2.2.3 as a result of gaseous effluents will be calculated using the dose factors of Regulatory Guide 1.109 or the methodology of the ODCM, as applicable. Doses calculated from environmental sample media will be based on the methodologies found in Regulatory Guide 1.109.

Unit 1 ODCM Revision 24 11 29 July 2003

3.3 Evaluation of Doses From Direct Radiation The dose contribution as a result of direct radiation shall be considered when evaluating whether the dose limitations of 40 CFR 190 have been exceeded.

Direct radiation doses as a result of the reactor, turbine and radwaste buildings and outside radioactive storage tanks (as applicable) may be evaluated by engineering calculations or by evaluating environmental TLD results at critical receptor locations, site boundary or other special interest locations. For the evaluation of direct radiation doses utilizing environmental TLDs, the critical receptor in question, such as the critical residence, etc., will be compared to the control locations. The comparison involves the difference in environmental TLD results between the receptor location and the average control location result.

3.4 Doses to Members of the Public Within the Site Boundary The Radioactive Effluent Release Report shall include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary as defined by Figure 5.1-1 of the Technical Specifications. A member of the public, as defined in Part I, would be represented by an individual who visits the site's Energy Center for the purpose of observing the educational displays or for picnicking and associated activities.

Fishing is a major recreational activity in the area and on the Site as a result of the salmonoid and trout populations in Lake Ontario. Fishermen have been observed fishing at the shoreline near the Energy Center from April through December in all weather conditions. Thus, fishing is the major activity performed by members of the public within the site boundary. Based on the nature of the fishermen and undocumented observations, it is conservatively assumed that the maximum exposed individual spends an average of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months per week fishing from the shoreline at a location between the Energy Center and the Unit 1 facility. This estimate is considered conservative but not necessarily excessive and accounts for occasions where individuals may fish more on weekends or on a few days in March of the year.

The pathways considered for the evaluation include the inhalation pathway, the ground dose pathway with the resultant whole body and skin dose and the direct radiation dose pathway with the associated whole body dose. The direct radiation dose pathway, in actuality, includes several pathways. These include: the direct radiation gamma dose to an individual from an overhead plume, a gamma submersion plume dose (as applicable), possible direct radiation dose from the facility and a ground plane dose (deposition). Because the location is in close proximity to the site, any beta plume submersion dose is felt to be insignificant.

Other pathways, such as the ingestion pathway, are not applicable since these doses are included under calculations for doses to members of the public outside of the site boundary.

In addition, pathways associated with water related recreational activities, other than fishing, are not applicable here. These include swimming, boating and wading which are prohibited at the facility.

Unit 1 ODCM Revision 24 1 30 July 2003

3A Doses to Members of the Public Within the Site Boundary (Cont'd)

The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium and particulates) in the effluent for the appropriate time period. The radionuclide concentrations are then multiplied by the appropriate X/Q value, inhalation dose factor, air intake rate, and the fractional portion of the year in question. Thus, the inhalation pathway is evaluated using the following equation adapted from Regulatory Guide 1.109. Table 3-23 presents the reference for the parameters used in the following equation.

NOTE: The following equation is adapted from equations C-3 and C4 of Regulatory Guide 1.109. Since many of the factors are in units of pCi/n 3 , m3 /sec., etc.,

and since the radionuclide decay expressions have been deleted because of the short distance to the receptor location, the equation presented here is not identical to the Regulatory Guide equations.

Dja = E i[(Q)F (XtQ)(DFA)-ij(BR)at1 Where:

Dja= The maximum dose from all nuclides to the organ j and age group (a) in mremlyr.

Ci = The average concentration in the stack release of nuclide i for the period in pCi/m 3 .

F = Unit 1 average stack flowrate in m3 /sec.

X/Q = The plume dispersion parameter for a location approximately 0.50 miles west of NMP-1; the plume dispersion parameter is 8.9E-06 sec/in 3 (stack) and was obtained from the C.T. Main five year average annual X/Q tables. The stack (elevated) X/Q is conservative when based on 0.50 miles because of the close proximity of the stack and the receptor location.

(DFA). = The dose factor for nuclide i, organ j, and age group a in mrem per pCi (Reg. Guide 1.109, Table E-7).

(BR)a = Annual air intake for individuals in age group a in m3 per year (obtained from Table E-5 of Regulatory Guide 1.109).

t = Fractional portion of the year for which radionuclide i was detected and for which a dose is to be calculated (in years).

Unit 1 ODCM Revision 24 1 31 July 2003

3.4 Doses to Members of the Public Within the Site Boundary (Cont'd)

The ground dose pathway (deposition) will be evaluated by obtaining at least one soil or shoreline sediment sample in the area where fishing occurs. The dose will then be calculated using the sample results, the time period in question, and the methodology based on Regulatory Guide 1.109 as presented in Section II-3.1. The resultant dose may be adjusted for a background dose by subtracting the applicable off-site control soil or shoreline sediment sample radionuclide activities. In the event it is noted that fishing is not performed from the shoreline, but is instead performed in the water (i.e., the use of waders), then the ground dose pathway (deposition) may not be evaluated.

The direct radiation gamma dose pathway includes any gamma doses from an overhead plume, potential submersion in the plume, possible direct radiation from the facility and ground plane dose (deposition). This general pathway will be evaluated by average environmental TLD readings. At least two environmental TLDs will be utilized at one location in the approximate area where fishing occurs. The TLDs will be placed in the field on approximately the beginning of a calendar quarter and removed on approximately the end of the calendar quarter. For the purposes of this evaluation, TLD data from quarters 2, 3, and 4 will be utilized.

The average TLD readings will be adjusted by the average control TLD readings. This is accomplished by subtracting the average quarterly control TLD value from the average fishing location TLD value. The applicable quarterly control TLD values will be utilized after adjusting for the appropriate time period (as applicable). In the event of loss or theft of the TLDs, results from a TLD or TLDs in a nearby area may be utilized.

Unit 1 ODCM Revision 24 1 32 July 2003

4.0 ENVIRONMENTAL MONITORING PROGRAM 4.1 Sampling Stations The current sampling locations are specified in Table 5-1 and Figures 5.1-1, 5.1-2. The meteorological tower is shown in Figure 5.1-1. The location is shown as TLD location 17.

The Radiological Environmental Monitoring Program is a joint effort between the owners and operators of the Nine Mile Point Unit 1 and the James A. FitzPatrick Nuclear Power Plant.

Sampling locations are chosen on the basis of historical average dispersion or deposition parameters from both units. The environmental sampling location coordinates shown on Table 5-1 are based on the NMP-2 reactor centerline.

The average dispersion and deposition parameters have been calculated for a 5 year period, 1978 through 1982. These average dispersion or deposition parameters for the site are used to compare results of the annual land use census.

If it is determined that sample locations required by Part I are unavailable or new locations are identified that yield a significantly higher (e.g. 50%) calculated D/Q value, actions will be taken as required by Controls DLCO 3.6.20 and DLCO 3.6.22, and the Radiological Environmental Monitoring program updated accordingly.

4.2 Interlaboratory Comparison Program Analyses shall be performed on samples containing known quantities of radioactive materials that are supplied as part of a Commission approved or sponsored Interlaboratory Comparison Program, such as the EPA Crosscheck Program. Participation shall be only for those media, e.g., air, milk, water, etc., that are included in the Nine Mile Point Environmental Monitoring Program and for which crosscheck samples are available. An attempt will be made to obtain a QC sample to program sample ratio of 5% or better. The site identification symbol or the actual Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission staff may evaluate the results.

Specific sample media for which EPA Cross Check Program samples are available include the following:

- gross beta in air particulate filters

- garnma emitters in air particulate filters

- gamma emitters in milk

- gamma emitters in water

- tritium in water 131 in water Unit 1 ODCM Revision 24 1 33 July 2003

4.3 Capabilities for Thermoluninescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements required by Table D 4.6.20-1, footnote b of Part I are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use. In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabilities may be determined by the vendor who supplies the TLDs. Required detection capabilities are as follows:

4.3.1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting from an exposure rate of 10 mR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less than 7.5%. A total of at least 5 TLDs shall be evaluated.

4.3.2 Reproducibility shall be determined by giving TLDs repeated exposures equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The average of the relative standard deviations of the responses shall be less than 3.0%. A total of at least 4 TLDs shall be evaluated.

4.3.3 Dependence of exposure interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constant. This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures. For these tests, the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85. At least 6 TLDs shall be evaluated.

4.3A Energy dependence shall be evaluated by the response of TLDs to photons for several energies between approximately 30 keV and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25% for photons with energies greater than 80 keV and shall not be enhanced by more than a factor of two for photons with energies less than 80 keV. A total of at least 8 TLDs shall be evaluated.

4.3.5 The directional dependence of the TLD response shall be determined by comparing the response of the TLD exposed in the routine orientation with respect to the calibration source with the response obtained for different orientations. To accomplish this, the TLD shall be rotated through at least two perpendicular planes. The response averaged over all directions shall not differ from the response obtained in the standard calibration position by more than 10%. A total of at least 4 TLDs shall be evaluated.

4.3.6 Light dependence shall be determined by placing TLDs in the field for a period equal to the field cycle under the four conditions found in ANSI N545, section 4.3.6. The results obtained for the unwrapped TLDs shall not differ from those obtained for the TLDs wrapped in aluminum foil by more than 10%. A total of at least 4TLDs shall be evaluated for each of the four conditions.

Unit 1 ODCM Revision 24 1 34 July 2003

4.3.7 Moisture dependence shall be determined by placing TLDs (that is, the phosphors packaged for field use) for a period equal to the field cycle in an area where the exposure rate is known to be constant. The TLDs shall be exposed under two conditions: (1) packaged in a thin, sealed plastic bag, and (2) packaged in a thin, sealed plastic bag with sufficient water to yield observable moisture throughout the field cycle. The TLD or phosphor, as appropriate, shall be dried before readout. The response of the TLD exposed in the plastic bag containing water shall not differ from that exposed in the regular plastic bag by more than 10%. A total of at least 4 TLDs shall be evaluated for each condition.

4.3.8 Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is less than 10 uR/hr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3). The average exposure inferred from the responses of the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least 3 TLDs shall be evaluated.

Unit 1 ODCM Revision 24 II 35 July 2003

TABLE 1-1 Average Energy Per Disintegration ISOTOPiE &mev/disd(4)

Ar-41 1.294 (3) 0.464 (3)

Kr-83m 0.00248 (1) 0.0371 (1)

Kr-85 0.0022 (1) 0.250 (1)

Kr-85m 0.159 (1) 0.253 (1)

Kr-87 0.793 (1) 1.32 (1)

Kr-88 1.95 (1) 0.377 (1)

Kr-89 2.22 (2) 1.37 (2)

Kr-90 2.10 (2) 1.01 (2)

Xe-131M 0.0201 (1) 0.143 (1)

Xe-133 0.0454 (1) 0.135 (1)

Xe-133m 0.042 (1) 0.19 (1)

Xe-135 0.247 (1) 0.317 (1)

Xe-135m 0.432 (1) 0.095 (1)

Xe-137 0.194 (1) 1.64 (1)

Xe-138 1.18 (1) 0.611 (1)

(1) ORNL-4923, Radioactive Atoms - Supplement I, M.S. Martin, November 1973.

(2) NEDO-12037, "Summary of Gamma and Beta Emitters and Intensity Data"; M.E. Meek, R.S.

Gilbert, January 1970. (The average energy was computed from the maximum energy using the ICRP II equation, not the 1/3 value assumption used in this reference).

(3) NCRP Report No. 58, "A Handbook of Radioactivity Measurements Procedures"; 1978 (4) The average energy includes conversion electrons.

Unit 1 ODCM Revision 24 II 36 July 2003

TABLE 2-1 A, VALUES. LIQUID*

RADWASTE TANK INFANT mrem - nl hr - gtCi NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 2.90E-1 2.90E-1 2.90E-1 2.90E-1 2.90E-1 2.90E-1 Cr51 1.29E-2 8.39E-3 1.83E-3 1.63E-2 3.75E-1 Cu 64 1.13E-1 5.23E-2 - 1.91E1 2.32 Mn 54 1.87E+1 4.23 4.14 6.86 FE 55 1.31E+1 8.44 2.26 -- 4.13 1.07 Fe 59 2.84E+1 4.96E+1 1.96E+1 -- 1.47E+1 2.37E+1 Co 58 3.34 8.34 8.33 Co 60 1.02E+1 2.40E+1 -- 2.42E+1 Zn 65 1.72E+1 5.91E+1 2.73E+1 -- 2.87E+1 5.OOE+1 Sr 89 2.32E+3 6.66E+1 - 4.77E+1 Sr 90 1.74E+4 4.43E+3 - 2.17E+2 Zr 95 1.91E-1 4.66E-2 3.30E-2 - 5.02E-2 2.32E+1 Mn 56 -- 2.40E-4 4.15E-5 - 2.07E-4 2.18E-2 Mo 99 -- 2.34E+1 4.57 - 3.50E+1 7.71 Na 24 2.37 2.37 2.37 2.37 2.37 2.37 2.37 1131 3.03E+1 3.54E+1 1.57E+1 1.17E+4 4.17E+1 1.28 I 133 4.22 6.15 1.80 1.12E+3 7.23 1.04 Ni 65 1.33E-3 1.51E-4 6.85E-5 1.15E-2 I 132 1.58E-4 3.21E-4 1.14E-4 1.50E-2 3.58E-4 2.60E-4 Cs 134 3.54E+2 6.60E+2 6.67E+1 1.70E+2 6.97E+1 1.79 Cs 136 4.05E+1 1.19E+2 4.45E+1 4.75E+1 9.71E+1 1.81 Cs 137 4.91E+2 5.75E+2 4.07E+1 1.54E+2 6.24E+1 1.80 Ba 140 1.50E+2 1.50E-1 7.74 3.57E-2 9.23E-2 3.69E+1 Ce 141 7.21E-2 4.40E-2 5.17E-3 1.36E-2 2.27E+1 Nb 95 3.85E-2 1.59E-2 9.18E-3 1.14E-2 1.34E+1 La 140 1.18E-2 4.67E-3 1.20E-3 5.48E+1 Ce 144 2.79 1.14 1.57E-1 4.62E-1 1.60E+2

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

Unit 1 ODCM Revision 24 I 37 July 2003

TABLE 2-2 Ah. VALUES - LIQUID*

RADWASTE TANK CHILD mrem - m!

hr 1 Ci NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 4.39E-1 4.39E-1 4.39E-1 4.39E-1 4.39E-1 4.39E-1 Cr 51 2.13E-2 2.13E-2 1.40 7.86E-1 2.30E-1 1A2 7.3 1E+1 Cu 64 2.5 1E-6 2.70 1.63 2.51E-6 6.52 2.51E-6 1.27E+2 Mn 54 6.92 3.38E+3 9.06E+2 6.92 9.53E+2 6.92 2.84E+3 Fe 55 9.21E+2 4.88E+2 1.51E+2 2.76E+2 9.05E+1 Fe 59 1.30E+3 2.11E+3 1.05E+3 1.34 1.34 6.12E+2 2.19E+3 Co 58 1.89 7.46E+1 2.24E+2 1.89 1.89 1.89 4.26E+2 Co 60 1.12E+2 3.28E+2 7.48E+2 1.12E+2 1. 12E+2 1.12E+2 1.31E+3 Zn 65 2.15E+4 5.73E+4 3.56E+4 3.85 3.61E+4 3.85 1.01E+4 Sr 89 3.26E+4 1.10E-4 9.32E+2 1.10E-4 1.10E-4 1.10E4 1.26E+3 Sr 90 4.26E+5 1.08E+5 5.74E+3 Zr 95 1.70 1.33 1.32 1.23 1.38 1.23 1.08E+2 Mn 56 - 1.65E-1 3.73E-2 2.OOE-1 2.39E+1 Mo 99 5.35E-3 9.57E+1 2.37E+1 5.35E-3 2.04E+2 5.35E-3 7.91E+1 Na 24 1.52E+2 1.52E+2 1.52E+2 1.52E+2 1.52E+2 1.52E+2 1.52E+2 1131 2.09E+2 2.10E+2 1.19E+2 6.94E+4 3.45E+2 5.60E-2 1.87E+1 I 133 3.39E+1 4.19E+1 1.59E+1 7.78E+3 6.98E+1 1.38E-4 1.69E+1 Ni 65 2.67E-1 2.51E-2 1A7E-2 3.08 1132 6.13E-3 1.13E-2 5.18E-3 5.22E-1 1.72E-2 1.32E-2 Cs 134 3.68E+5 6.04E+5 1.27E+5 3.54E+1 1.87E+5 6.72E+4 3.29E+3 Cs 136 3.52E+4 9.67E+4 6.26E+4 6.21E-1 5.15E+4 7.68E+3 3.40E3+3 Cs 137 5.15E+5 4.93E+5 7.28E+4 5.37E+1 1.61E+5 5.78E+4 3.14E+3 Ba 140 3.61E+2 3.96E-1 2.11E+1 7.96E-2 1.82E-1 2.68E-1 1.83E+2 Ce 141 1.50E-1 1.07E-1 6.99E-2 6.34E-2 8.24E-2 6.34E-2 5.40E+1 Nb 95 5.21E+2 2.03E+2 1.45E+2 6.39E-1 1.91E+2 6.39E-1 3.75E+5 La 140 1.50E-1 5.93E-2 2.68E-2 1.03E-2 1.03E-2 1.03E-2 1.36E+3 Ce 144 5.00 1.81 6.06E-1 3.58E-1 1.16 3.58E-1 3.801E+2

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

Unit 1 ODCM Revision 24 1 38 July 2003

TABLE 2-3 Awt VALUES - LIQUID*

RADWASTE TANK TEEN mnrem - ml hr PCi NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 3.28E-1 3.28E-1 q IRPA 14IMP-1 3.28E-1 3.28E-1 Cr 51 1.02E-1 1.02E-1 1.39 8.16E-1 3.84E-1 1.94 2.16E+2 Cu 64 1.20E-5 2.89 1.36 1.20E-5 7.32 1.20E-5 2.24E+2 Mn 54 3.31E+1 4.34E+3 8.87E+2 3.31E+1 1.32E+3 3.31E+1 8.86E+3 Fe 55 6.94E+2 4.92E+2 1.1SE+2 3.12E+2 2.13E+2 Fe 59 1.07E+3 2.49E+3 9.64E+2 6.41 6.41 7.89E+2 5.87E+3 Co58 9.03 9.82E+1 2.15E+2 9.03 9.03 9.03 1.24E+3 Co 60 5.36E+2 7.96E+2 1.12E+3 5.36E+2 5.36E+2 5.36E+2 3.93E+3 Zn 65 2.10E+4 7.28E+4 3.40E+4 1.84E+1 4.66E+4 1.84E+1 3.08E+4 Sr 89 2.44E+4 5.24E14 6.98E+2 5.24E-4 5.24E-4 5.24E-4 2.90E+3 A ICZ,"..&

Sr 90 1.1SE+5 1.3 1E+4 Zr 95 6.20 6.00 5.97 5.90 6.04 5.90 2.28E+2 Mn 56 - 1.81E-1 3.22E-2 2.29E-1 1.19E+1 Mo 99 2.56E-2 9.22E+1 1.76E+1 2.56E-2 2.1 1E+2 2.56E-2 1.65E+2 Na 24 1.39E+2 1.39E+2 1.39E+2 1.39E+2 1.39E+2 1.39E+2 1.39E+2 1131 1.55E+2 2.17E+2 1.16E+2 6.31E+4 3.73E+2 2.68E-1 4.30E+1 1133 2.53E+1 4.29E+1 1.31E+1 5.99E+3 7.52E+1 6.60E-4 3.25E+1 Ni 65 2.08E-1 2.66E-2 1.21E-2 1.44 1132 4.90E-2 1.28E-2 4.60E-3 4.32E-1 2.02E-2 5.59E-3 Cs 134 3.05E+5 7.18E+5 3.33E+5 1.69E+2 2.28E+5 8.73E+4 9.10E+3 Cs 136 2.98E+4 1.17E+5 7.88E+4 2.97 6.38E+4 1.01E+4 9.44E+3 Cs 137 4.09E+5 5.44E+5 1.90E+5 2.57E+2 1.85E+5 7.21E+4 7.99E+3 Ba 140 2.35E+2 4.10E-1 1.55E+1 3.81E-1 4.79E-1 5.75E-1 3.63E+2 Ce 141 3.46E-1 3.32E-1 3.07E-1 3.04E-1 3.17E-1 3.04E-1 8.16E+1 Nb 95 4.44E+2 2.48E+2 1.18E+2 3.06 2.40E+2 3.06 1.05E+6 La 140 1.57E-1 1.02E-1 6.35E-2 4.94E-2 4.94E-2 4.94E-2 3.05E+3 Ce 144 3.99 2.65 1.83 1.71 2.27 1.71 5.74E+2

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

Unit 1 ODCM Revision 24 1 39 July 2003

TABLE 2-4 Akt VALUES - LIQUID*

RADWASTE TANK ADULT mrem - ml hr - uCl NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 -- 4.45E-1 4.45E-1 4.45E-1 4.45E-1 4.45E-1 4.45E-1 Cr51 1.82E-2 1.82E-2 1.27 7.64E-1 2.93E-1 1.67 3.14E+2 Cu 64 - 2.75 1.29 -- 6.94 -- 2.35E+2 Mn 54 5.94 4.38E+3 8.41E+2 5.94 1.31E+3 5.94 1.34E+4 Fe 55 6.64E+2 4.58E+2 1.07E+2 -- -- 2.56E+2 2.63E+2 Fe 59 1.03E+3 2.43E+3 9.31E+2 1.15 1.15 6.79E+2 8.09E+3 Co 58 1.62 9.15E+1 2.03E+2 1.62 1.62 1.62 1.82E+3 Co 60 9.60E+1 2.57E+2 6.71E+2 9.60E+1 9.60E+1 9.60E+1 4.99E+3 Zn 65 2.3 1E+4 7.36E+4 3.32E+4 3.30 4.92E+4 3.30 4.63E+4 Sr 89 2.25E+4 9.39E-5 6.45E1+2 9.39E-5 9.39E-5 9.39E-5 3.60E+3 Sr 90 5.60E+5 -- 1.37E+5 -- - -- 1.62E+4 Zr95 1.36 1.15 1.12 1.06 1.21 1.06 3.06E+2 Mn 56 - 1.73E-1 3.07E-2 - 2.20E-1 -- 5.52 Mo 99 4.58E-3 8.70E+1 1.66E+1 4.58E-3 1.97E+2 4.58E-3 2.02E+2 Na 24 1.35E+2 1.35E+2 1.35E+2 1.35E+2 1.35E+2 1.35E+2 1.35E+2 1131 1.45E+2 2.07E+2 1.19E+2 6.79E+4 3.55E+2 4.80E-2 5.47E+1 1133 2.35E+1 4.09E+1 1.25E+1 6.02E+3 7.14E+1 1.18E-4 3.68E+1 Ni 65 1.93E-1 2.51E-2 1.14E-2 - -- - 6.36E-1 1132 4.68E-3 1.25E-2 4.38E-3 4.38E-1 2.OOE-2 - 2.35E-3 Cs 134 2.98E+5 7.08E+5 5.79E+5 3.03E+1 2.29E+5 7.61E+4 1.24E+4 Cs 136 2.96E+4 1.17E1+5 8.42E+4 5.32E-1 6.51E+4 8.93E+3 1.33E+4 Cs 137 3.82E+5 5.22E+5 3.42E+5 4.60E+1 1.77E+5 5.90E+4 1.02E+4 Ba 140 2.24E+2 3.49E-1 1.47E+1 6.83E-2 1.64E-1 2.29E-1 4.61E+2 Ce 141 9.53E-2 8.20E-2 5.75E-2 5.44E-2 6.72E-2 5.44E-2 1.06E+2 Nb 95 4.39E+2 2.44E+2 1.32E+2 5.47E-1 2.41E+2 5.47E-1 1.48E+6 La 140 1.11E-1 6.03E-2 2.24E-2 8.84E-3 8.84E-3 8.84E-3 3.78E+3 Ce 144 2.48 1.22 4.24E-1 3.07E-1 8.47E-1 3.07E-1 7.37E+2

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

Unit 1 ODCM Revision 24 II 40 July 2003

TABLE 2-5 Awt VALUES. LIQUID*

EMERGENCY CONDENSER VENT INFANT mremrn ml hr.ICi NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 - 7.43E-4 7.43E-4 7.43E-4 7.43E-4 7.43E-4 7.43E-4 Cr51 -- - 3.30E-5 2.15E-5 4.70E-6 4.18E-5 9.61E-4 Cu 64 -- 2.89E-4 1.34E-4 - 4.89E-4 - 5.94E-3 Mn 54 - 4.79E-2 1.08E-2 -- 1.06E-2 - 1.76E-2 Fe 55 3.35E-2 2.16E-2 5.78E-3 -- -- 1.06E-2 2.75E-3 Fe 59 7.29E-2 1.27E-1 5.02E-2 - -- 3.76E-2 6.08E-2 Co 58 -- 8.58E-3 2.14E-2 - -- - 2.14E-2 Co 60 -- 2.60E-2 6.15E-2 - - - 6.19E-2 Zn 65 4.42E-2 1.52E-1 6.99E-2 -- 7.35E-2 -- 1.28E-1 Sr 89 5.95 -- 1.71E-1 -- - -- 1.22E-1 Sr 90 4A6E+1 -- 1.14E+1 -- -- -- 5.57E-1 Zr 95 4.90E-4 1.19E-4 8.47E-5 - 1.29E-4 - 5.95E-2 Mn 56 -- 6.17E-7 1.06E-7 - 5.30E-7 - 5.60E-5 Mo 99 - 6.OOE-2 1.17E-2 -- 8.97E-2 - 1.98E-2 Na 24 6.07E-3 6.07E-3 6.07E-3 6.07E-3 6.07E-3 6.07E-3 6.07E-3 I 131 7.77E-2 9.16E-2 4.03E-2 3.01E+1 1.07E-1 - 3.27E-3 1133 1.08E-2 1.58E-2 4.62E-3 2.87 1.85E-2 -- 2.67E-3 Ni 65 3.41E-6 3.86E-7 1.76E-7 - - - 2.94E-5 1132 4.05E-7 8.22E-7 2.93E-7 3.85E-5 9.17E-7 - 6.66E-7 Cs 134 9.08E-1 1.69 1.71E-1 -- 4.36E-1 1.79E-1 4.60E-3 Cs 136 1.04E-1 3.06E-1 1.14E-1 -- 1.22E-1 2.49E-2 4.64E-3 Cs 137 1.26 1.47 1.04E-1 -- 3.95E-1 1.60E-1 4.61E-3 Ba 140 3.85E-1 3.85E-4 1.99E-2 -- 9.15E-5 2.37E-4 9A7E-2 Ce 141 1.85E-4 1.13E-4 1.33E-5 - 3.48E-5 - 5.82E-2 Nb 95 9.88E-5 4.07E-5 2.35E-5 - 2.92E-5 - 3.43E-2 La 140 3.03E-5 1.20E-5 3.08E-6 - -- - 1.41E-1 Ce 144 7.16E-3 2.93E-3 4.02E-4 - 1.19E-3 - 4.11E-1

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

Unit 1 ODCM Revision 24 1 41 July 2003

TABLE 2-6 A1.wVALUES - LIQUID*

EMERGENCY CONDENSER VENT CHILD mnrem - ml hr - MCi NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 I A4E.1 A'P 1 SAAi SI I I SAAP- 1.44E-1 1.44E-1 Cr51 3.78E-5 3.78E-5 1.37 7.581E-1 2.07E-1 1.38 7.24E+1 Cu 64 2.63 1.59 6.35 -- 1.23E+2 Mn 54 1.23E-2 3.36E+3 8.95E+2 1.23E-2 9.42E+2 1.23E-2 2.82E+3 Fe 55 9.04E+2 4.79E+2 1.49E+2 2.71E+2 8.88E+1 Fe 59 1.28E+3 2.07E+3 1.03E+3 2.38E-3 2.38E-3 6.00E+2 2.15E1+3 Co 58 3.36E-3 7.01E+1 2.15E+2 3.36E-3 3.36E-3 3.36E-3 4.09E+2 Co 60 1.99E-1 2.08E+2 6.14E+2 1.99E-1 1.99E-1 1.99E-1 1.15E+3 Zn 65 2.15E+4 5.73E+4 3.56E+4 6.84E-3 3.61E+4 6.84E-3 1.01E+4 Sr 89 3.07E+4 8.78E+2 - 1.19E+3 Sr 90 4.01E+5 1.02E+5 -5.40E+3 Zr 95 3.01E-1 6.78E-2 6.06E-2 2.19E-3 9.61E-2 2.19E-3 6.84E+1 Mn 56 - 1.65E-1 3.73E-2 2.OOE-1 -- 2.39E+1 Mo 99 - 8.16E+1 2.02E1+1 1.74E+2 -- 6.75E+1 Na 24 1.50E+2 1.50E+2 1.50E+2 1.50E+2 1.50E+2 1.50E+2 1.50E+2 1131 1.86E+2 1.87E+2 1.06E+2 6.19E+4 3.08E+2 1.67E+1 I 133 3.08E+1 3.81E+1 1.44E+1 7.07E+3 6.35E+1 1.53E+1 Ni 65 2.66E-1 2.50E-2 1.46E-2 3.07 I 132 6.01E-3 1.10E-2 5.08E-3 5.12E-1 1.69E-2 1.30E-2 Cs 134 3.68E+5 6.04E+5 1.27E+5 6.29E-2 1.87E+5 6.71E+4 3.25E+3 Cs 136 3.51E+4 9.66E+4 6.25E+4 1.101E-3 5.14E+4 7.67E+3 3.40E+3 Cs 137 5.14E+5 4.92E+5 7.27E+4 9.55E-2 1.60E1+5 5.77E+4 3.08E+3 Ba 140 2.48E+2 2.17E-1 1.45E+1 1A2E-4 7.09E-2 1.30E-1 1.26E+2 Ce 141 3.08E-2 1.54E-2 2.39E-3 1.131E-4 6.83E-3 1.13E-4 1.91E+1 Nb 95 5.21E+2 2.03E+2 1.45E+2 1.14E-3 1.90E+2 1.14E-3 3.75E+5 La 140 1.31E-1 4.59E-2 1.55E-2 1.83E-5 1.83E-5 1.83E-5 1.28E+3 Ce 144 1.64 5.15E-1 8.81E-2 6.36E-4 2.85E-1 6.36E-4 1.34E+2

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

Unit 1 ODCM Revision 24 1 42 July 2003

TABLE 2-7 Aht VALUES - LIQUID*

EMERGENCY CONDENSER VENT TEEN mrem - ml hr - gCi NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 1.74E-1 1.74E-1 1.74E-1 1.74E-1 1.74E-1 1.74E-1 Cr 51 1.81E-4 1.81E-4 1.28 7.12E-1 2.81E-1 1.83 2.15E+2 Cu 64 2.86 1.35 7.24 - 2.22E+2 Mn 54 5.89E-2 4.29E+3 8.52E+2 5.89E-2 1.28E+3 5.89E-2 8.81E+3 Fe 55 6.89E+2 4.88E+2 1.14E+2 3.10E+2 2.11E+2 Fe 59 1.05E+3 2.46E+3 9.50E+2 1.14E-2 1.14E-2 7.76E+2 5.82E+3 Co 58 1.61E-2 8.78E+1 2.02E+2 1.61E-2 1.61E-2 1.61E-2 1.21E+3 Co 60 9.53E-1 2.57E+2 5.78E+2 9.53E-1 9.53E-1 9.53E-1 3.34E+3 Zn 65 2.10E+4 7.28E+4 3.39E+4 3.28E-2 4.66E+4 3.28E-2 3.08E+4 Sr 89 2.38E+4 6.8 1E+2 -- 2.83E+3 A rAr . C Sr 90 1.12E+5 -- 1.27E+4 Zr 95 2.56E-1 8.80E-2 6.38E-2 1.05E-2 1.24E-1 1.05E-2 1.79E+2 Mn 56 - 1.81E-1 3.22E-2 2.29E-1 -- 1.19E+1 Mo 99 - 8.57E+1 1.63E+1 1.96E+2 -- 1.54E+2 Na 24 1.38E+2 1.38E+2 1.38E+2 1.38E+2 1.38E+2 1.38E+2 1.38E+2 1131 1.47E+2 2.06E+2 1.10E+2 6.00E+4 3.54E+2 4.77E-4 4.07E+1 1133 2.42E+1 4.11E+1 1.25E+1 5.74E+3 7.21E+1 3.11E+1 Ni 65 2.08E-1 2.66E-2 1.21E-2 - 1.44 1132 4.86E-3 1.27E-2 4.56E-3 4.29E-1 2.00E-2 5.54E-3 Cs 134 3.05E+5 7.18E+5 3.33E+5 3.01E-1 2.28E+5 8.71E+4 8.93E+3 Cs 136 2.98E+4 1.17E+5 7.87E+4 5.28E-3 6.38E+4 1.O1E-i4 9.43E+3 Cs 137 4.09E+5 5.44E+5 1.89E+5 4.57E-1 1.85E+5 7.19E+4 7.73E+3 Ba 140 1.96E+2 2.47E-2 1.27E+1 6.77E-4 8.23E-2 1.62E-1 3.03E+2 Ce 141 2.43E-2 1.64E-2 2.36E-3 5.40E-4 8.02E-3 5.40E-4 4.54E+1 Nb 95 4.41E+2 2.45E+2 1.15E+2 5.43E-3 2.37E+2 5.431E-3 1.05E+6 La 140 1.05E-1 5.17E-2 1.38E-2 8.78E-5 8.78E-5 8.78E-5 2.96E+3 Ce 144 1.27 5.28E-1 7.12E-2 3.04E-3 3.17E-1 3.04E3-3 3.19E+2

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

Unit I ODCM Revision 24 1143 July 2003

TABLE 2-8 Aw VALUES - LIQVID*

EMERGENCY CONDENSER VENT ADULT inrem - ml hr - ACi NUCLIDE BONE LIVER T BODY THYROID KIDNEY LUNG GI-TRACT H3 -- 2.27E-1 2.27E-1 2.27E-1 2.27E-1 2.27E-1 2.27E-1 Cr 51 3.24E-5 3.24E-5 1.24 7.43E-1 2.74E-1 1.65 3.12E+2 Cu 64 - 2.72 1.28 -- 6.86 -- 2.32E+2 Mn 54 1.06E-2 4.37E+3 8.33E+2 1.06E-2 1.30E+3 1.06E-2 1.34E+4 Fe 55 6.58E+2 4.55E+2 1.06E+2 - - 2.54E+2 2.61E+2 Fe 59 1.02E+3 2.41E+3 9.22E+2 2.04E-3 2.04E-3 6.72E+2 8.02E+3 Co 58 2.88E-3 8.83E+1 1.98E+2 2.88E-3 2.88E-3 2.88E-3 1.79E+3 Co 60 1.71E-1 2.56E+2 5.65E+2 1.71E-1 1.71E-1 1.71E-1 4.81E+3 Zn 65 2.3 1E+4 7.36E+4 3.32E+4 5.87E-3 4.92E+4 5.87E-3 4.63E+4 Sr 89 2.18E+4 -- 6.27E+2 -- - - 3.50E+3 Sr 90 5.44E+5 -- 1.34E+5 - - 1.57E+4 Zr 95 2.40E-1 7.81E-2 5.35E-2 1.88E-3 1.22E-1 1.88E-3 2.42E+2 Mn 56 - 1.73E-1 3.07E-2 -- 2.20E-1 -- 5.52 Mo 99 - 8.04E+1 1.53E+1 -- 1.82E+2 -- 1.86E+2 Na 24 1.34E+2 1.34E+2 1.34E+2 1.34E+2 1.34E+2 1.34E+2 1.34E+2 I 131 1.37E+2 1.96E+2 1.12E+2 6.43E+4 3.36E+2 -- 5.17E+1 1133 2.25E+1 3.91E+1 1.19E+1 5.75E+3 6.82E+1 -- 3.51E+1 Ni 65 1.93E-1 2.50E-2 1.14E-2 - - - 6.36E-1 1132 4.64E-3 1.24E-2 4.34E-3 4.34E-1 1.98E-2 - 2.33E-3 Cs 134 2.98E+5 7.08E+5 5.79E+5 5.39E-2 2.29E+5 7.61E+4 1.24E+4 Cs 136 2.96E+4 1.17E+5 8.42E+4 9.46E-4 6.51E+4 8.92E+3 1.33E+4 Cs 137 3.82E+5 5.22E+5 3.42E+5 8.19E-2 1.77E+5 5.89E+4 1.01E+4 Ba 140 1.84E+2 2.32E-1 1.21E+1 1.21E-4 7.88E-2 1.33E-1 3.79E+2 Ce 141 2.21E-2 1.50E-2 1.78E-3 9.67E-5 7.OOE-3 9.67E-5 5.68E1+1 Nb 95 4.38E+2 2.44E+2 1.31E+2 9.73E-4 2.41E+2 9.73E-4 1.48E+6 La 140 9.90E-2 4.99E-2 1.32E-2 1.57E-5 1.57E-5 1.57E-5 3.66E+3 Ce 144 1.17 4.89E-1 6.33E-2 5.45E-4 2.90E-1 5.45E-4 3.95E+2

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

Unit 1 ODCM Revision 24 I 44 July 2003

TABLE 3-1 Critical Receptor Dispersion Paranmeters' For Ground Level and Elevated Releases ELEVATED ELEVATED GROUND' GROUND' LOCATION DIR MILES X1O (sec/m) D/Om 2 X/O(sec/mn Residences E (98c) 1.4 1.8 E-07b 5.2 EW9b 4.02 E-07 8.58 E-09 Dairy Cows' SE (130) 2.6 2.2 E-08c 7.0 E-10c 6.00 E-08 1.64 E-09 Milk Goats' SE (1300) 2.6 2.2 E-08c 7.0 E-10c 6.00 E-08 1.64 E-09 Meat Animals ESE (1150) 1.8 5.1 E-08c 1.7 E49c 1.16 E-07 3.54 E-09 Gardens E (970) 1.8 1.0 E-07c 3.5 E-09 2.53 E-07 5.55 E-09 Site Boundary ENE (67Y) 0.4 2.4 EO6bd 4.4 EM8c d 6.63 E-06 6.35 E-08

a. These values will be used in dose calculations beginning in April 1986 but may be revised periodically to account for changes in locations of farms, gardens or critical residences.

b. Values based on 5 year annual meteorological data (C.T. Main, Rev. 2)

c. Values based on 5 year average grazing season meteorological data (C.T. Main Rev. 2)

d. Value are based on most restrictive X/Q land-based sector (ENE). (C.T. Main, Rev. 2)

e. Values are based on average annual meteorological data for the year 1985.

f. Conservative location based on past dairy cow and goat milk history.

Unit 1 ODCM Revision 24 1I45 July 2003

TABLE 3-2 Gamma Air and Whole Body Plume Shine Dose Factors*

For Noble Gases Gamma Whole Gamma Air Bi Body Vi mradhr mrem/vr Nuclide p1Ci/sec jiCi/sec Kr-85 2.23E-6 Kr-85m 1.75E-3 1.68E-3 Kr-87 1.02E-2 9.65E-3 Kr-88 2.23E-2 2.17E-2 Kr-89 2.50E-2 1.71E-2 Kr-83m 2.26E-6 Xe-133 2.91E-4 1.75E-4 Xe-133m 2.27E-4 1.87E-4 Xe-135 2.62E-3 2.50E-3 Xe-135m 5.20E-3 4.89E-3 Xe-137 2.30E-3 2.20E-3 Xe-138 1.54E-2 1.03E-2 Xe-131m 1.74E-5 1.47E-6 Ar-41 1.64E-2 1.57E-2

Calculated in accordance with Regulatory Guide 1.109. (See Appendix B.)

Unit 1 ODCM Revision 24 B146 July 2003

TABLE 3-3 IMMERSION DOSE FACTORS FOR NOBLE GASES*

Nuclide K, (Y-Body)** Lj(P-Skin)** M(dY.AIr)*** n,(o-Air)***

Kr 83m 7.56E-02 1.93E1 2.88E2 Kr 85m 1.17E3 1A6E3 1.23E3 1.97E3 Kr 85 1.61E1 1.34E3 1.72E1 1.95E3 Kr 87 5.92E3 9.73E3 6.17E3 1.03E4 Kr 88 1.47E4 2.37E3 1.52E4 2.93E3 Kr 89 1.66E4 1.01E4 1.73E4 1.06E4 Kr 90 1.56E4 7.29E3 1.63E4 7.83E3 Xe 131m 9.15E1 4.76E2 1.56E2 1.11E3 Xe 133m 2.51E2 9.94E2 3.27E2 1.48E3 Xe 133 2.94E2 3.06E2 3.53E2 1.05E3 Xe 135m 3.12E3 7.11E2 3.36E3 7.39E2 Xe 135 1.81E3 1.86E3 1.92E3 2A6E3 Xe 137 1.42E3 1.22E4 1.51E3 1.27E4 Xe 138 8.83E3 4.13E3 9.21E3 4.75E3 Ar 41 8.84E3 2.69E3 9.30E3 3.28E3

From, Table B-l.Regulatory Guide 1.109 Rev. 1

- mrem/yr per ILCi/m 3.

- - mrad/yr per tCi/m3 .

Unit 1 ODCM Revision 24 1147 July 2003

TABLE 3-4 DOSE AND DOSE RATE RI VALUES - INHALATION- INFANT' mrmiemi pCifm I

NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H3* 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 C 14* 2.65E4 5.31E3 5.31E3 5.31E3 5.31E3 5.3 1E3 5.31E3 Cr 51 8.95E1 5.75E1 1.32E1 1.28E4 3.57E2 Mn 54 2.53E4 4.98E3 4.98E3 1.00E6 7.06E3 Fe 55 1.97E4 1.17E4 3.33E3 8.69E4 1.09E3 Fe 59 1.36E4 2.35E4 9A8E3 1.02E6 2.48E4 Co58 1.22E3 1.82E3 7.77E5 1.llE4 Co 60 8.02E3 1.18E4 4.51E6 3.19E4 Zn 65 1.93E4 6.26E4 3.11E4 3.25E4 6.47E5 5.14E4 Sr 89 3.98E5 1.14E4 2.03E6 6.40E4 Sr 90 4.09E7 2.59E6 1.12E7 1.3 lE5 Zr 95 1.15E5 2.79E4 2.03E4 3.11E4 1.75E6 2.17E4 Nb 95 1.57E4 6.43E3 3.78E3 4.72E3 4.79E5 1.27E4 Mo 99 1.65E2 3.23E1 2.65E2 1.35E5 4.87E4 I-131 3.79E4 4.44E4 1.96E4 1.48E7 5.18E4 1.06E3 1133 1.32E4 1.92E4 5.60E3 3.56E6 2.24E4 2.16E3 Cs 134 3.96E5 7.03E5 7.45E4 1.90E5 7.97E4 1.33E3 Cs 137 5.49E5 6.12E5 4.55E4 1.72E5 7.13E4 1.33E3 Ba 140 5.60E4 5.60E1 2.90E3 1.34E1 1.60E6 3.84E4 La 140 5.05E2 2.00E2 5.15E1 1.68E5 8.48E4 Ce 141 2.77E4 1.67E4 1.99E3 5.25E3 5.17E5 2.16E4 Ce 144 3.19E6 1.21E6 1.76E5 5.38E5 9.84E6 1.48E5 Nd 147 7.94E3 8.13E3 5.00E2 3.15E3 3.22E5 3.12E4

rnrenmyr per tLCi/m 3 .

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

Unit 1 ODCM Revision 24 1148 July 2003

TABLE 3-5 DOSE AND DOSE RATE RI VALUES - INHALATION - CHILD mre/

uCilm NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H3* - 1 12F 1.12E3 1 12F3 1 1t12E 1.12E3 1.12E3 C 14* 3.59E4 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 Cr 51 1.54E2 8.55E1 2.43E1 1.70E4 1.08E3 Mn 54 4.29E4 9.51E3 1.00E4 1.58E6 2.29E4 Fe 55 4.74E4 2.52E4 7.77E3 1.11E5 2.87E3 Fe 59 2.07E4 3.34E4 1.67E4 1.27E6 7.07E4 Co 58 1.77E3 3.16E3 1.11E6 3.44E4 Co 60 1.31E4 2.26E4 7.07E6 9.62E4 Zn 65 4.26E4 1.13E5 7.03E4 7.14E4 9.95E5 1.63E4 Sr 89 5.99E5 1.72E4 2.16E6 1.67E5 Sr 90 1.01E8 6.44E6 1.48E7 3.43E5 Zr 95 1.90E5 4.18E4 3.70E4 5.96E4 2.23E6 6.11E4 Nb 95 2.35E4 9.18E3 6.55E3 8.62E3 6.14E5 3.70E4 Mo 99 1.72E2 4.26E1 3.92E2 1.35E5 1.27E5 1131 4.81E4 4.81E4 2.73E4 1.62E7 7.88E4 2.84E3 1133 1.66E4 2.03E4 7.70E3 3.85E6 3.38E34 5.48E3 Cs 134 6.51E5 1.01E6 2.25E5 3.30E5 1.21E5 3.85E3 Cs 137 9.07E5 8.25E5 1.28E5 2.821E5 1.04E5 3.62E3 Ba 140 7.40E4 6.48E1 4.33E3 2.1 1E1 1.74E6 1.02E5 La 140 6.44E2 2.25E2 7.55E1 1.83E5 2.26E5 Ce 141 3.92E4 1.95E4 2.90E3 8.55E3 5.44E5 5.66E4 Ce 144 6.77E6 2.12E6 3.61 ES 1.17E6 1.20E7 3.89E5 Nd 147 1.08E4 8.73E3 6.81E2 4.81E3 3.28E5 8.21E4

mrem/yr per jjCi/m3.

Unit 1 ODCM Revision 24 I 49 July 2003

TABLE 3-6 DOSE AND DOSE RATE R VALUES - INHALATION - TEEN mrem/vr ttCi/m NUCLIDE BONE LIVER T. BODI t' THYROID KIDNEY LUNG GI-LLI H3* 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 C 14* 2.60E4 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 Cr51 1.35E2 7.50E1 3.07E1 2.10E4 3.00E3 Mn 54 5.11E4 8.40E3 1.27E4 1.98E6 6.68E4 Fe 55 3.34E4 2.38E4 5.54E3 1.24E5 6.39E3 Fe 59 1.59E4 3.70E4 1.43E4 1.53E6 1.78ES Co 58 2.07E3 2.78E3 1.34E6 9.52E4 Co60 1.51E4 1.98E4 8.72E6 2.59E5 Zn 65 3.86E4 1.34E5 6.24E4 8.64E4 1.24E6 4.66E4 Sr 89 4.34E5 1.25E4 2A2E6 3.71E5 Sr 90 1.08E8 6.68E6 1.65E7 7.65E5 Zr 95 1.46E5 4.58E4 3.15E4 6.74E4 2.69E6 1.49E5 Nb 95 1.86E4 1.03E4 5.66E3 1.00E4 7.51ES 9.68E4 Mo 99 1.69E2 3.22E1 4.11E2 1.54E5 2.69E5 I131 3.54E4 4.91E4 2.64E4 1.46E7 8.40E4 6.49E3 1133 1.22E4 2.05E4 6.22E3 2.92E6 3.59E4 1.03E4 Cs 134 5.02E5 1.13E6 5.49E5 3.75E5 1A6E5 9.76E3 Cs 137 6.70E5 8.48E5 3.11ES 3.04E5 1.21E5 8.48E3 Ba 140 5.47E4 6.70E1 3.52E3 2.28E1 2.03E6 2.29E5 La 140 4.79E2 2.36E2 6.26E1 2.14E5 4.87E5 Ce 141 2.84E4 1.90E4 2.17E3 8.88E3 6.14E5 1.26E5 Ce 144 4.89E6 2.02E6 2.62E5 1.21E6 1.34E7 8.64E5 Nd 147 7.86E3 8.56E3 5.13E2 5.02E3 3.72E5 1.82E5

mrern/yr per gCi/m3.

Unit 1 ODCM Revision 24 II 50 July 2003

TABLE 3-7 DOSE AND DOSE RATE RI VALUES - INHALATION - ADULT mrenlvr ILCi/m NUCLIDE BONE LIVER T. BODI .r THYROID KIDNEY LUNG GI-LLI H3* 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 1.26E3 C 14* 1.82E4 3.41E3 3.41E3 3.41E3 3.41E3 3.41E3 3.41E3 Cr 51 1.00E2 5.95E1 2.28E1 1.44E4 3.32E3 Mn 54 3.96E4 6.30E3 9.84E3 1.40E6 7.74E4 Fe 55 2.6E4 1.70E4 3.94E3 7.21E4 6.03E3 Fe 59 1.18E4 2.78E4 1.06E4 1.02E6 1.88E5 Co 58 1.58E3 2.07E3 9.28E5 1.06E5 Co 60 1.15E4 1.48E4 5.97E6 2.85E5 Zn 65 3.24E4 1.03E5 4.66E4 6.90134 8.64ES 5.34E4 Sr 89 3.04E5 8.72E3 1.40E6 3.50E5 Sr 90 9.92E7 6.10E6 9.60E6 7.22E5 Zr 95 1.07E5 3.44E4 2.33E4 5.421E4 1.77E6 1.5OE5 Nb 95 1.41E4 7.82E3 4.21E3 7.74E3 5.05E5 1.04E5 Mo 99 1.21E2 2.30E1 2.911E2 9.12E4 2.48E5 I131 2.52E4 3.58E4 2.05E4 1.19E7 6.13E4 6.28E3 1133 8.64E3 1.48E4 4.52E3 2.15E6 2.581E4 8.88E3 Cs 134 3.73E5 8A8E5 7.28E5 2.87E35 9.76E4 1.04E4 Cs 137 4.78E5 6.21E5 4.28E5 2.221E5 7.52E4 8.40E3 Ba 140 3.90E4 4.90E1 2.57E3 1.67E1 1.27E6 2.18E5 La 140 3.44E2 1.74E2 4.58E1 1.36E5 4.58E5 Ce 141 1.99E4 1.35E4 1.53E3 6.261E3 3.62E5 1.20E5 Cc 144 3.43E6 1.43E6 1.84E5 8.A8E5 7.78E6 8.16E5 Nd 147 5.27E3 6.10E3 3.65E2 3.56E3 2.21E5 1.73E5

mrenfyr per ACi/m 3 .

Unit 1 ODCM Revision 24 1151 July 2003

TABLE 3-8 DOSE AND DOSE RATE Ri VALUES - GROUND PLANE ALL AGE GROUPS 2

-mrem/r ILCLSeC NUCLIDE TOTAL BODY SKIN H3 C 14 Cr51 4.65E6 5.5OE6 Mn 54 1.40E9 1.64E9 Fe 55 Fe 59 2.73E8 3.20E8 Co 58 3.80E8 4.45E8 Co 60 2.15E10 2.53E10 Zn 65 7.46E8 8.57E8 Sr 89 2.16E4 2.51E4 Sr 90 Zr 95 2.45E8 2.85E8 Nb 95 1.36E8 1.61E8 Mo 99 3.99E6 4.63E6 1131 1.72E7 2.09E7 1133 2.39E6 2.91E6 Cs 134 6.83E9 7.97E9 Cs 137 1.03E10 1.20E10 Ba 140 2.05E7 2.35E7 La 140 1.92E7 2.18E7 Ce 141 1.37E7 1.54E7 Ce 144 6.96E7 8.07E7 Nd 147 8.46E6 l.0E7 Unit 1 ODCM Revision 24 1152 July 2003

TABLE 3-9 DOSE AND DOSE RATE It VALUES - COW MILK - INFANT m2 -mremlvr tCUsec NUCLIDE BONE LIVER T. BODY THYROID KU)NEY LUNG GI-LLI 2.38E3 2.38E3 2.38E3 2.3i iE3 2.38E3 2.38E3 C 14' 3.23E6 6.89E5 6.89E5 6.89E5 6.89,E5 6.89E5 6.89E5 Cr 51 8.35E4 5.45E4 1.199E4 1.06E5 2.43E6 Mn 54 2.51E7 5.68E6 5.566E6 9.21E6 Fe 55 8.43E7 5.44E7 1.45E7 2.66E7 6.91E6 Fe 59 1.22E8 2.13E8 8.38E7 6.29E7 1.02E8 Co 58 1.39E7 3.46E7 3.46E7 Co 60 5.90E7 1.39E8 1.40E8 Zn 65 3.53E9 1.21E10 5.58E9 5.87E9 1.02E10 Sr 89 6.93E9 1.99E8 1.42E8 Sr 90 8.19E10 2.09E10 1.02E9 Zr 95 3.85E3 9.39E2 6.66E2 1.01E3 4.68E5 Nb 95 4.21E5 1.64E5 1.17E5 1.54E5 3.03E8 Mo 99 1.04E8 2.03E7 1.55E8 3.43E7 1131 6.81E8 8.02E8 3.53E8 2.64Ell 9.37E8 2.86E7 1133 8.52E6 1.24E7 3.63E6 2.26E9 1.46E7 2.10E6 Cs 134 2.41E10 4.49E10 4.54E9 1.16E10 4.74E9 1.22E8 Cs 137 3.47E10 4.06E10 2.88E9 1.09E10 4AlE9 1.27E8 Ba 140 1.21E8 1.21E5 6.22E6 2.87E4 7.42E4 2.97E7 La 140 2.03E1 7.99 2.06 9.39E4 Ce 141 2.28E4 1.39E4 1.64E3 4.28E3 7.18E6 Ce144 1.49E6 6.10E5 8.34E4 2.46E5 8.54E7 Nd 147 4.43E2 4.55E2 2.79E1 1.76E2 2.89E5

'mrem/yr per gCdm3 .

Unit 1 ODCM Revision 24 1153 July 2003

TABLE 3-10 DOSE AND DOSE RATE RI VALUES - COW MILK - CHILD

-Imremlvr VCsec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 3 1.57E3 1.57E3 1.57E3 1.57E3 1.57E3 1.57E3 C 14 1.65E6 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 3.29ES Cr 51 5.27E4 2.93E4 7.99E3 5.34E4 2.80E6 Mn 54 1.35E7 3.59E6 3.78E6 1.13E7 Fe 55 6.97E7 3.07E7 1.15E7 2.09E7 6.85E6 Fe 59 6.52E7 1.06E8 5.26E7 3.06E7 1.10E8 Co 58 6.94E6 2.13E7 4.05E7 Co 60 2.89E7 8.52E7 1.60E8 Zn 65 2.63E9 7.00E9 4.35E9 4AlE9 1.23E9 Sr 89 3.64E9 1.04E8 1.41E8 Sr 90 7.53E10 1.91E10 1.01E9 Zr 95 2.17E3 4.77E2 4.25E2 6.83E2 4.98E5 Nb 95 1.86E5 1.03E4 5.69E4 1.00E 4.42E8 Mo 99 4.07E7 1.01E7 8.69E7 3.37E7 1131 3.26E8 3.28E8 1.86E8 1.08El 1 5.39E8 2.92E7 1133 4.04E6 4.99E6 1.89E6 9.27E8 8.32E6 2.01E6 Cs 134 1.50E10 2.45E10 5.18E9 7.61E9 2.73E9 1.32E8 Cs 137 2.17E10 2.08E10 3.07E9 6.78E9 2.44E9 1.30E8 Ba 140 5.87E7 5.14E4 3A3E6 1.67E4 3.07E4 2.97E7 La 140 9.70 3.39 1.14 9.45E4 Ce 141 1.15E4 5.73E3 8.51E2 2.51E3 7.15E6 Ce 144 1.04E6 3.26E5 5.55E4 1.80E5 8.49E7 Nd 147 2.24E2 1.81E2 1.40E1 9.94E1 2.87E5

'mrem/yr per pCi/m3.

Unit 1 ODCM Revision 24 1154 July 2003

TABLE 3-11 DOSE AND DOSE RATE Ra VALUES - COW MILK - TEEN m nkmrem/vr MLCUsec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 C 14. 6.70E5 1.34E5 1.34E5 1.34E5 1.34E5 1.35E5 1.34E5 Cr51 2.58E4 1.44E4 5.66E3 3.69E4 4.34E6 Mn 54 9.01E6 1.79E6 2.69E6 1.85E7 Fe 55 2.78E7 1.97E7 4.59E6 1.251E7 8.52E6 Fe 59 2.81E7 6.57E7 2.54E7 2.071E7 1.55E8 Co58 4.55E6 1.05E7 6.27E7 Co 60 1.86E7 4.19E7 2.42E8 Zn 65 1.34E9 4.65E9 2.17E9 2.97E9 1.97E9 Sr 89 1.47E9 4.21E7 1.75E8 Sr 90 4.45E10 1.OE10O 1.25E9 Zr 95 9.34E2 2.95E2 2.03E2 4.33E2 6.80E5 Nb 95 1.86E5 1.03E5 5.69E4 1.OOES 4.42E8 Mo 99 2.24E7 4.27E6 5.12E7 4.01E7 I131 1.34E8 1.88E8 1.01E8 5.49E00 3.24E8 3.72E7 I133 1.66E6 2.82E6 8.59E5 3.93E8 4.94E6 2.13E6 Cs 134 6.49E9 1.53E10 7.08E9 4.85E9 1.85E9 1.90E8 Cs 137 9.02E9 1.20E10 4.18E9 4.08E9 1.591E9 1.71E8 Ba 140 2.43E7 2.98E4 1.57E6 1.01E4 2.00E4 3.75E7 La 140 4.05 1.99 5.30E-1 1.14E5 Ce 141 4.67E3 3.12E3 3.58E2 1.47E3 8.91E6 Co 144 4.22E5 1.74E5 2.27E4 1.04E5 1.06E8 Nd 147 9.12E1 9.91E1 5.94EO 5.82EI 3.58E5 mrem/yr per pICi/m 3 .

Unit 1 ODCM Revision 24 1155 July 2003

TABLE 3-12 DOSE AND DOSE RATE R&VALUES. COW MILK - ADULT 2

mn -mremrfr ttci/sec NUCLIDE BONE LIVER T. BODY THYROID KII)NEY LUNG GI-LLI H3 7.63E2 7.63E2 7.63E2 7.6 3E2 7.63E2 7.63E2 C 14^ 3.63E5 7.26E4 7.26E4 7.26E4 7.245E4 7.26E4 7.26E4 Cr51 1.48E4 8.85E3 3.2(5E3 1.96E4 3.72E6 Mn 54 5A1E6 1.03E6 1.6)ilE6 1.66E7 Fe 55 1.57E7 1.08E7 2.52E6 6.04E6 6.21E6 Fe 59 1.61E7 3.79E7 1.45E7 1.06E7 1.26E8 Co 58 2.70E6 6.05E6 5.47E7 Co 60 1.10E7 2.42E7 2.06E8 Zn 65 8.71E8 2.77E9 1.25E9 1.85E9 1.75E9 Sr 89 7.99E8 2.29E7 1.28E8 Sr 90 3.15E10 7.74E9 9.11E8 Zr 95 5.34E2 1.71E2 1.16E2 2.69E2 5.43E5 Nb 95 1.09E5 6.07E4 3.27E4 6.00E4 3.69E8 Mo 99 1.24E7 2.36E6 2.81E7 2.87E7 I 131 7.41E7 1.06E8 6.08E7 3.47E10 1.82E8 2.80E7 1133 9.09E5 1.58E6 4.82E5 2.32E8 2.76E6 1.42E6 Cs 134 3.74E9 8.89E9 7.27E9 2.88E9 9.55E8 1.56E8 Cs 137 4.97E9 6.80E9 4A6E9 2.31E9 7.68E8 1.32E8 Ba 140 1.35E7 1.69E4 8.83E5 5.75E3 9.69E3 2.77E7 La 140 2.26 1.14 3.01E-1 8.35E4 Ce 141 2.54E3 1.72E3 1.95E2 7.99E2 6.58E6 Ce 144 2.29E5 9.58E4 1.23E4 5.68E4 7.74E7 Nd 147 4.74E1 5A8E1 3.28E0 3.20E1 2.63E5 mrem/yr per ACi/m3.

Unit 1 ODCM Revision 24 II56 July 2003

TABLE 3-13 DOSE AND DOSE RATE Ri VALUES - GOAT MILK - INFANT zn2 mremhVr PCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDINEY LUNG GI-LLI Hr3 6.33E3 6.33E3 6.33E3 6.333E3 6.33E3 6.33E3 C 14. 3.23E6 6.89E5 6.89E5 6.89E5 6.89 ES 6.89E5 6.89E5 Cr 51 1.00E4 6.56E3 1.43 E3 1.28E4 2.93E5 Mn 54 3.01E6 6.82135 6.677ES 1.11E6 Fe 55 1.10E6 7.08E5 1.891E5 3.46E5 8.98E4 Fe 59 1.59E6 2.78E6 1.09E6 8.21E5 1.33E6 Co 58 1.67E6 4.16E6 4.16E6 Co 60 7.08E6 1.67E7 1.68E7 Zn 65 4.24E8 1.45E9 6.701E8 7.04E38 1.23E9 Sr 89 1.48E10 4.241E8 3.04E8 Sr 90 .1.72E11 4.381E10 2.15E9 Zr 95 4.66E2 1.13E2 8.04E1 1.22E12 5.65E4 Nb 95 9.42E4 3.88E4 2.24E4 2.78E4 3.27E7 Mo 99 1.27E7 2.471E6 1.89E7 4.17E6 I131 8.17E8 9.63E8 4.231E8 3.16El 1 1.12E9 3.44E7 I133 1.02E7 1.49E7 4.36E36 2.71E9 1.75E7 2.52E6 Cs 134 7.23E10 1.35E1l1 1.36E10 3.47E310 1.421E10 3.66E8 Cs 137 1.04E11 1.22E1 1 8.63E9 3.271E10 1.321E10 3.81E8 Ba 140 1A5E7 1.45E4 7.48E5 3.44113 8.91133 3.56E6 La 140 2A3E0 9.59E-1 2.47E1-1 1.13E4 Ce 141 2.74E3 1.67E3 1.96E12 5.14E12 8.62E5 Ce144 1.79E5 7.32E4 1.00E4 2.96E34 1.03E7 Nd 147 5.32E1 5.47E1 3.35E0 2.1 1E1 3.46E4 mrenhyr per pCi/m3.

Unit 1 ODCM Revision 24 II57 July 2003

TABLE 3-14 DOSE AND DOSE RATE R&VALUES - GOAT MILK. CHILD m -mrernlvr gCVsec NUCLIDE BONE LIVER T. BODY THYROID KII)NEY LUNG GI-LLI H-37 4.17E3 4.17E3 4.17E3 4.1, 7E3 4.17E3 4. 17E C 14' 1.65E6 3.29E5 3.29E5 3.29E5 3.25)E5 3.29E5 3.29E5 Cr 51 6.34E3 3.52E3 9.6, iE2 6.43E3 3.36E5 Mn 54 1.62E6 4.3 IE5 4.54i4E5 1.36E6 Fe 55 9.06E5 4.81E5 1.49E5 2.72115 8.91E4 Fe 59 8.52E5 1.38E6 6.86E5 3.99115 1.43E16 Co 58 8.35E5 2.56E6 4.87E16 Co 60 3.47E6 1.02E7 1.92E7 Zn 65 3.15E8 8.40E8 5.23E8 5.29E8 1.48E8 Sr 89 7.77E9 2.22E8 3.01E8 Sr 90 1.58E1 1 4.01E310 2.13E9 Zr 95 2.62E2 5.76E1 5.13EI 8.25E1 6.01E4 Nb 95 5.05E4 1.96E4 1.40E4 1.85E4 3.63E7 Mo 99 4.95E6 1.22E6 1.06E7 4.09E6 1131 3.91E8 3.94E8 2.24E8 1.30E1 1 6.46E8 3.50E7 1133 4.84E6 5.99E6 2.27E6 1.11E9 9.98E6 2.41E6 Cs 134 4.49E10 7.37E10 1.55E10 2.28E10 8.19E9 3.97E8 Cs 137 6.52E10 6.24E10 9.2 1E9 2.03E10 7.32E9 3.91E18 Ba 140 7.05E6 6.18E3 4.12E5 2.01E3 3.68E3 3.57E6 La 140 1.16E0 4.07E-1 1.37E-1 1.13E4 Ce 141 1.38E3 6.88E2 1.02E2 3.02E2 8.59E5 Ce 144 1.25E5 3.91E4 6.66E3 2.16E4 1.02E7 Nd 147 2.68E1 2.17E1 1.68E0 1.19E1 3.44E4 nmremtyr per gCi/m3.

Unit 1 ODCM Revision 24 1158 July 2003

TABLE 3-15 DOSE AND DOSE RATE R1 VALUES - GOAT MILK - TEEN e 2 -mremnvr ga/sec NUCLIDE BONE LIVER T. BODY THYROID KI DNEY LUNG GI-LLI H3 2.64E3 2.64E3 2.64E3 2.64E3 2.64E3 2.64E3 C 14* 6.70E5 1.34E5 1.34ES 1.34E5 1.34E5 1.35E5 1.34E5 Cr51 3.11E3 1.73E3 6.882E2 4.44E3 5.23E5 Mn 54 1.08E6 2.15E5 3.2 3E5 2.22E6 Fe 55 3.61E5 2.56E5 5.97E4 1.62E5 1.11E5 Fe 59 3.67E5 8.57E5 3.31E5 2.70E5 2.03E6 Co58 5.46E5 1.26E6 7.53E6 Co 60 2.23E6 5.03E6 2.91E7 Zn 65 1.61E8 5.58E8 2.60E8 3.57E8 2.36E8 Sr 89 3.14E9 8.99E7 3.74E8 Sr 90 9.36E10 2.31E10 2.63E9 Zr 95 1. 13E2 3.56EI 2.45E1 5.23E1 8.22E4 Nb 95 2.23E4 1.24E4 6.82E3 1.20E4 5.30E7 Mo 99 2.72E6 5.19E5 6.23E6 4.87E6 1131 1.61E8 2.26E8 1.21E8 6.59E10 3.89E8 4.47E7 I 133 1.99E6 3.38E6 1.03E6 4.72E8 5.93E6 2.56E6 Cs 134 1.95E10 4.58E10 2.13E10 1.46E10 5.56E9 5.70E8 Cs 137 2.71E10 3.60E10 1.25E10 1.23E10 4.76E9 5.12E8 Ba 140 2.92E6 3.58E3 1.88E5 1.21E3 2.41E3 4.50E6 La 140 4.86E-1 2.39E-1 6.36E-2 1.37E4 Ce 141 5.60E2 3.74E2 4.30E1 1.76E2 1.07E6 Cc 144 5.06E4 2.09E4 2.72E3 1.25E4 1.27E7 Nd 147 1.09EI 1.19EI 7.13E-1 6.99E0 4.29E4 rnrem/yr per pCi/m 3 .

Unit 1 ODCM Revision 24 1 59 July 2003

TABLE 3-16 DOSE AND DOSE RATE R

1 VALUES - GOAT MILK - ADULT m2 -mrern/vr ACusec NUCLI DE BONE LIVER T. BODY THYROID KIEDNEY LUNG GI-LLI H3 -- 2A.03'E 2.03E3 2.03E3 2.03P3E3 2l.03E 2.03E3 C 14' 3.63E5 7.26E4 7.26E4 7.26E4 7.266E4 7.26E4 7.26E4 Cr 51 1.78E3 1.06E3 3.922E2 2.36E3 4.48E5 Mn 54 6.50E5 1.24E5 1.933ES 1.99E6 Fe 55 2.04E5 1.41ES 3.28E4 7.85E4 8.07E4 Fe 59 2.10ES 4.95E5 1.90E5 1.38E5 1.65E6 Co 58 3.25E5 7.27E5 6.58E6 Co 60 1.32E6 2.91E6 2.48E7 Zn 65 1.05E8 3.33E8 1.51E8 2.23E8 2.10E8 Sr 89 1.70E9 4.89E7 2.73E8 Sr 90 6.62E10 1.63E10 1.91E9 Zr 95 6.45E1 2.07E1 1.40E1 3.25E1 6.56E4 Nb 95 1.31E4 7.29E3 3.92E3 7.21E3 4.42E7 Mo 99 1.51E6 2.87E5 3.41E6 3.49E6 1131 8.89E7 1.27E8 7.29E7 4.17E10 2.18E8 3.36E7 1133 1.09E6 1.90E6 5.79E5 2.79E8 3.31E6 1.71E6 Cs 134 1.12E10 2.67E10 2.18E10 8.63E9 2.86E9 4.67E8 Cs 137 1.49E10 2.04E10 1.34E10 6.93E9 2.30E9 3.95E8 Ba 140 1.62E6 2.03E3 1.06E5 6.91E2 1.16E3 3.33E6 La 140 2.71E-1 1.36E-1 3.61E-2 1.00E4 Cc 141 3.06E2 2.07E2 2.34E1 9.60E1 7.90E5 Cc 144 2.75E4 1.15E4 1A8E3 6.82E3 9.30E6 Nd 147 5.69E0 6.57E0 3.93E-1 3.84E0 3.15E4 mnrem/yr per pCi/m3.

Unit 1 ODCM Revision 24 1160 July 2003

TABLE 3-17 DOSE AND DOSE RATE RI VALUES. COW MEAT. CHILD m2 -mremnfvr ttCivsec NUCLIDE BONE LIVER T. BODY THYROID KIIDNEY LUNG GI-LLI 2.34E2 2.34E2 2.34E2 2.3, 4E2 2.34E2 2.34E2 C 14 5.29E5 1.06E5 1.06E5 1.06E5 1.0 5E5 1.06E5 1.06E5 Cr 51 4.55E3 2.52E3 6.91DE2 4.61E3 2.41E5 Mn 54 5.15E6 1.37E6 1.4 4116 4.32E6 Fe 55 2.89E8 1.53E8 4.74E7 8.66E7 2.84E7 Fe 59 2.04E8 3.30E8 1.65E8 9.58E7 3.44E8 Co 58 9A1E6 2.88E7 5.49E7 Co 60 4.64E7 1.37E8 2.57E8 Zn 65 2.38E8 6.35E8 3.95E8 4.00E8 1.12E8 Sr 89 2.65E8 7.57E6 1.03E7 Sr 90 7.01E9 1.78E9 9.44E7 Zr 95 1.51E6 3.32E5 2.95E5 4.751E5 3.46E8 Nb 95 4.10E6 1.59E6 1.14E6 1.501E6 2.95E9 Mo 99 5.42E4 1.34E4 1.161E5 4.48E4 1131 4.15E6 4.18E6 2.37E6 1.38E9 6.861E6 3.72E5 I133 9.38E-2 1.16E-1 4.39E-2 2.15E1 1.931E-1 4.67E-2 Cs 134 6.09E8 1.00E9 2.11E8 3.10E8 1.1 lE8 5.39E6 Cs 137 8.99E8 8.60E8 1.27E8 2.80E8 1.01E8 5.39E6 Ba 140 2.20E7 1.93E4 1.28E6 6.271E3 1.15E4 1.11E7 La 140 2.80E-2 9.78E-3 3.30E-3 2.73E2 Ce 141 1.17E4 5.82E3 8.64E2 2.55E3 7.26E6 Ce144 1.48E6 4.65E5 7.91E4 2.57E15 1.21E8 Nd 147 5.93E3 4.80E3 3.72E2 2.64133 7.61E6

mrem/yrper 1gCi/m3.

Unit 1 ODCM Revision 24 1161 July 2003

TABLE 3-18 DOSE AND DOSE RATE R. VALUES - COW MEAT - TEEN

_' -mremfvr

,pCVsec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 37 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 C 14 2.81E5 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 Cr 51 2.93E3 1.62E3 6.39E2 4.16E3 4.90E5 Mn 54 4.50E6 8.93E5 1.34E6 9.24E6 Fe 55 1.5OE8 1.07E8 2.49E7 6.77E7 4.62E7 Fe 59 1.15E8 2.69E8 1.04E8 8.47E7 6.36E8 Co 58 8.05E6 1.86E7 1.11E8 Co 60 3.90E7 8.80E7 5.09E8 Zn 65 1.59E8 5.52E8 2.57E8 3.53E8 2.34E8 Sr 89 1.40E8 4.01E6 1.67E7 Sr 90 5.42E9 1.34E9 1.52E8 Zr 95 8.50E5 2.68E5 1.84E5 3.94E5 6.19E8 Nb 95 2.37E6 1.32E6 7.24E5 1.28E6 5.63E9 Mo 99 3.90E4 7.43E3 8.92E4 6.98E4 1131 2.24E6 3.13E6 1.68E6 9.15E8 5.40E6 6.20E5 I133 5.05E-2 8.57E-2 2.61E-2 1.20E1 1.50E-1 6.48E-2 Cs 134 3.46E8 8.13E8 3.77E8 2.58E8 9.87E7 1.01E7 Cs 137 4.88E8 6.49E8 2.26E8 2.21E8 8.58E7 9.24E6 Ba 140 1.19E7 1.46E4 7.68E5 4.95E3 9.81E3 1.84E7 La 140 1.53E-2 7.51E-3 2.OOE-3 4.31E2 Ce 141 6.19E3 4.14E3 4.75E2 1.95E3 1.18E7 Ce 144 7.87E5 3.26E5 4.23E4 1.94E5 1.98E8 Nd 147 3.16E3 3.44E3 2.06E2 2.02E3 1.24E7 nmrem/yr per Ci/m3 .

Unit 1 ODCM Revision 24 1162 July 2003

LBLE 3-19 DOSE AlND DOSE RATE R VALUES - ( COW MEAT - ADULT mW,mremmvr gCilsec NUCLIDE BONE LIVER -

T. BODY THYROID KUII )NEY LUNG GI-LLI H37 3.25E2 3.25E2 3.25E2 3.2' 5E2 3.25E2 3.25E2 C 14* 3.33E5 6.66E4 6.66E4 6.66E4 6.6( 5E4 6.66E4 6.66E4 Cr51 3.65E3 2.18E3 8.0: 1E2 4.84E3 9.17E5 Mn 54 5.90E6 1.13E6 -- 1.7( 'E6 1.81E7 Fe 55 1.85E8 1.28E8 2.98E7 7.14E7 7.34E7 Fe 59 1.44E8 3.39E8 1.30E8 9.46E7 1.13E9 Co 58 1.04E7 2.34E7 2.12E8 Co 60 5.03E7 1.11E8 9.45E8 Zn 65 2.26E8 7.19E8 3.25E8 4.81E8 4.53E8 Sr 89 1.66E8 4.76E6 2.66E7 Sr 90 8.38E9 2.06E9 2.42E8 Zr 95 1.06E6 3.40E5 2.30E5 5.34E5 1.08E9 Nb 95 3.04E6 1.69E6 9.08E5 1.67E6 1.03E10 Mo 99 4.71E4 8.97E3 1.07E5 1.09E5 1131 2.69E6 3.85E6 2.21E6 1.26E9 6.61E6 1.02E6 1133 6.04E-2 1.05E-1 3.20E-2 1.54E1 1.83E-1 9.44E-2 Cs 134 4.35E8 1.03E9 8.45E8 3.35E8 1.11E8 1.81E7 Cs 137 5.88E8 8.04E8 5.26E8 2.73E8 9.07E7 1.56E7 Ba 140 1.44E7 1.81E4 9.44E5 6.15E3 1.04E4 2.97E7 La 140 1.86E-2 9.37E-3 2.48E-3 6.88E2 Ce 141 7.38E3 4.99E3 5.66E2 2.32E3 1.91E7 Ce 144 9.33E5 3.90E5 5.01E4 2.31E5 3.16E8 Nd 147 3.59E3 4.15E3 2.48E2 2.42E3 1.99E7 mrem/yr per AlCi/m 3.

Unit 1 ODCM Revision 24 1163 July 2003

TABLE 3-20 DOSE AND DOSE RATE RI VALUES - V EGETATION - CHILD ni2-mrem/yr pCifsec NUCLIDE BONE LIVER -

T. BODY THYROID KID NEY LUNG GI-LLI H 37 4.01E3 4.01E3 4.01E3 4.01] E3 4.01E3 4.01E3 C 14. 3.50E6 7.01E5 7.01E5 7.01ES 7.01] ES 7.011ES 7.01E5 Cr51 1.17E5 6.49E4 1.77]E4 1.18E5 6.20E6 Mn 54 6.65E8 1.77E8 -- 1.86] 5.58E8 Fe 55 7.63E8 4.05E8 1.25E8 2.29E8 7.50E7 Fe 59 3.97E8 6.42E8 3.20E8 1.86E8 6.69E8 Co58 6.45E7 1.97E8 3.76E8 Co 60 3.78E8 1.12E9 2.10E9 Zn 65 8.12E8 2.16E9 1.35E9 1.36E9 3.80E8 Sr 89 3.59E10 1.03E9 1.39E9 Sr 90 1.24E12 3.15E11 1.67E10 Zr 95 3.86E6 8.50E5 7.56E5 1.22E6 8.86E8 Nb 95 1.02E6 3.99E5 2.85E5 3.75E5 7.37E8 Mo 99 7.70E6 1.91E6 1.65E7 6.37E6 1131 7.16E7 7.20E7 4.09E7 2.38E10 1. 18E8 6.41E6 1133 1.69E6 2.09E6 7.92E5 3.89E8 3.49E6 8.44E5 Cs 134 1.60E10 2.63E10 5.55E9 8.15E9 2.93E9 1.42E8 Cs 137 2.39E10 2.29E10 3.38E9 7.46E9 2.68E9 1.43E8 Ba 140 2.77E8 2.43E5 1.62E7 7.90E4 1.45E5 1.40E8 La 140 3.25E3 1.13E3 3.83E2 3.16E7 Ce 141 6.56E5 3.27E5 4.85E4 1.43E5 4.08E8 Ce144 1.27E8 3.98E7 6.78E6 2.21E7 1.04E10 Nd 147 7.23E4 5.86E4 4.54E3 3.22E4 9.28E7 mrenmlr per gCi/m3.

Unit 1 ODCM Revision 24 11 64 July 2003

TABLE 3-21 DOSE AND DOSE RATE Ri VALUES - VEGETATION - TEEN m -rmremfvr iiCI/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3 C 14' 1.45E6 2.91E5 2.91E5 2.91E5 2.91E5 2.91E5 2.91E5 Cr 51 6.16E4 3.42E4 1.35E4 8.79E4 1.03E7 Mn 54 4.54E8 9.01E7 1.36E8 9.32E8 Fe 55 3.10E8 2.20E8 5.13E7 1A0E8 9.53E7 Fe 59 1.79E8 4.18E8 1.61E8 1.32E8 9.89E8 Co58 4.37E7 1.01E8 6.02E8 Co60 2.49E8 5.60E8 3.24E9 Zn 65 4.24E8 1.47E9 6.86E8 9.41E8 6.23E8 Sr 89 1.51E10 4.33E8 1.80E9 Sr 90 7.51Ell 1.85E1 1 2.11E10 Zr 95 1.72E6 5.44E5 3.74E5 7.99E5 1.26E9 Nb 95 4.80E5 2.66E5 1A6E5 2.58E5 1. 14E9 Mo 99 5.64E6 1.08E6 1.29E7 1.01E7 1131 3.85E7 5.39E7 2.89E7 1.57E10 9.28E7 1.07E7 I133 9.29E5 1.58E6 4.80ES 2.20E8 2.76E6 1.19E6 Cs 134 7.10E9 1.67E10 7.75E9 5.31E9 2.03E9 2.08E8 Cs 137 1.O1E10 1.35E10 4.69E9 4.59E9 1.78E9 1.92E8 Ba 140 1.38E8 1.69E5 8.91E6 5.74E4 1.14E5 2.13E8 La 140 1.81E3 8.88E2 2.36E2 5.10E7 Ce 141 2.83E5 1.89E5 2.17E4 8.89E4 5.40E8 Cc 144 5.27E7 2.18E7 2.83E6 1.30E7 1.33E10 Nd 147 3.66E4 3.98E4 2.3863 2.34E4 1.44E8 mnren/yr per pCi/m3 Unit 1 ODCM Revision 24 H 65 July 2003

TABLE 3-22 DOSE AND DOSE RATE RFVALUES - VEGETATION -ADULT M2 -Mremfvr ICisec NUCLIDE BONE LIVER -

T. BODY THYROID KID] WNY LUNG GI-LLI H3 2.26E3 2.26E3 2.26E3 2.261E33 2.26E3 2.26E3 C 14' 8.97E5 1.79E5 1.79E5 1.79E5 1.791ES1 1.79E5 1.79E5 Cr 51 4.64E4 2.77E4 1.021E4 6.15E4 1.17E7 Mn 54 3.13E8 5.97E7 9.311E71 9.58E8 Fe 55 2.00E8 1.38E8 3.22E7 7.69E7 7.91E7 Fe 59 1.26E8 2.96E8 1.13E8 8.27E7 1.02E9 Co 58 3.08E7 6.90E7 6.24E8 Co 60 1.67E8 3.69E8 3.14E9 Zn 65 3.17E8 1.01E9 4.56E8 6.75E8 6.36E8 Sr 89 9.96E9 2.86E8 1.60E9 Sr 90 6.05El 1 1.48E1 1 1.75E10 Zr 95 1.18E6 3.77E5 2.55E5 5.92E5 1.20E9 Nb 95 3.55E5 1.98E5 1.06ES 1.95E5 1.20E9 Mo 99 6.14E6 1.17E6 1.39E7 1.42E7 I131 4.04E7 5.78E7 3.31E7 1.90E10 9.91E7 1.53E7 1133 1.00E6 1.74E6 5.30E5 2.56E8 3.03E6 1.56E6 Cs 134 4.67E9 1.1lE10 9.08E9 3.59E9 1.19E9 1.94E8 Cs 137 6.36E9 8.70E9 5.70E9 2.95E9 9.81E8 1.68E8 Ba 140 1.29E8 1.61E5 8.42E6 5.49E4 9.25E4 2.65E8 La 140 1.98E3 9.97E2 2.63E2 7.32E7 Ce 141 1.97E5 1.33E5 1.5 E4 6.19E4 5.09E8 Ce 144 3.29E7 1.38E7 1.77E6 8.16E6 1.11E10 Nd 147 3.36E4 3.88E4 2.32E3 2.27E4 1.86E8

mrem/yr per plCi/m3 Unit 1 ODCM Revision 24 1 66 July 2003

TABLE 3-23 PARAMETERS FOR THE EVALUATION OF DOSES TO REAL MEMBERS OF THE PUBLIC FROM GASEOUS AND LIQUID EFFLUENTS Pathway Parameters Value Reference Fish U (kg/yr) - adult 21 Reg. Guide 1.109 Table E-5 Fish Daipj (mremlpCi) Each Radionuclide Reg. Guide 1.109 Table E-1 1 Shoreline U (hr/yr) 67 Reg. Guide 1.09

- adult 67 Assumed to be same as Adult

- teen Shoreline Ddipj Each Radionuclide Reg. Guide 1.109 (nirem/hr per pCi/m 2 ) Table E-6 Inhalation DFAij. Each Radionuclide Reg. Guide 1.109 Table E-7 Unit 1 ODCM Revision 24 1167 July 2003

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

Map Collection Site Tvpe of Samble Location (Env. Program No.) Location Radioiodine and 1 Nine Mile Point Road 1.8 mi@88 E Particulates (air) North (R-1)

Radioiodine and 2 Co. Rt. 29 & Lake Road (R-2) 1.1 mi @ 104' ESE Particulates (air)

Radioiodine and 3 Co. Rt. 29 (R-3) 1.5 mi @ 132 SE Particulates (air)

Radioiodine and 4 Village of Lycoming, NY (R-4) 1.8 mi @ 143' SE Particulates (air)

Radioiodine and 5 Montario Point Road (R-5) 16.4mi @42NE Particulates (air)

Direct Radiation (TLD) 6 North Shoreline Area (75) 0.lmi@5'N Direct Radiation (TLD) 7 North Shoreline Area (76) 0.1 mi @ 25 NNE Direct Radiation (TLD) 8 North Shoreline Area (77) 0.2 mi @ 45 NE Direct Radiation (TLD) 9 North Shoreline Area (23) 0.8 mi @ 70 ENE Direct Radiation (TLD) 10 JAF East Boundary (78) 1.0mi @90E Direct Radiation (TLD) 11 Rt. 29 (79) 1.1 mi @ 115'ESE Direct Radiation (TLD) 12 Rt. 29 (80) 1.4mi @ 133SE Direct Radiation (TLD) 13 Miner Road (81) 1.6 mi @ 159 SSE Direct Radiation (TLD) 14 Miner Road (82) 1.6 mi @ 18I'S Direct Radiation (TLD) 15 Lakeview Road (83) 1.2 mi @ 200 SSW Direct Radiation (TLD) 16 Lakeview Road (84) 1.1 mi @ 225 SW Direct Radiation (TLD) 17 Site Meteorological Tower (7) 0.7 mi @ 250 WSW Direct Radiation (TLD) 18 Energy Information Center (18) 0.4 mi @ 265 W

Map = See Figures 5.1-1 and 5.1-2 Unit 1 ODCM Revision 24 1168 July 2003

TABLE 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collection Site Type of Samnle Location (Env. Proeram No.) Location Direct Radiation (TLD) 19 North Shoreline (85) 0.2 mri C@294 WNW Direct Radiation (TLD) 20 North Shoreline (86) 0.1 mi ( @315'W Direct Radiation (TLD) 21 North Shoreline (87) 0.1 mi C@341 NNW Direct Radiation (TLD) 22 Hickory Grove (88) 4.5 mi C@97 E Direct Radiation (TLD) 23 LeavittRoad(89) 4.1miC @111 ESE Direct Radiation (TLD) 24 Rt. 104 (90) 4.2 mi ( @135 SE Direct Radiation (TLD) 25 Rt. 51A (91) 4.8 mi 0@156*SSE Direct Radiation (TLD) 26 Maiden Lane Road (92) 4.4 mi ( @183' S Direct Radiation (TLD) 27 Co. Rt. 53 (93) 4.4 mi 6@205' SSW Direct Radiation (TLD) 28 Co.Rt. 1 (94) 4.7 mi l @223' SW Direct Radiation (TID) 29 Lake Shoreline (95) 4.1 mi ( @237' WSW Direct Radiation (TLD) 30 Phoenix, NY Control (49) 19.8 mi @ 163' S Direct Radiation (TLD) 31 S. W. Oswego, Control (14) 12.6 mi @ 226' SW Direct Radiation (TLD) 32 Scriba, NY (96) 3.6 mri C@199' SSW Direct Radiation (TLD) 33 Alcan Aluminum, Rt. IA (58) 3.1 mri t @220' SW Direct Radiation (TLD) 34 Lycoming, NY (97) 1.8 mri ( @143' SE Direct Radiation (TLD) 35 New Haven, NY (56) 5.3 mi (I@123' ESE Direct Radiation (TLD) 36 W. Boundary, Bible Carnp (15) 0.9 mi Cz @237' WSW Direct Radiation (TLD) 37 Lake Road (98) 1.2 mi ( @101'E Surface Water 38 OSS Inlet Canal (NA) 7.6 mi (C @235' SW Surface Water 39 JAFNPP Inlet Canal (NA) 0.5 ml (C @70' ENE (NA) = Not applicable

Map = See Figures 5.1-1 and 5.1-2 Unit 1 ODCM Revision 24 I 69 July 2003

TABLE 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collection Site Type of Samile Location (Env. Program No.) Location Shoreline Sediment 40 Sunset Bay Shoreline (NA) 1.5 mii@ 86 E Fish 41 NMP Site Discharge Area (NA) 0.3 mii@ 315 NW (and/c'r)

Fish 42 NMP Site Discharge Area (NA) 0.6 mii @ 55 NE Fish 43 Oswego Harbor Area (NA) 6.2 ni i @ 235 SW Milk 44 Milk Location #50 8.2 ni i @ 93 E Milk 64 Milk Location #55 9.0 mii @ 95 E Milk 65 Milk Location #60 9.5 mii@ 9 E Milk 66 Milk Location #4 7.8 mi @i113 ESE Milk (CR) 77 Milk Location 13.9 mIni @ 1910 SSW (Summerville)

Food Product 48 Produce Location #6** 1.9 mri @ 141 SE (Bergenstock) (NA)

Food Product 49 ProduceLocation#1** 1.7nim i @ 96 E (Culeton) (NA)

Food Product 50 Produce Location #2** 1.9 mii @ 101'E (Vitullo) (NA)

Food Product 51 Produce Location #5** 1.5 mii @ 114 ESE (C.S. Parkhurst) (NA)

Food Product 52 ProduceLocation#3** 1.6mm I@ 84'E (C. Narewski) (NA)

Map = See Figures 5.1-1 and 5.1-2

- = Food Product Samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q.

(NA) - Not applicable CR - Control Result (location)

Unit 1 ODCM Revision 24 1170 July 2003

TABLE 5.1 (Cont'd)

NINE MILE POINT NUCLEAR STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM SAMPLING LOCATIONS

Map Collection Site TvPe of Sample Location (Env. Program No.) Location Food Product 53 Produce Location #4** 2.1 mrii @ 110 ESE (P. Parkhurst) (NA)

Food Product (CR) 54 Produce Location #7** 15.0 rni @223 SW (Mc Millen) (NA)

Food Product (CR) 55 Produce Location #8*

12.6 r ni @ 225 SW (Denman) (NA)

Food Product 56 Produce Location #9** 1.6 nti@ 171'S (O'Connor) (NA)

Food Product 57 Produce Location #10*

2.2 ni i @ 123 ESE (C. Lawton) (NA)

Food Product 58 Produce Location #11 ** 2.0 ni @ 112 ESE (C. R. Parkhurst) (NA)

Food Product 59 Produce Location #12*

1.9 mi @ 115 ESE (Barton) (NA)

Food Product (CR) 60 Produce Location #13** 15.6 r ni @225 SW (Flack) (NA)

Food Product 61 Produce Location #14** 1.9 m:i@ 95 E (Koeneke) (NA)

Food Product 62 Produce Location #15** 1.7 m i @ 136 SE (Whaley) (NA)

Food Product 63 ProduceLocation#16** 1.2 n i @207 SSW (Murray) (NA)

Food Product 67 Produce Location #17** 1.76 rni @ 97 E (Battles)

Map = See Figures 5.1-1 and 5.1-2

- = Food Product Samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q.

(NA) = Not applicable CR = Control Result (location)

Unit 1 ODCM Revision 24 1171 July 2003

Unit 1 ODCM II72 Revision 24 July 2003

SCALE Of MILES FIGURE 5.1-2 C U

_. J===

U.S. A St-to Hghway .

LEGEND NINE MILE POINT .

U"

.5'1 2 County Reeds.....................

Town Read ...

Cn...ly Lien.........................

Town Line9 .. . .. ____

OFF-SITE MAP *_ 4 City& Vilhge Ln ...

Railhonds .. ..............

ENVIRONMENTAL SAMPLE ...

A w .(10/2001)

LOCATION LeUtds 43yWN.2 Lonituds 7'3crW.

et Os.lo Couny lit. O.w-p NY.

aLn.dA... MSqf.l e L A KE O N T A R IO

- l w A I

I I

A.-

!,.)%

-I'.

tQ 9,; . a RIP',,,

i 11 41 c 7-1 BO'K 0

i! I 6i

--A

%~

%, * ~Comersu

- I I . n lL.&.

,ZI9 i

I..d GranyOAIA i

i L

5.1-2a 1 C-4 FIGURE 20 M 79

(d) ,p(e)

A TT. R 0

. z JAMES A.FiTZPATRICK

- r(b) NUgCLEAR POWER ME i PLANT P~~

00 PMtPNT

/ASO xE aeve, I ,','

NULA TTO. _L ENEG Niagara Mohawk Power Corporation retains ownership hn certain transmnission line and switchyard facilities within the exclusion~area boundary. Access and usage are conlrolled by Nine Mile Point Nucdear Station, LL.C by AgreemenL p 1,4 1(2 FIGUR 5.131 -

SCALE-MILES SITE BOUNDAR IES NINE MTE pOIN-UNT 1 Unit 1 ODCM Revision 24 II174 July 2003

APPENDIX A LIQUID DOSE FACTOR DERIVATION Unit 1 ODCM Revision 24 1 75 July 2003

Appendix A Liquid Effluent Dose Factor Derivation, Ajt At (mremthr per gCi/ml) which embodies the dose conversion factors, pathway transfer factors (e.g.,

bioaccumulation factors), pathway usage factors, and dilution factors for the points of pathway origin takes into account the dose from ingestion of fish and drinking water and the sediment. The total body and organ dose conversion factors for each radionuclide will be used from Table E-1 1 of Regulatory Guide 1.109. To expedite time, the dose is calculated for a maximum individual instead of each age group. The maximum individual dose factor is a composite of the highest dose factor Ait of each nuclide i age group a, and organ t, hence Al.. It should be noted that the fish ingestion pathway is the most significant pathway for dose from liquid effluents. The water consumption pathway is included for consistency with NUREG 0133.

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

A Ko[Uw(e i pw +Uf (BF) 1(e'AltPf )DFL)a 69.3 U6 W e-AItPs (1-el AItb )(DFS))]

Where:

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

Ko = Is the unit conversion factor, 1.14E5=lE6pCi/gCi x 1E3 ml/kg

-:- 8760 hr/yr.

UW Water consumption (l/yr); from Table E-5 of Reg. Guide 1.109.

Uf = Fish consumption (Kg/yr); from Table E-5 of Reg. Guide 1.109.

U. = Sediment Shoreline Usage (hr/yr); from Table E-5 of Reg. Guide 1.109.

(BF)i = Bioaccumulation factor for nuclide, i, in fish, (pCi/kg per pCi/l),

from Table A-1 of Reg. Guide 1.109.

(DFL)it = Dose conversion factor for age, nuclide, i, group a, total body or organ t, (niremlpCi); from Table E-11 of Reg. Guide 1.109.

(DFS)1 = Dose conversion factor for nuclide i and total body, from standing on contaminated ground (memlhr per pCi/mi); from Table E-6 of Reg. Guide 1.109.

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

Unit 1 ODCM Revision 24 1 76 July 2003

Appendix A (Cont'd)

D, Dilution factor from the near field area within one quarter mile of the release point to the shoreline deposit (taken at the same point where we take environmental samples 1.5 miles; unitless).

69.3 = conversion factor .693 x 100, 100 = K (LIkg-hr) *40*24 hr/dayl.693 in Im 2 -d, and Ks = transfer coefficient from water to sediment in L/kg per hour.

tPw9 tpf, = Average transit time required for each nuclide to tP. reach the point of exposure for internal dose, it is the total time elapsed from release of the nuclides to either ingestion for water (w) and fish (f) or shoreline deposit (s), (hr).

tb = Length of time the sediment is exposed to the contaminated water, nominally 15 yrs (approximate midpoint of facility operating life),

(hrs).

=decay constant for nuclide i (hr ').

W = Shore width factor (unitless) from Table A-2 of Reg. Guide 1.109.

Example Calculation For I-131 Thyroid Dose Factor for an Adult from a Radwaste liquid effluents release:

(DFS), = 2.80E-9 mrem/hr per pCi/m2 (DFL)ja = 1.95E-3 mrem/pCi t1W = 30 hrs. (w = water)

BFj = 15 pCilKg per pCi/L tpf = 24 hrs. (f = fish)

Uf = 21 Kg/yr tb = 1.314E5 hrs. (5.48E3 days)

Dw = 40 unitless Us = 730 yr D, = 12 unitless Ko = 1.14E5 (pCi/iCfl(ml/ka)

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

W =0.3 = 3.61E-3hfl tp = 5.5 hrs (s = Shoreline Sediment)

These values will yield an Awt Factor of 6.79E4 mrem-ml per jLCi-hr as listed in Table 24. It should be noted that only a limited number of nuclides are listed on Tables 2-1 to 2-8. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in a revision to the ODCM.

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

Unit 1 ODCM Revision 24 1177 July 2003

APPENDIX B PLUME SHINE DOSE FACTOR DERIVATION Unit 1 ODCM Revision 24 H178 July 2003

APPENDIX B For elevated releases the plume shine dose factors for gamma air (Bi) and whole body (Vi), are calculated using the finite plume model with an elevation above ground equal to the stack height. To calculate the plume shine factor for gamma whole body doses, the gamma air dose factor is adjusted for the attenuation of tissue, and the ratio of mass absorption coefficients between tissue and air. The equations are as follows:

Gamma Air Bi =l KE I. Where: K' = conversion factor (see R9 -V, below for actual value).

. = mass absorption coefficient (cm 2 /g; air for Bi, tissue for Va)

E = Energy of gamma ray per disintegration (Mev)

V, = average wind speed for each stability class (s), "

R = downwind distance (site boundary, m) e = sector width (radians) s = subscript for stability class

1. = I function = 1, + k12 for each stability class.

(unitless, see Regulatory Guide 1.109)

=Fraction of the attenuated energy that is actually absorbed in air (see Regulatory Guide 1.109, see below for equation)

Whole Body

- atd Vi = 1.llSFBie Where: tdj = tissue depth (g/cm 2 )

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

Where all other parameters are defined above.

'K = conversion factor = [3.7 E10 dis] 1.6 E-6 erg]

Ci-sec Mev = 0.46 L1293 q J L100 erg J g-rad 2

k =

Where: p = mass attenuation coefficient (cn 2 /g; air for B1, tissue for VI) 18 = defined above Unit 1 ODCM Revision 24 11 79 July 2003

APPENDIX B (Cont'd)

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

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

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

David C. Kocher.

The mass absorption (WE) and attenuation (fl) coefficients were calculated by multiplying the mass absorption (p.4&) and mass attenuation (Wp) coefficients given in the Radiation Health Handbook by the air density equal to 1.293 E-3 glcc or the tissue density of 1 glcc where applicable. The tissue depth is 5glcm2 for the whole body.

The downwind distance is the site boundary.

Unit 1 ODCM Revision 24 11n80 July 2003

APPENDIX B (Cont'd)

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

-DATA E = 2.22MeV k = p!:-6 = .871 K = 0.46 P. = 2.943 E-3m'1 Pa VF = 5.55 m/sec P = 5.5064E-3m'l R = 644m.

0 = 0.39 ez = 19m....... vertical plume spread taken from "Introduction to Nuclear Engineering, John R. LaMarsh

-I Function Ue. = 0.06 11 = 0.33 12 = 0.45 I = I, + kI2 = 0.33 + (0.871) (0.45) = 0.72

=

r dis. 1 0.46L.Ci-sec)(Mev/ergs. (2. 943E-3m1) (2.22Mev) (.72)

(T%) (g/m?) (ergs) (5.55 mWs) (.39) (644m)

(g-rad)

= 1.55(-6) rad/s (3600 s/hr' (24 h/d) (365 d/y) (lE3mrad/rad)

Ci/s (lE6pCi)

Ci

= 2.76(-2) mrad/yr pCi/sec -(.0253 cm2 /g) (5g/cm2 )

Vi = 1.11 (.7) .76(-2) mrad/yr [e I itCi/sec

= 1.89(-2) mrad/yr PCi/sec NOTE: The above calculation is for the F stability class only. For Table 3-2 and procedure values, a weighted fraction of each stability class was used to determine the Bi and V1 values.

Unit 1 ODCM Revision 24 1 81 July 2003

APPENDIX C ORGAN DOSE PARAMETERS FOR IODINE 131 and 133, PARTICULATES AND TRITIUM Unit 1 ODCM Revision 24 1182 July 2003

APPENDIX C ORGAN DOSE PARAMETERS FOR IODINE - 131 AND - 133, PARTICULATES AND TRITIUM This appendix contains the methodology which was used to calculate the organ dose factors for 1-131, 1-133, particulates, and tritium. The dose factor, Rj, was calculated using the methodology outlined in NUREG-0133. The radioiodine and particulate ODCM Part I (Control DLCO 3.6.15) is applicable to the location in the unrestricted area where the combination of existing pathways and receptor age groups indicates the maximum potential exposure occurs, i.e., the critical receptor. Washout was calculated and determined to be negligible. RI values have been calculated for the adult, teen, child and infant age groups for all pathways. However, for dose compliance calculations, a maximum individual is assumed that is a composite of highest dose factor of each age group for each organ and pathway. The methodology used to calculate these values follows:

C. 1 Inhalation Pathway Ri (I) = KK' (BR)a(DFA)ija where:

Rfi) - dose factor for each identified radionuclide i of the organ of interest (units = mrem/yr per p.Ci/m 3 );

K' = a constant of unit conversion, 1E6 pCi/pCi (BR)a = Breathing rate of the receptor of age group a, (units = m3 /yr);

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

The breathing rates (BR). for the various age groups, as given in Table E-5 of Regulatory Guide 1.109 Revision 1, are tabulated below.

Age Group (a) Breathing Rate (Lm 3/r)

Infant 1400 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DFA)1j. for the various age groups are given in Tables E-7 through E-10 of Regulatory Guide 1.109 Revision 1.

Unit 1 ODCM Revision 24 1 83 July 2003

APPENDIX C (Cont'd)

C.2 Ground Plane Pathway Ri(G) = K'K' '(SF) (DFG) i ,

(1-e

)

Where:

R4(G) = Dose factor for the ground plane pathway for each identified radionuclide i for the organ of interest (units = m2 -mrem/yr per pCi/sec)

K' = A constant of unit conversion, 1E6 pCi/4Ci K = A constant of unit conversion, 8760 hr/year

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

t = The exposure time, sec, 4.73E8 sec (15 years)

(DFG)I = The ground plane dose conversion factor for radionuclide i; (units = mrem/hr per pCi/m2 )

SF = The shielding factor (dimensionless)

A shielding factor of 0.7 is discussed in Table E-15 of Regulatory Guide 1.109 Revision 1. A tabulation of DFG, values is presented in Table E-6 of Regulatory Guide 1.109 Revision 1.

Unit 1 ODCM Revision 24 1 84 July 2003

APPENDIX C (Cont'd)

C.3 Grass-(Cow or Goat)-Milk Pathway

-xith -)itf Ri(C) = K'Qf(U..) Fp (r) (DFL)i+/-t f + (-fpf.) (e )e

(;i + Aj) Y_

Where:

Ri(C) = Dose factor for the cow milk or goat milk pathway, for each identified radionuclide i for the organ of interest, (units = m2-mrem/yr per pCi/sec)

K = A constant of unit conversion, 1E6 pCi4LCi Qf = The cow's or goat's feed consumption rate, (units = Kg/day-wet weight)

UP= The receptor's milk consumption rate for age group a, (units = liters/yr)

Yp= The agricultural productivity by unit area of pasture feed grass, (units = kg/m2)

Y.= The agricultural productivity by unit area of stored feed, (units = kg/m2)

Fm = The stable element transfer coefficients, (units = pCi/liter per pCi/day) r = Fraction of deposited activity retained on cow's feed grass (DFL)Iat = The ingestion dose factor for nuclide i, age group a, and total body or organ t (units = mrem/pCi)

= The radiological decay constant for radionuclide i, (units=sec -1)

= The decay constant for removal of activity on leaf and plant surfaces by weathering equal to 5.73E-7 sec -1 (corresponding to a 14 day half-life) tf = The transport time from pasture to cow or goat, to milk, to receptor, (units = sec) th = The transport time from pasture, to harvest, to cow or goat, to milk, to receptor (units = sec)

Unit 1 ODCM Revision 24 11 85 July 2003

APPENDIX C (Cont'd) fp= Fraction of the year that the cow or goat is on pasture (dimensionless) f= Fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless)

Milk cattle and goats are considered to be fed from two potential sources, pasture grass and stored feeds.

Following the development in Regulatory Guide 1.109 Revision 1, the value of f, is considered unity in lieu of site specific information. The value of fp is 0.5 based on 6 month grazing period. This value for fp was obtained from the environmental group.

Table C-I contains the appropriate values and their source in Regulatory Guide 1.109 Revision 1.

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

Therefore, the RT(C) is based on X/Q:

RT(C) = K'K FmQfUap(DFL)at 0.75(0.5/H)

Where:

RT(C) = Dose factor for the cow or goat milk pathway for tritium for the organ of interest, (units = mrem/yr per gCi/m3 )

K = A constant of unit conversion, 1E3 g/kg H = Absolute humidity of the atmosphere, (units = g/m 3 )

0.75 = The fraction of total feed that is water 0.5 = The ratio of the specific activity of the feed grass water to the atmospheric water Other values are given previously. A site specific value of H equal to 6.14 g/m 3 is used. This value was obtained from the environmental group using actual site data.

Unit 1 ODCM Revision 24 1 86 July 2003

APPENDIX C (Cont'd)

C.4 Grass-Cow-Meat Pathway R(C) = K'Qf(U) Ff(r) (DFL)+/- fPfu + ( (_e (xi + kw) LYP Y R1(M) = Dose factor for the meat ingestion pathway for radionuclide i for any organ of interest, (units = m2 -mrem/yr per jICi/sec)

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

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

All other terms remain the same as defined for the milk pathway. Table C-2 contains the values which were used in calculating R1(M).

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

RT{M) = K' KFfQfUap(DFL)iat [0.75(0.5/H)]

Where:

RT(M) = Dose factor for the meat ingestion pathway for tritium for any organ of interest, (units =

mrem/yr per ILCi/m 3)

All other terms are defined above.

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

+ {FL)*t -AitL Usa ige Ri(V) = K' r (DUL) iatiLFLe + U'F t Yv O., + km)

Unit 1 ODCM Revision 24 1 87 July 2003

APPENDIX C (Cont'd)

Where:

R1 (V) = Dose factor for vegetable pathway for radionuclide i for the organ of interest, (units = m2 -mremtyr per ACi/sec)

K' = A constant of unit conversion, 1E6 pCi/pCi UL. = The consumption rate of fresh leafy vegetation by the receptor in age group a, (units = kg/yr)

US= The consumption rate of stored vegetation by the receptor in age group a (units = kg/yr)

FL = The fraction of the annual intake of fresh leafy vegetation grown locally F5 = The fraction of the annual intake of stored vegetation grown locally tL = The average time between harvest of leafy vegetation and its consumption, (units

= sec) th = The average time between harvest of stored vegetation and its consumption, (units = sec)

Y= The vegetation areal P density, (units = kg/n 2 )

All other factors have been defined previously.

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

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

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

RT(V) = KIK [TfafL + Ua fj(DFL)iat 0.75(0.5/H)

Where:

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

All other terms are defined in preceeding sections.

Unit 1 ODCM Revision 24 11 88 July 2003

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

Qf (kgtday) 50 (cow) Table E-3 6 (goat) Table E-3 r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 (DFL)Ij. (mrem/pCi) Each radionuclide Tables E- l1to E-14 Fm (pCi/liter per pCi/day) Each stable element Table E-1 (cow)

Table E-2 (goat)

Y. (kg/n 2 ) 2.0 Table E-15 Yp (kg/M2 ) 0.7 Table E-15 th (seconds) 7.78 x 106 (90 days) Table E-15 tf (seconds) 1.73 x 105 (2 days) Table E-15 U8p(liters/yr) 330 infant Table E-5 330 child Table E5 400 teen Table E-5 310 adult Table E-5 Unit 1 ODCM Revision 24 I 89 July 2003

TABLE C-2 Parameters for the Grass-Cow-Meat Pathway Reference Parameter Value (Reg. Guide 1.109 Rev. 1) r 1.0 (radioiodines) Table E-15 0.2 (particulates) Table E-15 Ff (pCi/Kg per pCi/day) Each stable element Table E-1 Uap (Kg/yr) 0 infant Table E-5 41 child Table E-5 65 teen Table E-5 110 adult Table E-5 (DFL).j. (mrem/pCi) Each radionuclide Tables E-11 to E-14 Yp (kg/m 2 ) 0.7 Table E-15 Y. (kg/rM2 ) 2.0 Table E-15 th (seconds) 7.78E6 (90 days) Table E-15 tf (seconds) 1.73E6 (20 days) Table E-15 Qf (kg/day) 50 Table E-3 Unit 1 ODCM Revision 24 1190 July 2003

TABLE C-3 Parameters for the Vegetable Pathway Reference Parameter Value (Reg. Guide 1.109 Rev. 1) r (dimensionless) 1.0 (radioiodines) Table E-1 0.2 (particulates) Table E-1 (DFL)ja (mremlpCi) Each radionuclide Tables E-11 to E-14 UL)a (kg/yr) - infant 0 Table E-5

- child 26 Table E-5

- teen 42 Table E-5

- adult 64 Table E-5 UV)a (kg/yr) - infant 0 Table E-5

- child 520 Table E-5

- teen 630 Table E-5

- adult 520 Table E-5 tL (seconds) 8.6E4 (1 day) Table E-15 th (seconds) 5.18E6 (60 days) Table E-15 Y, (kg/m) 2.0 Table E-15 Unit 1 ODCM Revision 24 1191 July 2003

APPENDIX D DIAGRAMS OF RADIOACTIVE LIQUID AND GASEOUS EFFLUENT TREATMENT SYSTEMS AND MONITORING SYSTEMS Unit 1 ODCM Revision 24 I 92 July 2003

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ATTACHMENT 13 Process Control Program (PCP)

is NINE MILE POINT NUCLEAR STATION UNIT 1 RPCP REVISION 07 UNIT I RADWASTE PROCESS CONTROL PROGRAM TECHNICAL SPECIFICATION REQUIRED Approved by: Ii..f'-1 3 L.A. Hopkins Date THIS IS A FULL REVISION Effective Date: 12/11/2003

TABLE OF CONTENTS SECTION PAGE 1.0 PURPOSE . . . . . . . .. . . . . . . . . . . . . . . . . . . . .. 1 2.0 RESPONSIBILITIES ..... . ... .. .. ......... .. 1 3.0 PROGRAM .. . .. .. .. . . . . .. . . . . 1 4.0 RADIOACTIVE WASTES ..... . . . . .. . .......... . 3 5.0 DEFINITIONS ..... . .. .. .. . . ... . .. . . .. . . . . . 7

6.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . .. 7 ATTACHMENT 1: UNIT 1 RADWASTE PROCESS CONTROL PROGRAM IMPLEMENTING PROCEDURES ..... . . . . . . . . . . . . . . . . . . . 10 ATTACHMENT 2: SOLID WASTE SOURCES .... . . . . . . . . . . . . . . . 11 Page i RPCP Rev 07

1.0 PURPOSE To describe the methods for processing, packaging, transporting, and storing low-level radioactive waste and provide assurance of complete stabilization of various radioactive wastes in accordance with applicable NRC & DOT regulations and guidelines.

2.0 RESPONSIBILITIES 2.1 The Plant Manager is responsible for:

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

2.1.2 Reviewing and approving changes to the Unit 1 Radwaste Process Control Program in accordance with the applicable Technical Specification.

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

2.3 The Supervisor Radwaste is responsible for overall implementation of the Radwaste Process Control Program.

2.4 Oerators are responsible to process and package waste in accordance with applicable Waste Handling Procedures (WHP's).

3.0 PROGRAM

3.1 System Description

3.1.1 General

a. The Solid Waste Management System (SWMS) implemented by the procedures identified in the Unit 1 Radwaste Process Control Program Implementing Procedures (Attachment 1) collects, reduces the volume, dewaters and packages wet and dry types of radioactive waste in preparation for shipment off-site for further processing or disposal at a licensed burial site. The processing and storage methods used for interim storage are consistent with the present waste form stability requirements.

b. Types of solid waste sources are identified in Solid Waste Sources (Attachment 2).

Page 1 RPCP Rev 07

3.1.1 (Cont)

c. Bead resins, powdered resins and charcoal are dewatered using approved vendor equipment in:

1. Vendor certified polyethylene containers, or

2. Carbon steel liners, or a

3. High Integrity Container (HIC)

d. Concentrated wastes are processed off-site to dryness by an approved vendor.

e. Evaporator bottoms are transferred to a liner in the Radwaste Truck Bay for off-site processing by an approved vendor.

f. Dry solid trash is collected in the Radwaste Facility, sorted, and sent off-site for further separation and processing.

3.1.2 Ventilation Systems

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

b. The Radwaste Solidification and Storage Building (RSSB)

Ventilation System provides filtered, conditioned outside air to selected areas in the RSSB.

Recirculation fans continuously filter and condition the air, and exhaust fans, taking a suction on the truck bays, exhaust the air to the Turbine Building stack. (The system maintains the building at a pressure below atmospheric to help prevent any unmonitored air leakage to the environment.)

3.1.3 Crane

a. All liner movements are completed using a remote controlled/operated crane. The movements are facilitated by the use of remote controlled cameras and monitors.

Page 2 RPCP Rev 07

3.1.3 (Cont)

b. Liners are moved when required using a ceiling grid coordinated system for placement of the liner.

c. When liners stored in the RSSB storage area are to be shipped, the liners scheduled for shipment are moved to the East-West Truck Bay and then loaded for transportation.

4.0 RADIOACTIVE WASTES 4.1 Waste Processing System The Supervisor Radwaste shall ensure:

4.1.1 Radioactive waste is processed using approved equipment with approved procedures.

4.1.2 Radioactive waste may be processed using approved vendor equipment and procedures.

4.1.3 Radioactive wastes are disposed of in the applicable approved containers.

4.1.4 Radioactive waste is transferred into shipping casks in accordance with approved procedures.

4.1.5 Waste is transferred between units and placed in interim storage in accordance with approved procedures.

4.2 Solid Dry Radioactive Wastes (SDRW)

The Supervisor Radwaste shall ensure:

4.2.1 Low Specific Activity (LSA) Solid Dry Radioactive Waste (SDRW) is collected and prepared in accordance with the applicable procedure, meeting 10CFR61, Sub Part D, Technical Requirements for Land Disposal Facilities and Final Waste Classification and Waste Form Technical Position Papers requirements.

4.2.2 SDRW is examined for liquids or items that could compromise the integrity of the package or violate the burial site license and/or criteria. These items are removed or separated.

Page 3 RPCP Rev 07

4.2.3 SDRW` is shipped in containers meeting the transport requirements of 49CFR173.427, Transport Requirements for Low Specific Activity (LSA) Radioactive Materials.

4.2.4 Waste precluded from disposal in LSA boxes or drums, due to radiation limits, is disposed of in the applicable containers.

4.2.5 Waste segregation and volume reduction processing techniques are used for waste generated during operation, maintenance, and modifications.

4.2.6 Scrap metal is separated from waste, when possible, for on-site or off-site decontamination.

NOTE: Vendor services may be used for waste segregation and further volume reduction processes.

4.2.7 Waste is placed in interim storage in accordance with approved procedures.

4.3 Waste Classification/Characterization 4.3.1 The Supervisor Radwaste shall ensure:

a. The minimum waste classification/characteristic requirements identified in IOCFR61.56, Waste Characteristics, are satisfied.

b. The radionuclide concentration determination methods and frequency are conducted in accordance with approved procedures.

4.3.2 The General Supervisor Chemistry shall ensure the chemical and radionuclide content of waste is determined in accordance with the applicable Chemistry procedures.

4.3.1 The Manager Radiation Protection shall ensure classification of waste is performed in accordance with approved procedures.

4.4. Administrative Controls 4.4.1 The Supervisor Radwaste is responsible for overall administrative control of the Radwaste Process Control Program, ensuring:

Page 4 RPCP Rev 07

4.4.1 (C~ont)

( a. Changes to the Process Control Program (PCP:) shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the change(s) was made. The submittal shall contain information as described in the Offsite Dose Calculation Manual (ODCM) section D 6.9.1.e, NReporting Requirements".

b. Shipping manifests are completed and tracked to satisfy the requirements of 10CFR20.2006, Transfer for Disposal and Manifests, in accordance with Waste Handling Procedures.

c. Temporary storage of solid radioactive material awaiting shipment in an area other than a designated storage area is done in accordance with the applicable radioactive material storage procedures.

d. Interim storage of low level waste is performed in accordance with approved procedures.

4.4.2 The Nuclear Division Quality Assurance Program assures effective implementation of the Process Control Program, as follows:

NOTE: The Manager, Nuclear QA, Operations has the authority to stop work when significant conditions adverse to quality exist and require corrective action.

a. Under the cognizance of the Safety Review and Audit Board (SRAB), the Process Control Program and implementing procedures for processing and packaging of radioactive waste are audited at least once every 24 months as required by the UFSAR Section B.2.2.16.

b. QA audits waste classification records to ensure compliance with 10CFR20.2006, Transfer for Disposal and Manifests.

c. QA Inspectors performing Radwaste inspections receive training in Department of Transportation and NRC Radwaste Regulatory requirements.

d. Management reviews results of QA audits.

Page 5 RPCP Rev 07

4.4.3 The Nuclear Division Training Program assures personnel responsible for Implementation of the Process Control Program are effectively trained in accordance with the applicable training procedures as follows:

a. Qualification as a Radwaste Operator requires satisfactory completion of the Radwaste Operations Unit 1 Initial Training Program and participation in continued training. This includes:

1. Demonstrating an acceptable level of skill and familiarity associated with Radwaste operations by achieving an average grade of 80 percent or above on written examinations.

2. Receiving on-the-job training in accordance with applicable training procedures.

3. Continued training conducted on a cyclical basis and includes a fundamental review of system modifications, revisions or changes to procedures, and changes or experiences in the nuclear industry.

4. Individuals that demonstrate a significant deficiency in a given area of knowledge and/or proficiency (as identified during continued training) are placed in a remedial training program as directed by approved training procedures.

4.4.4 Training records and Waste Management records are maintained in accordance with applicable Quality Assurance procedures.

4.4.5 Solid Radioactive Waste Specification

a. This Specification implements the requirements of 10CFR part 50.36a and General Design Criteria 60 Of Appendix A to 10CFR part 50. The process parameters included in establishing the process control program may include, but are not limited to waste type, waste pH, waste/liquid/solidification agent/catalyst ratios, waste oil content, waste principal chemical constituents and mixing and curing times.

b. The solid radwaste system shall be used in accordance with the Process Control Program to process wet radioactive wastes to meet shipping and burial ground requirements.

c. With the provisions of the process control program not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

Page 6 RPCP Rev 07

-4.4.5 (Cont)

d. The process control program shall be used to verify the solidification of at least one representative test specimen from at least every tenth batch of each type of wet radioactive waste (e.g., filter sludges and evaporator bottoms).

1. If any test specimen fails to verify solidification, the solidification of the batch may then be resumed using the alternative solidification parameters determined by the process control program.

2. If the initial test specimen from a batch of waste fails to verify solidification, the process control program shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive-initial test specimens demonstrate solidification.

5.0 DEFINITIONS 5.1 The applicable Radwaste packaging, processing, and transportation definitions will be used in accordance with 49CFR171 and 49CFR Sub Part I.

5.2 Solidification Solidification shall be the conversion of wet or liquid waste into a form that meets shipping and burial ground requirements.

6.0 REFERENCES

6.1 Licensee Documentation 6.1.1 Unit 1 Technical Specifications, Section 6.6.3, Radioactive Effluent Release Report.

6.1.2 Unit 1 Offsite Dose Calculation Manual (ODCM).

6.1.3 Nine Mile Point Unit 1 Operating License No. DPR-63 (Docket No. 50-220) 6.1.4 QATR-1, Quality Assurance Program Topical Report for Nine Mile Point Nuclear Station Operations, Section 17.0, Quality Assurance Records Page 7 RPCP Rev 07

6.1.5 UFSAR,Section XII.A, Radioactive Wastes

( 6.1.6 UFSAR, Section 111.1, RSSB 6.1.7 Safety Evaluation 92-049, Rev. 04, Interim Storage 6.1.8 Offsite Dose Calculation Manual (ODCM) section D 6.9.1.e, Reporting Requirements 6.2 Standards. Regulations, and Codes 6.2.1 10CFR20, Standards for Protection Against Radiation 6.2.2 10CFR61, Sub Part D, Technical Requirements for Land Disposal Facilities and Final Waste Classification and Waste Form Technical Position Papers 6.2.3 10CFR61.55, Waste Classification 6.2.4 10CFR61.56, Waste Characteristics 6.2.5 10CFR71, Packaging and Transportation of Radioactive Material, (Refer to applicable S-RPIPs for the packaging and transportation of radioactive material) 6.2.6 49CFR173, Shippers - General Requirements for Shipment and Packagings, (Refer to applicable S-RPIPs for the packaging

( and transportation of radioactive material) 6.2.7 49CFR173.427, Transport Requirements for Low Specific Activity (LSA) Radioactive Materials 6.2.8 NUREG-0133, Section 3.5, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants 6.2.9 NUREG-0473, Sections 3.11.3 and 6.14, Draft Radiological Effluent Technical Specifications for Boiling Water Reactors 6.2.10 NUREG-0800, Section 11.4, Standard Review Plan for Solid Waste Management Systems 6.3 Policies. Programs, and Procedures 6.3.1 NDD-LPP, Licenses, Plans, and Programs 6.3.2 NDD-OPS, Operations 6.3.3 NDD-RMP, Radioactive Material Processing, Transport, and Disposal 6.3.4 NIP-ECA-01, Deviation/Event Report 6.3.5 NIP-PRO-03,-Preparation and Review of Technical Procedures Page 8 RPCP Rev 07

6.3.6 NIP-RMG-01, Records Management

( 6.3.7 NIP-TQS-O1, Qualification and Certification 6.3.8 GAP-ALA-01, Site ALARA Program 6.3.9 GAP-INV-02, Control of Material Storage Areas 6.3.10 GAP-OPS-01, Administration of Operations 6.3.11 GAP-RPP-01, Radiation Protection Program 6.3.12 GAP-RPP-02, Radiation Work Permit 6.3.13 GAP-RMP-O1, Interim Storage of Low-Level Radioactive Waste 6.4 Supplemental References 6.4.1 Vendor Training and Requalification Procedure 6.4.2 Nuclear Regulatory Commission's Branch Technical Position of Waste Classification and Waste Form, May 1983 6.4.3 DER 1-94-0549 6.4.4 Structural Calculation S.2.3-R5252-Tank 01 6.4.5 Modification N1-91-033 6.4.6 Procedure N1-MFT-30 Page 9 RPCP Rev 07

ATTACHMENT 1: UNIT 1 RADWASTE PROCESS CONTROL PROGRAM IMPLEMENTING PROCEDURES Waste Handling Procedures (Nl-WHPs and S-WHPs)

Liouid Waste Processing Procedures (NI-LWPPs)

Radiation Protection Procedures (S-RPIPs)

Chemistry Technical Procedures (NI-CTPs)

Quality Assurance Audit and Surveillance Procedures (QAPs)

Nuclear Training Procedures (NTPs)

Generation Administrative Procedures (GAPs)

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Page 10 RPCP Rev 07

ATTACHMENT 2: SOLID WASTE SOURCES (Sheet I of 3) 1.0 RADWASTE FILTERS 1.1 Mechanical Radwaste filters filter resin and crud (backwash material) from the Waste Collector Sub-System.

1.2 When a filter reaches a predetermined differential pressure, the filter Is backwashed into the filter sludge tank, which is then processed via the clarifier to the thickener tanks.

2.0 RADWASTE DEMINERALIZER 2.1 The Radwaste Demineralizer is used as an ionic exchange media for processing high quality water from the Waste Collector Tanks.

2.2 When determined the resin can no longer be used, the depleted resin is transferred to the Spent Resin Tank.

3.0 CONDENSATE DEMINERALIZERS 3.1 The Condensate Demineralizers remove soluble and Insoluble impurities from the condensate water to maintain reactor feedwater purity.

3.2 After It is determined these resins can no longer be used, the depleted resin are transferred to the Radwaste Demineralizer or Spent Resin Tank.

4.0 THERMEX SYSTEM 4.1 Concentrate will be pumped to the Spent Resin Tank and dewatered or stored in a liner and eventually pumped to a transport liner in the Radwaste Truck Bay for off-site processing.

4.2 Exhausted resin and charcoal are either:

a. transferred to the Spent Resin Tank, mixed to a homogenous mixture and then transferred to a liner in the truck bay for dewatering, or

b. transfered to a liner in the Truckbay.

4.3 Exhausted Reverse Osmosis membranes will be processed as DAW.

5.0 FUEL POOL FILTER SLUDGE TANK This tank receives the exhausted powdered filter media (resins) from the Fuel Pool Cleanup System, which is subsequently pumped to the

( Filter Sludge Tank for processing.

Page 11 RPCP Rev 07

ATTACHMENT 2 (Co~nt)

(Sheet 2 of 3) f 6.0 CLEANUP FILTER SLUDGE TANK This tank receives the exhausted powdered filter media (resins) from the Reactor Cleanup System, which is subsequently pumped to the Filter Sludge Tank, Clarifier, or directly to a liner in the Radwaste Truck Bay for processing.

7.0 FILTER SLUDGE STORAGE TANK This tank receives waste from the Radwaste filters, Fuel Pool and Cleanup Sludge Tanks, Clarifier and Thickener Tank overflows, and Radwaste Floor Drain Sump #11. Tank discharge is to the Clarifier (Filter Sludge Thickener System) or directly to a liner in the Radwaste Truck Bay for processing.

8.0 FILTER SLUDGE THICKENER TANKS (CLARIFIER)

Waste from the Filter Sludge Storage Tank or the Cleanup Filter Sludge Tank is pumped to the Clarifier, mixf?d with a flocculent and drained to the Thickener Tanks. The Thickens?r Tanks are pumped to a liner in the Radwaste Truck Bay for processing 9.0 SPENT RESIN STORAGE TANK Exhausted resin from the Condensate Demineralizers, Radwaste Demineralizer, Cleanup Demineralizer, and THERMEX System are

- transferred to the Spent Resin Tank. The tank is subsequently pumped to a liner in the Radwaste Truck Bay for dewatering and further processing.

10.0 CONTAMINATED OIL Oil from sources within Unit 1 that becomes contaminated is stored in containers to be shipped off-site for processing and/or disposal.

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Page 12 RPCP Rev 07

ATTACHMENT 2 (Cont)

(Sheet 3 of 3)

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11.0 COMPACTIBLE SOLIDS 11.1 Compactible low level trash is shipped off-site for vendor separation and processing.

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

13.0 WASTE CONCENTRATOR 13.1 The Waste Concentrator processes low quality waste from the Floor Drain Collector System.

13.2 The Waste Concentrator is designed to concentrate waste to a 25% solid concentration, which may then be discharged to the #13 Concentrated Waste Tank for transfer to the Radwaste Truck Bay for vendor processing.

Page 13 RPCP Rev 07

ML041320372

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ML041320372

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SUBJECT:

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