Text

Part 1 of 2, R.E. Ginna Nuclear Power Plant - Annual Radioactive Effluent & Environmental Reports
	ML021430275
Person / Time
Site:	Ginna
Issue date:	05/15/2002
From:	Mecredy R; Rochester Gas & Electric Corp
To:	Clark R; NRC/NRR/DLPM/LPD1
References
	Download: ML021430275 (173)
	v • d • e

R fIN ' A Subsidiary of RGS Energy Group, Inc.

www.rge.com ROCHESTER GAS AND ELECTRIC CORPORATION ° 89 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 ° 776 546-2700 ROBERT C. MECREDY May 15, 2002 Vice President Nuclear Operations Mr. Robert Clark Project Directorate I USNRC Washington, DC 20555

Subject:

Annual Radioactive Effluent and Environmental Reports R.E. Ginna Nuclear Power Plant Docket No. 50-244

Dear Sirs:

This Annual Radioactive Effluent Release Report is being submitted in accordance with the requirements of Technical Specification Section 5.6.3 and 10CFR 50.36.

The Annual Radiological Environmental Operating Report is being submitted in accordance with the requirements of Technical Specification 5.6.2.

Very truly yours, Robert C. Mecredy xc: Mr. Robert Clark (Mail Stop O-8C2)

USNRC One White Flint North 11555 Rockville Pike Rockville, MD 20852 Regional Administrator - Region 1 USNRC 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector z-/&

Jay Dunkleberger State Emergency Management Office Building #22, First Sublevel State Campus Albany, NY 12226-5000 American Nuclear Insurers MAELU Suite 300S 29 S. Main Street West Hartford, CT 06107-2445 Attn: Thomas Wolff INPO 700 Galleria Parkway Atlanta, GA 30339-5957

Rochester Gas and Electric Corporation Inter-Office Correspondence January 24, 2001

Subject:

Offsite Dose Calculation Manual (ODCM) Revision 15 To: Distribution Attached is the Ginna Station Offsite Dose Calculation Manual (ODCM).

The ODCM has historically been controlled as an attachment to procedure CHA RETS-ODCM. To increase the visibility of this document it is being relocated to the Technical Specifications Supporting Documentation, Volume III.

The ODCM Revision 15 was made effective January 22, 2001 as part of CHA RETS-ODCM Revision 10. The ODCM will retain its current revision level and effective date, as there are no changes to the document as the result of the relocation.

Instructions for the necessary changes to the ITS for your controlled copy of the ODCM are as follows:

Volume Section Remove Insert II Programs Page 1 Page 1 Inl Programs Page 2 Page 2 inl ODCM -- ODCM Rev. 15 Please contact Tom Harding (extension 3384) if you have any questions.

Thomas L. Harding

.W

Program Documents and Locations Subject Reportability Matter Program Requirement TS # Description Expert Document(s) Location(s) 5.2.2.d Overtime Policy Program J. Widay IP-HPE-3 N/A 5.2.2.f STA Training Program S. Carter TR-C. 11 N/A 5.5.1 Offsite dose Calculation Manual F. Mis ODCM A-25.6 (ODCM) G. Jones 5.5.2 Primary Coolant Sources Outside G. Joss PT-39 PT-39 Containment Program 5.5.3 Post Accident Sampling Program G. Jones CHA-PASP N/A 5.5.4 Radioactive Effluent Controls F. Mis ODCM A-25.6 Program G. Jones 5.5.5 Component Cyclic or Transient P. Bamford RE-50.1 RE-50.1 Limit Program 5.5.6 Pre-Stressed Concrete Containment G. Joss PT-27.2 SR 3.6.1.2 Tendon Surveillance Program L. Sucheski 5.5.7 Inservice Testing (IST) Program G. Joss IST Program SR 3.4.10.1 G. Hermes Document SR 3.5.2.4 SR 3.5.3.1 SR 3.6.3.4 SR 3.6.6.7 SR 3.7.2.1 SR 3.7.2.2 SR 3.7.3.2 SR 3.7.5.2 SR 3.7.5.3 SR 3.7.5.4 SR 3.7.7.2 May also impact other TS components.

5.5.8 Steam Generator Tube Surveillance F. Klepacki ISI Program SR 3.4.13.1 Program Document 5.5.9 Secondary Water Chemistry B. Dahl CHA-SEC-GUIDE N/A Program 5.5.10 Ventilation Filter Testing Program G. Joss A-1040 SR 3.6.6.10 (VFTP) SR 3.6.6.11 SR 3.7.9.2 SR 3.7.10.3 5.5.11 Explosive Gas and Storage Tank G. Jones CHA-EGSTRM A-25.6 Radioactivity Monitoring Program

-I- Revision 2

Subject Reportability Matter Program Requirement TS # Description Expert Document(s) Location(s) 5.5.12 Diesel Fuel Oil Testing Program B. Dali CHA-DFOTP SR 3.8.3.2 5.5.13 Technical Specification Bases T. Harding IP-LPC-2 N/A Control Program 5.5.14 Safety Function Determination P.Sidelinger A-52.3 LCO 3.0.3 Program (SFDP) M. Flaherty 5.5.15 Containment Leakage Rate Testing G. Joss PT-22 SR 3.6.1.1 Program PT-23 SR 3.6.2.1 I- _SR 3.6.3.5 Revision 2

ODCM R.E. Ginna Nuclear Power Plant Offsite Dose Calculation Manual ODCM Revision 15 Responsible Manager:

PORC Review Date: //i/qlKzool Effective Date: 1*2.1 o/2 IL Controlled Copy No.

TABLE OF CONTENTS

................................................................. I DEFINITIO NS 1.0 LIQUID EFFLUENTS ...................................................... 5 1.1 C oncentration ...................................................... 5 1.2 Dose (10 CFR 50 Appendix I) ............................................. 6 1.3 D ose (40 CFR Part 190) ................................................ 7 1.4 Liquid Effluents Release Points ........................................ 12 1.5 Liquid Effluents Monitor Setpoints ...................................... 12 1.6 Liquid Effluent Release Concentrations .................................. 16 1.7 Liquid Effl uent Dose ................................................. 17 2.0 GASEOUS EFFLUENTS ................................................... 20 2.1 C oncentration ..................................................... 20 2.2 Dose R ate ........................................................ 20 2.3 Dose (10 CFR Part 50, Appendix I) ......................................... 22 2.4 Dose (40 CFR Part 190) ............................................... 25 2.5 Gaseous Effluent Release Points ...................................... 29 2.6 Gaseous Effluent Monitor Setpoints ..................................... 29 2.7 Gaseous Effluent Dose Rate .......................................... 35 2.8 Gaseous Effluent Doses ............................................. 37 3.0 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION ................... 42 3.1 Liquid Effluent Monitors ............................................. 42 3.2 Gaseous Effluent Monitors ............................................ 45 3.3 Radiation Accident Monitoring Instrumentation ............................ 49 3.4 Area Radiation Monitors .............................................. 52 4.0 RADWASTE TREATMENT ................................................. 53 4.1 Liquid Radwaste Treatment ........................................... 53 4.2 Gaseous W aste Treatment ........................................... 54 4.3 Solid Radioactive Waste ............................................. 55 4.4 Configuration C hanges .............................................. 55 4.5 Process Control Program ............................................ 56 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING ............................. 56 5.1 Monitoring Program ................................................. 56 5.2 Environmental Monitor Sample Locations ................................ 62 5.3 Land Use Census ................................................ 73 5.4 Interlaboratory Comparison Program .................................... 74 6.0 REPORTING REQUIREMENTS ........................................... 75 6.1 Annual Radiological Environmental Operating Report ....................... 75 6.2 Radioactive Effluent Release Report .................................... 76 6.3. S pecial R eports .................................................... 77

7.0 REFERENCES

.......................................................... 79 i

LIST OF TABLES Table 1-1 Radioactive Liquid Waste Sampling and Analysis Program ........................ 9 Table 2-1 Radioactive Gaseous Waste Sampling and Analysis Program .................... 27 Table 2-2 Meteorological Data and Locations of Receptors for Set Point Calculations ........ 34 Table 2-3 Dose Factors For Noble Gases and Daughters ................................. 39 Table 2-4 Dose Parameters for Radionuclides and Radioactive Particulate, Gaseous Effluents 40 Table 2-5 Pathway Dose Factors Due to Radionuclides Other Than Noble Gases * ........... 41 Table 3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation ......................... 43 Table 3.1-2 Radioactive Liquid Effluent Monitoring Surveillance Requirements ................ 44 Table 3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation ....................... 46 Table 3.2-2 Radioactive Gaseous Effluent Monitoring Surveillance Requirements ............. 48 Table 3.3-1 Radiation Accident Monitoring Instrumentation ................................ 50 Table 3.3-2 Radioactive Accident Monitoring Surveillance Requirements ..................... 51 Table 5-1 Radiological Environmental Monitoring Program ............................... 59 Table 5-2 Direction and Distance to Sample Points ...................................... 63 Table 5-3 Detection Capabilities for Environmental Sample Analysis ....................... 68 Table 5-4 Reporting Levels for Radioactivity Concentrations in Environmental Samples ...... 69 Table 5-5 D/Q and X/Q 5 Year Rolling Average 1995 - 1999 - Plant Vent ..................... 70 Table 5-6 D/Q and X/Q 5 Year Rolling Average 1995 - 1999 - Containment Vent ............... 71 Table 5-7 D/Q and XIQ 5 Year Rolling Average 1995 - 1999 - Air Ejector ..................... 72 Table 6-1 Environmental Radiological Monitoring Program Summary ...................... 78 ii

LIST OF FIGURES 64 Figure 5-1 Location of Onsite Air Monitors and Post Accident TLDs ........................

Location of Farms for Milk Samples and Ontario Water District Intake ............. 65 Figure 5-2 Location of Offsite TLDs .................................................... 66 Figure 5-3 Location of Offsite Air Monitors .............................................. 67 Figure 5-4 iii

DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these controls.

Terms used in these controls and not defined herein have the same definition as listed in the Technical specifications and/or the Technical Reference Manual. If a conflict in definition exists, the definition in the Technical Specifications takes precedence.

ACTION ACTION shall be that part of a Control that prescribes required actions to be taken under designated conditions, within specified completion times.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock display, and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.

This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, 1977.

ODCM Page 1 Revision 1 5

FREQUENCY NOTATION shall correspond The FREQUENCY NOTATION specified for the performance of Surveillance Requirements to the intervals defined as follows:

NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N/A Not applicable P Completed prior to each release FUNCTIONAL TEST sensor as An FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the The FUNCTIONAL TEST practicable to verify OPERABILITY of alarm, interlock display and/or trip functions.

and/or Trip Setpoints such that the shall include adjustments, as necessary, of the alarm, interlock display setpoints are within the required range and accuracy.

LOWER LIMIT OF DETECTION The LOWER LIMIT OF DETECTION (LLD) is the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% probability with only 5%

probability of falsely concluding that a blank observation represents a "real" signal. The LLD is defined as a priori(before the fact) limit representing the capability of a measurement system and not as a posteriori(after the fact) limit for a particular measurement, the minimum detectable activity (MDA).

MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant.

This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program (REMP). The ODCM shall also contain descriptions of the Radioactive Effluent Controls and Radiological Environmental Monitoring Program and descriptions of the information that shall be included in the Annual Radiological Environmental Operating Report and the Annual Radioactive Effluent Release Report.

ODCM Page 2 Revision 1 5

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s) and when all necessary attendant instrumentation, controls, electric power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1-1 of Technical Specifications.

PURGE - PURGING PURGE or PURGING shall be any 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.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

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.

SURVEILLANCE REQUIREMENT SURVEILLANCE REQUIREMENTS shall be met during the OPERATIONAL MODES or other conditions specified for individual CONTROLS unless otherwise stated in an individual SURVEILLANCE REQUIREMENT. Each SURVEILLANCE REQUIREMENT shall be performed within the specified time interval with:

1. A maximum allowable extension not to exceed 25% of the surveillance interval, but

2. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

ODCM Page 3 Revision 15

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

VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, 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.

WASTE GAS HOLDUP SYSTEM A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

ODCM Page 4 Revision 1 5

1.0 RADIOACTIVE LIQUID EFFLUENTS 1.1 Concentration (10 CFR 20)

CONTROLS C.1.1 The release of radioactive liquid effluents shall be such that the concentration in the circulating water discharge does not exceed ten times the concentration values specified in Appendix B, Table 2, Column 2 to 10 CFR Part 20.1001 - 20.2402. For dissolved or entrained noble gases, the total activity due to dissolved or entrained noble gases shall not exceed 2 E-04 pCi/ml.

APPLICABILITY: At all times.

ACTION: If the concentration of radioactive material in the circulating water discharge exceeds ten times the concentration values of Appendix B, Table 2, Column 2 of 10CFR20, measures shall be initiated to restore the concentration to within these limits as soon as practicable.

ACTION: If the concentration when averaged over one hour exceeds twenty times the applicable concentrations specified in Appendix B of 10CFR Part 20, Table 2, Column 2, at the point of entry to receiving waters, submit to the commission a special report within 30 days.

SURVEILLANCE REQUIREMENTS S.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 1-1. The results of pre-release analyses shall be used with the calculational methods in Section 1.6 to assure that the concentration at the point of release is limited to the values in C.1.1 BASES This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section II.A design objectives of Appendix 1,10 CFR 50, to a MEMBER OF THE PUBLIC, and (2) the limits of Appendix I, 10 CFR 20, 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 MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

ODCM Page 5 Revision 15

1.2 Dose (10 CFR 50 Appendix I)

CONTROLS in liquid C.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials effluents released to UNRESTRICTED AREAS shall be limited:

and

1. during any calendar quarter to < 1.5 mrem to the total body and to < 5 mrem to any organ,

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

APPLICABILITY: At all times.

ACTION: Whenever the calculated dose resulting from the release of radioactive materials in liquid effluents exceeds the quarterly limits of C.1.2.1, a Special Report shall be submitted to the Commission within thirty days which includes the following information:

1. identification of the cause for exceeding the dose limit;

2. corrective actions taken and/or to be taken to reduce the releases of radioactive material in liquid effluents to assure that subsequent releases will remain within the above limits;

3. The results of the radiological analyses of the nearest public drinking water source, and an evaluation of the radiological impact due to licensee releases on finished drinking water with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.

ACTION: During any month when the calculated dose to a MEMBER OF THE PUBLIC exceeds 1/48 the annual limit (0.06 mrem to the total body or 0.2 mrem to any organ), projected cumulative dose contributions from liquid effluents shall be determined for that month and at least once every 31 days for the next 3 months.

SURVEILLANCE REQUIREMENTS S.1.2.1 Post-release analyses of samples composited from batch releases shall be performed in accordance with Table 1-1. The results of the post-release analyses shall be used with the calculational methods in Section 1.6 to assure that the dose commitments from liquids are limited to the values in C.1.2 BASES This control is provided to implement the requirements of Sections II.A, Ill.A, and IV.A of Appendix I, 10 CFR 50. This control implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV A of Appendix I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as reasonably achievable". Also, with Lake Ontario drinking water supplies potentially affected by plant operations, there is reasonable assurance that the operation of the plant will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CR 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational models and data, such that the actual exposure of a MEMBER OF THE PUBLIC appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor ODCM Page 6 Revision 1 5

Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I", Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", April 1977.

1.3 Dose (40 CFR Part 190)

CONTROLS C.1.3 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

APPLICABILITY: At all times.

ACTION: With the calculated doses from the release of radioactive materials in liquid effluents exceeding twice the limits of C.1.2, prepare and submit to the Commission within 30 days a Special Report that defines the corrective actions to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and include scheduling for achieving conformance with the above limits. Calculations which include direct radiation contributions from the unit and from any radwaste storage shall be performed to determine total dose to a member of the public. This Special Report, as defined in 10 CFR 20.405(c) 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 releases covered by this report. 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 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190.

Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

SURVEILLANCE REQUIREMENTS S.1.3.1 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with SURVEILLANCE REQUIREMENTS S.1.2.1 at least once every 31 days, in accordance with the methodology and parameters of Section 1.7 of the ODCM.

S.1.3.2 Cumulative dose contributions from direct radiation from the unit and from radwaste storage shall be determined from environmental thermoluminescent dosimeter (TLD) data at least quarterly.

ODCM Page 7 Revision 15

BASES been incorporated into This control is provided to meet the dose limitations of 40 CFR 190 that have requires the preparation and submittal of a Special 10 CFR 20 by 46FR18525. The specification due to releases of radioactivity and to radiation from uranium Report whenever the calculated doses to the whole body or any organ, except the thyroid, which shall fuel cycle sources exceed 25 mrems to 75 mrems. It is highly unlikely that the resultant dose to a MEMBER be limited to less than or equal within twice the dose OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the plant remains The Special Report will design objectives of Appendix I, and if direct radiation doses are kept small.

dose to a MEMBER OF describe a course of action that should result in the limitation of the annual it may be THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, negligible. If the dose to assumed that the dose contributions from other uranium fuel cycle sources is of 40 CFR 190, the Special any MEMBER OF THE PUBLIC is estimated to exceed the requirements resulting in violation of 40 CFR Report with a request for a variance, (provided the release conditions provisions of 40 CR 190.11 and 10 190 have not already been corrected), in accordance with the he requirements of 40 CFR 190 until CFR20.405c, is considered to be a timely request and fulfils not NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does dose limitation of 10 CFR 20, as addressed in C.1.1 apply in any way to the other requirements for and C.2.2. 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.

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Table 1-1 Radioactive Liquid Waste Sampling and Analysis Program Page 1 of 3 Liquid Release Sampling Minimum Type of Lower Limit Type Frequency Analysis Activity Analysis of Detection Frequency (LLD) (AzCi/ml)

(a)

Batch Release (b)

P P Principal Gamma Emitters 5 E-07 Each Batch Each Batch (d) and 1-131 1 E-06 or Gross Beta-gamma

5 E-07 Batch Waste Release Tanks P M Dissolved and Entrained One Batch/M 1 E-05 Gases (Gamma Emitters)

P M H-3 1 E-05 Each Batch Composite (c) Gross Alpha 1 E-07 P Q Sr-89 Sr-90 5 E-08 Each Batch Composite (c) Fe-55 1 E-06 Continuous Release (e)

Retention Tank Continuous W Principal Gamma Emitters 5 E-07 Composite (c) (d) 1-131 1 E-06 Service Water Gross (CV Fan Cooler and Continuous M or S **Grab Beta-gamma 5 E-07 SFP HX lines)IIII If gross beta is performed for batch releases, then a weekly composite shall also be analyzed for Principal Gamma Emitters and 1-131.

Service water samples shall be taken and analyzed once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ifalarm setpoint is reached on continuous monitor.

ODCM Page 9 Revision 15

Table 1-1 (continued) Page 2 of 3 Radioactive Liquid Waste Sampling and Analysis Program Table Notation net (a) The LLD is the smallest concentration of radioactive material in a sample that will yield a count above system background that will be detected with 95% probability with only 5%

probability of falsely concluding that a blank observation represents a "real" signal. The LLD is defined as an a priori before the fact) limit representing the capability of a measurement system and not as an a posterioriafter the fact) limit for a particular measurement, the minimum detectable activity (MDA).

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

4 .6 6 )(Sb)

LLD =(

(Y)(E)(V)(2.22 E+06)[exp(-X)]

Where: LLD is the lower limit of detection as defined above as UCi 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 disintegration V is the sample size in units of mass or volume E is the counting efficiency Y is the fractional radiochemical yield when applicable 2.22 E+06 is the number of disintegrations per minute per 4Ci Xis the decay constant t is time elapsed since sample time The value of Sb used in the calculation of the LLD for a particular measurement system shall be based on the actual observed variance of the background counting rate or the counting rate of the blank samples, as appropriate, rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contribution of other radionuclides normally present in the samples. Typical values of E, V, and Y should be used in the calculation.

The background count rate is calculated from the background counts that are determined to be within +/- one FWHM energy band about the energy of the gamma ray peak used for the quantitative analysis for this radionuclide.

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. When circumstances result in LLDs higher than required, the reasons shall be documented in the Annual Radioactive Effluent Report.

ODCM Page 10 Revision 1 5

Table 1-1 (continued) Page 3 of 3 Radioactive Liquid Waste Sampling and Analysis Program (b) A batch release is the discharge of liquid wastes of a discrete volume.

(c) A composite sample is one is which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. Decay corrections are calculated from the midpoint of the sampling period.

(d) The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides:

Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-134, Cs-137 and Ce-141.

This list does not mean that only these nuclides are to be detected and reported. Other nuclides which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should be reported as less than the LLD and should not be reported as being present at the LLD level.

The less than values should not be used in the required dose calculations. When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Annual Radioactive Effluent Release Report.

(e) A continuous release is the discharge of liquid wastes of a non-discrete volume; e.g. from a volume or system that has an input flow during the continuous release. Decay corrections will be calculated based on all samples collected during the release-ODCM Page 11 Revision 15

1.4 Liquid Effluents Release Points There are three normal release points for liquid radioactive effluents from the plant that empty into the discharge canal. These are the Radwaste Treatment Discharge, Retention Tank discharge and the All Volatile Treatment Discharge. Each of these is a monitored release line that can be isolated before the release reaches the discharge canal. There is also a release point for the service water lines used for cooling the heat exchangers that is a monitored release line but is not isolatable. If there is an alarm on the service water monitor, it is necessary to sample each heat exchanger separately to determine which has a leak and then isolate the affected heat exchanger. The pressure of the service water system flow would normally force water from the clean service water side into the contaminated side of the heat exchanger. Dilution of liquid effluent is provided by the discharge canal. The discharge canal flow is 1.7 E+05 gpm for each circulating water pump. During operating periods, two circulating water pumps are in operation. During shutdown periods, one circulating water pump is operated. If neither circulating water pump is operable, dilution is provided by operation of one to three service water pumps which provide 5.3E+03 gpm each.

1.5 Liquid Effluents Monitor Setpoints Alarm and/or trip setpoints for radiation monitors on each liquid effluent line are required. Precautions, limitations and setpoints applicable to the operation of Ginna Station liquid effluent monitors are provided in plant procedures P-9 and CH-RETS-RMS. Setpoint values are calculated to assure that alarm and trip actions occur prior to exceeding ten times the effluent concentration of Appendix B, Table 2, Column 2 of 10 CFR 20.1001 - 20.2402 at the release point to the unrestricted area.

For added conservatism, liquid effluent release rates are administratively set so that only fractions of the applicable maximum effluent concentrations can be reached in the discharge canal.

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The Calculated alarm and trip action setpoints for each radioactive liquid effluent line monitor and flow determination must satisfy the following equation:

Equation(1): ef < C F+f Where: C = the effluent concentration which implements ten times 10 CFR 20 limit for unrestricted areas, in /Ci/ml.

c = the setpoint of the radioactivity monitor measuring the radioactivity concentration in the discharge line prior to dilution and subsequent release, in gCi/ml.

F= the dilution water flow as determined prior to the release point, in volume per unit time.

f = the liquid waste flow as measured at the discharge point, in volume per unit time, in the same units as F.

Liquid effluent batch releases from Ginna Station are discharged through a liquid waste disposal monitor. The liquid waste stream (f) is diluted by (F) in the plant discharge canal before it enters Lake Ontario.

The limiting batch release concentration (c) corresponding to the liquid waste monitor setpoint is calculated from the above expression. Since the value of (f) is very small in comparison to (F), and tritium can not be accounted for, the expression becomes:

CF Equation(2): c CF x 0.4 (1-TCF) f ODCM Page 1 3 Revision 1 5

Where: C = 10 x the allowable concentration of Cs-1 37 as given in Appendix B, Table 2, Column 2 of 10 CFR 20, 1x10 5 . This value is normally more restrictive than the calculated mixed isotopic release concentration. A weighted average, excluding Sb-125, from 1998 data indicated a release concentration of 1.15 x 10i 4Ci/ml. (See DA-RP-99-078). This should be reviewed as an annual basis, and the more conservative value between the two be utilized.

F = the dilution flow assuming operation of only 1 circulating water pump (170,000 gpm).

c= The limiting batch release concentration corresponding to the liquid monitor setpoint f = the maximum waste effluent discharge rate through the designated pathway.

0.4 = A conservatism based on the possibility of 2 liquid discharges occurring simultaneously, minus 0.2 (Total Instrumental Uncertainty) eg: (limit - 0.2 TIU/2)

TCF = Tritium Correction Factor, based on concentration of diluted RCS H-3, divided by H-3 E.C.L. The maximum UFSAR value is 7.5 AzCi/ml. The monitor setpoint against gamma emitters must be reduced proportional to the ratio of tritium in the waste stream, to which the monitor will not respond.

The limiting release concentration (c) is then converted to a set-point count rate by the use of the monitor calibration factor determined per procedure CH-RETS-RMS. The expression becomes:

c (uCi/ml)

Equation (3): Setpoint (cpm) =

Cal Factor (uCi/ml/cpm)

Example (Liquid Radwaste Monitor R-18):

Assuming, for example, that the maximum pump effluent discharge rate (f) is 90 gpm and the RCS tritium concentration is 3 tCi/ml, then the limiting batch release concentration (c) would be determined as follows:

33 c(Ci / ml) < 1E- 05 (QCi/ml)x 170,000 (gpm) x 0.4x K-1890j 90 gpm L1E 2 c :< 6.39 E-3 (uCi/ml)

ODCM Page 14 Revision 15

The monitor R-18 alarm and trip setpoint (in cpm) is then determined utilizing the monitor calibration factor calculated in plant procedure CH-RETS-RMS. Assuming a calibration factor of 1.2 E-8 (uCi/m[)

cpm and a limiting batch release concentration determined above, the alarm and trip setpoint for monitor R-1 8 would be:

6.39E-3 (aCi/=ml) =5.33E + 05cpm above background 1.2 E 8/.Ci /ml cpm The setpoint values for the containment Fan Cooler monitor (R-16), Spent Fuel Pit Heat Exchanger Service Water Monitors (R-20A and R-20B), Steam Generator Blowdown Monitor (R-1 9), the Retention Tank Monitor (R-21, and the All volatile Treatment Waste Discharge Monitor (R-22) are calculated in a similar manner using equation (2), substituting appropriate values of (f) and the corresponding calibration factor.

Effluent Monitor Warning alarm setpoints are set at one-half of the trip setpoint. With all calculations equal this is a warning that 20% of the release limit has been reached at a single release point.

ODCM Page 1 5 Revision 1 5

1.6 Liquid Effluent Release Concentrations based Liquid batch releases are controlled individually and each batch release is authorized the discharge canal. Plant procedures upon sample analysis and the existing dilution flow in for sampling and CH-RETS-LIQ-RELEASE and CH-RETS-LIQ-COMP establish the methods for each batch analysis of each batch prior to release. A release rate limit is calculated before it is based upon analysis, dilution flow and all procedural conditions being met entering the discharge canal is authorized for release. The waste effluent stream if the preselected continuously monitored and the release will be automatically terminated and monitor setpoint is exceeded. A release may continue subject to grab sample analysis monitoring in accordance with Table 3.1-1.

analysis, If gross beta analysis is performed for each batch release in lieu of gamma isotopic then a weekly composite for principal gamma emitters and 1-131 is performed. Additional monthly and quarterly composite analyses are performed as specified in Table 1-1.

The equations used to calculate activity are:

Gamma Spectroscopy Equation (4):

uCi/gm Act. = peak area counts - bkgd counts (C Time)(Efl(Vol)(Decay)(3.7 E+04)

Gross Beta/Gamma Equation (5):

Where: C Time = seconds of count time uCi/gm Act. =Total counts - bkgd counts (C Time)(Efj)(Vol)(Decay)(3.7 E+04)

Eff = counting efficiency in count per sec disintegrations per sec vol = volume in milliliters decay = decay correction factor, exp(- A*r) 3.7 E+04 = conversion constant, in disintegration per sec MCi ODCM Page 1 6 Revision 15

1.7 Liquid Effluent Dose The dose contribution received by the maximally exposed individual from the ingestion of Lake Ontario fish and drinking water is determined using the following methodology. These calculations will assume a near field dilution factor of 1.0 in evaluating the fish pathway dose, and a dilution factor of 20 between the plant discharge and the Ontario Water District drinking water intake located 1.1 miles away (Figure 5-2). The dilution factor of 20 was derived from drift and dispersion studies documented in reference 4.

Dose contributions from shoreline recreation, boating and swimming have been shown to be negligible in the Appendix I dose analysis, reference 5, and do not need to be routinely evaluated. Shoreline sediment samples downstream from the plant will be collected at least annually for the Radiological Environmental Monitoring Program, as a conservatism. Also, there is no known human consumption of shellfish from Lake Ontario; therefore, this pathway is not taken into consideration in dose calculations.

The dose contribution to an individual will be determined to ensure that it complies with the specification of C. 1.2. Offsite receptor doses will be determined for the limiting age group and organ, unless census data show that actual offsite individuals are the limiting age group.

The following expression in used to calculate ingestion pathway dose contributions for the total release period from all radionuclides identified in liquid effluents released to unrestricted areas Equation (6): D, = 1X[A,,At CIU Fi]

I i i Where: D, = the cumulative dose commitment to the total body or any organ, T, from the liquid effluents for the summation of the total time period in mrem.

Z is bor total number oj hours oJ release.

I At = the length of the jth time period over which C,j and F, are averaged for all liquid releases in hours.

C, = the average concentration of radionuclide i in undiluted liquid effluent during time period Atý from any liquid release in ptCi/ml.

A,, = the site-related ingestion dose commitment factor to the total body or any organ, T,for each identified principal gamma and beta emitter in mrem/hr per/uCi/ml.

See equation (7).

F, = the discharge canal dilution factor for C,, during any liquid effluent release.

Defined as the ratio of the maximum undiluted liquid waste flow during release to unrestricted receiving waters. The dilution factor will depend on the number of circulation pumps operating and, during icing conditions, the percentage opening of the recirculating gate. Reference curves are presented in plant procedure CH-RETS-LIQ-RELEASE.

ODCM Page 1 7 Revision 15

Equation (7): A,* = ko (UI/Dw +UF BF,) DFi Where: A,, = The site-related ingestion dose commitment factor to the total body or to any organ, T, for each identified principal gamma and beta emitter in mrem/hr per

,.Ci/ml.

k, = units conversion factor, 1.14 E+05 = 1 E+06 pCi/4Ci x 1 E+03 ml/kg ÷ 8760 hr/yr U,, = a receptor person's water consumption by age group from table E-5 of Regulatory guide 1.109 Dw = dilution factor from the near field area of the release point to potable water intake.

The site specific dilution factor is 20. This factor is 1.0 for the fish ingestion pathway.

UF = a receptor person's fish consumption by age group from table E-5 of Regulatory Guide 1.109 BF = bioaccumulation factor for nuclide, i, in fish in pCi/kg per pCi/L, from table A-i of Regulatory Guide 1.109 DF,-- dose conversion factor for the ingestion of nuclide, i, for a receptor person in pre-selected organ, T, in mrem/pCi, from Tables E-11, E-12, E-13, E-14 of Regulatory guide 1.109 The monthly dose contribution from releases for which radionuclide concentrations are determined by periodic composite sample analysis may be approximated by assuming an average monthly concentration based on the previous monthly or quarterly composite analysis. However, in the Annual Radioactive Effluent Release Report the calculated dose contributions from these radionuclides shall be based on the actual composite analysis.

ODCM Page 18 Revision 1 5

Example:

Computing the dose to the whole body via the fish and drinking water pathways, assuming an initial Cs-137 discharge concentration of 3.0 E-04 ,Ci/ml:

Given the following discharge factors for example, where:

At,= 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the duration of the release Cj= 3.0 E-04 /, Oi/ml Fj = liquid waste flow = 20 qpm = 1.2 E-04 (dilution flow)(z) 170,000 gpm Z Near field dilution = 1.0 for Ginna D = 20 and, taking the following values from Regulatory Guide 1.109 which concern the critical receptor, which is considered to be the child in this case:

Uw = 510 I/year UF = 6.9 kg/year BF, = 2000 pCi/kg per pCi/I DF, = 4.62 E-05 mrem/pCi Then, the site-related ingestion dose commitment factor, Aj,, is calculated as follows:

Ai, mrem/hr = fko (Uw/Dw + UFBF,) DF, yCi/ml i 1.14 E+05 [510 + (6.9)(2000)] 4.62 E-05 20 Ai, = 7.28 E+04 mrem/hr per ACi/ml And, the whole body dose to the child is then:

D, mrem i(Aid(0t)(C)(F)

= (7.28 E+04)(1)(3.0 E-04)(1.2 E-04)

D, = 2.6 E-03 mrem to the whole body from Cs-137 The dose contribution from any other isotopes would then need to be calculated and all the isotopic contributions summed.

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2.0 RADIOACTIVE GASEOUS EFFLUENTS 2.1 Concentration (10 CFR 20)

CONTROLS the concentration at the C.2.1 The release of radioactive gaseous effluents shall be such that values specified in release point does not exceed ten times the effluent concentration Appendix B, Table 2, Column I of 10CFR Part 20.1001-20.2402.

APPLICABILITY: At all times.

specified in ACTION: If the concentration exceeds ten times the applicable concentration Appendix B, Table 2, Column I in an unrestricted area, submit to the Commission a special report within 30 days.

SURVEILLANCE REQUIREMENTS to be S.2.1 The radioactivity content of each batch release of radioactive gaseous waste and analysis discharged shall be determined prior to release by representative sampling in accordance with Table 2-1. The results of pre-release analyses shall be used with the the concentration at the calculational methods in Sections 2.4 and 2.5 to assure that gaseous waste point of release is limited to the values in C.2.1 and the dose rates from are limited to the values in C.2.2.

BASES This control is provided to ensure that the dose at any time at and beyond the SITE 20 to BOUNDARY from gaseous effluents will be within the annual dose limits of 10 CFR the UNRESTRICTED AREAS. The annual dose limits are the doses associated with 2, Column 1. Control of concentrations to concentrations of 10 CFR 20, Appendix B, Table within the limits of 10 CFR 20, Appendix B, Table 2, Column 1 at the release point is OF conservative against the dose limits associated with those concentrations to a MEMBER more restrictive than C.2.2.

THE PUBLIC at the SITE BOUNDARY. C.2.1 is intended to be 2.2 Dose Rate CONTROLS C.2.2 The instantaneous dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

(1) The dose rate for noble gases shall be :_500 mrem/yr to the total body and

3000 mrem/yr to the skin, and (2) The dose rate for 1-131, 1-133, tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days shall be :5 1500 mrem/yr to any organ.

Page 20 Revision 15 ODCM

APPLICABILITY: At all times.

Note: For unplanned release of gaseous wastes, compliance with C.2.2 may be calculated using annual average X/Q. Compliance with C.2.2 shall be determined by considering the applicable ventilation system flow rates. These flow rates shall be determined at the frequency required by Table 3.2-2.

ACTION: If the calculated dose rate of radioactive materials released in gaseous effluents from the site exceeds the limits of C.2.2, measures shall be initiated to restore releases to within limits as soon as practicable.

SURVEILLANCE REQUIREMENTS S.2.2.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters of Section 2.7 of the ODCM.

S.2.2.2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters of Section 2.7 of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 2-1.

Note: The effluent continuous monitors listed in Table 3.2-1 that have provisions for the automatic termination of gas decay tank, shutdown purge or mini-purge releases, shall be used to limit releases within the values established in C.2.2 when monitor setpoint values are exceeded.

BASES This control is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents will be within the annual dose limits of 10 CFR 20, Appendix B, Table 2, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR 20. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY.

Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. 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 whole 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.

ODCM Page 21 Revision 15

2.3. Dose (10 CFR 50, Appendix I)

CONTROLS areas at or beyond the C.2.3.1The air dose due to noble gases released in gaseous effluents to SITE BOUNDARY shall be limited to the following:

(1) During any calendar quarter to

5 mrad for gamma radiation and to _<10 mrad for beta radiation.

(2) During any calendar year to _<10 mrad for gamma radiation and to s: 20 mrad for beta radiation.

APPLICABILITY: At all times ACTION: Whenever the calculated dose to a MEMBER OF THE PUBLIC resulting from noble gases exceeds the limits of C.2.3.1, a Special Report shall be submitted to the Commission within thirty days which includes the following information:

(1) Identification of the cause for exceeding the dose limit.

(2) Corrective actions taken and/or to be taken to reduce releases of radioactive material in gaseous effluents to assure that subsequent releases will be within the above limits.

ACTION: During any month when the calculated dose to a MEMBER OF THE PUBLIC exceeds 1/48th the annual limits of C.2.3.1, (0.2 mrad gamma or 0.4 mrad beta),

projected cumulative dose contributions from gaseous effluents shall be determined for that month and at least once every 31 days for the next 3 months.

SURVEILLANCE REQUIREMENTS S.2.3.1.1 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters of Section 2.8 of the ODCM at least once every 31 days.

BASES This control is provided to implement the requirements of Sections II.B, III.A, and IV.A of Appendix 1, 10 CFR 50. The control implements the guides set forth in Section I.B of Appendix

1. 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 release of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as reasonably achievable". The SURVEILLANCE REQUIREMENTS implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM 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 10 CFR 50, Appendix I", Revision 1, October 1977 and Regulatory ODCM Page 22 Revision 15

Guide 1.111, "Methods for Estimating Atmospheric Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors", Revision I, July 1977. The ODCM equations for determining air doses at the SITE BOUNDARY are based on historical average atmospheric conditions.

CONTROLS C.2.3.2The dose to a MEMBER OF THE PUBLIC from 1-131, 1-133, tritium, and for all radioactive materials in particulate form with half-lives greater than eight days released with gaseous effluents from the site shall be limited to the following:

(1) during any calendar quarter to _<7.5 mrem to any organ.

(2) during any calendar year to _<15 mrem to any organ.

APPLICABILITY: At all times.

ACTION: Whenever the calculated dose to a MEMBER OF THE PUBLIC resulting from radionuclides other than noble gases exceeds the limits of C.2.3.2, a Special Report shall be submitted to the Commission within thirty days which includes the following information:

(1) Identification of the cause for exceeding the dose limit.

(2) Corrective actions taken and/or to be taken to reduce releases of radioactive material in gaseous effluents to assure that subsequent releases will be within the above limits.

ACTION: During any month when the calculated dose to a MEMBER OF THE PUBLIC exceeds 1/48th the annual limit of 2.3.2, (0.3 mrem), projected cumulative dose contributions from gaseous effluents shall be determined for that month and at least once every 31 days for the next 3 months.

SURVEILLANCE REQUIREMENTS S.2.3.2.1 Cumulative dose contributions for the current calendar quarter and current calendar year for lodine-131, lodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with methodology and parameters of Section 2.8 of the ODCM at least once every 31 days.

ODCM Page 23 Revision 15

BASES of This control is provided to implement the requirements of Sections II.C, Ill.A, and IV.A Appendix 1, 10 CFR 50. The control implements the guides set forth in Section I1.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time of implement the guides set forth in Section IV.A of Appendix I to assure that the release AREAS will be kept "as low as radioactive material in gaseous effluents to UNRESTRICTED reasonably achievable". The SURVEILLANCE REQUIREMENTS implement the requirements by in Section III.A of Appendix I that conformance with the guides of Appendix I be shown 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 the ODCM for calculating the doses due to the actual release rates of the subject materials 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 10 CFR 50, Appendix I", Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors", Revision I, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. He release rate controls for Iodine-131, Iodine-133, 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 at or beyond the SITE BOUNDARY. The pathways examined in development of the calculations were:

1.) Individual inhalation of airborne radionuclides 2.) Deposition of radionuclides onto green leafy vegetation with subsequent consumption by man 3.) Deposition of radionuclides onto grassy areas where milk animals and meat producing animals graze, followed by human consumption of that milk and meat 4.) Deposition of radionuclides on the ground followed by subsequent human exposure ODCM Page 24 Revision 15

2.4 Dose (40 CFR 190)

CONTROLS C.2.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

APPLICABILITY: At all times.

ACTION: With the calculated doses from the release of radioactive materials in gaseous effluents exceeding twice the limits of C.2.3.1 and C.2.3.2, prepare and submit to the Commission within 30 days a Special Report that defines the corrective actions to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and include scheduling for achieving conformance with the above limits. Calculations which include direct radiation contributions from the unit and from any radwaste storage shall be performed to determine total dose to a member of the public. This Special Report, as defined in 10 CFR 20.405(c) 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 releases covered by this report. 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 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

ACTION: This report shall include an analysis which demonstrates that radiation exposure to all MEMBERS OF THE PUBLIC from the plant are less than the 40 CFR 190 limits. Otherwise, the report shall request a variance from the commission to permit releases to exceed 40 CFR Part 190. Submittal of the report is considered a timely request by the NRC, and a variance is granted until staff action on the request is complete.

SURVEILLANCE REQUIREMENTS S.2.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with SURVEILLANCE REQUIREMENTS S.2.3.1, and S.2.3.2 at least once every 31 days, in accordance with the methodology and parameters of Section 2.8 of the ODCM.

S.2.4.2 Cumulative dose contributions from direct radiation from the unit and from radwaste storage shall be determined from environmental thermoluminescent dosimeter (TLD) data at least quarterly.

ODCM Page 25 Revision 15

BASES 190 that have been This control is provided to meet the dose limitations of 40 CFR requires the preparation and incorporated into 10 CFR 20 by 46FR1 8525. The specification to releases of radioactivity submittal of a Special Report whenever the calculated doses due 25 mrems to the whole body or any and to radiation from uranium fuel cycle sources exceed or equal to 75 mrems. It is highly organ, except the thyroid, which shall be limited to less than exceed the dose limits of unlikely that the resultant dose to a MEMBER OF THE PUBLIC will of Appendix I, and if 40 CFR 190 if the plant remains within twice the dose design objectives course of action that direct radiation doses are kept small. The Special Report will describe OF THE PUBLIC to within the should result in the limitation of the annual dose to a MEMBER be assumed that the dose 40 CFR 190 limits. For the purposes of the Special Report, it may MEMBER dose to any contributions from other uranium fuel cycle sources is negligible. If the the Special Report OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, violation of 40 CFR with a request for a variance, (provided the release conditions resulting in the provisions of 40 CR 190.11 and 190 have not already been corrected), in accordance with requirements of 40 CFR 10 CFR 20.405c, is considered to be a timely request and fulfils he 190, 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 20, as 10 CFR and does not apply in any way to the other requirements for dose limitation of PUBLIC addressed in C.1.1 and C.2.2.An individual is not considered a MEMBER OF THE out any operation that is part of the during any period in which he/she is engaged in carrying nuclear fuel cycle.

Page 26 Revision 15 ODCM

Table 2-1 Page 1 of 2 Radioactive Gaseous Waste Sampling and Analysis Program Gaseous Sampling Minimum Type of Lower Limit of Release Type Frequency Analysis Activity Detection Frequency Analysis (LLD)(gCilcc)(a)

Containment P P Principal 1 E-04 Purge Each Purge Gamma (bc) Emitters (e) 1 E-06 Grab Sample H-3 Auxiliary M (b) M (b) Principal 1 E-04 Building Grab Sample Gamma Ventilation Emitters (e) 1 E-06 H-3 Continuous (d) W (b) 1-131 1 E-12 Charcoal Sample 1-133 1 E-10 W (b) Principal 1 E-1 1 All Release Continuous (d)

Types as listed Particulate Gamma above Sample Emitters (e)

Continuous (d) M Gross Alpha 1 E-1 1 Composite Particulate Sample Continuous (d) Q Sr-89, Sr-90 1 E-1 1 Composite Particulate Sample Air Ejector M (b,f) M (b, f) Principal 1 E-04 Grab Sample Gamma Emitters (e) 1-131 (h) 1 E-06 H-3 (g)

All Release Continuous (d) Noble Gas Beta or Gamma 1 E-06 Types listed Monitor above Gas Decay P Each Tank P Principal 1 E-04 Tank Grab Sample Each Tank Gamma I Emitters (e)

W ODCM Page 2 7 Revision 1 5

PM a9nf')

Table 2-1 (continued) ,.

Radioactive Gaseous Waste Sampling and Analysis Program Notation (a) of Table 1-1.

(a) The lower limit of detection (LLD) is defined in Table sampler (b) Analyses shall also be performed when the monitor on the continuous reaches its setpoint.

per week when the reactor (C) Tritium grab samples shall be taken at least three times cavity is flooded.

shall be known for (d) The ratio of the sample flow rate to the sampled stream flow rate the time period covered by each dose or dose rate calculation made in accordance with C.2.1, C.2.2, C.2.3.1, & C.2.3.2 are (e) The principal gamma emitters for which the LLD specification will apply exclusively the following radionuclides:

Kr-85m, Xe-1 33, Xe-1 33m and Xe-1 35 for gaseous emissions 1-131, Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.

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

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

(f) Air ejector samples are not required during cold or refueling shutdowns.

is less (g) Air ejector tritium sample is not required if the secondary coolant activity than 1 E-04 /tCi/gm.

(h) Air ejector iodine samples shall be taken and analyzed weekly if the secondary coolant activity exceeds 1 E-04 4Ci/gm.

ODCM Page 28 Revision 15

2.5 Gaseous Effluent Release Points There are three release points continuously monitored for noble gases, containment vent, plant vent and air ejector. The containment vent and plant vent are also continuously monitored for radioiodines and particulates. Since the air ejector is a steam release point, continuous radioiodine and particulate monitoring is not required when the secondary coolant activity is less than 1 E-04 4Ci/gm. Flow rates through the vents are measured periodically. During shutdown, temporary trailers may be brought on site that also require monitoring and characterization of their releases, such as the CO2 decon trailer.

Quarterly plant measurements of one week duration for the particulate and iodine released in the steam by the air ejector demonstrate that sampling this pathway for particulate and iodine is not necessary since these releases are less than 0.1% of the Plant Vent. The releases are correlated to blowdown activity for determining activity in steam releases. During shutdown and startup, special systems are in use that may release small amounts of radioactivity in steam releases. This is accounted for by using operational data and activity in the source of the steam. Grab samples are obtained when practicable.

If an unmonitored release point is discovered, a calculation is performed to determine the potential radioactivity that is released. The calculation includes a conservative estimate of source term if sample data is not available, and a conservative estimate of flow rate and duration if measurement of flow and duration are not available. If the release is continuous, it is included in the monthly report that accounts for releases from the site for calculating doses to the general public.

2.6 Gaseous Effluent Monitor Setpoints Alarm and/or trip setpoints for specified radiation monitors are required on each noble gas effluent line from the plant. Precautions, limitations and setpoints applicable to the operation of Ginna Station gaseous effluent monitors are provided in plant procedures P-9 and CH-RETS-RMS. Setpoints are conservatively established for each ventilation noble gas monitor so that dose rates in unrestricted areas corresponding to 10 CFR Part 50 Appendix I limits will not be exceeded. Setpoints shall be determined so that dose rates from releases of noble gases will comply with C.2.2.(1). Table 6-2 provides the gaseous and particulate meteorological assumption used in development of the P-9 setpoints.

ODCM Page 29 Revision 15

effluent The calculated alarm and trip action setpoints for each radioactive gaseous monitor must satisfy the following equation:

Y-,Qiv Equation (8): c, <

(1)(k)(K)

Where: cv = setpoint in cpm from vent (v)

Qi,= release rate limit by specific nuclide (i) in /ACi/sec f = discharge flow rate in cfm k = units conversion factor in cc/sec/cfm K = calibration factor in 4Ci/cc/cpm The general methodology for establishing plant ventilation monitor setpoints is based upon a vent concentration limit in 4Ci/cc derived from site specific meteorology and vent release characteristics.

Additional radiation monitor alarm and/or trip setpoints are calculated for radiation monitors measuring radioiodines, radioactive materials in particulate form and to radionuclides other than noble gases. Setpoints are determined to assure that dose rates from the release of these effluents shall comply with C.2.2(2)

The release rate limit for noble gases shall be calculated by the following equation for total body dose:

Qj, [uCi/sec] Q , 500 mrem/yr Equation (9): (X/Q), ZK,Q,V i

of Note: The occupancy factor "T" of 1 is assumed. This may be modified following reviews the area in question.

and by the following equation for skin doses:

ODCM Page 30 Revision 15

Equation (10):

Q r1Qv,mse']

.C/sc]~ 3000

'*'v(X/Q), mrem/yr Y(LI+ 1.1M)Q, Where: Q,v = the release rate of radionuclide (i) from vent (v) which results in a dose rate of 500 mrem/yr to the whole body or 3000 mrem/yr to the skin of the critical receptor in pCi/sec.

K = the total body dose factor due to gamma emissions for each 3

identified noble gas radionuclide in mrem/yr per 4Ci/m from table 2-3.

Li = the skin dose factor due to beta emissions for each identified noble gas radionuclide in mrem/yr per /Ci/m 3 from table 2-3.

M, = the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad/yr per 4Ci/m 3 from Table 2-3. Unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose.

(X/Q)V the highest calculated annual average dispersion parameter for estimating the dose to the critical offsite receptor from vent release point (v) in sec/m 3 . The (X/Q),

is calculated by the method described in Regulatory Guide 1.111.

Noble gas monitor setpoints are conservatively set according to procedure P-9 to correspond to fractions of the applicable 10 CFR Part 20 dose limits for unrestricted areas. Fractions are small enough to assure the timely detection of any simultaneous discharges from multiple release points before the combined downwind site boundary concentration could exceed allowable limits. Additional conservatism is provided by basing these setpoints upon instantaneous downwind concentrations. Release rates during the remainder of a given year, combined with any infrequent releases at setpoint levels, would result in only a very small fraction of the 10 CFR Part 20 annual limits.

Historically, xenon-1 33 is the principal noble gas released from all vents and is appropriate for use as the reference isotope for establishing monitor setpoints. The whole body dose will be limiting, and the Xe-1 33 release rate limit is calculated by substituting the appropriate values into equation (9). After the release rate limit for Xe-133 is determined for each vent, the corresponding vent concentration limits are calculated based on applicable vent flow rates.

Annually-derived monitor calibration factors in pCi/cc per cpm are used to convert limiting vent concentrations to count rates.

ODCM Page 31 Revision 1 5

Exam*le: Plant Vent Monitor, R-14 assuming a measured Using Xe-1 33 as the controlling isotope for the setpoint and cpm above background, activity of 2.66 E-04 /.tCi/cc and a ratemeter reading of 4750 of 7.45 E+04 cfm, Ki from the efficiency can be calculated, using a measured vent flow of 2.7 E-06, the Release Table 2-3 of 2.94E+02 and a (X/Q), for the site boundary Rate Limit is calculated and then the setpoint determined.

Activity Xe-133 efficiency = Net ratemeter reading

= 5.67 E-08 uCilcc Xe-133 efficiency = 2.66 E-04 cpm 4750 Using Equation 9:

500 _<6.3 E+05 uCi/sec (2.94 E+02)(2.7 E-06)

Release Rate Limit Qv < 500 mremyr (Ki)(X/Q),

Using Equation 8:

Setpoint c <

(1)(k)(K) 6.3 E+05 uCi/sec (7.45 E+04 cfm) 472 c-l ) 5.67 E-08 UCi-c-cfm 0pm 3 c c :5 3.2 E+05 cpm Page 32 Revision 15 ODCM

Per procedure P-9, R-14 is set at 0.4 of this value or 1.28 E+05 cpm for normal operation. 40% of the release rate limit is a conservatism based on the possibility of two release points simultaneously at their setpoints for a total of 80% of the release rate limit.

Effluent Monitor Warning alarm setpoints are set at one-half of the trip setpoint. With all calculations equal this is a warning that 20% of the release limit has been reached at a single release point.

ODCM Page 3 3 Revision 15

Table 2-2 Meteorological Data and Locations of Receptors for Set Point Calculations Process Monitors Monitor Geographic Location Release Point X/Q D/Q Flow (Radioisotope) (Distance/Direction) (vent) (sec/m3) (sec/m3) (cfm)

R-1OA 0.5 - 1 mile ESE Containment 2.4E-8 11,000 (Radionuclide)

R-1OB 0.5 - 1 mile ESE Plant 3.OE-8 76,000 (Radionuclide)

R-11 (Cs-1 37) 0.5 - 1 mile ESE Containment 2.4E-8 11,000 R-12 (Xe-133) 0.5 -1 mile E Containment 1.6E-6 11,000 R-13 (Cs-1 37) 0.5 - 1 mile ESE Plant 3.OE-8 76,000 R-14 (Xe-1 33) 0.5 - 1 mile E Plant 2.7E-6 76,000 R-15 (Xe-133) 0.3 miles SSE Air Ejector 4 600 SPINGS Monitor Geographic Location Release Point Noble Gas Iodine Particulate (Radioisotope) (Distance/Direction) (vent) uCilmin UCilmin ACilmin R-12A 0.5 - 1 mile Containment 6.3E+7 3.4 80 R-14A 0.5 - 1 mile Plant 3.8E+7 2.7 64 R-15A 0.33 miles Air Ejector 1.OE+2 Further details found in procedure P-9.

1-131, 1-133, H-3 and particulates with half lifes greater than 8 days utilizes the following equation:

, ý1500D mrem / year

/ QvP, For Noble Gases:

Q _ 500 mremXv/X / year Pi Food and ground pathways in m2 mrem/year per MCi/sec Ki mrem/year per 4Ci/m3 Owv Release rate in units of kiCi/sec ODCM Page 34 Revision 15

2.7 Gaseous Effluent Dose Rate this manual are Gaseous effluent monitor setpoints as described in Section 2.6 of action to be established at concentrations which permit some margin for corrective CFR Part 20 limitations.

taken before exceeding offsite dose rates corresponding to 10 CH-RETS-PURGE Plant procedures CH-RETS-RMS-CV, CH-RETS-RMS-CV-ALT, CV, CH-RETS-SAMP-PV, CH-RETS-SAMP-PV-ALT, CH-RETS-PV-RELEASE, for sampling CH-RETS-AIR-H3 and CH-RETS-MINIPURGE establish the methods purge releases.

and analysis for continuous ventilation releases and for containment the methods for sampling and Plant procedure CH-RETS-GDT-REL establishes dose rate in unrestricted analysis prior to gas decay tank releases. The instantaneous be calculated areas due to unplanned releases of airborne radioactive materials may the following using annual average X/Q's. Dose rates shall be determined using expressions:

For noble gases:

Equation (12): D, = I[(Li + 1.1 M)(X/Q), Q,,,] -<3000 mrem/yr total gamma and beta dose to the skin half-lives greater For 1-131, 1-133, tritium and all radioactive materials in particulate form with than 8 days:

Equation (13): D, - P, (X/Q)VQl, *- 1500 mrem/yr to critical organ ODCM Page 35 Revision 15

Where:

K= the total body dose factor due to gamma emissions for each identified noble gas 3

radionuclide (J) in mrem/yr per #Ci/m from Table 2-2.

L = the skin dose factor due to beta emissions for each identified noble gas 3

radionuclide (i) in mrem/yr per ksCi/m from Table 2-2.

Mi the air dose factor due to gamma emissions for each identified noble gas 3

radionuclide (J) in mrad/yr per UCi/m from Table 2-2. Unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose.

P, the dose parameter for radionuclide (J) other than noble gases for the inhalation 3

pathway, in mrem/yr per 12Ci/m . The dose factors are based on the critical individual organ and the child age group. Pi is further defined as: Pi = 3 (10 6pCi/uCi)(BR)(DFA,) where BR is the breathing rate for a child in m /yr and DFA. is the dose factor for the child in mrem/pCi.

(X/Q), = the highest calculated annual average relative concentration for any area at or beyond 3

the unrestricted area boundary in sec/m .

W = the highest annual average dispersion parameter for estimating the dose to the 2

critical receptor in sec/m 3 for the inhalation pathway and in m for the food and ground pathways.

Qi = the release rate of radionuclide (i) from vent (v) in 12Ci/sec.

ODCM Page 36 Revision 15

2.8 Gaseous Effluent Doses The air dose in unrestricted areas due to noble gases released in gaseous effluents from the site shall be determined using the following expressions:

During any calendar year, for gamma air dose:

Equation (14): Dj, = 3.17 E-08 . [M(X/Q), Q1 3 <_ 10 mrad During any calendar quarter, for gamma air dose:

During any calendar year for beta air dose:

Equation (15): D*f = 3.17 E-08 E [Nr(X/Q)v Q1j]

20 mrad I

Durinq any calendar quarter, for beta air dose:

Equation (15A): D*f = 3.17 E-08 Y [N,(X/Q), Qji - 10 mrad Where:

Mi the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad/yr per /zCi/m 3 from Table 2-3 N, the air dose factor due to beta emissions for each identified noble gas radionuclide in 3

mrad/yr per /.ZCi/m from Table 2-3 (X/Q), for vent releases. The highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary in sec/mr.

D = the total gamma air dose from gaseous effluents in mrad.

D = the total beta air dose from gaseous effluents in mrad.

Q = the release of noble gas radionuclides, i, in gaseous effluents from vents in uCi.

Releases shall be cumulative over the time period.

3.17 E-08 = the inverse of the number of seconds in a year ODCM Page 3 7 Revision 15

The dose to an individual from 1-131, 1-133, tritium and all radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents released from the site to unrestricted areas shall be determined using the following expression:

dose during any calendar year:

Equation (16): D, = 3.17 E-08 Y_ [Ri W Q,] _ 15 mrem I

dose during any calendar quarter:

Equation (16A): D1 = 3.17 E-08 E [R1 WV Q1, _<7.5 mrem Where:

D = the total dose from 1-131, 1-133, tritium and all radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents in mrem.

2 Ri = the dose factor for each identified radionuclide (i) in m mrem/yr per M.Ci/sec or 3

mrem/yr per /.ZCi/m from Table 2-5.

W = the annual average dispersion parameter for estimating the dose to2an individual 3

at the critical location in sec/m for the inhalation pathway and in m- for the food and ground pathways.

Qr = the release of 1-131, 1-133, tritium and all radioactive materials in particulate form in gaseous effluents with half-lives greater than 8 days in MCi. Releases shall be cumulative over the desired time period as appropriate.

ODCM Page 38 Revision 15

Table 2-3 Dose Factors to the Child For Noble Gases and Daughters

Total Body Dose Factor Skin Dose Factor L, Gamma Air Dose Factor M, Beta Air Dose Factor N1 Radionuclides Ki (mrem/yr per yCilm3 ) (mradlyr per /Cilm 3 ) (mrad/yr per Cilmr3 )

(mremiyr per /Ci/m 3 )

Kr-83m 7.56E-02** ---- 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01 E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-135m 3.12E+03 7.11 E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03

The listed dose factors are for radionuclides that may be detected in gaseous effluents. These dose factors for noble gases and daughter nuclides are taken from Table B-1 of Regulatory Guide 1.109 (reference 3). A semi-infinite cloud is assumed.

7.56E-02 = 7.56 x 10.2 ODC"v Page 39 Rovicinn 1 S

Table 2-4 Dose Parameters for Radionuclides and Radioactive Particulate, Gaseous Effluents

P, Inhalation Pathways P, Food & Ground P, Inhalation Pathways P, Food & Ground Pathways Radionuclides (mremlyr per J.Ci/mW) Pathways Radionuclides (mremtyr per JCilm3) (in 2

mremlyr per MsCI/sec)

(m-2 mremlyr per

/tCi/sec)

H-3 6.5E+02 2.4E+03 Cd-115m 7.OE+04 4.8E+07 C-14 8.9E+03 1.3E+09 Sn-126 1.2E+06 1.1E+09 Cr-51 3.6E+02 1.1E+07 Sb-125 1.5E+04 1.1E+09 Mn-54 2.5E+04 1.1E+09 Te-127m 3.8E+04 7.4E+10 Fe--59 2.4E+04 7.OE+08 Te-129m 3.2E+04 1.3E+09 Co-58 1.1E+04 5.7E+08 Te-132 1.OE+03 7.2E+07 Co-60 3.2E+04 4.6E+09 Cs-134 7.OE+05 5.3E+10 Zn-65 6.3E+04 1.7E+10 Cs-136 1.3E+05 5.4E+09 R1J-86 1.9E+05 1.6E+10 Cs-137 6.1E+05 4.7E+10 Sr-89 4.0E+05 1.OE+10 Ba-140 5.6E+04 2.4E+08 Sr-90 4.1E+07 9.5E+10 Ce-141 2.2E+04 8.7E+07 Y-91 7.OE+04 1.9E+09 Ce-144 1.5E+05 6.5E+08 Zr-95 2.2E+04 3.5E+08 Np-239 2.5E+04 2.5E+06 Nb-95 1.3E+04 3.6E+08 1-131 1.5E+07 1,11E+12 Mo-99 2.6E+02 3.3E+08 1-133 3.6E+06 9.6E+09 Ru-103 1.6E+04 3.4E+10 Unidentified 4.1E+07 9.5E+ 10 Ru-1 06 1.6E+05 4.4E+1 1 Ag-110m I 3.3E+04 I 1.5E+10

The listed dose parameters are for radionuclides that may be detected in gaseous effluents. These and additional dose parameters for isotopes not included in Table 2-4 may be calculated using the methodology described in NUREG-0133, Section 5.2.1 (reference 2).

ODCM' ePae 40 Rpvision 15

Table 2-5 Pathway Dose Factors Due to Radionuclides Other Than Noble Gases

Inhalation Meat Ground Plane Cow-Milk-Child Leafy Vegetables Radionuclides Pathway R, Pathway R, Pathway R, Pathway R, Pathway R, (mrem/yr per /tCilmr) (m2"mremlyr per yCi/sec) (m2omrem/yr per /Ci/sec) (m 2*mremlyr per/uCilsec) (m2omrem/yr per

/t.Ci/sec)

H-3 1.12E+03 2.33E+02 0. 2.38E+03 2.47_E+02 Cr-51 1.70E+04 4.98E+05 5.31 E+06 5.75E+06 1.63E+06 Mn-54 1.57E+06 7.60E+06 1.56E+09 3.70E+07 5.38E+07 Fe-59 1.27E+06 6.49E+08 3.09E+08 4.01E+08 1.1OE+08 Co-58 1.1OE+06 9.49E+07 4.27E+08 7.01 E+07 4.55E+07 Co-60 7.06E+06 3.61E+08 2.44E+ 10 2.25E+08 1.54E+08 Zn-65 9.94E+05 1.05E+09 8.28E+08 1.99E+10 2.24E+08 Sr-89 2.15E+06 4.89E+08 2.42E+04 1.28E+10 5.39E+09 Sr-90 1.01 E+08 1.01E+10 0 1.19E+10 9.85E+10 Zr-95 2.23E+06 6.09E+08 2.73E+08 8.76E+05 1.13E+08 1-131 1.62E+07 2.60E+09 1.01E+07 4.95E+11 2.08E+ 10 1-133 3.84E+06 6.45E+01 1.43E+06 4.62E+09 3.88E+08 Cs-134 1.01E+06 1.42E+09 7.70E+09 6.37E+10 1.96E+09 Cs-136 1.71 E+05 5.06E+07 1.64E+08 6.61E+09 1.60E+08 Cs-137 9.05E+05 1.27E+09 1.15E+10 5.75E+10 1.80E+09 Ba-140 1.74E+06 5.OOE+07 2.26E+07 2.75E+08 2.03E+08 Ce-141 5.43E+05 1.45E+07 1.48E+07 1.43E+07 8.99E+07

Additional dose factors for isotopes not included in Table 2-5 may be calculated using the methodology described in NUREG-01 33, Section 5.3,1 (reference 2).

ODCMt Paep 41 Rpvijion 15

3.0 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION 3.1 Liquid Effluent Monitors CONTROLS C.3.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.1-1 shall be OPERABLE with their Alarm/Trip setpoints set to ensure that the limits of Control C.1.1 are not exceeded. The Alarm/Trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: At all times.

ACTION: With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

ACTION: With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.1-1.

Restore the inoperable instrumentation to OPERABLE status within 30 days or explain in the next Annual Radioactive Effluent Release Report, pursuant to Section 6.2 of the ODCM, why this inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS S.3.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 3.1-2.

BASES The radioactive liquid effluent instrumentation is is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the Alarm/Trip will occur prior to exceeding the limits of 10 CFR 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50.

ODCM Page 42 Revision 15

Table 3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation Minimum Channels Action Gross Activity Monitors (Liquid)

OPERABLE 1 1

a. Containment Fan Coolers (R-16) 1 2 b- Liquid Radwaste (R-18) 1(a) 3

c. Steam Generator Blowdown (R-19) 1 1 d Spent Fuel Pool Heat Exchanger (R-20A, R-20B)

e. Turbine Building Floor Drains (R-21) 1 1 1 2

f. High Conductivity Waste (R-22)

(a) Not required when steam generator blowdown is being recovered, i.e. not released.

Table Notation Action 1 - If the number of OPERABLE channels is less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months grab samples are analyzed for isotopic concentration or gross radioactivity (beta or Gamma) at a lower limit of detection (LLD) of at most 1 E-07 PCi/gm.

Action 2 - If the number of OPERABLE channels is less than required by the minimum Channels OPERABLE requirement, effluent releases from the tank may continue for up to 14 days, provided that prior to initiating a release:

Note: When counting 2 independent samples for agreement, doubling the acceptance criterion for low (<6E-05 1 iCi/ml) activity samples from 10% to 20% results in a consequence at the release point of < 1%. The expanded acceptable criterion for low activity samples is employed to compensate for increased impact of sampling and counting error on acceptance.

1. At least two independent samples of the tank's contents, taken at least 60 minutes apart, are analyzed and agree within 10% of total activity, (20% if total activity minus noble gases <6E-05 uCi/ml), and

2. At least two technically qualified members of the Facility Staff independently review and approve the analytical results, and

3. At least two technically qualified members of the Facility Staff independently verify the discharge line valving.

ODCM Page 43 Revision 15

Action 3 - When Steam Generator Blowdown is being released (not recycled) and the number of channels OPERABLE is less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for isotopic concentration or gross radioactivity (beta or gamma) at a lower limit of detection (LLD) of at most 1 E-07 pCi/gram:

1. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the concentration of the secondary coolant is > 0.01 uCi/gram (DOSE EQUIVALENT 1-131).

2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the concentration of the secondary coolant is _<0.01 uCi/gram (DOSE EQUIVALENT 1-131).

Table 3.1-2 Radioactive Liquid Effluent Monitoring SURVEILLANCE REQUIREMENTS Gross Activity Monitor (Liquid) Channel Source Functional Channel Check Check Test Calibration

a. Containment Fan Coolers (R-16) D(5) M(3) Q(2) R(4)

b. Liquid Rad waste (R-18) D(5) M(3) Q(1) R(4)

c. Steam Generator Blowdown (R-19) D(5) M(3) Q(1) R(4)

d. Spent Fuel Pool Heat Exchanger (R-20A, R- D(5) M(3) Q(2) R(4) 20B) 1

e. Turbine Building Floor Drains (R-21) D(5) M(3) Q(1) R(4)

f. High Conductivity Waste (R-22) D(5) M(3) Q(1) R(4)

Table Notation (1) The FUNCTIONAL Test shall also demonstrate that automatic isolation of this pathway and control room alarm will occur if any of the following conditions exist:

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

2. Power failure. (Verified in same functional test as Alarm/Trip setpoint)

(2) The FUNCTIONAL Test shall also demonstrate that control room alarm occurs if any of the following conditions exist.

1. Instrument indicates measured levels above the alarm setpoint.

2. Power failure. (Verified in same functional test as Alarm setpoint)

(3) This check may require the use of an external source due to high background in the sample chamber.

(4) Source used for the CHANNEL CALIBRATION shall be traceable to the National Institute for Standards and Technology (NIST) or shall be obtained from suppliers (e.g. Analytics) that provide sources traceable to other officially designated standards agencies.

(5) Applies only during releases via this pathway.

ODCM Page 44 Revision 15

3.2 Gaseous Effluent Monitors CONTROLS C.3.2 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.2-1 shall be OPERABLE with their Alarm/Trip setpoints set to ensure that the limits of Control C.2.2 are not exceeded. The Alarm/Trip setpoints of these channels meeting Control C.2.2 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: As shown in Table 3.2-1 Note: The Radioactive Effluent Monitoring Instrumentation may be removed from service for short periods of time without the instrumentation being considered inoperable for weekly grab filter or cartridge changes or quarterly valve stroke testing, with the exception of R10 A, R1 1, R1 2 skid. Preventative maintenance, calibrations, and moving filter replacements require instrumentation to be declared inoperable.

ACTION: With a radioactive gaseous effluent monitoring instrumentation channel Alarm/Trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.

ACTION: With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.2-1.

Restore the inoperable instrumentation to OPERABLE status within 30 days or, if not, explain in the next Annual Radioactive Effluent Release Report, pursuant to Section 6.2 of the ODCM, why this inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS S.3.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 3.2-2.

BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the Alarm/Trip will occur prior to exceeding the limits of 10 CFR 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50.

ODCM Page 45 Revision 15

Table 3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation Plant Ventilation (a) Minimum Channels Action OPERABLE

a. Iodine sampler (R-1OB or R-14A3) 1 1

b. Particulate Sampler (R-1 3 or R-1 4A1) 1 1

c. Noble Gas Activity (R-14 or R-14A5) 1 (b) 2

d. Containment Noble Gas Activity (R-12) or 1 (e) 3 Containment Particulate Sampler (R-1 1) (d)

Containment Purge (c) Minimum Channels Action OPERABLE

a. Iodine Sampler (R-10A or R-12A3) 1 1

b. Particulate Sampler (R-1 1 or R-12A1) 1 (f) 5

c. Noble Gas Activity (R-12 or R-12A5) 1 (f) 5 Air Ejector Monitor (g) Minimum Channels Action OPERABLE Noble Gas Activity (R15 or R15A5) 1 4 Note: Gaseous effluent monitors are not considered inoperable due to changes in ventilation flow.

Reduced flow in the ventilation makes the monitor setpoint more conservative.

Table Notation (a) Required at all times.

(b) Only radiation monitors R-13 and R-14 have isolation signals. If R-14A5 is being used to monitor gas releases, no gas decay tanks may be released.

(c) Required in MODES 5 and 6.

(d) The mini-purge system allows the release of Containment atmosphere through the plant vent. 10 CFR 100 type releases via mini-purge are limited by an isolation signal generated from Safety Injection. 10 CFR 20 releases through the mini-purge are considered to be similar to other plant ventilation releases and are monitored by R-14, R-13 and R-10B. R-1A may be used as a substitute since automatic isolation is available from the R-12 and R-1 1 monitors if the activity in Containment increases.

Therefore, either R-12 or R-11 is required to sample Containment during a mini-purge release. Automatic isolation of mini-purge for 1 OCFR part 20 type releases is considered unnecessary due to the low flow associated with mini-purge, the continuous monitoring from R-12 o R-1 1 and the original measurement before the purge begins. To ensure the Containment sample monitored by R-1 1 or R-1 2 is 2.

representative of the containment atmosphere, at least one recirculation fan is required to be in operation during mini-purge operation. Should R-11 and R-12 become inoperable, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months limit is chosen to be consistent with the generally accepted time for prompt action.

ODCM Page 46 Revision 15

Table 3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation Paqe 2 of 2 (e) If the R-1 OA, R-1 1, R-1 2 skid is not OPERABLE, it is possible to substitute the R-1 0B, R-13, R-14 skid when the R-14A skid is OPERABLE. The setpoints for the R-1OA, R 11, R-12 skid would be used. There would be no automatic containment isolation capability from the radioactive effluent monitoring instrumentation when using R-10B, R-13, R-14 skid for containment leakage measurements. This cannot be used if Containment Ventilation Isolation is required.

(f) If containment ventilation isolation instrumentation is required by LCO 3.3.5 for core alteration or movement of irradiated fuel in containment, R-12A skid cannot be used in place of the R-1OA, R-11, R-12 skid. Terminate the purge immediately.

(g) Required only when Air ejector is operating.

Action I - If the number of OPERABLE channels is less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided iodine and particulate samples are continuously collected with alternate sampling equipment. This should be completed within one hour.

Action 2 - If the number of OPERABLE channels is less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken and analyzed for isotopic activity at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Action 3- If the number of OPERABLE channels is less than required by the Minimum Channels OPERABLE requirement, or at least one containment fan cooler is not in operation, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months terminate any mini-purge in process.

Action 4- If the number of OPERABLE Channels is less than required by the Minimum Channels OPERABLE requirement and the Secondary Activity is _<1 E-04 aCi/gm, effluent releases may continue via this pathway provided grab samples are analyzed for isotopic concentration or gross radioactivity (beta or gamma) at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . If the secondary activity is > 1 E-04 4Ci/gm, effluent releases via this pathway may continue for up to 31 days provided grab samples are taken every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and analyzed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Action 5 - If the number of OPERABLE channels is less than required by the Minimum Channels Operable requirement, terminate the purge within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ODCM Page 47 Revision 15

Table 3.2-2 Radioactive Gaseous Effluent Monitoring SURVEILLANCE REQUIREMENTS Plant Ventilation Channel Source Functional Channel Check Check Test Calibration

a. Iodine Sampler (R-10B) W(6) N.A. M R(4)

b. Particulate Sampler (R-13) W(6) N.A. N.A R(4)

c. Noble Gas Activity (R-14) D(6) M Q(1) R(4)

d. Flow Rate Determination N.A. N.A. N.A. R(5)

Containment Purge Channel Source Functional Channel Check Check Test Calibration

a. Iodine Sampler (R-10A) W(6) N.A. M R(4)

b. Particulate Sampler (R-1 1) W(6) N.A. Q(1) R(4)

c. Noble Gas Activity (R-12) D(6) PR Q(1) R(4)

d. Flow Rate Determination N.A. N.A. N.A. R(5)

Air Ejector Monitor Channel Source Functional Channel Check Check Test Calibration Noble Gas Activity (R-1 5) D(6) M M(2) R(4)

Table Notation (1) The FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm occur if any of the following conditions exist:

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

2. Power failure. (Verified in same functional test as Alarm/Trip Setpoint)

(2) The FUNCTIONAL TEST shall also demonstrate that control room alarm occurs if any of the following conditions exist.

1. Instrument indicates measured levels above the alarm setpoint.

2. Power failure. (Verified in same functional test as Alarm Setpoint)

(3) This check may require the use of an external source due to high background .in the sample chamber.

(4) Source used for the Channel Calibration shall be traceable to the National Institute for Standards and Technology (NIST) or shall be obtained from suppliers (e.g. Amersham) that provide sources traceable to other officially designated standards agencies.

(5) Flow rate for main plant ventilation exhaust and containment purge exhaust are calculated by the flow capacity of ventilation exhaust fans in service and shall be determined at the frequency specified.

(6) Applies only during releases via this pathway.

ODCM Page 48 Revision 15

3.3 Radiation Accident Monitoring Instrumentation CONTROLS C.3.3 The radiation accident monitoring instrumentation channels shown in Table 3.3-1 shall be OPERABLE according to the following schedule:

APPLICABILITY:

(1) Containment Area Monitors (R-29, R-30) - Modes 1, 2, and 3 (2) Containment Purge (R-12A) - Modes 5 and 6 when the purge flanges are removed.

(3) Plant vent (R-14A) - All modes (4) Air Ejector (R-15A) - When air ejector is operating (5) A Main Steam Line (R-31) - Modes 1, 2, and 3 (6) B Main Steam Line (R-32) - Modes 1, 2, and 3 Note: The Radiation Accident Monitoring Instrumentation may be removed from service for short periods of time without the instrumentation being considered inoperable for weekly grab filter or cartridge changes. Preventative maintenance and calibrations require instrumentation to be declared inoperable.

ACTION: With less than the minimum number of radiation accident monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS S.3.3.1 Each radiation accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and FUNCTIONAL TEST at the frequencies shown in Table 3.3-2.

BASES Radiation accident monitoring instrumentation is provided to monitor, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the Alarm will occur prior to exceeding the limits of 10 CFR 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50.

ODCM Page 49 Revision 1 5

Table 3.3-1 Radiation Accident Monitoring Instrumentation Instrument Minimum Channels Action Operable Containment Area Monitors (R-29 and R-30) See Tech Spec 3.3.3 2 2 Noble Gas Effluent Monitors

a. Containment Purge (R-12A) 1* 1

b. Plant Vent (R-14A) 1 1

c. Air Ejector (R-1 5A) 1"* 1

d. A Main Steam Line (R-31) 1 1

e. B Main Steam Line (R-32) 1 1

Only when the shutdown purge flanges are removed; otherwise, instrumentation kept in STANDBY mode.

R-1 5A has a low activity alarm to ensure equipment is not accidently removed from service when the plant is operating. During shutdown, the channel is removed from scan on the PPCS to keep from receiving unnecessary alarms.

Table Notation Action 1 - With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirements, either restore the inoperable channel(s) to OPERABLE status within 30 days of the event, or if not restored, prepare and submit, within the following 14 days, a Special Report to the Commission outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. If the channel(s) is inoperable greater than 7 days but less than 30 days, report the cause of the inoperability and the actions taken in the Annual Radioactive Effluent Release Report.

Action 2 - Take action in accordance with Tech Spec Table 3.3.3-1 item 10.

ODCM Page 50 Revision 15

Table 3.3-2 Radiation Accident Monitoring Surveillance Requirements 1 T Fu nctiona Channel Radiation Accident Monitoring Channe Source Functiona Channel Check I Calibration Instrumentation I Check Test W(3) M M(1) R(2)

a. Containment Purge (R-12A)

D(3) M M(1) R(2)

b. Plant Vent (R-14A)

D(3) M M(1) R(2)

c. Air Ejector (R-1 5A)

M N.A. Q R

d. A Main Steam Line (R-31)

e. B Main Steam Line (R-32) M N.A. Q R Table Notation (1) The FUNCTIONAL TEST shall also demonstrate that control room alarm occurs if any of the following conditions exist.

1. Instrument indicates measured levels above the alarm setpoint.

2. Power failure.

(2) Source used for the CHANNEL CALIBRATION shall be traceable to the National Institute for Standards and Technology (NIST) or shall be obtained from suppliers (e.g. Analytics) that provide sources traceable to other officially designated standards agencies.

(3) Applies only during releases via this pathway.

ODCM Page 51 Revision 1 5

3.4 Area Radiation Monitors

a. Channel calibration, channel check, and a functional test of the area radiation monitors shall be performed as specified in Table 3-4.

Table 3-4 Area Radiation Monitor Surveillance Requirements Instrument Channel Functional Test Channel Check Calibration A. Control Room R-1 D M R R-2 D M R B. Containment C. Radiochemistry Lab R-3 D M R D. Charging Pump Room R-4 D M R E. Spent Fuel Pool R-5 D M R F. Nuclear Sample Room R-6 D M R G. Incore Detector Area R-7 D M R H. Drumming Station R-8 D M R I. Letdown Line Monitor R-9 D M R J. Component Cooling Water Heat Exchanger* R-17 D M R K. AVT A Mixed Bed R-23 N.A. Q N.A.

L. AVT B Mixed Bed R-24 N.A. Q N.A.

M. AVT C Mixed Bed R-25 N.A. Q N.A.

N. AVT D Mixed Bed R-26 N.A. Q N.A.

0. HCWT and LCWT R-27 N.A. Q N.A.

P. Resin Regeneration Tank R-28 N.A. Q N.A.

M. Nuclear Sample Room Wide Range Area Monitor R-33 N.A. Q N.A.

R. Containment Spray Pump Wide Range Area Monitor R-34 N.A. Q N.A.

S. PASS Panel Wide Range Area Monitor R-35 N.A. Q N.A.

While not an area monitor by strict definition, it serves as an indicator of internal leakage and provides an isolation signal for the internal component cooling system.

ODCM Page 52 Revision 15

4.0 RADWASTE TREATMENT 4.1 Liquid Radwaste Treatment System\

CONTROLS C.4.1 The liquid radwaste system shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent to UNRESTRICTED AREAS would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31 day period.

APPLICABILITY: At all times.

ACTION: With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System which could reduce the radioactive liquid waste discharged not in operation, prepare and submit to the Commission within 30 days a Special Report that includes the following information:

a.) Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability.

b.) Action(s) taken to return the inoperable equipment to OPERABLE status, and c.) Summary description of action(s) taken to prevent recurrence.

SURVEILLANCE REQUIREMENTS S.4.1.1 Doses due to liquid releases to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized.

S.4.1.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Controls C.1.1 and C.1.2.

BASES The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable. This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR 50, and the design objective given in Section II.D of Appendix I to 10 CFR 50. The specified limits governing the use of the appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix A to 10 CFR 50 for liquid effluents.

ODCM Page 53 Revision 15

4.2 Gaseous Radwaste Treatment System CONTROLS C.4.2 The Gaseous Radwaste Treatment System and the Ventilation Exhaust Treatment System shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases to areas at and beyond the SITE BOUNDARY would exceed:

a.) 0.2 mrad to air from gamma radiation, or b.) 0.4 mrad to air from beta radiation, or c.) 0.3 mrem to any organ of a MEMBER OF THE PUBLIC APPLICABILITY: At all times.

ACTION: With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days a Special Report that includes the following information:

a.) Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability, b.) Action(s) taken to restore the inoperable equipment to OPERABLE status, and c.) Summary description of action(s) taken to prevent recurrence.

SURVEILLANCE REQUIREMENTS S.4.2.1 Doses due to gaseous releases to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized.

S.4.2.2The installed GASEOUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM shall be considered OPERABLE by meeting Controls C.2.2, C.2.3.1, and C.2.3.2.

BASES 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 the 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 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 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C. of Appendix I to 10 CFR 50, for gaseous effluents.

ODCM Page 54 Revision 15

4.3 Solid Radioactive Waste CONTROLS C.4.3 The solid radwaste system shall be used as applicable in accordance with the Process Control Program for the solidification and packaging of radioactive waste to ensure meeting the requirements of 10 CFR 71 prior to shipment of radioactive wastes from the site.

APPLICABILITY: At all times.

ACTION: If the packaging requirements of 10 CFR 71 are not satisfied, suspend shipments of deficiently packaged solid radioactive wastes from the site until appropriate corrective measures have been taken.

4.4 Configuration Changes CONTROLS C.4.4 Major changes to the Radioactive Waste Treatment Systems, (Liquid, Solid, and Gaseous), shall be reported to the Commission by the inclusion of a suitable discussion or by reference to a suitable discussion of each change in the Annual Radioactive Effluent Release Report for the period in which the changes were made. Major changes to Radioactive Waste Treatment Systems, (Liquid, Gaseous and Solid), shall include the following:

(1) Changes in process equipment, components and structures from those in use (e.g., deletion of evaporators and installation of demineralizers);

(2) Changes in the design of Radioactive Waste Treatment Systems that could significantly alter the characteristics and/or quantities of effluents released; (3) Changes in system design which may invalidate the accident analysis (e.g.,

changes in tank capacity that would alter the curies released).

Note: The Radioactive Waste Treatment Systems, are those systems used to minimize the total activity released from the site.

Note: Changing the filters used, replacement resins or minor modifications (pipe or valve dimensions or manufacturers) due to maintenance activities would not be considered a major change.

APPLICABILITY: At all times.

ACTION: The discussion of each change shall contain:

(1) a summary, in accordance with 10 CFR 50.59, of the evaluation that led to the determination that the change could be made; (2) sufficient detailed information to support the reason for the change; (3) a detailed description of the equipment, components and processes involved and the interfaces with other plant systems; ODCM Page 55 Revision 15

(4) an evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents from those previously predicted; (5) an evaluation of the change which shows the expected maximum exposures to individuals in all UNRESTRICTED AREAS and to the MEMBERS OF THE PUBLIC from those previously estimated; (6) documentation of the fact that the change was reviewed and found acceptable by the Plant Operations Review Committee.

4.5 Process Control Program

a. The Process Control Program (PCP) shall be a document outlining the method of processing wet or dry solid wastes and for solidification of liquid wastes. It shall include the process parameters and evaluation methods used to assure meeting the requirements or 1 OCFR Part 71 prior to shipment of containers of radioactive waste from the site.

b. Licensee may make changes to the PCP and shall submit to the Commission with the Radioactive Effluent Release Report for the period in which any change(s) is made a copy of the new PCP and a summary containing:

(1) sufficiently detailed information to support the rationale for the change; (2) a determination that the change will not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and (3) documentation of the fact that the change has been reviewed and found acceptable by the onsite review function.

c. Licensee initiated changes shall become effective after review and acceptance by the Plant Operation Review Committee.

5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING 5.1 Monitoring Program CONTROLS C.5.1 The Radiological Environmental Monitoring Program (REMP) shall be conducted as specified in Table 5-1 at the locations given in Figures 5-1, 5-2, 5-3 and 5-4.

APPLICABILITY: At all times.

ACTION: If the radiological environmental monitoring program is not conducted as specified in Table 5-1, prepare and submit to the Commission, in the Annual Radiological 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 sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal availability, or to malfunction of automatic sampling equipment. If the latter, efforts shall be made to complete corrective action prior to the end of the next sampling period.

ODCM Page 56 Revision 15

Sampling periods for this specification are usually of one week duration. If continuous water sampling equipment is out of service, the 120 minute aliquot sampling period does not mean that grab samples must be taken every 120 minutes, but one grab sample once each week is sufficient until the automatic sampling equipment is restored to service.

ACTION: If the level of radioactivity as a result of plant effluents in an environmental sampling medium at one or more of the locations specified exceeds the reporting levels of Table 5-4 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from receipt of the laboratory analysis a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions 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 C.1.2, C.2.3.1, and C.2.3.2.

When more than one of the radionuclides in Table 5-4 are detected in the sampling medium, this report shall be submitted if:

concentration (1) + concentration (2) + *.o> 1.0 limit level (1) limit level (2)

When radionuclides other than those in Table 5-4 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC from all radionuclides is greater than the calendar year limit of Controls C.1.2, C.2.3.1, and C.2.3.2 This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

ACTION: With milk or fresh leafy vegetable samples unavailable from one or more of the sampling locations indicated on Figure 5-2, a discussion shall be included in the Annual Radiological Environmental Operating Report which identifies the cause for the unavailability of samples and identifies locations for obtaining replacement samples. If a milk or leafy vegetable sample location becomes unavailable, the location from which samples were unavailable may then be deleted provided that a comparable location is added to the radiological environmental monitoring program as described in the ODCM, unless no other sample location is available.

SURVEILLANCE REQUIREMENTS S.5.1 The radiological environmental samples shall be collected pursuant to Table 5-1 from the specific locations given in the table and figure(s) given in the ODCM, and shall be analyzed pursuant to the requirements of Table 5-1 and the detection capabilities required by Table 5-3.

ODCM Page 5 7 Revision 15

BASES The REMP 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 to MEMBERS OF THE PUBLIC resulting from plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR 50, and thereby supplements the RETS by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979.

ODCM Page 58 Revision 15

Table 5-1 Radiological Environmental Monitoring Program Page 1 of 2 EXPOSURE PATHWAY NUMBER OF SAMPLES SAMPLING AND TYPE AND FREQUENCY OF ANALYSIS AND/OR SAMPLE & SAMPLE LOCATIONS COLLECTION FREQUENCY (1)

1. AIRBORNE

a. Radioiodine 5 indicator Continuous operation of Radioiodine canister. Analyze within 7 days of collection for 1 control sampler with sample collection 1-131.

at least once per 10 days

b. Particulate 9 indicator Same as above Particulate sampler. Analyze for gross beta radioactivity 3 control > 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following filter change,(3) Perform gamma isotopic analysis on each sample for which gross beta activity is > 10 times the mean of offsite samples. Perform gamma isotopic analysis on composite (by location) sample at least once per 92 days.(4)

2. DIRECT 18 indicator TLDs at least quarterly Gamma dose quarterly.

RADIATION (2) 9 control 11 placed greater than 5 miles from plant site.

3. WATERBORNE

a. Surface(5) 1 control (Russell Composite* sample collected Gross beta and gamma isotopic analysis of each composite Station) over a period of < 31 days.(6) sample. Tritium analysis of one composite sample at least 1 indicator (Condenser once per 92 days.(4)

Water Discharge)

b. Drinking Same as above(6) 1 indicator (Ontario Same as above Water District Intake)

c. Shoreline 1 control (Russell Semi-annually Sediment Station) Gamma isotopic analysis of each sample(4) 1 indicator (Ontario Water District)

Composite sample to be collected by collecting an aliquot at intervals not exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

OF)CMV' Pagp 5q Rpvikion 1 5

Table 5-1 (continued) Page 2 of 2 Radiological Environmental Monitoring Program EXPOSURE PATHWAY NUMBER OF SAMPLES SAMPLING AND TYPE AND FREQUENCY OF ANALYSIS ANDIOR SAMPLE & SAMPLE LOCATIONS COLLECTION FREQU

4. INGESTION

a. Milk 1 control At least once per 15 days Gamma isotopic and 1-131 analysis of each 3 indicator sample.(4,7)

June thru October each of 3 farms 1 control At least once per 31 days Gamma isotopic and 1-131 analysis of each 1 indicator sample.(4,7)

November thru May one of the farms

b. Fish 4 control Twice during fishing season Gamma isotopic analysis on edible portions of 4 indicator (Off shore at including at least four species. each sample.(4,7)

Ginna)

Annual at time of harvest.

c. Food Products 1 control Sample from two of the Gamma isotopic analysis on edible portion of 2 indicator (On site) following: each sample.(4)

1. apples

2. cherries

3. Other crops grown on site by contract farmer 1 control 2 indicator At time of harvest. One sample Gamma isotopic analysis on edible portion of (Nearest offsite garden of: each sample.(4) within 5 miles in the 1. broad leaf vegetation

highest D/Q 2. other vegetable meteorological sector or onsite garden) *leaves from 3 different plant species nrncmO Page 60 Revision 15

Table Notation (1) Specific parameters of distance and direction sector from the centerline of the reactor, and additional description where pertinent, shall be provided for each and every sample location in Table 5-1 in a table and figures in the ODCM. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and malfunction of automatic sampling equipment. If specimens are unavailable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the next Annual Radiological Environmental Operating Report. It is recognized that, at times, it may not be possible or practicable to continue 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 appropriate substitutions made within 30 days in the radiological environmental monitoring program as described in the ODCM. Submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure(s) and table for the ODCM reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples for the pathway and justifying the selection of the new location(s) for obtaining samples.

(2) 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 purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in one packet are considered to be two or more dosimeters. Film badges shall not be used for measuring direct radiation.

The 39 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g. some sectors may be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

(3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 or more hours after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

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

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

(6) A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g. hourly) relative to the compositing period (e.g. monthly) in order to assure obtaining a representative sample.

(7) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the OCDM.

ODCM Page 61 Revision 15

5.2 Environmental Monitor Sample Locations distance to All sample locations are specified on Table 5-2, a list of direction and sample points. Indicator and control samples required by the environmental program are noted by an I or a C.

Figure 5-1 shows the onsite* sample locations for airborne particulates, radioiodine and direct radiation. Also indicated on Figure 5-1 is the onsite 2-7 vegetable garden, as well as the placement of post accident TLDs, locations and 13 - 24. TLD locations 2 - 7 are co-located with the air monitor samplers.

is the The onsite garden is located in the SE sector near the closest resident who real maximally exposed individual, rather than in the ESE sector which has the highest D/Q.

Figure 5-2 gives the location of the only milk herds within 5 miles of the plant. On this map is also included the Ontario Water District intake pumping station where lake water is sampled prior to treatment.

Figure 5-3 shows the offsite control locations for direct radiation as measured by TLDs.

Figure 5-4 shows the offsite sample locations for airborne particulates, and radioiodine. Sample stations 9 and 11 are situated near population centers, Webster and Williamson, located approximately 7 miles from the Ginna Site. TLD locations 8 - 12 are co-located with air monitor samplers.

Onsite refers to the area surrounding the Ginna Plant bounded by RG&E property lines. Offsite refers to the area beyond the immediate RG&E property.

ODCM Page 62 Revision 15

Table 5-2 Direction and Distance to Sample Points All directions given in degrees and all distances given in meters Air Sample Stations Direction Distance TLD Direction Distance Locations

2 1 87 320 #2 1 87 320

3 1 110 420 #3 1 110 420

4 1 140 250 #4 1 140 250

5 1 185 160 #5 1 185 160

6 1 232 225 #6 1 232 225

7 1 257 220 #7 1 257 220

8 C 258 19200 #8 C 258 19200

9 I 235 11400 #9 I 235 11400

10 C 185 13100 #10 C 185 13100

11 1 123 11500 # 11 1 123 11500

12 C 93 25100 # 12 C 93 25100

13 I 194 690 #13 I 292 230 Water Sample Locations Direction Distance # 14 I 292 770 Russell Station C 270 25600 # 15 I 272 850 Ontario Water Dist Intake I 70 2200 # 16 I 242 900 Circ Water Intake S 0 420 # 17 I 208 500 Circ Water Discharge I 15 130 #18 I 193 650 Deer Creek S 105 260 # 19 1 177 400

20 I 165 680

21 I 145 600

22 I 128 810

23 I 107 680

24 I 90 630

25 C 247 14350

26 C 223 14800 Milk Sample Locations Direction Distance # 27 C 202 14700 Farm A I 113 9500 #28 C 145 17700 Farm B I 242 5450 # 29 C 104 13800 Farm C I 156 4950 # 30 C 103 20500 Farm D C 132 21000 # 31 I 263 7280 Fish Samples # 32 I 246 6850 Indicator Samples Lake Ontario Discharge Plume # 33 I 220 7950 Control Samples Russell Station # 34 I 205 6850 Produce Samples # 35 I 193 7600 Indicator Samples Grown on property surrounding Plant # 36 I 174 5650 Control Samples Purchased from farms > 10 miles # 37 I 158 6000

38 I 137 7070

= Indicator Samples C Control or Background Samples #39 I 115 6630 S Supplemental Sample

40 1 1 87 6630 a

ODCM Page 63 Revision 15

Figure 5-1 Location of Onsite Air Monitors and Post Accident TLDs I

i

"* I I4 -

LAKE ONTAR IO Onsite Garden I -

I 1S0 -

ONSITE AIR HONITOR 0 POST- ACCIDENr TLD 0 200 400 600 Scale Meters S-.... "1II PJI I 4m, r A

Figure 5-2 Location of Farms for Milk Samples and Ontario Water District Intake

,East

\[_jJ Water Sample Station Clifton Si 10 KM Sample Station SMilk

"%M r -A 4 r ,--

Figure 5-3 Location of Offslte TLDs LI TLDs PERMANENTLY PLACED

-NIW onrht Pvirionn 1 r;

CD LL 0

0 0 C 4..0

__CD

Table 5-3 Detection Capabilities for Environmental Sample Analysis Lower Limit of Detection (LLD)

Airborne Food Shoreline Water Particulate Fish Milk Products Sediment Analysis (pCiIl) Or Gas (pCilkg) (pCill) (pCi/kg) (pCi/kg)

(pCi/rn) wet wet dry gross beta 4(a) 1 E-02 3-H 2000 (1 000)(a) 54-Mn 15 130 59-Fe 30 260 58, 60-Co 15 130 65-Zn 30 260 95-Zr-Nb 15(b) 131-1 1 7 E-02 1 60 134, 137-Cs 15(10)(a), 1 E-02 130 15 60 150 18 140-Ba-La 15(b) 15(b)

a. LLD for drinking water

b. Total for parent and daughter The LLD shall be calculated as described in Notation (a) to Table 1-1.

ODCM Page 68 Revision 15

Table 5-4 Reporting Levels for Radioactivity Concentrations in Environmental Samples Reporting Levels Airbourne Broad Leaf Analysis Water (pCi/I) Particulate or Fish Milk Vegetables Gas (pCi/m 3) (pCi/kg, (pCi/I) (pCi/kg, wet) wet)

H-3 2 E+04 Mn-54 1000 3 E+04 Fe-59 400 1 E+04 Co-58 1000 3 E+04 Co-60 300 1 E+04 Zn-65 300 2 E+04 Zr-Nb-95 400(a) 1-131 2 0.9 3 1 E+02 Cs-134 30 10 1 E+03 60 1 E+03 Cs-137 50 20 2 E+03 70 2 E+03 Ba-La-140 200(a) 1 300

a. Total for parent and daughter Decay correction in analysis of environmental samples is taken from the end of the sampling time not from the midpoint of the sample period.

12 ODCM Page 69 Revision 15

Table 5-5 D/Q and X/Q 5 Year Rolling Average 1995 - 1999 Plant Vent Distance to section boundary in meters:

804m 1609m 2416m 3218m 4022m 4827m 5632m 6436m 7240m 8045m D/Q Direction N 1.74E-09 8.20E-1 0 5.54E-10 3.36E-1 0 2.45E-10 1.85E-10 1.41E-10 1.13E-10 1.01E-10 3.86E-1 0 NNE 1.18E-09 6.28E-1 0 3.99E-1 0 2.75E-10 2.02E-10 1.53E-10 1.29E-10 9.54E-1 1 8.50E-1 1 2.26E-10 NE 1.74E-09 1.84E-06 6.26E-1 0 3.86E-1 0 2.83E-10 2.14E-10 1.64E-10 1.32E-10 1.09E-10 9.17E-1 1 ENE 2.99E-09 1.43E-09 8.56E-1 0 5.76E-1 0 4.25E-1 0 3.14E-10 2.39E-10 1.91E-10 1.58E-10 1.32E-10 E 5.11E-09 2.20E-09 1.23E-09 7.96E-1 0 5.69E-1 0 4.17E-10 5.09E-1 0 6.34E-10 4.74E-10 4.OOE-10 ESE 7.41 E-09 3.19E-09 1.67E-09 1.13E-09 9.34E-10 9.18E-10 7.27E-10 5.16E-10 4.26E-1 0 3.54E-10 SE 4.14E-09 1.93E-09 9.91 E-1 0 7.32E-1 0 7.05E-10 5.40E-1 0 4.OOE-1 0 3.05E-1 0 2.52E-10 2.09E-1 0 SSE 1.32E-09 6.71 E-1 0 3.72E-1 0 2.68E-10 2.58E-10 1.88E-10 1.38E-10 2.76E-1 0 8.94E-1 1 7.48E-1 1 S 2.15E-09 1.29E-09 7.37E-1 0 6.54E-1 0 4.95E-1 0 3.58E-1 0 2.61 E-1 0 2.02E-1 0 1.67E-1 0 1.39E-10 SSW 2.57E-09 1.48E-09 8.43E-1 0 5.50E-1 0 4.OOE-10 3.95E-1 0 2.87E-10 2.22E-1 0 1.83E-10 1.52E-10 SW 2.88E-09 1.53E-09 8.50E-1 0 5.66E-1 0 4.79E-10 4.41 E-1 0 3.20E-1 0 2.49E-1 0 2.05E-1 0 1.71E-10 WSW 2.21 E-09 1.18E-09 6.93E-10 4.73E-1 0 3.57E-1 0 3.04E-1 0 4.38E-10 3.39E-1 0 2.80E-10 2.33E-10 W 9.54E-1 0 5.40E-1 0 3.27E-1 0 2.21 E-10 1.61E-10 1.20E-10 1.76E-10 2.71 E-1 0 2.25E-10 1.87E-10 WNW 1.29E-10 9.58E-1 1 6.87E-1 1 4.91 E-11 1.18E-10 2.83E-1 1 2.23E-1 I 1.82E-11 1.51 E-11 1.27E-11 NW 4.80E-1 0 3.03E-10 2.03E-1 0 1.41 E-10 1.05E-10 8.01E-11 6.25E-1 1 5.05E-11 4.20E-1 1 3.52E-1 1 NNW 1.37E-09 7.06E-1 0 4.40E-10 3.01E-10 2.21 E-1 0 1.73E-10 1.29E-10 1.03E-1 0 8.59E-1 1 7.19E-11 X/Q N 8.56E-08 9.42E-08 9.19E-08 8.1OE-08 6.99E-08 6.15E-08 5.38E-08 5.41 E-08 6.17E-08 1.20E-07 NNE 7.17E-08 8.06E-08 8.23E-08 7.45E-08 7.21 E-08 5.83E-08 5.28E-08 4.81 E-08 6.98E-08 1.33E-07 NE 8.27E-08 9.48E-08 9.36E-08 8.33E-08 7.23E-08 6.94E-08 5.63E-08 5.05E-08 4.57E-08 4.18E-08 ENE 1.05E-07 1.16E-07 1.06E-07 8.89E-08 7.41 E-08 6.26E-08 5.35E-08 4.66E-08 4.13E-08 3.70E-08 E 1.91 E-07 1.81 E-07 1.53E-07 1.16E-07 9.09E-08 7.32E-08 8.82E-08 7.67E-08 6.51 E-08 5.61 E-08 ESE 2.43E-07 2.13E-07 1.70E-07 1.35E-07 1.11E-07 9.27E-08 7.19E-08 5.86E-08 4.96E-08 4.27E-08 SE 1.47E-07 1.38E-07 1.15E-07 1.12E-07 9.67E-08 7.43E-08 5.79E-08 5.33E-08 5.21 E-08 3.44E-08 SSE 6.06E-08 6.56E-08 5.66E-08 5.38E-08 4.55E-08 3.40E-08 2.64E-08 2.16E-08 1.83E-08 1.58E-08 S 1.06E-07 1.49E-07 1.27E-07 9.80E-08 7.1OE-08 5.27E-08 4.09E-08 3.34E-08 2.83E-08 2.42E-08 SSW 1.06E-07 1.59E-07 1.54E-07 1.04E-07 7.61 E-08 6.96E-08 5.35E-08 4.35E-08 3.68E-08 3.16E-08 SW 1.06E-07 1.39E-07 1.43E-07 1.18E-07 1.01E-07 9.76E-08 7.60E-08 6.22E-08 5.27E-08 4.53E-08 WSW 1.13E-07 1.40E-07 1.33E-07 1.23E-07 1.20E-07 1.30E-07 1.47E-07 1.20E-07 1.02E-07 8.78E-08 W 7.19E-08 1.07E-07 9.56E-08 7.99E-08 6.66E-08 5.67E-08 9.77E-08 9.14E-08 7.77E-08 6.68E-08 WNW 6.07E-09 1.64E-08 1.96E-08 1.87E-08 1.68E-08 1.49E-08 1 .33E-08 1.20E-08 1.08E-08 9.88E-09 NW 1.99E-08 3.49E-08 3.64E-08 3.24E-08 2.80E-08 2.42E-08 2.11E-08 1.86E-08 1.66E-08 1.50E-08 NNW 6.23E-08 6.98E-08 6.67E-08 5.74E-08 4.86E-08 4.15E-08 3.58E-08 3.20E-08 2.80E-08 2.53E-08 ODCM v Page 70 Revision 15

Table 5-6 D/Q and X/Q 5 Year Rolling Average 1995- 1999 Containment Vent Distance to section boundary in meters:

804m 1609mn 2416in 3218in 4022m 4827m 56321n 6436ni 7240mi 8045ni D/Q Direction N 1.88E-08 5.95E-09 2.88E-09 1.85E-09 1.31 E-09 9.45E-10 6.86E-10 5.31 E- 10 4.42E-10 3.90E-10 NNE 1.86E-08 5.88E-09 2.85E-09 1.83E-09 1.29E-09 9.35E-10 6,79E-10 5.25E-1 0 4.39E-10 3.90E-10 NE 1.99E-08 6.30E-09 3.05E-09 1.96E-09 1.38E-09 1.OOE-09 7.27E-1 0 5.62E-10 4,64E-10 3.86E-10 ENE 1.98E-08 6.28E-09 3.04E-09 1.95E-09 1.38E-09 1.08E-09 7.24E-10 5.60E-10 4.62E-10 3.84E-1 0 E 1.99E-08 6.30E-09 3.05E-09 1.96E-09 1.38E-09 1.OOE-09 7.41 E-1 0 5.75E-10 4.75E-10 3.95E-10 ESE 1.78E-08 5.66E-09 2.74E-09 1.77E-09 1.27E-09 9.19E-10 6.67E-1 0 5.16E-10 4.11E-10 3.54E-10 SE 1.01E-08 3.23E-09 1.57E-09 1.05E-09 7.51 E-1 0 5.43E-1 0 3.94E-10 3.05E-10 2.52E-10 2.09E-10 SSE 3.66E-09 1.18E-09 5.75E-10 3.92E-10 2.85E-10 2.06E-1 0 1.50E-10 1.16E-10 9.56E-1 1 7.94E-1i1 S 6.65E-09 2.14E-09 1.07E-09 7.06E-10 4,99E-10 3.60E-1 0 2.62E-10 2.02E-10 1.67E-10 1.39E-10 SSW 7.05E-09 2.28E-09 1.17E-09 7.53E-10 535E-10 3.95E-10 2.87E-10 2.22E-10 1.83E-10 1.52E-10 SW 7.77E-09 2.50E-09 1.22E-09 7.94E-10 5.98E-10 4.43E-10 3.22E-10 2.49E-10 2.05E-10 1.71 E-10 WSW 1.04E-08 3 32E-09 1.61 E-09 1.04E-09 7.44E-10 5.64E-1 0 4.39E-10 3.39E-10 2.80E-10 2.33E-10 W 8,42E-09 2.68E-09 1.30E-09 8.33E-10 5.89E-1 0 4.27E-10 3.46E-10 2.72E-10 2.25E-10 1.87E-10 WNW 2.68E-09 1.18E-09 4.16E-10 2.67E-10 1.89E-10 1.36E-10 9.92E-1 1 7.67E-1 1 6.34E-1 1 5.27E-1 1 NW 5.20E-09 1.66E-09 8,05E-10 5.16E-10 3.65E-10 2.64E-10 1.92E-10 1.48E-10 1.23E-10 1.02E-10 NNW 1.13E-08 3.58E-09 1.74E-09 1.12E-09 7.88E-10 5.70E-10 4.14E-10 3.20E-10 2.65E-10 2.20E-1 0 X/Q N 1.73E-06 6.24E-07 3.58E-07 2.44E-07 1.52E-07 1.42E-07 1.15E-07 9.67E-08 9.34E-08 9.86E-08 NNE 2.15E-06 7.57E-07 4.37E-07 3.01E-07 2.26E-07 1.78E-07 1.46E-07 1.24E-07 1.26E-07 1.48E-07 NE 1.94E-06 7.OOE-07 3.99E-07 2.70E-07 2.OOE-07 1.55E-07 1.25E-07 1.05E-07 9.02E-08 7.88E-08 ENE 1.20E-06 4.40E-07 2.46E-07 1.64E-07 1.19E-07 9.14E-08 7.26E-08 6.03E-08 5.17E-08 4.50E-08 E 1.05E-06 3.91E-07 2.18E-07 1.44E-07 1.03E-07 7.84E-08 6.58E-08 5.39E-08 4.59E-08 3.96E-08 ESE 8.27E-07 3.15E-07 1.83E-07 1.24E-07 8.99E-08 6.76E-08 5.27E-08 4.32E-08 3.67E-08 3.16E-08 SE 5.82E-07 2.44E-07 1.56E-07 1.17E-07 8.36E-08 6.27E-08 4.88E-08 4.OOE-08 3.39E-08 2.92E-08 SSE 3.27E-07 1.42E-07 8.76E-08 6.27E-08 4.44E-08 3.31 E-08 2.57E-08 2.09E-08 1.77E-08 1.52E-08 S 5.09E-07 2.29E-07 1.40E-07 8.96E-08 6.92E-08 4.71E-08 3.65E-08 2.98E-08 2,52E-08 2.16E-08 SSW 4.64E-07 2.44E-07 1.61 E-07 1.03E-07 7.31 E-08 5.49E-08 4.27E-08 3.49E-08 2.95E-08 2.54E-08 SW 4.99E-07 2,52E-07 1.95E-07 1.36E-07 1.OOE-07 7.59E-08 5.94E-08 4.87E-08 4.13E-08 3.56E-08 WSW 9.88E-07 3.99E-07 2.57E-07 1.99E-07 1.61E-07 1.37E-07 1.11E-07 9.16E-08 7.79E-08 6.73E-08 W 9.24E-07 3.62E-07 2.15E-07 1.49E-07 1.1OE-07 8.62E-08 8.29E-08 6.83E-08 5.82E-08 5.03E-08 WNW 3.25E-07 1.26E-07 7.51 E-08 5.22E-08 3.92E-08 3.08E-08 2.51E-08 2.11 E-08 1.83E-08 1.60E-08 NW 5,27E-07 1.98E-07 1.14E-07 7.80E-08 5.78E-08 4.50E-08 3.62E-08 3.03E-08 2.62E-08 2.29E-08 NNW 9.39E-07 3.46E-07 1.98E-07 1.34E-07 9.89E-08 7.65E-08 6.13E-08 5.12E-08 4.41 E-08 3.85E-08 ODCM " Pane 71 Revision 19

Table 5-7 D/Q andX/Q 5 Year Rolling Average 1995 - 1999 Air Ejector Distance to section boundary in meters:

804m 1609m 2416m 3218m 4022m 4827m 5632m 6436m 7240m 8045m D/Q Direction N 1.74E-09 8.20E-10 5.54E-10 3.36E-1 0 2.45E-10 1.85E-10 1.41 E-1 0 1.13E-10 1.01E-10 3.86E-1 0 N 2.02E-08 6.38E-09 3.09E-09 1.98E-09 1.40E-09 1.01E-09 7.34E-1 0 5.68E-10 4.69E-10 3.90E-10 NNE 2.07E-08 6.55E-09 3.17E-09 2.03E-09 1.44E-09 1.04E-09 7.54E-10 5.83E-10 4.81 E-10 4.00E-10 NE 2.11E-08 6.66E-09 3.22E-09 2.07E-09 1.46E-09 1.06E-09 7.67E-10 5.93E-10 4.89E-1 0 4.07E-10 ENE 2.05E-08 6.49E-09 3.14E-09 2.01E-09 1.42E-09 1.03E-09 7.47E-10 5.77E-10 4.77E-10 3.96E-10 E 2.04E-08 6.46E-09 3.13E-09 2.01E-09 1.42E-09 1.02E-09 7.43E-10 5.75E-1 0 4.75E-10 3.95E-10 ESE 1.84E-08 5.80E-09 2.81 E-09 1.80E-09 1.27E-09 9.19E-10 6.67E-10 5.16E-10 4.26E-10 3.54E-1-0 SE 1.08E-08 3.43E-09 1.66E-09 1.06E-09 7.51E-10 5.43E-10 3.94E-10 3.05E-1 0 2.52E-10 2.09E-1 0 SSE 4.12E-09 1.30E-09 6.30E-1 0 4.04E-10 2.85E-10 2.06E-10 1.50E-10 1.16E-10 9.56E-1 1 7.94E-11 S 7.19E-09 2.27E-09 1.1OE-09 7.06E-10 4.99E-10 3.60E-1 0 2.62E-10 2.02E-1 0 1.67E-10 1.39E-10 SSW 7.89E-09 2.49E-09 1.21 E-09 7.74E-10 5.47E-10 3.95E-10 2.87E-10 2.22E-10 1.83E-10 1.52E-10 SW 8.85E-09 2.80E-09 1.35E-09 8.68E-1 0 6.13E-10 4.43E-10 3.22E-10 2.49E-10 2.05E-10 1.71E-10 WSW 1.21E-08 3.82E-09 1.85E-09 1.18E-09 8.37E-10 6.05E-10 4.39E-10 3.39E-10 2.80E-10 2.33E-10 W 9.68E-09 3.06E-09 1.48E-09 9.49E-10 6.71 E-10 4.85E-10 3.52E-10 2.72E-10 2.25E-10 1.87E-10 WNW 3.28E-09 1.04E-09 5.54E-10 3.22E-10 2.51E-10 1.64E-10 1.19E-10 9.22E-1 I 7.62E-11 6.33E-11 NW 5.88E-09 1.86E-09 8.99E-1 0 5.77E-10 4.07E-10 2.94E-10 2.14E-10 1.65E-10 1.37E-10 1.13E-10 NNW 1.22E-08 3.84E-09 1.86E-09 1.19E-09 8.43E-1 0 6.09E-10 4.42E-10 3.42E-1 0 2.82E-10 2.35E-10 X/Q N 2.34E-06 8.13E-07 4.56E-07 3.06E-07 2.24E-07 1.72E-07 1.37E-07 1.13E-07 9.72E-08 8.43E-08 NNE 3.01 E-06 1.02E-06 5.81 E-07 3.94E-07 2.91 E-07 2.25E-07 1.80E-07 1.49E-07 1.28E-07 1.11E-07 NE 2.48E-06 8.70E-07 4.88E-07 3.27E-07 2.40E-07 1.84E-07 1.47E-07 1.21E-07 1.04E-07 9.01E-08 ENE 1.51E-06 5.37E-07 2.94E-07 1.92E-07 1.39E-07 1.05E-07 8.25E-08 6.79E-08 5.79E-08 5.01E-08 E 1.28E-06 4.52E-07 2.44E-07 1.58E-07 1.13E-07 8.50E-08 6.65E-08 5.46E-08 4.65E-08 4.01E-08 ESE 9.59E-07 3.28E-07 1.75E-07 1.13E-07 8.09E-08 6.08E-08 4.75E-08 3.90E-08 3.31 E-08 2.85E-08 SE 7.73E-07 2.65E-07 1.42E-07 9.20E-08 6.57E-08 4.95E-08 3.87E-08 7.93E-05 2.70E-08 2.33E-08 SSE 4.47E-07 1.54E-07 8.18E-08 5.27E-08 3.75E-08 2.80E-08 2.18E-08 1.79E-08 1.52E-08 1.68E-08 S 6.59E-07 2.27E-07 1.21E-07 7.75E-08 5.49E-08 4.11E-08 3.20E-08 2.61 E-08 2.21 E-08 1.90E-08 SSW 6.43E-07 2.22E-07 1.19E-07 7.73E-08 5.52E-08 4.16E-08 3.25E-08 2.67E-08 2.26E-08 1.95E-08 SW 7.75E-07 2.65E-07 1.45E-07 9.61 E-08 6.96E-08 5.31 E-08 4.19E-08 3.46E-08 2.95E-08 2.55E-08 WSW 1.49E-06 5.11E-07 2.86E-07 1.91 E-07 1.40E-07 1.08E-07 8.58E-08 7.11E-08 6.08E-08 5.27E-08 W 1.29E-06 4.52E-07 2.51 E-07 1.67E-07 1.22E-07 9.30E-08 7.37E-08 6.09E-08 5.21E-08 4.51 E-08 WNW 5.27E-07 1.88E-07 1.05E-07 6.99E-08 5.10E-08 3.91 E-08 3.10E-08 2.56E-08 2.19E-08 1.90E-08 NW 7.90E-07 2.79E-07 1.54E-07 1.02E-07 7.39E-08 5.63E-08 4.45E-08 3.67E-08 3.14E-08 2.71E-08 NNW 1.28E-06 4.51 E-07 2.49E-07 1.64E-07 1.19E-07 9.06E-08 7.15E-08 5.90E-08 5.04E-08 4.36E-08 ODCMo Page 72 Revision 15

5.3 Land Use Census CONTROL C.5.3 A Land Use Census shall be conducted annually between June 1 and October 1, and shall identify within a distance of 5 miles the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden of greater than 500 square feet producing broad leaf vegetation. (In lieu of a garden census, broad leaf vegetation sampling of at least three different kinds of vegetation may be performed in an onsite garden located in the meteorological sector with the highest average annual growing season deposition parameter (D/Q) OR another location with a higher D/Q than the location of the maximally exposed individual.)

APPLICABILITY: At all Times.

ACTION: With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Surveillance S.2.3.2.1 of the ODCM, identify the new location(s) in the next Annual Radioactive Effluent Release Report.

ACTION: With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Control C.5.1, add the new location(s) within 30 days to the REMP described in the ODCM, if permission from the owner to collect samples can be obtained and sufficient sample volume is available. The sampling location(s), excluding Control location(s), having the lowest calculated dose or dose commitment(s),

via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted.

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

SURVEILLANCE REQUIREMENTS S.5.3 The Land Use Census shall be conducted between June 1 and October 1 of each year using a method that will best provide the necessary information such as by door-to-door survey, vehicular survey, aerial survey, or by consulting local agricultural authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Control C.6.1 of the ODCM.

BASES This specification is provided to ensure that changes in the use of areas at or beyond the SITE BOUNDARY are identified and that modifications to the REMP given in the ODCM are made if required by the results of this census. Information from methods such as the door-to door survey, vehicular survey, aerial survey, or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR 50. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored, since a garden of this size is the minimum ODCM Page 73 Revision 15

required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1)20% of the garden was used for growing broad yield leaf vegetation (i.e.similar to lettuce and cabbage), and (2) there was a vegetation of 2kg/M 2.

5.4 Interlaboratory Comparison Program CONTROL C.5.4 Analyses shall be performed on all radioactive materials supplied as part of an Interlaboratory Comparison Program, that correspond to samples required by the REMP, and that has been approved by the Commission, if such a program exists.

APPLICABILITY: At all times.

ACTION: With analyses not performed as required above, report the corrective actions taken to prevent recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Control C.6.1 of the ODCM.

SURVEILLANCE REQUIREMENTS S.5.4.1 The Interlaboratory Comparison Program is described in and implemented by procedure CHA-QC-INTERLAB. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Control C.6.1 of the ODCM.

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

ODCM Page 74 Revision 15

6.0 REPORTING REQUIREMENTS 6.1 Annual Radiological Environmental Operating Report An Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 15 of each year.

The Annual Radiological Environmental Operating Report shall include summaries, interpretations, and analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with background (control) samples and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The report shall also include the results of the Land Use Census as required.

This report shall include any new location(s) identified by the Land Use Census which yield a calculated dose or dose commitment greater than those forming the basis of Control C.5.1 The report shall also contain a discussion which identifies the causes of the unavailability of milk or leafy vegetable samples and identifies locations for obtaining replacement samples in accordance with Control C.5.1 The Annual Radiological Environmental Operating Report shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the tables and figures of Section 5.0 of the ODCM, the summarized and tabulated results of these analyses and measurements shall be in the format of Table 6-1, derived from the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.

The report shall also include the following:

a. a summary description of the radiological environmental monitoring program including a map of all sampling locations keyed to a table giving distances and directions from the reactor centerline; and

b. the results of the licensee participation in an Interlaboratory Comparison Program, and the corrective actions taken if the specified program is not being performed as required by Control C.5.4.

c. a discussion of all deviations from the sampling schedule specified in Table 5-1.

d. a discussion of any environmental sample measurements that exceed the reporting levels but are not the result of plant effluents, as required in the second ACTION of C.5.1.

e. a discussion of all analyses in which the required LLD was not achievable.

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6.2 Annual Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year of operation shall be submitted prior to May 15 each year. This report shall include a summary, on a quarterly basis, of the quantities of radioactive liquid and gaseous effluents and solid waste released as outlined in Regulatory Guide 1.21, Revision 1, with data summarized on a quarterly basis following the format of the Appendix thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: classes of solid wastes (as defined by 10 CFR 61), type of container (e.g. LSA, Type A, Type B, etc.) and solidification agent or absorbent (e.g., Portland cement).

The Radioactive Effluent Release Report shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each of the previous four calendar quarters as outlined in Regulatory Guide 1.21, Revision 1. In addition, the site boundary maximum noble gas gamma air and beta air doses shall be evaluated. The assessment of radiation doses shall be performed in accordance with controls 1.2 and 2.3. This same report shall include an annual summary of hourly meteorological data collected over the previous calendar year. Alternatively, the licensee has the option of retaining this summary on site in a file that shall be provided to the NRC upon request. The Radioactive Effluent Release Report shall include a discussion which identifies the circumstances which prevented any required detection limits for effluent sample analyses being met. This report shall also include an assessment of the radiation doses from radioactive gaseous and liquid effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY during the report period. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the ODCM.

The Annual Radioactive Effluent Release Report shall also include an assessment of radiation doses to the likely maximum exposed MEMBER OF THE PUBLIC from reactor releases, including doses from effluent releases and direct radiation, for the previous calendar year to demonstrate compliance with 40 CFR 190.

This report shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

This report shall include any changes made during the reporting period to the Offsite Dose Calculation Manual (ODCM). Licensee may make changes to this ODCM and shall submit to the Commission, with the Radioactive Effluent Release Report for the period in which any change(s) is made, a copy of the new ODCM and a summary containing:

a. sufficiently detailed information to support the rationale for the change;

b. a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c. documentation of the fact that the change has been reviewed and found acceptable by the Plant Operations Review Committee.

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Licensee initiated changes shall become effective after review and acceptance by the Plant Operations Review Committee on a date specified by the licensee.

This report shall include any changes made during the reporting period to the Process Control Program (PCP). This report shall include a discussion of any major changes to the radioactive waste treatment systems.

The Radiological Environmental Operating Report and the Annual Radioactive Effluent Release Report will be prepared and submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555 and a copy to the Regional Administrator of the USNRC, Region I.

6.3 Special Reports Guidance is given for each of these reports in the applicable Control. The following general guidelines are included here for calculating dose to an exposed individual or the MEMBERS OF THE PUBLIC for preparation of Special Reports:

a. The maximally exposed real MEMBER OF THE PUBLIC will generally be the same individual considered in the ODCM.

b. Dose contributions to the maximally exposed individual need only be considered to be those resulting from the Ginna plant itself. All other uranium fuel cycle facilities or operations are of sufficient distance to contribute a negligible portion of the individual's dose.

c. For determining the total dose to the maximally exposed individual from the major gaseous and liquid effluent pathways and from direct radiation, dose evaluation techniques used in preparing the Special Report will be those described in the ODCM, or other applicable methods where appropriate.

d. The contribution from direct radiation may be estimated by effluent dispersion modeling or calculated from the results of the environmental monitoring program for direct radiation.

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Table 6-1 Environmental Radiological Monitoring Program Summary ROCHESTER GAS AND ELECTRIC CORPORATION R.E. GINNA NUCLEAR POWER PLANT - DOCKET NO. 50-244 WAYNE, NEW YORK LOCATION WITH HIGHEST ANNUAL MEAN PATHWAY SAMPLED TYPE AND TOTAL NUMER INDICATOR LOCATIONS UNIT OF MEASUREMENT CONTROL LOCATIONS OF ANALYSES LLD MEAN (1) RANGE NAME, DISTANCE MEAN (1) RANGE MEAN (1) RANGE AND DIRECTION AIR: Particulate Gross Beta

..................... p.C...uM). G m Sn......................

Iodine Gamma Scan DIRECT RADIATION:

TLD Gamma (mrem/QUARTER)

WATER: Drinking Gross Beta (pCi/Liter) Gamma Scan Iodine Surface Gross Beta (pCi/liter) Gamma Scan Iodine Rainfall Gross Beta (pCi/m 2/day)

MILK: Iodine (pCi/Liter) Gamma Scan FISH: Gamma Scan (pCi/Kg)

VEGETATION: Gamma Scan (pCi/Kg) I (1) Mean and range based on detectable measurements only. Fraction of detectable measurements at specified locations in parentheses.

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7.0 REFERENCES

License

1. R. E. Ginna Nuclear Power Plant Unit No. 1, Appendix A to Operating No. DPR-1 8, Technical Specifications, Rochester Gas and Electric Corporation, Docket 50-244

2. USNRC, Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, NUREG-0133 (October, 1978).

3. USNRC, Calcu!ation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1, Regulatory Guide 1.109, Revision 1 (October 1977).

4. R. E. Ginna Nuclear Power Plant, Updated Final Safety Analysis Report.

5. R. E. Ginna Nuclear Power Plant, Calculations to Demonstrate Compliance with the Design Objectives of 10 CFR Part 50, Appendix 1, Rochester Gas and Electric Corporation, (June, 1977).

6. USNRC, Methods for Estimating Atmospheric Transport and dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, Regulatory Guide 1.111, Revision 1 (July, 1977).

7. R. E. Ginna Nuclear Power Plant, Incident Evaluation, Ginna Steam Generator Tube Failure Incident January 25, 1982, Rochester Gas and Electric Corporation, (April 12, 1982).

8. Pelletier, C. A., et .al., Sources of Radioiodine at Pressurized Water Reactors, EPRI NP-939 (November 1978).

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Rochester Gas and Electric Corporation Inter-Office Correspondence July 16, 2001

Subject:

Offsite Dose Calculation Manual (ODCM) Revision 16 To: Distribution Attached is the revised Ginna Station Offsite Dose Calculation Manual (ODCM).

The changes included within Revision 16 are summarized as follows:

1) add additional requirements for R1 5A being inoperable

2) add surveillance for R29 and R30

3) add data for air ejector releases

4) surveillance clarification

5) fix previous errors Revision 16 of the ODCM is considered effective July 19, 2001. Instructions for the necessary changes to the ITS for your controlled copy of the ODCM are as follows:

Volume Section Remove Insert III ODCM ODCM Rev. 15 ODCM Rex'. 16 Please contact Tom Harding (extension 3384) if you have any questions.

Thomas L. Harding

ODCM R.E. Ginna Nuclear Power Plant Offsite Dose Calculation Manual ODCM Revision 16 Responsible Manager: S PORC Review Date: 7-/2 o/

Effective Date: 7-/2 c /

Controlled Copy No.

Record Category: 4.43.5

TABLE OF CONTENTS DEFINITIO NS .................................................................... I 1.0 LIQUID EFFLUENTS ......................................................... 5 1.1 Concentration ......................................................... 5 1.2 Dose (10 CFR 50 Appendix I) .............................................. 6 1.3 Dose (40 CFR Part 190) .................................................. 7 1.4 Liquid Effluents Release Points .......................................... 12 1.5 Liquid Effluents Monitor Setpoints ........................................ 12 1.6 Liquid Effluent Release Concentrations .................................... 16 1.7 Liquid Effl uent Dose ................ .................................. 17 2.0 GASEOUS EFFLUENTS ..................................................... 20 2.1 C oncentration ........................................................ 20 2.2 Dose R ate .......................................................... 20 2.3 Dose (10 CFR Part 50, Appendix I) ........................................... 22 2.4 Dose (40 CFR Part 190) . ................................................ 25 2.5 Gaseous Effluent Release Points ........................................ 29 2.6 Gaseous Effluent Monitor Setpoints ...................................... 29 2.7 Gaseous Effluent Dose Rate ............................................ 35 2.8 Gaseous Effluent Doses ................................................ 37 3.0 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION .................... 42 3.1 Liquid Effl uent Monitors ................................................ 42 3.2 Gaseous Effluent Monitors ............................................. 45 3.3 Radiation Accident Monitoring Instrumentation .............................. 49 3.4 Area Radiation Monitors ............................................... 53 4.0 RADWASTE TREATMENT ................................................... 54 4.1 Liquid Radwaste Treatment System ...................................... 54 4.2 Gaseous Radwaste Treatment System .................................... 55 4.3 Solid Radwaste System ................................................ 56 4.4 Configuration Changes ................................................ 56 4.5 Process Control Program .............................................. 57 5.0 RADIOLOGICAL ENVIRONMENTAL MONITORING .............................. 58 5.1 Monitoring Program ................................................... 58 5.2 Environmental Monitor Sample Locations .................................. 63 5.3 Land Use Census .................................................... 74 5.4 Interlaboratory Comparison Program ..................................... 75 6.0 REPORTING REQUIREMENTS............................................76 6.1 Annual Radiological Environmental Operating Report ........................ 76 6.2 Radioactive Effluent Release Report ...................................... 77 6.3. S pecial R eports ...................................................... 78

7.0 REFERENCES

................. ...................................... 80

LIST OF TABLES Table 1-1 Radioactive Liquid Waste Sampling and Analysis Program .................... .9 Table 2-1 Radioactive Gaseous Waste Sampling and Analysis Program ................. 27 Table 2-2 Meteorological Data and Locations of Receptors for Set Point Calculations ....... 34 Table 2-3 Dose Factors For Noble Gases and Daughters ............................. 39 Table 2-4 Dose Parameters for Radionuclides and Radioactive Particulate, Gaseous Effluents 40 Table 2-5 Pathway Dose Factors Due to Radionuclides Other Than Noble Gases * ........ 41 Table 3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation ....................... 43 Table 3.1-2 Radioactive Liquid Effluent Monitoring Surveillance Requirements .............. 44 Table 3.2-1 Radioactive Gaseous Effluent Monitoring Instrumentation ..................... 46 Table 3.2-2 Radioactive Gaseous Effluent Monitoring Surveillance Requirements ............ 48 Table 3.3-1 Radiation Accident Monitoring Instrumentation .............................. 50 Table 3.3-2 Radioactive Accident Monitoring Surveillance Requirements ................... 52 Table 5-1 Radiological Environmental Monitoring Program ............................ 60 Table 5-2 Direction and Distance to Sample Points .................................. 64 Table 5-3 Detection Capabilities for Environmental Sample Analysis ....................

69 Table 5-4 Reporting Levels for Radioactivity Concentrations in Environmental Samples .....

70 Table 5-5 D/Q and X/O 5 Year Rolling Average 1995 - 1999 - Plant Vent .................

71 Table 5-6 D/Q and X/Q 5 Year Rolling Average 1995 - 1999 - Containment Vent ...........

72 Table 5-7 D/Q and X/Q 5 Year Rolling Average 1995 - 1999 - Air Ejector ................. 73 Table 6-1 Environmental Radiological Monitoring Program Summary .................... 79 ii

LIST OF FIGURES Figure 5-1 Location of Onsite Air Monitors and Post Accident TLDs ...................... 65 Figure 5-2 Location of Farms for Milk Samples and Ontario Water District Intake ........... 66 Figure 5-3 Location of Offsite TLDs ............................................... 67 Figure 5-4 Location of Offsite Air Monitors .......................................... 68 iii

DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these controls.

Terms used in these controls and not defined herein have the same definition as listed in the Technical specifications and/or the Technical Reference Manual. If a conflict in definition exists, the definition in the Technical Specifications takes precedence.

ACTION ACTION shall be that part of a Control that prescribes required actions to be taken under designated conditions, within specified completion times.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock display, and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1 134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, 1977.

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FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined as follows:

NOTATION FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

N/A Not applicable P Completed prior to each release FUNCTIONAL TEST An FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock display and/or trip functions. The FUNCTIONAL TEST shall include adjustments, as necessary, of the alarm, interlock display and/or Trip Setpoints such that the setpoints are within the required range and accuracy.

LOWER LIMIT OF DETECTION The LOWER LIMIT OF DETECTION (LLD) is the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95%

probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. The LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori(after the fact) limit for a particular measurement, the minimum detectable activity (MDA).

MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

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OFFSITE DOSE CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program (REMP). The ODCM shall also contain descriptions of the Radioactive Effluent Controls and Radiological Environmental Monitoring Program and descriptions of the information that shall be included in the Annual Radiological Environmental Operating Report and the Annual Radioactive Effluent Release Report.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s)and when all necessary attendant instrumentation, controls, electric power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component , or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1-1 of Technical Specifications.

PURGE - PURGING PURGE or PURGING shall be any 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.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

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.

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SURVEILLANCE REQUIREMENT SURVEILLANCE REQUIREMENTS shall be met during the OPERATIONAL MODES or other conditions specified for individual CONTROLS unless otherwise stated in an individual SURVEILLANCE REQUIREMENT. Each SURVEILLANCE REQUIREMENT shall be performed within the specified time interval with:

1. A maximum allowable extension not to exceed 25% of the surveillance interval, but

2. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.

UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM shall be 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 iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents.

Engineered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, 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.

WASTE GAS HOLDUP SYSTEM A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

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1.0 RADIOACTIVE LIQUID EFFLUENTS 1.1 Concentration (10 CFR 20)

CONTROLS C. 1.1 The release of radioactive liquid effluents shall be such that the concentration in the circulating water discharge does not exceed ten times the concentration values specified in Appendix B, Table 2, Column 2 to 10 CFR Part 20.1001 - 20.2402. For dissolved or entrained noble gases, the total activity due to dissolved or entrained noble gases shall not exceed 2 E-04 pCi/ml.

APPLICABILITY: At all times.

ACTION: If the concentration of radioactive material in the circulating water discharge exceeds ten times the concentration values of Appendix B, Table 2, Column 2 of 1 OCFR20, measures shall be initiated to restore the concentration to within these limits as soon as practicable.

ACTION: If the concentration when averaged over one hour exceeds ten times the applicable concentrations specified in Appendix B of 10CFR Part 20, Table 2, Column 2, at the point of entry to receiving waters, submit to the commission a special report within 30 days.

SURVEILLANCE REQUIREMENTS S.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 1-1. The results of pre-release analyses shall be used with the calculational methods in Section 1.6 to assure that the concentration at the point of release is limited to the values in C.1.1 BASES This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within (1) the Section I.A design objectives of Appendix I, 10 CFR 50, to a MEMBER OF THE PUBLIC, and (2) the limits of Appendix B, 10 CFR 20, 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 MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

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1.2 Dose (10 CFR 50 Appendix I)

CONTROLS C.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS shall be limited:

1. during any calendar quarter to < 1.5 mrem to the total body and to < 5 mrem to any organ, and

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

APPLICABILITY: At all times.

ACTION: Whenever the calculated dose resulting from the release of radioactive materials in liquid effluents exceeds the quarterly limits of C.1.2.1, a Special Report shall be submitted to the Commission within thirty days which includes the following information:

1. identification of the cause for exceeding the dose limit;

2. corrective actions taken and/or to be taken to reduce the releases of radioactive material in liquid effluents to assure that subsequent releases will remain within the above limits;

3. The results of the radiological analyses of the nearest public drinking water source, and an evaluation of the radiological impact due to licensee releases on finished drinking water with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.

ACTION: During any month when the calculated dose to a MEMBER OF THE PUBLIC exceeds 1/48 the annual limit (0.06 mrem to the total body or 0.2 mrem to any organ), projected cumulative dose contributions from liquid effluents shall be determined for that month and at least once every 31 days for the next 3 months.

SURVEILLANCE REQUIREMENTS S.1.2.1 Post-release analyses of samples composited from batch releases shall be performed in accordance with Table 1-1. The results of the post-release analyses shall be used with the calculational methods in Section 1.6 to assure that the dose commitments from liquids are limited to the values in C.1.2 BASES This control is provided to implement the requirements of Sections iI.A, Ill.A, and IV.A of Appendix I, 10 CFR 50 . This control implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the ODCM Page 6 Revision 16

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 "as low as reasonably achievable". Also, with Lake Ontario drinking water supplies potentially affected by plant operations, there is reasonable assurance that the operation of the plant will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational models and data, such that the actual exposure of a MEMBER OF THE PUBLIC appropriate pathways is unlikely to be substantially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I", Revision 1, October 1977, and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I", April 1977.

1.3 Dose (40 CFR Part 190)

CONTROLS C.1.3 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

APPLICABILITY: At all times.

ACTION: With the calculated doses from the release of radioactive materials in liquid effluents exceeding twice the limits of C.1.2, prepare and submit to the Commission within 30 days a Special Report that defines the corrective actions to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and include scheduling for achieving conformance with the above limits. Calculations which include direct radiation contributions from the unit and from any radwaste storage shall be performed to determine total dose to a member of the public. This Special Report, as defined in 10 CFR 20.405(c) 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 releases covered by this report. 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 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

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SURVEILLANCE REQUIREMENTS S.1.3.1 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with SURVEILLANCE REQUIREMENTS S.1.2.1 at least once every 31 days, in accordance with the methodology and parameters of Section 1.7 of the ODCM.

S.1.3.2 Cumulative dose contributions from direct radiation from the unit and from radwaste storage shall be determined from environmental thermoluminescent dosimeter (TLD) data at least quarterly.

BASES This control is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46FR1 8525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources 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. It is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the plant remains within twice the dose design objectives of Appendix I, and if direct radiation doses are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose contributions from other uranium fuel cycle sources is negligible. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance, (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CR 190.11 and 10 CFR20.405c, is considered to be a timely request and fulfils he requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR 20, as addressed in C.1.1 and C.2.2. 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.

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Table 1-1 Radioactive Liquid Waste Sampling and Analysis Program Page 1 of 3 Batch Release (b)

.7 r r 5 E-07 P P Principal Gamma Emitters Each Batch Each Batch (d) and 1-131 1 E-06 or Gross Beta-gamma

5 E-07 Batch Waste Release Tanks P M Dissolved and Entrained 1 E-05 One Batch/M Gases (Gamma Emitters) _

P M H-3 1 E-05 Each Batch Composite (c) Gross Alpha 1 E-07 P Q Sr-89 Sr-90 5 E-08 Each Batch Composite (c) Fe-55 1 E-06 Continuous Release (e)

Retention Tank Continuous W Principal Gamma Emitters 5 E-07 Composite (c) (d) 1-131 1 E-06 Service Water Gross (CV Fan Cooler Continuous M or S **Grab Beta-gamma 5 E-07 and SFP HX lines)

If gross beta is performed for batch releases, then a weekly composite shall also be analyzed for Principal Gamma Emitters and 1-131.

Service water samples shall be taken and analyzed once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if alarm setpoint is reached on continuous monitor.

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Table 1-1 (continued) Page 2 of 3 Radioactive Liquid Waste Sampling and Analysis Program Table Notation (a) The LLD is the smallest concentration of radioactive material in a sample that will yield a net count above system background that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal. The LLD is defined as an a prioribefore the fact) limit representing the capability of a measurement system and not as an a posterioriafter the fact) limit for a particular measurement, the minimum detectable activity (MDA).

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

(4.66)(St,)

LLD LLD = ___________ __

(Y)(E)(V)(2.22 E+06)[exp(-?t)]

Where: LLD is the lower limit of detection as defined above as MCi per unit mass or volume S, is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate as counts per disintegration V is the sample size in units of mass or volume E is the counting efficiency Y is the fractional radiochemical yield when applicable 2.22 E+06 is the number of disintegrations per minute per !Ci Ais the decay constant t is time elapsed since sample time The value of S, used in the calculation of the LLD for a particular measurement system shall be based on the actual observed variance of the background counting rate or the counting rate of the blank samples, as appropriate, rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contribution of other radionuclides normally present in the samples. Typical values of E, V, and Y should be used in the calculation.

The background count rate is calculated from the background counts that are determined to be within +/- one FWHM energy band about the energy of the gamma ray peak used for the quantitative analysis for this radionuclide.

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable.

When circumstances result in LLDs higher than required, the reasons shall be documented in the Annual Radioactive Effluent Report.

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Table 1-1 (continued) Page 3 of 3 Radioactive Liquid Waste Sampling and Analysis Program (b) A batch release is the discharge of liquid wastes of a discrete volume.

(c) A composite sample is one is which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released. Decay corrections are calculated from the midpoint of the sampling period.

(d) The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides:

Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-134, Cs-137 and Ce-141.

This list does not mean that only these nuclides are to be detected and reported. Other nuclides which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should be reported as less than the LLD and should not be reported as being present at the LLD level. The less than values should not be used in the required dose calculations. When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Annual Radioactive Effluent Release Report.

(e) A continuous release is the discharge of liquid wastes of a non-discrete volume; e.g. from a volume or system that has an input flow during the continuous release. Decay corrections will be calculated based on all samples collected during the release.

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1.4 Liquid Effluents Release Points There are three normal release points for liquid radioactive effluents from the plant that empty into the discharge canal. These are the Radwaste Treatment Discharge, Retention Tank discharge and the All Volatile Treatment Discharge. Each of these is a monitored release line that can be isolated before the release reaches the discharge canal. There is also a release point for the service water lines used for cooling the heat exchangers that is a monitored release line but is not isolatable. If there is an alarm on the service water monitor, it is necessary to sample each heat exchanger separately to determine which has a leak and then isolate the affected heat exchanger. The pressure of the service water system flow would normally force water from the clean service water side into the contaminated side of the heat exchanger. Dilution of liquid effluent is provided by the discharge canal. The discharge canal flow is 1.7 E+05 gpm for each circulating water pump. During operating periods, two circulating water pumps are in operation. During shutdown periods, one circulating water pump is operated. If neither circulating water pump is operable, dilution is provided by operation of one to three service water pumps which provide 5.3E+03 gpm each.

1.5 Liquid Effluents Monitor Setpoints Alarm and/or trip setpoints for radiation monitors on each liquid effluent line are required.

Precautions, limitations and setpoints applicable to the operation of Ginna Station liquid effluent monitors are provided in plant procedures P-9 and CH-RETS-RMS. Setpoint values are calculated to assure that alarm and trip actions occur prior to exceeding ten times the effluent concentration of Appendix B, Table 2, Column 2 of 10 CFR 20.1001 20.2402 at the release point to the unrestricted area. For added conservatism, liquid effluent release rates are administratively set so that only fractions of the applicable maximum effluent concentrations can be reached in the discharge canal.

ODCM Mdy age 14 Rvso1 Revision 16

The Calculated alarm and trip action setpoints for each radioactive liquid effluent line monitor and flow determination must satisfy the following equation:

Equation(1): f < C F+f Where: C = the effluent concentration which implements ten times 10 CFR 20 limit for unrestricted areas, in/uCi/ml.

c = the setpoint of the radioactivity monitor measuring the radioactivity concentration in the discharge line prior to dilution and subsequent release, in ACi/ml.

F = the dilution water flow as determined prior to the release point, in volume per unit time.

f = the liquid waste flow as measured at the discharge point, in volume per unit time, in the same units as F.

Liquid effluent batch releases from Ginna Station are discharged through a liquid waste disposal monitor. The liquid waste stream (f) is diluted by (F) in the plant discharge canal before it enters Lake Ontario.

The limiting batch release concentration (c) corresponding to the liquid waste monitor setpoint is calculated from the above expression. Since the value of (f) is very small in comparison to (F), and tritium can not be accounted for, the expression becomes:

CF Equation(2): c < CF x 0.4 (1-TCF) f ODCM Page 13 Revision 16

Where: C = 10 x the allowable concentration of Cs-1 37 as given in Appendix B, Table 2, Column 2 of 10 CFR 20, 1x10 5 . This value is normally more restrictive than the calculated mixed isotopic release concentration. A weighted average, excluding Sb-1 25, from 1998 data indicated a release concentration of 1.15 x 10-5 /.Ci/mI.

(See DA-RP-99-078). This should be reviewed as an annual basis, and the more conservative value between the two be utilized.

F = the dilution flow assuming operation of only 1 circulating water pump (170,000 gpm).

c = The limiting batch release concentration corresponding to the liquid monitor setpoint f = the maximum waste effluent discharge rate through the designated pathway.

0.4 = A conservatism based on the possibility of 2 liquid discharges occurring simultaneously, minus 0.2 (Total Instrumental Uncertainty) eg: (limit - 0.2 TIU/2)

TCF = Tritium Correction Factor, based on concentration of diluted RCS H-3, divided by H-3 E.C.L. The maximum UFSAR value is 7.5 gCi/ml. The monitor setpoint against gamma emitters must be reduced proportional to the ratio of tritium in the waste stream, to which the monitor will not respond.

The limiting release concentration (c) is then converted to a set-point count rate by the use of the monitor calibration factor determined per procedure CH-RETS-RMS. The expression becomes:

c (uCi/mi)

Equation (3): Setpoint (cpm) =

Cal Factor (uCi/ml/cpm)

Example (Liquid Radwaste Monitor R-18):

Assuming, for example, that the maximum pump effluent discharge rate (f) is 90 gpm and the RCS tritium concentration is 3 uCi/ml, then the limiting batch release concentration (c) would be determined as follows:

c (pCi/ ml)< IE - 05 (uCi / ml) x 170.000(gpm) x 0.4 x c

90 gpn

< 6.39 E-3 (uCi/ml)

K1- 1890)

IE-2 IL ODCMV Page 14 Revision 16

The monitor R-18 alarm and trip setpoint (in cpm) is then determined utilizing the monitor calibration factor calculated in plant procedure CH-RETS-RMS. Assuming a calibration factor of 1.2 E-8 (uCi/ml) cpm and a limiting batch release concentration determined above, the alarm and trip setpoint for monitor R-18 would be:

1.29E -- 83uCi 6.39E 3 (pCi // ml)=

ml) 533E +05cpm above background cpm The setpoint values for the containment Fan Cooler monitor (R-16), Spent Fuel Pit Heat Exchanger Service Water Monitors (R-20A and R-20B), Steam Generator Blowdown Monitor (R-19), the Retention Tank Monitor (R-21, and the All volatile Treatment Waste Discharge Monitor (R-22) are calculated in a similar manner using equation (2), substituting appropriate values of (f) and the corresponding calibration factor.

Effluent Monitor Warning alarm setpoints are set at one-half of the trip setpoint. With all calculations equal this is a warning that 20% of the release limit has been reached at a single release point.

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1.6 Liquid Effluent Release Concentrations Liquid batch releases are controlled individually and each batch release is authorized based upon sample analysis and the existing dilution flow in the discharge canal. Plant procedures CH-RETS-LIQ-RELEASE and CH-RETS-LIQ-COMP establish the methods for sampling and analysis of each batch prior to release. A release rate limit is calculated for each batch based upon analysis, dilution flow and all procedural conditions being met before it is authorized for release. The waste effluent stream entering the discharge canal is continuously monitored and the release will be automatically terminated if the preselected monitor setpoint is exceeded. A release may continue subject to grab sample analysis and monitoring in accordance with Table 3.1-1.

If gross beta analysis is performed for each batch release in lieu of gamma isotopic analysis, then a weekly composite for principal gamma emitters and i-131 is performed.

Additional monthly and quarterly composite analyses are performed as specified in Table 1-1.

The equations used to calculate activity are:

Gamma Spectroscopy Equation (4):

uCi/gm Act. peak area counts - bkgd counts (C Time)(Eff)(Vol)(Decay)(3.7 E+04)

Gross Beta/Gamma Equation (5):

Where: C Time = seconds of count time uCi/gmn Act. Total counts - bkgd counts (C Time)(Efi)(Vol)(Decav)(3.7 E+04)

Eff = counting efficiency in count oer sec disintegrations per sec vol volume in milliliters decay = decay correction factor, exp(- A T) 3.7 E+04 conversion constant, in disinteqration per sec gCi ODCM Page 16 Revision 16

1.7 Liquid Effluent Dose The dose contribution received by the maximally exposed individual from the ingestion of Lake Ontario fish and drinking water is determined using the following methodology. These calculations will assume a near field dilution factor of 1.0 in evaluating the fish pathway dose, and a dilution factor of 20 between the plant discharge and the Ontario Water District drinking water intake located 1.1 miles away (Figure 5-2). The dilution factor of 20 was derived from drift and dispersion studies documented in reference 4.

Dose contributions from shoreline recreation, boating and swimming have been shown to be negligible in the Appendix I dose analysis, reference 5, and do not need to be routinely evaluated. Shoreline sediment samples downstream from the plant will be collected at least annually for the Radiological Environmental Monitoring Program, as a conservatism. Also, there is no known human consumption of shellfish from Lake Ontario; therefore, this pathway is not taken into consideration in dose calculations.

The dose contribution to an individual will be determined to ensure that it complies with the specification of C.1.2. Offsite receptor doses will be determined for the limiting age group and organ, unless census data show that actual offsite individuals are the limiting age group.

The following expression in used to calculate ingestion pathway dose contributions for the total release period from all radionuclides identified in liquid effluents released to unrestricted areas Equation (6): Dr = Y-[A,, Y At. C.. F.]

j i

Where: D, = the cumulative dose commitment to the total body or any organ, T, from the liquid effluents for the summation of the total time period in mrem.

1 is jor total number oj hours oj release.

Atj = the length of the jth time period over which C,j and F, are averaged for all liquid releases in hours.

C= the average concentration of radionuclide i in undiluted liquid effluent during time period At, from any liquid release in .zCi/ml.

A= the site-related ingestion dose commitment factor to the total body or any organ, T,for each identified principal gamma and beta emitter in mrem/hr per ptCi/ml. See equation (7).

Fj = the discharge canal dilution factor for C,j during any liquid effluent release.

Defined as the ratio of the maximum undiluted liquid waste flow during release to unrestricted receiving waters. The dilution factor will depend on the number of circulation pumps operating and, during icing conditions, the percentage opening of the recirculating gate. Reference curves are presented in plant procedure CH-RETS-LIQ-RELEASE.

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Equation (7): Ai- = k0 (UID +UF BF) DF Where: A = The site-related ingestion dose commitment factor to the total body or to any organ, T, for each identified principal gamma and beta emitter in mrem/hr per PCi/ml.

k, = units conversion factor, 1.14 E+05 = 1 E+06 pCi/,uCi x 1 E+03 ml/kg ÷ 8760 hr/yr U,= a receptor person's water consumption by age group from table E-5 of Regulatory guide 1.109 D= dilution factor from the near field area of the release point to potable water intake.

The site specific dilution factor is 20. This factor is 1.0 for the fish ingestion pathway.

UF = a receptor person's fish consumption by age group from table E-5 of Regulatory Guide 1.109 BF, = bioaccumulation factor for nuclide, i, in fish in pCi/kg per pCi/L, from table A-1 of Regulatory Guide 1.109 DF,, = dose conversion factor for the ingestion of nuclide, i, for a receptor person in pre-selected organ, T, in mrem/pCi, from Tables E-11, E-12, E-13, E-14 of Regulatory guide 1.109 The monthly dose contribution from releases for which radionuclide concentrations are determined by periodic composite sample analysis may be approximated by assuming an average monthly concentration based on the previous monthly or quarterly composite analysis. However, in the Annual Radioactive Effluent Release Report the calculated dose contributions from these radionuclides shall be based on the actual composite analysis.

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Example:

Computing the dose to the whole body via the fish and drinking water pathways, assuming an initial Cs-1 37 discharge concentration of 3.0 E-04 gCi/ml:

Given the following discharge factors for example, where:

At, = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the duration of the release Ci = 3.0 E-04 kCi/ml Fj = liquid waste flow = 20 qpm = 1.2 E-04 (dilution flow)(z) 170,000 gpm Z = Near field dilution = 1.0 for Ginna D = 20 and, taking the following values from Regulatory Guide 1.109 which concern the critical receptor, which is considered to be the child in this case:

UThn 510s/year U, 6.9 kg/year BFj 2000 pCi/kg per pCi/I DF, 4.62 E-05 mrem/pQi Then, the site-related ingestion dose commitment factor, Ai, , is calculated as follows:

A,, mrem/hr = Eko(Uw/Dw + UF BF )DF, aCi/ml 1.14 E+05 [510 + (6.9)(2000)] 4.62 E-05 20 A = 7.28 E+04 mrem/hr per ItlCi/ml And, the whole body dose to the child is then:

D, mrem = j

= (7.28 E+04)(1)(3.0 E-04)(1.2 E-04)

D, = 2.6 E-03 mrem to the whole body from Cs-137 The dose contribution from any other isotopes would then need to be calculated and all the isotopic contributions summed.

z.

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2.0 RADIOACTIVE GASEOUS EFFLUENTS 2.1 Concentration (10 CFR 20)

CONTROLS C.2.1 The release of radioactive gaseous effluents shall be such that the concentration at the release point does not exceed ten times the effluent concentration values specified in Appendix B, Table 2, Column I of 10CFR Part 20.1001-20.2402.

APPLICABILITY: At all times.

ACTION: If the concentration exceeds ten times the applicable concentration specified in Appendix B, Table 2, Column I in an unrestricted area, submit to the Commission a special report within 30 days.

SURVEILLANCE REQUIREMENTS S.2.1 The radioactivity content of each batch release of radioactive gaseous waste to be discharged shall be determined prior to release by representative sampling and analysis in accordance with Table 2-1. The results of pre-release analyses shall be used with the calculational methods in Sections 2.4 and 2.5 to assure that the concentration at the point of release is limited to the values in C.2.1 and the dose rates from gaseous waste are limited to the values in C.2.2.

BASES This control is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents will be within the annual dose limits of 10 CFR 20 to UNRESTRICTED AREAS. The annual dose limits are the doses associated with the concentrations of 10 CFR 20, Appendix B, Table 2, Column 1. Control of concentrations to within the limits of 10 CFR 20, Appendix B, Table 2, Column 1 at the release point is conservative against the dose limits associated with those concentrations to a MEMBER OF THE PUBLIC at the SITE BOUNDARY. C.2.1 is intended to be more restrictive than C.2.2.

2.2 Dose Rate CONTROLS C.2.2 The instantaneous dose rate due to radioactive materials released in gaseous effluents from the site shall be limited to the following values:

(1) The dose rate for noble gases shall be _<500 mrem/yr to the total body and _*3000 mrem/yr to the skin, and (2) The dose rate for 1-131, 1-133, tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days shall be _<1500 mrem/yr to any organ.

ODCM Page 20 Revision 16

APPLICABILITY: At all times.

Note: For unplanned release of gaseous wastes, compliance with C.2.2 may be calculated using annual average X/Q. Compliance with C.2.2 shall be determined by considering the applicable ventilation system flow rates. These flow rates shall be determined at the frequency required by Table 3.2-2.

ACTION: If the calculated dose rate of radioactive materials released in gaseous effluents from the site exceeds the limits of C.2.2, measures shall be initiated to restore releases to within limits as soon as practicable.

SURVEILLANCE REQUIREMENTS S.2.2.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters of Section 2.7 of the ODCM.

S.2.2.2 The dose rate due to radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters of Section 2.7 of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 2-1.

Note: The effluent continuous monitors listed in Table 3.2-1 that have provisions for the automatic termination of gas decay tank, shutdown purge or mini-purge releases, shall be used to limit releases within the values established in C.2.2 when monitor setpoint values are exceeded.

BASES This control is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents will be within the annual dose limits of 10 CFR 20, Appendix B, Table 2, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table 2 of 10 CFR 20. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. 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 whole 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.

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2.3 Dose (10 CFR 50, Appendix I)

CONTROLS C.2.3.1The air dose due to noble gases released in gaseous effluents to areas at or beyond the SITE BOUNDARY shall be limited to the following:

(1) During any calendar quarter to _<5 mrad for gamma radiation and to _<10 mrad for beta radiation.

(2) During any calendar year to _<10 mrad for gamma radiation and to *< 20 mrad for beta radiation.

APPLICABILITY: At all times ACTION: Whenever the calculated dose to a MEMBER OF THE PUBLIC resulting from noble gases exceeds the limits of C.2.3.1, a Special Report shall be submitted to the Commission within thirty days which includes the following information:

(1) Identification of the cause for exceeding the dose limit.

(2) Corrective actions taken and/or to be taken to reduce releases of radioactive material in gaseous effluents to assure that subsequent releases will be within the above limits.

ACTION: During any month when the calculated dose to a MEMBER OF THE PUBLIC exceeds 1/48th the annual limits of C.2.3.1, (0.2 mrad gamma or 0.4 mrad beta),

projected cumulative dose contributions from gaseous effluents shall be determined for that month and at least once every 31 days for the next 3 months.

SURVEILLANCE REQUIREMENTS S.2.3.1.1 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters of Section 2.8 of the ODCM at least once every 31 days.

BASES This control is provided to implement the requirements of Sections l1.B, Ill.A, and IV.A of Appendix 1,10 CFR 50. The control implements the guides set forth in Section I.B of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the release of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as reasonably achievable". The SURVEILLANCE REQUIREMENTS implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM 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 ODCM Page 22 Revision 16

Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I",

Revision I, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors",

Revision I, July 1977. The ODCM equations for determining air doses at the SITE BOUNDARY are based on historical average atmospheric conditions.

CONTROLS C.2.3.2The dose to a MEMBER OF THE PUBLIC from 1-131, 1-133, tritium, and for all radioactive materials in particulate form with half-lives greater than eight days released with gaseous effluents from the site shall be limited to the following:

(1) during any calendar quarter to

7.5 mrem to any organ.

(2) during any calendar year to _<15 mrem to any organ.

APPLICABILITY: At all times.

ACTION: Whenever the calculated dose to a MEMBER OF THE PUBLIC resulting from radionuclides other than noble gases exceeds the limits of C.2.3.2, a Special Report shall be submitted to the Commission within thirty days which includes the following information:

(1) Identification of the cause for exceeding the dose limit.

(2) Corrective actions taken and/or to be taken to reduce releases of radioactive material in gaseous effluents to assure that subsequent releases will be within the above limits.

ACTION: During any month when the calculated dose to a MEMBER OF THE PUBLIC exceeds 1/48th the annual limit of 2.3.2, (0.3 mrem), projected cumulative dose contributions from gaseous effluents shall be determined for that month and at least once every 31 days for the next 3 months.

SURVEILLANCE REQUIREMENTS S.2.3.2.1 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with methodology and parameters of Section 2.8 of the ODCM at least once every 31 days.

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BASES This control is provided to implement the requirements of Sections II.C, IIl.A, and IV.A of Appendix 1, 10 CFR 50. The control implements the guides set forth in Section 11.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the release of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as reasonably achievable". The SURVEILLANCE REQUIREMENTS implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of the subject materials 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 10 CFR 50, Appendix I",

Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors",

Revision I, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. He release rate controls for Iodine-131, Iodine-133, 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 at or beyond the SITE BOUNDARY. The pathways examined in development of the calculations were:

1.) Individual inhalation of airborne radionuclides 2.) Deposition of radionuclides onto green leafy vegetation with subsequent consumption by man 3.) Deposition of radionuclides onto grassy areas where milk animals and meat producing animals graze, followed by human consumption of that milk and meat 4.) Deposition of radionuclides on the ground followed by subsequent human exposure ODCM Page 24 Revision 16

2.4 Dose (40 CFR 190)

CONTROLS C.2.4 The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

APPLICABILITY: At all times.

ACTION: With the calculated doses from the release of radioactive materials in gaseous effluents exceeding twice the limits of C.2.3.1 and C.2.3.2, prepare and submit to the Commission within 30 days a Special Report that defines the corrective actions to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and include scheduling for achieving conformance with the above limits.

Calculations which include direct radiation contributions from the unit and from any radwaste storage shall be performed to determine total dose to a member of the public. This Special Report, as defined in 10 CFR 20.405(c) 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 releases covered by this report. 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 40 CFR 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.

ACTION: This report shall include an analysis which demonstrates that radiation exposure to all MEMBERS OF THE PUBLIC from the plant are less than the 40 CFR 190 limits.

Otherwise, the report shall request a variance from the commission to permit releases to exceed 40 CFR Part 190. Submittal of the report is considered a timely request by the NRC, and a variance is granted until staff action on the request is complete.

SURVEILLANCE REQUIREMENTS S.2.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with SURVEILLANCE REQUIREMENTS S.2.3.1, and S.2.3.2 at least once every 31 days, in accordance with the methodology and parameters of Section 2.8 of the ODCM.

S.2.4.2 Cumulative dose contributions from direct radiation from the unit and from radwaste storage shall be determined from environmental thermoluminescent dosimeter (TLD) data at least quarterly.

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BASES This control is provided to meet the dose limitations of 40 CFR 190 that have been incorporated into 10 CFR 20 by 46FR18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources 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. It is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the plant remains within twice the dose design objectives of Appendix I, and if direct radiation doses are kept small. The Special Report will describe course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR 190 limits. For the purposes of the Special Report, it may be assumed that the dose contributions from other uranium fuel cycle sources is negligible. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance, (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfils he requirements of 40 CFR 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR 20, as addressed in C.1.1 and C.2.2.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.

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Table 2-1 Radioactive Gaseous Waste Sampling and Analysis Program Page 1 of 2 Gaseous Sampling Minimum Type of Activity Lower Limit of Release Type Frequency Analysis Analysis Detection Frequency (LLD)(uiCi/cc)(a)

Containment P P Principal Gamma 1 E-04 Purge Each Purge (b,c) Emitters (e)

Grab Sample H-3 1 E-06 Auxiliary Building M (b) M (b) Principal Gamma 1 E-04 Ventilation Grab Sample Emitters (e)

H-3 1 E-06 Continuous (d) W (b) 1-131 1 E-12 Charcoal Sample 1-133 1 E-10 All Release Types Continuous (d) W (b) Principal Gamma 1 E-11 as listed above Particulate Emitters (e)

Sample Continuous (d) M Gross Alpha 1 E-1 1 Composite Particulate Sample Continuous (d) Q Sr-89, Sr-90 1 E-1 1 Composite Particulate Sample Air Ejector M (b,f) M (b, f) Principal Gamma 1 E-04 Grab Sample Emitters (e) 1-131 (h) 1 E-12 H-3 (g) 1 E-06 All Release Types Continuous (d) Noble Gas Beta or Gamma 1 E-06 listed above Monitor Gas Decay Tank P Each Tank P Principal Gamma 1 E-04 Grab Sample J Each Tank Emitters (e)

ODCM Page 27 Revision 16

Table 2-1 (continued) Page 2 of 2 Radioactive Gaseous Waste Sampling and Analysis Program (a) The lower limit of detection (LLD) is defined in Table Notation (a) of Table 1-1.

(b) Analyses shall also be performed when the monitor on the continuous sampler reaches its setpoint.

(c) Tritium grab samples shall be taken at least three times per week when the reactor cavity is flooded.

(d) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with C.2.1, C.2.2, C.2.3.1, & C.2.3.2.

(e) The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides:

Kr-85m, Xe-1 33, Xe-133m and Xe-1 35 for gaseous emissions 1-131, Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.

This list does not mean that only these nuclides are to be detected and reported. Other nuclides which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances result in LLDs higher than required, the reasons shall be documented in the Annual Radioactive Effluent Release Report.

(f) Air ejector samples are not required during cold or refueling shutdowns.

(g) Air ejector tritium sample is not required ifthe secondary coolant activity is less than 1 E-04

,Ci/gm.

(h) Air ejector iodine samples shall be taken and analyzed weekly ifthe secondary coolant activity exceeds 1 E-04 4Ci/gm.

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2.5 Gaseous Effluent Release Points There are three release points continuously monitored for noble gases, containment vent, plant vent and air ejector. The containment vent and plant vent are also continuously monitored for radioiodines and particulates. Since the air ejector is a steam release point, continuous radioiodine and particulate monitoring is not required when the secondary coolant activity is less than 1 E-04 gCi/gm. Flow rates through the vents are measured periodically. During shutdown, temporary trailers may be brought on site that also require monitoring and characterization of their releases, such as the CO 2 decon trailer.

Quarterly plant measurements of one week duration for the particulate and iodine released in the steam by the air ejector demonstrate that sampling this pathway for particulate and iodine is not necessary since these releases are less than 0.1% of the Plant Vent. The releases are correlated to blowdown activity for determining activity in steam releases.

During shutdown and startup, special systems are in use that may release small amounts of radioactivity in steam releases. This is accounted for by using operational data and activity in the source of the steam. Grab samples are obtained when practicable.

If an unmonitored release point is discovered, a calculation is performed to determine the potential radioactivity that is released. The calculation includes a conservative estimate of source term if sample data is not available, and a conservative estimate of flow rate and duration if measurement of flow and duration are not available. If the release is continuous, it is included in the monthly report that accounts for releases from the site for calculating doses to the general public.

2.6 Gaseous Effluent Monitor Setpoints Alarm and/or trip setpoints for specified radiation monitors are required on each noble gas effluent line from the plant. Precautions, limitations and setpoints applicable to the operation of Ginna Station gaseous effluent monitors are provided in plant procedures P-9 and CH-RETS-RMS. Setpoints are conservatively established for each ventilation noble gas monitor so that dose rates in unrestricted areas corresponding to 10 CFR Part 50 Appendix I limits will not be exceeded. Setpoints shall be determined so that dose rates from releases of noble gases will comply with C.2.2.(1). Table 6-2 provides the gaseous and particulate meteorological assumption used in development of the P-9 setpoints.

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The calculated alarm and trip action setpoints for each radioactive gaseous effluent monitor must satisfy the following equation:

EQi, Equation (8): c < I (1)(k)(K)

Where: c, = setpoint in cpm Qi = release rate limit by specific nuclide (i) in 4Ci/sec from vent (v) f = discharge flow rate in cfm k = units conversion factor in cc/sec/cfm K = calibration factor in 4Ci/cc/cpm The general methodology for establishing plant ventilation monitor setpoints is based upon a vent concentration limit in gCi/cc derived from site specific meteorology and vent release characteristics.

Additional radiation monitor alarm and/or trip setpoints are calculated for radiation monitors measuring radioiodines, radioactive materials in particulate form and to radionuclides other than noble gases. Setpoints are determined to assure that dose rates from the release of these effluents shall comply with C.2.2(2)

The release rate limit for noble gases shall be calculated by the following equation for total body dose:

Equation (9): Q,. [uCi/sec] <_VQ 500 mrem/vr S(X/Q), 7K Q.0.

Note: The occupancy factor "T" of 1 is assumed. This may be modified following reviews of the area in question.

and by the following equation for skin doses:

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Equation (10):

V [uCi/sec] E, 3000 mrem/yr (X/Q), (L + 1.1M)Q Where: Q,, = the release rate of radionuclide (i) from vent (v) which results in a dose rate of 500 mrem/yr to the whole body or 3000 mrem/yr to the skin of the critical receptor in uCi/sec.

K = the total body dose factor due to gamma emissions for each identified noble gas radionuclide in mrem/yr per kCi/m 3 from table 2-3.

L = the skin dose factor due to beta emissions for each identified noble gas radionuclide in mrem/yr per UCi/m 3 from table 2-3.

M = the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad/yr per .zCi/m 3 from Table 2-3. Unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose.

(X/Q), = the highest calculated annual average dispersion parameter for estimating the dose to the critical offsite receptor from vent release point (v) in sec/m 3. The (X/Q)v is calculated by the method described in Regulatory Guide 1.111.

Noble gas monitor setpoints are conservatively set according to procedure P-9 to correspond to fractions of the applicable 10 CFR Part 20 dose limits for unrestricted areas.

Fractions are small enough to assure the timely detection of any simultaneous discharges from multiple release points before the combined downwind site boundary concentration could exceed allowable limits. Additional conservatism is provided by basing these setpoints upon instantaneous downwind concentrations. Release rates during the remainder of a given year, combined with any infrequent releases at setpoint levels, would result in only a very small fraction of the 10 CFR Part 20 annual limits.

Historically, xenon-1 33 is the principal noble gas released from all vents and is appropriate for use as the reference isotope for establishing monitor setpoints. The whole body dose will be limiting, and the Xe-133 release rate limit is calculated by substituting the appropriate values into equation (9). After the release rate limit for Xe-1 33 is determined for each vent, the corresponding vent concentration limits are calculated based on applicable vent flow rates. Annually-derived monitor calibration factors in kzCi/cc per cpm are used to convert limiting vent concentrations to count rates.

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Example: Plant Vent Monitor, R-14 Using Xe-1 33 as the controlling isotope for the setpoint and assuming a measured activity of 2.66 E-04 /bCi/cc and a ratemeter reading of 4750 cpm above background, the efficiency can be calculated, using a measured vent flow of 7.45 E+04 cfm, K,from Table 2-3 of 2.94E+02 and a (X/Q)v for the site boundary of 2.7 E-06, the Release Rate Limit is calculated and then the setpoint determined.

Xe-133 efficiency = Activity Net ratemeter reading 2.66 E-04 uCi/c_

Xe-133 efficiency - - 5.67 E-08 -i 4750 cpm Using Equation 9:

500 (2.94 E+02)(2.7 E-06) - 6.3 E+05 uCi/sec Release Rate Limit Q9. 500 mrei/r (K)(X/Q),

Using Equation 8:

Setpoint c <_

(t)(k)(K)

(6.3 E+05 uCi/sec cc/se/c (7.45 E+04 cfm) 472 c/sec 5.67 E-08 c_ 3.*CrF+A5 epm cpm ODCM Page 32 Revision 16

Per procedure P-9, R-14 is set at 0.4 of this value or 1.28 E+05 cpm for normal operation.

40% of the release rate limit is a conservatism based on the possibility of two release points simultaneously at their setpoints for a total of 80% of the release rate limit.

Effluent Monitor Warning alarm setpoints are set at one-half of the trip setpoint. With all calculations equal this is a warning that 20% of the release limit has been reached at a single release point.

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Table 2-2 Meteorological Data and Locations of Receptors for Set Point Calculations Process Monitors Monitor Geographic Location Release Point XJQ DIQ Flow (Radioisotope) (DistancelDirection) (vent) (sec/m3 ) (m 2-mrem/yr (cfm) per uCi/sec)

R-10 A 0.5 - 1 mile ESE Containment 2.4E-8 11,000 (Radioiodine)

R-10B 0.5 - 1 mile ESE Plant 3.OE-8 76,000 (Radioiodine)

R-1 1 (Cs-1 37) 0.5 - 1 mile ESE Containment 2.4E-8 11,000 R-12 (Xe-133) 0.5 - 1 mile E Containment 1.6E-6 11,000 R-13 (Cs-137) 0.5 - 1 mile ESE Plant 3.OE-8 76,000 R-14 (Xe-1 33) 0.5 - 1 mile E Plant 2.7E-6 76,000 R-1 5 (Xe-1 33) 0.3 miles SSE Air Ejector 1.3E-5 600 SPINGS Monitor Geographic Location Release Point Noble Gas Iodine Particulate (Radioisotope) (Distance/Direction) (vent) utCi/min /tCi/min /zCi/min R-12A 0.5 - 1 mile E Containment 6.3E+7 3.4 80 R-14A 0.5 - 1 mile E Plant 3.8E+7 2.7 64 R-15A 0.33 miles SSE Air Ejector 2.64E+5 Further details found in procedure P-9.

1-131, 1-133, H-3 and particulates with half lifes greater than 8 days utilizes the following equation:

1500 mrem / Year D/ OP, For Noble Gases:

500 mrem / year X/QOK, Pj Food and ground pathways in m2 mrem/year per 4Ci/sec 3

Ki mrem/year per kCi/m W Qi11 Release rate in units of uCi/sec ODCM Page 34 Revision 16

2.7 Gaseous Effluent Dose Rate Gaseous effluent monitor setpoints as described in Section 2.6 of this manual are established at concentrations which permit some margin for corrective action to be taken before exceeding offsite dose rates corresponding to 10 CFR Part 20 limitations. Plant procedures CH-RETS-RMS-CV, CH-RETS-RMS-CV-ALT, CH-RETS-PURGE-CV, CH-RETS-SAMP-PV, CH-RETS-SAMP-PV-ALT, CH-RETS-PV-RELEASE, CH-RETS AIR-H3 and CH-RETS-MINIPURGE establish the methods for sampling and analysis for continuous ventilation releases and for containment purge releases. Plant procedure CH-RETS-GDT-REL establishes the methods for sampling and analysis prior to gas decay tank releases. The instantaneous dose rate in unrestricted areas due to unplanned releases of airborne radioactive materials may be calculated using annual average X/Q's.

Dose rates shall be determined using the following expressions:

For noble gases:

Equation (12): D, = [(L, + 1.1 M)(X/Q) Q <3000 mrem/Qr total gamma and beta dose to the skin For 1-131, 1-133, tritium and all radioactive materials in particulate form with half-lives greater than 8 days:

Equation (13): D, = P1 (X/Q),Q *_ 1500 mrem/yr to critical organ i2 ODCM Page 35 Revision 16

Where:

K, the total body dose factor due to gamma emissions for each identified noble gas radionuclide (i) in mrem/yr per ;Ci/m3 from Table 2-2.

L, the skin dose factor due to beta emissions for each identified noble gas radionuclide (i) in mrem/yr per uCi/m 3 from Table 2-2.

M = the air dose factor due to gamma emissions for each identified noble gas radionuclide (i) in mrad/yr per yCi/m3 from Table 2-2. Unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose.

P, the dose parameter for radionuclide (") other than noble gases for the inhalation pathway, in mrem/yr per UCi/M 3. The dose factors are based on the critical individual organ and the child age group. P, is further defined as: Pi =

(10 6pCi/uCi)(BR)(DFAý) where BR is the breathing rate for a child in m3/yr and DF* is the dose factor for the child in mrem/pCi.

(X/Q), the highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary in sec/m 3.

W = the highest annual average dispersion parameter for estimating the dose to the critical receptor in sec/m 3 for the inhalation pathway and in m-2 for the food and ground pathways.

QIV the release rate of radionuclide (i) from vent (v) in gCi/sec.

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2.8 Gaseous Effluent Doses The air dose in unrestricted areas due to noble gases released in gaseous effluents from the site shall be determined using the following expressions:

Durina any calendar year, for aamma air dose:

Equation (14): Dvy = 3.17 E-08 * [M(X/Q)v Qj.]_ 10 mrad During any calendar quarter, for gamma air dose:

Durina any calendar year for beta air dose:

Equation (15): D,3 = 3.17 E-08 " [N.(X/Q)v Qj. 20 mrad During any calendar quarter, for beta air dose:

Equation (1SA): D P = 3.17 E [N(X/Q), Q Qj] _ 10 mrad Where:

M, the air dose factor due to gamma emissions for each identified noble gas radionuclide in mrad/yr per 1aCi/m 3 from Table 2-3 N, = the air dose factor due to beta emissions for each identified noble gas radionuclide in mrad/yr per UCi/M3 from Table 2-3 (X/Q), = for vent releases. The highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary in sec/m3.

D = the total gamma air dose from gaseous effluents in mrad.

D = the total beta air dose from gaseous effluents in mrad.

Q = the release of noble gas radionuclides, i, in gaseous effluents from vents in /2Ci.

Releases shall be cumulative over the time period.

3.17 E-08 = the inverse of the number of seconds in a year ODCM Page 37 Revision 16

The dose to an individual from 1-131, 1-133, tritium and all radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents released from the site to unrestricted areas shall be determined using the following expression:

dose during any calendar year:

Equation (16): D, = 3.17 E-08 Y [R1 W Q.] _ 15 mrem dose during any calendar quarter:

Equation (16A): D, = 3.17 E-08 Yi [R W Q14 _<7.5 mrem Where:

D, the total dose from 1-131, 1-133, tritium and all radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents in mrem.

R, the dose factor for each identified radionuclide (i) in m 2 mrem/yr per /,Ci/sec or mrem/yr per UCi/M 3 from Table 2-5.

W, the annual average dispersion parameter for estimating the dose to an individual at the critical location in sec/m3 for the inhalation pathway and in m2 for the food and ground pathways.

QIV the release of 1-131, 1-133, tritium and all radioactive materials in particulate form in gaseous effluents with half-lives greater than 8 days in uCi. Releases shall be cumulative over the desired time period as appropriate.

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Table 2-3 Dose Factors to the Child For Noble Gases and Daughters

Total Body Dose Factor Skin Dose Factor L, Gamma Air Dose Factor M, Beta Air Dose Factor N, Radionuclides K, (mrem/yr per yCi/m 3) (mradlyr per ptCilm3 ) (mrad/yr per UCi/m 3)

(mrem/yr per yCi/m3)

Kr-83m 7.56E-02** ---- 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1._11_E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-133m 2.51 E+02 9.94E+02 3.27E+02 1.48E+03 Xe- 135m 3.12E+03 7.11 E+02 3.36E+03 7.39E+02 Xe-135 1.81 E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51 E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03

The listed dose factors are for radionuclides that may be detected in gaseous effluents. These dose factors for noble gases and daughter nuclides are taken from Table B-1 of Regulatory Guide 1.109 (reference 3). A semi-infinite cloud is assumed.

7.56E-02 = 7.56 x 10-2 ODCM " P~np Vq

Table 2-4 Dose Parameters for Radionuclides and Radioactive Particulate, Gaseous Effluents

P, Inhalation P, Food & Ground P, Inhalation Pathways P, Food & Ground Radionuclide Pathways Pathways Radionuclides (mrem/yr per ACi/m 3) Pathways s (mrem/yr per kCi/m3 ) (mi2

mrem/yr per yCi/sec) (n 2

mrem/yr per

/pCi/sec)

H-3 6.5E+02 2.4E+03 Cd-115m 7,OE+04 4.8E+07 C-14 8.9E+03 1.3E+09 Sn-126 1.2E+06 1.1 E+09 Cr-51 3.6E+02 1.1E+07 Sb-125 1.5E+04 1.1 E+09 Mn-54 2.5E+04 1.1E+09 Te-127m 3.8E+04 7.4E+10 Fe--59 2.4E+04 7.OE+08 Te-129m 3.2E+04 1.3E+09 Co-58 1.1E+04 5.7E+08 Te-132 1.OE+03 7.2E+07 Co-60 3.2E+04 4.6E+09 Cs-134 7.OE+05 5.3E+10 Zn-65 6.3E+04 1.7E+10 Cs-136 1.3E+05 5.4E+09 Rb-86 1.9E+05 1.6E+10 Cs-137 6.1E+05 4.7E+10 Sr-89 4.OE+05 1.OE+10 Ba-140 5.6E+04 2.4E+08 Sr-90 4.1E+07 9.5E+10 Ce-141 2.2E+04 8.7E+07 Y-91 7.OE+04 1.9E+09 Ce-144 1.5E+05 6.5E+08 Zr-95 2.2E+04 3.5E+08 Np-239 2.5E+04 2.5E+06 Nb-95 1.3E+04 3.6E+08 1-131 1.5E+07 1.1E+12 Mo-99 2.6E+02 3.3E+08 1-133 3.6E+06 9.6E+09 Ru-IU3 1.6E+04 3.4E+10 I Inid*_ntifi*=d A I r-.4.N'7 3.4E+1O U nirIentifior'r A I C+/-AL7 9.5E+10 Ru-106 1.6E+05 4 41=+1 1 Ag-ib1m [ g 0.3E+044 3.3 E+04 (1.5E+10 41E+1 1

The listed dose parameters are for radionuclides that may be detected in gaseous effluents. These and additional dose parameters for isotopes not included in Table 2-4 may be calculated using the methodology described in NUREG-0133, Section 5.2.1 (reference 2).

Table 2-5 Pathway Dose Factors Due to Radionuclides Other Than Noble Gases

Inhalation Meat Ground Plane Cow-Milk-Child Leafy Vegetables Radionuclides Pathway R, Pathway R, Pathway R, Pathway R, Pathway R, (mremlyr per (m 2.mrem/yr per (m 2omrem/yr per (m 2'mremlyr per (m 2°mremlyr per MMCi/m 3 ) yCi/sec) pCi/sec) ACi/sec) pCilsec)

H-3 1.12E+03 2.33E+02 0. 2.38E+03 2.47E+02 Cr-51 1.70E+04 4.98E+05 5.31 E+06 5.75E+06 1.63E+06 Mn-54 1.57E+06 7.60E+06 1.56E+09 3.70E+07 5.38E+07 Fe-59 1.27E+06 6.49E+08 3.09E+08 4.01E+08 1.10E+08 Co-58 1.10E+06 9.49E+07 4.27E+08 7.01E+07 4.55E+07 Co-60 7.06E+06 3.61E+08 2.44E+10 2.25E+08 1.54E+08 Zn-65 9.94E+05 1.05E+09 8.28E+08 1.99E+10 2.24E+08 Sr-89 2.15E+06 4.89E+08 2.42E+04 1.28E+10 5.39E+09 Sr-90 1.01E+08 1.01E+10 0 1.19E+10 9.85E+10 Zr-95 2.23E+06 6.09E+08 2.73E+08 8.76E+05 1.13E+08 1-131 1.62E+07 2.60E+09 1.01E+07 4.95E+ I 2.08E+10 1-133 3.84E+06 6.45E+01 1.43E+06 4.62E+09 3.88E+08 Cs-134 1.01E+06 1.42E+09 7.70E+09 6.37E+10 1.96E+09 Cs-136 1.71E+05 5.06E+07 1.64E+08 6.61E+09 1.60E+08 Cs-137 9.05E+05 1.27E+09 1.15E+10 5.75E+10 1.80E+09 Ba-140 1.74E+06 5.OOE+07 2.26E+07 2.75E+08 2.03E+08 Ce-141 5.43E+05 1.45E+07 1.48E+07 1.43E+07 8.99E+07

Additional dose factors for isotopes not included in Table 2-5 may be calculated using the methodology described in NUREG-0133, Section 5.3.1 (reference 2).

n - A -

3.0 RADIOACTIVE EFFLUENT MONITORING INSTRUMENTATION 3.1 Liquid Effluent Monitors CONTROLS C.3.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.1-1 shall be OPERABLE with their Alarm/Trip setpoints set to ensure that the limits of Control C.1.1 are not exceeded. The Alarm/Trip setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: At all times.

ACTION: With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

ACTION: With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.1-1.

Restore the inoperable instrumentation to OPERABLE status within 30 days or explain in the next Annual Radioactive Effluent Release Report, pursuant to Section 6.2 of the ODCM, why this inoperability was not corrected in a timely manner.

SURVEILLANCE REQUIREMENTS S.3.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 3.1-2.

BASES The radioactive liquid effluent instrumentation is is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm/Trip setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the Alarm/Trip will occur prior to exceeding the limits of 10 CFR 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR 50.

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Table 3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation Gross Activity Monitors (Liquid) Minimum Channels Action OPERABLE

a. Containment Fan Coolers (R-16) 1 1

b. Liquid Radwaste (R-18) 1 2

c. Steam Generator Blowdown (R-19) 1(a) 3

d. Spent Fuel Pool Heat Exchanger 1 1 (R-20A, R-20B)

e. Turbine Building Floor Drains (R-21) 1 1

f. High Conductivity Waste (R-22) 1 2 (a) Not required when steam generator blowdown is being recovered, i.e. not released.

Table Notation Action 1 - If the number of OPERABLE channels is less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months grab samples are analyzed for isotopic concentration or gross radioactivity (beta or Gamma) at a lower limit of detection (LLD) of at most 1 E-07 AzCi/gm.

Action 2 - If the number of OPERABLE channels is less than required by the minimum Channels OPERABLE requirement, effluent releases from the tank may continue for up to 14 days, provided that prior to initiating a release:

Note: When counting 2 independent samples for agreement, doubling the acceptance criterion for low (<6E-05 MCi/mi) activity samples from 10% to 20% results in a consequence at the release point of < 1%. The expanded acceptable criterion for low activity samples is employed to compensate for increased impact of sampling and counting error on acceptance.

1. At least two independent samples of the tank's contents, taken at least 60 minutes apart, are analyzed and agree within 10% of total activity, (20% if total activity minus noble gases <6E-05 uCi/ml), and

2. At least two technically qualified members of the Facility Staff independently review and approve the analytical results, and

3. At least two technically qualified members of the Facility Staff independently verify the discharge line valving.

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Action 3 - When Steam Generator Blowdown is being released (not recycled) and the number of channels OPERABLE is less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for isotopic concentration or gross radioactivity (beta or gamma) at a lower limit of detection (LLD) of at most 1 E-07 MCi/gram:

1. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months when the concentration of the secondary coolant is

> 0.01 uCi/gram (DOSE EQUIVALENT 1-131).

2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the concentration of the secondary coolant is _ 0.01 gCi/gram (DOSE EQUIVALENT 1-131).

Table 3.1-2 Radioactive Liquid Effluent Monitoring SURVEILLANCE REQUIREMENTS Gross Activity Monitor (Liquid) Channel Source Functional Channel Check Check Test Calibration

a. Containment Fan Coolers (R-16) D(5) M(3) Q(2) R(4)

b. Liquid Rad waste (R-18) D(5) M(3) Q(1) R(4)

c. Steam Generator Blowdown (R-19) D(5) M(3) Q(1) R(4)

d. Spent Fuel Pool Heat Exchanger D(5) M(3) Q(2) R(4)

(R-20A, R-20B) I