Text

Enclosure 2, Offsite Dose Calculation Manual for Calvert Cliffs Nuclear Power Plant, Revision 7.
	ML042030129
Person / Time
Site:	Calvert Cliffs
Issue date:	07/15/2004
From:	; Constellation Energy Group
To:	; Office of Nuclear Reactor Regulation
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ENCLOSURE (2)

OFFSITE DOSE CALCULATION MANUAL FOR CALVERT CLIFFS NUCLEAR POWER PLANT, Revision 7 Calvert Cliffs Nuclear Power Plant, Inc.

July 15,2004

. - I OFFSITE DOSE CALCULATION MANUAL Revision 7 Calvert Cliffs Nuclear Power Plant DATE P-Aeo'S DATE:

DATE: -> ' >S REVlEWEDBY:

POSRO O3.O4 a Meeting Nunber DATE: X AA/t7 32 APPROVED BY: &A DATE: ektk/b7 Du ontGenaral Manager V

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 2 of 319 EXECUTIVE

SUMMARY

OF CHANGES REV CHG PAGE DESCRIPTION 7 0 6 Updated organization titles for Fort Smallwood Personnel.

6 A. Added TS 5.5.1 and 5.5.4 to establish and emphasize link to TS.

C. Changed 'radiation effluent" to 'effluent radiations.

7 Deleted "and recommendations made by CE", this change caused the conjunction to be moved in the sentence.

15,134 Changed reference from TE-001 to STP-M-461-1 and STP-M.461-2, or added reference to STP-MI461-1 and STP-M-461-2 10, 31,42,43,48, 52, 53, 54, Removed reference to Improved TS.

55,58,110,204,209,210.

211,214,215,219 18.22,23,60,88 Reworded for clarity.

37 Corrected TS and TRM reference.

56 Changed "four" to "three" to reflect the correct number of pathways.

59 Corrected System Description references.

63, 73, 84, 92, 93, 97, 100, Deleted reference to Attachment 3. Attachment 3 was removed 102 during a previous revision.

95 Deleted entire paragraph because CCNPP is committed to the old 10 CFR 20.

105 Added new substep 10 to list of potential release pathways, IR4-

__ 004-004.

109 Added new step 13 to describe Gland Steam Exhaust Condenser pathway, IR4-004-004.-

216,217 Changed sector designation from "south" to "south-east" as a result of Global Positioning Satellite analysis of site.

219 Removed paragraph describing NRC TLD cross-check program. It was discontinued by the NRC.

222 Updated "Test and Equipment Unit" to "Electrical and Control Section".

229 Replaced drawing with new for clarity. Updated drawing to eliminate Outfall 002 pathway to the bay. Outfall 02 is now directed to circulating water via Outfall 001.

234 Added reference to STP-M-462-1 and STP-M-462-2 as replacements for TE-001 plant vent stack flow rate determination.

316,317,318 Changed sector designations and updated distances as a result of GPS and satellite imaging.

OFFSITE DOSE CALCULATION MANUAL .,.Rev.

Page 7 k I ~ 3 of 319 LIST OF EFFECTIVE PAGES - . ;

PAGES REVISION/CHANGE 1-319 7 PROCEDURAL ALTERATIONS PAGES REVISION/CHANGE 4,6,7,10,18,22-23,31,37 07/00 I 42-43,48,52-56,58-60 63,73,84,88,92-93,95,97 100,102,105,109-110,134, 204 209-211,214-217,219,222 229,234,316,317,318

Rev. 7

-OFFSITE DOSE CALCULATION MANUAL BPage 4 ofC319 TABLE OF CONTENTS Part 1.0: PURPOSE AND APPLICABILITY/SCOPE ............................................................... 6 Part 2.0: DEFINITIONS AND REFERENCES ............................................................... 8 Developmental References .............................................................. 13 Performance References .............................................................. 15 PArt 3.0/4.0: CONTROLS AND SURVEILLANCE REQUIREMENTS ................................................... 16 Monitoring Instrumentation .............................................................. 18 Radioactive Gaseous Effluent .............................................................. 18 Radioactive Liquid Effluent .............................................................. 23 Radioactive Effluents .............................................................. 28 Liquid Effluents .............................................................. 28 Gaseous Effluents: .............................................................. 34 Total Dose .............................................................. 41 Radiological Environmental Monitoring .............................................................. 42 Program .............................................................. 42 Land Use Census .............................................................. 54 Interlaboratory Comparison Program .............................................................. 55 Part 5.0: CALCULATIONAL METHODOLOGIES .................... .......................................... 56 Radioactive Liquid Effluents .............................................................. 56 Release Pathways [B527] .............................................................. 56 Types of Liquid Releases .............................................................. 58 Processing Equipment .............................................................. 58 Liquid Effluent Radiation Monitor Alarm and Fixed Setpoints .............................................................. 59 Limits On Radionuclide Concentrations In Liquid Effluents .............................................................. 89 Limits On Cumulative Total Body Doses And Cumulative Organ Doses For Liquid Effluents .................................. 95 Limits For The Liquid Waste Processing System .............................................................. 102 Radioactive Gaseous Effluents .............................................................. 105 Release Pathways [B527] .............................................................. 105 Types of Gaseous Releases .............................................................. 110 Processing Equipment .............................................................. 110 Gaseous Effluent Radiation Monitors And Setpoints ............................................................... 111 Annual Total Body Dose Rate Due To Noble Gases In Gaseous Effluents ............................................................ 152 Annual Skin Dose Rate Due To Noble Gases In Gaseous Effluents .............................................................. 160 Annual Organ Dose Rates Due To lodines And Particulates In Gaseous Effluents ............................................... 166 Cumulative Gamma Air Doses Due To Noble Gases In Gaseous Effluents ........................................................... 174 Cumulative Beta Air Doses Due To Noble Gases In Gaseous Effluents .............................................................. 182 Cumulative Organ Doses Due To lodines And Particulates In Gaseous Effluents ................................................. 189 Limits For The Gaseous RadWaste Processing System .............................................................. 196 Limits On Total Annual Dose - Gases, Liquids, and Uranium Fuel Cycle Sources ............................................... 198 Special Exceptions and Assumptions For Calculation of Doses and Dose Rates .................................................. 202

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 5 of 319

- I ---- ------- ------- ------ ------- ------- - -- -----

TABLE OF CONTENTS (Continued)

Radiological Environmental Monitoring Program ..................................... 204 Radiological Environmental Monitoring Program ........................... ............................................................. ............. .

Land Use Census.......................................................................... ............................................................. 209 .

Interlaboratory Comparison Program............................................ .

Annual Radiological Environmental Operating Report ................. 219 Independent Spent Fuel Storage Installation Monitoring Program 915 Radioactive Effluent Release Report .219 Introduction .. .............................. ............................................................... :.. 21 9 Responsibilities.. ..............................................................I....................................................................................... .214 21 9 An Report Contents. ..................................................................................................................................................... clu Report Submittal .................................................................................................................................................... -- -

Administration Of The ODCM ............................. 221 Introduction .................. . . . . . . : . 221 Controls On The ODCM.221 Surveillance Requirements......................... 222 Responsible Company Organizations ......................... 222 Initiating Conditions.222 Calculation Methodologies ............. ;223 Corrective actions ........................... .223 Bases.223 ATTACHMENTS:

Attachment 1 Final Grading And Drainage Plan ............................................................ ........ 224 2.......................

Attachment 2 Sources of Liquid Radioactive Waste ............................................................ 225 2..............................

Attachment 3......... 228 Attachment 4 Block Diagram of Liquid Radioactive Waste Systems ........ 229 Attachment 5 Radionuclides inTypical Radwaste Discharges ........ 230 Attachment 6 Liquid Effluent Dose Factors (mrem/hr per uCi/ml) ........ 231 Attachment 7 General Information Related to Gaseous Releases via the Main Vents ........ 234 Attachment 8 General Information Related to Gaseous Releases via Pathways other than the Main Vents .235 Attachment 9 Block Diagram of Gaseous Radioactive Waste Systems .236 Attachment 10 Noble Gas Dose Factors ........................... 237 Attachment 11 Empirical Derivation Of Site-Specific Dose Factors .238 Attachment 12 Gaseous Effluent Pathway Dose Factors .243 Attachment 13 Environmental Monitoring Sites for REMP .309 Attachment 14 Radiological Environmental Monitoring Program Surveillances for Direct Radiation .310 Attachment 15 Radiological Environmental Monitoring Program Surveillances for Airborne Activity .311 Attachment 16 Radiological Environmental Monitoring Program Surveillances for Waterborne Activity . 312 Attachment 17 Radiological Environmental Monitoring Program Surveillances for Ingestible Activity .313 Attachment 18 Map of Environmental Monitoring Sites for REMP (5 Km Radius)............................................314 Attachment 19 Map of Environmental Monitoring Sites for REMP (10 Mile Radius) .315 Attachment 20 Environmental Monitoring Sites for ISFSI ............................................. : ;; ;.

316 Attachment 21 Map of Environmental Monitoring Sites for ISFSI .317 Attachment 22 Map of Environmental Monitoring Sites for ISFSI .318 Attachment 23 Effluent Radiation Monitors .319

OFFSITE DOSE CALCULATION MANUAL PRage 76of319 PART 1.0: PURPOSE AND APPLICABILITYISCOPE PURPOSE A. The ODCM lists the radiological effluent controls established by Technical Specifications l 5.5.1 and 5.5.4 at Calvert Cliffs Nuclear Power Plant (CCNPP) for the purpose of ensuring the amount of radioactive materials released to the environment are as low as reasonably achievable.

B. The ODCM defines parameters and methodologies for calculating projected doses and dose rates resulting from the release of radioactive materials in liquid and gaseous effluents.

C. The ODCM defines parameters and methodologies for calculating alarm and fixed setpoints for Technical Specification related effluent radiation monitoring systems.

D. The ODCM defines and establishes controls for the Radiological Environmental Monitoring Program.

APPLICABILITY/SCOPE A. The information presented in this manual is applicable to any division, department, section, or unit within the Calvert Cliffs Nuclear Power Plant (CCNPPI) which is either wholly or partly responsible for performing any of the activities listed in the PURPOSES section of this procedure. Responsible organizations include, but are not limited to, the following:

1. CCNPPI, CHEMISTRY SECTION

2. Constellation Generation Group, Technical Services, Analytical Chemistry Unit Cu

3. CCNPPI, Electrical and Controls (E&C) Section

4. CCNPPI, Operations Section

5. CCNPPI, Radiation Safety Section B. This manual is applicable to the determination of alarm and fixed setpoints for the following radioactive gaseous effluent monitoring Instrumentation:

1. O-RE-2191

2. 1/2-RE-5415

3. 1/2-RE-5416

-OFFSITE DOSE CALCULATION MANUAL Rev. 7 C. This manual is applicable to the determination of alarm and fixed setpoints for the following radioactive liquid effluent monitoring instrumentation:

1. O-RE-2201

2. 1/2-RE-4014

3. 1/2-RE-4095 -

D. This manual is applicable to the determination of the offsite doses and/or offsite- dose rates due to the following: ...

1. Radioactive material in gaseous waste discharged from CCNPP

2. Radioactive material in liquid waste discharged from CCNPP

3. Radioactive material contained in outside storage tanks at CCNPP E. This manual is applicable to the determination of the radiological effects on the environment due to the presence of Calvert Cliffs Nuclear Power Plant.

F. The ODCM is based on Technical Specifications and CCNPPls interpretation of industry standards arid practices. '

OFFSITE DOSE CALCULATION MANUAL RPge 7of319 PART 2.0: DEFINITIONS AND REFERENCES DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Controls.

ABNORMAL AND/OR UNANTICIPATED RADIOACTIVE RELEASE:

Any unplanned or uncontrolled release of radioactive material from the SITE BOUNDARY.

ACTION:

That part of a control which prescribes remedial measures required under designated conditions.

BATCH RELEASE:

A BATCH RELEASE is the discharge of liquid (or gaseous) waste of a discrete volume. (NUREG-0133, page 14)

CHANNEL CALIBRATION:

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION 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.

CHANNEL FUNCTIONAL TEST:

a. An analog CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

b. A bistable CHANNEL FUNCTIONAL TEST shall be injection of a simulated signal into the channel sensor to verify OPERABILITY including alarm and/or trip functions.

COMPOSITE SAMPLE:

A COMPOSITE SAMPLE is a combination of individual samples obtained at intervals that are very short (e.g., hourly) in relation to the compositing time interval (e.g., monthly) to assure obtaining a representative sample. The sample volume should be proportionate to the volume of fluid, either liquid or gas, flowing through the system.

CONTINUOUS RELEASE A CONTINUOUS RELEASE is the discharge of liquid (or gaseous) wastes of a nondiscreet volume; e.g., from a volume or system that has an input flow during the CONTINUOUS RELEASE. (NUREG-0133, P. 14.)

Page 9 of 319

-OFFSITE DOSE CALCULATION MANUAL Page------

CONTINUOUS SAMPLING FREQUENCY A CONTINUOUS SAMPLING FREQUENCY is a sampling arrangement that provides for uninterrupted sampling flow under normal operating conditions. During periods of sampling interruption, the requirement for continuous sampling is considered to be met, provided the interruption is less than one hour AND plant conditions are such that no adverse trend in effluents would be present during the period of interruption. Interruptions in continuous sampling during changing plant conditions OR interruptions in continuous sampling for greater than one hour must be evaluated to determine if an ODCM violation has occurred.

DOSE EQUIVALENT IODINE-131:

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per 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 Ill of TID-14844, Calculation of Distance Factors for Power and Test Reactor Sites."'

FREQUENCY NOTATION:

The frequency notation specified for the performance of Surveillance Requirements shall correspond to the intervals defined in the following Table; Frequency Notation Table 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 6 months R At least once per 18 months

- S/U Prior to each reactor startup P Completed prior to each release NA. Not applicable Refueling Interval At least once per 24 months

-OFFSITE DOSE CALCULATION MANUAL Re 7 GAMMA ISOTOPIC ANALYSIS:

A GAMMA ISOTOPIC ANALYSIS is a analytical method of measurement used for the identification and quantification of gamma-emitting radionuclides.

GASEOUS RADWASTE PROCESSING SYSTEM:

A GASEOUS RADWASTE PROCESSING SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

GSC:

GSC stands for General Supervisor Chemistry.

LIQUID RADWASTE PROCESSING SYSTEM:

A LIQUID RADWASTE PROCESSING SYSTEM is the system installed and designed to reduce radioactive liquid effluents. The minimum components necessary for reducing liquid radioactive effluents is either 13 or 14 Reactor Coolant Waste Ion Exchange; however, other equipment in the system may be used in the maintenance of ALARA for liquid radioactive effluents.

LOWER LIMIT OF DETECTION:

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.

MEMBERS OF THE PUBLIC:

MEMBERS OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, 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 and fixed setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain the radioactive effluent controls, radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Reports required by Technical Specification 5.6.2 and 5.6.3.

OPERABILITY:

A system, subsystem, train, component or device shall be operable or have OPERABILITY when it is capable of performing its specified function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other required auxiliary equipment that are required for the system, subcomponent or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE:

An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified below.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 11of319 Operational Modes

- REACTIVITY '--% RATED THERMAL AVERAGE COOLANT MODE CONDITION, Kff ; POWER* TEMPERATURE POWER OPERATION 20.99 >5% - Ž300 OF STARTUP Ž0.99 Ž2300 OF HOT STANDBY <0.99, . . 0.. 2300 0F HOT SHUTDOWN <0.99 0 300 °F > T WM> 200 OF COLD SHUTDOWN <0.99 ' ' °0 *200 IF REFUELING** *0.95 0 *140 °F

Excluding decay heat.

Reactor vessel head unbolted or removed and fuel In the vessel.

PROCESS CONTROL PROGRAM:.

PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 71 and Federal and State and Local regulations governing the disposal of the radioactive waste.

PROJECTED DOSE:

A reasonable estimate of dose expected as a result of future radioactive releases.

PURGE OR PURGING:

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

RATED THERMAL POWER:

RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWt.

SIMULTANEOUS RELEASES:

Simultaneous liquid releases are discharges of liquid radwaste which occur at the same time.

Simultaneous gaseous releases are discharges of gaseous radwaste which occur at the same time excluding main vent discharges.

SITE BOUNDARY:

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

(The SITE BOUNDARY is depicted on Attachment 18, "Environmental Monitoring Sites")

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.

THERMAL POWER:

The THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

-oOFFSITE DOSE CALCULATION MANUAL Page eg12o3912 of 319 TYPICAL RADWASTE RELEASES (OR TYPICAL RADWASTE EFFLUENTS):

TYPICAL RADWASTE RELEASES are defined as (1) all releases conducted during any calendar quarter while either Unit 1 or Unit 2 is in mode 1, and also includes (2) all releases conducted during any calendar quarter following mode 1 operation of either Unit 1 or Unit 2.

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 of/for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST PROCESSING SYSTEM:

A VENTILATION EXHAUST PROCESSING SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or High Efficiency Particulate Air (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 Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING:

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

WASTE GAS HOLDUP SYSTEM:

See GASEOUS RADWASTE PROCESSING SYSTEM.

0-X-

Rev. 7 OFFSITE DOSE CALCULATION MANUAL - - Rev. 13

, Page 7 of 319 REFERENCES DEVELOPMENTAL REFERENCES

1. NUREG-Ol 33, "Preparation Of Radiological Effluent Technical Specification For Nuclear Power Plants", Boegli, J.S., R. R. Bellamy, W. L. Britz, and R. L.

Waterfield, (October 1978).

2. Regulatory Guide 1.109, "Calculation Of Annual Doses To Man From Routine Release Of Reactor Effluents For The Purpose Of Compliance With 10 CFR Part 50, Appendix I," Revision 1, (October 1977).

3'. Regulatory Guide 1.111,"Methods For Estimating Atmospheric Transport And Dispersion Of Gaseous Effluents In Routine Releases From Light-Water-Cooled Reactors," Revision 1, (July 1977).

4. CCNPP System Description Number 14B, "Reactor Coolant Waste Processing System Description."

5. Updated Final Safety Analysis Report, Chapter 11, "Waste Processing And Radiation Protection."

6. CCNPP System Description Number 14D, "Miscellaneous Liquid Waste Processing System Description."

7.01-17D, "Miscellaneous Waste Processing System"

8.01-17C, "Reactor Coolant Waste Processing System"

9. Title 10 of the Code of Federal Regulations, Part 20, Jan 1990 and May 1991

10. Title 10 of the Code of Federal Regulations, Part 50

11. Calvert Cliffs Nuclear Power Plant Semi-Annual Radiological Effluent Release Report (1986, 1987, 1988).

12. Radioactive Decay Data Tables, David C. Kocher,1 981.

13. Radiological Health Handbook, Bureau of Radiological Health, Jan. 1970.

14. TE-001, "Main Vent Stack Flow Measurement"

15. ETP-87-16, "Main Vent Stack Flow Measurement"

16. Verification And Validation Of The Gaseous Release Permit Portion Of The CCNPP EMS Computer Code, J. S. Bland And Associates, July 10, 1990.

17. Land Use Survey InThe Vicinity Of The Calvert Cliffs Nuclear Power Plant September 1991, (for the year 1990), Baltimore Gas And Electric Company, Environmental Programs Section, C. Key, B. E. Helbing

9 DOSE CALCULATION

->OFFSITE MANUAL Page 14 of 319 page14f1

18. CP-607, Revision 1, "Offsite Dose Calculation Manual"

19. NO-1-201, Calvert Cliffs Operating Manual

20. TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites"

21. "Plant Data Book", BGE CCNPP Units 1 and 2, Bechtel Power Corporation, Volume 1,Job 6750.

22. 50.59 Log No. 82-B-999-028-ROO, Safety Analysis No. 2, FCR 82-1053, Supplement 1 (Component Cooling System) [B527]

23. 50.59 Log No. 90-0-074-01 1-R2, Activity MASE 90-7 (Plant Nitrogen System)

[B527]

24. 50.59 Log No. 90-0-029-045-R1, Activity MASE 90-6 (Plant Heating System)

[B527]

25. 50.59 Log No. 90-0-037-044-Ri, Activity MASE 90-5 (Demineralized Water System) [B527]

26. 50.59 Log No. 90-B-012-043-Ri, Activity MASE 90-4 (Service Water System)

[B527]

27. 50.59 Log No. 90-0-027-037-R2, Activity MASE 90-8 (Auxiliary Boiler System)

[B527]

28. Bechtel Power Corporation, Calculation Number M-90-20, "Allowable Radioactive Contamination Levels in the Plant Heating System" [B527]

29. Bechtel Power Corporation, Calculation Number M-90-18, "Allowable Radioactive Contamination Levels in the Nitrogen System Header" [B527]

30. Bechtel Power Corporation, Calculation Number M-90-04, "Allowable Radioactive Contamination Levels in Auxiliary Boiler Water" [B527]

31. Bechtel Power Corporation, Calculation Number M-90-21, "Allowable Radioactive Contamination Levels in the Demineralized Water System" [B527]

32. Bechtel Power Corporation, Calculation Number M-90-19, "Allowable Radioactive Contamination Levels in the Service Water System" [B527]

33. Bechtel Power Corporation, Job Number 11865, Calculation Number 7.4.3-15, "Allowable Radioactive Contamination Levels in the Condensate Storage Tank"

[B527]

34. EN-1-100, Engineering Service Process Overview

35. 01-8A, Blowdown System

-- OFFSITE DOSE CALCULATION MANUAL Page 15 of 319 Page15f1

36. NUREG-1301, "OFFSITE DOSE CALCULATION MANUAL: Standard Radiological Effluent Controls for Pressurized Water Reactors", W. W. Meinke, and T. H. Essig, (Published April 1991).

37. Radiological Environmental Monitoring Program Annual Report for the Calvert Cliffs Nuclear Power Plant Units 1 and 2 January 1 - December 31, 1991 ",

Baltimore Gas And Electric Company, March 1992

38. 50.59 Log No. 90-B-037-120-R2, Activity MASE 90-11 (Condensate Storage Tank) [B527]

39.' Regulatory Guide 4.13, Performance Testing and Procedural Specifications for Thermoluminescence Dosimetry; Environmental Applications PERFORMANCE REFERENCES

1. STP-M-461-1 and STP-M-461-2, Main Vent Stack Flow Measurement CD

2. Technical Specifications

3. Technical Requirements Manual

4. Regulatory Guide 4.13, Performance Testing and Procedural Specifications for Thermoluminescence Dosimetry; Environmental Applications

5. TE-006, Containment Purge Exhaust System HEPA Filter Test

?- '

-OFFSITE DOSE CALCULATION MANUAL Rev. 7 Page 16 of 319 PART 3.0 / 4.0: CONTROLS AND SURVEILLANCE REQUIREMENTS CONTROLS 3.0.1 Compliance with the Controls contained in the succeeding controls is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Control, the associated ACTION requirements shall be met.

3.0.2 Noncompliance with a Control shall exist when the requirements of the Control and associated ACTION requirements are not met within the specified time intervals. If the Control is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.

3.0.3 When a Control is not met, except as provided in the associated ACTION requirements, within one hour ACTION shall be initiated to place the unit in a MODE in which the control does not apply by placing it, as applicable, in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Control. Exceptions to these requirements are stated in the individual controls.

3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Control are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual controls.

3.0.5 When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of its applicable Control, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system(s), subsystem(s), train(s),

component(s) and device(s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , action shall be initiated to place the unit in a MODE in which the applicable Control does not apply by placing it, as applicable in:

1. At least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2 At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and

3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

This specification is not applicable in MODES 5 or 6.

-OFFSITE DOSE CALCULATION MANUAL Rev. 17 7 of 319 Page SURVEILLANCE l, 4.0.1 Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Control unless otherwise stated in an individual Surveillance Requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25 percen't of the specified surveillance interval.

4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Control 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Control. The time limits of the ACTION requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. However, this time of applicability may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement(s) associated with the Control have been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements.

-OFFSITE DOSE CALCULATION MANUAL Page 18 of 319 age18o39 MONITORING INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT CONTROLS 3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their alarm or alarm-with-automatic-termination setpoints set to ensure that the limits of Control 3.11.2.1 are not exceeded. The setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: As shown in Table 3.3-12.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm setpoint less conservative than required by the above Control, without delay suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12.

Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.9 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 operations at the frequencies shown in Table 4.3-11.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 19 of 319 TABLE 3.3-12 - - ;

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

.1 . . .. . ...

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor - Providing 35 Alarm and Automatic Termination of Release

b. Effluent System Flow Rate Measuring I 36 Device

2. MAIN VENT SYSTEM

a. Noble Gas Activity Monitor '-I

37

b. Iodine Sampler I

38

c. Particulate Sampler I

38

d. Tritium Sampler I

38 At all times I

OFFSITE DOSE CALCULATION MANUAL Rev. 7 Page 20 of 319 TABLE 3.3-12 (Continued)

ACTION STATEMENTS ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment:

a. Using the main vent monitor as a backup and recording RMS readings every 15 minutes during the release, or

b. Provided that prior to initiating the release, at least two independent samples of the tank's contents are analyzed, and at least two technically qualified members of the Facility Staff independently verify the release rate calculations and two qualified operators verify the discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 36- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 37- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided either (1)grab samples are taken and analyzed for gross activity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or (2) an equivalent monitor is provided.

ACTION 38- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected as required in Table 4.11-2 with auxiliary sampling equipment.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL- Page 21 of 319 TABLE 4.3-11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL -- MODES IN CHANNEL SOURCE CHANNEL- WHICH INSTRUMENT CHECK

FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST REQUIRED

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor UP PP R(3) *

- Providing Alarm and Automatic Termination of Release

b. Effluent System Flow Rate NA R NA

Measuring Device

2. MAIN VENT SYSTEM M: - ' R (3)

a. Noble Gas Activity Monitor D

b. Iodine Sampler W NA NA NA

c. Particulate Sampler W NA^ - - - NA NA

d. Tritium Sampler W NA NA NA *

A all times ot erth n he the line is v u o t an l At all times other-than when the line is valued out and locked I.. . i

OFFSITE DOSE CALCULATION MANUAL Page 22 of 319 TABLE 4.3-11 (Continued)

TABLE NOTATION The CHANNEL FUNCTIONAL TEST shall also demonstrate the automatic isolation of this pathway and/or Control Room alarm annunciation occurs if the appropriate following condition(s) exists:

1. Instrument indicates measured levels above the alarm-with-automatic-termination setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

The CHANNEL FUNCTIONAL TEST shall also demonstrate that Control Room alarm annunciation occurs if any of the following conditions exists:

1. Instrument indicates measured levels above the alarm setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards traceable to NIST or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system within its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration can be used.

(4) The CHANNEL CHECK shall consist of verifying indication of flow during periods of release and shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during periods in which effluent releases are made.

OFFSITE DOSE CALCULATION MANUAL Page 23 of 319 RADIOACTIVE LIQUID EFFLUENT CONTROLS 3.3.3.10 The radioactive liquid effluent monitoring instrumentation channels shown in Table >

3.3-13 shall be OPERABLE with their alarm or alarm-with-automatic-termination setpoints set to j ensure that the limits of Control 3.11.1.1 are not exceeded. The setpoints of these channels shall be determined and adjusted In accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY: As shown In Table 3.3-13.

ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Control, without delay lO 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.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.

Exert best efforts to return the instruments to OPERABLE status within 30 days and, if unsuccessful, explain in the next Annual Radioactive Effluent Release Report why the inoperability was not corrected in a timely manner.

c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.10 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 operations at the frequencies shown in Table 4.3-12.

OFFSITE DOSE CALCULATION MANUAL Re 7 Page 24 of 319 TABLE 3.3-13 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM INSTRUMENT CHANNELS APPLICABILITY ACTION OPERABLE

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line 1 (1) 28

b. Steam Generator Blowdown Effluent Line 1 (2) 29

2. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line 1 (1) 30

b. Steam Generator Blowdown Effluent Line 1 (2) 30 Note: (1) At all times.

(2) At all times while process flow is present.

S OFFSITE DOSE CALCULATION MANUAL TABLE 3.3-13 (Continued)

Rev. 7 Page 25 of 319 ACTION STATEMENTS ACTION 28 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may continue provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Control 4.11.1.1.1, and

b. At least two technically qualified members of the Facility Staff independently verify the release rate calculations and two qualified operators verify the discharge valve line up.

ACTION 29- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for gross radioactivity (beta or gamma) at the LOWER LIMIT OF DETECTION defined in Table 4.11-1:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcurie/gram DOSE EQUIVALENT 1-131.

b. At least once per 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie/gram DOSE EQUIVALENT 1-131.

ACTION 30- With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump performance curves may be used to estimate flow.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 26 of 319 TABLE 4.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONA INSTRUMENT CHECK CHECK CALIBRATION L TEST

1. GROSS RADIOACTIVITY MONITORS PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE

a. Liquid Radwaste Effluent Line D P R(2)

b. Steam Generator Blowdown Effluent Line D P R(2)

2. FLOW RATE MEASUREMENT DEVICES

a. Liquid Radwaste Effluent Line D(3) NA R NA

b. Steam Generator Blowdown Effluent Line D () NA R NA

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 27of319 TABLE 4.3-12 (Continued)

TABLE NOTATION (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and/or Control Room alarm annunciation occur if the appropriate following condition(s) exists:

1. Instrument indicates measured levels above the alarm or fixed setpoint.

2. Circuit failure.

3. Instrument indicates a downscale failure.

(2)

The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards traceable to NIST or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system within its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration can be used.

'3' (3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during periods in which effluent releases are made.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 28 of 319 RADIOACTIVE EFFLUENTS LIQUID EFFLUENTS:

Concentration CONTROLS 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.

APPLICABILITY: At all times.

ACTION:

a. With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, without delay restore the concentration to within the above limits.

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1.

4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Control 3.11.1.1.

Rev. 7.

OFFSITE DOSE CALCULATION MANUAL Page 29 of 319 TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LIQUID AMPIN G ANALYSIS TYPE OF ACTIVITY LOWER LIMIT OF RELEASE TYPE FREQUENCY F

FREQENCY(LLD) ANALYSIS DETECTION m (MCIIml)

A. Batch Waste P 1P Principal Gamma 5x10' Releasesb Each Batch Each Batch Emittersc 131 . .1x10 Mo-99, Ce-144 . 2x104 P M H-3 1x10o Each Batch Composited Gross Alpha 1x10' P Q Each Batch Composited Sr-89, Sr-90 5x104 B. Turbine Building M M Principal Gamma 5x10' Sump Emittersc 1-131 - 1x10o MO-99, Ce-144 2x10

Q OFFSITE DOSE CALCULATION MANUAL Pgev 700f319 TABLE 4.11-1 (Continued)

TABLE NOTATION a The LLD is defined, for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD = 4.66Sb_____ _ __ __ _

E

V

2.22 x 106

Y

exp(-,At)

Where:

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

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

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

OFFSITE DOSE CALCULATION MANUAL- Page 1 of 319 TABLE 4.11-1 (Continued)

TABLE NOTATION b Prior to sampling Reactor Coolant Waste and Miscellaneous Waste for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.

The principal gamma emitters for which the LLD control applies exclusively are 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 considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3. - lS d-A COMPOSITE SAMPLE is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged.in which the method of sampling employed results in a specimen that is representative of the liquids released.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 32 of 319 LIQUID EFFLUENTS:

Dose CONTROLS 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRESTRICTED AREAS shall be limited:

a. During any calendar quarter to less than or equal to 3.0 mrems to the total body and to less than or equal to 10 mrems to any organ, and

b. During any calendar year to less than or equal to 6 mrems to the total body and to less than or equal to 20 mrems to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission pursuant to 10 CFR 50.4 within 30 days a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective ACTIONS that have been taken to reduce the releases and the proposed corrective ACTIONS to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

OFFSITE

(~~). DOSE CALCULATION MANUAL -Page Rev. 733 of 319 LIQUID EFFLUENTS:

Liquid Radwaste Processinq System CONTROLS 3.11.1.3 The LIQUID RADWASTE PROCESSING SYSTEM shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the calculated doses due to the liquid effluent to UNRESTRICTED AREAS exceeds 0.36 mrem to the total body or 1.20.

mrem to any organ in a 92 day period.

APPLICABILITY: At all times.

ACTION:

a. With radioactive liquid waste being discharged without processing and in excess of the above limits, prepare and submit to the Commission pursuant to 10 CFR 50.4 within 30 days a Special Report that includes the following information:

1. Explanation of why liquid radwaste was being discharged without processing, identification of any Inoperable equipment or subsystems, and the reason for the inoperability,

2. ACTION(s) taken to restore the inoperable equipment to OPERABLE status, and

3. -Summary description of ACTION(s) taken to prevent a recurrence.

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.3 Doses over a 92-day period due to liquid releases to UNRESTRICTED AREAS shall be calculated at least once per 31 days in accordance with the methodology and parameters in the ODCM.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 34 of 319 GASEOUS EFFLUENTS:

Dose Rate CONTROLS 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see figure in UFSAR Chapter 1) shall be limited to the following:

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

b. For iodine-1 31 and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to 1500 mrems/yr to any organ.

APPLICABILITY: At all times.

ACTION:

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

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.1.1 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 in the ODCM.

4.11.2.1.2 The dose rate due to iodine-131 and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL- Pge 35of319 TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM GASEOUS SAMPLING MINIMUM - TYPE OF LOWER LIMIT RELEASE TYPE FREQUENCY ANALYSIS ACTIVITY OF DETECTION FREQUENCY ANALYSIS (LLD)' UCiml)

A. Waste Gas p P Principal Gamma I x 104 Storage'Tank Each Tank Each Tank Emittersb

- - Grab Sample (Gaseous Emissions Only)

B. Containment P P Principal Gamma 1 x 104 Purge and Vent Each Batch Each Batch Emittersb Grab Sample (Gaseous Emissions Only)

C. Main Vent Mc Mc Principal Gamma 1x 104 Grab Sample Emittersb (Gaseous Emissions Only)

Continuousd M H-3 1x 104 Continuousd W 1-131 1 x 10.12 Charcoal Sample Continuousd. W Principal Gamma I x 10l Particulate Emittersb Sample' (1-131, others)

Continuousd M Gross Alpha I x 10 Composite Particulate Sample Continuousd Q Sr-89, Sr9O I x 10.11 Composite Particulate Sample Continuousd Noble Gas Noble Gases Ix 104 Monitor Gross Beta or Gamma D. Incinerated 0 i1 p P Principal Gamma 5 x 107 Each Batch ach Batchh Emitters

OFFSITE DOSE CALCULATION MANUAL Rage 367of39 TABLE 4.11-2 (Continued)

TABLE NOTATION a The LLD is defined, for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD = 4.66 Sb E e V

2.22x1060 Y

exp (-AA t)

Where:

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

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

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

OFFSITE DOSE CALCULATION MANUAL Page 37 of 319 TABLE 4.11-2 (Continued)

TABLE NOTATION b The principal gamma emitters for which the LLD control applies exclusively are the-following radionuclides: Kr-87, Kr-88, Xe-1 33, Xe-1 33m, Xe-1 35, and Xe-I 38 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-1 34, Cs-137, Ce-141 and Ce-144 for particulate emissions. This list does not mean, that only these nuclides are to be considered. Other gamma peaks that are identifiable,' together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3.

CD)

Sampling and analysis shall also be performed following shutdown, STARTUP, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1)analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than alfactor of 5, and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 5.

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 Controls 3.11.2.1, 3.11.2.2 and 3.11.2.3.

e Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing, or after removal from sampler. When sample collection time is less than seven days, the corresponding LLDs may be increased by a proportional factor.

Collect sample and analyze daily for total Curie content per Technical Requirements Manual 15.11.2.1 when the Reactor Coolant System specific l )

activity of Xe-1 33 is greater than 150 uCi/ml.

g Incinerated oil may be discharged via points other than the main vent (e.g.,

Auxiliary Boiler). Releases shall be accounted for based on pre-release grab sample data.

h Samples for incinerated oil releases shall be collected from and representative of filtered oil in liquid form.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 38 of 319 GASEOUS EFFLUENTS:

Dose - Noble Gases CONTROLS 3.11.2.2 The air dose due to noble gases released in gaseous effluents to areas at and beyond the SITE BOUNDARY (see figure in UFSAR Chapter 1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation and,

b. During any calendar year: Less than or equal to 20 mrads for gamma radiation and less than or equal to 40 mrads for beta radiation.

APPLICABILITY: At all times.

ACTION:

a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission pursuant to 10 CFR 50.4 within 30 days a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective ACTIONS that have been taken to reduce the releases and the proposed corrective ACTIONS to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.2 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 in the ODCM at least once per 31 days.

OFFSITE DOSE CALCULATION MANUAL Rev. 7 Page 39 of 319 GASEOUS EFFLUENTS:

Dose - lodine-131 & Radionuclides in Particulate Form CONTROLS 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131 and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released to areas at and beyond the SITE BOUNDARY (see figure in UFSAR Chapter 1) shall be limited to the following:

a. During any calendar quarter: Less thanbr equal to 15 mrems to.an organ and,

b. During any calendar year: Less than or equal to 30 mrems to any organd.

c. Less than 0.1% of the limits of 3.11.2.3(a) and (b)as a result of burning contaminated oil.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of iodine-131 and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission pursuant to 10 CFR 50.4 within 30 days a Special Report that identifies the cause(s) for exceeding the liimit and defines the corrective ACTIONS that have been taken to reduce the releases and the proposed corrective ACTIONS to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.3 Cumulative dose contributions for the current calendar quarter and the current calendar year for iodine-131 and radionuclides in particulate form with half lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

OFFSITE DOSE CALCULATION MANUAL PRage 470of319 GASEOUS EFFLUENTS:

Gaseous Radwaste Processing System CONTROLS 3.11.2.4 The GASEOUS RADWASTE PROCESSING SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the gaseous effluent air doses due to gaseous effluent releases, to areas at and beyond the SITE BOUNDARY (see figure in UFSAR Chapter 1) exceeds 1.20 mrad for gamma radiation and 2.4 mrad for beta radiation in a 92 day period. The VENTILATION EXHAUST PROCESSING SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the calculated doses due to gaseous effluent releases, to areas at and beyond the SITE BOUNDARY (see figure in UFSAR Chapter

1) exceeds 1.8 mrem to any organ in a 92 day period.

APPLICABILITY: At all times.

ACTION:

a. With gaseous waste being discharged without processing and in excess of the above limits, prepare and submit to the Commission pursuant to 10 CFR 50.4 within 30 days a Special Report that includes the following information:

1. Explanation of why gaseous radwaste was being discharged without processing, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. ACTION(s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of ACTION(s) taken to prevent a recurrence.

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.4 Doses over a 92-day period due to gaseous releases shall be calculated at least once per 31 days in accordance with the methodology and parameters in the ODCM.

OFFSITE DOSE CALCULATION MANUAL PRge 41 of319 TOTAL DOSE CONTROLS 3.11.4 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 total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

APPLICABILITY: At all times.

ACTION:

a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Control 3.11.1.2.a, 3.11.1.2.b, 3.11.2.2.a, 3.11.2.2.b, 3.11.2.3.a, or 3.11.2.3.b, 6calculatiofis shall be made including direct radiation contributions from the reactor units and outside storage tanks to determine whether the above limits of Control 3.11.4 have been exceeded. If such is the case,'prepare and submit to the Commission pursuant to 10 CFR 50.4 within 30 days a Special Report that defines the corrective ACTION to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. 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 Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff ACTION on the request is complete.

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS' 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Controls 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.

4.11.4.2 Cumulative dose contributions from direct radiation from the reactor units and outside storage tanks shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in Control 3.11.4.a.

OFFSITE DOSE CALCULATION MANUAL Rage 42of319 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM CONTROLS 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Technical Specification l 5.6.2, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant effluents in an environmental sample at a specified location exceeding the reporting levels of Table 3.12-2, prepare and submit to the Commission pursuant to 10 CFR 50.4 within 30 days after receiving the sample 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 3.11.1.2, 3.11.2.2, and 3.11.2.3. When more than one of the radionuclides in Table 3.12-2 are detected in the sample this report shall be submitted if:

concentration (1) + concentration (2) + , 1.0 reporting level (1) reporting level (2)

When radionuclides other than those in Table 3.12-2 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 is equal to or greater than the calendar year limits of Controls 3.11.1.2, 3.11.2.2 and 3.11.2.3. 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.

gOFFSITE DOSE CALCULATION T DS MANUAL .Page Rev. 7fof 319 43

c. With fresh leafy vegetable samples unavailable from one or more of the sample locations required by Table 3.12-1, identify locations for obtaining replacement samples and add them to the radiological environmental monitoring program within'30 days. The specific locations from which samrples were unavailable may then be deleted from the monitoring program. Pursuant to Technical l>

Specification 5.6.2, identify the cause of the unavailability of samples and identify l O the new location(s) for obtaining the replacement samples in the next Annual Radiological Environmental Operating Report.

d. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS A 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the table and figure(s) in'the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 44 of 319 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE SAMPLING AND TYPE AND EXPOSURE PATHWAY SAMPLES AND SAMPLE COLLECTION FREQUENCY AND/OR SAMPLE LOCATIONSa FREQUENCY OF ANALYSIS

1. DIRECT RADIATION" 23 routine monitoring stations At least Quarterly Gamma dose at least (DR1 - DR23) either with two or more quarterly dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:

an inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDARY (DRI - DR09);f an outer ring of stations, one in each meteorological sector in the 6- to 8-km range from the site (DR10 - DR18);

the remaining stations (DR1 9 -

DR23) to be placed in special interest areas such as population centers, nearby residences, schools, and in 1area to serve as a control station.

Rev. 7 OFFSITE DOSE 1 ........

CALCULATION MANUAL Page 45 of 319 TABLE 3.12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOUREPATWAY NUMBER OF REPRESENTATIVE I -SAMPLING AND - TYPE AND EXPOUREPATWAY SAMPLES AND SAMPLE COLLECTION FREQUENCY AND/OR SAMPLE LOCATIONS' 1 FREQUENCY OF ANALYSIS

2. AIRBORNE -

Radioiodine and Samples from 5 locations (Al-A5): Continuous sampler RadioiodineCanister Particulates operation with 1-131 analysis 3 samples (A1-A3) from close to sample collection weekly.

the 3 SITE BOUNDARY locations, weekly, or more in different sectors of the highest frequently if required Particulate Sampler:

calculated annual average ground- by dust loading. Gross beta level D/Q. radioactivity analysis following filter I sample (A4) from the vicinity of a change; Gamma community having the highest isotopic analysisd of calculated annual average ground- composite level D/Q. (by location) quarterly I sample (A5) from a control location, as for example 15-30 km distant and in the least prevalent wind direction.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 46 of 319 TABLE 3.12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE PATHWAY NUMBER OF REPRESENTATIVE SAMPLING AND TYPE AND AND/OR SAMPLE SAMPLES AND SAMPLE COLLECTION FREQUENCY LOCATIONSa FREQUENCY OF ANALYSIS

3. WATERBORNE

a. Surface 1 sample at intake area (Wal) COMPOSITE GAMMA ISOTOPIC 1 sample at discharge area (Wa2) SAMPLE over 1- ANALYSISd monthly.

month period' Composite for tritium analysis quarterly

b. Sediment from 1 sample from downstream area with Semiannually GAMMA ISOTOPIC shoreline existing or potential recreational value ANALYSISd (Wbl) semiannually

Rev. 7 OFFSITE DOSE CALCULATION MANUAL - Page 47 of 319

.1 TABLE 3.12-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATI' VE SAMPLING AND TYPE AND EXPOSURE PATHWAY SAMPLES AND SAMPLE COLLECTION FREQUENCY AND/OR SAMPLE LOCATIONS' FREQUENCY OF ANALYSIS

4. INGESTION

a. Fish and 3 samples of commercially and/or - Sample in season, or - GAMMA ISOTOPIC Invertebrates recreationally important species (2 fi sh semiannually if they - ANALYSISd on

species and 1 invertebrate species) in are not seasonal edible portions vicinity of plant discharge area (lal-la3).

3 samples of same species in areas not influenced by plant discharge (la4-la6).

b. Food Products Samples of 3 different kinds of broad Monthly during Gamma isotopicd and leaf vegetation grown near the SITE growing season 1-131 analysis BOUNDARY at 2 different locations of highest predicted annual average, ground level DIQ (Ib1-Ib6).f 1 sample of each of the similar broad Monthly during Gamma isotopicd and leaf vegetation grown 15-30 km' distant -growing season 1-131 analysis in the least prevalent wind direction (Ib7-lb9).

OFFSITE DOSE CALCULATION MANUAL Page 48of319 TABLE 3.12-1 (Continued)

TABLE NOTATION a The code in parenthesis, e.g., DR1, Al, defines generic sample locations in this control that can be used to identify the specific locations in the map(s) and table in the ODCM. Specific parameters of distance and direction sector from the central point between the two containment buildings and additional description where pertinent, is provided for each sample location in Table 3.12-1, and in a table and figure(s) in the ODCM. Referto NUREG-0133, "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants, October 1978", and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. 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 unobtainable 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 Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2. 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 X 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. Pursuant to Technical Specification 5.6.2, 4 identify the cause of the unavailability of samples for that pathway and identify C the new location(s) for obtaining samples in the next Annual Radiological Environmental Operating Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 49of319 TABLE 3.12-1 (Continued)

TABLE NOTATION 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 a packet are considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct radiation. 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. Due to the geographical limitations, 9 sectors are monitored around the Calvert Cliffs Nuclear Power Plant.

c Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, GAMMA ISOTOPIC ANALYSIS shall be performed on the individual samples.

GAMMA ISOTOPIC ANALYSIS means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

e 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.

f. Exception to these locations is in the South Sector where DR7, Al, 1b4, 1b5, and 1b6 are located approximately 0.7 km from the release point. This location is conservative with respect to the site boundary, which is located approximately 2.1 km from the release point.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 50 of 319 TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS AIRBORNE FISH & FOOD ANALYSIS WATER PARTICULATE INVERTEBRATES MILK PRODUCTS (PC ill) OR GASES(P!I (pCi/M 3 ) (pCilkg, wet) (pCi/kg, wet)

H-3 20,000^

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400 1-131 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-1 37 50 20 2,000 70 2,000 Ba-La-140 200 300 t For drinking water samples. This is a 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/I may be used.

OFFSITE-DOSE ~CALCULATION. I MANUAL

.I I . I Rev.

Page 751 of 319 TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSISab LOWER LIMIT OF DETECTION (LLD)C AIRBORNE FISH &

WATER - PARTICULATE INVERTEBRATES MILK FOOD SEDIMENT (pCil) OR GASES (pCI/kg, wet) (pCi) (pCikg wet) (pCilkgdry)

(PCIIM 3) (C/ wt Gross Beta 4 0.01 H-3 2,000' Mn-54 15 130 Fe-59 30 260 Co-58,60 15 130 Zn-65 30 260 Zr-Nb-95 15 1-131 1d 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15

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

OFFSITE DOSE CALCULATION MANUAL RPge 72of319 TABLE 4.12-1 (Continued)

TABLE NOTATION a This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2.

b Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.

The LLD is defined, for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD = 4.66 Sb E V 2.22 eYe exp(-AA t)

Where:

LLD is the "Apriori" LOWER LIMIT OF DETECTION as defined above, as picocuries per unit mass or volume, Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate, as counts per minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable, X is the radioactive decay constant for the particular radionuclide, and

OFFSITE.DOSE CALCULATION . I MANUAL Rev Page 553 of 316 TABLE 4.12-1 (Continued)

TABLE NOTATION At for environmental samples is the elapsed time between sample collection, or end of the sample collection period,)and time of counting.

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

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a nosteriori (after the fact) limit for a particular measurement. Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2. Ia 4D 0

d LLD for drinking water samples. If no drinking water pathway exists, the LLD of GAMMA ISOTOPIC ANALYSIS may be used.

,. , ._ I<. . i jr.,

. , T . . , t

- t OFFSITE DOSE CALCULATION MANUAL PRage 754Of319 RADIOLOGICAL ENVIRONMENTAL MONITORING:

LAND USE CENSUS CONTROLS 3.12.2 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles), the location in each of the 9 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden of greater than 50 m2 (500 ft2) producing broad leaf vegetation. (For elevated releases as defined in Regulatory Guide 1.1 11, Revision 1, July 1977, the land use census shall also identify within a distance of 5 km (3 miles) the locations in each of the 9 meteorological sectors of all milk animals and all gardens of greater than 50 m2 producing broad leaf vegetation).

APPLICABILITY: At all times.

ACTION:

a. 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 Control 4.11.2.3, identify the new location(s) in the next Annual Radiological Environmental Operating Report, pursuant to Technical Specification 5.6.2.

b. With a land use census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Control 3.12.1, add the new location(s) to the radiological environmental monitoring program within 30 days. The sampling location(s), excluding the control station location, 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. Pursuant to Technical Specification 5.6.2, identify the new location(s) in the next Annual Radiological Environmental Operating Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.2 The land census shall be conducted during the growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2.

Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted DWQs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 3.12-1.4b shall be followed, including analysis of control samples.

OFFSITEIDOSE CALCULATION MANUAL - . .-. !

Rev. 55 Page' 7 of319 RADIOLOGICAL ENVIRONMENTAL MONITORING:

INTERLABORATORY COMPARISON PROGRAM CONTROLS 3.12.3 Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

a. With analyses not being performed as required above, report the corrective ACTIONS taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Technical Specification IJa 5.6.2. Qz 0

b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS -

4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM. 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 Technical 0 Specification 5.6.2.' I -

w

. I .

I x. . .. .

- ... . . I, I . . . . . I -

I;

OFFSITE DOSE CALCULATION MANUAL PRage 576of319 PART 5.0: CALCULATIONAL METHODOLOGIES RADIOACTIVE LIQUID EFFLUENTS RELEASE PATHWAYS IB5271

1. Introduction a) Radioactive liquid waste generated as a result of operating the Calvert Cliffs Nuclear Power Plant (CCNPP) may be released to the Chesapeake Bayl.

b) There are three pathways by which waste water, non-radioactive and radioactive, l may be discharged from the site to the bay:

(1) Outfall 001, (2) Outfall 003, (3) Outfall 004.

c) A diagram which shows the location of each outfall is included as Attachment 1.

2. Outfall 001 a) Water from the Chesapeake Bay is pumped through the condensers and is discharged to the Chesapeake Bay through the circulating water discharge conduits, known as outfall 001.

b) The liquid radioactive waste is mixed with and diluted by the circulating water prior to exiting the discharge conduit.

c) The circulating water discharge conduit extends 850 feet into the Chesapeake bay.

d) The circulating water discharge conduit may accept liquid radioactive waste from various sources. Sources which may contribute radioactive material to the circulating water discharge conduit are tabulated in Attachment 2.

3. Outfalls 003 and 004 a) There are two other potential pathways for the release of radioactive liquids to the bay. These pathways are designated outfall outfall 003 and outfall 004.

b) Sources which could potentially contribute radioactive material to each of these outfalls are tabulated in Attachment 2.

c) Outfall 002 was modified in 2001 to be discharged via outfall 001.

4. Unmonitored release paths not shown on Attachment 3 should be evaluated and added to the ODCM as necessary.

The federal controls and administrative limits associated with the release of radioactive materials from CCNPP are discussed elsewhere in this document.

OFFSITE DOSE CALCULATION MANUAL PRage 57Of319

5. Safety evaluations have been conducted for operating the following systems after they have become contaminated:

a) component cooling water system (1) In accordance with applicable safety evaluations (Ref. 22), continued operation of this system is allowed as long as the concentration of -

radionuclides in the component cooling water system is less than 3,000 MPCs.

b) plant heating system (1) In accordance with applicable safety evaluations (Ref. 24 and Ref. 28),

continued operation of this system is allowed as long as the concentration of radionuclides Inthe plant heating system is less than 0.3 MPCs.

c) auxiliary boiler system (1) In accordance with applicable safety evaluations (Ref. 27 and Ref. 30) continued operation of this system is allowed as long as the concentration of radionuclides in the auxiliary boiler steam drum is less than 96 MPCs.

d) demineralized water system (1) In accordance with applicable safety evaluations (Ref. 25 and Ref. 31),

continued operation of this system is allowed as long as the concentration of radionuclides in the demineralized water system is less than 0.3 MPCs.

e) service water system (1) In accordance with applicable safety evaluations (Ref. 26 and Ref. 32),

continued operation of this system is allowed as long as the concentration of radionuclides in the service water system is less than 0.1 MPCs.

f) condensate storage tank (1) In accordance with applicable safety evaluations (Ref. 33), continued operation of this system is allowed as long as the concentration of radionuclides in the condensate storage tank is less than 1.0 MPCs.

OFFSITE DOSE CALCULATION MANUAL PRage 578of319 TYPES OF LIQUID RELEASES

1. Liquid radwaste discharges have been classified as CONTINUOUS or BATCH as shown on Attachment 2.

PROCESSING EQUIPMENT

1. Simplified Flow Diagram a) An overview of the liquid waste processing system, including major equipment and (normal) flow paths, is outlined on Attachment 4.

b) There is no processing equipment for wastes discharged through outfalls 003 and 004; however, the waste shall be analyzed for radioactivity in accordance with the analysis frequencies contained in approved CHEMISTRY SECTION procedures.

2. Modifications a) Licensee initiated major changes to the Liquid Radioactive Waste System shall be reported to the Commission in the Radioactive Effluent Release Report for the period in which the modification to the waste system is completed. The discussion of each change shall contain:

(1) A description of the equipment, components and processes involved; and (2) Documentation of the fact that the change, including the safety analysis, was reviewed and found acceptable by the onsite review function.

The report shall also include changes to the ODCM, in accordance with Technical Specification 5.6.3.

b) A "major' change or modification includes, but is not limited, to the removal or permanent bypass of any of the following:

(1) degassifier (2) reactor coolant waste receiver tank (3) reactor coolant waste monitor tank (4) letdown filter (5) reactor coolant waste ion exchanger (6) miscellaneous waste ion exchanger (7) miscellaneous waste filter (8) miscellaneous waste receiver tank (9) miscellaneous waste monitor tank (10) evaporator

OFFSITE DOSE CALCULATION MANUAL PRaev 59 Of 319

3. Detailed Description a) A detailed description of the liquid waste processing system is beyond the scope of the ODCM.

b) For more information on the Miscellaneous Liquid Waste Processing System see FSAR and System Description SD-071A, "Miscellaneous Liquid Waste li Processing System Description."

c) :For more information on the Reactor Coolant Waste Processing System see System Description SD-071B, "Reactor Coolant Waste Processing System l$

Description."

d) For more information see the Updated Final Safety Analysis Report, Chapter 11, "Waste Processing And Radiation Protection."

LIQUID EFFLUENT RADIATION MONITOR ALARM AND FIXED SETPOINTS

1. Liquid Waste Discharge Radiation Monitor (0-RE-2201) a) General description (1) number of radiation elements: one (2) type of radiation element: in-line scintillation detector (3) output: analog (4) radiation indicator: 0-RI-2201 (5) units for radiation indicator: counts per minute (6) supplier: Westinghouse b) Functions of 0-RE-2201 (1) continuously measure the activity contained in liquid waste discharge line (Control 3.3.3.10)

(2) continuously indicate (via 0-RI-2201) the activity of liquids contained in liquid waste discharge line (Control 3.3.3.10)

(3) alarm (via 0-RI-2201) prior to exceeding 10 CFR 20, Appendix B, Table II, Column 2 limits (per Control 3.11.1.1)

(4) automatically terminate discharges from the liquid waste processing system by closing the discharge isolation valves (MWS-2201-CV, and MWS-2202-CV) whenever the radiation indicator (0-RI-2201) exceeds the fixed setpoint.

c) OPERABILITY of 0-RE-2201 (1) This monitor shall be operable (or have OPERABILITY) when it is capable of performing its specified function(s)

OFFSITE DOSE CALCULATION MANUAL PRage 60of319 (2) The functions of 0-RE-2201 are listed in paragraph (b)above.

d) Monitors equivalent to 0-RE-2201 (1) There are no equivalent monitors for 0-RE-2201.

e) Radiological effluent controls for O-RE-2201 (1) Liquid waste discharges via this pathway may continue if any one of the following two conditions are satisfied (per Control 3.3.3.10).

(a) The liquid waste discharge monitor, 0-RE-2201, is operable and o the alarm-with-automatic-termination setpoint for this monitor is l D set to ensure the concentrations of radioactive materials released ° in liquid effluents to UNRESTRICTED AREAS are less than the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2, or (b) two independent samples are analyzed in accordance with Control 4.11.1.1.1; AND at least two technically qualified members of the Facility Staff independently verify the release rate calculations; AND two qualified operators verify the discharge valve line up.

f) Surveillances for 0-RE-2201 (1) Control 4.3.3.10 requires demonstrating the OPERABILITY of 0-RE-2201 by satisfying the checks, calibrations, and tests listed below:

(a) CHANNEL CHECK within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (b) SOURCE CHECK prior to each release (c) CHANNEL CALIBRATION within the past 18 months (d) CHANNEL FUNCTIONAL TEST within the past 6 six months g) Setpoints for 0-RI-2201 (1) There are three radiation alarm setpoints associated with, or otherwise related to, the liquid waste discharge monitor.

(a) 0-RI-2201 fixed high radiation alarm and automatic termination setpoint (b) 0-RI-2201 adjustable plant computer high radiation alarm and manual termination setpoint (c) 0-RI-2201 low radiation alarm setpoint (2) In order to simplify the setpoint terminology, eliminate ambiguity, and minimize the possibility of misinterpretation, the ODCM will refer to these setpoints as follows:

(a) The 0-RI-2201 fixed high radiation alarm and automatic termination setpoint will be referred to as the fixed setpoint.

OFFSITE DOSE CALCULATION MANUAL,

. . . . ., Rev.

at - I Page -617 of 319 "

(b) The O-RI-2201 adjustable plant computer high radiation alarm and manual termination setpoint will be referred t6 as the'adjustable setpoint.

(c) The O-RI-2201 low radiation alarm setpoint will be referred to as the low setpoint.

(3) Each of these alarm setpoints is described below.

h) Fixed setpoint for O-RI-2201 (1) General information (a) This setpoint is not adjusted for each release.

(b) Whenever the fixed setpoint is exceeded, discharges from the liquid waste processing system will be automatically suspended.

(c) The fixed setpoint corresponds to the maximum concentration of radionuclides allowed in liquid waste discharged from the liquid waste processing system.

(d) The current value for the fixed setpoint is specified in the CCNPP Alarm Manual.

(e) The CCNPP Alarm Manual refers to this alarm and fixed setpoint as the O-RI-2201 Liquid Waste Discharge High'Alarm.

(f) The fixed setpoint is integral to the liquid release discharge monitor, as purchased from the supplier.

(g) The fixed setpoint is administratively controlled by EN-1 -i 00.

(h) The fixed setpoint shall be' calculated as described below 1.

(2) Calculating the fixed setpoint for O-RI-2201 (a) The fixed setpoint for O-RI-2201 shall be calculated as follows:

I Per Control 3.3.3.10.

OFFSITE DOSE CALCULATION MANUAL PRage 62of319 FIXED SETPOINT FOR 0-RI-2201 Snx < Kf{(FdIFu)X[(Al.)(ei)]+Bkg } Eq. IL' Sfix = the fixed setpoint for O-RI-2201 (cpm)

Ksf = a constant, actually a safety factor, which is the ratio of the CCNPP activity limit to the 10 CFR 20 MPC limit (unitless)

The safety factor chosen shall be less than or equal to 1.00. This ensures the fixed setpoint is always less than or equal to the limits of 10 CFR 20.

A safety factor of 1.00 will yield a fixed setpoint which corresponds to 1.0 MPC.

A safety factor of 0.500 will yield a fixed setpoint which corresponds to 0.5 MPCs.

It is recommended that a safety factor of 0.5 be used for calculating the fixed setpoint, however, other values-not to exceed 1.00--may be used as directed by the General Supervisor Chemistry.

The use of a "safety margin" is in accordance with the provisions of NUREG-0133, section 4.1.1, which states, "The alarm and trip setpoints ... should correspond to a value(s) which represents a safe margin of assurance that the instantaneous liquid release limit of 10 CFR Part 20 is not exceeded."

This safety margin will prevent minor fluctuations in the nominal circulating water flow rate (and other statistical aberrations) from adversely impacting the calculated fixed setpoint.

Fd = the dilution water flow rate (gpm) prior to the point of release to UNRESTRICTED AREAS The dilution water flow rate is actually the sum of the minimum circulating water flow rate, the minimum salt water flow rate, and the maximum undiluted radwaste flow rate.

The dilution water flow rate shall be calculated in accordance with equation 2L.

I This formula has been derived from NUREG-0133, Addendum, page AA-1.

OFFSITE DOSE CALCULATION MANUAL. ~ Rev. 7 of31 9 Page '63 DILUTION WATER FLOW RATE Fd = Fe. + F, + Fu Eq. 2Ll FC = the minimum circulating water system flow rate necessary to conduct liquid releases A minimum of two circulating water pumps (one circulating water pump per conduit accepting radwaste) shall be operable when discharging liquid radwaste -- via this monitor -

to outfall 001.

Additional circulating water pumps may be required as specified in approved Chemistry Procedures.

Ifa release of liquid radwaste is to be conducted using less than minimum circulating water flow rate necessary to conduct liquid releases, then the release shall not be'allowed until a new setpoint has been calculated and entered into 0-RI-2201 (unless the existing setpoint is conservative), or otherwise comply with the ACTION requirements of Control 3.3.3.10.

Fsw = the minimum salt water system flow rate necessary to release liquid radwaste Ifthe minimum salt water flow rate available for liquid releases is unknown, difficult to predict, or may decrease during a liquid release, a minimum salt water flow rate of 0 gpm should be used when calculating the fixed setpoint.

Ifa release of liquid radwaste is to be conducted using less than the minimum salt water flow rate necessary to conduct liquid releases, the release shall not be allowed until a new setpoint has been calculated and entered into 0-RI-2201 (unless the existing setpoint is conservative), or otherwise comply with the ACTION requirements of Control 3.3.3.10.

Fu = maximum undiluted radwaste flow rate (gpm)

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment2.

Since the flow rate of undiluted liquid waste (120 gpm maximum) is insignificant relative to the circulating water flow rate (200,000 gpm per circulating water pump), the flow rate of undiluted liquid waste need not be'considered when calculating the dilution water flow rate.

Since the maximum undiluted radwaste flow rate is used for calculating the fixed setpoint, a flow setpoint is not required-for the flow measuring device (0-FE-2199) in the effluent line-to verify compliance with Control 3.3.3.10.

AiLn = specific activity limit for the release of radionuclide, i, to UNRESTRICTED AREAS (calculated In accordance with 10 CFR 20, Appendix B, Table II, Note I as described below; microcuries per milliliter).

In order to calculate a'meaningful and accurate fixed setpoint, the specific radionuclides, i, chosen for calculating the fixed setpoint should correspond to those radionuclides found in TYPICAL RADWASTE RELEASES from CCNPP.

Attachment 5 provides further guidance for determining the identity of those radionuclides, i, to be used to calculate a fixed setpoint.

Values for AiLn shall be calculated, as described below, for each radionuclide.

OFFSITE DOSE CALCULA XTIONLTIO MANUAL MAUAL Rev.

Page 7 64 of 319

-_=-

SPECIFIC ACTIVITY LIMIT FOR RADIONUCLIDE i AIL. = (fl) (ATt) Eq. 3L fi = a fraction which represents the relative activity contribution of nuclide i to the average total effluent activity (unitless)

The average total effluent activity does not include tritium or dissolved and entrained noble gases.

This value may be obtained using the guidance provided on Attachment 5.

ATLf = the sum of the total specific activities of all radionuclides found in a TYPICAL LIQUID RADWASTE RELEASE (microcuries/cm 3 )

This value corresponds to 1 MPC.

ATLn shall be calculated as shown below.

TOTAL SPECIFIC ACTIVITY EQUIVALENT TO 1 MPC

[(f 1) (ATL.)/AA,]1 = I Eq. 4L' AiLt = the specific activity limit for radionuclide, i, as obtained from 10 CFR 20, Appendix B, Table 11,Column 2 (microcuries/cm 3 )

For all the radionuclides found in TYPICAL RADWASTE EFFLUENTS, use the value from 10 CFR 20, Appendix B, Table II, Column 2.

1 = the MPC limit (MPCs) for UNRESTRICTED AREAS This value is based on the MPC limit from 10 CFR 20, Table II, Note 1.

e, = absolute detector efficiency for nuclide, i (cpm/microcuries per milliliter)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

Bkg = an approximation of the detector background prior to initiating the liquid release (cpm)

Instead of using an approximation of the detector background, a value of 0 cpm may be used as the detector background if so desired.

(3) Documenting the fixed setpoint for 0-RI-2201 (a) Whenever the fixed setpoint is calculated, the specific values chosen for each of the parameters shall be documented in accordance with EN-1-100.

(4) Changing the fixed setpoint for 0-RI-2201 This formula has been derived from 10 CFR 20, Table II, Notes 1, 2, and 3.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL, Page 65of319 (a) If the fixed setpoint calculated in accordance with equation 1L exceeds the maximum range of the monitor, the fixed setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(b) The fixed setpoint for 0-RI-2201 may be established at values lower than the maximum allowable setpoint, if desired.

(c) A setpoint change should be initiated whenever any of the parameters identified in equation 1L have changed.

(d) The fixed setpbint should not be changed unless one of the following occurs:

i) the relative activity' of any radionuclide in TYPICAL LIQUID EFFLUENTS, f1, has changed by greater than 10%, and the new radionuclide mixture yields a fixed setpoint which is 10% (or more) lower than the current fixed setpoint,m ii) the minimum dilution water flow rate is not available for a liquid release, iii) the values listed in 10 CFR 20, Table II, column 2 have changed, iv) the radiation monitor has been recently calibrated, repaired, or otherwise altered, or v) the monitor is not conservative in its function (see "Functions of 0-RE-2201" earlier in this section).

(e) EN-1 -100 contains the administrative 2controls associated with changing and approving the fixed setpoint.

As determined in accordance with Attachment 5.

OFFSITE DOSE CALCULATION MANUAL PRage 676of319 i) The adjustable setpoint for O-RI-2201 (1) General information (a) The adjustable setpoint is calculated and adjusted prior to each release from the liquid waste processing system.

(b) The adjustable setpoint is based on the specific activities of the radionuclides present in the undiluted liquid waste (as determined by radiochemical analysis per Control 4.1 1.1.1.1).

(c) Whenever the adjustable setpoint is exceeded, discharges from the liquid waste processing system will be manually suspended.

(d) See 01-17C, "Reactor Coolant Waste Processing System", or 01-170, "Miscellaneous Waste Processing System," for a full list of operator ACTIONS taken in response to this alarm.

(e) The adjustable setpoint corresponds to the maximum concentration of radionuclides anticipated or expected in discharges from the liquid waste processing system.

(f) The value for the adjustable setpoint is recorded on the liquid release permit in accordance with CHEMISTRY SECTION procedures.

(g) This alarm is not integral to the liquid release discharge monitor, as purchased from the supplier.

(h) This alarm is generated by the plant computer, which monitors output from O-RI-2201, and provides an alarm to plant operators when the adjustable setpoint has been exceeded.

(i) A value for the adjustable setpoint shall be calculated prior to each release as shown below.

(2) Calculating the adjustable setpoint for O-RI-2201 (a) The adjustable setpoint is based on the specific activities of the radionuclides in the undiluted liquid waste (as determined by radiochemical analysis per Control 4.11.1.1.1), and shall be calculated as shown below.

Rev. 7

- OFFSITE DOSE CALCULATION MANUAL Page 67 of319 ADJUSTABLE SETPOINT FOR 0-RI-2201 SadJ 5 1.50[ (Al,)(el) Bkg] Eq.5Ll Sadj = the adjustable setpoint for 0-RI-2201 (cpm) 1.50 = a constant, actually a safety factor, which allows for fluctuations in radiation monitor response (unitless)

This safety factor helps ensure the release is not unnecessarily terminated due to (1)electronic anomalies which cause spurious monitor responses, (2) statistical fluctuations in disintegration rates, (3)statistical fluctuations in detector efficiencies, (4) errors associated with sample analysis, or (5)errors associated with monitor calibrations.'

Alt! = specific activity of radionuclide, i, in the undiluted waste stream (microcuries per milliliter) el = absolute detector efficiency for nuclide, I (cpm/microcuries per milliliter)

The detector efficiency for each radionuclide may be calculated frorri data collected during calibration of the radiation monitor. -

Bkg = an approximation of the detector background (cpm)

(3) Documenting the adjustable setpoint for 0-RI-2201 (a) Whenever the adjustable setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with approved CHEMISTRY SECTION procedures.

(4) Changing the adjustable setpoint for O-RI-2201 (a) In all cases, the adjustable setpoint shall be set to a value which is less than or equal to the fixed setpoint.

(b) If the adjustable setpoint exceeds the maximum range of the monitor, the setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(c) CHEMISTRY SECTION procedures contain administrative controls associated with calculating and approving an adjustable setpoint.

(d) Whenever 0-RI-2201 is operable, the calculated value for the adjustable setpoint shall be entered into the plant computer prior to each release from the liquid waste processing system.

The t analysis errors' and 'calibration errors" refer to errors which are within established quality assurance and quality control limits.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 68 of 319 (e) Plant Operating Instructions contain administrative controls associated with entering the adjustable setpoint into the plant computer.

j) The low setpoint for 0-RI-2201 (1) This setpoint is integral to the liquid release discharge monitor, as purchased from the supplier.

(2) The current value for the low setpoint is specified in the CCNPP Alarm Manual.

(3) The low setpoint may be used to determine the OPERABILITY of this monitor (per Control 4.3.3.10, CHANNEL FUNCTIONAL TEST).

(4) The alarm generated by the low setpoint may be used to terminate a release in the event 0-RE-2201 fails (i.e., downscale failure or circuit failure) in accordance with Control 4.3.3.10.

(5) The low setpoint calculations are not described in the ODCM.

(6) Changes to the low setpoint are controlled by EN-1 -100.

2. Steam Generator Blowdown Effluent Radiation Monitors (1/2-RE-4095) a) General description of 1/2-RE-4095 (1) number of radiation elements: one per unit (2) Type of radiation elements: in-line scintillation detectors (3) output: analog (4) Designations for radiation indicators (a) 1-RI-4095 (b) 2-RI-4095 (5) units for radiation indicator: counts per minute (6) supplier: Westinghouse

OFFSITE DOSE CALCULATION DOSE, MANUAL AL -Rev.

Page 769 of 319 b) Functions of 1-2-RE-4095 - - -

(1) continuously measure the activity contained in steam generator blowdown effluent line (Control 3.3.3.10)

(2) continuously indicate (via 112-RI-4095) the activity of liquids contained in the steam generator blowdown effluent line (Control 3.3.3.10)

(3) alarm (via 1/2-RI-4095) prior to exceeding 10 CFR 20, Appendix B, Table II, Column 2 limits (per Control 3.11.1.1).

(4) automatically terminate steam generator blowdown releases to UNRESTRICTED AREAS when the radiation indicator (1/2-RI-4095) exceeds the fixed setpoint c) OPERABILITY of 1/2-RE-4095

- (1) -This monitor shall be operable (or have OPERABILITY) when it is capable of performing its specified function(s).

(2) The' functions of this monitor are listed in section (b) above.

(3) It should be noted that if the steam generator blowdown processing system heat exchangers are bypassed, it is possible for blowdown flow to bypass 1/2-RE-4095. If blowdown flow is allowed to bypass 1/2-RE-4095, the minimum channels OPERABILITY requirement of Control 3.3.3.10 may not be satisfied.'

d) Monitors equivalent to 1/2-RE-4095 (1) 1/2-RE-4014 is normally considered the PRIMARY MONITOR for measuring activity released via the steam generator blowdown processing system, and 1/2-RE-4095 is normally considered the BACKUP MONITOR for measuring activity released via the steam generator blowdown processing system.

- (2) In the event PRIMARY MONITOR (1/2-RE-4014) is inoperable or otherwise unavailable, the BACKUP MONITOR (1/2-RE-4095) may fulfill the measuring, indicating, and alarming functions normally provided by the PRIMARY MONITOR.

(3) 1/2-RE-4014, Steram Generator Blowdown Tank Radiation Monitor is considered to be "e4uivalent" monitor to 1/2-RE-4095 as specified below.

(4) 1/2-RE-4014, Steam Generator Blowdown Tank Radiation Monitor, may perform measurement, indication, alarm, and isolation functions (see "Functions of 1/2-RE-4014" earlier in this section) which limit the concentration of radioactive materials released to UNRESTRICTED AREAS in accordance with Control 3.11.1.1 as long as the following conditions are satisfied:

(a) the OPERABILITY of 1/2-RE-4014 must be demonstrated in accordance with Control 4.3.3.10, Table 3.3-13(lb), and

OFFSITE DOSE CALCULATION MANUAL PRaev 70of319 (b) the blowdown ion exchangers are isolated (or the decontamination factors for all radionuclides are verified to be greater than or equal to one for the duration of the release), and (c) the specific activities of radionuclides in the blowdown tank radiation monitor are representative of the activities of the radionuclides in the blowdown effluent line.

e) Radiological effluent controls for 1/2-RE-4095 (1) Steam generator blowdown releases via this pathway may continue if any one of the following two conditions are satisfied (per Control 3.3.3.10):

(a) A steam generator blowdown monitor (either 1/2-RE-4095 or 1/2-RE-4014) is OPERABLE (see "OPERABILITY of 1/2-RE-4095"; and "OPERABILITY of 1/2-RE-4014" earlier in this section) AND the setpoint for this monitor is set to ensure the concentrations of radioactive materials released in liquid effluents to UNRESTRICTED AREAS are less than the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2, or (b) a grab sample is collected AND analyzed in accordance with the provisions described below:

i) analysis shall determine EITHER the gross gamma activity or gross beta activity of the sample ii) sampling and analysis shall be completed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if the specific activity of the steam generator blowdown is greater than 0.01 microcuries per gram IODINE DOSE EQUIVALENT iii) sampling and analysis shall be completed at least once per 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if the specific activity of the steam generator blowdown is less than or equal to 0.01 microcuries per gram IODINE DOSE EQUIVALENT iv) the LOWER LIMITS OF DETECTION (LLDs) for the gross gamma and/or gross beta analyses are in accordance with the LLDs of Control 4.11.1.1.1

r~. DOSE CALCULATION MANUAL OFFSITE Rev. 7 Page 71 of 319 f) Surveillances for 1/2-RE-4095 (1) Control 4.3.3.10 requires demonstrating the OPERABILITY of 1/2-RE-4095 by satisfying the checks, calibrations, and tests listed below (a) CHANNEL CHECK within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (b) SOURCE CHECK prior to each release (c) CHANNEL CALIBRATION within the past 18 months (d) CHANNEL FUNCTIONAL TEST within the past 6 six months g) Setpoints for.1/2-RI-4095 (1) There are three radiation alarm setpoints associated with the Steam Generator Blowdown Effluent Radiation Monitor.

(a) 1/2-RI-4095 fixed high radiation alarm and automatic termination setpoint (b) 1/2-RI-4095 adjustable plant computer high radiation alarm and manual termination setpoint (c) 1/2-RI-4095 low radiation alarm setpoint (2) In order to simplify the setpoint terminology, eliminate ambiguity, and minimize the possibility of misinterpretation, the ODCM will refer to these setpoints as follows:

(a) The 1/2-RI-4095 fixed high radiation alarm and automatic termination setpoint will be referred to as the fixed setpoint.

(b) The 1/2-RI-4095 adjustable plant computer high radiation alarm and manual termination setpoint will be referred to as the adjustable setpoint.

(c) The 1/2-RI-4095 low radiation alarm setpoint will be referred to as the low setpoint.

(3) Each of these alarm setpoints are described below.

h) The fixed setpoin't for 1/2-RI-4095 (1) General Information

.-(a) The fixed setpoint is not adjusted for each release.

(b) Whenever the fixed setpoint is exceeded, discharges from the steam generator blowdown processing system to UNRESTRICTED AREAS will be automatically suspended.

OFFSITE DOSE CALCULATION MANUAL PRaev 72of319 (c) The fixed setpoint corresponds to the maximum concentration of radionuclides allowed in liquid waste discharged from the steam generator blowdown processing system.

(d) The current value for the fixed setpoint is specified in the CCNPP Alarm Manual.

(e) The CCNPP Alarm Manual refers to this setpoint as the 1/2-RI-4095 Steam Generator Blowdown Recovery High Alarm.

(f) This setpoint is integral to the steam generator blowdown discharge monitor, as purchased from the manufacturer.

(g) The fixed setpoint is administratively controlled by EN-1-1 00.

(h) The fixed setpoint shall be calculated as described below'.

(2) Calculating the fixed setpoint for 1/2-RI-4095 (a) The fixed setpoint for 1/2-RI-4095 shall be calculated in accordance with equation IL.

FIXED SETPOINT FOR 1/2-RI-4095 I-U K~f ((FdIF.) ,[ (AiLl) (el) ] +Bkg ) Eq.IL 2 Sfix = the fixed setpoint for 1/2-RI-4095 (cpm)

Ksf = a constant, actually a safety factor, which is the ratio of the CCNPP activity limit to the 10 CFR 20 MPC limit (unitless)

The safety factor chosen shall be less than or equal to 1.00. This ensures the fixed setpoint is always less than or equal to the limits of 10 CFR 20.

A safety factor of 1.00 will yield a fixed setpoint which corresponds to 1.0 MPC.

A safety factor of 0.500 will yield a fixed setpoint which corresponds to 0.5 MPCs.

It is recommended that a safety factor of 0.5 be used for calculating the fixed setpoint, however, other values--not to exceed 1.00--may be used as directed by the General Supervisor Chemistry.

The use of a "safety margin" is in accordance with the provisions of NUREG-01 33, section 4.1.1, which states, "The alarm and trip setpoints ... should correspond to a value(s) which represents a safe margin of assurance that the instantaneous liquid release limit of 10 CFR Part 20 is not exceeded."

Per Control 3.3.3.10.

2 Equation IL has been derived from NUREG-0133, Addendum, page AA-1.

-OFFSITE DOSE CALCULATION MANUAL OFFITEMANAL 'Rev.

Page~7 73 of 319 This safety margin will prevent minor fluctuations in the nominal circulating water flow rate (and other statistical aberrations) from adversely impacting the calculated fixed setpoint.

Fd = the dilution water flow rate (gpm) prior to the point of release to UNRESTRICTED AREAS the dilution water flow rate is actually the sum of the minimum circulating water flow rate, the minimum salt water flow rate, and the maximum undiluted radwaste flow rate.

The dilution water flow rate shall be calculated in accordance with equation 2L.

DILUTION WATER FLOW RATE Fd = Fw + Fw + FU Eq.2L FCW = the minimum circulating water system flow rate necessary to conduct liquid releases A minimum of two circulating water pumps (one circulating water pump per conduit accepting radwaste) shall be operable when discharging liquid radwaste - via this monitor -- to outfall 001.

The flow rate of each circulating water pump is 200,000 gallons per minute.

Additional circulating water pumps may be required as specified in approved CHEMISTRY SECTION procedures.

If a release of liquid radwaste from the steam generator blowdown system is to be conducted using less than minimum circulating water flow rate necessary to conduct liquid releases, the release shall not be allowed until a new setpoint has been calculated and entered into 1/2-RI-4095 (unless the existing setpoint is conservative), or otherwise comply with the ACTION requirements of Control 3.3.3.10.

Fsw = the minimum salt water system flow rate necessary to release liquid radwaste If the minimum salt water flow rate available for liquid releases is unknown, difficult to predict, or may decrease during a liquid release, a minimum salt water flow rate of 0 gpm should be used

- when calculating the fixed setpoint.

The maximum flow rate for one salt water pump Is listed on Attachment 2.

If a release of liquid radwaste is to be conducted using less than the minimum salt water flow rate necessary to conduct liquid releases, the release shall not be allowed until a new setpoint has been calculated and entered into 1/2-RI-4095 (unless the existing setpoint is conservative),

or otherwise comply with the ACTION requirements of Control 3.3.3.10.

FU = maximum undiluted radwaste flow rate (gpm)

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 2.

Since the steam generator blowdown flow rate (225 gpm maximum) is insignificant relative to the circulating water flow rate (200,000 gpm per circulating water pump), the steam generator blowdown flow rate need not be considered when calculating the dilution water flow rate.

OFFSITE DOSE CALCULATION MANUAL PRaev 74of319 Since the maximum undiluted radwaste flow rate is used for calculating the fixed setpoint, a flow setpoint is not required--for the flow measuring device in the effluent line--to verify compliance with Control 3.3.3.10.

Aln = specific activity limit for the release of radionuclide, i, to UNRESTRICTED AREAS (calculated in accordance with 10 CFR 20, Appendix B, Table II, Note 1 as described below; microcuries per milliliter)

In order to calculate a meaningful and accurate fixed setpoint, the specific radionuclides, i, chosen for calculating the fixed setpoint should correspond to those radionuclides found in TYPICAL RADWASTE RELEASES from CCNPP.

Attachment 5 provides further guidance for determining which radionuclides, i, should be used to calculate a fixed setpoint.

Values for AiL shall be calculated, as described below, for each radionuclide.

SPECIFIC ACTIVITY LIMIT FOR EACH RADIONUCLIDE i AiLn = (hf) (AT,,n) Eq. 3L fj = a fraction which represents the relative activity contribution of nuclide i to the average total effluent activity (unitless)

The average total effluent activity does not include tritium or dissolved and entrained noble gases.

This value may be obtained using the guidance provided on Attachment 5.

ei = absolute detector efficiency for nuclide, i (cpm/microcuries per milliliter)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

ATLn = the sum of the total specific activities of all radionuclides found in a TYPICAL LIQUID RADWASTE RELEASE (microcuries/cm3 )

This value corresponds to 1 MPC.

ATLn shall be calculated as shown below.

TOTAL SPECIFIC ACTIVITY EQUIVALENT TO 1 MPC

[(f)(ATL.)/AiLt = 1 Eq.4L' AiLt = the specific activity limit for radionuclide, i, as obtained from 10 CFR 20, Appendix B, Table II, Column 2 (microcuries/cm 3 )

1 This formula has been derived from 10 CFR 20, Table II, Notes 1, 2, and 3.

Rev. 7 e0k. OFFSITE DOSE CALCULATION MANUAL Page 75 of 319 For all the radionuclides found in TYPICAL RADWASTE EFFLUENTS, use the vale from 10 CFR 20, Appendix B, Table II, Column 2.

1 = the MPC limit (MPCs) for UNRESTRICTED AREAS This value is based on the MPC limit from 10 CFR 20, Table II, Note 1.

Bkg = an approximation of the detector background (cpm)

Instead of using an approximation of the detector background, a value of 0 cpm may be used as the detector background if so desired.

(3) Documenting the fixed setpoint for 1/2-RI-4095 (a) Whenever the fixed setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with EN-1-100.

(4) Changing the fixed setpoint for 1/2-RI-4095 (a) If the fixed setpoint calculated in accordance with equation 1L exceeds the maximum range of the monitor, the fixed setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(b) The fixed setpoint may be established at values lower than the maximum allowable setpoint, if desired.

(c) A setpoint change should be initiated whenever any of the parameters identified in equation 1L have changed.

OFFSITE DOSE CALCULATION MANUAL PRaev 76of319 (d) The fixed setpoint should not be changed unless one of the following occurs:

I) the relative activity' of any radionuclide in TYPICAL LIQUID EFFLUENTS has changed by greater than 10%, and the new radionuclide mixture yields a fixed setpoint which is 10% (or more) lower than the current fixed setpoint, ii) the minimum dilution water flow rate is not available for a liquid release, iii) the values listed in 10 CFR 20, Table II, column 2 have changed, iv) the radiation monitor has been recently calibrated, repaired, or otherwise altered, or v) the monitor is not conservative in its function (see "Functions of 1/2-RE-4095" earlier in this section).

(e) EN-1-1 00 contains the administrative controls associated with changing and approving the fixed setpoint.

i) The adjustable setpoint for 1/2-RI-4095 (1) General information (a) The adjustable setpoint is calculated and adjusted prior to each release from the steam generator blowdown processing system.

(b) The adjustable setpoint is based on the specific activities of the radionuclides present in the undiluted liquid waste (as determined by radiochemical analysis per Control 4.11.1.1.1).

(c) Whenever the adjustable setpoint is exceeded, discharges from the steam generator blowdown processing system will be manually suspended.

(d) See 01-8A for a full list of operator actions taken in response to this alarm.

(e) The adjustable setpoint corresponds to the maximum concentration of radionuclides anticipated or expected in discharges from the steam generator blowdown processing system.

(f) The value for the adjustable setpoint is recorded on the liquid release permit in accordance with CHEMISTRY SECTION procedures.

(g) This alarm is not integral to the steam generator blowdown effluent monitor, as purchased from the supplier.

As determined in accordance with Attachment 5.

. 4 OFFSITE DOSE CALCULATION MANUAL . ... . ... Page' 7

- Rev. 77 of 319

- (h) This alarm is generated by the plant computer which monitors output from 1/2-RI-4095, and provides an alarm to plant operators when the adjustable setpoint has been exceeded.

(i) A value for the adjustable setpoint shall be calculated prior to each release'as shown below.

(2) Calculating the adjustable setpoint for 1/2-RI-4095 (a) The adjustable setpoint is based on the specific activities of radionuclides In the undiluted liquid waste (as determined by radiochemical analysis per Control 4.11.1.1.1), and shall be calculated as'shown below.

ADJUSTABLE SETPOINT FOR 1/2-RI-4095 Sad] < 2.50 [(Al.) (el) + Bkg] Eq. 5L' Where, 5

adj = the adjustable setpoint for 1/2-RI-4095 (cpm) 2.50 = a constant, actually a safety factor, which allows for fluctuation in radiation monitor response (unitless)

This safety factor helps ensure the release is not unnecessarily terminated due to (1) electronic anomalies which cause spurious monitor responses, (2) statistical fluctuations in disintegration rates, (3) statistical fluctuations in detector efficiencies, (4)errors associated with sample analysis, or (5) errors associated with monitor calibrations.2 AiUl = specific activity of radionuclide, i, in the undiluted waste stream (microcuries per milliliter) ei = absolute detector efficiency for nuclide, i (cpm/microcuries per milliliter)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

Bkg = an approximation of the detector background prior to initiating the liquid release (cpm)

(3) Documenting the adjustable setpoint, 1/2-RI-4095 (a) Whenever the adjustable setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with approved CHEMISTRY SECTION procedures.

(4) Changing the adjustable setpoint for 1/2-RI-4095 2 This formula may be derived from NUREG-0133, Addendum, page AA-1.

2 The "analysis errors" and "calibration errors" refer to errors which are within established quality assurance and quality control limits.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 78 of 319 (a) ir the adjustable setpoint exceeds the maximum range or the monitor, the setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(b) In all cases, the adjustable setpoint shall be set to a value which is less than or equal to the fixed setpoint.

(c) Chemistry procedures contain administrative controls associated with calculating and approving an adjustable setpoint.

(d) Whenever 1/2-RI-4095 is operable, the calculated value for the adjustable setpoint shall be entered into the plant computer prior to each release from the steam generator blowdown processing system.

(e) Plant Operating Instructions (e.g., 01-8A) contain administrative controls associated with entering the adjustable setpoint in the plant computer.

j) The low setpoint for 1/2-RI-4095 (1) This alarm is integral to the steam generator blowdown effluent monitor, as purchased from the manufacturer.

(2) The current value for the low alarm setpoint is specified in the CCNPP Alarm Manual.

(3) The low alarm setpoint may be used to determine OPERABILITY of this monitor (per Control 4.3.3.10, CHANNEL FUNCTIONAL TEST).

(4) The low alarm setpoint may be used to terminate a release in the event 1/2-RE-4095 fails (i.e., downscale failure or circuit failure).

(5) Changes to the low alarm setpoint are controlled by EN-1-1 00.

(6) The ODCM does not address the calculations associated with the low alarm setpoint.

3. Steam Generator Blowdown Tank Radiation Monitors (1/2-RE-4014) a) General description of 1/2-RE-4014 (1) number of radiation elements: one per unit (2) Type of radiation element: off-line scintillation detector (3) output: analog (4) Designations for radiation indicators (a) 1-RI-4014 (b) 2-RI-4014

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 79 of 319 (5) units for radiation indicator: counts per minute-

-.(6) supplier: Westinghouse (7) A 3 gallon per minute sample is cooled, passed through the detector, and pumped back into the steam generator blowdown tank.

b) Functions of 1/2-RE-4014 (1) continuously measure the activity contained in an off-line sample of the steam generator blowdown tank (Control 3.3.3.10)

(2) continuously indicate (via 112-RI-4014) the specific activity in an off-line sample of the steam generator blowdown tank (Control 3.3.3.10)

(3) 'alarm (via 1/2-RI-4014) prior to exceeding the 10 CFR 20, Appendix B, Table 11,Column 2 limits (per Control 3.11.1.1)

(4) automatically terminate steam generator blowdown releases to UNRESTRICTED AREAS when the radiation indicator (1/2-RI-4014) exceeds the fixed alarm setpoint.

c) OPERABILITY of 1/2-RE-4014 (1) This monitor shall be operable (or have OPERABILITY) when it is capable of performing its specified function(s)

(2) The functionsof this monitor are listed in section (b) above.

OFFSITE DOSE CALCULATION MANUAL PRage 870of319 d) Monitors equivalent to 1/2-RE-4014 (1) 1/2-RE-4014 is normally considered the PRIMARY MONITOR for measuring activity released via the steam generator blowdown processing system, and 1/2-RE-4095 is normally considered the BACKUP MONITOR for measuring activity released via the steam generator blowdown processing system.

(2) In the event PRIMARY MONITOR (1/2-RE-4014) is inoperable or otherwise unavailable, the BACKUP MONITOR (1/2-RE-4095) may fulfill the measuring, indicating, and alarming functions normally provided by the PRIMARY MONITOR.

(3) 1/2-RE-4095, Steam Generator Blowdown Effluent Radiation Monitor, performs measurement, indication, alarm, and isolation functions equivalent to 1/2-RE-4014 (see "Function of 1/2-RE-4014" earlier in this section), unless 1/2-RE-4095 is bypassed as described below.

(4) It should be noted that if the steam generator blowdown processing system heat exchangers are bypassed, it is possible for blowdown flow to bypass 1/2-RE-4095 (the steam generator blowdown effluent radiation monitor). If blowdown flow is allowed to bypass 1/2-RE-4095, the minimum channels OPERABILITY requirement of Control 3.3.3.10 may not be satisfied (unless 1/2-RE-4014 is operable).

e) Radiological effluent controls for 1/2-RE-4014 (1) Steam generator blowdown releases via this pathway may continue if any one of the following two conditions are satisfied (per Control 3.3.3.10):

(a) A steam generator blowdown monitor (either 1/2-RE-4095 or 1/2-RE-4014) is OPERABLE (see "OPERABILITY of 1/2-RE-4095"; and "OPERABILITY of 1/2-RE-4014" earlier in this section) AND the alarm setpoint for this monitor is set to ensure the concentrations of radioactive materials released in liquid effluents to UNRESTRICTED AREAS are less than the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2, or (b) a grab sample is collected AND analyzed in accordance with the provisions described below:

i) analysis shall determine EITHER the gross gamma activity or gross beta activity of the sample ii) sampling and analysis shall be completed at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if the specific activity of the steam generator blowdown is greater than 0.01 microcuries per gram IODINE DOSE EQUIVALENT iii) sampling and analysis shall be completed at least once per 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if the specific activity of the steam generator blowdown is less than or equal to 0.01 microcuries per gram IODINE DOSE EQUIVALENT

OFFSITE DOSE CALCULATION MANUAIL Page 87of 319 iv) - the LOWER LIMITS OF DETECTION for the gross gamma and/or gross beta analyses are in accordance with the LLDs of Control 4.11.1.1.1 f) Surveillances for 1/2-RE-4014 (1) Control 4.3.3.10 requires demonstrating the OPERABILITY of 1/2-RE-4014 by satisfying the checks, calibrations, and tests listed below (a) CHANNEL CHECK within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (b) SOURCE CHECK prior to each release (c) CHANNEL CALIBRATION within the past 18 months (d) CHANNEL FUNCTIONAL TEST within the past 6 six months g) Setpoints for 1/2-RI-4014 (1) There are three radiation alarm setpoints associated with the Steam Generator Blowdown Tank Radiation Monitor (a) 112-RI-4014 fixed high radiation alarm and automatic termination setpoint (b) 1/2-RI-4014 adjustable plant computer high radiation alarm and manual termination setpoint (c) 112-RI-4014 low radiation alarm setpoint (2) In order to simplify the setpoint terminology,'eliminate ambiguity, and minimize the possibility of misinterpretation, the ODCM will refer to these setpoints as follows:

(a) The 1/2-RI-4014 fixed high radiation alarm and automatic termination setpoint will be referred to as the fixed setpoint.

(b) The 1/2-RI-4014 adjustable plant computer high radiation alarm and manual termination setpoint will be referred to as the adjustable setpoint.

(c) The 1/2-RI-4014 low radiation alarm setpoint will be referred to as the low setpoint.

OFFSITE DOSE CALCULATION MANUAL PRage 82of319 (3) Each of these alarms is described below.

h) The fixed setpoint for 1/2-RI-4014 (1) General Information (a) The fixed setpoint is not adjusted for each release.

(b) Whenever the fixed setpoint is exceeded, discharges from the steam generator blowdown processing system to UNRESTRICTED AREAS will be automatically suspended.

(c) The fixed setpoint corresponds to the maximum concentration of radionuclides allowed in liquid waste discharged from the steam generator blowdown processing system.

(d) The current value for the fixed setpoint is specified in the CCNPP Alarm Manual.

(e) The CCNPP Alarm Manual refers to the fixed setpoint as the 1/2-RI-4014 Steam Generator Blowdown High Alarm.

(f) The fixed setpoint is integral to the liquid release discharge monitor, as purchased from the manufacturer.

(g) The fixed setpoint is administratively controlled by EN-1 -100.

(h) The fixed setpoint shall be calculated as described below'.

(2) Calculating the fixed setpoint for 1/2-RI-4014 (a) The fixed setpoint for 1/2-RI-4014 shall be calculated as follows:

I Per Control 3.3.3.10.

OFFSITE DOSE CALCULATION MANUAL PRage 73of319 IXED SETPOINT FOR 1I2-RI 4014

.S c

-x Ksf((Fd/F.)Y[(AjL.)(ei)]+Bkg } Eq.1L 1'

Where, Sfix = the fixed setpoint for 1/2-RI-4014 (cpm)

KS! - a constant, actually a safety factor, which is the ratio of the CCNPP activity limit to the 10 CFR 20 MPC limit (unitless)

The safety factor chosen shall be less than or equal to 1.00. This ensures the fixed setpoint is always less than or equal to the limits of 10 CFR 20.

A safety factor of 1.00 will yield a fixed setpoint which corresponds to 1.0 MPC.

A safety factor of 0.500 will yield a fixed setpoint which corresponds to 0.5 MPCs.

It is recommended that a safety factor of 0.5 be used for calculating the fixed setpoint, however, other values-not to exceed 1.00--may be used as directed by the General Supervisor Chemistry.

The use of a "safety margin" is in accordance with the provisions of NUREG-0133, SECTION 4.1.1, which states, "The alarm and trip setpoints ... should correspond to a value(s) which represents a safe margin of assurance that the instantaneous liquid release limit of 10 CFR Part 20 is not exceeded."

This safety margin will prevent minor fluctuations in the nominal circulating water flow rate (and other statistical aberrations) from adversely impacting the calculated fixed setpoint.

Fd = the dilution water flow rate (gpm) prior to the point of release to UNRESTRICTED AREAS The dilution water flow rate is actually the sum of the minimum circulating water flow rate, the minimum salt water flow rate, and the maximum undiluted radwaste flow rate The dilution water flow rate shall be calculated in accordance with equation 2L.

Equation IL has been derived from NUREG-01 33, Addendum, page AA-1.

OFFSITE DOSE CALCULATION MANUAL PRage 84of319 DILUTION WATER FLOW RATE Fd = Fc e uEq.2Ll FCW = the minimum circulating water system flow rate necessary to conduct liquid releases A minimum of two circulating water pumps (one circulating water pump per conduit accepting radwaste) shall be operable when discharging liquid radwaste -- via this monitor -- to outfall 001.

The flow rate of each circulating water pump is 200,000 gallons per minute.

Additional circulating water pumps may be required as specified in approved CHEMISTRY SECTION procedures.

If a release of liquid radwaste from the steam generator blowdown system is to be conducted using less than minimum circulating water flow rate necessary to conduct liquid releases, the release shall not be allowed until a new setpoint has been calculated and entered into 1/2-RI-4014 (unless the existing setpoint is conservative), or otherwise comply with the ACTION requirements of Control 3.3.3.10.

Fsw = the minimum salt water system flow rate necessary to release liquid radwaste If the minimum salt water flow rate available for liquid releases is unknown, difficult to predict, or may decrease during a liquid release, a minimum salt water flow rate of 0 gpm should be used when calculating the fixed setpoint.

The maximum flow rate for one salt water pump is listed on Attachment 2.

If a release of liquid radwaste is to be conducted using less than the minimum salt water flow rate necessary to conduct liquid releases, the release shall not be allowed until a new setpoint has been calculated and entered into 1/2-RI-4014 (unless the existing setpoint is conservative),

or otherwise comply with the ACTION requirements of Control 3.3.3.10.

Fu = maximum undiluted radwaste flow rate (gpm)

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 2. l Since the steam generator blowdown flow rate (225 gpm maximum) is insignificant relative to the circulating water flow rate (200,000 gpm per circulating water pump), the steam generator blowdown flow rate need not be considered when calculating the dilution water flow rate.

Since the maximum undiluted radwaste flow rate is used for calculating the fixed setpoint, a flow setpoint is not required--for the flow measuring device in the effluent line--to verify compliance with Control 3.3.3.10.

AiLn = specific activity limit for the release of radionuclide, i, to UNRESTRICTED AREAS (calculated in accordance with 10 CFR 20, Appendix B, Table II, Note 1 as described below; microcuries per milliliter)

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 85 of 319 In order to calculate a meaningful and accurate fixed setpoint, the specific radionuclides, i,- -

chosen for calculating the fixed setpoint should correspond to those radionuclides found in TYPICAL RADWASTE RELEASES from CCNPP.

Attachment 5 provides further guidance for determining which radionuclides, i, should be used to calculate a fixed setpoint. An acceptable alternative is to assume an isotopic mix which results in a more conservative setpoint.,

Values for AiLn shall be calculated, as described below, for each radionuclide.

SPECIFIC ACTIVITY LIMIT FOR RADIONUCLIDE, i AILn = (fa)(ATLn) - Eq. 3L T fj = a fraction which represents the relative activity contribution of nuclide i to the average total effluent activity.(unitless)

The average total effluent activity does not include tritium or dissolved and entrained noble gases.

This value may be obtained using the guidance provided on Attachment 5.

ATLn = the sum of the total specific activities of all radionuclides found in a TYPICAL LIQUID RADWASTE RELEASE (microcuries/cm 3 )

This value corresponds to I MPC.

Calculate the value of ATL as shown below.

TOTAL SPECIFIC ACTIVITY CORRESPONDING TO 1 MPC AT THE SITE BOUNDARY l [(fj)(ATL.)1AiuLt =.I Eq. 4L' AiLt = the specific activity limit for radionuclide, i, as obtained from 10 CFR 20, Appendix B, Table II, Column 2 (microcuries/cm 3 ) ;

For all the radionuclides found in TYPICAL RADWASTE EFFLUENTS, use the value from 10 CFR 20, Appendix B, Table II, Column 2.

I This formula has been derived from 10 CFR 20, Table II,Notes 1,2,and 3.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 86 of 319 1 = --------------------------------------------------------------------------------------------------

the MPC limit (MPCs) for UNRESTRICTED AREAS This value is based on the MPC limit from 10 CFR 20, Table II, Note 1.

eB = absolute detector efficiency for nuclide, i (cpm/microcuries per milliliter)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

Bkg = an approximation of the detector background (cpm)

Instead of using an approximation of the detector background, a value of 0 cpm may be used as the detector background if so desired.

(3) Documenting the fixed setpoint for 1/2-RI-4014 (a) Whenever the fixed setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with EN-1-100.

(4) Changing the fixed setpoint for 1/2-RI-4014 (a) If the fixed setpoint calculated in accordance with equation 1L exceeds the maximum range of the monitor, the fixed setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(b) The fixed setpoint may be established at values lower than the maximum allowable setpoint, if desired.

(c) A setpoint change should be initiated whenever any of the parameters identified in equation 1L (identified in this section of the ODCM) have changed.

(d) The fixed setpoint should not be changed unless one of the following occurs:

i) the relative activity' of any radionuclide in TYPICAL LIQUID EFFLUENTS has changed by greater than 10%, and the new radionuclide mixture yields a fixed setpoint which is 10% (or more) lower than the current fixed setpoint, ii) the minimum dilution water flow rate is not available for a liquid release, iii) the values listed in 10 CFR 20, Table II, column 2 have changed, iv) the radiation monitor has been recently calibrated, repaired, or otherwise altered, or I As determined in accordance with Attachment 5.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL 'Page 87 of 319 '

- v) the monitor is not conservative in its function (see "Functions of 1/2-RE-4014" earlier Inthis section).

(e) 'EN-1-100 contains the administrative controls associated with changing and approving the fixed setpoint.

i) The adjustable setpoint for 1/2-RI-4014 (1) General information (a) The adjustable setpoint is calculated and adjusted prior to each release from the steam generator blowdown processing system.

(b) The adjustable setpoint is based on the specific activities of the radionuclides present in the undiluted liquid waste'(as determined by radiochemical analysis per Control 4.11.1.1.1).

(c) Whenever the adjustable setpoint is exceeded, discharges from the steam generator blowdown processing system will be manually suspended.

(d) See 01-BA for a full list of operator ACTIONS taken in response to this alarm.'

(e) The adjustable setpoint corresponds to the maximum concentration of radionuclides anticipated or expected in discharges from the steam generator blowdownr processing system.

(e) The value for the adjustable setpoint is recorded on the liquid release permit in accordance with CHEMISTRY SECTION procedures.

(g) This alarm is not Integral to the steam generator blowdown tank radiation monitor, as purchased from the supplier. -

(h) This alarm is geherated by the plant computer which monitors output from 1/2-RI-4014, and provides an alarm to plant operators when the 1/2-RI-4014 adjustable setpoint has been exceeded.

- (i) A value for the adjustable setpoint shall be calculated prior to each release as shown below.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 88 of 319 (2) Calculating the adjustable setpoint for 1/2-RI-4014 (a) The adjustable setpoint is based on the specific activity of the radionuclides in the undiluted liquid waste (as determined by radiochemical analysis per Control 4.11.1.1.1), and is calculated as shown below.

ADJUSTABLE SETPOINT FOR 1/2-RI-4014 Sadi < 2.50 [7 (Ail) (e1) + Bkg] Eq. 5L' Sadj = the adjustable setpoint for 1/2-RI-4014 (cpm) 2.50 = a constant, actually a safety factor, which allows for fluctuation in radiation monitor response (unitless)

This safety factor helps ensure the release is not unnecessarily terminated due to (1) electronic anomalies which cause spurious monitor responses, (2) statistical fluctuations in disintegration rates, (3) statistical fluctuations in detector efficiencies, (4) uncertainties associated with sample S.23 analysis, and (5)uncertainties associated with monitor calibrations.2 0 Ai, = specific activity of radionuclide, i, in the undiluted waste stream (microcuries per milliliter) el = absolute detector efficiency for nuclide, i (cpm/microcuries per milliliter)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

Bkg = an approximation of the detector background (cpm)

(3) Documenting the adjustable setpoint for 1/2-RI-4014 (a) Whenever the adjustable setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with approved CHEMISTRY SECTION procedures.

(4) Changing the adjustable setpoint for 1/2-RI-4014 (a) If the adjustable setpoint exceeds the maximum range of the monitor, the setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(b) In all cases, the adjustable setpoint shall be set to a value which is less than or equal to the fixed setpoint.

0 1 Equation 5L has been derived from NUREG-0133, Addendum, page AA-1.

2 The "analysis uncertainties" and "calibration uncertainties" refer to uncertainties which are within established -10 quality assurance and quality control limits.

r~. DOSE CALCULATION MANUAL

- OFFSITE Rev.

Page 7 89 of 319 (c) Chemistry procedures contain administrative controls associated with calculating and approving an adjustable setpoint.

(d) Whenever.1/2-RI-4014 is operable, the calculated value for the adjustable setpoint shall be entered into the plant computer prior to each release from the steam generator blowdown and processing system.

(e) Plant Operating Instructions (e.g., 01-8A) contain administrative controls associated with entering the adjustable setpoint in the plant computer.

j) The low alarm setpoint for 112-RI-4014 (1) This alarm is integral to the steam generator blowdown tank monitor, as purchased from the manufacturer.

(2) The current value for the low alarm setpoint is specified in the CCNPP Alarm Manual.

(3) Changes to the low alarm setpoint are controlled by EN-1 -100.

(4) The ODCM does not address the calculations associated with the low alarm setpoint.

LIMITS ON RADIONUCLIDE CONCENTRATIONS IN LIQUID EFFLUENTS

1. Introduction a) 10 CFR 20, Appendix B, specifies concentration limits associated with the release of radioactive materials to UNRESTRICTED AREAS.

b) Radiological effluent controls have been established to implement the requirements of 10 CFR20.20 --

c) These radiological effluent controls are described below.

2. Radiological Effluent Controls a) The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases (per Control 3.11.1.1)..

OFFSITE DOSE CALCULATION MANUAL PRage 970of319 b) It should be noted that NUREG-0133 specifies that the concentration of radioactive materials in liquid effluents to UNRESTRICTED AREAS shall be limited to 2 E-4 microcuries per milliliter for dissolved or entrained noble gases, this control has not been incorporated into the CCNPP Technical Specifications, and as a result, the ODCM does not include calculations for same.

c) The routine surveillances which are performed to verify compliance with these radiological effluent controls are described below.

3. Surveillance Requirements a) Routine surveillances for BATCH RELEASES (1) Sample each "BATCH" of radioactive liquid waste prior to its release (per Control 4.11.1.1.1).

(2) Determine the concentrations of principle gamma emitters (including, but not limited to, Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-134, Cs-137, Ce-141, I-131, Mo-99, and Ce-144) in the "BATCH" sample, prior to the "BATCH" release (per Control 4.11.1.1.1).

(3) Determine the concentration of tritium contained in a monthly COMPOSITE SAMPLE at least once per month (per Control 4.11.1.1.1).

(4) Determine the concentrations of Sr-89 and Sr-90 contained in a quarterly COMPOSITE SAMPLE at least once per quarter (per Control 4.11.1.1.1).

(5) Calculate the concentrations of radionuclides in the receiving waters at the point the liquid radioactive waste is released to UNRESTRICTED AREAS (per Control 4.11.1.1.2).

b) Routine surveillances for CONTINUOUS RELEASES (1) Sample CONTINUOUS RELEASES of radioactive liquid waste at least once per month (per Control 4.11.1.1.1).

(2) Determine the concentrations of principle gamma emitters (including, but not limited to, Mn-54, Fe-59, Co-58, Co-60, Zn-65, Cs-1 34, Cs-1 37, Ce-141, I-131, Mo-99, and Ce-144) in the undiluted, CONTINUOUS waste stream sample (per Control 4.11.1.1.1).

(3) Calculate the concentrations of radionuclides in the receiving waters at the point the liquid radioactive waste is released to UNRESTRICTED AREAS (per Control 4.11.1.1.2).

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 91 of 319

4. Responsible Plant Organizations - , --

a) The CHEMISTRY SECTION is responsible for performing the sampling, analysis, and calculations required by the above surveillances. -

b) The conditions which initiate the required surveillances are contained in the following section.

5. Initiating Conditions a) The surveillances for BATCH RELEASES shall be performed prior to each BATCH RELEASE' and may be performed more often as specified in approved CHEMISTRY SECTION procedures. ;

b) The surveillances for CONTINUOUS RELEASES shall be performed at least monthly 2 (until the CONTINUOUS RELEASE has been terminated) and, if activity is identified3 in the CONTINUOUS waste stream, the surveillances may be performed more often as specified in approved CHEMISTRY SECTION procedures.

6. Calculation Methodology a) At CCNPP, there are two methodologies for calculating the concentrations of radionuclides in the receiving waters, at the point of release to UNRESTRICTED AREAS, resulting from the discharge of liquid waste.

(1) The rigorous method shall be used IF a computer system and the appropriate software are available.

(2) The simplified method may be used IF a computer system and the appropriate software are NOT available.

(3) These methods, as well as additional supporting information, are presented in the following sections.

b) Rigorous method (1) Solution of the following equation may prove too rigorous for routine use unless a computer system and appropriate software are available.

(2) If a computer system and the appropriate software are available, ensure the concentrations of radionuclides in UNRESTRICTED AREAS are less than one MPC by verifying the following inequality is true:

Per Control 4.11.1.1.1.

2 Per Control 4.11.1.1.1.

The LOWER LIMITS OF DETECTION shall conform to Control 4.1 .1.1.1,Table 4.11-1.

OFFSITE DOSE CALCULATION MANUAL PRage 92of319 LIMIT ON CONCENTRATIONS OF RADIONUCLIDES IN UNRESTRICTED AREAS (RIGOROUS METHOD)

(Fu/Fd) v (AuI/AiLt) < I Eq. 6L' l Fd = the actual dilution water flow rate (gpm) prior to the point of release to UNRESTRICTED AREAS The actual dilution water flow rate is the sum of the circulating water flow rate, the salt water flow rate, and the undiluted radwaste flow rate.

The dilution water flow rate shall be calculated in accordance with equation 2L.

DILUTION WATER FLOW RATE Fd = Fcw + Fsw + Fu Eq. 2L Fcs, = the actual circulating water system flow rate The flow rate of each circulating water pump is 200,000 gallons per minute.

FSW = the salt water system flow rate If the actual salt water flow rate is unknown or otherwise not readily available, a salt water flow rate of 0 gpm may be used in equation 2L.

The maximum flow rate for one salt water pump is listed on Attachment 2.

Fu = actual undiluted radwaste flow rate (gpm)

If the actual undiluted radwaste flow rate is less than 1% of the total dilution water flow rate, an undiluted radwaste flow rate of 0 gpm may be used in equation 2L.

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 2. _Y AiLt = the specific activity limit for radionuclide, i, as obtained from 10 CFR 20, Appendix B, Table II, Column 2 (microcuries/cm 3 )

A.u = the specific activity of nuclide, i, in the undiluted liquid radwaste (microcuries per milliliter) 1 = the MPC limit (MPCs) for UNRESTRICTED AREAS This value is based on the MPC limit from 10 CFR 20, Table II, Note 1.

c) Simplified method I Equation 6L has been derived from 10 CFR 20, Appendix B, Table II, Note 1.

Rev. 7 g F OFFSITE DOSE CALCULATION MANUAL - age 93 of 319 (1) Whenever a computer system and appropriate software are unavaila61e to perform the rigorous calculations described in the previous section, ensure the concentrations of radionuclides in UNRESTRICTED AREAS are less than one MPC by verifying the following inequality is true.

LIMIT ON CONCENTRATIONS OF RADIONUCLIDES IN UNRESTRICTED AREAS

-(SIMPLIFIED METHOD) -

{FU/[(Fd)(Ai1310L)(K.f)]} A;U < 1 Eq.7L0 Fu = actual undiluted radwaste flow'rate (gpm)

If the actual undiluted radwaste flow rate is less than 1%of the total dilution water flow rate, an undiluted radwaste flow rate of 0 gpm may be used in equation 2L.

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 2.

Fd = the actual dilution water flow rate (gpm) prior to the point of release to UNRESTRICTED AREAS The actual dilution water flow rate is the sum of the circulating water flow rate, the salt water flow rate, and the undiluted radwaste flow rate.

DILUTION WATER FLOW RATE Fd = F. + F.w + F. Eq.2L FCW = the actual circulating water system flow rate The flow rate of each circulating water pump is 200,000 gallons per minute.

Fsw ~the actual salt water system flow rate If the actual salt water flow rate is unknown or otherwise not readily available a salt water flow rate of 0 gpm may be used in equation 2L.

The maximum flow rate for one salt water pump is listed on'Attachment 2.

A113ILt = the specific activity limit for 1-131 corresponding to the limiting concentration specified in 10 CFR 20, Appendix B, Table II, Column 2 This value is 3E-7 microcuries per milliliter.

Au = the specific activity of nuclide, i, in the undiluted liquid radwaste (microcuries per milliliter)

Ksf = a constant, actually a safety factor, which is the ratio of the CCNPP activity limit to the activity limit of 10 CFR 20, Appendix B, Table II, Column 2, (unitless) (per Control 3.11.1.1)

The safety factor chosen shall be less than or equal to 1.00. This ensures the activity is always less than or equal to the activity limit of 10 CFR 20, Appendix B, Table II, Column 2.

I This equation has been derived from 10 CFR 20, Appendix B,Table II,Note 1.

OFFSITE DOSE CALCULATION MANUAL Page 794of319 A safety factor of 1.00 will yield a activity which corresponds to the 10 CFR 20, Appendix B, Table 11,Column 2 activity limit.

A safety factor of 0.500 will yield an activity which corresponds to one-half the activity limit of 10 CFR 20, Appendix B, Table II, Column 2.

It is recommended that a safety factor of 1.0 be used for calculating the activity, however, other values--not to exceed 1.00--may be used as directed by the General Supervisor Chemistry.

The particular value selected for the safety factor is somewhat arbitrary, however a safety factor does provide plant personnel with a degree of administrative control over the use of simplified equations for generating radioactive liquid release permits. This administrative control is designed to minimize the possibility of violating 10 CFR 20, Appendix B, Table II, Column 2 limits whenever simplifying assumptions are used.

The use of a safety factor is consistent with the ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as is reasonably achievable.

This safety factor has been included in equation 7L to account for any potential nonconservatism associated with applying the 1-131 MPC limit to all radionuclides identified in the liquid release. Such nonconservatism could conceivable be present whenever radionuclides which have an MPC value higher than the 1-131 MPC limit are present in a liquid release.

d) Once the rigorous or simplified calculations have been completed, the calculation results are compared to the applicable limits and corrective ACTIONS are initiated as described below.

7. Corrective ACTIONS a) CHEMISTRY SECTION surveillance procedures shall contain and/or reference administrative and/or Technical Specification limits for concentration of radionuclides in liquid effluents and shall specify corrective ACTIONS to be initiated when these limits are exceeded.

b) Refer to Control 3.11.1.1 for ACTIONS to be taken in the event the concentrations of radionuclides in UNRESTRICTED AREAS exceed one MPC.

n~7~ Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 95 of 319

- LIMITS ON CUMULATIVE TOTAL BODY DOSES AND CUMULATIVE ORGAN DOSES FOR LIQUID EFFLUENTS

1. Introduction a) Appendix Ito 10 CFR 50 specifies total body dose limits and organ dose limits associated with the release of radioactive liquids to UNRESTRICTED AREAS.

b) Radiological effluent controls have been established to implement the requirements i of 10 CFR 50, Appendix I. . -

-- c) These radiological effluent controls are described below.

2. Radiological Effluent Controls a) - The total body dose to MEMBERS OF THE PUBLIC, from liquid waste discharged to UNRESTRICTED AREAS, shall be less than 3 mrems per calendar quarter and 6 mrems per calendar year (Control 3.11.1.2).

b) The organ doses to MEMBERS OF THE PUBLIC, from liquid waste discharged to UNRESTRICTED AREAS, shall be less than 10 mrems per calendar quarter and 20 mrems per calendar year (Control 3.11.1.2).

c) The routine surveillances which are performed to verify compliance with these radiological effluent controls are described below.

3. Surveillance Requirement(s) a) Cumulative total body doses to MEMBERS OF THE PUBLIC in UNRESTRICTED AREAS--for the current calendar month, the calendar quarter, the current calendar year, and the previous 92 days-shall be calculated, in accordance with equation 8L, at least once per 31 days (per Control 4.11.1.2 and 4.11.1.3).

b) Cumulative organ doses to MEMBERS OF THE PUBLIC in UNRESTRICTED AREAS--for the current calendar month, the current calendar quarter, the current calendar year, and the previous 92 days-shall be calculated, in accordance with equation 8L, at least once per 31 days (per Control 4.11.1.2 and 4.11.1.3).

.4. Responsible Plant Organizations * - ;

a) The CHEMISTRY SECTION is responsible for performing the surveillances, listed above, whenever the appropriate initiating conditions, listed below, are present.

OFFSITE DOSE CALCULATION MANUAL PRage 976f319

5. Initiating Conditions a) Perform the surveillances, listed above, at least once per 31 days (per Control 4.11.1.2).

b) For BATCH RELEASES, perform the surveillances, listed above, prior to each BATCH RELEASE of radioactive liquid waste in accordance with approved CHEMISTRY SECTION procedures.

c) For CONTINUOUS RELEASES, perform the surveillances, listed above, at least monthly (until the CONTINUOUS RELEASE has been terminated) and, if activity is identified' in the CONTINUOUS waste stream, the surveillances may be performed more often as specified in approved CHEMISTRY SECTION procedures.

d) Whenever the correct initiating conditions are present, the calculations required by the above mentioned surveillance(s) shall be completed in accordance with the methodology listed in the next section.

6. Calculation Methodology a) The cumulative total body dose and the cumulative organ doses (for the current calendar month, current calendar quarter, previous 92 days, and current calendar year) shall be calculated as follows:

CUMULATIVE TOTAL BODY OR ORGAN, o, DOSE FROM LIQUID RELEASES, r l DTL = Z Dor Eq. 8L DTOL = the sum total of the total body or organ, o, dose for all liquid releases discharged during the applicable time interval Dor = the organ or total body dose resulting from release, r Calculate the values of Dor for each liquid release as described below.

b) At CCNPP, two methods exist for calculating Dor (i.e., the total body and organ doses resulting from any single release of radioactive liquid to an UNRESTRICTED AREA).

(1) The rigorous method shall be used IF a computer system and the appropriate software are available.

(2) The simplified method may be used IF a computer system and the appropriate software are NOT available.

(3) These methods, as well as additional supporting information, are presented in the following sections.

c) Rigorous method The LOWER LIMITS OF DETECTION shall conform to Control 4.11.1.1.1, Table 4.11-1.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Pageg 97 of 319' (1) Solution of the following equation may prove too rigorous for routine use unless a computer system and appropriate software are available.

(2) If a computer system and the appropriate software are available, the dose commitments due to each release of radioactive liquid to an UNRESTRICTED AREA shall be calculated in accordance with the following equation:

DOSE TO THE TOTAL BODY OR ORGAN, o, FROM A LIQUID RELEASE (RIGOROUS EQUATION)

D. = (VUI(60)(Fd)]} 7{(Aq.)(C1,)) Eq. 9L.l

Where, Fd = actual dilution water flow rate (gpm) prior to the point of release to UNRESTRICTED AREAS The actual dilution water flow rate is the sum of the circulating water flow rate, the salt water flow rate, and the undiluted radwaste flow rate.

The dilution water flow rate shall be calculated in accordance with equation 2L.

DILUTION WATER FLOW RATE Fd Fcw + Fs. + Fu Eq. 2L FCW = actual circulating water system flow rate The flow rate of each circulating water pump is 200,000 gallons per minute.'

= actual salt water system flow rate If the actual salt water flow rate is unknown or otherwise not readily available a salt water flow rate of 0 gpm may be used in equation 2L.

The maximum flow rate for one salt water pump is listed on Attachment 2.

Fu = actual undiluted radwaste flow rate (gpm)

If the actual undiluted radwaste flow rate is less than 1%of the total dilution water flow rate, an undiluted radwaste flow rate of 0 gpm may be used in equation 2L. '

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 2.

Vu = volume of undiluted radwaste (gallons) 4 Aju = specific activity of nuclide, I, in the undiluted liquid radwaste (microcuries per mrilliliter)

CIO = liquid 'release dose factor for nuclide, i, 'and organ, o (mremlhr pernmicrocurie/mIj)

Equation 9L has been derived from NUREG-0133, 4.3.

S--- OFFSITE DOSE CALCULATION MANUAL Rev. 7 Page 98 of 319 The liquid release dose factors for principal gamma and beta emitters are listed in Attachment 6.

The liquid release dose factors for principal gamma and beta emitters were obtained in accordance with the methodology of NUREG-0133, section 4.3.1 (for salt water sites).

60 = a constant, the number of minutes per hour (3) In the event a computer system is unavailable, a simplified equation may be used to calculate the total body and organ dose commitments due to individual liquid releases.

(4) The simplified method is presented below.

d) Simplified methods (1) Whenever a computer system is unavailable to perform the rigorous total body dose calculations described in the previous section, the total body dose commitments--due to each release of radioactive liquid to an UNRESTRICTED AREA--may be calculated in accordance with equation 10L.

(2) Whenever a computer system is unavailable to perform the rigorous organ dose calculations described in the previous section, the dose commitments--

to the maximum exposed organ, due to each release of radioactive liquid to an UNRESTRICTED AREA--may be calculated in accordance with equation 11L.

TOTAL BODY DOSE FROM A LIQUID RELEASE (SIMPLIFIED EQUATION)

Dtb = { [(58.6)(VU)]/[(Ksf)(Fd)1}) YAlu Eq.10L' Dtb = dose commitment, to the total body, due to each release of radioactive liquid to an UNRESTRICTED AREA 58.6 = a conversion constant which includes:

1) 1.33E4 mrem/hr per microcurie/mI (the total body dose factor for Cs-134)

2) 1.000 hr/60.00 min

3) 1.000 gallons/3.785 liters Vu = volume of undiluted radwaste (liters)

Ksf = a constant, actually a safety factor, which is the ratio of the CCNPP total body dose limit to the total body dose limit of Control 3.11.1.2, (unitless)

The safety factor chosen shall be less than or equal to 1.00. This ensures the total body dose is always less than or equal to the total body dose limit of Control 3.11.1.2.

1 Equations 1OL has been derived from NUREG-0 133, 4.3.

OFFSITE DOSE CALCULATION MANUAL 99 of 319

A --------------------------------------------------------------------------------------------------

safety factor of 1.00 will yield a total body dose which corresponds to the total body dose limit of Control 3.11.1.2.

A safety factor of 0.500 will yield an total body dose which corresponds to one-half the total body dose limit of Control 3.11.1.2.

It is recommended that a safety factor of 1.0 be used for calculating the total body dose, however, other values-not to exceed 1.00-may be used as directed by the General Supervisor Chemistry.

The particular value selected for the safety factor is somewhat arbitrary, however a safety factor does provide plant personnel with a degree of administrative control over the use of simplified equations for generating radioactive liquid release permits. This administrative control is designed to minimize the possibility of violating Control 3.11.1.2 when simplifying assumptions are used.

The use of a safety factor is consistent with the ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as is reasonably achievable.

This safety factor has been included in equation 1OL to account for any potential nonconservatism associated with applying the Cs-1 34 total body dose conversion factor to all radionuclides identified in the liquid release. Such nonconservatism could conceivable be present whenever radionuclides having a dose conversion factor greater than that of Cs-1 34 are present in a liquid release.

Fd = actual dilution water flow rate (gpm) prior to the point of release to UNRESTRICTED AREAS The actual dilution water flow rate is the sum of the circulating water flow rate, the salt water flow rate, and the undiluted radwaste flow rate.

The dilution water flow rate shall be calculated in accordance with equation 2L.

Rev. 7 OFFSITE DCISE CALCULATION MANUAL Page 100 of 319 DILUTION WATER FLOW RATE Fd = Fw + Fw + Fu Eq. 2L FCW = actual circulating water system flow rate The flow rate of each circulating water pump is 200,000 gallons per minute.

FSW = actual salt water system flow rate If the actual salt water flow rate is unknown or otherwise not readily available, a salt water flow rate of 0 gpm may be used in equation 2L.

The maximum flow rate for one salt water pump is listed on Attachment 2.

FU = actual undiluted radwaste flow rate (gpm)

If the actual undiluted radwaste flow rate is less than 1% of the total dilution water flow rate, an undiluted radwaste flow rate of 0 gpm may be used in equation 2L.

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 2.

AN Au = specific activity of nuclide, i, in the undiluted liquid radwaste (microcuries per milliliter)

DOSE TO THE ORGAN, o, FROM A LIQUID RELEASE (SIMPLIFIED EQUATION)

D. = { (449)(VU)]/[(K.f)(Fd)1} YAj. Eq. 1IL 1 Do = the dose commitment to the maximum exposed ORGAN due to each release of radioactive liquid waste to an UNRESTRICTED AREA 449 = a conversion constant which includes:

1) 1.02E5 mrem/hr per microcurie/ml (the dose factor for 1-131 to the thyroid)

2) 1.000 hr/60.00 min

3) 1.000 gallons/3.785 liters Vu = volume of undiluted radwaste (liters)

Ksf = a constant, actually a safety factor, which is the ratio of the CCNPP organ dose limit to the organ dose limit of Control 3.11.1.2, (unitless)

The safety factor chosen shall be less than or equal to 1.00. This ensures the organ dose is always less than or equal to the organ dose limit of Control 3.11.1.2.

Equations 11L has been derived from NUREG-0133, 4.3.

Rev. 7

- OFFSITE DOSE CALCULATION MANUAL Page 101 of 319 A safety factor of 1.00 will yield a organ dose which corresponds to the organ dose limrit of Control 3.11.1.2.

A safety factor of 0.500 will yield an organ dose which corresponds to one-half the organ dose limit of Control 3.11.1.2.

It is recommended that a safety factor of 1.0 be used for calculating the organ dose, however, other values-not to exceed 1.00--may be used as directed by the General Supervisor Chemistry.

The particular value selected for the safety factor is somewhat arbitrary, however a safety factor

'does provide plant personnel with a degree of administrative control over the use of simplified equations for generating radioactive liquid release permits. This administrative control is designed to minimize the possibility of violating Control 3.11.1.2 when simplifying assumptions are used.

The use of a safety factor is consistent with the ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as is reasonably achievable.

This' safety factor has been included in equation I IL to account for any potential nonconservatism associated with applying the 1-131 thyroid dose conversion factor to all radionuclides identified in the liquid release. Such nonconservatism could conceivable be present whenever radionuclides having a dose conversion factor greater than that of 1-131 are present In a liquid release.

Fd = dilution water flow rate (gpm) prior to the point of release to UNRESTRICTED AREAS The actual dilution water flow rate is the sum of the circulating water flow rate, the salt water flow rate, and the undiluted radwaste flow rate.

The dilution water flow rate shall be calculated in accordance with' equation 2L.

DILUTION WATER FLOW RATE Fd FCW + F.w + FU Eq. 2L Fa, = actual circulating water system flow rate' The flow rate of each circulating water pump is 200,000 gallons per minute.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 102 of 319 Fsw salt water system flow rate If the actual salt water flow rate is unknown or otherwise not readily available, a salt water flow rate of 0 gpm may be used in equation 2L.

The maximum flow rate for one salt water pump is listed on Attachment 2.

Fu = actual undiluted radwaste flow rate (gpm)

If the actual undiluted radwaste flow rate is less than 11% of the total dilution water flow rate, an undiluted radwaste flow rate of 0 gpm may be used in equation 2L.

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 2.

0 Au = specific activity of nuclide, i, in the undiluted liquid radwaste (microcuries per milliliter)

7. Corrective ACTIONS a) CHEMISTRY SECTION surveillance procedures shall contain and/or reference administrative and/or Control limits for cumulative total body doses or cumulative organ doses resulting from liquid effluents and shall specify corrective actions to be initiated when these limits are exceeded.

b) Refer to Control 3.11.1.2 for actions to be taken in the event the calculated cumulative total body dose exceeds 3 mrem per calendar quarter or 6 mrem per calendar year.

c) Refer to Control 3.11.1.2 for actions to be taken in the event the calculated cumulative organ dose-for any organ-exceeds 10 mrem per calendar quarter or 20 mrem per calendar year.

LIMITS FOR THE LIQUID WASTE PROCESSING SYSTEM

1. Introduction a) 10 CFR 50.36a requires licensees to maintain and use the equipment installed in the liquid waste processing system for the purpose of controlling effluents to the environment.

b) Radiological effluent controls have been established to implement the requirements of 10 CFR 50.36a.

c) Different documents (e.g., Technical Specifications or NUREG-0472) may reference this effluent control using a variety of synonyms. Examples of some of these synonyms include terms such as the PROJECTED DOSE, the 92-DAY PROJECTED DOSE, the 92-DAY CUMULATIVE DOSE, the DOSE IN A 92-DAY PERIOD, or the DOSE FOR THE PREVIOUS 92 DAYS. The ODCM methodology for calculating this dose - by whatever name - is based on a simple sum of the dose contributions for the previous 92 days. As result, all the terms listed above are considered as equivalent quantities for purposes of this effluent control at CCNPP. For simplicity, the ODCM generally uses the phrase udose for the previous 92 days' when referencing this effluent control.

d) These radiological effluent controls are described below.

OFFSITE DOSE CALCULATION MANUAL PRage 103of319'

2. Radiological EffluentControls a) The liquid radwaste processing system shall be used to reduce the quantity of radioactive materials in'liquid waste released to the environment whenever the total body dose'to MEMBERS OF THE PUBLIC, from liquid waste discharged to UNRESTRICTED AREAS, is greater than 0.36 mrem for the previous 92 days (per Control 3.11.1.3).

b) The liquid radwaste processing system shall be used to reduce the quantity of radioactive materials Inliquid waste released to the environment whenever the organ doses to MEMBERS OF THE PUBLIC, from liquid waste discharged to UNRESTRICTED AREAS, is greater than 1.20 mrem for the previous 92 days (per Control 3.11.1.3).

c) The routine surveillances which are performed to verify compliance with this radiological effluent controls are described below.

3. Surveillance Requirement(s) a) The previous 92-day, cumulative, total body dose--to MEMBERS OF THE PUBLIC in UNRESTRICTED AREAS-shall be calculated, as described in the ODCM, at least once per 31 days (per Control 4.11.1.3).

b) The previous 92-day, cumulative, organ doses-to MEMBERS OF THE PUBLIC in UNRESTRICTED AREAS-shall be calculated, as described in the ODCM, at least once per 31 days (per Control 4.11.1.3).

c) The CCNPP organization(s) responsible for performing these surveillances are identified in the next section.

4. Responsible Plant Organization(s) a) The CHEMISTRY SECTION is responsible for performing the surveillances, listed above, whenever the appropriate initiating conditions, listed below, are present.

5. Initiating Conditions a) Perform the surveillances, listed above, at least once per 31 days (per Control 4.11.1.3).

b) For BATCH RELEASES, perform the surveillances, listed above, prior to each BATCH RELEASE of radioactive liquid waste in accordance with approved CHEMISTRY SECTION procedures.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 104 of 319 c) For CONTINUOUS RELEASES, perform the surveillances, listed above, at least monthly (until the CONTINUOUS RELEASE has been terminated) and, if activity is identified' in the CONTINUOUS waste' stream, the surveillances may be performed more often as specified in approved CHEMISTRY SECTION procedures.

d) Whenever the correct initiating conditions a're present, the calculations required by the above mentioned surveillance(s) shall be completed in accordance with the methodology contained in the following section.

6. Calculation Methodology a) The cumulative total body dose and the cumulative organ doses for the previous 92 days shall be calculated in accordance with equation 8L found in the section, "Limits On Cumulative Total Body Doses And Cumulative Organ Doses For Liquid Effluents", found elsewhere in the ODCM.

7. Corrective ACTIONS a) CHEMISTRY SECTION surveillance procedures shall contain/and or reference administrative and/or Control limits for cumulative total body doses or cumulative organ doses resulting from liquid effluents and shall specify corrective ACTIONS to be initiated when these limits are exceeded.

b) Refer to Control 3.11.1.3 for ACTIONS to be taken in the event the calculated cumulative total body dose exceeds 0.36 mrem for the previous 92 days.

c) Refer to Control 3.11.1.3 for ACTIONS to be taken in the event the calculated cumulative organ dose-for any organ-exceeds 1.2'mrem for the previous 92 days.

I The LOWER LIMITS OF DETECTION shall conform to Control 4.11.1.1.1, Table 4.1 1-1.

OFFSITE DOSE CALCULATION MANUAL PRage 105 of 319

- RADIOACTIVE GASEOUS EFFLUENTS RELEASE PATHWAYS rB5271

1. Introduction a) Radioactive gaseous waste generated from operation of CCNPP may be released to the atmosphere.

b) By design (i.e., in the absence of primary-to-secondary leaks), there are 2 pathways by which waste gas from the site may be discharged to the atmosphere. These pathways are listed below. General information related to each of these potential release pathways is contained on Attachment 7.

(1) Unit- main vent stack (2) Unit 2 main vent stack c) Depending on plant conditions, (e.g., primary-to-secondary leaks) a potential exists for the release of radioactive materials from other pathways. Examples of these pathways are listed below. General information related to each of these potential release pathways is contained on Attachment 8.

(1) auxiliary boiler deaerator (2) steam generator atmospheric steam dump system (3) plant nitrogen system (4) turbine building ventilation exhaust (5) emergency air lock (6) plant compressed air (7) main steam line penetrations (8) containment equipment hatch (9) auxiliary feed pumps (10) gland steam exhaust condenser d) All of these pathways are described below.-

2. Unit 1 Main Vent Stack a) Dilution air and radioactive gaseous waste are discharged to the atmosphere through the Unit I main vent stack. '

b) The radioactive gaseous waste is mixed with and diluted by the outside air and building exhausts prior to exiting the Unit 1 main vent stack.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 106 of 319 c) The Unit 1 main vent stack is secured to the Unit 1 reactor containment building.

d) The top of the Unit 1 main vent stack is at elevation 203.5 feet (mean sea level, MSL), and as such is 10.1 feet above the top of the reactor containment building dome. As a result, the Unit 1 main vent stack does not qualify as a "free-standing" stack greater than 80 meters tall'.

e) The Unit 1 main vent stack is designed to accept gaseous radioactive waste from various sources. Sources which may contribute radioactive material to the Unit 1 main vent stack are tabulated in Attachment 7.

3. Unit 2 Main Vent Stack a) Dilution air and radioactive gaseous waste are discharged to the atmosphere through the Unit 2 main vent stack.

b) The Unit 2 main vent stack is designed to accept radioactive gaseous waste from various sources.

c) The radioactive gaseous waste is mixed with and diluted by the outside air and building exhausts prior to exiting the Unit 2 main vent stack.

d) The Unit 2 main vent stack is secured to the Unit 2 reactor containment building.

e) The top of the Unit 2 main vent stack is at elevation 203.5 feet (MSL), and as such is 10.1 feet above the top of the reactor containment building dome. As a result, the Unit 2 main vent stack does not qualify as a "free-standing" stack greater than 80 meters tall.'

f) The Unit 2 main vent stack is designed to accept gaseous radioactive waste from various sources. Sources which may contribute radioactive material to the Unit 2 main vent stack are tabulated in Attachment 7.

4. Auxiliary Boiler Deaerator a) Radioactive gases may be vented from the auxiliary boiler deaerator during periods of primary to secondary leakage.

b) Steam from the Moisture Separator Reheater (MSR) may be used in the deaerator.

In the event of a primary to secondary leak, the MSR steam could become contaminated. Therefore, a potential exists for the release of radioactive gases in steam discharged from the auxiliary boiler deaerator.

c) The discharge of steam is accomplished via a relief vent, 0-VBV-1891, which allows excess pressure to be vented to atmosphere.

d) In the event the auxiliary boiler deaerator were to become contaminated, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known:

(1) the MSR steam activity obtained from a sample, I As defined by Regulatory Guide 1.111

-j ~_ Rev. 7 ByQV'~ OFFSITE DOSE CALCULATION MANUAL page 107of319

_~~~~~~~~~~Pg 107 of _319 _______ _____

(2) the duration of the discharge, (3) the estimated steam discharge flow rate, and (4) the measured or average annual meteorological conditions.

e) In accordance with applicable safety evaluations', continued operation of this system is allowed as long as the concentration of radionuclides in the auxiliary boiler steam drum is less than 96 MPCs.

5. Steam Generator Atmospheric Steam Dump System a) Radioactive gases are not normally vented from this pathway.

b) Radioactive gases may be vented from the steam generator atmospheric steam dump system during periods of primary to secondary leakage.

c) If a primary to secondary leak is present and the steam dump valves are opened, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known (per UFSAR, 10.1.2.2): - !. *

(1) the specific activity of a main steam sample as determined by GAMMA ISOTOPIC ANALYSIS, (2) the duration of discharge, (3) the estimated steam discharge flow rate, and (4) the measured or average annual meteorological conditions.

d) The total capacity of the atmospheric steam dump valve is 5 percent of steam flow with the reactor at full power (per UFSAR, 10.1.2.2).

6. Plant Nitrogen System a) Radioactive gases are not normally vented from this pathway.

b) Nitrogen is supplied to various components which contain radioactive materials (e.g.,

VCT).

c) In the event the plant nitrogen system were to become contaminated, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known:

(1) the specific activity of the gas in the plant nitrogen system as determined by GAMMA ISOTOPIC ANALYSIS, (2) the pressure of the nitrogen system, (3) the volume of the nitrogen system, and See 50.59 Log No. 90-0-027-037-RI.

OFFSITE DOSE OFFITERev.

CALCULATION MANUAL Page 7108 of 319 (4) the measured or average annual meteorological conditions.

d) It should be noted that the amount of radioactivity released could be estimated based on knowledge of other related parameters.

e) In accordance with applicable safety evaluations', continued operation of this system is allowed as long as the concentration of radionuclides is less than 13,400 MPCs.

7. Turbine Building Exhaust a) Radioactive gases are not normally vented from this pathway.

b) In the event radioactive gases were to be released through the turbine building exhaust, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known:

(1) the specific activity of the turbine building air, (2) the duration of the discharge, (3) the estimated flow rate during the discharge, and (4) the measured or average annual meteorological conditions.

8. Emergency Air Lock a) Radioactive gases are not normally vented from this pathway.

b) Inthe event radioactive gases were to be released through the emergency air lock, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known:

(1) the containment air activity obtained from a sample, (2) the volume of the air lock (9.558 cubic meters),

(3) the measured or average annual meteorological conditions.

9. Plant Compressed Air a) Radioactive gases are not normally vented from this pathway.

b) In the event the plant compressed air system were to become contaminated, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known:

(1) the specific activity of the compressed air system, (2) the pressure of the compressed air system, (3) the volume of the compressed air system, and See 50.59 Log No. 90-0-074-01 1-R1.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 109 of 319 (4) -- the measured or average annual meteorological conditions. --

10. Main Steam Line Penetrations a) Radioactive gases are not normally vented from this pathway.

b) This penetration is cooled by outside air.:

c) Gases may be released to the atmosphere through safety vents to the roof at elevation 91.5 feet.

d) See UFSAR 9.8.2.3.

11. Steam Driven Auxiliary Feed Pumps a) Radioactive gases are not normally vented from this pathway.

b) In the event radioactive gases were to be released through the auxiliary feed pumps, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known:

(1) the activity in the steam, (2) the volume of steam released.

12. Containment Equipment Hatch a) Radioactive gases are not normally vented from this pathway.

b) In the event radioactive gases were to be released through the containment equipment hatch, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known:

(1) the containment air activity obtained from a'sample, (2) the volume of the air released, (3) the measured or average annual meteorological conditions.

13. Gland Steam Exhaust Condenser a) Radioactive gases are not normally vented from this pathway.

b) In the event radioactive gases were to be released through the gland steam exhaust condenser, the amount of radioactivity released and the resulting doses/dose rates at the SITE BOUNDARY can be estimated if the following parameters are known:

(1) activity in the gland steam exhaust condenser, (2) the flow rate through the gland steam exhaust condenser, (3)' the duration of the leak, and - i (4) the measured or average annual meteorological conditions.

Rev. 7

->OFFSITE DOSE CALCULATION MANUAL Page 110of319

14. Other unmonitored release paths should be evaluated and added to the 00CM as necessary.

TYPES OF GASEOUS RELEASES

1. All gaseous radwaste releases are classified as either BATCH RELEASES or CONTINUOUS RELEASES.

2. The definition of BATCH RELEASE is included in the definitions section of the ODCM.

3. The definition of CONTINUOUS RELEASE is included in the definitions section of the ODCM.

4. Gaseous radwaste discharges have been classified as CONTINUOUS or BATCH as shown on Attachments 7 and 8.

PROCESSING EQUIPMENT

1. Simplified Flow Diagram a) An overview of the gaseous waste processing system, including major equipment and (normal) flow paths, is outlined on Attachment 9.

2. Modifications a) Licensed initiated major changes to the gaseous waste processing system shall be reported to the Commission in the Radioactive Effluent Release Report for the period S in which the modification to the waste system was completed (per Technical Specification 5.6.3). The discussion of each change shall contain:

(1) A description of the equipment, components and processes involved; and (2) Documentation of the fact that the change, including the safety analysis, was reviewed and found acceptable by the onsite review function.

b) A "major" change or modification includes, but is not limited to, the removal or permanent bypass of any of the following:

(1) waste gas decay tank (2) waste gas surge tank (3) degassifier (4) HEPA filter (5) charcoal filter

3. Detailed Description a) A detailed description of the gaseous waste processing system is beyond the scope of the ODCM.

l

OFFSITE DOSE CALCULATION MANUAL Page 171of 319 b) For more information on the Waste Gas System,see the CCNPP Syster Description Number 14A, 'Waste Gas System."

c) For more information on the Waste Gas System, see the CCNPP Updated Final Safety Analysis Report, Chapter 11, 'Waste Processing And Radiation Protection."

GASEOUS EFFLUENT RADIATION MONITORS AND SETPOINTS

1. Wide Range Gas Monitor (1-RE-5416) a) General description (1) The Wide Range Gas Monitor (WRGM) contains 3 radiation elements (a) low-range noble gas detector i)' Designation of radiation element: 1-RE-5416 ii) type of radiation element: Off-line scintillation iii) output: digital iv) Radiation indicator: 1-RIC-5415 v) units for radiation indicator are user programmable and are normally set to microcuries per cubic centimeter or microcuries per second vi) supplier: Sorrento Electronics (formerly General Atomics)

(b) mid-range, noble gas detector i) Designation of radiation element: 1-RE-5417 ii) type of radiation element: Solid state iii) This noble gas monitor is used to measure the release of radioactivity from unit 1 main vent in the event of an accident.

(UFSAR, 11 .2.3.2.12) iv) setpoints for the mid-range detector will not be addressed in the ODCM (c) high-range, noble gas detector i) Designation of radiation element: 1-RE-5418 ii) type of radiation element: Solid state iii This detector is used to measure the release of radioactivity from unit 1 main vent in the event of an accident. (UFSAR, 11.2.3.2.12)

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page II2of319 iv) setpoints for the high-range detector will not be addressed in the ODCM (2) The low range detector will be the only detector addressed further in the ODCM.

b) Functions of 1-RE-5416 (1) continuously measure the release rate of noble gases emanating from the Unit 1 main vent stack (Control 4.11.2.1.1 or 4.11.2.1.2, Table 4.11-2)

(2) continuously indicate (via 1-RIC-5415) the release rate of noble gases emanating from the Unit I main vent stack (Control 4.11.2.1.1 or Control 4.11.2.1.2, Table 4.11-2)

(3) alarm (via 1-RIC-5415) prior to exceeding the site-boundary, noble-gas, total-body-dose-rate limit of 500 mr/yr (per Control 3.11.2.1.a)

(4) alarm (via 1-RIC-5415) prior to exceeding the site-boundary, noble-gas, skin-dose-rate limit of 3000 mr/yr (per Control 3.11.2.1.a)

C) OPERABILITY of 1-RE-5416 (1) This monitor shall be operable (or have OPERABILITY) when it is capable of performing its specified function(s).

(2) The functions of this monitor are listed in section (b) above.

d) Monitors equivalent to 1-RE-5416 (1) 1-RE-5415 [the 'Westinghouse Plant Vent Stack Monitor"] has the capability of providing the measurement and alarm functions of 1-RE-5416 during times when 1-RE-5416 is declared inoperable.

(2) 1-RE-5415 provides redundant monitoring [for 1-RE-5416] at the low end of the concentration ranges (UFSAR 11.2.3.2.12).

(3) In the event 1-RE-5415 is inoperable or otherwise unavailable, 1-RE-5416 may fulfill the measuring, indicating, and alarming functions normally provided by 1-RE-5415.

e) Radiological effluent controls for 1-RE-5416 (1) Control 3.3.3.9 states that releases via the plant vent stack may continue if any one of the following three conditions are satisfied:

(a) 1-RE-5415 is operable AND the alarm setpoint for 1-RI-5415 is set to ensure the annual dose rates due to noble gases at the SITE BOUNDARY are less than 500 mr/yr to the total body and are less than 3000 mr/yr to the skin (per Control 3.11.2.1.a), or

OFFSITE DOSE CALCULATION MANUAL ~RAMIIA J Page 7113 of 319 Rev.

(b) an "equivalent monitor" is operable AND the alarm trip setpoint for the "equivalent monitor" is set to ensure annual dose rates due to noble gases at the SITE BOUNDARY are less than 500 mr/yr to the total body and are less than 3000 mr/yr to the skin (per Control 3.11.2.1.a),

or (c) grab samples are obtained and analyzed for gross activity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with Controls 3.11.2.1.a, 4.11.2.1.1, and 4.11.2.1.2 (per Control 4.3.3.9, Table 3.3-12, ACTION 37).

f) Surveillances for 1-RE-5416 (1) Control 4.3.3.9 requires demonstrating the OPERABILITY of 1-RE-5416 by satisfying the checks, calibrations, and tests listed below:

(a) CHANNEL CHECK within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (b) SOURCE CHECK within the past 31 days (c) CHANNEL CALIBRATION within the past 18 months (d) CHANNEL FUNCTIONAL TEST within the past 6 six months g) Setpoints for 1-RIC-5415 (1) Requirements and commitments (a) The alarm and fixed setpoints ... shall be determined and adjusted in accordance with the methodology and parameters of the ODCM.

(Control 3.3.3.9)

(b) The method for calculating fixed or adjustable setpoints shall be provided in the ODCM (per NUREG-0133, 5.1.1).

(2) There are four radiation alarm setpoints associated with, or otherwise related to, the WRGM.

(a) 1-RIC-5415 fixed high-high radiation alarm setpoint (b) 1-RIC-5415 fixed high radiation alarm setpoint (c) 1-RIC-5415 adjustable plant computer high radiation alarm setpoint (d) 1-RIC-5415 adjustable plant computer alert setpoint.

(3) In order to simplify the setpoint terminology, eliminate ambiguity, and minimize the possibility of misinterpretation, the ODCM will refer to these setpoints as follows (a) The 1-RIC-5415 fixed high-high radiation alarm setpoint will be referred to as the fixed high alarm setpoint (b) The 1-RIC-5415 fixed high radiation alarm setpoint will be referred to as the fixed alert setpoint

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 114 of 319 (c) The 1-RIC-5415 adjustable plant computer high radiation alarm setpoint will be referred to as the adjustable setpoint (d) The 1-RIC-5415 adjustable plant computer alert setpoint will be referred to as the alert setpoint.

(4) Each of these alarm setpoints are described below.

h) Fixed high alarm setpoint for 1-RIC-5415 (1) General information (a) The fixed high alarm setpoint is not adjusted for each release.

(b) Whenever the fixed high alarm setpoint is exceeded, an alarm will be generated.

(c) The current value for the fixed high alarm setpoint is specified in the CCNPP Alarm Manual.

(d) The CCNPP Alarm Manual' refers to the fixed high alarm setpoint as the Unit 1 Wide Range Noble Gas Radiation Monitor high alarm setpoint.

(e) The fixed high alarm setpoint is integral to the WRGM, as purchased from the supplier.

(f) The fixed high alarm setpoint is administratively controlled by EN 100.

(g) The fixed high alarm setpoint shall be calculated as described below.

(2) Calculating the fixed high alarm setpoint for 1-RIC-5415 (a) The fixed high alarm setpoint for 1-RIC-5415 (WRGM) shall be calculated in accordance with equation 1G. 2 FIXED HIGH ALARM SETPOINT FOR 1-RIC-5415 ISfihh 54 [c'/(x/Q)] [Fd.I(Fdi +Fd 2 )J [ATL. Eq.1IG 3

Where, Sfixhh = the fixed high alarm setpoint for 1-RIC-5415 (microcuries per second) c' = a conversion constant (1E6 cubic centimeters per cubic meter)

I The CCNPP Alarm Manual is controlled by NO-1-201.

2 The alarm and trip setpoints ... shall be determined and adjusted in accordance with the methodology and parameters of the ODCM. (Control 3.3.3.9).

Equation IG has been derived from NUREG-0133, 5.2.1.

Rev. 7 i? OFFSITE DOSE CALCULATION MANUAL Page 115of319

- x/Q = the highest calculated annual average relative concentration for any area at or beyond the UNRESTRICTED AREA boundary (2.2E-6 seconds per cubic meter)

Unit 1 main vent releases are considered "long-term", releases', and as such, the highest historical annual average dispersion factor, (x/Q), is used in the setpoint calculations.

The highest annual average dispersion factor (x/Q) is 2.2E-6 (UFSAR, 2.3.6.3) for purposes of routine, long-term concentrations (e.g., routine noble gas releases)

The maximum annual average on-shore concentrations occur in the southeast sector at a distance of 1300 meters for purposes of routine, long-term concentrations (e.g., routine noble gas releases) (UFSAR, 2.3.6.3)

FdX = the estimated main vent stack (diluted gaseous radwaste) flow rate for'unit, x,'(cubic meters per second)

The estimated main vent stack flow rates for Unit I and Unit 2 are defined below.

Fdl =the estimated main vent stack flow rate for Unit i (cubic'meters per second)

Since the main vent stack flow rate will vary depending on the configuration of air dampers and the input gas streams, nominal main vent stack flow rate is used to calculate the fixed high alarm setpoint.

Use the nominal Unit 1 main vent stack flow rate listed on Attachment 7.

The main vent stack flow rate shall be determined, in accordance with approved procedures, at least once per 6'months (+ 25%). The Test and Equipment Unit shall be responsible for performing this test. The results of the main vent flow rate test shall be evaluated to ensure the main vent flow rates used in the ODCM are an accurate reflection of the true main vent flow rates. The Radiological Effluent Technical Specifications (RETS) Program Manager is responsible for modifying the (main vent flow rates used in the) ODCM in the event the main vent flow rate for either Unit I or Unit 2 has increased to a value which is greater than the maximum discharge flow rates listed on Attachment 7.

Fd2 = the estimated main vent stack (diluted gaseous radwaste) flow rate for unit 2 (cubic meters per second)

Since the main vent stack flow rate will vary depending on the configuration of air dampers and the input gas streams, nominal main vent stack flow rate is used to calculate the fixed high alarm setpoint.

Use the nominial Unit 2 main vent stack flow rate listed on Attachment 7.

The main vent stack flow rate shall be determined, in accordance with approved procedures, at least once per 6 months (+/- 25%). The Test and Equipment Unit shall be responsible for performing this test. The results of the main vent flow rate test shall be evaluated to ensure the main vent flow rates used in the ODCM are an accurate reflection of the true main vent flow rates. The RETS Program Manager is responsible for modifying the (main vent flow rates used in the) ODCM in the event the main vent flow rate for either Unit 1 or Unit 2 has increased to a value which is greater than the maximum discharge flow rates listed on Attachment 7.

NUREG-0133, 3.3

OFFSITE DOSE CALCULATION MANUAL PRage 1716of319 ATLn '= the sum of the total specific activities of all radionuclides found in TYPICAL GASEOUS RADWASTE RELEASES (microcuries/cm 3 )

Calculate A+, in accordance with equation 2G.

SPECIFIC ACTIVITY CORRESPONDING TO THE SITE BOUNDARY LIMIT Z[(fi)(ATLn)jAiU < LMPC Eq. 2G1 l

Where, fi = a fraction which represents the relative activity contribution of noble gas radionuclide i to the total noble gas activity for TYPICAL GASEOUS EFFLUENTS (unitless)

This value may be obtained using the guidance provided on Attachment 5.

AiLt = the specific activity limit for radionuclide, i, as obtained from 10 CFR 20, Appendix B, Table II, Column 1 (microcuries/cm 3 )

For all the radionuclides found in TYPICAL RADWASTE EFFLUENTS, use the value from 10 CFR 20, Appendix B, Table II, Column 1. An acceptable alternative is to assume an isotopic mix which results in a more conservative setpoint.

LMPc = the site MPC limit (MPCs) for UNRESTRICTED AREAS The value chosen for LMPc in this equation is 2. The basis for this limit is 10 CFR 50.72.

It has been shown2 that, for the radionuclides present in TYPICAL GASEOUS EFFLUENTS from CCNPP, the 2 MPC limit is more restrictive than the limits of Control 3.11.2.1 (a).

It should be noted that by using "2" as the MPC limit (10 CFR 50.72), instead of using the limits of Control 3.11.2.1(a), a safety factor has been incorporated into equation 2G.

An alarm setpoint corresponding to 2 MPCs serves to initiate a determination of whether the "4-hour NRC notification" (specified in 10 CFR 50.72) is required.

(3) Documenting the fixed high alarm setpoint (a) Whenever the fixed high alarm setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with EN-1 -100.

This equation has been derived from 10 CFR 20, Appendix B, Table II, Note 1.

2 Addendum To Setpoint Calculations For WRGM Monitors 1-RIC-5415 and 2-RIC-5415, R.L. Conatser, December 10, 1991.

Rev. 7 g OFFSITE DOSE CALCULATION MANUAL Page 117of319

-7(4) Changing the fixed high alarm setpoint for 1-RIC-5415 - -

(a) If the fixed high alarm setpoint calculated in accordance with equation 1G exceeds the maximum range of the monitor, the fixed high setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(b) The fixed high alarm setpoint may be established at values lower than the maximum allowable setpoint, if desired.

(c) A setpoint change should be initiated whenever any of the parameters identified in the setpoint calculation equations (identified in this section of the ODCM) have changed.

(d) The fixed high alarm setpoint should not be changed unless one of the following occurs:

i) the relative activity' of any radionuclide in TYPICAL GASEOUS EFFLUENTS has'changed by greater than 10%,

and the new radionuclide mixture yields a fixed setpoint which is 10% (or more) lower than the current fixed setpoint, ii) the historical maximum annual average atmospheric dispersion factor has changed, iii) the MPC limit at the SITE BOUNDARY, (LMPc) has changed, iv) the Unit I or Unit 2 main vent stack flow rate has changed by greater than or equal to 10%2, v) the values listed in 10 CFR 20, Table II, column 1 have changed, vi) the radiation monitor has been recently calibrated, repaired, or otherwise'altered, or vii) the monitor is not conservative in its function (see "Functions of 1-RE-5416" earlier in this section).

(e) EN-1-100 contains the administrative controls associated with changing and approving fixed high alarm setpoint.

As determined in accordance with Attachment 5.'

2 As determined by analysis of the TE-001 and ETP-87-16 test results.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 118of319 i) Fixed alert setpoint for 1-RIC-5415 (1) General information (a) The fixed alert setpoint is not adjusted for each release.

(b) Whenever the fixed alert setpoint is exceeded, an alarm will be generated.

(c) The CCNPP Alarm Manual does not reference this setpoint.

(d) The fixed alert setpoint is integral to the WRGM, as purchased from the supplier.

(e) The current value for the fixed alert setpoint is specified in the CCNPP Setpoint File.

(f) The fixed alert setpoint is administratively controlled by EN-1 -100.

(g) The fixed alert setpoint shall be calculated as described below'.

(2) Calculating the fixed alert setpoint for 1-RIC-5415 (a) The fixed alert setpoint for 1-RIC-5415 shall be calculated as described below:

FIXED ALERT SETPOINT FOR 1-RIC-5415 I SnXh 5 K,, [ Snxhh I Eq. 3G I Where, Srixh = the fixed alert setpoint for I-RIC-5415 (microcuries per second)

Srihh = the fixed high alarm setpoint for 1-RIC-5415 (microcuries per second)

Ksf = a constant, actually a safety factor, which is the fraction of the fixed high setpoint (unitless).

The safety factor chosen shall be less than or equal to 1.00. This ensures the fixed alert setpoint is always less than or equal to the fixed high alarm setpoint.

A safety factor of 1.00 will yield a fixed alert setpoint which corresponds to the fixed high alarm setpoint.

A safety factor of 0.100 will yield a fixed alert setpoint which corresponds to one-tenth the fixed high alarm setpoint.

The alarm and trip setpoints ...shall be determined and adjusted inaccordance with the methodology and parameters of the ODCM. (Control 3.3.3.9).

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 119 of 319 It is recommended that a safety factor of 0.1 be used for calculating the fixed alert setpoint,t however, other values-not to exceed 1.00-may be used as directed by the General Supervisor Chemistry.

The particular value selected for the safety factor is somewhat arbitrary, however a value less than 1.00 does provide plant personnel with'adequate time to respond to changing plant conditions and to initiate corrective ACTIONS so as to minimize the possibility of violating either the 10 CFR 50.72 limit or the Control 3.3.3.9 limits.

The use of the safety factor is consistent with ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as is reasonably achievable.

The use of a "safety margin" is in accordance with the provisions of NUREG-0133 which states that "... the alarm and trip setpoints ... should correspond to a value(s) which represents a safe margin of assurance that the instantaneous gaseous release limit of Control 3.11.2.1 (a)will not be exceeded." (per NUREG-0133, 5.1.1).

This safety margin will prevent minor fluctuations in the nominal plant vent stack flow rates, errors in monitor efficiencies, and other statistical aberrations from adversely impacting the calculated fixed alert setpoint.

(3) Documenting the fixed alert setpoint (a) Whenever the fixed alert setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with EN-1-100.

(4) Changing the fixed alert setpoint for 1-RIC-5415 (a) A setpoint change should be initiated whenever any of the parameters identified in equation 3G have changed.

(b) The fixed alert setpoint should be changed whenever the fixed high setpoint is changed.

(c) The fixed alert setpoint should be changed if the value of the safety factor is changed.

(d) See EN-1 -100 for a description of activities associated with setpoint changes and setpoint approvals.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 120 of 319 j) Adjustable alarm setpoint for 1-RIC-5415 (1) General information (a) Whenever this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9), the adjustable setpoint is calculated and adjusted prior to each release of a WGDT, each containment vent, and each containment purge discharged via the main vent.

(b) The adjustable setpoint is based on the specific activities of the radionuclides present in either the WGDT or the containment building, whichever is applicable. (The radionuclide concentrations are determined by radiochemical analysis in accordance with applicable CHEMISTRY SECTION procedures as required by Control 4.11.2.1.2).

(c) Whenever the adjustable setpoint is exceeded, the WGDT, PURGE, or vent discharge via the main vent will be manually suspended.

(d) Refer to the Alarm Manual for a full list of operator ACTIONS taken in response to this alarm.

(e) The adjustable setpoint corresponds to the maximum concentration of radionuclides anticipated or expected when discharging a WGDT, a containment vent, or a containment purge via the main vent. For containment purges during outages, system evolutions may cause containment atmosphere activity to increase above what is normally expected for short periods of time.

(f) The value for the adjustable setpoint is recorded on the gaseous release permit in accordance with applicable CHEMISTRY SECTION procedures.

(g) This alarm is not integral to the main vent radiation monitor, as purchased from the supplier.

(h) This alarm is generated by the plant computer which monitors output from 1/2-RIC-5415, and provides an alarm to plant operators when the 1/2-RIC-5415 adjustable setpoint has been exceeded.

(i) When this monitor is satisfying the minimum channels operable requirement (per Control 3.3:3.9), a value for the adjustable alarm setpoint shall be calculated prior to each release of a WGDT, each containment vent, and each containment purge as shown below.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 121of 319 (2)- Calculating the adjustable setpoint for 1/2-RIC-5415 -7 (a) The adjustable setpoint is based on the specific activity of the radionuclides in the undiluted gaseous waste (as determined by radiochemical analysis per Control 4.11.2.1.2), and the alarm setpoint is calculated as shown below.

ADJUSTABLE SETPOINT FOR 1/2-RIC-5415 Sdj < (KS)(Fdx)(c'){[(F.IFd.) (AlU)(eI)]+Bkg} Eq.29G1 Sadj = the adjustable setpoint for 1/2-RIC-5415 (microcuries per second)

Ksf = a constant, actually a safety factor, which allows for fluctuation in radiation monitor response (unitless)

This safety factor helps ensure the release is not unnecessarily terminated due to (1)electronic anomalies which cause spurious monitor responses, (2) statistical fluctuations in disintegration rates, (3) statistical fluctuations in detector efficiencies, (4) errors associated with sample analysis, (5)errors associated with monitor calibrations2 , and (6)anticipated short term variations in containment activity (applicable to containment purges only).

It is recommended that a safety factor of 10 for containment purge releases be used for calculating the adjustable setpoint. However, other values for purge releases - not to exceed 10 -- may be used as directed by the General Supervisor Chemistry. A safety factor of 1.5 shall be used for all other gaseous releases.

The particular value selected for the safety factor is somewhat arbitrary, however a value less than or equal to 10 does provide plant personnel with adequate time to respond to changing plant conditions and to initiate corrective ACTIONS so as to minimize the possibility of violating either the 10 CFR 50.72 limit or the Control 3.3.3.9 limits.

The use of the safety factor is consistent with ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as Is reasonably achievable.

The use of a "safety margin" is in accordance with the provisions of NUREG-0133 which states that ". . . the alarm and trip setpoints ... should correspond to a value(s) which represents a safe margin of assurance that the instantaneous gaseous release limit of Control 3.11.2.1 (a)will not be exceeded."

(per NUREG-0133, 5.1.1).

Equation 29G has been derived from NUREG-0133, Addendum, page AA-1.

2 The 'analysis errors" and "calibration errors" refer to errors which are within established quality assurance and quality control limits.

OFFSITE DOSE CALCULATION MANUAL Rev.

Page 7 122 of 319 This safety~margin will prevent minor fluctuations in the nominal plant vent stack flow rates, errors in monitor efficiencies, and other statistical aberrations from adversely impacting the calculated adjustable setpoint. Additionally for a special case of containment purges during outages, the safety, factor,allows for short term variations in activity created as a result of system evolutions in containment.

Fu = maximum undiluted radwaste flow rate (cubic meters per second)

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 7.

FdX = the estimated main vent stack (diluted gaseous radwaste) flow rate for unit x (cubic meters per second)

Since the main vent stack flow rate will vary depending on the reactor unit, the configuration of air dampers, and the input gas streams, nominal main vent stack flow rate is used to calculate the adjustable setpoint.

Use the nominal main vent stack flow rate, for thei appropriate unit, listed on Attachment 7.

The main vent stack flow rate shall be determined, in accordance with approved procedures, at least once per 6 months (+/- 25%). The Test and Equipment Unit shall be responsible for performing this test. The results of the main vent flow rate test shall be evaluated to ensure the main vent flow rates used in the 0DCM are an accurate reflection of the true main vent flow rates. The RETS Program Manager is responsible for modifying the (main vent flow rates used in the) 00CM in the event the main vent flow rate for either Unit I or Unit 2 has increased to a value which is greater than the maximum discharge flow rates listed on Attachment 7.

A = specific activity of radionuclide, i, in the undiluted waste stream, either the WGDT or the containment building (microcuries per cubic centimeter) el = absolute detector efficiency for nuclide, i (microcuries Xe-I 33 equivalent per microcuries nuclide i)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

Bkg =an approximation of the detector background (microcuries per cubic centimeter)

C, c= a conversion constant (IE6 cubic centimeters per cubic meter)

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 123 of 319 (3) Documenting the adjustable setpoint for 1/2-RIC-5415 (a) Whenever the adjustable setpoint is calculated, the specific values chosen for each-of the variables shall be documented in accordance with approved CHEMISTRY SECTION procedures (e.g., CP-604).

(4) Changing the adjustable setpoint for 1/2-RIC-5415 (a) In all cases, the adjustable setpoint shall be set to a value which is less than or equal to the fixed setpoint.

(b) If the adjustable setpoint exceeds the maximum range of the monitor, the setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(c) CHEMISTRY SECTION procedures (e.g., CP-604) contain administrative controls associated with calculating and approving an adjustable setpoint.

(d) Whenever this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9 ) the calculated value for the adjustable setpoint shall be entered into the plant computer prior to each release of a WGDT, a containment vent, or a containment purge via the main vent.

k) Alert setpoint for 1-RIC-5415 (1) General information (a) The alert setpoint is applicable to containment purges only.

(b) Whenever this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9 ), the alert setpoint is calculated and adjusted prior to each containment purge discharged via the main vent.

(c) The alert setpoint is based on the specific activities of the radionuclides present in the containment building. (The radionuclide concentrations are determined by radiochemical analysis in accordance with applicable CHEMISTRY SECTION procedures as required by Control 4.11.2.1.2).

(d) Whenever the alert setpoint is exceeded the PURGE via the main vent may continue.

OFFSITE DOSE CALCULATION MANUAL Rev.

Page 7124 of 319 (e) The alert setpoint corresponds to a level of activity which indicates additional source term(s) may be present, and as a result, additional notifications and/or actions are required to identify the source and to accurately account for the activity discharged.

(f) The value for the alert setpoint is recorded on the gaseous release permit in accordance with applicable CHEMISTRY SECTION procedures.

(g) This alarm is not integral to the main vent radiation monitor, as purchased from the supplier.

(h) This alarm is generated by the plant computer which monitors output from 1/2-RIC-5415, and provides an alarm to plant operators when the 1/2-RIC-5415 alert setpoint has been exceeded.

(i) When this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9), a value for the alert setpoint shall be calculated prior to each containment purge via the main vent as shown below.

(2) Calculating the alert setpoint for 1/2-RIC-5415 (a) The alert setpoint is based on the specific activity of the radionuclides in the undiluted gaseous waste (as determined by radiochemical analysis per Control 4.11.2.1.2), and the setpoint is calculated as shown below.

ALERT SETPOINT FOR 1/2-RIC-5415 I S.10't < (1.50)(Fd.)(r.'){[(F./Fd.)Y;(Al.)(eq)]+Bkg) Eq.29G' Salert = the alert setpoint for 1/2-RIC-5415 (microcuries per second) 1.50 = a constant, actually a safety factor, which allows for fluctuation in radiation monitor response (unitless)

This safety factor helps ensure the release is not unnecessarily terminated due to (1) electronic anomalies which cause spurious monitor responses, (2) statistical fluctuations in disintegration rates, (3) statistical fluctuations in detector efficiencies, (4) errors associated with sample analysis, and (5) errors associated with monitor calibrations2 .

Equation 29G has been derived from NUREG-0133, Addendum, page AA-1.

2 The "analysis errors" and "calibration errors" refer to errors which are within established quality assurance and quality control limits.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 125 of 319

-- The use of the safety factor is consistent with ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as is reasonably achievable.

The use of a "safety margin" is in accordance with the provisions of NUREG-0133 which states that "... the alarm and trip setpoints ... should correspond to a value(s) which represents a safe margin of assurance that the instantaneous gaseous release limit of Control 3.11.2.1 (a)will not be exceeded." (per NUREG-0133, 5.1.1).

This safety margin will prevent minor fluctuations in the nominal plant vent stack flow rates, errors in monitor efficiencies, and other statistical aberrations from adversely impacting the calculated alert setpoint.

Fu = maximum undiluted radwaste flow rate (cubic meters per second)

Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 7.

FdX = the estimated main vent stack (diluted gaseous radwaste) flow rate for unit x (cubic meters per second)

Since the main vent stack flow rate will vary depending on the reactor unit, the configuration of air dampers, and the input gas streams, nominal main vent stack flow rate is used to calculate the alert setpoint.

Use the nominal main vent stack flow rate, for the appropriate unit, listed on Attachment 7.

The main vent stack flow rate shall be determined, in accordance with approved procedures, at least once per 6 months La 25%). The Test and Equipment Unit shall be responsible for performing this test. The results of the main vent flow rate test shall be evaluated to ensure the main vent flow rates used in the ODCM are an accurate reflection of the true main vent flow rates. The RETS Program Manager is responsible for modifying the (main vent flow rates used in the) ODCM in the event the main vent flow rate for either Unit I or Unit 2 has increased to a value which is greater than the maximum discharge flow rates listed on Attachment 7.

Aju = specific activity of radionuclide, i, in the containment building (microcuries per cubic centimeter) el = absolute detector efficiency for nuclide, i (microcuries Xe-133 equivalent per microcuries nuclide i)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

Bkg = an approximation of the detector background (microcuries per cubic centimeter) c' = a conversion constant (lE6 cubic centimeters per cubic meter)

(3) Documenting the alert setpoint for 1/2-RIC-5415 (a) Whenever the alert setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with approved CHEMISTRY SECTION procedures (e.g., CP-604).

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 126 of 319 (4) Changing the alert setpoint for 1/2-RIC-5415 (a) In all cases, the alert setpoint shall be set to a value which is less than or equal to the fixed setpoint.

(b) If the alert setpoint exceeds the maximum range of the monitor, the setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(c) CHEMISTRY SECTION procedures (e.g., CP-604) contain administrative controls associated with calculating and approving an alert setpoint.

(d) Whenever this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9) the calculated value for the alert setpoint shall be entered into the plant computer prior to each containment purge via the main vent.

2. Wide Range Gas Monitor (2-RE-5416) a) all information related to 1-RE-5416 is applicable to the Unit 2 WRGM with the following exceptions(s) b) Monitors equivalent to 2-RE-5416 (1) 2-RE-5415 [the "Westinghouse Plant Vent Stack Monitor"] has the capability of providing the measurement and alarm functions of 2-RE-5416 during times when 2-RE-5416 is declared inoperable (2) 2-RE-5415 provides redundant monitoring [for 2-RE-5416] at the low end of the concentration ranges (UFSAR 11.2.3.2.12)

3. Westinghouse Plant Vent Stack Monitor (1-RE-5415) a) The Westinghouse Plant Vent Stack Monitor contains 2 radiation elements (1) 1-RE-5414 (a) particulate detector (b) off-line scintillation detector (c) analog output (d) supplies signals to radiation indicator 1/2-RI-5414 (e) values displayed by 1/2-RI-5414 are in units of counts per minute (f) the detector manufacturer is Westinghouse (2) 1-RE-5415 (a) noble gas detector

OFFSITE DOSE CALCULATION MANUAL

-Page Rev. 7 127 of 319 (b)i off-lineGMTube (c) ^ analog output (d) supplies signals to-radiation indicator 1/2-RI-5415 (e) values displayed by 1/2-RI-5415 are In units of counts per minute (f) the detector manufacturer is Westinghouse b) Functions of 1-RE-5414 (1) The functions of 1-RE-5414 are mentioned here only as a basis for excluding this radiation element from the setpoint controls of Control 3.3.3.9.

(2) This monitor (the particulate monitor) was retired in place.

c) Functions of 1-RE-54151 (1) continuously measure the activity (cpm) of noble gases emanating from the Unit I main vent stack (Control 4.11.2.1.2, Table 4.11-2)

(2) continuously indicate (via 1-RI-5415) the activity (cpm) of noble gases emanating from the Unit 1 main vent stack (Control 4.11.2.1.2, Table 4.11-2)

(3) alarm (via 1-RI-5415) prior to exceeding the site-boundary, noble-gas, total-body-dose-rate limit of 500 mr/yr (per Control 3.11.2.1.a)

(4) alarm (via 1-RIC-5415) prior to exceeding the site-boundary, noble-gas, skin-dose-rate limit of 3000 mr/yr (per Control 3.11.2.1.a)

This (radiation element) monitors noble gases. Other radiation elements monitor particulates inthis waste stream.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 128 of 319 d) OPERABILITY of 1-RE-5415 (1) This monitor shall be operable (or have OPERABILITY) when it is capable of performing its specified function(s).

(2) The functions of 1-RE-5415 are listed in section (c) above.

e) Monitors equivalent to 1-RE-5415 (1) The Wide Range Gas Monitor (i.e., 1-RE-5416) has the capability of providing the measurement and alarm functions of 1-RE-5415 during times when 1-RE-5415 is declared inoperable.

(2) 1-RE-5415 provides redundant monitoring [for 1-RE-5416] at the low end of the concentration ranges (UFSAR 11.2.3.2.12).

(3) In the event 1-RE-5415 is inoperable or otherwise unavailable, 1-RE-5416 may fulfill the measuring, Indicating, and alarming functions normally provided by 1-RE-5415.

(4) The absence of a radiation element dedicated to measuring the particulate activity in the Wide Range Gas Monitor does not preclude the use of 1-RE-5416 as a backup fdr'1-RE-5415. This is mentioned only as a basis for excluding 1/2-RE-5414 from' the setpoint'controls of Control 3.3.3.9 (see "Functions of 1-RE-5414" earlier in this section).

f) Radiological effluent controls for 1-RE-5415 (1) Control 3.3.3.9 states that releases via the plant vent stack may continue if any one of the following three conditions are satisfied (a) 1-RE-5415 is operable AND the alarm setpoint for 1-RI-5415 is set to ensure the annual dose rates due to noble gases at the SITE BOUNDARY are less than 500 mr/yr to the total body and are less than 3000 mr/yr to the skin (per Control 3.11.2.1.a), or (b) an "equivalent monitor' (see section (e)above) is operable AND the alarm setpoint for the "equivalent monitor" is set to ensure annual dose rates due to noble gases at the SITE BOUNDARY are less than 500 mr/yr to the total body and are less than 3000 mr/yr to the skin (per Control 3.11.2.1.a), or (c) grab samples are obtained and analyzed for gross activity at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in accordance with Controls 3.11.2.1.a, 4.11.2.1.1, and 4.11.2.1.2 (per Control 4.3.3.9, Table 3.3-12, ACTION 37).

OFFSITE DOSE CALCULATION MANUAL PRaev 29 of 3197

--- (2) .Control 3.11.2.1.b (i.e., dose rates due to lodines and particulates at the SITE BOUNDARY) is not applicable to noble gas detector or to the setpoints related to the noble gas detector 1-RE-5415. As a result, the 1500 my/yr organ dose limit is not included as a radiological effluent control in this section of the ODCM.

g) Surveillances for 1-RE-5415 (1) Control 4.3.3.9 requires demonstrating the OPERABILITY of 1-RE-5415 by satisfying the checks, calibrations, and tests listed below:

(a) CHANNEL CHECK within the past 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (b) SOURCE CHECK within the past 31 days (c) CHANNEL CALIBRATION within the past 18 months (d) CHANNEL FUNCTIONAL TEST within the past 6 six months h) Setpoints for 1-RI-5415 (1) Requirements and commitments (a) The alarm and fixed setpoints ... shall be determined and adjusted in accordance with the methodology and parameters of the ODCM.

(Control 3.3.3.9)

(b) The method for calculating fixed or adjustable setpoints shall be provided in the ODCM. (NUREG-0133, 5.1.1)

(2) There are four alarms associated with, or otherwise related to, 1-RE-5415.

(a) 1-RI-5415 fixed high radiation alarm setpoint (b) 1-RI-5415 adjustable plant computer high radiation alarm setpoint (c) 1-RI-5415 low radiation alarm setpoint (d) 1-RI-5415 adjustable plant computer alert setpoint.

(3) In order to simplify the setpoint terminology, eliminate ambiguity, and minimize the possibility of misinterpretation, the ODCM will refer to these

setpoints as follows .

(a) The 1-RI-5415 fixed high radiation alarm setpoint will be referred to as the fixed setpoint.

(b) The 1-RI-5415 adjustable plant computer high radiation alarm setpoint will be referred to as the adjustable setpoint.

(c) The 1-RI-5415 low radiation alarm setpoint will be referred to as the low setpoint.

OFFSITE DOSE CALCULATION MANUAL PRage 130 of 319 (d) The 1-RI-5415 adjustable plant computer alert setpoint will be referred to as the alert setpoint.

(4) Each of these alarm setpoints are described below.

I) The fixed setpoint for 1-RI-5415 (1) General information (a) The fixed setpoint is not adjusted for each release.

(b) Whenever the fixed setpoint is exceeded, an alarm will be generated.

(c) The current value for the fixed setpoint is specified in the CCNPP Alarm Manual.

(d) The CCNPP Alarm Manual refers to this setpoint as the 1-RI-5415 High Alarm Setpoint.

(e) The fixed setpoint is integral to the Main Vent (Westinghouse) RMS as purchased from the supplier.

(f) The fixed setpoint is administratively controlled by EN-1-100.

(g) The fixed setpoint shall be calculated as described below'.

(2) Calculating the fixed setpoint for 1-RI-5415 (a) The fixed setpoint for 1-RI-5415 (plant vent stack monitor) shall be calculated in accordance with equation 4G.

THE FIXED SETPOINT FOR 1-RI-5415 I fl {K~fI[(x/Q) (Fd +Fdz) I) Y[(el)(AIL.)] Eq. 4G2 Srix = the fixed setpoint for 1-RI-5415 (counts per minute)

Ksf = a constant, actually a safety factor, which is the ratio of the CCNPP activity limit to the MPC limit, LMPc, used in equation 2G (unitless)

The safety factor chosen shall be less than or equal to 1.00. This ensures the fixed setpoint is always less than or equal to the MPC limit, LMpc, used in equation 2G.

A safety factor of 1.00 will yield a fixed setpoint which corresponds to the MPC limit, LMpc, in equation 2G.

The alarm and trip setpoints ...shall be determined and adjusted in accordance with the methodology and parameters of the ODCM. (Control 3.3.3.9).

2 Equation 4G has been derived from NUREG-0133, 5.2.1, (the 500 mr/yr equation).

Rev. 7

-0? OFFSITE DOSE CALCULATION MANUAL- Page 131 of 319 A safety factor of 0.500 will yield a fixed setpoint which corresponds to one-half the MPG limit, LMPC, in equation 2G.

It is recommended that a safety factor of 1.0 be used for calculating the fixed setpoint, however, other values--not to exceed 1.00-may' be used as directed by the General Supervisor Chemistry.

The particular value selected for the safety factor is somewhat arbitrary, however a safety factor does provide plant personnel with adequate time to respond to changing plant conditions and to initiate corrective ACTIONS so as to minimize the possibility of violating either the 10 CFR 50.72 limit or the Control 3.3.3.9 limits.

The use of a safety factor is consistent with the ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as is reasonably achievable.

The use of a "safety margin" is in accordance with the provisions of NUREG-Ol33, section 5.1.1, which states that "... the alarm and trip setpoints ... should correspond to a value(s) which represents a safe margin of assurance that the Instantaneous gaseous release limit of Control 3.11.2.1 (a)will not be exceeded."

This safety margin will prevent minor fluctuations In the nominal plant vent stack flow rates, errors in detector efficiencies, and other statistical aberrations from adversely impacting the calculated fixed setpoint.

x/Q - the highest calculated annual average relative concentration for any area at or beyond the UNRESTRICTED AREA boundary (2.2E-6 seconds per cubic meter)

Unit 1 and Unit 2 main vent releases are considered "long-term" releases1 , and as such, the highest historical annual average dispersion factor, (x/Q), is used in the setpoint calculations.

The highest annual average dispersion factor (x/Q) is 2.2E-6 (UFSAR, 2.3.6.3) for purposes of routine, long-term concentrations (e.g., routine noble gas releases).

The maximum annual average on-shore concentrations occur in the southeast sector at a distance of 1300 meters for purposes of routine, long-term concentrations (e.g., routine noble gas releases) (UFSAR, 2.3.6.3).

NUREG-0133, 3.3

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 132 of 319 Fd1 = the estimated main vent stack flow rate for Unit 1 (cubic meters per second)

Since the main vent stack flow rate will vary depending on the configuration of air dampers and the input gas streams, nominal main vent stack flow rate is used to calculate the fixed setpoint.

Use the nominal Unit 1 main vent stack flow rate listed on Attachment 7.

The main vent stack flow rate shall be determined, In accordance with approved procedures, at least once per 6 months (+/-25%). The Test and Equipment Unit shall be responsible for performing this test. The results of the main vent flow rate test shall be evaluated to ensure the main vent flow rates used in the ODCM are an accurate reflection of the true main vent flow rates. The RETS Program Manager is responsible for modifying the (main vent flow rates used in the) ODCM in the event the main vent flow rate for either Unit I or Unit 2 has increased to a value which is greater than the maximum discharge flow rates listed on Attachment 7.

Fd2 = the estimated main vent stack flow rate for unit 2 (cubic meters per second)

Since the main vent stack flow rate will vary depending on the configuration of air dampers and the input gas streams, nominal main vent stack flow rate is used to calculate the fixed setpoint.

Use the nominal Unit 2 main vent stack flow rate listed on Attachment 7.

The main vent stack flow rate shall be determined, in accordance with approved procedures, at least once per 6 months (+/- 25%). The Test and Equipment Unit shall be responsible for performing this test. The results of the main vent flow rate test shall be evaluated to ensure the main vent flow rates used in the ODCM are an accurate reflection of the true main vent flow rates. The RETS Program Manager is responsible for modifying the (main vent flow rates used in the) ODCM in the event the main vent flow rate for either Unit 1 or Unit 2 has increased to a value which is greater than the maximum discharge flow rates listed on Attachment 7.

ei = absolute detector efficiency for nuclide, i (cpm/microcuries per milliliter)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

AiLn = the specific activities of radionuclide, i, found in TYPICAL GASEOUS RADWASTE RELEASES (calculated in accordance with 10 CFR 20, Appendix B, Table II, Note 1 as described below; microcuries per milliliter)

Calculate AjLn in accordance with equation 5G.

- I tt ; Rev. 7 9NO OFFSITE DOSE CALCULATION MANUAL. -Pag 13 f 1 SPECIFIC ACTIVITY LIMIT FOR NUCLIDE I IN A RADIONUCLIDE MIXTURE Ajn = (f 1 )(ATn) Eq.5G fi = a fraction which represents the relative activity contribution of noble gas radionuclide i to the total noble gas activity for TYPICAL GASEOUS EFFLUENTS (unitless)

This value may be obtained using the guidance provided on Attachment 5.

ATLn= the sum of the total specific activities of all noble gas radionuclides found in TYPICAL GASEOUS RADWASTE RELEASES (microcuries/cm 3 )

Calculate AT~n in accordance with equation 2G.

SPECIFIC ACTIVITY CORRESPONDING TO THE SITE BOUNDARY LIMIT 7 [(fI)(ATLn)J/AiLt < LMPC Eq. 2G1 LMPc = the MPC limit The value chosen for LMPC in this equation is 2. The basis for this limit is 10 CFR 50.72.

It has been shown2 that, for the radionuclides present in TYPICAL GASEOUS EFFLUENTS from CCNPP, the 2 MPC limit is more restrictive than the limits of Control 3.3.3.9.

It should be noted that by using "2" as the MPC limit (10 CFR 50.72), instead of using the limits of Control 3.11.2.1(a), a safety factor has been incorporated into equation 2G.

The use of 2 MPCs as a safety margin is consistent with the provisions of NUREG-0133, section 5.1.1, which states that, "... in all cases, conservative assumptions may be necessary in establishing these setpoints to account for system variables, ... the variability in release flow, and the time lag between alarm and final isolation of radioactive effluents."

An alarm setpoint corresponding to 2 MPCs serves to initiate a determination of whether the "4-hour NRC notification" (specified in 10 CFR 50.72) is required.

The use of a limiting specific activity equivalent to 2 MPCs is consistent with the provisions of 10 CFR 20.

AILt = the specific activity limit for radionuclide, i, as obtained from 10 CFR 20, Appendix B, Table II, Column 1 (microcuries/cm 3 )

For all the radionuclides found in TYPICAL RADWASTE EFFLUENTS, use the value from 10 CFR 20, Appendix B, Table II, Column 1. An acceptable alternative is to assume an isotopic mix which results in a more conservative setpoint.

Equation 2G has been derived from 10 CFR 20, Appendix B,Table II,Note 1..

2 Addendum To Setpoint Calculations For WRGM Monitors 1-RIC-5415 and 2-RIC-5415, R.L. Conatser, December 10, 1991.

OFFSITE DOSE CALCULATION MANUAL Paev 137 f 1 (3) The low setpoint for 1-RI-5415 (a) The ODCM does not address the calculations associated with the low setpoint.

(b) The low setpoint is specified in the CCNPP Alarm Manual.

(c) The low setpoint may be used to determine OPERABILITY of this monitor (in accordance with the provisions of Control 4.3.3.9, Table 4.3-11, Note 2).

(4) Adjusting the fixed setpoint for 1-RI-5415 (a) If the fixed setpoint calculated in accordance with equation 4G exceeds the maximum range of the monitor, the fixed setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

(b) The fixed setpoint may be established at values lower than the maximum allowable setpoint, if desired.

(c) A setpoint change should be initiated whenever any of the parameters identified in equation 4G have changed.

(d) The fixed setpoint should not be changed unless one of the following occurs:

I) the relative activity' of any radionuclide in TYPICAL GASEOUS EFFLUENTS has changed by greater than 10%,

and the new radionuclide mixture yields a fixed setpoint which is 10% (or more) lower than the current fixed setpoint, ii) the historical maximum annual average atmospheric dispersion factor has changed, iii) the MPC limit at the SITE BOUNDARY, (presently 2 MPCs) has changed, iv) the estimated Unit 1 main vent stack flow rate or Unit 2 main vent stack flow rate has changed by greater than or equal to 1 Q%2, v) the values listed in 10 CFR 20, Table II, column 1 have changed, vi) the radiation monitor has been recently calibrated, repaired, or otherwise altered, or As determined in accordance with Attachment 5. 0 2 As determined by analysis of the TE-001, STP-M-461-1, STP-M-461-2, and ETP-87-16 test results.

Ca

Rev. 7

@>F OFFSITE DOSE CALCULATION MANUAL Page 135 of 319 vii) - the monitor is not conservative in its function (see section "Functions of 1/2-RE-5415" earlier in this section).

(e) EN-1 -100 contains the administrative controls associated with changing and approving the fixed setpoint.

j) Adjustable setpoint for 1/2-RI-5415 (1) General information (a) Whenever this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9), the adjustable setpoint is calculated and adjusted prior to each release of a WGDT, each containment vent, and each containment purge discharged via the main vent.

(b) The adjustable setpoint is based on the specific activities of the radionuclides present in either the WGDT or the containment building, whichever is applicable. (The radionuclide concentrations are determined by radiochemical analysis in accordance with applicable CHEMISTRY SECTION procedures as required by Control 4.1 1.2.1.2).

(c) Whenever the adjustable setpoint is exceeded, the WGDT, PURGE, or vent discharge via the main vent will be manually suspended.

(d) Refer to the Alarm Manual for a full list of operator ACTIONS taken in response to this alarm.

(e) The adjustable setpoint corresponds to the maximum concentration of radionuclides anticipated or expected when discharging a WGDT, a containment vent, or a containment purge via the main vent. For containment purges during outages, system evolutions may cause containment atmosphere activity to increase above what is normally expected for short periods of time.

(e The value for the adjustable setpoint is recorded on the gaseous release permit in accordance with applicable CHEMISTRY SECTION procedures.

(g) This alarm is not integral to the'main vent radiation monitor, as purchased from'the supplier.

(h) This alarm is'generated by the plant computer which monitors output from 1/2-RI-5415, and provides an alarm to plant operators when the 1/2-RI-5415 adjustable setpoint has been exceeded.' '

(i) When this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9), a value for the adjustable setpoint shall be calculated prior to each release of a WGDT, each containment vent, and each containment purge as shown below.

(2) - Calculating the adjustable setpoint for.1/2-RI-5415

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 136 of 319 (a) The adjustable alarm setpoint is based on the specific activity of the radionuclides in the undiluted gaseous waste (as determined by radiochemical analysis per Control 4.11.2.1.2), and the alarm setpoint is calculated as shown below.

ADJUSTABLE SETPOINT FOR 1/2-RI-5415 Sd 1 5 (K.,) (FU / Fd.) [a,(Ai.) (el) + Bkg Eq. 27G' Sadj = the adjustable setpoint for 1/2-RI-5415 (cpm)

Ksf = a constant, actually a safety factor, which allows for fluctuation in radiation monitor response (unitless)

This safety factor helps ensure the release is not unnecessarily terminated due to (1) electronic anomalies which cause spurious monitor responses, (2)statistical fluctuations in disintegration rates, (3) statistical fluctuations in detector efficiencies, 14) errors associated with sample analysis, (5) errors associated with monitor calibrations , and (6)anticipated short term variations in activity (this applicable to containment purges only).

It is recommended that a safety factor of 10 for containment purge releases be used for calculating the adjustable setpoint. However, other values for purge releases - not to exceed 10 -- may be used as directed by the General Supervisor Chemistry. A safety factor of 1.5 shall be used for all other gaseous releases.

The particular value selected for the safety factor is somewhat arbitrary, however a value less than or equal to 10 does provide plant personnel with adequate time to respond to changing plant conditions and to initiate corrective ACTIONS so as to minimize the possibility of violating either the 10 CFR 50.72 limit or the Control 3.3.3.9 limits.

The use of the safety factor is consistent with ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as is reasonably achievable.

The use of a "safety margin" is in accordance with the provisions of NUREG-0133 which states that "... the alarm and trip setpoints ... should correspond to a value(s) which represents a safe margin of assurance that the instantaneous gaseous release limit of Control 3.11.2.1(a) will not be exceeded." (per NUREG-0133, 5.1.1).

This safety margin will prevent minor fluctuations in the nominal plant vent stack flow rates, errors in monitor efficiencies, and other statistical aberrations from adversely impacting the calculated adjustable setpoint. Additionally for a special case of containment purges during outages, the safety factor allows for short term variations in activity created as a result of system evolutions in containment.

Fu = maximum undiluted radwaste flow rate (cubic meters per second)

Equation 27G has been derived from NUREG-01 33, Addendum, page AM-1.

2 The "analysis errors" and "calibration errors" refer to errors which are within established quality assurance and quality control limits.

Rev.

OFFSITE DOSE CALCULATION MANUAL Page 7137 of 319 Values of maximum undiluted radwaste flow rates for various waste streams are tabulated in Attachment 7.

FdX = the estimated main vent stack (diluted gaseous radwaste) flow rate for unit x (cubic meters per second)

Since the main vent stack flow rate will vary depending on the reactor unit, the configuration of air dampers, and the input gas streams, nominal main vent stack flow rate is used to calculate the adjustable setpoint.

Use the nominal main vent stack flow rate, for the appropriate unit, listed on Attachment 7.

The main vent stack flow rate shall 'be determined, in accordance with approved procedures, at least once per 6 months (+/- 25%). The Test and Equipment Unit shall be responsible for performing this test. The results of the main vent flow rate test shall be evaluated to ensure the main vent flow rates used in the ODCM are an accurate reflection of the true main vent flow rates. The RETS Program Manager is'responsible for modifying the (main vent flow rates used in the) ODCM in the event the' main vent flow rate for either Unit I or Unit 2 has increased to a value which is greater than the maximum discharge flow rates listed on Attachment 7.

Aiu = specific activity of radionuclide, i, in the undiluted waste stream, either the WGDT or containment building as applicable (microcuries per milliliter) e; = absolute detector efficiency for nuclide, i (cpm/microcuries per milliliter)

The detector efficiency for each radionuclide may be calculated from data collected during calibration of the radiation monitor.

Bkg = an approximation of the detector background (cpm)

(3) Documenting the adjustable setpoint for 1/2-RI-5415 (a) Whenever the adjustable setpoint is calculated, the specific values chosen for each of the variables shall be documented in accordance with approved CHEMISTRY SECTION procedures (e.g., CP-604).

(4) Changing the adjustable setpoint for 1/2-RI-5415 (a) In all cases, the adjustable alarm setpoint shall be set to a value which is less than or equal to the fixed setpoint.

(b) If the adjustable setpoint exceeds the maximum range of the monitor, the setpoint shall be adjusted to a value which falls within the normal operating range of the monitor.

- (c) CHEMISTRY SECTION procedures (e.g., CP-604) contain administrative controls associated with calculating and approving an adjustable setpoint.

(d) Whenever this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9 ) the calculated value for the adjustable setpoint shall be entered into the plant computer prior to

OFFSITE DOSE CALCULATION MANUAL PRage 38 of 319 each release of a WGDT, a containment vent, or a containment purge via the main vent.

k) Alert setpoint for 1/2-RI-5415 (1) General information (a) The alert setpoint is applicable to containment purges only.

(b) Whenever this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9), the alert setpoint is calculated and adjusted prior to each containment purge discharged via the main vent.

(c) The alert setpoint is based on the specific activities of the radionuclides present in the containment building. (The radionuclide concentrations are determined by radiochemical analysis in accordance with applicable CHEMISTRY SECTION procedures as required by Control 4.11.2.1.2).

(d) Whenever the alert setpoint is exceeded, the PURGE via the main vent may continue.

(e) The alert setpoint corresponds to a level of activity which indicates additional source term(s) may be present, and as a result, additional notifications and/or actions are required to identify the source and to accurately account for the activity discharged.

(f) The value for the alert setpoint is recorded on the gaseous release permit in accordance with applicable CHEMISTRY SECTION procedures.

(g) This alarm is not integral to the main vent radiation monitor, as purchased from the supplier.

(h) This alarm is generated by the plant computer which monitors output from 1/2-RI-5415, and provides an alarm to plant operators when the 1/2-RI-5415 alert setpoint has been exceeded.

(I) When this monitor is satisfying the minimum channels operable requirement (per Control 3.3.3.9), a value for the alert setpoint shall be calculated prior to each containment purge as shown below.

(2) Calculating the alert setpoint for 1/2-RI-5415 (a) The alert setpoint is based on the specific activity of the radionuclides in the undiluted gaseous waste (as determined by radiochemical analysis per Control 4.11.2.1.2), and the setpoint is calculated as shown below.

ALERT SETPOINT FOR 1/2-RI-5415

'6A Rev.7 OFFSITE DOSE CALCULATION MANUAL Page 139of-319 S.1.,t 1.50 (FU/FdX)Z (AuJ ) (e) + Bkg] Eq. 27G' Saert = the alert setpoint for 1/2-RI-5415 (cpm) 1.50 = a constant, actually a safety factor,'which allows for fluctuation in radiation monitor response (unitless).

This safety factor helps ensure the release is not unnecessarily terminated due to (1) electronic anomalies which cause spurious monitor responses, (2)statistical fluctuations in disintegration rates, (3) statistical fluctuations in detector efficiencies, (4) errors associated with sample analysis, and (5) errors'associated with monitor calibrations.2 1 Equation 27G has been derived from NUREG-0133, Addendum, page AA-1.

2 The "analysis errors' and "calibration errors" refer to errors which are within established quality assurance and quality control limits.

Rev. 7 OFFSITE DOSE CALCULATION MANUAL Page 140 of 319 The use of the safety factor is consistent with ALARA philosophy that licensees should make every reasonable effort to maintain radiation exposures, and releases of radioactive materials in effluents to UNRESTRICTED AREAS, as low as is reasonably achievable.