Text

Revs 1,1A & 2,Vol 4 to Safstor Odcm
	ML20205G918
Person / Time
Site:	Humboldt Bay
Issue date:	12/31/1998
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20205G895	List: HBL-99-006, Forwards Annual Radioactive Effluent Release Rept, Covering Period Jan-Dec 1998.Encl 2 Contains Revs 1 & 2 to Safstor ODCM as Required by Specification Section 4.2 of Safstor Offsite Dose Calculation Manual; ML20205G900; ML20205G918;
References
	GL-89-01, GL-89-1, PROC-981231, NUDOCS 9904070419
	Download: ML20205G918 (120)
	v • d • e

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 l- REVISION 1, I A & 2

[ TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL EFFEC.DATE 1998 PAGE i CHANGES TO THE ODCM DURING 1998 I

(Procedure Classification - Quality Related)

INTRODUCTION The SAFSTOR Off-site Dose Calculation Manual (ODCM)is provided to support implementation of I the Humboldt Bay Power Plant (HBPP) Unit 3 radiological emuent controls and radiological environmental monitoring. The ODCM is divided into two parts, Pan 1 - Specifications and Part II -

Calculational Methods and Parameters.

Pad I contains the specifications for liquid and gaseous radiological emuents (RETS) developed in !

accordance with NUREG-0473, Draft Radiological Effluent Technical Specipcations - BWR, by License Amendment Request (LAR) 96-02 and the radiological environmental monitoring program (RE: Both the RETS and the REMP were relocated from the Technical Specifications by LAR 96-02 in accordance with the provisions of Generic Letter 89-01, Implementation ofProgrammatic Controlsfor Radiological Effluent Technical Specipcations in the Administrative Controls Section of the TechnicalSpecifications and the Relocation ofProceduralDetails ofRETS to the Offsite Dose Calculation Manual or to the Process ControlProgram, issued by the NRC in January,1989.

I Implementation of the LA.R revised the instantaneous liquid concentration limits based on "old" 10 CFR 20 maximum permissible concentrations (MPCs) to 10 times the "new" 10 CFR 20, Appendix B, 1 Table 2, Column 2 emuent concentration limits (ECLs) and replaced the gaseous emuent instantaneous concentration limits at the site boundary with annual dose rate limits equating to the doses associated with the annual average concentrations of"old" 10 CFR 20, Appendix B, Table II, Column 1. The LAR also established limits for doses to members of the public from radiological emuents based on the as low as reasonably achievable (ALARA) design objectives of 10 CFR 50, Appendix 1 as applicable to a nuclear power plant which has been chut down in excess of 20 years and is in SAFSTOR Decommissioning. These dose limits were established fohawing the guidance of NUREG-0133, Preparation ofRadiological Effluent Technical Specificationsfor Nuclear Power Plants, and NUREG-0473. This guidance was modified, as appropriate, to reflect the SAFSTOR decommissioning licensing basis contained in the HBPP S AFSTOR Decommissioning Plan, the Environmental Report submitted as Attachment 6 to the HBPP SAFSTOR licensing amendment request and NUREG-1166, HBPP FinalEnvironmentalStatement.

The ODCM contains the REMP required by Technical Specification VII.G. This program consists of monitoring stations and sampling programs based on the SAFSTOR Decommissioning Plan and the Environmental Report which established baseline conditions for soil, biota and sediments. The REMP also includes requirements to participate in an interlaboratory comparison program.

Part 11 of the ODCM contains the calculational methods developed, following the above guidance, to be used in determining the dose to members of the public resulting from routine radioactive emuents l

released from HBPP during the SAFSTOR period. Part 11 also contains the methodology used to 9904070419 990331 PDR ADOCK 05000133 R PDR

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE ii determine emuent nanitor alarm / trip setpoints which assure that releases of radioactive materials remain within specified concentrations.

The ODCM also contains the Process Control Program (PCP) for solid radioactive wastes which is required by Technical Specification VII.E.1.j. The ODCM also contains administrative controls ;

regarding the content of the Annual Radiological Environmental Monitoring Report and the Annual Radioactive Emuent Release Report which are required by Technical Specifications Vll.J.1 and VII.J.3 and administrative controls regarding major changes to radioactive waste treatment systems.

9/24/98 The ODCM shall become effective after review by the Plant Staff Review Committee and approval by Em98 the Plant Manager in accordance with Technical Specification Section Vll.O. Changes to the ODCM shall be documented and records of reviews performed shall be retained. This documer.:ation shall contain suflicient information to support the change (including analyses or evaluations), and a determination that the change will maintain the level of radioactive emuent control required by the regulations listed in Technical Specification VII.F and not adversely impact the accuracy or reliability of emuent, dose, or setpoint calculations.

Changes sha!! be submitted to the NRC in the form of a complete and legible copy of the entire ODCM as part of, or concurrent with, the Annual Radioactive Emuent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed.

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U NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE iii TABLE OF CONTENTS PART I - SPECIFICATIONS Section Title Page 1.0 DEFINITIONS I-1 2.0 SPECIFICATIONS I-6 2.1 Radioactive Liquid Emuent Monitoring Instrumentation I-6 2.2 Radioactive Gaseous Emuent Monitoring Instrumentation 1-9 2.3 Liquid Emuent - Concentration 1-12 2.4 Liquid Emuent - Dose I-16 2.5 Liquid Waste Treatment I-17 2.6 Gaseous Emuents - Dose Rate 1-18 2.7 Gaseous Emuents: Dose - Noble Gases 1-22 2.8 Gaseous Emuents: Dose - Tritium and Radionuclides in Pa ticulate Form I-23 2.9 Solid Radioactive Waste I-24 2.10 Total Dose I-25

2. I 1 REMP Monitoring Program 1-26 2.12 REMP Interlaboratory Comparison Program 1-39 3.0 SPECIFICATION BASES I-40 3.1 Radioactive Liquid Emuent Monitoring Instmmentation Basis 1-40 3.2 Radioactive Gaseous Emuent Monitoring Instrumentation Basis I-40 3.3 Liquid Emuent Concentration Basis 1-40 3.4 Liquid Emuent Dose Basis 1-41 3.5 Liquid Waste freatment Basis 1-41 3.6 Gaseous Emuents Dose Rate Basis I-41 3.7 Gaseous Emuents: Noble Gases Dose Basis 1-42 3.8 Gaseous Emuents: Tritium and Radionuclides in Particulate Form Dose Basis 1-43 3.9 Solid Radioactive Waste Basis 1-43 3.10 Total Dose Basis 1-43 3.11 REMP Monitoring Program Basis 1-44 3.12 REMP Interlaboratory Comparison Program Basis 1-45 l 4.0 ADMINISTRATIVE CONTROLS 1-46 l

4.1 Annual Radiological Environmental Monitoring Report 1-46

-4.2 Annual Radioactive Emuent Release Report I-50 4.3 Special Reports 1-51 4.4 Major Changes to Radioactive Waste Treatment Systems 1-51

i l NUCLEAR POWER GENERATION NUMBER ODCM f' HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE iv PART II- CAI,CULATIONAI, METilODS AND PARAMETERS Section Title Paue 1.0 EFFLUENT MONITOR SETPOINT C ALCULATIONS 11-1 1.1 Liquid EfIluent Monitors 11-1 1.2 Gaseous Efiluent Monitor 11-3 2.0 LIQUID EFFLUENT DOSE CALCULATIONS 11-6 2.1 Calendar Quarter 11-6 2.2 Calendar Year 11-6 2.3 Liquid Effluent Dose Calculation Methodology 11-7 3.0 LIQUID WASTE TREATMENT 11-1 0 3.1 Treatment Requirements 11- 1 0 3.2 Treatment Capabilities 11-1 0 i 4.0 GASFOUS EFFLUENT DOSE CALCULATIONS 11-1 4 4.1 Dose Rate 11-1 4 4.2 Dose - Noble Gases 11-1 5 4.3 Dose - Tritium and Radionuclides in Particulate Form 11-1 7

) 1 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE 11-3 0 l 5.1 Whole Body Dose 11-3 0 5.2 Skin Dose 11-3 0 5.3 Dose to Other Organs 11-3 0 5.4 Dose to the Thyroid 11-31 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING 11-3 2 SOLIDIFICATION 7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED 11-3 3 IN HIGH INTEGRITY CONTAINERS j 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED 11-3 4 RESINS AND OTHER WET WASTES 9.0 PROGRAM CHANGES 11-3 5 j

r; NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE y PART II- CALCULATIONAL METIIODS AND PARAMOTERS Section Title Page APPENDIX A - SAFSTOR BASELINE CONDITIONS A-1 APPENDIX B - BASIS FOR INSTANTANEOUS X/Q VALUE B-1 APPENDIX C - Kr-85 MONITOR CA.LIBRATION C-1

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i NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

' REVISION 1, I A & 2

TITLC SAFSTOR OFFSiTE DOSE CALCULATION MANUAL PAGE vi LIST OF TABLES - PART I Table Title Page 1-1 Frequency Notation I-5 2-1 RadioacLve Liquid Efiluent Monitoring Instrumentation I-7 2-2 Radioactive Liquid Efiluent Monitoring Instrumentation Surveillance Requirements 1-8 2-3 Radioactive Gaseous Efiluent Monitoring Instrumentation I-10 2-4 Radioactive Gaseous Efiluent Monitoring Instnimentaten Surveillance I 11 Requirements 2-5 Radioactive Liquid Waste Sampling and Analysis Program I-13 2-6 Radioactive Gaseous Waste Sampling and Analysis Program I-19 2-7 HBPP Radiological Environmental Monitoring Program I-28 2-8 Reporting Levels frr Radioactivity Concentrations In Environmental Samples I-30 2-9 Detection Capabilities for Environmental Sample Analysis Lower Limits Of I-31 Detection (LLD) 2-10 Distances and Directions To Environmental Monitoring Stations l'33 4-1 Radiological Environmental Monitoring Progran: Annual Report Summary - I-48 Example m ..

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFS?TE DOSE CALCULATION MANUAL PAGE vii LIST OF TABLES - PART II Table Title Pagg 9/24/98 l l-1 Liquid EfIluent Monitor Alarm Setpoints 11-3 2-1 Ingestion Dose Factors for Adult Age Group 11-8 2-2 Ingestion Dose Factors for Teen Age Group 11-8 2-3 Ingestion Dose Factors for Child Age Group 11-8 2-4 Bioaccumulation Factors for Saltwater Environment 11-9 2-5 Average Individual Foods Consumption for Various Age Groups 11-9 2-6 Maximum Individual Foods Consumption for Variaus Age Groups 11-9 4-1 Inhalation Dose Factors for Adult Age Group 11-25 4-2 Inhalation Dose Factors for Teen Age Group 11-2 6 4-3 Inhalation Dose Factors for Child Age Group 11-2 6 4-4 Inhalation Dose Factors for Infant Age Group 11-2 6 4-5 External Dose Factors for Standing on Contaminated Grour;d 11-2 7 4-6 Average Individual Foods Consumption for Various Age Groups 11-2 7 4-7 Maximum Individual Foods Consumption for Various Age Groups 11-2 7 4-8 Ingestion Dose Factors for Adult Age Group 11-2 8 4-9 Ingestion Dose Factors for Teen Age Group 11-2 8 4-10 Ingestion Dose Factors for Child Age Grono 11-2 8 4-11 Ingestion Dose Factors for Infant Age G~ y 11-2 9 4-12 Stable Element Transfer Data For Cow-h.W Path 11-2 9 4-13 Stable Element Transfer Data For Cow-Meat Path 11-2 9 i,

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r NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLA o VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE viii i

LIST OF FIGURES - PART I l Fieure Title Page 2-1 HBPP Onsite TLD Locations 1-34 2-2 HBPP Onsite Monitoring Well Locations 1-35 2-3 HBPP OITsite Sampling Locations 1-36 2-4 HBPP Offsite Sampling Locations (Continued) 1-37 2-5 HBPP Offsite Sampling Locations (Continued) I-38 i

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l NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-1 PART I - SPECIFICATIO'NS 1.0 DEFINITIONS 1.1 ACTION l ACTION shall be that part of a control that prescribes remedial measures required under designated conditions.

1.2 CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with 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. CHANNEL CALlBRATION may ve performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

1.3 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.

l 1.4 CHANNEL FUNCTIONAL TEST !

a. Analog channels - one injection of a simulated or actual signal into the channc! as close to the sensor as practicable to verify OPERABILITY including required alarms, interlocks, display, and trip functions.

b. Bistable channels - the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including alarm and trip functions.

1.5 ENVIRONMENTAL REPORT Submitted as Attachment 6 to the SAFSTOR license amendment request, the ENVIRONMENTAL REPORT established baseline radiological environmental conditions for soil, biota and sediments. In accordance with the NRC approved SAFSTOR Decommissioning Plan, these baseline conditions will only need to be reestablished prior to DECON if a significant release during SAFSTOR occurs as the result of an accident.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-2 1.6 FREQL'ENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1-1.

1.7 INDEPENDENT VERIFICATION

! INDEPENDENT VERIFICATION is a separate act of confirming or substantiating that an activity or condition has been completed or implemented, in accordance with specified requirements, by an individual not associated with the original determination that the activity. or condition was completed or implemented in accordance with specified requirements.

l 1.8 INSTANTANEOUS CONCENTRATION INSTANTANEOUS CONCENTRATION is the concentration averaged over one hour of radioactive materials in effluents.

1.9 LIQUID RADWASTE TREATMENT SYSTEM The LIQUID RADWASTE TREATMENT SYSTEM shall be any available equipment (e.g., filters, evaporators, demineralizers, or contiactor services) capable of reducing the ;

quantity of radioactive material, in liquid efiluents, prior to discharge.

1.10 MEMBER OF THE PUBLIC

- MEMBER OF THE PUBLIC means an individual in any area located beyond the boundary of the restricted area controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials and within, at, or beyond the SITE BOUNDARY. However, an individual is not a member of the public during any period in which the individual receives an onsite occupational dose.

L11 OFFSITE DOSE CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANUAL contains the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program.

The ODCM also contains the Radioactive Efiluent Controls and Radiological Environmental Monitoring Program and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Elliuent Release Reports. The ODCM also contains the Process Control Program (PCP) for solid radioactive wastes which is required by Technical Specification VII.E.1.j.

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT /OLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-3 1.12 OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s). Implicit in this dermition shall be the a sumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment, that are required for the system, subsystem, train, component or device ta perform its function (s), are also capable of performing their related support function (s).

1.13 PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that 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 Parts 20,61, and 71, State regulations, buiial ground requirements, and other requirements governing the disposal of solid radioactive waste.

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

1.15 SITE BOUNDARY l

The SITE BOUNDARY shall be the bounday of the unrestricted area used in the offsite dose calculations for gaseous and liquid efHuents as defined in Technical Specification ll.B.

1.16 SOLIDIFICATION SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with defmite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

1.17 SOURCE CHECK A SOUPCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLOT BAY POWER PLANT VOLUME 4

.7EVISION 1, I A & 2

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-4 1.18 UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area located beyond the boundary of the restricted area controlled by the licensee for purposes of protection ofindividuals from exposure to radiation and radioactive materials acd within, at, or beyond the SITE BOUNDARY.

1.19 URANIUM FUEL CYCLE As defined in 40 CFR Part 190.02(b)," URANIUM FUEL CYCLE means the operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal l sites, transportation of any radioactive material in support of these operations, and the l reuse of recovered non-uranium special nuclear and by-product materials from the cycle."

1.20 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and l installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to release to the environment (such a system is not considered to have any effect on noble gas efiluents).

1.21 VENTING ;

VENTING is the controlled process of discharging air or gas from a confinement to maintain temperaturc, 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. '

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-5 Table 1-1 1 FREQUENCY NOTATION 2

NOTATION FREOUENCY EXTENSION PERIOD ,

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . None gg W ,

At least once per 7 days. 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months M At least once per 31 days. 7 davs Q At least once per 92 days. 22 days SA At least once per 184 days. 45 days A At least once per 365 days. 91 days 4 P Complcted prior to each release.

N.A. Not applicable.

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.e extension period for a frequency of a week or longer is 25% with a maximum tolerance of 325% for three MS8 consecutive periods.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1,1 A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-6 2.0 SPECIFICATIONS 9GG8 l 2.1 RAD;OACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION i 9G4/98 l LIMITING CONDITIONS 9G4/98 l 2.1.1 The radioactive liquid emuent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of specification 2.3 are not exceeded.

APPLICABILITY: At all times ACTION:

a. With a cadioactive liquid emuent monitoring instrumentation channel alarm / trip setpoint ;

less conservative than required above, without delay suspend the release of radioactive liquid emuents monitored by the affected channel, or change the setpoint so that it is acceptably conservative, or declare the channel inoperable.

b. With one or more radioactive liquid effluent monitoring instrumentation channels inoperable, take the ACTION shown in Table 2-1. For the instrumentation covered by items 1 and 2 of the table, exert best efforts to return the inoperable instrument (s) to OPERABLE status within 30 days. If the affected instrument (s) cannot be returned to OPERABLE status within 30 days, provide information on the reasons for inoperability and lack of timely corrective action in the next Radioactive Emuent Release Report.

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! SURVEILLANCE REQUIREMENTS 1

gnesl 2.1.2 Each radioactive liquid emuent monitoring instrumentation channel shall be I

demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHFCK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 2-2.

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-7 Table 2-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1, Gross Radioactivity Monitors Providing Automatic Termination '

ofRelease

a. Process Water Monitor 1 21 5/7/98 l

2. Flow Rate Measurement Devices

a. None Table Notation ACTION 21 With less than the required number of OPERABLE channels, effluent releases via this pathway may continue, provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Specification 2.3.1, and

b. An INDEPENDENT VERIFICATION of release rate calculations is performed, and

c. An INDEPENDENT VERIFICATION of discharge valve lineup is performed.

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Otherwise, suspend releases of radioactive materials via this pathway.

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m NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-8 Table 2-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Channel Source Channel Functional Instrument Check Check Calibration Test

1. Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Process Water Monitor D Q A Q(1)(2) 5!7/98 l

2. Flow Rate Measurement Devices

a. None Table Notation (1) Alarm functions and background readings shall be checked weekly. If a background reading exceeds the equivalent of 5 x 10 5 micro-Ci/ml of Cs-137, the cause will be investigated and remedial measures taken to reduce the background reading.

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alarm setpoint.

5(1/98 9/24S8 l b. Circuit failure.

ERR 8 9n4a8 l c. Instrument indicates a downscale failure.

5088 l ST//98 l f/34/98 I l

I NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

, REVISION 1, I A & 2 l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-9 9G4/98 l 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 904/98 l LIMITING CONDITIONS 9G4/98 l 2.2.1 The radioactive gaseous emuent monitoring instrumentation channels shown in Table 2-3 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of specification 2.6 are not exceeded.

APPLICABILITY: Whenever the ventilation system is in operation.

ACTION:

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

b. With one or more radioactive gaseous emuent monitoring instrumentation channels inoperable, take the ACTION shown in Table 2-3. For the instrumentation covered, exert best effons to return the inoperable instrument (s) to OPERABLE status within 30 days. If the affected instrument (s) cannot be returned to OPERABLE status within 30 days, provide information on the reasons for inoperability and lack of timely corrective action in the next Radioactive Emuent Release Report.

SURVEILLANCE REQUIREMENTS 9G4/98 l 2.2.2 Each radioactive gaseous emuent 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 2-4.

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r NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1,1 A & 2 i TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 10 Table 2-3 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION

1. Stack Gas Monitoring System maa l a. Noble Gas Activity Monitor 1 23,24

b. Iodine Sampler

N.A.

i c. Particulate Sampler 1 23,25 mas d. Emuent System Flow Rate Monitor 1 26

e. Sampler Flow Rate Monitor *

1 Table Notation ,

l ACTION 23 The monitor may be taken out of service for calibration or maintenance, but shall be returned to service as soon as practicable within the 30 day period allowed by ACTION 2.2.b. )

, ACTION 24 With the number of channels OPERABLE less than that required by the Minimum Channels OPERABLE requirement, emuent releases via this pathway may continue for up to 20 days provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for noble gas activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

g33 ACTION 25 With the number of channels OPERABLE less than that required by the Minimum i Channels OPERABLE requirement, emuent releases via this pathway may continue for up to 30 days provided samples are continuously collected as required in Table 2-6.

ACTION 26 With the number of channels OPERABLE less than that required by the Minimum Channels OPERABLE requirement, the emuent system default flow rate may be used for emuent calculations.

Not included in the stack gas monitoring system .

l- ** Loss of sampler flow would result in alarm and failure of Noble Gas activity monitor and ;

. mas particulate sampler.

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f NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL, PAGE I - 11 Table 2-4 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUnlENTATION SURVEILLANCE REQUIREMENTS Channel Channel Source Channel Functional Instrument Check Check Calibra"on Test

1. Stack Gas Monitoring System

a. Noble Gas Activity Monitor D M A Q(1)

b. lodine Sampler

N.A. N.A. N.A. N.A.

c. Particulate Sampler W N.A. N.A. N.A. j was d. Efiluent System Flow Rate Monitor W N.A. A N.A. I

e. Sampler Flow Rate . Monitor Q N.A. N.A. N.A.

Table Notation (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alarm setpoint .

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b. Circuit failure MS8 l c. Instmment imlicates a downscale failure. **

Not included in the stack gas monitoring system.

Ms8 l ** No downscale failure provided on this instmment. J l

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F l NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME' 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 12

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8G4/98 l 2.3 LIOUID EFFLUENT- CONCENTRATION t 9G4S8 l LIMITING CONDITIONS l I 9G4/98 l 2.3.1 The instantaneous concentration of radioactive material released beyond the SITE BOUNDARY shall be less than or equal to 10 times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2.

l APPLICABILITY: At all times.

ACTION With the instantaneous concemration of radioactive mate:ials released beyond the SITE BOUNDARY exceeding the above limits, without delay restore the concentration of radioactive materials being released beyond the SITE BOUNDARY to within the above limits. i l

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l SU~RVEILLANCE REQUIREMENTS i

9G4/98 l 2.3.2 Radioactive liquid wastes shall be sampled and analyzed in accordance with the sampling l

and analysis program of Table 2-5. j l \

l 9G4/98 l 2.3.3 The results of the radioactivity analyses shall be used with the calculational meti ods in Part 11 of the ODCM to assure that the concentrations of radioactive material released to 9G4S8 l .Humboldt Bay are maintained within the limits of Specification 2.3.1.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE : - 13 Table 2-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PdOGRAM Minimum Lower Limit Sampling Analysis Type of Activity of Detecdon Liquid Release Type Frequency Frequency Analysis (LLD)

(pCi/ml)'

A. Batch Waste Release Tanks

P P Principal Gamma 5 x 10

l. Treated Waste Hold Tank (2) Each Batch Each Batch Emitters *

2. Waste Receiver Tanks (3) P M H-3 1 x 10

6 Each Batch Composite Gross Alpha 1 x 10-7 P Sr-90 5 x 10'"

Q 6 Each Batch Composite B. Plant Continuous Releases d

D W Principal Gamma 5 x 10

l. Caisson Sump Grab Sample Compositeh Emitters

M88 D M H-3 1 x 10

6 Grab Sample Composite Gross Alpha 1 x 10'7 D Q Sr-90 5 x 10'"

6 Grab Sample Composite Table Notation a The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal .

W/98 l

For a particular measurement system (which may include radiochemicr.1 separation):

sh LLD =

6 (E)(V)(2.22 x 10 )(e-w) y gnes For low background counting, the methodology of NUREG-4007 is used as follows:

_LD = 3 + 4.66 sb 6

(E)(V)(2.22 x 10 )(e-w) y 1

L

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 14 Table 2-5 (Continued)

Table Notation Where:

LLD is the lower limit of detection as defined above (as microcurie per unit mass or volume),

s3i s 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 10' is the number of disintegrations per rainute per microcurie, Y is the fractional radiochemical yield (when applicable),

1 is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant efiluents, not environmental samples).

gams 3 is a term introduced to account for minor deviations of the Poisson distribution from the Normal distribution for low count rates.

8088l Typical values of E, V, Y, and At shall be used in the calculation.

The LLD is defincd as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (afler the fact) limit for a particular measurement.

b. A composite sample is one in which the quantity ofliquid sampled is proportional to the quantity ofliquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

c. A batch release is the discharge ofliquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed to assure representative sampling.

d. A continuous release is the discharge ofliquid wastes of a nondiscrete volume; e.g., from a volume or system that has an input flow during the continuous release.

r NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 15

e. The principal gamma emitters for which the LLD specification applies exclusively are Co-60 and Cs-137. This does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be Srl/98 l identified and reported. Nuclides which are not detected for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level t~or that

! nuclide. The "less than" values shall not be used in the required dose calculations.

l l

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g NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 16 9G4/98 l 2.4 LIOUID EFFLUENT- DOSE 9/24/98 l LIMITING CONDITIONS 9/24/98 l 2.4.1 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid efiluents released beyond the SITE BOUNDARY shall be limited as follows:

i j a. During any calendar quarter to less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ.

b. During any calendar year to less than or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid efiluents exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a Special Report gg4fgg l pursuant to Administrative Control 4.3, which includes: j l

a. Identification of the cause for exceeding the limit (s);

1

b. Corrective action taken to reduce the release of radioactive materials in liquid efiluents !

during the remainder of the current calendar quarter and during the remainder of the ,

current calendar year so that the dose or dose commitment to a MEMBER OF THE l PUBLIC from this source is less than or equal to 3 mrem total body and less than or equal to 10 mrem to any organ during the calendar year. j i

r 4 SURVEILLANCE REQUIREMENTS 9/24/98 l 2.4.2 Baseline Comparison. Cumulative activity contributions from liquid efiluents shall be compared with the baseline conditions established by the Environmental Report submitted to the NRC as Attachment 6 to the SAFSTOR licensing amendment request at least once per 31 days. IF the comparison indicates that the activity released will exceed the Environmental Report baseline release for the current calendar quarter, THEN a dose calculation shall be performed.

NUCLEAR POWER GENERATION NUMBER ODCM l HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1,1 A & 2 l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 17

! i 9GM8 l ' 2.5 I IOUID WASTE TREATMENT 9GM8 l LIMITING CONDITIONS __

9UM8 l 2.5.1 The LIQUID RADWASTE TREATMENT SYSTEM shall be used, as appropriate, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to liquid effluents discharged to Humboldt Bay would exceed the action levels of 0.06 mrem whole body or 0.2 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

When radioactive liquid waste, in excess of the above action levels, is discharged without prior treatment, prepare and submit to the Commission within 30 days, a Special Report pursuant to snesl Administrative Control 4.3, which includes the folbwing information:

a. Identification ofinoperable equipment arid the reasons for inoperability.

b. Actions taken to restore the inoperable e quipment to OPERABLE status.

c. Actions taken to prevent recurrence.

1 i

SURVEILLANCE REQUIREMENTS enasl 2.5.2 Baseline Comparison. Activity contributions from liquid effluents shall be compared with the baseline conditions established by the Environmental Report at least once per 31 days. IF the comparison indicates that the activity released will exceed the Environmental Report baseline release, THEN a dose calculation shall be performed.

)

1 NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 18 i

l 94488 l Lp GASEOUS EFFLUENTS- DOSE RATE 9G4S8 l LIMITING CONDITIONS l

9G4/98 l 2.6.1 The dose rate at or beyond the SITE BOUNDARY, due to radioactive materials released in gaseous efiluents, shall be limited as follows:

a. Ncble gases: less than or equal to 500 mrem / year total body and less than or equal to 3000 mrem / year to the skin.

b. Tritium and radioactive particulates with half-lives ofgreater than 8 days: less than or equal to 1500 mrem / year to any organ.

APPLICABILITY: At all times.

ACTION:

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

l l ,

l SURVEILLANCE REQUIREMENTS l

90488 l 2.6.2 The dose rate due to noble gases in gaseous effluents shall be determined to be within the i above limits. This determination has been established by the Environmental Report.

gn43, l 2.6.3 The dose rate due to radioactive materials specified above, other than noble gases, in gaseous efiluents shall be determined to be within the above limits by obtaining l- representative samples and performing analyses i cecordance with Table 2-6 and comparing cumulative activity released with the Environmental Report baseline ,

conditions.

l j .-

1 NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 )

REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 19 Table 2-6 ,

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit Sampling Analysis Type of Activity ofDetection Sr/88 l Gaseous Release Type Frequency Frequency Analysis (LLD)

(pCi/ml)*

Plant Stack 7 Q* Noble Gas (Kr-85) 1 x 10" Grab Sample d

Continuous W" Principal Ganuna 1 x 10'"

Particulate Emitters

Sample d

Continuous M Gross Alpha 1 x 10'"

]

Composite -

Particulate l

Sample Continuous' Q Sr-90 1 x 10'"

Composite Particulate Sample d 4 Continuous Noble Gas Noble Gas 1 x 10 Monitor Gross Beta Table Notation The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

g33 l

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

Sb LLD = f I

(E)(V)(2.22 x 10')(e~") Y 9/24/98 For low background counting, the methodology of NUREG-4007 is used as follows:

b LLD = f (E)(V)(2.22 x 10')(e-") Y l

r NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 20 Table 2-6 (Continued)

Table Notation Where:

LLD is the lower limit of detection as defmed above (as microcurie per unit mass or volume),

s6 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 10' is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

1 A is the radioactive decay constant for the particular radionuclide, and i At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

3 is a term introduced to account for minor deviations of the Poisson distribution from M8 the Normal distribution for low count rates.

ma l Typical values of E, V, Y, and At shall be used in the calculation.

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.

6 Analyses shall also be performed following an occurrence which could alter the mixture of radionuclides. !

i 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).

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 the Specifications 2.6, 2.7, and 2.8.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 21 Table 2-6 (Continued)

Table Notation The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr 85 for gaseous emissions and Co-60 and Cs-137 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. )

Other peaks which are measurable and identifiable, together with the above nuclides, shall also I sm98 l be identified and reported. Nuclides which are not detected for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that i nuclide. The "less than" values shall not be used in the required dose calculations. j l

l l

1 NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 j REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 22 i

9G4/98 l 2.7 GASEOUS EFFLUENTS: DOSE - NOBLE GASES 9G4/98 l LIMITING CONDITIONS 9/24/98 l 2.7.1 The air dose at or beyond the SITE BOUNDARY due to radioactive noble gases released in gaseous effluents shall be limited to :

l

a. During any calendar quarter: less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year: less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad beta radiation.

APPLICABILITY: At all times.

ACTION:

With the calculated air dose from radioactive noble gases in gaseous efiluents exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a Special Report, gg4;gg l pursuant to Administrative Control 4.3, which includes:

a. Identification of the cause for exceeding the limit (s).

b. Corrective action taken to reduce the release of radioactive noble gases in gases effluents during the remainder of the current calendar quarter and during the remainder of the current calendar year so that the average dose during the calendar year is less than or equal to 10 mrad gamma and 20 mrad beta radiation.

SURVEILLANCE REQUIREMENTS 9/24/98 l 2.7.2 Compliance with these Specifications for normal SAFSTOR conditions has been established on a licensing basis by the Environmental Report and NUREG-1166, Final Environmental Statementfor Decommissioning Humboldt Bay Power Plant, Unit No. 3, issued by the NRC. If an accident involving spent fuel occurs during the SAFSTOR period, the noble gas activity released in gaseous effluents shall be compared with the baseline conditions established by the Environmental Report. IF the comparison indicates that the activity released will exceed the Environmental Report baseline release, THEN a dose calculation shall be performed.

l L

p l NUCLEAR POWER GENERATION

~~~

MjMBER ODCM ,

HUMBOLDT BAY POWER PLANT VOLUME 4 l REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 23 l

9/24/98 l 2.8 GASEOUS EFFLUENTS: DOSE-TRITIUM AND RADIONUCI,lDES IN PARTICULATE FORM l

9G4/98 l LIMITING CONDITIONS l

9/24/98 l 2.8.1 The dose to a MEMBER OF THE PUBLIC from the release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous emuents released beyond the SITE BOUNDARY shall be limited as follows:

a. During any calendar quarter: less than or equal to 7.5 mrem to any organ, and

b. During any calendar year: less than or equal to 15 mrem to any organ.

APPLICABILITY: At all times .

ACTION:

With the calculated dose from the release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous emuents exceeding any of the above limits, prepare j 9/24/98 and submit to the Commission, within 30 days, a Special Report, pursuant to Administrative Control 4.3, which includes:

a. Identification of the cause for exceeding the limit (s).

b. Corrective action taken to reduce the release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous emuents during the remainder of the current calendar quarter and during the remainder of the current calendar year so that the average dose to any organ is less than or equal to 15 mrem.

I SURVEILLANCE REQUIREMENTS l

9/24/98 l 2.8.2 .B_aseline Congarison. Cum .lative Activity contributions from tritium and radioactive l materials in particulate form in gaseous emuents shall be compared with the baseline !

conditions established by the Environmental Report at least once per 31 days. IF the l comparison indicates that the activity released will exceed the Environmental Report baseline release, THEN a dose calculation shall be performed.

L NUCLEAR POWER GENERATION NUMBER ODCM

HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 24 l 9/24/98 l 2.9 SOLID RADIOACTIVE WASTE 9/2438 l LIMITING CONDITIONS endas l 2.9.1 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.

APPLICABILITY: At alltimes.

ACTION: ,

i With the provisions of the a PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.

1 SURVEILLANCE REQUIREMENTS 9/24/98 l 2.9.2 The PROCESS CONTROL PROGRAM, as defined in Section 1.0, shall be used to verify that processed wet radioactive wastes (e.g., filter sludges, spent resins and evaporator bottoms) meet the shipping and burial ground requirements with regard to i solidification and dewatering. )

1 l

i l

l J

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 l

REVISION 1,1 A & 2

(

I TITLE SAFS'IOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 25 l

944/98 l 2.10 TOTAL DOSE

! 944/98 l LIMITING CONDITIONS 9GG8 l 2.10.1 The calendar year dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem).

APPLICABILITY: At all time s.

ACTION:

With the calculated doses from the release of radioactive materials in liquid or gaseous effluents 9n4/98 l exceeding twice the limits of Specification 2.4.1.a,2.4.1.b, 2.7.1.a,2.7.1.b,2.8.1.a, or 2.8.1.b, calculations should be made, which include direct radiation contributions from the reactor, to determine whether the above limits of Specification 2.10 have been exceeded. If such is the !

goesl case, prepare and submit to the Commission within 30 days, pursuant to Administrative Control '

4.3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits arid includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all efiluent 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.

t.

SURVEILLANCE REQUIREMENTS 9G4/98 l 2.10.2 DOSE CALCULATIONS - Annual dose contributions from Jiquid and gaseous efiluents shall be calculated in accordance with dose calculation methodology provided for Specifications 2.4.1,2.7.1, and 2.8.1.

l

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 26 sne8 l 2.11 REMP MONITORING PROGRA,M 90a8 l LIMITING CONDITIONS

, 9GE8 l 2.11.1 The radiological environmental monitoring program shall be conducted as specified in Table 2-7.

APPLICABILITY: At all times.

ACTION:

a. With the radiological environmental monitoring program not being conducted as specified in Table 2-7, prepare arJ submit to the Commission, in the Annual Radiological Environmental Monitoring Report, a description of the reasons for not conducting the progtam as required and the plans for preventing a recurrence.

b. With the level of radioactivity, resulting from plant efiluents, in an environmental sampling medium exceeding the reporting levels of Table 2-8 when averaged over any calendar quarter, prepare and submit to the Commission, within 30 days of obtaining analytical results from the affected sampling period, a Special Report pursuant to gnesl Administrative Control 4.3, which includes an evaluation of any release conditions,

, environmental factors or other aspects which caused the limits of Table 2-8 to be exceeded. When more than one of the radicnuclides in Table 2-8 are detected in the sampling medium, this report shall be submitted if:

93 , 3 concentration (1) concentration (2)

+ .2 1.0 reporting level (1) reporting level (2)

This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Monitoring Report.

When radionuclides other than those in Table 2-8 are detected and are the result ofplant eflluents, this repon shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is greater than or equal to the calendar year limits of Specifications 2.4, 2.7, and 2.8. This report is not required if the measured level of radioactivity was not the result of plant ellluents; however, in such an event, the condition shall be reported and desciibed in the Annual Radiological En ironmental Monitoring Report.

I NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1,1 A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 27 SURVEILLANCE REQUIREMENTS 9/34/98 l 2.11.2 The radiological environmental monitoring samples shall be collected pursuant to Table 2-7 from the locations given in Tables 2-7 and 2-10 and Figures 2-1,2-2,2 .',2-4 and 2-5 and shall be analyzed pursuant to the requirements of Tables 2-7 and 2-9.

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l Table 2-8 I I

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Analysis Water (pCi/L)

H-3 20,000*

Co-60 300 4

Cs-137 50 l

For drinking water samples. This is the i 40CFR141 value. If no drinking water pathway exists, a value of 30,000 pCi/L may .

be used. I i~

l

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 31 Table 2-9 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS"""

LOWER LIMITS OF DETECTION (LLD)")

Airborne Food Water Particulate or Fish Milk Products Sediment Analysis (pCi/L) Gases (pCi/m') (pCi/kg, wet) (pCi/L) (pCi/kg, wet) (pCi/kg, dry)

Gross Beta 4 0.01 H-3 2000*

Co-60 15 130 Cs-137 18 0.06 150 18 80 180 Table Notations

")

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 l the Annual Radiological Environmental Operating Report.

(M Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13, Revision 1, July 1977.

") The LLD is defined, for purposes of these specifications, as the smallest concentration of l

radioactive material in a sample that will yield a net count, above system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD = 4.66Sb E x V x 2.22 x Y x exp(-At)

For low background counting, the methodology of NUREG-04007 is sed as follows:

' "8 LLD = 3 + 4.66sb E x V x 2.22 x Y x exp(-At)

Where:

LLD =

the "a priori" lower limit of detection as defined above (as pCi per unit mass or volume)

S, =

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

E = the counting efliciency (as counts per transformation)

V = the sample size (in units of mass or volume) 2.22 =

the number of transformations per minute per pico-curie

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1 - 32 Table 2-9 (Continued)

Table Notations (Continued)

Y =

the fractional radiochemical yield (when applicable)

=

A the radioactive decay constant for the particular radionuclide 9GG8l At =

the elapsed time between sample collection (or end of the sample collection i period) and time of counting ;

9GG8 3 =

is a term introduced to account for minor deviation of the Poison distribution from the Normal distribution for low count rates ,

The value of S3 used in the calculation of the LLD for a detection system will be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance.

In calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background will include the typical contributions of other radionuclides normally present in the samples (e.g., potassium 40 in milk samples).

Analyses will be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence ofinterfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable. In such cases, the contributing factors will be identified and described in the Annual Environmental Radiological Operating Report.

Typical values of E, V, Y and t should be used in the calculation. It should be recognized that the LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a nosteriori (afler the fact) limit for a particular measurernent.

For surface water samples, a value of 3000 pCi1 may be used.

1 i

r NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 i TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 33 _

Table 2-10 DISTANCES AND DIRECTIONS TO ENVIRONMENTAL MONITORING STATIONS Radial Direction Radial Distance Station Hy from Plant No. Code Station Name Sector Degrees (Miles)

1 AO King Salmon Picnic Area W 270 0.3

2 A 1742 Wood, Fortuna SSE 158 11.2 3 A0 Humboldt Hill Road at Bret Harte Lane SSE 158 0.9 4 A Wood and K Street, Eureka NNE 42 4.0 5 0 Redwood Avenue, Arcata NE 45 12.3 6 A Table Bluff and Clough Road S 180 5.7 7 A College of the Redwoods S 180 2.6

8. A Humboldt rlill Road near TV Station SSE 170 1.8 9 A 2376 Harbor View Drive SSE 165 1.6 10 A B Street, Fields Landing SSW 200 1.2 11 A Whittier Court & Irving Humboldt Hill SSE 175 1.1 12 A Bell Hill Road and Sauters SSW 195 0.7

14 A South Bay School Parking Lot S 180 0.4 16 AO Elk Piver Road /PG&E Gas Reg /Pedrotti Dairy ENE 72 1.4 17 A Bassinrd Road at Grauer's Lane E 90 2.0 18 A 6418 Elk River Road ESE 112 2.0 19 A 539' Noe /. venue NE 45 1.9 21 AO PGt.E Well 2, HH Road ESE 128 0.5 22 A Station B - 14th Street NNE 23 4.0 24 A PG&E Office,6th and L Street NNE 32 5.0

25 A Irying Drive, ilumboldt 11111 SSE 175 1.3 27 A 6700 Berta Road ESE 125 1.9 28 A 7200 Berta Road SSE 142 2.1 29 A Vista Road. Humboldt Hill SSE 148 1.5 31 A King Salmon Road East of Freeway SSE 170 0.4 32 A Loma Road and Volpis SSW 185 0.5 33 AO 110 kV Line No.1 Well ESE 110 0.1 34 A King Salmon Road and RR Track SSW 185 0.3 36 A Plant Entrance Road WSW 230 0.2 9/24/98 l 45 A Humboldt Substation (T17) ENE 61 5.9 48 0 Holgerson Daity S 180 5.1 55 O HBPP Outfall Canal NNW 338 0.1 56 O 1000 ft North of Outfall Canal Discharu NE 45 0.2 57 0 1000 ft South of Outfall Canal Discharge W 270 0.2 59 O Hookton Chan: -1 SW 225 0.8 I 65 O Coast Ovster Con.pany NNE 23 4.6 I I Table Notations l Code: A Dosimetry Station O Air Particulate Station O Biological Station Note:

Quality Related Station

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 34 Figure 2-1 IIHPP ONSITE TLD LOCATIONS

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NUCLEAR POWER GENERATION NUMBER ODCM l

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1- 39 9/2GB l 2.12 REMP INTERLABORATORY COMPARISON PROGRAM 9/24/98 l 2.12.1 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission.

i APPLICABILITY: At all times.

l I

ACTION: l With analyses not being perform ed as required above, report the corrective actions taken to prevent a recurrence to the Con mission in the Annual Radiological Environmental Monitoring l Repon.

SURVEILLm CE AEQUIREMENTS ;

9/2G8 l 2.12.2 A summary of the results obtained from this program shall be included in the Annual Radiological Environmental Monitoring Report pursuant to Specification 4.1.

l 1

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l NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 40 3.0 SPECIFICATION BASES l l

3.1 Radioactive Liauid Emuent Monitoring _ Instrumentation Basis The radioactive liquid emuent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases ofliquid emuents. The alarm / trip setpoints for these instruments shall be calculated in accordance with Part 11 of the ODCM to ensure that the alarm / trip will occur prior to exceeding 10 times the emuent concentration limits of 10 CFR Part 20 for releases to Humboldt Bay. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60,63 and 64 of Appendix A to 10 CFR Part 50.

3.2 Radioactive Gaseous Emuent Monitoring Instrumentation Basis The radioactive gaseous emuent instrumentation is provided to monitor the releases of radioactive materials in gaseous emuents during actual or potential releases of gaseous emuents from the plant stack. The alarm setpoints for these instruments are calculated in accordance with Part 11 of the ODCM to ensure that the alarm will occur prior to exceeding a radioactive material concentration corresponding to gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY ofless than or equal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60,63, and 64 of Appendix A to 10 CFR Part 50.

3.3 Liauid Emuent Concentration Basis This specification is provided to ensure that the instantaneous concentration of radioactive materials released in liquid waste emuents beyond the SITE BOUNDARY will be less than 10 times the emuent concentration limits specified in 10 CFR Pan 20.

The specification provides operational flexibility for releasing liquid emuents in concentrations to follow the Section ll.A and ll.C design objectives of Appendix 1 to 10 CFR 50. This limitation provides reasonable assurance that the levels of radioactive materials released to Humboldt Bay will result in exposures within (1) the Section ll.A design objectives of Appendix 1,10 CFR 50, to a MEMBER OF THE PUBIC and (2) the limits of 10 CFR 20.1302 to the population. This specification does not affect the requirement to comply with the annual limitations of 10 CFR 20.1301(a).

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1,1 A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 41 SPECIFICATION BASES l

l 3.4 Liauid Emuent Dose Basis This specification is provided to implenx requirements af Sections II.A. lil-A and IV.A of Appendix I,10 CFR Part 50 ~ niting Condition for Operation implements the guides set forth in Section ll.A o.

x 1.. The ACTION statement provides the required operating flexibility and at th, . , 'me implements the guides set forth in l Stetion IV.A of Appendix I to assure that ths leases of radioactive material in liquid emuents will be kept "as low as is reasonably achievable" (ALARA). The dose calculations in the OFFSITE DOSE CALCULATION MANUAL (ODCM) implement the requirements in Section Ill.A of Appendix 1 that conformance with the guides of l Appendix 1 be shown by calculational procedures based on models and data, such that l the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the OFFSITE l DOSE CALCULATIONAL MANUAL (ODCM) for calculating the doses due to the actual release rates of radioactive materials in liquid emuents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Emuents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, " Estimating Aquatic Dispersion of Emuents from Accidental and Routine Reactor Releases for the Purpose ofImplementing Appendix I," April 1977.

3.5 Liquid Waste Treatment Basis The requirement that these systems be used when specified provides assurance that the releases of radioactive materials in liquid emuents will be kept "as low as reasonably achievable" (ALARA). This specification implements the requirements of 10 CFR Part

, 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design j objective given in Section ll.D of Appendix I to 10 CFR Part 50. The specified limits i governing the use of appropriate portions of the liquid radwaste treatment system were I selected as one quarter of the dose design objectives (on a monthly basis) set forth in l Section II.A of Appendix 1,10 CFR Part 50, for liquid emuents (3 mrem /yr; 10 mrem /yr to any organ).

3.6 Gaseous Effluents Dose Rate Basis I

This specification provides reasonable assurance that radioactive material discharged in j gaseous emuents will not result in the exposure of a MEMBER OF THE PUBLIC in an i UNRESTRICTED AREA either within or outside the SITE BOUNDARY in excess of the design objectives of Appendix 1 to 10 CFR 50. The annual dose rate limits are the doses associated with the annual average concentrations of"old" 10 CFR 20, Appendix B, Table II, Column 1. The specification provides operational flexibility for releasing gaseous emuents to satisfy the Section II.A and II.C design objectives of Appendix I to 4

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I 42 SPECIFICATION BASES I 10 CFR 50. For a MEMBER OF THE PUBLIC who may at times be within the SITE

)

i BOUNDARY, the period of occupancy (which is bounded by the maximum occupational period while working in Units 1 or 2) will be sumciently low to compensate for the reduced atmospheric dispersion of gaseous emuents relative to that for the SITE -

. BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin. This specification does not affect the requirement to comply with the annual limitations of 10 CFR. 20.1301(a).

The dose assessment contained in NUREG-1166 has established that neither the routine release of tritium and radioactive particulates with half-lives of greater than 8 days during SAFSTOR nor the occurrence of an analyzed accident during SAFSTOR will exceed the 1500 mrem / year dose rate limit.

The only tritium source term is the spent fuel pool. Assuming a conservative tritium spent fuel pool concentration of 1.0E-4 microcuries/ml, an evaporation rate of 50 gal per day and a ventilation flow rate of 32,000 cfm, the airborne tritium concentration is weil below the required LLD of 1.0E-6 microcuries/ml. Therefore tritium is not sampled in the plant stack efiluent stream.

3.7 Gaseous Efiluents: Noble Gases Dose Basis This Specification is provided to implement the requirements of Sections II.B, Ill.A and IV.A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation implements )

the guides set forth in Section II.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in ,

Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous l efiluents will be kept "as low as reasonably achievable" (ALARA). The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established for calculating the doses due to the actual release rates of radioactive noble gases in gaseous efiluents are consistent with the methodology ;

provided in Regulatory Guide 1.109, " Calculational of Anr.ual Doses to Man from j Routine Releases of Reactor Efiluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1," Revision 1, October 1977 and Regulatory Guide 1.111, 4

" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Efiluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. The luations provided for determining the air doses at and beyond the SITE BOUNDARY

, based upon the historical average atmospheric conditions.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I- 43 l_ SPECIFICATION BASES l

l 3.8 Gaseous EfIluents: Tritium and R. adionuclides in Particulate Form Dose Basis This specification is provided to implement the requirements of Sections ll.C, Ill.A, and IV.A of Appendix 1,10 CFR Pan 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix 1 to assure that the releases of radioactive materials in gaseous effluent will be kept "as low as is reasonably achievable" (ALARA). The calculational methods specified in the Surveillance Requirements implement the requirements in i Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The methods for calculating the dose due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of i Reactor EfIluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, ;

Appendix 1," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for l Estimating Atmospheric Transport and Dispersion of Gaseous EfIluents in Routine ;

Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations J also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for radioactive paniculates with half-life greater than eight days are dependent on the existing radionuclide pathways to man, in areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these calculations were: 1) Individualinhalation of airborne radionuclides, 2) deposition of radionuclides onto green leaf vegetation with subsequent consumption by man,3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) ,

deposition on the ground with subsequent exposure of man.

3.9 Solid Radioactive Waste Basis This Specification ensures that radioactive wastes that are transponed from the site shall meet the solidification requirements specified by the burial ground licensee of the respective states to which the radioactive material will be shipped. It also implements the requirements of 10 CFR Pan 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3.10 Total Dose Basis

- This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from l

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, l A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANU'.L PAGE I - 44 1

SPECIFICATION HASES l

plant radioactive effluents exceed twice the design objective doses of Appendix 1. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For i the purposes of the Special Repon, it may be assumed that the dose commitment to the l

MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Repon with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR part 190.11 and 10 CFR Part 20.2203a4, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Spech. cations 2.3,2.4,2.6,2.7 and 2.8. An individualis not considered a 1 MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is pan of the nuclear fuel cycle.

3.11 REMP Monitorina Program Basis The quality-related portion of the REMP satisfies the requirements in 10 CFR Parts 20 I and 50 that radiological environmental monitoring programs be established to provide data on measurable levels of radiation and radioactive materials in the site environs. It supplements the S AFSTOR Environmental Report baseline environmental conditions by conducting onsite and ofTsite environmental monitoring to evaluate routine conditions during SAFSTOR and to document any increased nuclide concentrations and/or radiation levels resulting from accidents during SAFSTOR.

The non quality-related portion of the HBPP REMP fulfills commitments for environmental monitoring made to the state of California and conducts additional environmental monitoring which PG&E/HBPP has elected to continue from the REMP which was being implemented prior to approval of the SAFSTOR Decommissioning Plan. Normally, non quality-related environmental monitoring (including sample collection and analysis) is conducted in accordance with the programmatic controls established for the quality-related environmental monitoring; however, this monitoring is not subject to the program requirements for radiological environmental monitoring contained in the NRC Radiological Assessment Branch's Branch Technical Position which was issued as Generic Letter 79-65 nor is it subject to the HBPP Decommissioning Quality Assurance Program requirements including adherence to Regulatory Guide 4.15, Quality Assurancefor RadiologicalMonitoring Programs (Normal Operations)-

Effluent Streams and the Environment.

l

g NUCLEAR POWER GENERATION NUMBER ODCM

.HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 45 SPECIFICATION HASES The SAFSTOR Environmental Report, submitted to the NRC as Attachment 6 to the SAFSTOR license amendment request, established baseline conditions for soil, biota and serliments. In accordance with the NRC approved SAFSTOR Decommissioning Plan, these baseline conditions will only need to be reestablished prior to DECON if a l significant release during SAFSTOR occurs as the result of an accident.

l The LLD's required by Table 2-9 are considered optimum for routine environmental l measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141, 3.12 BE)_tP Interlaboratory Comnarison Program Basis The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample rnatrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

i l

l l

l 1

i

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I 46 4.0 ADMINISTRATIVE CONTROLS j MS8 l 4.1 Annual SAFSTOR Radiolouical Environmental Monitorine Report 1

MSa l A report on the SAFSTOR Radiological Environmental Monitoring Program shall be prepared annually in accordance with the NRC Branch Technical Position and submitted to the N~RC within 90 days of January 1 of each year. This report should be included as gg, a separate section to the Annual Radiological Monitoring Report required by Technical Specification Vll.J.l.

was l The Annual Radiological Environmental Report should include:

a. Summaries, interpretations, and an analysis of trends of the results of the quality related Radiological Environmental Monitoring Program activities for the report period.

b. A comparison with the baseline environmental conditions established in the Decommissioning Environmental Report.

c. The results of analysis of quality related environmental samples and of quality related environmental radiation measurements taken during the period pursuant to the locations specified in Table 2-7 summarized and tabulated in the format of Table 4-1, Radiological Environmental Monitoring Program Annual Report Summary, or equivalent. )

d. A summary description of the SAFSTOR Radiological Environmental Monitoring Program.

e. Legible maps covering all sampling locations keyed to a table giving distances and l directions from Unit 3.

1 \

f. The results oflicensee participation in the Interlaboratory Comparison Program and l the corrective action taken if the specified program is not being performed as required in accordance with Specification 2.12.

g. The reason for not conducting the quality related portion of the Radiological l Environmental Monitoring Program as required, and discussion of all deviations from

! the quality related sampling schedule of Table 2-7, including plans for preventing a l

recurrence in accordance with Specification 2.11.

h. A discussion of quality related environmental sample measurements that exceed the reporting levels of Table 2-8, Reporting Levels for Radioactivity Concentrations in Environmental Samples, but are not the result of plant effluents (i.e., demonstrated i by comparison with a control station or the SAFSTOR Environmental Report).

)

y i NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 47 W8 l i. A discussion of all analyses in which the LLD required by Table 2-9 was not achievable.

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1 NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1,1 A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-50 4.2 Annual Radioactive Efiluent Release Report This report shall be submitted prior to April 1 of each year as required by SAFSTOR Technical Specification Vll.J.3. The following information shall be included:

a. A summary of the quantities of radioactive liquid and gaseous efiluents and solid waste released from the plant as outlined in Regulatory Guide 1.21, Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluentsfrom Light-Water-Cooled Nuclear Power Plants, (Rev.1,1974) with data summarized on a quarterly basis i' following the format of Appendix B thereof.

b. For each type of solid waste shipped off-site:

1. Container Volume

2. Total Curie Quantity (specified as measured or estimated)

3. Principal Radionuclides (specified as measured or estimated)

4. Type of Waste (e.g., spent resin, compacted dry waste)

5. Solidification Agent (e.g., cement)

c. A list and description of unplanned releases beyond the SITE BOUNDA.RY.

d. Information on the reasons for inoperability and lack of timely corrective action for i any radioactive liquid or gaseous monitoring instrumentation inoperable for greater than 30 days in accordance with Specifications 2.1 and 2.2.

i

e. A summary description of changes made to:

1. Process Control Program (PCP) 1

2. Radioactive Waste Treatment Systems l

f. A complete, legible copy of the entire ODCM if any change to the ODCM was made

! during the reporting period. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

t I

L

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - 51 4.3 Special Reports The originals of Special Reports shall be submitted to the Document Control Desk with a copy sent to the Regional Administrator, NRC Region IV, within the time period specified for each report. These reports shall be submitted covering the activities identified below to the requirements of the applicable Specification.

a. Radioactive Efiluents - Specifications 2.4, 2.5, 2.7, 2.8 and 2.10, l

b. Radiological Environmental Monitoring - Specification 2.11, l

4.4 Maior Changes to Radioactive Waste Treatment Systems

a. Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid) shall be reported to the NRC in the Annual Radioactive Efiluent Release Report for the period in which the evaluation was reviewed. The changes shall be reviewed and concurred with by the Plant Staff Review Committee and approved by ma the Plant Manager.

i

b. The following information shall be available for review:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59,

2. Suflicient information to totally support the reason for the change, j

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

4. A evaluation of the change which shows the predicted releases of radioactive !

materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously estin,ated in the Environmental Report submitted to the ;

NRC as Attachment 6 to the SAFSTOR license amendment request,

5. An evaluation of the change which shows the expected maximum exposures to an individualin the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the Environmental ' Report,

6. An estimate of the exposure to plant personnel as a result of the change, and

7. Documentation of the fact that the change was reviewed and approved in accordance with plant procedures.

l L

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE , 11-1 PART II- CALCULATIONAL METIIODS AND PARAMETERS 1.0 EFFLUENT MONITOR SETPOINT CALCULATIONS 1.1 LIQUID EFFLUENT MONITORS Specification 2.1 requires that the process water monitor and the caisson sump monitor be set to alarm to ensure that the limits of specification 2.3 are not exceeded (the instantaneous concentration ofradioactive material released to UNRESTRICTED AREAS shall be less than or equal to 10 times the concentrations specified in 10 CFR Part 2 ' Appendix B, Table 2, Column 2).

1.1.1 The alarm setpoint (countrate) for each monitor is calculated as:

r s p3 A x 10 x (ECLc) x K x 0.85 +B

=

(1-1)

(Fi + F 2s _

where:

A =

The alarm setpoint, counts per minute, of the process water monitor or the caisson sump monitor.

Fi = Flow rate past the process water monitor.

=

F2 Flow rate past the caisson sump monitor.

F3 =

Flow rate of the efiluent canal into Humboldt Bay (F1 + F2 + circulating !

water flow - minimum flow with one Unit 1 or Unit 2 circulating water pump in operation is 12,500 gpm).

K =

Calibration factor for the monitor, with units of cpm per micro-Ci/ml.

Baseline calibration of the process water monitor (on 9/20/88) found this factor to be within *l5% of 3.06 x 10' cpm per micro-Ci/ml.

0.85 = Conservatism factor (85 percent of the Specification 2.3 concentration 9/M98 l limits to allow for 15% monitor calibration uncertainty).

l B =

The monitor background reading (prior to any discharge) in counts per minute.

ECLc = Composite EfIluent Concentration Limit (ECL) for the mix of f

radionuclides (micro-Ci/ml).

I l

i l l l i l l 1 j

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 1.1.2 The composite ECL for the mix of radionuclides is calculated as follows:

[ C, [f i ECL c = *

( )

C' f i 1, ECL 3

[ , 'ECL, l.

, where:

ECL;= ECL for radionuclide "i" from 10 CFR 20, Appendix B, Table 2, Column l

2 (micro-Ci/ml).

C; = Concentration of radionuclide "i" in the mixture.

I' f; =

Fraction of radionuclide "i" in the mixture.

l.1.3 Table 2-2 of Spccification 2.1 requires that if a background reading exceeds the I equivalent of 5 x 10 micro-Ci/ml of Cs-137, the cause will be investigated and remedial measures taken to reduce the background reading. Therefore, the maximum background allowable (Bmo, cpm) is:

B,m

= K x (5 x 10~5) cpm (1-3) 1.1.4 The most conservative background limit is calculated as if the calibration factor was 2.60 x 10' cpm per micro-Ci/ml (-15% tolerance). This background limit would be 13,005 cpm. It is plant policy to use a background limit (slightly lower) at 13,000 cpm to ensure that this limit is satisfied. Note that if the background setting exceeds 13,000 cpm, the monitor should be declared INOPERABLE until the background has been reduced.

1.1.5 For continuous direct caisson sump discharges, the monitor should be set to alarm at or below 7.5 times the Cs-137 ECL from 10 CFR 20, Appendix B, Table 2, column 2 (75 percent of the Specification 2.3 limit for Cs-137), assuming no circulating water pump flow and that no liquid radwaste discharge is in progress (i.e., Equation i-1 is solved with Fi = 0 and F3 = F2).

1.1.6 If the Technical Specification alarm setting is calculated for -15% tolerance, and for zero background, the Technical Specification alarm setting would be 26,000 cpm. It is plant policy to maintain the Technical Specification alarm setting at or below 25,000 cpm, to ensure that this Technical Specification limit is satisfied.

Refer to section 1.1.7 for the administrative (lower) alarm settings.

m 1.1.7 For routine liquid radwaste batch discharges, it is plant policy to set the process monitor alarm no higher than necessary in order to provide protection against inadvertent releases. With at least one circulator operating, the alarm should be set according to the following table, and in any case, no higher than 25,000 cpm.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-3 The table is based approximately on the on the sum of twice the typical background' and 130% of the predicted countrate for the batch4 , with the alarm l point rounded up to the next higher meter mark (for more precise settings).'

Table 1-1 Liquid Ellluent Monitor Alarm Setpoints Predicted Process Monitor Alarm Setting Equivalent Cs-137 Concentration Reading (Net cpm) (cpm) (micro-Ci/mi for 600 cpm background)

Up to 2,692 5000 1.4 x 10

2,692 up to 3,461 6000 1.8 x 10

gagg 3,461 up to 4,231 7000 2. I x 10

4,231 up to 5,000 8000 2.4 x 10

5,000 up to 5,769 9000 2.7 x 10

5,769 up to 6,538 10,000 3.1 x 10

6,538 up to 10,385 15,000 4.7 x 10

10,385 up to 14,231 20,000 6.3 x 10*

14.231 up to 18,077 25,000 S.0 x 10

l.2 GASEOUS EFFLUENT MONITOR 1.2.1 Specification 2.2 requires that the Stack Gas Monitoring System be set to alarm to ensure that the limits of specification 2.6 are not exceeded (the dose rate at or beyond the SITE BOUNDARY, due to noble gases released in gaseous efiluents, shall be limited to less than or equal to 500 mrem / year total body and less than or equal to 3000 mrem / year to the skin).

P 1.2.2 Therefore, the alarm setpoiat for this limiting condition is the lesser of Am or Asg calculated for Kr-85. Am is calculated as:

r a 500 x C, x C 2 goes Am = ,s +B (1-4)

Fx 1 xD 3 xK

< <Qs >

and Asg is calculated as 90r./98 'This table is based on a nominal background of 625 cpm. As of 4/15/97, the background reading is about 600 cpm.

The extra 25% provides an allowance related to the uncertainty of reading the background.

d See section 2.4 of ODCM-IV. The 30% tolerance is for a combination of analytical and process nunitor uncertainties 9GM8 and a 10% margin between the ratemeter and chart recorder.

9c4/90 l 5 Each decade is marked at 1,1.5, 2, 2.5, 3, 4, 5, 6, 7, 8 and 9

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-4 e >

3000 x C, x C 2 A 3x =

s +B (1-5)

Fx ,5 x L + (1.1.x M) x K

< <Q) '

s where:

AnT

=

The alarm setpoint, cpm, for the stack noble gas radioactivity monitor, measuring the radioactivity concentration in the stack (prior to release) based on total body dose.

Asg =

The alarm setpoint, cpm, for the stack noble gas radioactivity monitor, measuring the radioactivity concentration in the stack (prior to release) based on skin dose.

500 =

Whole body dose limit in mrem / year.

3000 = Skin dose limit in mrem / year.

4 Ci =

Conversion factor,10 micro-Ci/pico-Ci.

C2 = Conversion factor,104 m'/cc.

F = The flowrate, cubic meters per second, of the Unit No. 3 ventilation system discharge to the stack. This parameter is nominally 14.6 cubic meters per second (31,000 cfm) for the 250 foot stack and i1.8 cubic SU"8 meters per second (25,000 cfm) for the 50 foot stack.

- The " instantaneous" atmospheric dispersion parameter, seconds per cubic Q

meter.

= 4 1.46 x 10 seconds / cubic meter for releases from the 250 foot stack.

Taken from Appendix B.

= 4

,, 6.52 x 10 seconds / cubic meter for releases from the 50 foot stack at a f?cw rate of 25,000 cfm. Taken from Table 1 of Calculation N-238c,

" Calculate the Effect of HBPP Unit 3 Stack Reconfiguration on Safstor Routine Radioactive E91uents Program and Accident Dose Analysis."

=

9G4/98 l Dra The total body dose factor due to gamma exposure in a semi infinite cloud, mrem / year per pico Curie / cubic meter. The value of this parameter 4

is given in Table B-1 of Regulatory Guide 1.109 as 1.61 x 10 for Kr-85.

9GM8 l L =

The skin dose factor due to beta exposure in a semi infinite cloud, mrem / year per pico Curie / cubic meter. The value of this parameter is 4

given in Table B-1 of Regulatorf Guide 1.109 as 1.34 x 10 for Kr-85.

L . . - ----

! NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 I TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-5 9/2M8 l M =

The air dose factor due to gamma exposure in a semi infmite cloud, mrad / year per pico Curie / cubic meter. The value of this parameter is 4

given in Table B-1 of Regulatory Guide 1.109 as 1.72 x 10 for Kr-85.

The associated factor of 1.1 is a unit conversica from mrad to mrem.

K =

Calibration factor for the monitor. As discussed in Appendix C, the calibration factor is 3.1 x 10'" micro-Ci/cc per cpm.

B =

The monitor background reading due to ambient background radiation and natural radioactive noble gasses, cpm. This parameter is generally not significant, since the typical reading is 20

10 cpm.

1.2.3 Using the parameters above, the alarm point for the stack monitors should be set 9/2e8 at or below 33,400 cpm for the 250 foot stack and 9,200 cpm for the 50 foot stack. It is plant policy to set it at 1,000 cpm. Note that changes to these values affect EPIP R-6 (Volume 3), EDOI H-11 (Volume 2) and STP 3.16 4 (Volume 6).

l

r NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

[ REVlSION 1, I A & 2 ;

l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-6 i 2.0 LIQUID EFFLUENT DOSE CALCULATIONS

' 2.1 CALENDAR QUARTER l

l 9tae3 l Specification 2.4.1.a requires that the dose or dose commitment to a MEMBER OF THE l

PUBLIC from radioactive materials in liquid emuents released to UNRESTRICTED AREAS during any calendar quarter shall be limited to less than or equal to 1.5 mrem to the total body and less than or equal to 5 mrem to any organ .

9/2M8 l 2.1.1 Compliance with Specification 2.4.1.a has been established on a licensing basis by the Environmental Report submitted to the NRC as Attachment 6 to the SAFSTOR licensing amendment request and NUREG-1166, Fina/ Environmental Statementfor Decommissioning Humboldt Bay Power Plant, Unit No. 3, issued by the NRC .

2.1.2 These reports have demonstrated that neither the routine release of radioactive materials in liquid emuents duiing SAFSTOR nor the occurrence of an analyzed accident during SAFSTOR would exceed the dose specification of Specification gaw8l 2.4.1.a.

2.1.3 Therefore, calculation of dose due to the release of radioactive materials in liquid emuents during any calendar quarter is not necessary.

2.1.4 IF a comparison performed at least once per 31 days indicates that the activity due to the release of radioactive materials in liquid emuents will exceed the Environmental Report baseline release for the current calendar quarter, THEN a dose calculation for the current calendar quarter shall be performed.

2.2 CALENDAR YEAR 904/98 l Specification 2.4.1.b requires that the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid emuents released to UNRESTRICTED AREAS during any calendar year shall be limited to less than or equal to 3 mrem to the total body and less than or equal to 10 mrem to any organ .

9G4/98 l 2.2.1 Compliance with Specification 2.4.1.b has been established on a licensing basis by the Environmental Report and NUREG-1166.

l l

2.2.2 These reports have demonstrated that the routine release of radioactive materials in liquid emuents during SAFSTOR will not exceed the dose specification of i

9G4/98 l Specification 2.4.1.b.

2.2.3 Therefore, calculation of dose during any calendar year due to the routine release I of radioactive materials in lignid emuents is not necessary.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-7 l 2.2.4 IF a comparison ir.dicates that the activity due to the release of radioactive l

materials in liquid emuents will exceed the Environmental Report baseline release for the current calendar year, THEN a dose calculation for the current calendar year shall be performed.

2.3 LIQUID EFFLUENT DOSE CALCULATION METHODOLOGY l

The dose contribution to the total body and each individual organ (bone, liver, kidney, lung and GI-LLI) of the maximum and average exposed individual (adult, teen, child, and infant) will be calculated for the nuclides detected in emuents. The dose to an organ of an individual from the release of a mixture of radionuclides will be calculated as follows:

D=[ Ci x DF x {(Brisa.i x Ura)+(Bav.i x Umv)}.

,-i (2-1) where:

D =

The dose commitment, mrem per year, to an organ (or to the whole body) due to consumption of aquatic foods.

Ci =

The average diluted emuent concentration, pico-Curie / liter, for radionuclide, i. This will be estimated by dividing the total activity of the nuclide discharged during the quarter, pico-Curies, by the total citeulating water discharge flow during the quarter, liters. Note that the resulting dose commitment is the annual dose for the case of four quarters with this average concentration.

DF -

The dose conversion factor, mrem /pico-Curie for the nuclide, organ, and age group being calculated. This factor is taken from Tables 2-1,2-2, and 2-3.

BFish,i =

The bioaccumulation factor, pico-Curie / kilogram per pico-Curie / liter, in fish for the radionuclide in question. This value is taken from Table 2-4.

BInv.i = The bioaccumulation factor, pico-Curie / kilogram per pico-Curie / liter, in invertebrates for the radionuclide in question. This value is taken from Table 2-4.

UFish = Usage factor (consumption) of fish, kilogram / year, for the age group and individual (average or maximum)in question. This factor is derived from Table 2-5 or 2-6.

Utnv = Usage factor ofinvertebrates, kilogram / year, for the applicable age group and individual (average or maximum). This factor is from Table 2-5 or 2-6.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11 8 The total exposure to an organ (or whole body) is found from the summation of the contributions of each of the individual nuclides calculated. Note that the infant age group is not considered to consume either fish or other seafc,ed, and exposure to this age group need therefore not be calculated.

Table 2-1 3 Ingestion Dose Factors for Adult Age Group (mrem /pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-11 i Organ Nuclide Bone Liver Total Body ' Kidney Lung GI-LLI H-3 No Data 1.05 x 10-7 1.05 x 10~ / 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 Co-60 No Data 2.14 x 10-0 4.72 x 1C-6 No Data No Data 4.02 x 10-3 Sr-90 7.58 x 10-3 No Data IE x 10-3 No Data No Data 2.19 x 10-4 Cs-137 7.97 x 10-5 1.09 x 10-4 7.14 x 10-5 3.70 x 10-5 1.23 x 10-3 2.11 x 10-6 Y-90 9.62 x 10-9 No Data 2.58 x 10- No Data No Data 1.02 x 10-4 10 Table 2-2 Ingestion Dose Factors for Teen Age Group )

(mrem /pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-12 Organ l Nuclide Bone Liver Total Body Kidney Lung GI-LLI !

H-3 No Data 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 !

Co-60 No Data 2.81 x 10-6 6.33 x 10-0 No Data No Data 3.66 x 10-5 I Sr-90 8.30 x 10-3 No Data 2.05 x 10-3 No Data No Data 2.33 x 10-4 Cs-137 1.12 x 10-4 1.49 x 10-4 5.19 x 10-5 5.07 x 10-5 1.97 x 10-3 2.12 x 10-6 Y-90 1.37 x 10-6 No Data 3.69 x 10- No Data No Data 1.13 x 10-4 10 Table 2-3 Ingestion Dose Factors for Child Age Group <

(mrem /pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-13 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI II-3 No Data 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 Co-60 No Data 5.29 x 10-6 1.56 x 10-5 No Data No Data 2.93 x 10-3 Sr-90 1.70 x 104 No Data 4.31 x 10-3 No Data No Data 2.29 x 10-4 Cs-137 3.27 x 10-4 3.13 x 10-4 4.62 x 10-3 1.02 x 10-4 3.67 x 10-3 1.96 x 10-6 Y-90 4.11 x 10-5 No Data 1.10 x IOC No Data No Data 1.17 x 10-4

NUCLEAR POWER GENERATION WUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-9 I

Table 2-4 l Bioaccumulation Factors for Saltwater Environment (pCi/kg per pCi/ liter) l Selected Nuclides from Regulatory Guide 1.109, Table A-1 l Element Fish Invertebrate H 9.0 x 10-1 9.3 x 10-1 Co 1.0 x 102 1.0 x 103 Sr 2.0 2.0 x 101 "

Cs 4.0 x 101 2.5 x 101 Y 2.5 x 101 1.0 x 103 i

Table 2-5 Average Individual Foods Consumption for Various Age Groups (kilo-gram / year or liters / year) sna8l From Regulatory Guide 1.109, Table E-4 Other Seafood Fruits and Age Group Fish (Invertebrates) Vegetables Milk Meat Adult 6.9 1.0 190 110 95 Teen 5.2 0.75 240 200 59 Child 2.2 0.33 200 170 37 Infant 0 0 0 0 0

Table 2-6 Maximum Individual Foods Consumption for Various Age Groups (kilo-gram / year or liters / year) 9GM8l From Regulatory Guide 1.109. Table E-5 Other Seafood Fruits and Age Group Fish (Invertebrates) Vegetables Milk Meat Adult 21 5.0 520 310 110 Teen 16 3.8 630 400 65 Child Infant 6.9 0

]70 520 0

330 330 41 0

l

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 0 l

! 3.0 LIQUID WASTE TREATMENT l

3.1 TREATMENT REQUIREMENTS l 3.1.1 ODCM Specification 2.5 Specification 2.5 requires that liquid radwaste shall be treated, as required, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to liquid efiluents discharged to UNRESTRICTED AREAS would exceed 0.06 mrem whole bociy or 0.2 mrem to any organ. ;

3.1.2 NPDES Waste Discharge Requirement NPDES Permit No. CA0005622, issued by the California Regional Water Qnlity 1 Control Board - North Coast Region, requires that the discharge ofliquid wastes "shall not cause bottom deposits in the receiving waters." The permit also identifies i

Discharge Serial No. 001E (liquid low level radioactive waste) that indicates that the M88 waste may be treated prior to discharge. The permit does not mandate treatment.

3.2 TREATMENT CAPABILITIES J

9tama l 3.2.1 Liquid Waste Collection System Liquid waste is collected in either the turbine building drain tank (TBDT), reactor equipment drain tank (REDT), reactor caisson sump or radwaste building sump.

a. Turbine Building Drain Tank The TBDT, turbine building ficor drain pump and TBDT pumps are located at elevation -14 feet in the reactor caisson in a shielded vault beneath the new fuel storage vault. The contents of the 3,000 gallon capacity tank may be pumped to l a radwaste receiver tank or drained to the REDT via the caisson floor drain I system.

b. Reactor Equipment Drain Tank The REDT and associated REDT pumps aie located at the -66 foot level of the i reactor caisson access shafl. The contents of this 500 gallon capacity tank are pumped automatically to t'ne radwaste treatment system using either of the two REDT pumps.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11 11

c. Reactor Caisson Sump The reactor caisson sump and its associated reactor caisson sump pumps are located at the -66 foot level of the access shaft. The sump, which collects onesl groundwater in-leakage, has a capacity of 50 gallons. The pump may transfer its contents automatically through a liquid efIluent monitor to the Discharge Canal, 9n28l or may be valved to the radwaste treatment system if necessary for compliance with Specification 2.5 due to groundwater contamination.

d. Radwaste Building Sump The radwaste building sump tank, with a capacity of 250 gallons, is located beneath the radwaste building floos and receives liquids from drains in the vicinity of the radwaste building. The sump pump is located on the operating floor of the radwaste building (elevation +12 feet) over the sump tank. This pump automatically maintains the level of the tank and discharges to one of the waste receiver tanks.

3.2.2 Liquid Waste Treatment System The liquid waste treatment system processes, stores and provides for disposal of radioactively contaminated wastes and other liquid wastes that are potentially radioactively contaminated. These wastes are first collected by the radwaste collection system and are then pumped to the radwaste building on the nonh side of the refueling building. The major components of the liquid waste treatment system which are available for use to comply with Specification 2.5 include the:

waste receiver tanks (3) e radwaste concentrator e radwaste demineralizer e resin disposal tank e concentrated waste tanks (2) e waste hold tanks (2) e radwaste filters (2) e concentrator dnp receiver tank

a. Waste Receiver and Waste Hold Tanks The three 7,500 gallon carbon steel n dwaste receiver tank zie for wastes coming from the radwaste collection system. Two 7.500 carbon steel waste hold tanks are for storing treated wastes for retreatment or disposal. The tanks are located in an external section of the radwaste building, but are within the prefabricated steel radwaste enclosure.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 s

REVISION 1, I A & 2

TITLE SAFSTOR OFFSITE DOSE CALCUL ATION MANUAL PAGE 11-1 2 l

b. Radwaste Concentrator l The radwaste concentrator was designed to concentrate 7,500 gallons per week.

l' The concentrator consists of a vessel about 14 feet high and 24 inches in l diameter with a 40 square foot, callandria-type evaporating section near the l bottom. Steam from the Unit 1 or Unit 2 auxiliary steam system is fed to the j callandria outside of the tubes. Evaporation takes place within the tubes. The l l concentrator is located in a shielded cubicle in the radwaste building. l l

i Concentrator vapor goes to a condenser which is cooled with water from an independent cooling loop, and the condensate goes to the drip receiver tank for collection for further treatment or disposal. The concentrated radwaste is discharged to one of the two concentrated waste storage tanks. ,

Concentration by evaporation is generally the most appropaate method for l treatment ofliquids containing high total dissolved solids (TDS).

l

c. Radwaste Demineralizer The radwaste demineralizer is a single, mixed bed unit with a nominal flow of 20 I sms gpm and a flow capacity of 50 gpm. The demineralizer tank is 24 inches in diameter and was designed for 75 psig in accordance with the ASME Code.

There are no provisions for regeneration, spent resins are sluiced to the resin disposal tank. The demineralizer is located in a shielded cubicle in the radwaste building.

Demineralization is generally not an appropriate method to treat high TDS liquids.

d. Resin Disposal Tank This 10,000 gallon tank is located in an individual shielded vault within the radwaste building. It is accessed through a hatch in the top of the vault. All spent resins from the various demineralizers on site are routed to this tank. i

e. Concentrated Waste Tanks Two 5,000 gallon storage tanks are located in a shielded vault in the radwaste building. These tanks receive concentrated wastes from the concentrator. These tanks have no inherent means for draining and must be pumped down through access ports in the top of the tank.

I' NUCLEAR POWER GENERATION NUMBUR ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 3

f. Radwaste Filters Two radwaste filters are available in the radwaste building. These are cartridge-type filters which can remove particles down to 25 microns in diameter.

g. Concentrator Drip Receiver Tank l

l A concentrator drip receiver tank is provided to collect the condensed vapors

! from the concentrator. The concentrator drip receiver pump either recirculates l

water in the tank for sample mixing purposes, or it discharges to the treated <

waste pump discharge header for final disposition. I l

3.2.3 Mobile Liquid Waste Treatment Systems Various mobile liquid waste treatment systems are available from vendors for use if necessary. These include systems such as high pressure filtration, demineralization, reverse osmosis and solidification.

Mobile liquid waste treatment systems are available for trc. at of both high and j low TDS liquids.

I l

l I

l_

l l

(

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 l REVibON 1, I A & 2 l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-14 L 4.0 GASEOUS EFFLUENT DOSE CALCULATIONS l

4.1 DOSE RATE 4.1.1 Noble Gases l 9GMBl Specification 2.6.1.a requires that the dose rate at or beyond the SITE BOUNDARY, due to noble gases released in gaseous emuents, shall be limited to less than or equal to 500 mrem / year total body and less than or equal to 3000 mrem / year to the skin.

9GG8l a. Compliance with Specification 2.6.1.a has been established on a licensing basis by the Environmental Report and NUREG-1166.

b. These repons have demonstrated that neither the routine release of noble gases during SAFSTOR nor the occurrence of an analyzed accident involving spent fuel assemblies during SAFSTOR would exceed the dose rate specification of gna8l Specification 2.6.1.a.

l

c. Therefore, further methodology for the determination of dose rate due to noble gases is not necessary.

4.1.2 Tritium and Radioactive Particulates 9GS8l Specification 2.6.1.b requires that the dose rate at or beyond the SITE BOUNDARY, due to tritium and radioactive particulates with half-lives of greater than 8 days released in gaseous emuents, shall be limited to less than or equal to 1500 mrem / year to any organ.

Sams l a. Compliance with Specification 2.6.1.b has been established on a licensing basis ]

by the baseline gaseous emuent releases established in the Environmental Report and dose assessment contained in NUREG-1166.

b. These reports have demonstrated that neither the routine release of tritium and radioactive particulates with half-lives of greater than 8 days during SAFSTOR sras nor the occurrence of an analyzed accident during SAFSTOR will exceed the 90G8 dose rate specification of Specification 2.6.1.b. '

c. Therefore, funher methodology for the determination of dose rate due tritium l

and radioactive paniculates with half-lives of greater than 8 days released in gaseous emuents is not necessary.

t

{

l

[

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 5 4.2 DOSE - NOBLE GASES 4.2.1 Calendar Quarter 9Ga8l Specification 2.7.1.a requires that the air dose in UNRESTRICTED AREAS during any calendar quarter due to radioactive noble gases released in gaseous emuents shall be limited to less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation.

9GG8l a. Compliance with Specification 2.7.1.a has been established on a licensing basis by the Environmental Report and NUREG-1166.

b. These reports have demonstrated that the routine release of noble gases in gaseous emuents during SAFSTOR will not exceed the dose specification of 9Ga8l Specification 2.7.1.a.

c. Therefore, calculation of dose during any calendar quarter due to radioactive noble gases released in gaseous emuents is not necessary for the routine release of noble gases during SAFSTOR.

d. IF a comparison performed following an accident involving spent fuel indicates that the noble gases released in gaseous emuents will exceed the Environmental Report baseline release for the current calendar quarter, TIIEN a dose calculation for the current calendar quarter shall be performed.

4.2.2 Calendar Year 90#8l Specification 2.7.1.b requires that the air dose in UNRESTRICTED AREAS during any calendar year due to radioactive noble gases released in gaseous emuents shall be limited to less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation. ;

9GM8l a. Compliance with Specification 2.7.1.b has been established on a licensing basis by the Environmental Report and NUREG-1166.

b. These reports have demonstrated that the routine release of noble gases in gaseous emuents during SAFSTOR will not exceed the dose specification of IG4/98 l Specification 2.7.1.b.

c. Therefore, calculation of dose during any calendar year due to radioactive noble I gases released in gaseous emuents is not necessary for the routine release of noble gases during SAFSTOR.

\

d. IF a comparison performed following an accident involving spent fuelindicates that the radioactive noble gases released in gaseous emuents will exceed the

1.

L NUCLEAR POWER GENERATION NUMBER ODCM l HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 ll TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 6

, Environmental Report baseline release for the current calendar year, THEN a

! dose calculation for the currert calendar year shall be performed.

4.2.3 Noble Gas Dose Calculation Methodology .

! Both dose to the whole body (gamma dose) and dose to the skin (beta dose) due to the release of radioactive noble gas efiluents are calculated. However, due to the decay time since last operation, Kr-85 is the only radioactive noble gas that remains in the fuel. The equations for calculating the maximum hypothetical radiation exposure at an offsite location are as follows:

Dws = Q x (y/Q) x K (4-1)

Ds = Q x (y/Q) x 'L + (1.1 x M)] (4-2) where:

Dwa = Whole body (gamma) dose, mrem / year.

Ds =

Skin (beta + gamma) dose, mrem / year x/Q =

The atmospheric dispersion parameter, seconds per cubic meter.

= 4 1.4 x 10 seconds per cubic meter for releases from the 250 stack.

Calculated based on historical meteorological monitoring data for comparison of the plant with 10 CFR 50, Appendix I In that 6

calculation, the largest vahie of X/Q was used M"8 = 1.0 x 10-5 seconds per cubic meter for releases from the 50 foot stack with a flowrate of 25,000 cfm. Taken from Table 1 of Calculation N-238c, " Calculate the Effect of HBPP Unit 3 Stack Reconfiguration on Safstor Routine Radioactive Effluents Program and . Accident Dose Analysis."

9GG8l K = The total body dose factor due to gamma exposure in a semi-infinite cloud, mrem / year per pico-Curie / cubic meter. The value of this parameter is given in Table B-1 of Regulatory Guide 1.109 as 1.61 x 10-5 for Kr-35.

9GM8l L =

The skin dose factor due to beta exposure in a semi-infinite cloud,

mrem / year per pico-Curie / cubic meter. The value of this parameter is given in Table B-1 of Regulatory Guide 1.109 as 1.34 x 10-3 for Kr-85.

f

' Taken from Table A-6 of" Analysis of Compliance with 10 CFR 50, Appendix 1, for the Humboldt Bay Unit 3 Nuclear Unit," November 1976. This report was prepared, but not submitted. RMS 3633-3134.

I

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 7 l 9G#8l M =

The air dose factor due to gamma exposure in a semi-infinite cloud, mrad / year per pico-Curie / cubic meter. The value of this parameter is given in Table B-1 of Regulatory Guide 1.109 as 1.72 x 10-5 for Kr-85. The associated factor of 1.1 is a unit conversion from mrad to mrem.

Sm98 l Q =:

The average release rate of Kr-85 in gaseous releases, l p;co-Curies /sec.

Note that this is the exposure to a hypothetical individual continuously located at the maximum ground level exposure location.

4.3 DOSE - TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM 4.3.1 Calendar Quarter l sn4/98 l Specification 2.8.1.a requires that the organ dose to a MEMBER OF THE PUBLIC from the release of tritium and radioactive materials in particulate form with half.

lives greater than 8 days in gaseous emuents released to UNRESTRICTED AREAS shall be limited to less than or equal to 7.5 mrem during any calendar quarter.

904/98 l a. Compliance with Specification 2.8.1.a has been established on a licensing basis by the Environmental Report and NUREG-1166.

b. These reports have demonstrated that the routine release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous emuents during SAFSTOR will not exceed the dose specification of l 904/98 l Specification 2.8.1.a.

c. Therefore, calculation of dose during any calendar quarter due to tritium and

! radioactive materials in particulate form with half-lives greater than 8 days

- released in gaseous emuents is not necessary for the routine release of noble gases during SAFSTOR.

d. IF a comparison performed at least once per 31 days indicates that the tritium and radioactive materials in particulate form with half-lives greater than 8 days

released in gaseous emuents will exceed the Environmental Report baseline

release for the current calendar quarter, THEN a dose calculation for the current l calendar quarter shall be performed.

l 4.3.2 Calendar Year

, gn4;98 l Specification 2.8.1.b requires that the organ dose to a MEMBER OF THE PUBLIC from the release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous emuents released to UNRESTRICTED AREAS shall be limited to less than or equal to 15 mrem during any calendar year.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 8 9/24/98 l a. Compliance with Specification 2.8.1.b has been established on a licensing basis by the Environmental Report and NUREG-1166.

b. These reports have demonstrated that the routine release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous emuents during SAFSTOR will not exceed the dose specification of 9/34/98 l Specification 2.8. l.b.

c. . Therefore, calculation of dose during any calendar year due in tritium and radioactive materials in particulate form with half-lives greater than 8 days released in gaseous emuents is not necessary for the routine release of noble gases during SAFSTOR.

d. IF a comparison indicates that the tritium and radioactive materials in particulate form with half-lives greater than 8 days released in gaseous emuents will exceed the Environmental Report baseline release for the current calendar year, THEN a dose calculation for the current calendar year shall be performed. ]

4.3.3 Particulate Organ Dose Calculation Summation Methodology The releases of radioactive materials in particulate form with half-lives greater than !

8 days in gaseous emuents will be essentially limiad to Cs 137, Co-60, and Sr-90.

The annual dose commitment will be calculated for any organ of an individual age group as follows:

n D=[ ,Q, x (Rai + Re.i + Rm.u + Rm.i + Rv.s.i),

i-i (4-3) where:

D =

Annual dose commitment, mrem / year.

=

Qi The average release rate of the nuclide in question, pico- '

l Curies /second.

RInh,i =

The dose factor for the inhalation pathway for the radionuclide, i, in units of mrem / year per pico-Curie /sec.

rop,i = The dose factor for the ground plane (direct exposure from deposition) pathway for the radionuclide, i, in units of mrem / year per pico-Curie /sec.

RMeat.i = The dose factor for the grass-cow-meat pathway for the radionuclide, l

i, in units of mrem / year per pico-Curie /sec.

l 1

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 i- TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-19 RMilk,i =

The dose factor for the grass-cow-milk pathway for the radionuclide, i, in units of mrem / year per pico-Curie /sec.

Rveg,i =

The dose factor for the pathway of deposition on vegetation for the radionuclide, i, in units of mrem / year per pico-Curie /sec . l l

l In general, the calculations for these pathways give results that represent trivial radiation exposure. The values calculated for typical anticipated SAFSTOR releases range from about 0.002 mrem / year (fruit / vegetable consumption pathway) to less

~

l l than 1 x 104 mrem / year (for direct radiation exposure from material deposited on l the ground).

l 4.3.4 Particulate Inhalation Pathway Dose Calculation Methodology I Ria.i = (y/Q) x BR. x DFu. (4-3a) l where:

I =

9ar.198 X/Q The atmospheric dispersion parameter, seconds / cubic meter. j l =

1.4 x 104 seconds / cubic meter for releases from the 250 foot stack. I

= 4

, 1.0 x 10 seconds / cubic meter for releases from the 50 foot stack.

l l bra =

The breathing rate of the receptor age group (a), cubic meters per year. The values to be used are 1400,3700,8000, and 8000 cubic meters / year for the infant, child, teen, and adult age groups, respectively.

DFi,a =

The organ (or total boo)) inhalation dose factor, mrem /pico-Curie, for the receptor age group, r., for the radionuclide, i. The dose l factors are given in Tables 4-1,4-2,4-3, and 4-4.

4.3.5 Particulate Ground Plane Pathway Dose Calculation Methodology Roe.i = (D/Q) x SF x DF4 x K x W (4-3b) 1 where:

K = unit conversion constant,8760 hr/yr.

DFi = The ground plane dose conversion factor for radionuclide, i, in mrem /hr per pCi/m2 from Table 4-5.

SF = The shielding factor (dimensionless). Table E-15 of Regulatory unas l Guide 1.109 suggests values of 0.7 for the maximum individual.

l

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE II-20

=

D/Q The atmospheric deposition factor, with units ofinverse square meters.

= 3.0 x 10* inverse square meters for releases from the 250 foot stack.

Calculated based on historical meteorological monitoring data for comparison of the plant with 10 CFR 50, Appendix 1. In that gg, calculation, the largest value of D/Q was used7.

= 4 3.0 x 10 inverse square meters for releases from the 50 foot stack.

Taken from Table 1 of Calculation N-238c," Calculate the Effect of HBPP Unit 3 Stack Reconfiguration on Safstor Routine Radioactive Efiluents Program and Accident Dose. Analysis."

W =

Weathering factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.109, for a 14 day removal half-life. In this equation, W has the value of 1.74 x 106 seconds.

4.3.6 Particulate Grass-Cow-Milk Pathway Dose Calculation Methodology Rwu = (D/Q) x < Y >

(4-3c) where:

=

QF The cow's vegetation consumption rate. This is given as 50 kg/ day per Regulatory Guide 1.109, ' Table E-3.

=

Ua The receptor's milk consumption rate, liters / year for the age group in question. See Tables 4-6 and 4 7 Y =

The agricultural productivity by unit area of pasture. This parameter j is 0.7 kg/m2 per Regulatory Guide 1.109, Table E-15.

DFi,a =

The ingestion dose factor for radionuclide, i, for the receptor in age group (a), in units of mrem /pico-Curie, from Tables 4-8,4-9,4-10, or 4-11. i Fm =

The fraction of the cow's intake of a nuclide which appears in a liter of milk, with units of days / liter. This parameter is given by Table 4-12.

D/Q

=

The atmospheric deposition factor, with units ofinverse square meters.

8N8 = 3.0 x 10-9 inverse square meters for releases from the 250 foot stack.

4

= 3.0 x 10 inverse square meters for releases from the 50 foot stack.

' Taken from Table A-6 of" Analysis of Compliance with 10 CFR 50, Appendix 1, for the Humboldt Bay Unit 3 Nuclear Unit," November 1976. This report was prepared but not submitted. RMS 3633-3134. ,

I

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 1 W =

Weathering factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.109, for a 14 day removal half-life. In this equation, W has the value of1.74 x 10 6seconds.

4.3.7 Particulate Grass-Cow-Meat Pathway Dose Calculation Methodology

'Qr x U. x Fr x DFi.. x W' Ru.i = (D/Q) x s Y >

(4-3d) where:

= The cow's vegetation consumption rate of 50 kg/ day per Regulatory QF Guide 1.109, Table E-3.

Un =

The receptor's meat consumption rate, kilogram / year. Refer to Tables 4-5 and 4-7.

Y =

The agricultural productivity by unit area of pasture. This parameter is 0.7 kg/m2 per Regulatory Guide 1.109, Table E-15.

DFi,a =

The ingestion dose factor for radionuclide, i, for the receptor in age group (a), in mrem /pCi, from Tables 4-8,4-9, or 4-10, as appropriate. Note that this path is not considered to apply to the infant age group.

Fr =

The fraction of the animal's intake of a nuclide which finally appears in meat, days / kilogram. This parameter is given in Table 4-13.

=

D/Q The atmospheric deposition factor, with units ofinverse square meters.

= 3.0 x 10-9 inverse square meters for releases from the 250 foot stack.

"8 = 3.0 x 10 inverse square meters for releases from the 50 foot stack.

W = Weathering factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.109, for a 14 day removal half-life. In this equation, W has the value of 1.74 x 106seconds.

4.3.8 Particulate Vegetation Pathway Dose Calculation Methodology

' Ur x DFi.. x W' Rw. s = (D/Q) x

< Y >

(4-3 e) f where:

l t

E NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-22 l

! UT =

The total consumption rate of fmits and vegetables, kilogram / year.

l This parameter is determined with the default values from Regulatory l

Guide 1.109, as reproduced in Tables 4-6 and 4-7.

=

l D/Q The atmospheric deposition factor, with units ofinverse square i meters.

9/24 S 8 = 3.0 x 10-9 inverse square meters for releases from the 250 foot stack.

l =

3.0 x 10-' inverse square meters for releases from the 50 foot stack.

W =

Weathering factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.109, for a 14 day removal half-life. In this equation, W has the value of 1.74 x 106seconds.

Y =

The agricultural productivity by unit area of pasture. This parameter is 0.7 kg/m2per Regulatory Guide 1.109, Table E-15.

Note: this equation probably overestimates exposures, since it assumes that all of

. the deposition on a plant remains on the plant, while the Regulatory Guide allows a factor of 0.25. Also, the quantities assumed consumed include grain (none is grown in the vicinity of the plant), as well as vegetables and fruit grown in other areas (imported to Humboldt county).

4.3.9 Tritium Organ Dose Calculation Methodology The annual dose commitment may be calculated for any organ of an individual age group as follows:

D = Qto x (Ra m + Ror.m + Rw.t m + Rm.m + Rv.5 m) (4-4) where:

D = Annual dose commitment, mrem / year.

Q113

= The average release rate of H-3, pico-Curies /second.

Rinh,II3 = The dose factor for the inhalation pathway for H-3, mrem / year per pico-Curie /sec.

RMeat,113 = The dose factor for the grass-cow-meat pathway for H-3, mrem / year i per pico-Curie /sec.

l- RMilt,113 = The dose factor for the grass-cow-milk pathway for h-3, mrem / year per pico-Curie /sec.

Rveg.113 = The dose factor for the vegetation consumption pathway, mrem / year per picc Curie /sec.

l-

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 l TITLE SAFSTOR OFFSITE DOSE CAL.CULATION MANUAL PAGE 11-2 3 This pathway results in trivial offsite calculated radiation exposures. A very I

conservative assumption of Tritium release is that Spent Fuel Pool water at 1 x 10-2 micro-Curies /ml H-3 is lost to the stack at a rate of 50 gallons / day. With this assumption, the calculated maximum offsite exposure is 0.0013 mrem / year.

4.3.10 Tritium Inhalation Pathway Dose Calculation Methodology Rina.n3 =

( x BR. x DFiu.. (4-4a)

I where:

=

X/Q The atmospheric dispersion parameter, seconds / cubic meter.

=

9/24/98 1.4 x 10-6 seconds / cubic meter for releases from the 250 foot stack.

= 4 1.0 x 10 seconds / cubic meter for releases from the 50 foot stack.

bra =

The breathing rate of the receptor age group (a), cubic meters per year. The values to be used are 1400,3700,8000, and 8000 cubic l meters / year for the infant, child, teen, and adult age groups, respectively.

DFlu,a =

The organ (or total body) inhalation dose factor for the receptor age group, a, for H-3. This is given in units of mrem /pico-Curie by Tables 4-1,4-2,4-3, and 4-4.

4.3.11 Tritium Grass-Cow-Milk Pathway Dose Calculation Methodology The concentration of tritium in milk is based on the airborne concentration rather than the deposition:

/ 'O'75 x 0'5' Ruau.it3 = x x Qg x U. x Fm x DF. (4-4b)

N < H >

wher'e:

= The cow's vegetation consumption rate. This is 50 kg/ day per QF Regulatory Guide 1.109, Table E-3.

Ua = The receptor's milk consumption rate for age group, a, from Regulatory Guide 1.109. See Tables 4-6 or 4-7.

i DF. =

The ingestion dose factor for H-3, for the reference group, mrem /pico-Curie, from Tables 4-8,4-9,4-10, and 4-11.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2

_ TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-24 Fm =

The fraction of the cow's intake of a nuclide which appears in a liter of milk, with units of days / liter. This parameter is given by Table 4-12.

0.75 =

The fraction of total feed that is water.

0.5 =

The ratio ofspecific activity of the feed grass to the atmospheric water.

H =

Absolute humidity of the atmosphere,0.008 kilograms / cubic meter, according to Regulatory Guide 1.109.

=

X/Q The atmospheric dispersion parameter, seconds / cubic meter.

=

9/24/98 1.4 x 10-6 seconds / cubic meter for releases from the 250 foot stack.

=

1.0 x 10'5 seconds / cubic meter for releases from the 50 foot stack.

4.3.12 Tritium Grass-Cow-Meat Pathway Dose Calculation Methodology

'O'75 x 0'5' 9n4/98 RMeat,113 = C x x QF x Un x FM x DFa (4-4c)

< H >

Equation (C-9) from Regulatory Guide 1.109 where:

=

QF The cow's vegetation consumption rate: 50 kg/ day per Regulatory Guide 1.109, Table E-3.

Ua = The receptor's meat consumption rate. See Table 4-6 and Table 4-7.

DF. =

The ingestion dose factor for H-3, for the receptor in age group (a),

in mrem /pCi, from Tables 4-8 through 4-11.

9/24/98 l FM = The fraction of the animal's intake of H-3 which appears in a kilogram of meat, with units of days / kilogram. This parameter is given by Table 4-13.

l i

0.75 = The fraction of total feed that is water.

0.5 =

The ratio of specific activity of the feed grass to the atmospheric water.

H = Absolute humidity of the atmosphere,0.008 kilograms / cubic meter, i according to Regulatory Guide 1.109. !

l

= The atmospheric dispersion parameter, seconds / cubic meter.

9/24/98 l x/Q i

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 5

=

gg, 1.4 x 104 seconds / cubic meter for releases from the 250 foot stack.

=

1.0 x 10 seconds / cubic meter for releases from the 50 foot stack.

4.3.13 Tritium Vegetation Pathway Dose Calculation Methodology The concentration of tritium is based on the airborne concentration rather than the deposition: i

'0.75 x 0.5' x x Ur x DF.

Rvou = ( < H >

(4-4d) i

\

where:

UT = The total consumption rate of fruits and vegetables, kilogram / year.

This parameter is given in Tables 4-6 and 4-7.

H = Absolute humidity of the atmosphere,0.008 gm/m' per Regulatory Guide 1.109.

0.75 =

The fraction of total feed that is water.

0.5 = The ratio of specific activity of H-3 in the feed grass to the specific activity in atmospheric water.

DFa = The ingestion dose factor for H-3, for the receptor in age group (a), i in mrem /pCi, from Tables 4-8 through 4-11. I

= The atmospheric dispersion parameter, seconds / cubic meter.

x/Q 9/24/98 = 1.4 x 104 seconds / cubic meter for releases from the 250 foot stack.

= 1.0 x 10~5 seconds / cubic meter for releases from the 50 foot stack.

l Table 4-1 Inhalation Dose Factors for Adult Age Group (mrem /pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-7 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.58 x 10-7 1.58 x 10-7 1.58 x 10-7 1.58 x 10-7 1.58 x 10-7 Co-60 No Data 1.44 x 10-6 1.85 x 10-0 No Data 7.46 x 10-4 3.56 x 10-5 Sr-90 1.24 x 10-2 No Data 7.62 x 10-4 No Data 1.20 x 10-3 9.02 x 10-5 Cs-137 5.98 x 10-5 7.76 x 10-5 5.35 x 10-5 2.78 x 10-5 9.40 x 10-6 1.05 x 10-0 Y-90 2.61 x 10-7 No Data 7.01 x 10-9 No Data 2.12 x 10-5 6.32 x 10-5

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-26 Table 4-2 Inhalation Dose Factors for Teen Age Group (mrem /pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109 Table E-8 Organ Nuclide Bone Liver- Total Body Kidney Lung GI-LLI H-3 No Data 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 Co-60 ' No Data 1.89 x 10-6 2.48 x 10-6 No Data 1.09 x 10-3 3.24 x 10-3 Sr-90 1.35 x 10-2 No Data 8.35 x 10-4 No Data 2.06 x 10-3 9.56 x 10-3 Cs-137 8.38 x 10-3 1.06 x 10-4 3.89 x 10-3 3.80 x 10-3 1.51 x 10-3 1.06 x 10-6 Y-90 3.73 x 10-7 No Data 1.00 x 10-5 No Data 3.66 x 10-3 6.99 x 10-3 Table 4-3 Inhalation Dose Factors for Child Age Group (mrem /pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-9 '

Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 Co-60 No Data 3.55 x 10-6 6.12 x 10-6 No Data 1.91 x 10-3 2.60 x 10-3 Sr-90 2.73 x 10-2 No Data 1.74 x 10-3 No Data 3.99 x 10-3 9.28 x 10-3 Cs-137 2.45 x 10-4 2.23 x 10-4 3.47 x 10-3 7.63 x 10-3 2.81 x 10-3 9.78 x 10-7 Y-90 1.11 x 10-6 No Data 2.99 x 10-B No Data 7.07 x 10-3 7.24 x 10-3 Table 4-4 Inhalation Dose Factors for Infant Age Group (mrem /pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-10 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 4.62 x 10-7 4.62 x 10-7 4.62 x 10-7 4.62 x 10-7 4.62 x 10-7 Co-60 No Data 5.73 x 10-6 8.41 x 10-6 No Data 3.22 x 10-3 2.28 x 10-3 Sr-90 2.92 x 10-2 No Data 1.85 x 10-3 No Data 8.03 x 10-3 9.36 x 10-3 Cs-137 3.92 x 10-4 4.37 x 10-4 3.25 x 10-3 1.23 x 10-4 5.09 x 10-3 9.53 x 10-7 Y-90 2.35 x 10-6 No Data 6.30 x 10-5 No Data 1.92 x 10-4 7.43 x 10-3 i

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11 27 Table 4-5 External Dose Factors for Standing on Contaminated Ground (mrem / hour per pico-Curie / square meter)

Selected Nuclides from Regulatory Guide 1.109, Table E-6 Total Nuclide Skin Body 9/2G8 l H-3 0 0 Co-60 2.00 x 10-5 1.70 x 10-6 sng 8l Sr-90 2.60 x 10-12 2.20 x 10-12 Cs-137 4.90 x 10-9 4.20 x 10-9 Y-90 2.60 x 10-12 2.20 x 10-12 l

Table 4-6 Average Individual Foods Consumption for Various Age Groups (kilo-gram / year or liters / year) l 9GG8l From Regulatory Guide 1.109, Table E-4 Other Seafood Fruits and Age Group Fish (Invertebrates) Vegetables Milk Meat !

Adult 6.9 1.0 190 110 95 Teen 5.2 0.75 240 200 59 Child 2.2 0.33 200 170 37 l Infant 0 0 0 0 0 Table 4-7 Maximum Individual Foods Consumption for Various Age Groups (kilo-gram / year or liters / year) gnesl From Regulatory Guide 1.109, Table E-5 Other Seafood Fruits and Age Group Fish (Invertebrates) Vegetables Milk Meat Adult 21 5.0 520 310 110 Teen 16 3.8 630 400 65 Child 6.9 1.7 520 330 41 Infant 0 0 0 330 0 1

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11 28 Table 4-8 Ingestion Dose Factors for Adult Age Group (mrem /pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-11 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 Co-60 No Data 2.14 x 10-0 4.72 x 10-6 No Data No Data 4.02 x 10-3 Sr-90 7.58 x 10-3 ' No Data 1.86 x 10-3 No Data No Data 2.19 x 10-4 Cs-137 7.97 x 10-3 1.09 x 10-4 7.14 x 10-3 3.70 x 10-3 1.23 x 10-3 2.11 x 10-6 Y-90 9.62 x 10-9 No Data 2.58 x 10- No Data No Data 1.02 x 10-4 10

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Table 4-9 Ingestion Dose Factors for Teen Age Group (mrem /pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-12 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 Co-60 No Data 2.81 x 10-0 6.33 x 10-6 No Data No Data 3.66 x 10-3 Sr-90 8.30 x 10-3 No Data 2.05 x 10-3 No Data No Data 2.33 x 10-4 Cs-137 1.12 x 10-4 1.49 x 10-4 5.19 x 10-3 5.07 x 10-3 1.97 x 10-3 2.12 x 10-6 Y-90 1.37 x 10-6 No Data 3.69 x 10- No Data No Data 1.13 x 10-4 10 Table 4-10 Ingestion Dose Factors for Child Age Group (mrem /pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-13 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 Co-60 No Data 5.29 x 10-6 1.56 x 10-3 No Data No Data 2.93 x 10-3 Sr-90 1.70 x 10-2 No Data 4.31 x 10-3 No Data No Data 2.29 x 10-4 Cs-137 3.27 x 10-4 3.13 x 10-4 4.62 x 10-3 1.02 x 10-4 3.67 x 10-3 1.96 x 10-6 Y-90 4.11 x 10-8 No Data 1.10 x 10-9 No Data No Data 1.17 x 10-4

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-29 Table 4-11 Ingestion Dose Factors for Infant Age Group (mrem /pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109. Table E-14 Organ l

Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 3.08 x 10-7 3.08 x 10-7 3.08 x 10-7 3.08 x 10-7 3.08 x 10-7 l Co-60 No Data 1.08 x 10-3 2.55 x 10-5 No Data No Data 2.57 x 10-5 Sr-90 1.85 x 10-2 No Data 4.71 x 10-3 No Data No Data 2.31 x 10-4 Cs-137 5.22 x 10-4 6.11 x 10-4 4.33 x 10-3 1.64 x 10-4 6.64 x 10-3 1.91 x 10-6 ,

Y-90 8.69 x 104 No Data 2.33 x 10-9 No Data No Data 1.20 x 10-4 Table 4-12 88488 l Stable Element Transfer Data For Cow-Milk Pathway (days / liter)

Selected Nuclides from Regulatory Guide 1.109, Table E-1 Element Fm H 1.0 x 10-2 Co 1.0 x 10-3 Sr 8.0 x 10-4 Cs 1.2 x 10-2 Y 1.0 x 10-3 Table 4-13 904/98 l Stable Element Transfer Data For Cow-Meat Pathway (days / kilo-gram)

Selected Nuclides from Regulatory Guide 1.109. Table E-1 I Element Fr H 1.2 x 10-2 Co 1.3 x 10-2 Sr 6.0 x 10-4 Cs 4.0 x 10-3 Y 4.6 x 10-3 I

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l NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

> REVISION 1, I A & 2 l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-3 0 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE 5.1 WHOLE BODY DOSE

Specification 2.10 limits the whole body dose equivalent from the Uranium fuel to no more than 25 mrem / year. The whole body dose is determined by summing the calculated ,

doses from the following:

a. Stack Noble gas releases, using equation (4-1).

b. Stack Particulate releases, using equation (4-3). I

c. Stack Tritium releases, using equation (4-4).

d. Liquid releases, using equation (2-1).

To this calculated exposure is added potential direct radiation exposure to an individual at the site boundary. The only portion of the site boundary where there is significant direct radiation is near the radwaste facilities at the [PG&E] North edge of the site. Due I

to the possibility that an individual at the shoreline (fishing, bird watching, etc.) may use the path at the brow of the cliff for access, the TLD stations along the path are used to estimate an annual radiation exposure. The time period used for this estimate is 67 hours7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months / year, given by Table E-5 of Regulatory Guide 1.109, as the maximum time for shoreline recreation for the Teen age group.

5.2 SKIN DOSE Specification 2.10 limits the dose to any organ (thyroid excepted) to less than or equal to 25 mrem / year. The dose to the skin is determined by suniming the calculated doses from the following:

a. Stack Noble gas releases, using equation (4-2).

b.- Stack Tritium releases, using equation (4-4). (For H-3, the exposure to all organs is essentially equal, so the whole body value may be used for skin.)

c. Liquid Tritium releases, using equation (2-1). (Use whole body value, as above, for H-3).

d. The potential direct radiation exposure to an individual at the site boundary base on TLD stations, as determined in Section 5.1 above.

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NUCLEAR POWER GENERATION NUM'.sER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-31 5 DOSE TO OTHER ORGANS Specification 2.10 limits the dose to any organ (thyroid excepted) to less than or equal to 25 mrem / year. The dose to any individual other than skin organ is determined by summing the calculated doses from the following:

a. . Stack Noble gas releascs, using equation (4-1).

b. Stack Tritium releases, using equation (4-4).

c. Liquid Tritium releases, using equation (2-1).

d. The potential direct radiation exposure to an individual at the site boundary base on TLD stations, as determined in Section 5.1 above.

5.4 DOSE TO THE THYROID Specification 2.10 limits the dose to the thyroid to less than or equal to 75 mrem / year.

Since Unit 3 has not operated since July 2,1976, there is an insufficient radioactive iodine source term remaining onsite to approach this limit. Therefore, calculation of dose to the thyroid is not required.

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11 32 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING SOLIDIFICATION i I

6.1 SCOPE l This section pertains to radioactive waste containing a total specific activity which exceeds the burial ground criteria for solidification, or which exceeds the concentration limits for Class A waste as defmed in 10 CFR 61. These wastes must be stabilized by was l solidification and contain no freestanding liquids prior to shipment ofTsite for land burial, or else be packaged in a high integrity container in accordance with Section 7.0.

6.2 PROGRAM ELEMENTS was l For the land burial disposal of radioactive waste requiring solidification, HBPP shall implement the following steps:

was l 6.2.1 Contract vendor solidification service may be utilized. The contract vendor solidification service may consist of solidification by the contractor or supply of materials, procedures and process control program (PCP) for HBPP solidification. j l I i 6.2.2 This vendor service shall include transmittal to HBPP of copies of their solidification procedure and PCP prior to performing the solidification.

! 6.2.3 The process parameters included in the PCP may include, but are not limited to, waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil j content, waste principal chemical constituents and mixing and curing times. ;

MSB l 6.2.4 The vendor solidification procedure and PCP shall be incorporated into a Plant Manual procedure that will be effective during the solidification process. This procedure will identify all Plant interfaces with the vendor's equipment (e.g., flush water, fire protection, shielding requirements, etc.), as well as identify the actions to be taken if excess free standing liquids are observed. This procedure shall require at least one representative test specimen from at least every tenth batch of waste processed to ensure solidification. The procedure should also include the actions to be taken if the test specimen fails to solidify.

was l 6.2.5 This procedure shall be reviewed per plant procedures for adequacy in meeting applicable State, Federal, Department of Transportation and burial ground regulatory requirements and approved by the Plant Manager or designee prior to its implementation. This review shall ensure that the stability requirements of 10 CFR 61.56(b) for wastes exceeding Class A concentrations are met by the vendor solidification program.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-33 7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED IN llIGil INTEGRITY CONTAINERS 7.1 SCOPE This section pertains to radioactive waste containing specific activity which exceeds the burial ground criteria for solidification, or which exceeds the concentration limits for Class A waste as defined in 10 CFR 61. These wastes must be stabilized by packaging in dewatered form in a high-integrity container which meets burial ground and regulatory requirements, or else be solidified in accordance with Section 6.0.

7.2 PROGRAM ELEMENTS wasl For land burial disposal of radioactive waste requiring a high-integrity container, HBPP shall implement the foltawing step s:

7.2.1 .A contract vendor high-integrity container shall be used.

was 7.2.2 The container shall be demonstrated to have been approved or have a current Certificate of Compliance prior to acceptance for use by HBPP. This shall include provision by the vendor to HBPP of documentation reflecting this authorization.

7.2.3 The material placed in the high-integrity container shall meet all applicable burial ground and regulatory waste form requirements for waste which is packaged in this manner.

7.2.4 The above criteria shall be met by following Plant Manual procedures which will be reviewed and approved by the Plant Manager or designee in accordance with Plant Manual administrative procedures prior to implementation at the time of packaging and disposal.

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NUCLEAR POWER GENERATION NUMBER ODCM ,

HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1,1 A & 2 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-3 4 l 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED RESINS i AND OTIIER WET WASTES 8.1 SCOPE gg This section pertains to bead-type spent radioactive demineralizer resin and other wet wastes shipped for land burial which contain a total specific activity less than the burial ground criteria for solidification, and which does not exceed the concentration limits for Class A waste as defined in 10 CFR 61.

8.2 . PROGRAM ELEMENTS 8.2.1 The dewatered resin or wet wastes must meet the requirements of 10 CFR 61.56 or those of the burial ground (whichever is more restrictive) for freestanding, noncorrosive liquid.

8.2.2 For bead resins, the preceding criterion will be met by following approved Plant Manual procedures for dewatering resin.

8.2.3 Liquid waste, that will not be tiermal treated to remove freestanding liquid, must i majl be solidified.

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4 REVISION 1, I A & 2 l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-3 5 9.0 PROGRAM CIIANGES 9.1 PURPOSE OF THE OFFSITE DOSE CALCULATION MANUAL The Offsite Dose Calculation Manual was developed to support the implementation of the Radiological Emuent Technical Specifications required by 10 CFR 50, Appendix 1, and 10 CFR 50.36. The purpose of the manualis to provide the NRC with sumcient information relative to efiluent monitor setpoint calculations, emuent related dose calculations, and environmental monitoring to demonstrate compliance with radiological effluent controls.

9.2 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL ,

i It is recognized that changes to the ODCM may be required during the SAFSTOR period. All changes shall be reviewed and approved by the PSRC and the Plant Manager prior to implementation. The NRC shall be informed of all changes to the ODCM by providing a description of the change (s) in the first Annual Radioactive EfIluent Release Report following the date the change became efTective. Records of the reviews performed on change to the ODCM should be documented and retained for the duration J i

of the possession only license. I 10.0 P_ROCEDURE OWNER 10.1 Sr. Radiation Protection Engineer 1

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ODCM l

APPENDIX A Revision 1, I A & 2 l Page A 1 i

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1 APPENDIX A SAFSTOR BASELINE CONDITIONS I

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ODCM APPENDIX A Revision 1, I A & 2 Page A-2 1.0 LIQUID AND GASEOUS EFFLUENTS 1.1 LIQUID EFFLUENTS Baseline levels of radioactive materials contained in liquid effluents during the SAFSTOR period were established in the Environmental Report submitted as Attachment 6 to the SAFSTOR license amendment request. These values are presented for cumulative annual release and average monthly discharge in Table A-1.

1.2 GASEOUS EFFLUENTS Baseline levels of radioactive materials contained in gaseous effluents established in the Environmental Report are presented for cumulative annual and average monthly release in Table A-2.

Table A-1 Baseline Liquid Emuent Activity

Type of Activity Annual Release Monthly Average Release (Curies) (Curies)

Tritium 8.6E-2 7.2E-3 Principal Gamma Emitters (total) 1.85E-1 1.54E-2 Grl/98 lStrontium-90 3.28E-4 2.73E-5 l Table A-2 Baseline Gaseous Emuent Activity Type of Activity Annual Release Monthly Average Release I (Curies) (Curies) l Tritium <4.0E-2 <3.3 E-3 28 3.16E-4 2 63E-5 Particulate Gamma Emitters (total)

Strontium-90 3.38E-6 2.82E-7

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! ODCM APPENDIX B j Revision 1, I A & 2 l Page B-1 i

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APPENL)IX B BASIS FOR INSTANTANEOUS X/Q VALUE !

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ODCM A."SENDIX B Revision 1, I A & 2 Page B-2 1.0 BASIS FOR INSTANTANEOUS X/O VALUE 1.1 The " annual average" value of this parameter was calculated (on the basis of historical meteorological monitoring data) for comparison of the plant operating releases with 10 CFR 50, Appendix 1. This calculation was performed using the guidance of Regulatory Guide 1.109. For that calculation, the largest value ofX/Q was found to be 4 x 10 4 seconds / cubic meter.

1.2 That value of the parameter is not appropriate for the monitoring required by the SAFSTOR Technical Specifications, because it does not represent a "real-time" dispersion coeflicient. The appropriate value (for determining a stack monitor alarm setpoint) will be based on the atmospheric models of Regulatory Guide 1.145, Atmospheric Dispersion Modelsfor Potential Accident Consequence Assessments at Nuc/ car Power Plants. These models are intended to estimate "real time" conditions, rather than " annual average" values.

1.3 A conservative assumption used for calculating this dispersion coeflicient is that the release from the stack is during " fumigation" conditions (Pasquill stability class F), and that the plume proceeds downwind toward the rising terrain of Humboldt Hill. The dispersion coefficient is calculated for the fumigation condition (using the highest terrain at any particular distance to decrease the " effective stack height"), until the calculated ground level concentration is the same as the plume centerline concentration. From that distance, the dispersion is calculated for the plume centerline concentration. This calculation generally follows the guidance for the use of equations 4 and 5 of Regulatory )

Guide 1.145. l l

l.4 The calculations include the effects of the stack efIluent vertical velocity, which l increases the effective stack height. This plume rise is calculated from equation 5.1 of l " Meteorology and Atomic Energy - 1968" (except that the thermal buoyancy term is 1 neglected):

l h.= (d) x [w/u]" (B-1) where:

( h.= Plume rise, meters d= Stack nozzle diameter, meters w= Stack gas exit velocity, meters /sec u= Wind velocity, meters /sec L

L ODCM APPENDIX B Revision 1, I A & 2 Page B-3 1.5 The calculation values are a wind speed of 2 meters /second (commonly assumed for fumigation conditions). The stack nozzle diameter of 36 inches (0.914 meters) and stack flow rate of 31,000 cfm (14.6 cubic meters /sec result in a gas exit velocity of 22.3 meters /second, l Plume Rise = 0.914 x (22.3/2)" = 26.7 meters 1.6 Accordingly, the following calculations use an effective stack height of 103 meters, rather than the physical stack height of 76 meters (250 feet).

1.7 The concentration calculated for fumigation conditions at ground levelis:

1 X/Q = (B-2) 2.507 x U, x s, x h.

I where:

U, = 2 meters /second wind speed, a " reasonable assumption" for the fumigation condition. i a

sy = Lateral plume spread, meters, from Figure 1 of the Regulatory Guide 1.145.

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h. = Effective stack height, meters. Note that the ground elevation at the base of the stack is at +12 feet (4 meters, approximately), so that the effective stack elevation is 107 meters. To account for the rising terrain, the relative elevation is used for the stack effective height.

Distance Ground El. h. sy X/Q j (meters) (meters) (meters) (meters) (seconds / cubic meter) 500 5 102 20 9.78 x 104 1000 17 90- 38 5.83 x 104 4

1500 66 41 54 9.01 x 10 1700 91 16 62 2.01 x 10" 2000 104 3 70 9.50 x 10" 1.8 The concentration at the plume centerline is:

1 X/Q = (B-3) 3.1416 x U, x s, x s, l where:

i U, = 2 meters /second wind speed, a " reasonable assumption" for the fumigation condition.

sy = Lateral plume spread, meters, from Fi;;ure 1 of the Regulatory Guide.

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ODCM APPEND!X B

l. Revision 1, I A & 2 Page B-4 l s, =

Vertical plume spread, meters, from Figure 2 of the Regulatory Guide.

l Distance sy s, X/Q (meters) (meters) (meters) (seconds / cubic meter) j 500 20 8.3 9.59 x 10" 1000 38 13.5 3.10 x 10"

_ 2000 70 21 1.08 x 10" 4

3000 99 28 5.74 x 10 .

i 1.9 Based on the assumption that the offsite concentration is not higher than that calculated I l

with either of the two mathematical models, the highest offsite ground level i i

concentration occurs where the concentration curve for the fumigation condition crosses the curve for the plume centerline condition. At this location, X/Q is 1.46 x 10" seconds / cubic meter. i l

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p-ODCM APPENDIX B l' Revision 1, I A & 2 Page B-5 Figure i 153 u .

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e Fumigation

& Centerline g164;i\

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.155 500 1000- 1500 2000 2500 3000 Dstance Downwind (meters)

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ODCM APPENDIX C Revision 1, I A & 2 Page C-1 APPENDIX C Kr-85 MONITOR CALIBRATION l

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ODCM l APPENDIX C l Revision 1, I A & 2 Page C-2 1.0 Kr-85 MONITOR CALIBRATION 1.1 The original calibration factor was based on the manufacturer's calibration.

This calibration was re-examined after a test was performed to determine the effects of sample line pressure drop on the calibration of the stack sampler / monitor. This section documents the results of that test and review?

1.1.1 The two detector chambers were found to have essentially identical reduced pressures. The pressures in chambers 'A' and 'B' were -2.176 and -2.203 in. Hg. (relative to atmospheric pressure), respectively.

1.1.2 The effect of changing the stack particulate sample filter from 'diny' to

' clean' was small, with a pressure drop difference of 0.009 in. Hg. The chamber 'A' pressure was measured (relative to atmospheric pressure) at -2.167 for the ' clean' filter condition and at -2.176 in. Hg. for the

' dirty' filter conditions.

l 1.1.3 The true system flowrate was found to differ slightly from the flowrate indicated on the Photohelic Gauge when the system was set up in the then normal S.T.P. flow calibration configuration (flow calibrator inlet at atmospheric pressure), but the calibration was accurate at the normal system conditions (approximately 2 in. Hg. vacuum). The test pressure / flow measurement results are summarized below:

Photohelic Gauge C-812 Air Flow Chamber 'A' Vacuum Indicated Calibrator (in. Hg.) Flowrate (cfm) Flowrate (cfm) 0.275 2.2 1.9 1.001 2.2 2.15 1.995 2.2 2.2 3.020 2.2 2.2 1.2 The Kr-85 monitoring system was originally calibrated v.ith Kr-85 gas standards. The standard certificate concentrations we? rw foi :he gas at

'STP'(Standard Temperature & Pressure), but the calib, % vas performed at ' ambient' conditions, without any correction. AccordL e vendor of the radioactive standard gas, STP conditions are 760 mm3H . ad 0 C (273 K). The system calibration conditions were ' Ambient' temperature (recorded as 70 F, or 294 K) and ' Atmospheric' pressure (exact barometric pressure not recorded), at Indianapolis, IN. Since the elevation was about 800', the absolute atmospheric pressure could have ranged from about 29.0 to 29.6 in.

l Hg. Assuming that the absolute pressure was 29.3 in. Hg. (744 mm Hg.),

the concentration of the gas in the chambers at the actual calibration t

' After the Technical Review Group meeting of 4/14/93, a test procedure was developed to determine the effects of sample line pressure drop. The test was perfonned on 5/18/93.

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ODCM APPENDIX C Revision 1, I A & 2 Page C-3 i i

conditions would have been lower by a factor of 0.909 due to the lower l

pressure and higher temperature:

l 60 294 = 02 l

1.3 The following table summarizes the original calibration results, with the

, assumption that the absolute pressure for the calibration was 29.3 in. Hg.: I Gas Gas Detector Detector Detector 'A' Detector 'B' Concentration Concentration 'A' Net 'B' Net Calibration Calibration at STP at Original Countrate Countrate Factor Factor ;

( Ci/cc) Calibration (cpm) (cpm) ( Ci/cc per (pCi/cc per Conditions epm) cpm)

(pCi/cc) I 1.84E-6 1.67E-6 6.08El 6.42El 2.75E-8 2.61 E-8 l 1.66E-5 1.51 E-5 4.88E2 4.98E2 3.09E-8 3.03 E-8 l 1.67E-4 1.52E-4 5.10E3 5.39E3 2.98E-8 2.82E-8 j l 1.67E-3 1.52E-3 5.26E4 5.46E4 2.89E-8 2.78E-8 I 1.09E-2 9.91 E-3 3.36E5 3.38E5 2.95E-8 2.93 E-8 f 1.4 The effect of the sample line pressure drop (see section 6.1.1) is to reduce the density of the gas in the detector chambers relative to the density of the gas leaving the stack, thereby making the system read lower than it would if the l gas in the chambers was at atmospheric pressure. The correction factor for

this effect is about 1.08 (29.92/27.74). If this correction is applied to the l average of the 10 measurements above, the resulting calibration would be l 1.08 x 2.88E-8 = 3.11E-8 pC/cc per epm. This is essentially the same value j as the one originally established (3.lE-8), so the error produced by

neglecting the sample line pressure drop effectively canceled out the error i

resulting from incorrectly interpreting the original calibration.

1.5. The flow control system calibration (S.T.P. 3.16.7) was revised so that the Photobelic Gauge metering system flowrate is checked at the operating absolute pressure condition.

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