ML20132C739

Rev 0 to Safstor Odcm

ML20132C739
Person / Time
Site:	Humboldt Bay
Issue date:	12/09/1996
From:	PACIFIC GAS & ELECTRIC CO.
To:
Shared Package
ML20132C729	List: ML20132C723 ML20132C739
References
PROC-961209, NUDOCS 9612180447
Download: ML20132C739 (102)

Text

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ATTACHMENT D 1

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HBPP OFFSITE DOSE CALCUUtTION MANUAL o

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9612190447 961209 PDR ADOCK 05000133 l

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TITLE SAFSTOR OFFSITE DOSE CALCULATION EFFEC.DATE MANUAL PAGE i

APPROVED BY JN i

DATE

@ * 'lGNT MANAGER (Procedure Classification - Quality Related)

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

Calculational Methods and Parameters.

Part I contains the specifications for liquid and gaseous radiological effluents (RETS) developed in accordance with NUREG-0473, Draft Radiological Effhient Technical Specifications - BWR, by License Amendment Request (LAR) 96-02 and the radiological environmental monitoring program (REMP). 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 RadiologicalEffluent Technical Specipcations in the Adminisn ative 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.

Implementation of the LAR 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, j

Table 2, Column 2 efiluent 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 effluents 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 shut down in excess of 20 years and is in SAFSTOR Decommissioning. These dose limits were established following the guidance of NUREG-0133, Preparation ofRadiological Efpuent Technical Specipcationsfor Nuclear Power Plants, and NUREG-0473. This guidance was modified, as appropriate, to reflect the SAFSTOR decommissioning licensing basis contained in the HBPP SAFSTOR Decommissioning Plan, the Environmental Repon submitted as Attachment 6 to the HBPP SAFSTOR licensing amendment request and NUREG-1166, HBPP Final Environmental Statement.

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 Repon which established baseline conditions for soil, biota and sediments. The REMP also includes requirements to panicipate 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 efiluents released from HBPP during the SAFSTOR period. Pan II also contains the methodology used to

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i TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE ii determine effluent monitor alarm / trip setpoints which assure that releases ofradioactive 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 contrc!s regarding the content of the Annual Radiological Environmental Monitoring Report and the Annual Radioactive Effluent Release Report which are required by Technical Specifications VIIJ.1 and VII.J3 and administrative controls regarding major changes to radioactive waste treatment systems.

The ODCM shall become effective after the review and approval of the Plant Staff Review Committee l

and approval by the Plant Manager in accordance with Technical Specification Section VII..D.1.f.(1).

Changes to the ODCM shall be documented and records of reviews performed shall be retained. This documentation shall contain sufficient information to support the change (including analyses or evaluations), and a determination that the change will maintain the level of radioactive effluent control required by the regulations listed in Technical Specification VII.F and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

Changes shall 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 Efiluent 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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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE iii TABLE OF CONTENTS PART I - SPECIFICATIONS S.ection Title e

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1.0 DEFINITIONS I-l 2.0 SPECIFICATIONS I-6 2.1 Radioactive Liquid Emuent Monitoring Instrumentation I-6 2.2 Radioactive Gaseous Emuent Monitoring Instrumentation I-9 2.3 Liquid Emuent - Concentration I-12 2.4 Liquid Emuent - Dose 1-15 2.5 Liquid Waste Treatment I-16 2.6 Gaseous Emuents - Dose Rate 1-17 2.7 Gaseous Emuents: Dose - Noble Gases I-20 2.8 Gaseous Efiluents: Dose - Tritium and Radionuclides in Particulate Form I-21 2.9 Solid Radioactive Waste I-22 2.10 Total Dose 1-23 2 11 REMP Monitoring Program 1-24 2.12 REMP Interlaboratory Comparison Program I-37 3.0 SPECIFICATION BASES 1-B38 3.1 Radioactive Liquid Emuent Monitoring Instrumentation Basis I-B38 3.2 Radioactive Gaseous Emuent Monitoring Instrumentation Basis I-B38 3.3 Liquid Emuent Concentration Basis I-B38 3.4 Liquid Emuent Dose Basis I-B39 3.5 Liquid Waste Treatment Basis I-B39 3.6 Gaseous Emuents Dose Rate Basis I-B39 3.7 Gaseous Emuents: Noble Gases Dose Basis I-B40 3.8 Gaseous Emuents: Tritium and Radionuclides in Particulate Form Dose Basis I-B40 3.9 Solid Radioactive Waste Basis I-B41 3.10 Total Dose Basis I-B41 3.11 REMP Monitoring Program Basis I-B42 3.12 REMP Interlaboratory Comparison Program Basis I-B43 4.0 ADMINISTRATIVE CONTROLS 1-44 4.1

/,anual Radiological Environmental Monitoring Repon I-44 42 Annual Radioactive Emuent Release Report 1-48 4.3 Special Repons 1-49 44 Major Changes to Radioactive Waste Treatment Systems I-49

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T!TLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE iv PART II-CALCULATIONAL METHODS AND PARAMETERS ~

Section Title P_ase 1.0 EFFLUENT MONITOR SETPOINT CALCULATIONS II-l 1.1 Liquid Effluent Monitors 11-1 1.2 Gaseous Effluent Monitor II-2 P

2.0 LIQUID EFFLUENT DOSE CALCULATIONS II-5 2.1 Calendar Quarter 4

11-5 2.2 Calendar Year 11-5 2.3 Liquid Effluent Dose Calculation Methodology II-6 I

4 3.0 LIQUID WASTE TREATMENT II-9 4

3.1 Treatment Requirements 11 9 3.2 Treatment Capabilities 11 9 4.0 GASEOUS EFFLUENT DOSE CALCULATIONS 11-13 I

4.1 Dose Rate 11-1 3 J

4.2 Dose - Noble Gases 11-1 4 4.3 Dose - Tritium and Radionuclides in Particulate Form II-16 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE II-29 4

l 5.1 Whole Body Dose 11-2 9 i

5.2 Skin Dose II-29 l

5.3 Dose to Other Organs II-29 5.4 Dose to the Thyroid 11-3 0 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING 11-3 1 SOLIDIFICATION

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7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED II-32 IN HIGH INTEGRITY CONTAINERS 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED 11-3 3 RESINS AND OTHER WET WASTES 9.0 PROGRAM CHANGES II-34

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE v

PART U - CALCULATIONAL METHODS AND PARAMETERS 1

Section Title Pne APPENDIX A - SAFSTOR BASELINE CONDITIONS A-1 APPENDIX B - BASIS FOR INSTANTANEOUS X/Q VALUE B-1 APPENDIX C - Kr-85 MONITOR CALIBRATION C-1 d

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE vi l

LIST OF TABLES - PART I i

Igbje Title Egge l-1 Frequency Notation I-5 2-1 Radioactive Liquid Emuent Monitoring Instmmentation I-7 2-2 Radioactive Liquid EEuent Monitoring Instmmentation Surveillance Requirements I-8 2-3 Radioactive Gaseous Emuent Monitoring Instmmentation I-10 2-4 Radioactive Gaseous EEuent Monitoring Instmmentation Surveillance I-I1 Requirements 2-5 Radioactive Liquid Waste Sampling and Analysis Program I-13 2-6 Radioactive Gaseous Waste Sampling and Analysis Program I-l 8 2-7 HBPP Radiological Environmental Monitoring Program I-25 2-8 Reporting Levels for Radioactivity Concentrations In Environmental Samples I-28

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2-9 Detection Capabilities for Environmental Sample Analysis Lower Limits Of I-29 i

Detection (LLD) 2-10 Distances and Directions To Environmental Monitoring Stations I-31 41 Radiological Environmental Monitoring Program Annual Repon Summary -

I-46 Example i

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE vii 4

LIST OF TABLES - PART II Table Title Pace 2-1 Ingestion Dose Factors for Adult Age Group II-7 2-2 Ingestion Dose Factors for Teen Age Group II-7 2-3 Ingestion Dose Factors for Child Age Group 11-7 2-4 Bioaccumulation Factors for Saltwater Environment 11-8 2-5 Average Individual Foods Consumption for Various Age Groups 11-8 2-6 Maximum Individual Foods Consumption for Various Age Groups II-8 4-1 Inhalation Dose Factors for Adult Age Group II-24 4-2 Inhalation Dose Factors for Teen Age Group I1-25 4-3 Inhalation Dose Factors for Child Age Group 11-25 4-4 Inhalation Dose Factors for Infant Age Group II-25 4-5 External Dose Factors for Standing on Contaminated Ground 11-2 6 4-6 Average Individual Foods Consumption for Various Age Groups 11-26 4-7 Maximum Individual Foods Consumption for Various Age Groups II-26 4-8 Ingestion Dose Factors for Adult Age Group 11-2 7 4-9 Ingestion Dose Factors for Teen Age Group I1-27 4-10 Ingestion Dose Factors for Child Age Group 11-2 7 4-11 Ingestion Dose Factors for Infant Age Group 11-2 8 4-12 Stable Element Transfer Data For Cow-Milk Path 11-28 4-13 Stable Element Transfer Data For Cow-Meat Path 11-2 8

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE siii 4

i LIST OF FIGURES - PART I Figure Title P_ age 2-1 IEPP Onsite TLD Locations 1-32 2-2 HBPP Onsite Monitoring Well Locations I-33 2-3 IEPP Offsite Sampling Locations' I-34 4

2-4 HBPP Offsite Sampling Locations (Continued)

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2-5 HBPP Offsite Sampling Locations (Continued)

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-1 PART I - SPECIFICATIONS 1.0 DEFINITIONS 1.1 ACTION 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 channelincluding th sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. CHANNEL CALIBRATION may be 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 shallinclude, 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 1.4 CHANNEL FUNCTIONAL TEST Analog channels - one injection of a simulated or actual signal into the channel as a.

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.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-2 1.6 FREQUENCY 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.

1.8 INSTANTANEOUS CONCENTRATION INSTANTANEOUS CONCENTRATION is the concentration averaged over one hour -

of radioactive materials in effluents.

l 1.9 LIQUID RADWASTE TREATMENT SYSTEM The LIQUID RADWASTE TREATMENT SYSTEM shall be any available equipment (e.g., filters, evaporators, demineralizers, or contractor services) capable of reducing the quantity of radioactive material, in liquid effluents, prior to discharge.

1.10 MEMBER OF THE PUBLIC MEMBER OF THE PUBLIC means an individual in any area locsted beyond the i

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 i

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.

1.11 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 ofgaseous and liquid efnuent monitoring Alarm Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program. The ODCM also contains the Radioactive EfDuent Controls and Radiological Environmental Monitoring Program and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Radioactive Effluent Release Repons. 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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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-3 i-5 j

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 definition shall be the assumption that all necessary attendant instrumentation, controls, i

normal and emergency electrical power sources, cooling or seal water, lubrication or 3

other auxiliary equipment, that are required for the system, subsystem, train, component or device to perform its function (s), are also capable ofperforming their related support function (s).

l.13 PROCESS CONTROL PROGRAM s

i The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, j-sampling, analyses, tests, and determmations to be made to ensure that processing and

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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

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i with 10 CFR Parts 20,61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

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1.14 PURGE - PURGING 4

PURGE or PURGING is the controlled process of discharging air or gas from a '

3 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 I

confinement.

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1.15 SITE BOUNDARY The SITE BOUNDARY shall be the boundary of the unrestricted area used in the offsite j

dose calculations for gaseous rad liquid effluents as deSned in Technical Specification 11.B.

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

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

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-4 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 ofprotection ofindividuals from exposure to radiation and radioactive materials and within, at, or beyond the SITE BOUNDARY.

1.19 URANIUM FUEL CYCLE l

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

1.20 VENTILATION EXHAUST TREAThENT SYSTEM A VENTILATION EXHAUST TREAThENT SYSTEM is any system designed and i

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 effluents).

1.21 VENTING VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, huinidity, 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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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-5 i

Table 1-1 FREQUENCY NOTATION NOTATION FREOUENCY D

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

A At least once per 365 days.

P Completed prior to each release.

N.A.

Not applicable.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-6 2.0 SPECIFICATIONS RADIOACTIVE LIOliTD EFFLilENT MONITORING INSTRllMENTATION 2.1 The radioactive liquid emuent monitoring instrumentation channels shown in Table 2-1 shall be l

OPERABLE with their alarm / trip setpoints set to ensure that the limits of specification 2.3 are not exceeded.

APPLICABILITY: At all times ACTION:

With a radioactive liquid emuent monitoring instrumentation channel alarm / rip setpoint a.

t 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 emuent 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.

i SURVEILLANCE REQUIREMENTS 2.1.1 Each radioactive liquid 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 2-2.

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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 b.

Caisson Sump Discharge Monitor 1

22 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:

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

2.3.1, and

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

c.

Otherwise, suspend releases of radioactive materials via this pathway.

ACTION 22 With less than the required number of OPERABLE channels, effluent releases via this pathway may continue, provided that:

With less than the required number of OPERABLE channels, effluent releases sia a.

this pathway may continue, provided that grab samples are collected and analyzed for gross radioactivity, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-8 Table 2-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Channel Source Channel Ft.netional Instrument Check Check Calibration Test 1.

Gross Radioactivity Monitors Providing Alarm and Automatic Termination ofRelease a.

Process Water Monitor D

P A

Q(1)(2) b.

Caisson Sump Discharge Monitor D

M A

Q(1)(2) 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:

Instmment indicates measured levels above the alarm setpoint.

a.

b. Circuit failure.

c. Instrument indicates a downscale failure.

d. Instrument controls not set in operate mode.

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_ TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION i

2.2 The radioactive gaseous effluent monitoring instmmentation channels shown in Table 2-3 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of specification 2.6 -

i are not exceeded.

MPLICABILITY: Whenever the ventilation system is in operation.

ACTION:

With a radioactive gaseous effluent monitoring instmmentation channel alarm / trip a.

setpoint less conservative than required above, without delay suspend the release of radioactive gaseous effluents 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 effluent monitoring instmmentation channels inoperable, take the ACTION shown in Table 2-3. For the instrumentation covered, 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 Effluent Release Report.

SURVEILLANCE REQUIREMENTS

'2.2.1 Each radioactive gaseous effluent monitoring instmmentation 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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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-10 Table 2-3 i

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1.

Stack Gas Monitoring System a.

Noble Gas Activity Monitor 1

23 b.

Iodine Sampler

• N.A.

c.

Particulate Sampler 1

d.

Effluent System Flow Rate Monitor

• N.A.

e.

Sampler Flow Rate Monitor

• N.A.

Table Notation 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.

• Not included in the stack gas monitoring system.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-11 Table 2-4 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Channel Source Channel Functional Instrument

[h ek Check Calibratiqn ICE t

Stack Gas Monitoring System 1.

a.

Noble Gas Activity Monitor D

M A

Q(1) b.

Iodine Sampler

• N.A.

N.A.

N.A.

N.A.

t c.

Particulate Sampler W

N.A.

N.A.

N.A.

d.

Efiluent System Flow Rate Monitor

• N.A.

N.A.

N.A.

N.A.

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Sampler Flow Rate Monitor

• N.A.

N.A.

N.A.

N.A.

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

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Instmment indicates measured levels above the alarm setpoint.

a.

b. Circuit failure.

a

c. Instmment indicates a downscale failure.

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• Not included in the stack gas monitoring system.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-12 4

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LIOUID EFFLUENT-CONCENTRATION 2.3 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 j

B, Table 2, Column 2.

APPLICABILITY: At all times.

ACTION:

i With the instantaneous concentration ofradioactive materials 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.

SURVEILLANCE REQUIREhENTS 2.3.1 Radioactive liquid wastes shall be sampled and analyzed in accordance with the sampling and analysis program of Table 2-5.

2.3.2 The results of the radioactivity analyses shall be used with the calculational methods in Part II of the ODCM to assure that the concentrations of radioactive material released to Humboldt Bay are maintained within the limits of Specification 2.3.

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_ TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-13 Table 2-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit Sampling Analysis Type of Activity ofDetection Liquid Release Type Frequency Frequency Analysis (LLD)

(pCi/ml)*

A. Batch Waste Release Tanks' P'

P Principal Gamma 5 x 10 7

1. Treated Waste Hold Tank (2)

Each Batch Each Batch Emitters' I

2. Waste Receiver Tanks (3)

P M

H-3 1 x 10~'

6 Each Batch Composite Gross Alpha 1 x 10

P Q

Sr-89, St-90 5 x 10

Each Batch Composite' B. Plant Continuous Releases

• Continuous

• W Principal Gamma 5 x 10

l. Caisson Sump Composite

• Emitters' Continuous

• Q Sr-89, Sr-90 5 x 10~'

Composite

• Table Notation The LLD' is the smallest concentration of radioactive material in a sample that will be detected a

with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD =

(E)(V)(2.22 x 10')(e-*)

Where:

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

s is the standard deviation of the background counting rate or of the counting rate of a e

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,

l NUCLEAR POWER GENERATION NUMTER ODCM HOMBOLDT BAY. POWER PLANT VOLUME 4

4 REVISION 0

]..

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-14 q

Table 2-5 (Continued) i

}

Table Notation Y is the fractional radiochemical yield (when applicable),

I 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 j

plant effluents, not environmental samples).

i i

Typical values of E, V, Y, and t shall be used in the calculation.

)

The LLD is defmed as an A PHati(before the fact) limit representing the capability of a measurement system and not as an A costeriori (after the fact) limit for a panicular measurement.

l' 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.

i To be representative of the quantities and concentrations ofradioactive materials in liquid

'I L

c.

j effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

d.

A batch release is the discharge ofliquid wastes of a discrete volume. Prior to sampling for i

analyses, each batch shall be isolated, and then thoroughly mixed to assure representative

{

sampling.

A continuous release is the discharge ofliquid wastes of a nondiscrete volume; e.g., from a e.

j volume or system that has an input flow during the continuous release.

l f.

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 j

peaks which are measurable and identifiable, together with the above nuclides, shall also be i

identified and reported. Nuclides which are below the LLD 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 j

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

1 i

i i

s j

4 N

f i

1 NUCLEAR POWER GENERATION NUMBER' ODCM li HOMBOLDT BAY POWER PLANT VOLUME 4

REVISION 0

_ TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-15 4

i I

LIOUID EFFLUENT-DOSE i

..e.,..

+...

m y +km w.w<> %.. - +

l I

I 2.4 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released beyond the SITE BOUNDARY shall be limited as follows:

n During any calendar quarter to'less than or equal to 1.5 mrem to the total body and less a.

i i

than or equal to 5 mrem to any organ.

j b.

During any calendar year to less than or equal to 3 mrem to the total body and to less i

than or equal to 10 mrem to any organ.

1 3

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release ofradioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a Special Report pursuant to Specification 4.3, which includes:

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

a.

b.

Corrective action taken to reduce the release of radioactive materials in liquid effluents 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 PUBLIC from this source is less than or equal to 3 mrcm total body and less than or equal to 10 mrem to any organ during the calendar year.

SURVEILLANCE REQUIREMENTS 2.4.1 Baseline Comoarison. Cumulative activity contributions from liquid effluents 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.

NOCLEAR POWER GENERATibN NUMSER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-16 L10UID WASTE TREATMENT

(

2.5 The LIQUID RADWASTE TREATMENT SYSTEM shall be used, as appropriate, to reduce radioactive mate 96 in liquid wastes prior to their discharge, when projected monthly doses due to liquid effluents s scharged to Humboldt Bay would exceed the action levels of 0.06 mrem whole body or 0.2 mrem to any organ.

APPLIC. ABILITY: At all times.

i t

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 Specification 4.3, which includes the following information:

Identification ofinoperable equipment and the reasons for inoperability.

a.

b.

Actions taken to restore the inoperable equipment to OPERABLE status.

c.

Actions taken to prevent recurrence.

t 4

1

'l 4

SURVEILLANCE REQUIREhENTS 2.5.1 Baseline Comoarison. Activity contributions from liquid effluents shall be compared with the baseline conditions established by the Environmental Repon 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.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-17 GASEOUS EFFLUEN15 - DOSE RATE i

~

2.6 The dose rate at or beyond the SITE BOUNDARY, due to radioactive materials released in gaseous effluents, shall be limited as follows:

a.

Noble 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 of greater than 8 days: less than or equal to 1500 mrem / year to any organ.

1 APPLICABILITY: At all times.

i ACTION:

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

0 i

4 SURVEILLANCE REQUIREMENTS i

2.6.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits. This determmation has been established by the Environmental Report.

2.6.2 The dose rate due to radioactive materials specified above, other than noble gases, in gaseous effluents shall be determined to be within the above limits by obtaining representative samples and performing analyses in accordance with Table 2-6 and comparing cumulative activity released with the Environmental Report baseline conditions.

NUCLEAR POWER GENERATION NUMZER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-18 Table 2-6 i

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRA I

Minimum Lower Limit Sampling Analysis Type of Activity ofDetection Liquid Release Type Frequency Frequency Analys.2 (LLD)

(pCi/ml)*

Plant Stack Q'

Q*

Noble Gas (Kr-85) 1 x 10" Grab Sample Continuous' W'

Principal Gamma 1 x 10'"

Particulate Emitters

• Sample Continuous' M

Gross Alpha 1 x 10'"

Composite Particulate Sample Continuous' Q

Sr-89, Sr-90 1 x 10'"

Composite Particulate Sample Continuous' Noble Gas Noble Gas 1 x 10 4

Monitor Gross Beta Table Notation The LLD* is the smallest concentration ofradioactive materialin a sample that will be detected with 95% probability with 5% probability offalsely concluding that a blank observation represents a "real" signal.

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

i LLD = -

(E)(V)(2.22 x 10')(e-*)

Where:

LLD is the lower limit ofdetection as defined above (as microcurie per unit mass or volume).

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

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

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4 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-19 4

i Table 2-6 (Continued)

Table Notation E is the counting efficiency (as counts per disintegration),

4 V is the sample size (in units ofmass or volume),

)

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

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).

Typical values of E, V, Y, and t shall be used in the calculation.

3 The LLD is defined as an a pligIi (before the fact) limit representing the capability of a 4

measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

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

1 Samples shall be changed at least once per 7 days and analyses shall be completed within 48 j

hours after changing (or after removal from sampler).

t

-i 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.

4 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 l

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 be identified and reported. Nuclides which are below the LLD 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 j

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

NUCLEAR POWER GENERATibN NUM2ER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-20 GASEOUS EFFLUENTS: DOSE - NOBLE GASES i

2.7 The air dose at or beyond the SITE BOUNDARY due to radioactive noble gases released in gaseous efDuents shall be limited to:

During any calendar quarter: less thius or equal to 5 mrad for gamma radiation and less a.

than or equal to 10 mrad for beta radiation, and t

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 effluents exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a Special Report, pursuant to Specification 4.3, which includes:

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

a.

b.

Corrective action taken to reduce the release ofradioactive 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.

J l

SURVEILLANCE REQUIREhENTS 2.7.1 Compliance with these Specifications for normal SAFSTOR conditions has been established on a licensing basis by the Environmental Report and NUREG-1166, Final EnvironmentalStatementfor Decommissioning Humboldt Bay Power Plant, Unit No. 3, issued by the NRC. If an accident invoMng spent fuel occurs during the SAFSTOR period, the noble gas activity released in gaseous efDuents 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.

i

NUCLEAR POWER GENERATION NUM;ER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-21 GASEOUS EFFLUENTS: DOSE - TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM 2.8 The dose to a MEMBER OF THE PUBLIC from the release of tritium and radioactive materials l

in particulate form with half-lives greater than 8 days in gaseous effluents released beyond the SITE BOUNDARY shall be limited as follows:

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

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 efDuents exceeding any of the above limits, prepare and submit to the Commission, within 30 days, a Special Report, pursuant to Specification 4.3, which includes:

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

a.

b.

Corrective action taken to reduce the release of tritium and radioactive materials in paniculate form with half-lives greater than 8 days in gaseous efDuents 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.

SURVEILLANCE REQUIREhENTS 1

a 2.8.1 Baseline Comoarison. Cumulative Activity contributions from tritium and radioactive materials in paniculate form in gaseous efDuents shall be compared with the baseline conditions established by the Environmental Repon at least once per 31 days. IF the comparison indicates that the activity released will exceed the Environmental Repon baseline release, THEN a dose calculation shall be performed.

NUCLEAR POWER GENERATION NUMZER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

1 REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-22 SOLID RADIOACTIVE WASTE 2.9 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 all times.

3

{

ACTION:

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.

i l

i i

4 I

4 SURVEILLANCE REQUIREMENTS 2.9.1 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 solidification and dewatering.

NUCLEAR POWER GENERATION NUMBER ODChi HUMBOLDT BAY POWER PLANT VOLUME 4

l REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION hfANUAL PAGE I-23 TOTAL DOSE 2.10 The calendar year dose or dose commitment to any hEMBER 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 times.

ACTION:

With the calculated doses from the release of radioactive materials in liquid or gaseous efIluents exceeding twice the limits of Specification 2.4a,2.4b,2.7a,2.7b,2.8a, or 2.8b, 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 case, prepare and submit to the Commission within 30 days, pursuant to Specification 4.3, a Special Report that defmes the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Pan 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 effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also de: obe levels ofradiation 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 Repon shall j

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.

1 SURVEILLANCE REQUIREMENTS 2.10.1 DOSE CALCULATIONS - Annual dose contributions from liquid and gaseous effluents shall be calculated in accordance with dose calculation methodology provided for Specifications 2.4.1,2.7.1, and 2.8.1.

l 1

NOCLEAR POWER GENERATION NUMBER ODCM Hi)MBOLDT BAY POWER PLANT VOLUME 4

REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-24 REMP MONITORING PROGRAM 1

2.11 The radiological environmental monitoring program shall be conducted as specified in Table 2-7.

APPLICABILITY: At all times.

ACTION:

1 With the radiological environmental monitoring program not being conducted as a.

specified in Table 2-7, prepare and submit to the Commission, in the Annual Radiological Environmental Monitoring Repon, a description of the reasons for not conducting the 4

program as required and the plans for preventing a recurrence.

1 b.

With the level of radioactivity, resulting from plant emuents, in an environmental sampling medium exceeding the reponing 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 Repon pursuant to Specification 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 radionuclides in Table 2-8 are detected in the sampling medium, this repon shall be submitted if:

concentration (1) concentration (2)

.2 1.0 limit level (1) limit level (2)

This report is not required if the measured level ofradioactivity was not the result of plant emuents; however, in such an event, the condition shall be reponed and described in the Annual Radiological Emironmental Monitoring Repon.

When radionuclides other than those in Table 2-8 are detected and are the result of plant emuents, this report 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 repon is not required if the measured level of radioactivity was not the result of plant emuents; however, in such an event, the condition shall be reponed and described in the Annual Radiological Environmental Monitoring Repon.

SURVEILLANCE REQUIREMENTS 2.11.1 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 2-1,2-2,2-3,2-4 and 2-5 and shall be analyzed pursuant to the requirements of Tables 2-7 and 2-9.

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION O

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-25 Tahic 2-7 IIBPP RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PROGRAM DESCRIPTION PROGRAM RASIS Esposure Pathway Number of Sampics Sampling and CoIIcction Type and Frequency of SAFSTOR State of PG& E/HBPP and/9r Sampic and Locations

• Frequency Analysis DP California Elective (OR)

(NQR)

(NQR)

AIRIlORNE 2 cilsite locations m different Contmunus sampler operatmn with Gross beta radioactivity following X

set-tors sample collection at least once per 7 filter change

• days'"

Gamma isotopic" analysis on composite (by station) quarterl/"

DIRECT RADIATION

• 16 onsite stations with ILDs TLDs exchanged quarterl/"

Gamma exposure quartert/

X 1 ofTsite control station with TLD TLDs exchanged quarter!/"

Gamma exposure quarterly

• X 4 ofTsite stations with TLDs 1LDs exchanged quarterl/"

Gamma exposure quarterly" X

24 offsite stations with TLDs TLDs exchanged quartert/"

Gamma exposure quarteriv*

X(5 stations)

X(19 stations)

WATERBORNE Surface Water Discharge canal influent Continuous sampler operation with Gamma isotopic'" and Tritium X

sample collection wecki/"

analysis weekly

• Dip samples if sampler inoperable'"

Discharge canal influent Contirmous sampler operation with Sample submitted to the State X

sample collection monthlf" Department ofIlealth Services Dip samples if sampler inoperable'" monthl/"

Disharge canal influent Contimmm sampler operation with Gamma isotopic'" and Tritium X

sample collection monthiv'"

anahsis monthh*

Dip samples it sample, umpesable'"

Disharge canal clllnent Contima ns sampler operation with Gamma isotopic'" wrckly (TS)'"

X sample collection weekly'"

Gamma isotopic'" and Tritium Dip samples if sampler inoperable'" analysis wecktv*

Disharge canal etlinent Continmm s.unpler opivation with S,nnple submitted to the State X

sample collettion monthl/"

Department ofIlealth Services Dip samples if sampler inoperable'" monthlf" Discharge canal cfllnent Contimmus sampler operation with Gamma isotopic'" and Tritium X

sample collet tion monthh'"

anahxis monthly

• Dip samples il sampler inoperable'"

Discharge canal ellluent Continuous sampler operation with Sample submitted to the EPA as X

sample collectum quatterl/"

part of the EPA national Tritium Dip samples if sampler moperable'" monitoring program Grotmdwater 5 grotmdwater spent fuel pool Quarterl/"

Alpha. Heta, Tntium and gamma X

monitoring wells isotopic"' analysis (piarteilf Dnnking Water I hication (ItilPP) consisting of 2 Quarterly"'

Tritinm analysis and gamma X

wells (domestic water supply) isotopic analysis quarterly

• NUCLEAR POWER GENERATION NUMBER ODChi HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-26.

PROGRAM DESCRIPTION PROGRAM BASIS Esposure Pathway Number of Samples Sampling and Collection Type and Frequency of SAFSTOR State of PG&E/HRPP and/or Sample and Locations" Frequency Analysis DP California Elective (QR)

(NQR)

(NQR)

Sediment 3 locations located m liumboldt Quarterly'"

Gamma isotopic"' analysit X(1 station)

X(2 stations) llay quarterl/"

Algae 3 stations located in IInmboldt flav Quarterly, subiect to armlainhty"'

Gamma isotopic'danalysis quarterly X(1 station)

Xf2 stations)

INGESTION Milk Pedrotti Dairy htonthlv"""

Sample submitted to the State X

Department oflicalth Scnices Pedrotti Dairy Quarterl/"

Gamma isotopic"' analysis X

quarter 1/"

Ilolgerson Dairy Qnarterly"""

Sample submitted to the State X

Department ofIIcalth Services Ifolgerson Dairy Quarterl/"

Gamma isotopic'danahsis X

quarterl/"

I ish and lmertebrates I sample of fish from Station 5%

Quarterly, subicct to availabiht /"

Gamma isotopic"' analysis X

quartet ty'"

I sample of clams from Station 59 Quarterly, subicct to availaNiity'"

Gamma isr*yic"' analysis X

quarterly

I s:nnple of oysters from Station M

(.hiasteily. subiets to avail.ityhts"'

G.unma isotopit"' anahmis X

qu.u ten h*

TERRESTRIAL Sml 2 locations, one near the plant arni Quarterl/"

Gamma isimmic'" analysis X

one from a enntrol location gnarteth*

t Tahic Notations OR - Quality Related NQR - Non-Quality Related

"' Desiations are permitted from the required sampling schedule if specimens are unobtainabic due to circumstances such as harardous conditions. seasonal unavailability, mallimction of automatic sampling equipment and of her Icgitimate reasons. If specimens are unohtainabic due to sampling equipment malfinction, cflort shall be made to complete corrective action prior to the end of the next sampling period. All desiations from the quality-related sampling schedule shall be documented in the Annual Radiological Environmental Monitoring Report. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances. suitabic specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the REMP, and submitted in the next Anmial Radioactive Efiluent Release Report, including a revised figure (s) and table for the REMP reflecting the new location (s) with supporting information identifying the cause of the unavailability of sampics for that pathway andjustifying the section of the new location (sl for obtaining samples. Note: This reporting requirement applies only to the quality-related portion of the REMP.

NUCLEAR POWER GENERA lON NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION O

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-27 L

hhle Notations (Conti- -dl

• For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more

[

desimeters. Film badges should not be used as dosimeters for measuring direct radiation.

• Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attritmtable to the effluents from the facility.

i G

9

_ - _ _ -.,, - - _ _ - - _ - ~

NUCLEAR POWER GENERATION NUMHER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-28 i

Table 7-8 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRAubNS IN ENVIRONMENTAL SAMPLES Analysis Water (pCi/L) 3 H-3 20,000*

Co40 300 Cs-137 50

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

4 i

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4 l

}

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

REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-29 Table 2-9 DETECTION CAPABI.LITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS""M LOWER LIMITS OF DETECTION (LLD)M Airborne Food Water Particulate or Fish Milk Products Sediment Anal 3 sis (pCi/L)

Gases (pCi/m')

(pCi/kg, wet)

(pCi/L)

(pCi/kg, n et)

(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 identifiabic, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological i

Emironmental Operating Report.

• 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.

j

"' The LLD is defined, for purposes of these specifications. as the smallest concentration of radioactive material in a sample that will yield a net count, above system background. that will be detected with 95 percent probability with only 5 percent probability of falsely concludmg 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(dt)

Where:

LLD the "a priori" lower limit of detection as defined above (as pCi per umt 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 picoeurie

=

Y the fractional radiochemical yield (w hen applicable)

=

).

the radioactive decay constant for the particular radionuclide

=

the clapsed time between sample collection (or end of the sample collection period) and time of t =

counting

NUCLEAR POWER GENERATION NUMZER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUALPAGE I-30 Table Notations (Continued)

The value of Se used in the calculation of the LLD for a detection system will be based on observed variance of the background counting rate or of the counting rate of the blank appropriate) rather than on an unverified theoretically predicted variance. In calculating the L radionuclide determined by gamma ray spectrometry, the background will include the typic 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 pres interfering nuclides, or other uncontrollable circumstances may render these LLDs unachiev cases, the contributing factors will be identified and described in the Annual Environment Operating Report.

i Typical values of E, V, Y and t should be used in the calculation. It should be recognized tha defined as a priori (before the fact) limit representing the capability of a measuremen costeriori (after the fact) limit for a particular measurement.

• For surface water samples, a value of 3000 pCi/l may be used.

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-31 Table 2-10 DISTANCES AND DIRECTIONS TO ENVIRONMENTAL MONITORING STATIONS Fadial Direction Radial Distance Station By 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

AO Humboldt Hill Road at Bret Harte Lane SSE 158 0.9 4

A Wood and K Street, Eureka NNE 42 4.0 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 Hill 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 L1 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 River Road'PG&E Gas Reg /Pedrotti Dairy ENE 72 1.4 17 A

Bassford Road at Grauer's Lane E

90 2.0 18 4

6418 Elk River Road ESE 112 2.0 19 A

5399 Noe Avenue hE 45 1.9 21 AO PG&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

Irving Drive. Humboldt Hill SSE 175 1.3 27 A

6700 Bena 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 40 110 kV Line No. I 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 45 AD Humboldt Substation (T17)

ENE 61 5.9 48 O

Holgerson Dairy S

180 51 55 O

HBPP Outfall Canal NNW 338 0.1 56 0

1000 ft North of Outfall Canal Discharge NE 45 0.2 57 0

1000 ft South of Outfall Canal Discharge W

270 0.2 59 0

Hookton Channel SW 225 0.8 65 O

Coast Ovster Company NNE 23 46 Table Notations Code: A Dosimetry Station C Air Particulate Station O Biological Station Note.

• Qualit) Related Station

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NUMBER ODCM H'UMBOLDT BAY POWER PLANT VOLUME 4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-32 i

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_ TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-36 Figure 2-5 HBPP OFFSITE SAMPLING LOCATIONS (CONTINUED)

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NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION 0

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-37 REMP INTERLABORATORY COMPARISON PROGRAM 2.12 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission.

APPLICABILITY: At all times.

ACTION:

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

SURVEILLANCE REQUIREhENTS 2.12.1 A summary of the results obtained from this program shall be included in the Annual Radiological Emironmental Monitoring Report pursuant to Specification 4.1.

NUCLEAR POWER GENERATION NUMZER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - B38 3.0 SPECIFICATION BASES 4

3.1 Radioactive Liauid Emuent Monitorine Instrumentsion Basis The radioactive liquid effluent instrumentation is provided to mocitor and control, as applicable, the releases of radioactive materials in liquid efBuents during actual or potential releases ofliquid efDuents. The alarm / trip setpoints for these instmments shall j

be calculated in accordance with Part II of the ODCM to ensure that the alarm / trip will occur prior to exceeding 10 times the effluent 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 EfUuent Monitoring Instrumentation Basis The radioactive gaseous efDuent instrumentation is provided to monitor the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous efUuents from the plant stack. The alarm setpoints for these instruments are calculated in l

accordance with Part II 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 hEMBER 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 instmmentation is consistent with the requirements of General Design Criteria 60,63, and 64 of Appendix A to 10 CFR Part 50.

3.3 Liauid EfUuent Concentration Basis This specification is provided to ensure that the instantaneous concentration of radioactive materials released in liquid waste effluents beyond the SITE BOUNDARY will be less than 10 times the efDuent concentration limits specified in 10 CFR Part 20 The specification provides operational flexibility for releasing liquid efDuents in concentrations to follow the Section II.A and II.C design objectives of Appendix I 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 II.A design objectives of Appendix I,10 CFR 50, to a hEMBER 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

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - B39 SPECIFICATION BASES 3.4 Liauid Emuent Dose Basis This specification is provided to implement the requirements of Sections II.A. Ill-A and IV.A of Appendix I,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTIDN statement provides the required operating flexibility and at that same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable" (ALARA). The dose calculations in the OFFSITE DOSE CALCULATION hfANUAL (ODCM) implement i

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 i

the actual exposure of a hiEhfBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The equations specified in the OFFSITE DOSE CALCULATIONAL hfANUAL (ODChi) 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 hian 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 Efiluents from Accidental and Routine Reactor Releases for the Purpose ofImplementing Appendix I," April 1977.

3.5 Liauid 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 objective given in Section II.D of Appendix 1 to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were selected as one quaner of the dose design objectives (on a monthly basis) set forth in Section II.A of Appendix I,10 CFR Part 50, for liquid efIluents (3 mrem /yr; 10 mremlyr to any organ).

3.6 Gaseous Efiluents Dose Rate Basis This specification provides reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a hiEhfBER OF THE PUBLIC in an UNRESTRICTED AREA either within or outside the SITE BOUNDARY in excess of the design objectives of Appendix I 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 11, Column 1. The specification provides operational flexibility for releasing gaseous efIluents to satisfy the Section II. A and ll.C design objectives of Appendix 1 to

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION hiANUAL PAGE I-B40 SPECIFICATION BASES 10 CFR 50. For a AEMBER OF THE PUBLIC who may at times be within the SITE 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 ofgaseous 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).

3.7 Gaseous Emuents: Noble Gases Dose Basis This Specification is provided to implement the requirements of Sections II.B, III. 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 emuents 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 hEMBER OF TI-E 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 effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculational of Annual Doses to Man from Routine Releases of Reactor Efiluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111,

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

3.8 Gaseous Emuents Tritium and Radionuclides in Particulate Form Dose Basis This specification is provided to implement the requirements of Sections II.C, III. A, and IV.A of Appendix I,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix 1 to assure that the releases of radioactive materials in gaseous emuent will be kept "as low as is reasonably achievable" (ALARA). The calculational methods specified in the Surveillance Requirements implement the requirements in Section Ill.A of Appendix 1 that conformance with the guides of Appendix I be shown by

~_

N,UCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - B41 SPECIFICATION BASES 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 i

substantially underestimated. The methods for calculating the dose due to the actual release rates of the subject materials are consistent with the me,thodology 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 I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Efiluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheri: conditions. The release rate specifications for radioactive particulates 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 transported 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 Part 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 Repon whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. 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 the purposes of the Special Repon, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Pan 190, the Special Report 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

NUCLEAR POWER GENERATION NUMBER ODCM HUMBOLDT BAY POWER PLANT VOLUME 4

REVISION O

l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I - B42 SPECIFICATION BASES provisions of 40 CFR part 190.11 and 10 CFR Part 20.405c, 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 Specifications 2.3,2.4,2.6,2.7 and 2.8. An individualis not considered a 4

MEMBER OF THE PUBLIC during any period in which he/she is engaged in carryin2 out any operation that is part of the nuclear fuel cycle.

3.11 REMP Monitorine Procram Basis The quality-related ponion of the REMP satisfies the requirements in 10 CFR Parts 20 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 SAFSTOR Environmental Report baseline environmental conditions by conducting onsite and offsite 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 A ssurancefor Radiological Monitoring Programs (Normal Operations)--

Effluent Streams and the Environment.

The SAFSTOR Environmental Report, submitted to the NRC as Attachment 6 to the SAFSTOR license amendment request, established baseline conditions for soil, biota and sediments. In accordance with the NRC approved SAFSTOR Decommissioning Plan, these bcseline conditions will only need to be reestablished prior to DECON if a significant release during SAFSTOR occurs as the result of an accident The LLD's required by Table 2-9 are considered optimum for routine environmental measurements in industrial laboratories The LLD's for drinking water meet the requirements of 40 CFR 141.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION M ANUAL PAGE I-B43 SPECIFICATION BASES 3.12 REMP Interlaboratory Comoarison Program Basis 4

The requirement for panicipation 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 matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results 4

are reasonably valid.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-44 4.0 ADMINISTRATIVE CONTROLS 4.1 Annual Radiological Environmental Monitorine Reoort A report on the Radiological Environmental Monitoring Program shall be prepared annually in accordance with the NRC Branch Technical Position and submitted to the NRC within 90 days ofJanuary 1 of each year. This report should be included as a separate section to the Annual Facility Status and Survey Report required by Technical Specification VII.H.l.

The Annual Radiological Environmental Operating Report should include:

Summaries, interpretations, and an analysis of trends of the results of the quality a.

related Radiological Environmental Monitoring Program activities for the report period.

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

The results of analysis of quality related environmental samples and of quality related c.

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 Environment 11 Monitoring Program.

Legible maps covering all sampling locations keyed to a table giving distances and e.

directions from Unit 3.

f.

The results oflicensee participation in the Interlaboratory Comparison Program and 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 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 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 efDuents (i.e., demonstrated by comparison with a control station or the SAFSTOR Environmental Report).

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l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-45 i

i. A discussion of all analyses in which the LLD required by Attachment 6.6 was not achievable.

1 i

j 1

)

i

)

i O

~

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-46 Table 4-I RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM ANNUAL REPORT

SUMMARY

- EXAMPLE i

Name ofFacility Humboldt Bay Power Plant Unit 3 Docket No.

50-133, OL-DPR-7 Location of Facility Humboldt County, California Reporting Period January 1 -Dm.ui 31,1995 (County, State) l Location with Highest Annual Mean Medlem or Type and Total Number of Pathway Sampled Number of AllIndicator

Name, Control Locations Noureutine (Unit of Analyses 1.mwer Limit of b> cations Distante and Mean (f)"

Mean(f)"

Reported Messumnent)

Perfanned Detection (LLD)

Mean (f)" Range Direction Range Range Measurements AIRBORNE Radioiodine and Not Required N/A N/A N/A N/A Not Required N/A Particulates DIRECT RADIATION Onsite Direct radiation N/A 13.1 t o.1 mR Station T7 14.4 t o.7 mR (4/4)

Not Required 0

(64)

(64/64)

(14.0 - 15.7 mR)

(11.0 - 15.7 mR)

Offsite Direct radiation N/A 12.810.2 mR 1742 Wood, 12.6 i 0.4 mR (4/4)

Not Required 0

(16)

(16/16)

Fortuna (11.0 - 14.1 mR)

(10.6 - 14.1 mR)

I1.2 miles 158' WATERBORNE i

Surface Water Gamma isotopic Co 16 pCi/l

<MDA N/A N/A Not Required 0

Warse c=1 (52)

Cs-137 - I8 pCi/l (0/52) em-t)

Tritium 500 pCi/l

<MDA N/A N/A Not Required 0

(52)

(0/52)

Groundwater Gross Alpha 3 pCi/l 6 i3 pCi/l Monitoring Well 7 i 3 pCi/l(1/4)

<MDA (0/8) 0 of d aine w us)

(20)

(3/20)

No. 6 (N/A)

(N/A)

Gross Beta 4 pCi/l 6 i3 pCi/l Monitoring Well 7i 3 pCi/l(3/4) 7i3 pCi/l(6/8) 0 (20)

(9/20)

No.1 (6 - 8 pCi/l)

(6 - 8 pCi/l) s m.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-47 IAcation with Highest Annual Mean Medium or Type and Total Number of Pathway Sampled Number of All Indicator

Name, Contml Imestions Nonroatine (Unit of Analyses Iewer Limit of Locations Distance and Mean (f)*'

Mean(f)*'

Reported Measurement)

Perfomied Detection (LLD)

Mean (f)*) Range Direction Range Range Measurements Gamma isotopic Co 16 pCill

<MDA N/A N/A

<MDA (0/8) 0 (20)

Cs-137 - 18 pCi/l (0/20)

(N/A)

Tritium 500 pCi/l 9071220 pCi/l Menitoring Well 907i 220 pCi/l 907 i 220 pCi/l 0

(20)

(4/20)

No.1 I (4/4)

(4/8)

(831 - 1100 pCi/l)

(831 - 1100 pCi/I)

Drinking Water Not Required N/A N/A N/A N/A Not Required N/A Sediment Not Required N/A N/A N/A N/A Not Required N/A Algae Not Required N/A N/A N/A N/A Not Required N/A INGESTION Milk Not Required N/A N/A N/A N/A Not Required N/A Fish and Not Required N/A N/A N/A N/A Not Required N/A invertebrates TERRESTRIAL Soil Not Required N/A N/A N/A N/A Not Required N/A

(*' The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net count, abmt system background, that will be detected with 95 percent probability with only 5 percent probability of falsely concluding that a blank obsenation represents a "real" si nal.

F LLD is defined as the a priori lower limit of detection (as pCi per unit mass or volume) representing the capability of a measurement system and not a the a posteriori (after the fact) limit for a particular measurement. (Current literature defines the LLD as the detection capability for the instrumentation only, and the MDC, minimum detectable concentration, as the detection capability for a given instrument, procedure and type of sampic.)

• ) Mean and the tant,e based upon detectable measurements only. Fraction of detectable measurements at specified locations is indicated in parentheses (f);

e g., (10/12) means that 10 out of 12 samples contained detectable activity.

Not Required - not required by the HBPP Unit 3 Technical Specifications. Baseline emironmental conditions for this parameter were established in the SAFSTOR Emironmental Report as referenced by the SAFSTOR Decommissioning Plan.

N/A - Not applicable Note: The example data are based on the 1995 monitoring results and are prmided for illustrative purposes only.

.... - _ - - - - - - - - - - - - - ~ ~ ~~

\\

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-48 4.2 Ananal Radioactive Effluent Rc!== Reoort i

This report shall be submitted prior to April 1 of each year as required by SAFSTOR i

Technical Specification VII.J.3. The following information shall be included:

A summary of the quantities ofradioactive liquid and gaseous effluents and solid a.

waste released from the plant as outlined in Regulatory Guide 1.21, Measuring, i

Evaluating, and Reporting Radioactivity in Solid Wastes aruf Releases of Radioactive Materials in Liquid and Gaseous Effluentsfrom Light-Water-Cooled Nuclear Power Plants, (Rev.1,1974) with data summanzed on a quarterly basis 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., sptnt resin, compacted dry waste) j

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

A list and description ofunplanned releases beyond the SITE BOUNDARY.

c.

d. Information on the reasons for inoperability and lack of timely corrective action for any radioactive liquid or gaseous monitoring instnamentation inoperable for greater than 30 days in accordance with Speci6 cations 2.1 and 2.2.

A summary description ofchanges made to:

e.

1. Process Control Program (PCP)

N.

2. Radioactive Waste Treatment Systems 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.

l j

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l-

_ TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I-49 i

4.3 Soecial Reports The originals of Special Repons 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 repon. These reports shall be submitted covering the activities 3

i identified below to the requirements of the applicable Specification.

i l

a.

Radioactive Effluents - Specifications 2.4,2.5,2.7,2.8 and 2.10.

-l

b. Radiological Environmental Monitoring - Specification 2.11.

4.4 Maior Chanoes to Radioactive Waste Treatment Systems ii 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

j Report for the period in which the evaluation was reviewed. The changes shall be reviewed and approved by the Plant Staff Review Committee.

a f

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. Sufficient information to totally suppon the reason for the change,

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 estimated 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 individual in the UNRESTRICTED AREA and to the general population that i

diflir from those previously estimated in the Environmental Repon,

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

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

I

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 I

PART U - CALCULATIONAL METHODS AND PARAMETERS 1.0 EFFLUENT MONITOR SETPOINT CALCULATIONS l.1 LIQUID EFFLUENT MONITORS i

q 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 20, Appendix B, Table 2, Column 2).

4 1.1.1 The alarm setpoint (countrate) for each monitor is calculated as:

-r s

p' x 10 x (ECL ) x K x 0.85 B

(1-1)

A

=

c F

si+Fu

+

where:

A The alarm setpoint, counts per minute, of the process water monitor or

=

the caisson sump monitor.

F Flow rate past the process water monitor.

=

i F2 Flow rate past the caisson sump monitor.

=

F Flow rate of the efDuent canalinto Humboldt Bay (F1 + F2 + circulating

=

3 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 15% of 3.06 x 10' cpm per micro-Ci/ml.

0.85 Conservatism factor (85 percent of the Specification 2.3 concentration

=

limit).

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

=

minute.

ECLc= Composite Efnuent Concentration Limit (ECL) for the mix of radionuclides (micro-Ci/ml).

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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

i ECL C'

f (1-2)

=

c 1,ECL 1

'L,

, EC i

where:

ECL;= ECL for radionuclide "i" from 10 CFR 20, Appendix B, Table 2, Column 2 (micro-Ci/ml).

Concentration of radionuclide "i" in the mixture.

C;

=

f; Fraction of radionuclide "i" in the mixture.

=

1.1.3 Table 2-2 of Specification 2.1 requires that 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. Therefore, the maximum background allowable (Bm, cpm) is:

K x (5 x 10-5) cpm (1-3)

B_

=

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 1-1 is solved with F = 0 and F3 = F2).

i 1.2 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 effluents, 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).

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-3 l.2.2 Therefore, the alarm setpoint for this limiting condition is the lesser ofAra or Asg calculated for Kr-85. An is calculated as:

r 500 x CcxC A

2

+B

=

(1-4) w

, s Fx 8-xD,xK

<Q) and Asx is calculated as:

3000 x C, x C 2

+B (1-5)

=

,x

,s A

Fx 8-x 'L +(1.1 x M) x K

<Qs where:

4 1

Ara 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.

i Asn 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 doselimit in mrem / year.

4 4

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

=

4 Conversion factor,10 m'/cc.

C

=

2 i

i 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).

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

=

Q meter. As discussed in Appendix B, the value of this parameter is 1.46 x 10" seconds / cubic meter.

D The air dose factor due to gamma exposure in a semi infinite cloud,

=

4 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 for Kr-85.

4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 114 L

The air 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 for Kr-85.

4 The skin dose due to gamma exposure in a semi infinite cloud, mrad / year M

=

per pico Curie / cubic meter. The value of this parameter is given in Table 4

B-1 ofRegulatory Guide 1.109 as 1.72 x 10 f r Kr-85. The associated 9

factor of 1.1 is a unit conversion from mrad to mrem.

K Calibration factor for the monitor. As discussed in Appendix C, the

=

4 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 pomt for the stack monitors should be set at or below 33,400 cpm. 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).

4

_ _ _ _ _ _ _ _ ~ _ _ _ _

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-5 i

2.0 LIQUID EFFLUENT DOSE CALCULATIONS 2.1 CALENDAR QUARTER Speci6 cation 2.4.a requires that the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to UNRF' TRICTED 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.

2.1.1 Compliance with Specification 2.4.a has been established o~n a licensing basis by the Environmental Report submitted to the NRC as Attachment 6 to the SAFSTOR licensing amendment request and NUREG-1166, FinalEnvironmentalStatementfor.

Decommissioning Humboldt Bay Power Plant, Unit No. 3, issued by the NRC.

2.1.2 These repons have demonstrated that neither the routine release of radioactive materials in liquid effluents during SAFSTOR nor the occurrence of an analyzed accident during SAFSTOR would exceed the dose specification of Specification 2.4.a.

2.1.3 Therefore, calculation of dose due to the release of radioactive materials in liquid -

effluents 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 effluents will exceed the Environmental Report baseline release for the current calendar quarter, THEN a dose calculation for the current calendar quaner shall be performed.

2.2 CALENDAR YEAR Specification 2.4.b requires that the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents 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.

2.2.1 Compliance with Specification 2.4.b has been established on a licensing basis by the Environmental Report and NUREG-1166.

2.2.2 These reports have demonstrated that the routine release of radioactive materials in liquid effluents during SAFSTOR will not exceed the dose' specification of Specification 2.4.b.

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

2.2.4 IF a comparison indicates that the activity due to the release of radioactive materials in liquid effluents will exceed the Environmental Report baseline release for the

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-6 current calendar year, THEN a dose calculation for the current calendar year shall be performed.

2.3 LIQUID EFFLUENT DOSE CALCULATION METHODOLOGY 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 effluents. The dose to an organ of an individual from the release of a mixture of radionuclides will be' calculated as follows:

R D=[,Ci x DF x {(Bra. x Ure)+(Bav.i x Uav)}-

(2-1) i.i 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 effluent 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 circulating 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.

Bine,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.

Uinv 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.

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

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-7 1

4 j

not considered to consume either fish or other seafood, and exposure to this age group i

need therefore not be calculated.

l i

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

{

Selected Nuclides from Regulatory Guide 1.109, Table E-11 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI i

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 10-6 4.72 x 10-6 No Data No Data 4.02 x 10-5

\\

i j

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 10-5 1.09 x 10-4 7.14 x 10-5 3.70 x 10-5 1.23 x 10-5 2.11 x 10-6 Y-90 9.62 x 10-9 No Data 2.58 x 10-10 No Data No Data 1.02 x 10-4 Table 2-2 4

Ingestion Dose Factors for Teen Age Group i

(mrem /pico-Curieingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-12 Organ j

Nuclide Bone Liver Total Body Kidney Lung GI-LLI j

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 l

Co-60 No Data 2.81 x 10-6 6.33 x 10-6 No Data No Data 3.66 x 10-5

{

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 10-5 2.12 x 10-6 i

Y-90 1.37 x 10-8 No Data 3.69 x 10-10 No Data No Data 1.13 x 10-4 1

i i

Table 2-3 Ingestion Dose Factors for Child Age Group i

(mrem /pico-Curieingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-13 Organ

)

Nuclide Bone Liver Total Body Kidney Lung GI-LLI j

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 10-5 No Data No Data 2.93 x 10-5 Sr-90 1.70 x 10-2 No Data 4.31 x 10-3 No Data No Data 2.29 x 10-4 i

Cs-137 3.27 x 10-4 3.13 x 10-4 4.62 x 10-5 1.02 x 10-4 3.67 x 10-5 1.96 x 10-6 Y-90 4.jj x 10-8 No Data j.10 x 10-9 No Data No Data 1.17 x 10-4 1

i a

4

- - +,.,

m 4

w a

e-

,w-..

y

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-8 1

Table 2-4 Bioaccumulation Factors for Saltwater Environment i

(pCi/kg per pCi/ liter) i Selected Nuclides from Regulatory Guide 1.109, Table A-1 Element Fish Invertebrate H

9.0 x 10-1 9.3 x 10-1 Co 1.0 x 102 1,o 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)

Other Seafood Fruits and 1

Age Group Fish (Invenebrates)

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

i Table 2-6 Maximum Individual Foods Consumption for Various Age Groups (kilo-gram / year or liters / year)

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 l

0 0

0 330 0

4

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l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-9 3.0 LIQUID WASTE TREATMENT 3.1 TREATMENT REQUIREMENTS -

l 3.l'.1 ODCM Specification 2.5 1

Specification 2.5 requires that liquid radwaste shall be treated, as required, to reduce 4

radioactive materials in liquid wastes prior to their discharge, when projected

[

monthly doses due to liquid efBuents discharged to UNRE'STRICTED AREAS t

would exceed 0.06 mrem whole body or 0.2 mrem to any organ.

i 3.1.2 NPDES Waste Discharge Requirement NPDES Permit No. CA0005622, issued by the California Regional Water Quality l

Control Board - North Coast Region, requires that the discharge ofliquid wastes l

"shall not cause bottom deposits in the receiving waters." The permit also identifies j

Discharge Serial No. 001E (liquid low level radioactive waste) as being filtered j

liquids. Therefore, mimmum treatment ofliquid radwaste shall consist of filtration.

J 3.2 TREATMENT CAPABILITIES 1

j 3.2.1 Liquid Waste Collection i

j Liquid waste is collected in either the turbine building drain tank (TBDT), reactor i

equipment drain tank (REDT), reactor caisson sump or radwaste building sump.

l

a. Turbine Building Drain Tank

- The TBDT, turbine building floor 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 a radwaste receiver tank or drained to the REDT via the caisson floor drain system.

b. Reactor Equipment Drain Tank The REDT and associated REDT pumps are located at the -66 foot level of the reactor caisson access shaft. The contents of this 500 gallon capacity tank are pumped automatically to the radwaste treatment system using either of the two REDT pumps.

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 groundwater in-leakage, has a capacity of 50 gallons. The pump normally

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transfers its contents automatically through a liquid effluent monitor to the Discharge Canal, but may be valved to the radwaste treatment system if necessary of compliance with Specification 2.5 due to groundwater contamination.

d. Radwaste Building Sump The radwaste building sump tank, with a capacity of 2,50 gallons, is located.

beneath the radwaste building floor 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 j

waste receiver tanks.

i 3.2.2 Liquid Waste Treatment System 1

6

]

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 j

collection system and are then pumped to the radwaste building on the north 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:

l waste receiver tanks (3) radwaste concentrator e

radwaste demineralizer e

resin disposal tank e

concentrated waste tanks (2) e waste hold tanks (2) radwaste filters (2) concentrator drip receiver tank e

a. Waste Receiver and Waste Hold Tanks l

The three 7,500 gallon carbon steel radwaste receiver tanks are for wastes coming from the radwaste collection system. Two 7,500 carbon steel waste hold 4

i 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.

{

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

The concentrator consists of a vessel about 14 feet high and 24 inches in diameter with a 40 square foot, callandria-type evaporating section near the bottom. Steam from the Unit 1 or Unit 2 auxiliary steam system is fed to the l

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE II-11 callandria outside of the tubes. Evaporation takes place within the tubes. The concentrator is located in a shielded cubicle in the radwaste building.

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 evajioration is generally the most appropriate method for treatment ofliquids containing high total dissolved solids (TDS).

c. Radwaste Demineralizer The radwaste demineralizer is a single, mixed bed unit with 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.

e. Concentrated Waste Tanks i

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.

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.

I

g. Concentrator Drip Receiver Tank I

I A concentrator drip receiver tank is provided to collect the condensed vapors from the concentrator. The concentrator drip receiver pump either recirculates water in the tank for sample mixing purposes, or it discharges to the treated waste pump discharge header for final disposition.

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)

)

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

I

~

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

l d

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l TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 3 4.0 GASEOUS EFFLUENT DOSE CALCULATIONS 4.1 DOSE RATE 4.1.1 Noble Gases Specification 2.6.a requires that the dose rate at or beyond the SITE BOUNDARY, due to noble gases released in gaseous effluents, 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.

a. Compliance with Specification 2.6.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 Specification 2.6.a.

Therefore, further methodology for the determination of dose rate due to noble c.

gases is not necessary.

4.1.2 Tritium and Radioactive Paniculates Specification 2.6.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 effluents, shall be limited to less than or equal to 1500 mrem / year to any organ.

Compliance with Specification 2.6.b has been established on a licensing basis by a.

the baseline gaseous effluent releases established in the Environmental Repon 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

's nor the occurrence of an analyzed during SAFSTOR will' exceed the dose rate specification of Specification 2.6.b.

c. Therefore, further methodology for the determination of dose rate due tritium and radioactive paniculates with half-lives of greater than 8 days released in gaseous effluents is not necessary.

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i 4.2 DOSE - NOBLE GASES i

4.2.1 Calendar Quarter i

Specification 2.7.a requires that the air dose in UNRESTRICTED AREAS during any calendar quarter due to radioactive noble gases released in gaseous effluents shall be limited to less than or equal to 5 mrad for gamma radiation and less than or i

equal to 10 mrad for beta radiation.

i i

a.'

Compliance with Specification 2.7.a has been established on a licensing basis by j

the Environmental Report and NUREG-1166.

1

b. These reports have demonstrated that the routine release of noble gases in gaseous efDuents during SAFSTOR will not exceed the dose specification of Specification 2.7.a.

l

c. Therefore, calculation of dose during any calendar quarter due to radioactive noble gases released in gaseous effluents 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 efUuents will exceed the Environmental Report baseline release for the current calendar quaner, THEN a dose calculation for the current calendar quarter shall be performed.

4.2.2 Calendar Year Specification 2.7.b requires that the air dose in UNRESTRICTED AREAS during any calendar year due to radioactive noble gases released in gaseous efBuents 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.

a. Compliance with Specification 2.7.b has been established on a licensing basis by the Environmental Repon and NUREG-1166.

b. These reports have demonstrated that the routine release of noble gases in gaseous efBuents during SAFSTOR will not exceed the dose specification of Specification 2.7.b.

c. Therefore, calculation of dose during any calendar year due to radioactive noble gases released in gaseous effluents 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 efBuents will exceed the

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Environmental Report baseline release for the current calendar year, THEN a dose calculation for the current 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 effluents 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:

DWB Whole body (gamma) dose, mrem / year.

=

Ds Skin (beta + gamma) dose, mrem / year.

{

=

x/Q The atmospheric dispersion parameter, seconds per cubic meter.

=

This parameter was calculated (on the basis of historical meteorological monitoring data) for comparison of the plant with 10 CFR 50, Appendix I. In that calculation, the largest value ofX/Q l

was fand to be 1.4 x 104 K

The air 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 ofRegulatory Guide 1.109 as 1.61 x 10-5 for Kr-85.

L The air 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 S Table B-1 of Regulatory Guide 1.109 as 1.34 x 10-3 for Kr-85.

M The skin dose 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 ofRegulatory 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.

Q The average release of Kr-85 in gaseous releases, pico-Curies /sec.

=

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

i 4.3 DOSE - TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM 4.3.1 Calendar Quarter 4

Specification 2.8.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 effluents released to UNRESTRICTED AREAS shall be limited to less than or equal to 7.5 mrem during any calendar quarter,

a. Compliance with Speci6 cation 2.8.a has been established on a licensing basis by.

t the Environmental Report and NUREG-1166.

i l

b. These repons have demonstrated that the routine release of tritium and radioactive materials in paniculate form with half-lives greater than 8 days in gaseous effluents during SAFSTOR will not exceed the dose specification of Specification 2.8.a.

i

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

radioactive materials in particulate form with half-lives greater than 8 days released in gaseous effluents is not necessary for the routine release of noble i

gases during SAFSTOR.

i

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 3

released in gaseous effluents will exceed the Emironmental Report baseline release for the current calendar quarter, THEN a dose calculation for the current calendar quaner shall be p-*med.

4.3.2 Calendar Year Specification 2.8.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 effluents released to UNRESTRICTED AREAS shall be limited to less than or equal to 15 mrem during any calendar year.

Compliance with Specification 2.8.b has been established on a licensing basis by a.

the Emironmental 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 effluents during SAFSTOR will not exceed the dose specification of

^

Specification 2.8.b.

i

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TITLE S AFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE II-17 i

e

c. Therefore, calculation of dose during any calendar year due to tritium and radioactive materials in particulate form with half-lives greater than 8 days.

released in gaseous effluents 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 effluents 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 Paniculate Organ Dose Calculation Summation Methodology The releases ofradioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents will be essentially limited 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:

a D=[ -Q, x (Rinh,i + Ror i + Ru

t. + Re.i + Rw.i).

(4-3) s

..i where:

D Annual dose commitment, mrem / year.

=

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

=

Curies /second.

Rinh,i The dose factor for the inhalation pathway for the radionuclide, i, in

=

units of mrem / year per pico-Curie /sec.

RGP,i The dose factor for the ground plane (direct exposure from

=

deposition) pathway for the radionuclide, i, in units ofmrem/ year per pico-Curie /sec.

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

=

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

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.

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

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 8 than 1 x 104 mrem / year (for direct radiation exposure from material deposited on the ground).

4.3.4 Particulate Inhalation Pathway Dose Calculation Methodology Rm = (y/Q) x BR. x DFt.

(4-3a) where:

x/Q The atmospheric dispersion parameter, seconds / cubic meter... This

=

parameter was calculated (on the basis ofhistorical meteorological monitoring data) for comparison of the plant operating releases with 10 CFR 50, Appendix I. For that calculation, the largest value of x/Q was found to be 1.4 x 104 seconds / cubic meter.

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 body) inhalation dose factor, mrem /pico-Curie,

=

for the receptor age group, a, for the radionuclide, i. The dose factors are given in Tables 4-1,4-2,4-3, and 4-4.

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

K_

unit conversion constant, 8760 hr/yr.

=

DFi The ground plane dose conversion factor for radionuclide, i, in

=

2 mrem /hr per pCi/m from Table 4-5.

SF The shielding factor (dimensionless). Table E-15 ofRegulatory

=

Guide 1.109 suggests values of 0.7 and 0.5 for the maximum individual and for the average population, respectively.

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

=

meters. 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. For that calculation, the largest value of D/Q was found to be 3.0 x 10-9 inverse square meters.

i

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-1 9 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 4.3.6 Particulate Grass-Cow-Milk Pathway Dose Calculation Methodology Rum.i = (D/Q) x (4-3c)

Y 1

4 where:

l QF The cow's vegetation consumption rate. This is given as 50 kg/ day

=

l per Regulatory Guide 1.109, Table E-3.

1 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

=

2 is 0.7 kg/m 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.

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.

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

=

meters. This parameter was calculated (on the basis of historical meteorological monitoring data) for comparison of the plant operating releases with 10 CFR 50, Appendix I. For that calculation, the largest value of D/Q was found to be 3.0 x 10-9 inverse square meters.

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 seconds.

6 4.3.7 Particulate Grass-Cow-Meat Pathway Dose Calculation Methodology

'Qr x U. x Fr x DFt. x W' Rm.u = (D/Q) x (4-3d)

Y

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l TITLE SAFSTOR OFF3ITE DOSE CALCULATION MANUAL PAGE 11-20 where:

QF The cow's vegetation consumption rate of 50 kg/ day per Regulatory

=

Guide 1.109, Table E-3.

2 Ua 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

=

2 is 0.7 kg/m per Regulatory Guide 1.109, Table E-15.

2 DFi,a The ingestion dose factor for radionuclide, i, for the receptor in age

=

j group (a), in mrem /pCi, from Tables 4-8, 4-9, or 4-10, as j

appropriate. Note that this path is not considered to apply to the l

infant age group.

Fr The fraction of the animal's intake of a nuclide which finally appears 1

=

in meat, days / kilogram. This parameter is given in Table 4-13.

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

=

meters. This parameter was calculated (on the basis ofhistorical meteorological monitoring data) for comparison of the plant operating releases with 10 CFR 50, Appendix I. For that calculation, the largest value of D/Q was found to be 3.0 x 10-9 inverse square meters.

W Weathering factor. This is the reciprocal of the weathering time

=

constant given in Regulatory Guide 1.109, for a 14 day removal i

half-life. In this equation, W has the value of 1.74 x 10 seconds.

6 4.3.8 Particulate Vegetation Pathway Dose Calculation Methodology Rv.u = (D/Q) x (4-3e)

Y where:

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

=

This parameter is determined with the default values from Regulatory Guide 1.109, as reproduced in Tables 4-6 and 4-7.

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

=

meters. This parameter was calculated (on the basis ofhistorical meteorological monitoring data) for comparison of the plant operating releases with 10 CFR 50, Appendix 1. For that calculation,

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 1 the largest value of D/Q was found to be 3.0 x 10-9 inverse square meters.

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 10 seconds.

6 Y

The agricultural productivity by unit area ofpasture. This parameter

=

2 is 0.7 kg/m per Regulatory Guide 1.109, Table E-15.

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

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 = Qm x (Ra.m + RGP,H3 + Ru i.s3 + Ruiik.n3 + Rves m)

(4-4) where:

D Annual dose commitment, mrem / year.

=

QH3 The average release rate of H-3, pico-Curies /second.

=

Rinh,H3 The dose factor for the inhalation pathway for H-3, mrem / year per

=

pico-Curie /sec.

RMeat,H3 =

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

RMilk,H3 =

The dose factor for the grass-cow-milk pathway for H-3, mrem / year per pico-Curie /sec.

RVeg, H3 The dose factor for the vegetation consumption pathway, mrem / year

=

per pico-Curie /sec.

This pathway results in trivial offsite calculated radiation exposures. A very 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.

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE I1-22 4.3.10 Tritium Inhalation Pathway Dose Calculation Methodology Rm m =

x BR. x DFm..

(4-4a)

I where:

x/Q The atmospheric dispersion parameter, seconds / cubic meter. This

=

parameter was calculated (on the basis ofhistorical meteorological monitoring data) for comparison of the plant operating releases with 10 CFR 50, Appendix I. For that calculation, the largest value of x/Q was found to be 1.4 x 10-6 seconds / cubic meter.

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.

DFH3,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' Re.m = C x

x Qr x U. x Fm x DF.

(4-4b)

H where:

QF The cow's vegetation consumption rate. This is 50 kg/ day per

=

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.

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.

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.

=

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0.5 The ratin of specific 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. This

=

parameter was calculated (on the basis ofliistorical meteorological monitoring data) for comparison of the plant operating releases with 10 CFR 50, Appendix I. For that calculation, the largest value of x/Q was found to be 1.4 x 104 seconds / cubic meter.

4.3.12 Tritium Grass-Cow-Meat Pathway Dose Calculation Methodology

=(

x Qr x U. x Fi x DF.

(4-4c)

Ru i. n3 x

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.

Fr The fraction of the animal's intake of H-3 which appears in a

=

kilogram ofmeat, with units of days / kilogram. This parameter is given by Table 4-13.

l s

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 Absohite humidity of the atmosphere, 0.008 kilograms / cubic meter,

=

according to Regulatory Guide 1.109.

x/Q

=

The atmospheric dispersion parameter, seconds / cubic meter. This parameter was calculated (on the basis ofhistorical meteorological monitoring data) for comparison of the plant operating releases with

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 4 10 CFR 50, Appendix I. For that calculation, the largest value of x/Q was found to be 1.4 x 104 seconds / cubic meter.

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

'O'75 x 0'5' Rv.m = C x

x Ur x DF.

(4-4d)

H where:

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

=

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

Absolute humidity of the atmosphere,0.008 gm/m' per Regulatory H

=

Guide 1.109.

The fraction of total feed that is water.

0.75

=

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),

=

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

x/Q The atmospheric dispersion parameter, seconds / cubic meter. 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. For that calculation, the largest value of x/Q was found to be 1.4 x 104 seconds / cubic meter.

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

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 10-6 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 jo-6 1.05 x 10-6 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

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 5 Table 4-2 Inhalation Dose Factors for Teen Age Group (mrem /pico-Curieinhaled)

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 10-5 Sr-90 1.35 x 10-2 No Data 8.35 x 10-4 No Data 2.06 x 10-3 9.56 x 10-5 Cs-137 8.38 x 10-5 1.06 x 10-4 3.89 x 10-5 3.80 x 10-5 1.51 x 10-5 1.06 x 10-6

}

Y-90 3.73 x 10-7 No Data 1.00 x 10-8 No Data 3.66 x 10-5 6.99 x 10-5 i

Table 4-3

)

Inhalation Dose Factors for Child Age Group (mrem /pico-Curieinhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-9 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI 11-3 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 10-5 Sr-90 2.73 x 10-2 No Data 1.74 x 10-3 No Data 3.99 x 10-3 9.28 x 10-5 Cs-137 2.45 x 10-4 2.23 x 10-4 3.47 x 10-5 7.63 x 10-5 2.81 x 10-5 9.78 x 10-7 Y 90 1.11 x 10-6 No Data 2.99 x 10-8 No Data 7.07 x 10-5 7.24 x 10-5 Table 4-4 Inhalation Dose Factors for Infant Age Group (mrem /pico-Curieinhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-10 i

Organ l

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 10-5 Sr-90 2.92 x 10-2 No Data 1.85 x 10-3 No Data 8.03 x 10-3 9.36 x 10-5 j

Cs-137 3.92 x 10-4 4.37 x 10-4 3.25 x 10-5 1.23 x 10-4 5.09 x 10-5 9.53 x 10-7 Y-90 2.35 x 10-6 No Data 6.30 x 10-8 No Data 1.92 x 10-4 7.43 x 10-5 i

4

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 6 Table 4-5 i

Extemal Dose Factors for Standing on Contaminated Ground (mrem / hour per pico-Curie / square meter)

)

Selected Nuclides from Regulatory Guide 1.109, Table E-6 j

Total Nuclide Skin Body H-3 Nil Nil Co-60 2.00 x 10-8 1.70 x 10-8 Sr/Y-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 Table 4-6 Average Individual Foods Consumption for Various Age Groups (kilo-gram / year or liters / year)

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 4-7 Maximum Individual Foods Consumption for Various Age Groups (kilo-gram / year or liters / year)

Other Seafood Fruits and l

Age Group Fish (Invenebrates)

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

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 7 Table 4-8 i

Ingestion Dose Factors for Adult Age Group j

(mrem /pico-Curieingested)

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 10-6 4.72 x 10-6 No Data No Data 4.02 x 10-5 Sr-90 7.58 x 10-3 No Data 1.86 x 10-3 No Data No Data 2.19 x 10-4 i

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 10-5 2.11 x 10-6 Y-90 9.62 x 10-9 No Data 2.58 x 10-10 No Data No Data 1.02 x 10-4 Table 4-9 j

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

]

Selected Nuclides from Regulatory Guide 1.109, Table E-12 i

Orjzan 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 10-6 No Data No Data 3.66 x 10-5 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 10-5 2.12 x 10-6 Y-90 1.37 x 10-8 No Data 3.69 x 10-10 No Data No Data 1.13 x 10-4

]

Table 4-10 Ingestion Dose Factors for Child Age Group (mrem /pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-13 Orlzan 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 10-5 No Data No Data 2.93 x 10-5 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 10-5 1.02 x 10-4 3.67 x 10-5 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

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 8 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 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 Co-60 No Data 1,og x 10-5 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 10-5 1.64 x 10-4 6.64 x 10-5 1.91 x 10-6 Y-90 8.69 x 10-8 No Data 2.33 x 10-9 No Data No Data 1.20 x 10-4 j

Table 4-12 Stable Element Transfer Data For Cow-Milk Path (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 10-5 Table 4-13 Stable Element Transfer Data For Cow-Meat Path (days / kilo-gram) 4 Selected Nuclides from Regulatory Guide 1.109, Table E-1 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

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 9 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 summire the calculated doses from the following:

a.

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

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

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 indiviJrial at the site boundary. The only portion of the site boundary where there is significant dire.ct radiation is near the radwaste facilities at the [PG&E] North edge of the site. Due 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 <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> / 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 summing the calculated doses from the following:

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

a.

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.)

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

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.

5.3 DOSE TO OTHER ORGANS I

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-

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-3 0 Stack Noble gas releases, using equation (4-1).

a.

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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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-31 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING SOLIDIFICATION 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 defined in 10 CFR 61. These wastes must be stabilized by solidification and contain no freestanding liquids prior to shipment offsite for burial, or else be packaged in a high integrity container in accordance with Section 7.0.

6.2 PROGRAM ELEMENTS For the disposal of radioactive waste requiring solidification, HBPP shallimplement the following steps:

6.2.1 An NRC approved contract vendor solidification service shall 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.

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 content, waste principal chemical constituents and mixing and cunng times.

6.2.4 The vendor solidification procedure and PCP shall be incorporated into a temporary Plant Manual procedure that will be effective during the solidiScation 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 i

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.

6.2.5 This temporary 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 s! all ensure that the stability requirements of 10 CFR 61.56(b) for wastes exceeding Class A concentrations are met by the vendor solidi 6 cation program.

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TITLE S AFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE II-32

)

7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED IN HIGH 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 stabili' ed by packaging in z

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 PROGP.AM ELEMENTS For disposal of radioactive waste requiring a high-integrity container, HBPP shall implement the following steps:

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

7.2.2 The container shall be demonstrated to have been authorized by the NRC Division of Waste Management 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.

i 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.

I I

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TITLE S AFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-3 3 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED RESINS AND OTHER WET WASTES 8.1 SCOPE This section pertains to bead-type spent radioactive demineralizer resin and other wet wastes, such as absorbed oils, 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.

i 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 other than oil must be solidified or packaged in sufficient absorbent material to absorb twice the volume ofliquid l

8.2.4 Oil must be solidified.

j l

I

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TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1134 9.0 PROGRAM CHANGES 1

9.1 PURPOSE OF THE OFFSITE DOSE CALCULATION MANUAL 1

The Offsite Dose Calculation Manual was developed to support the implementation of the Radiological Effluent Technical Specifications required by 10 CFR 50, Appendix I, and 10 CFR 50.36. The purpose of the manualis to provide the NRC with sufficient information relative to effluent monitor setpoint calculations, effluent related dose calculations, and environmental monitoring to demonstrate compliance with radiological effluent controls.

9.2 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL 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 effective. Records of the reviews performed on change to the ODCM should be documented and retained for the duration of the possession only license.

10 0 PROCEDURE OWNER 10.1 Sr Radiation Protection Engineer

ODCM APPENDIX A Revision 0 Page A-1 APPENDIX A SAFSTOR BASELINE CONDITIONS l

1 l

i I

l ODCM APPENDIX A Revision 0

)

Page A 2 1

1.0 LIQUID AND GASEOUS EFFLUENTS 1.1 LIQUID EFFLUENTS Baseline levels of radioactive materials contained in liquid effluents during the S AFSTOR 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 I

Strontium-90 3.28E-2 2.73E-3 Table A-2 Baseline Gaseous Emuent Activity Type of Activity Annual Release Monthly Average Release (Curies)

(Curies)

Tritium

<l.0E-3

<8.3-5 Particulate Gamma Emitters (total) 3.16E-1 2.63E-2 Strontium-90 3.38E-6 2.82E-7

ODCM APPENDIX B Revisi:n 0 Page B-1 8

a APPENDIX B BASIS FOR INSTANTANEOUS X/Q VALUE i

i h

d i

4 1

ODCM APPENDIX E

~ Revision 0 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 l

CFR 50, Appendix 1. This calculation was performed using the guidance of Regulatory Guide 1.109. For that calculation, the largest value of X/Q was found to be 1.4 x 104 l

seconds / cubic meter.

1.2 That value of the parameter is not appropriate for the monito' ring required by the SAFSTOR Technical Specifications, because it does not represent a "real-time" dispersion coefficient. 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 Nuclear 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 coefficient is that the release from the stack is during " fumigation" conditions (Pasquill stability class F), and j

that the plume proceeds downwind toward the rising terrain of Humboldt Hill. The l

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.

1,4 The calculations include the effects of the stack effluent vertical velocity, which increases the effective stack height. This plume rise is calculated from equation 5.1 of

" Meteorology and Atomic Energy - 1968" (except that the thermal buoyancy term is neglected):

h.=

(d) x [w/u)"

(B-1) where:

h.=

Plume rise, meters d=

Stack nozzle diameter, meters w=

Stack gas exit velocity, meters /see u=

Wind velocity, meters /sec

4 g

ODCM APPENDIX B Revilon 0 Page B-3 1.5 The calculation values are a wind speed of 2 meters /second (commonly assumed for j

fumigation conditions). The stack nozzle diameter of 36 inches (0.914 meters) and stack j

flow rate of 31,000 cfm (14.6 cubic meters /sec result in a gas exit velocity of 22.3

{

meters /second.

Plume Rise 0.914 x (22.3/2)" = 26.7 meters

=

l 1.6 Accordingly, the following calculations use an effective stack height of 103 meters, rather i

than the physical stack height of 76 meters (250 feet).

1.7 1

The concentration calculated for fumigation conditions at ground level is:

l I

t X/Q =

2.507 x U, x s x h, (B-2) 4 y

where:

i U=

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

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

=

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 l

elevation is 107 meters. To account for the rising terrain, the relative elevation is used for the stack effective height.

Distance Ground El.

h.

s X/Q y

(meters)

(meters)

(meters)

(meters)

(sema4/ cubic meter) 500 5

102 20 9.78 x 10

1000 17 90 38 5.83 x 10-2 1500 66 41 54 9.01 x 10

I 1700 91 16 62 2.01 x 10 d

2000 104 3

70 9.50 x 10 d

l.8 The concentration at the plume centerline is:

1 X/Q =

3.1416 x U, x s, x s, (B-3) where:

U. =

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

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

s

=

y

1 ODCM l

APPENDIX B Revisiin 0 Page B-4 Vertical plume spread, meters, from Figure 2 of the Regulatory Guide.

s,

==

Distance s

s, X/Q y

(meters)

(meters)

(meters)

(Ws/ cubic meter) 500 20 8.3 9.59 x 10d 1000 38 13.5 3.10 x 10d 2000 70 21 1.08 x 10" 3000 99 28 5.74 x 10-5 1.9 Based on the assumption that the offsite concentration is not higher than that calculated with either of the two mathematical models, the highest offsite ground level 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 1

ODCM l

APPENDIX B Revision 0 Page B 5 Figure i 153 n

\\

r

\\

/

\\

I

\\

/

l

\\

/

\\

/

I X

1 s

. %.n f

l a c.ni.an.

(164; i

\\

a

\\

J' N

\\

/

'A N.

/

x v

m L

i I

1 E-5 500 1000 1500 2000 2500 3000 Dstance Downwind (meters) l t

l

\\

1 I

ODCM APPENDIX C Revision 0 Page C-1 l

t t

l APPENDIX C Kr-85 MONITOR CALIBRATION l

l t

l I

i l

l i

l l

l l

4 1

._. _ _ _.. ~ _

o a

ODCM APPENDIX C Revisisn 0 Page C-2 l.

1.0 Kr-85 MONITOR CALIBRATION l.1 The original calibration factor was based on the manufacturer's calibration. This 4

calibration was re-exammed after a test was performed to determine the effects of sample line pressure drop on the calibration ofthe stack sampler / monitor. This section documents the results of that test and review.1 i

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.

4 (relative to atmospheric pressure), respectively.

1.1.2 The effect of changing the stack particulate sample filter from ' dirty' 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.

1.1.3 The true system flowrate was found to differ slightly from the flowrate indicated on the Photobelic 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.

i vacuum). The test pressure / flow measurement results are summarized below:

Chamber 'A' Vacuum Photohelic GaugeIndicated C-812 Air Flow 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 with Kr-85 gas standards. The i

standard certificate concentrations were given for the gas at 'STP' (Standard Temperature

]

& Pressure), but the calibration was performed at ' ambient' conditions, without any correction. According to the vendor of the radioactive standard gas, STP conditions are l

760 mm Hg. and 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. 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 conditions would have been lower by a factor of 0.909 due to the lower pressure and higher temperature:

{74 27

= 0.909 8 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 performed on 5/18/93.

ODCM APPENDIX C -

Revision 0 Page C-3 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.:

Gas Gas Detector 'A' Detector 'B' Detector 'A' Detector 'B' C W on C W on Net Net Calibration

' Calibration at STP at Onginal Countrate Countrate Factor Factor (pCi/cc)

Calibration (cpm)

(cpm)

(pCi/cc per

( Ci/ccper Conditions '

cpm) cpm) i (pCi/cc) 1.84E-6 1.67E-6 6.08El 6.42El 2.75E-8 2.61E-8 1.66E-5 1.51E-5 4.88E2 4.98E2 -

3.09E-8 3.03E-8 1.67E-4 1.52E-4 5.10E3 5.39E3 2.98E-8 2.82E-8 1.67E-3 1.52E-3 5.26E4 5.46E4 2.89E-8 2.78E-8 j

1.09E-2 9.91E-3 3.36E5 3.38E5 2.95E-8 2.93E-8 1.4 The effect of the sample line pressure drop (see section 6.1.1) is to reduce the density of i

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 gas in the chambers was at atmospheric pressure. The correction factor for this effect is about 1.08 (29.92/27.74). Ifthis

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correction is applied to the average of the 10 measurements above, the resulting calibration

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would be 1.08 x 2.88E-8 = 3.1lE-8 pC/cc per epm. This is essentially the same value as

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the one originally established (3.lE 8), so the error produced by neglecting the sample line pressure drop effectively canceled out the error 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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