Text

Revision 15 to Safstor Offsite Dose Calculation Manual
	ML091040716
Person / Time
Site:	Humboldt Bay
Issue date:	01/12/2009
From:	; Pacific Gas & Electric Co
To:	; NRC/FSME
References
	Download: ML091040716 (107)
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Enclosure 2 PG&E Letter HBL-09-003 HUMBOLDT BAY POWER PLANT UNIT 3 SAFSTOR OFFSITE DOSE CALCULATION MANUAL INCLUDING CHANGES MADE DURING 2008 9

S S Nuclear Power Generation SECTION ODCM VOLUME 4 Humboldt Bay REVISION 15 EFFEC DATE 1-12-09 Power Plant PAGE i TITLE APPROVED BY SAFSTOR OFFSITE DOSE ORIGINAL SIGNED 10-2-08 CALCULATION MANUAL DIRECTORIPLANT MANAGER / DATE HB NUCLEAR (Procedure Classification - Quality Related)

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

Calculational Methods and Parameters.

Part I contains the specifications for liquid and gaseous radiological effluents (RETS) developed in accordance with NUREG-0473, Draft RadiologicalEffluent 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 of Programmatic Controlsfor RadiologicalEffluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of ProceduralDetails of RETS to the Offsite Dose CalculationManual or to the Process Control Program,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, Table 2, Column 2 effluent concentration limits (ECLs) and replaced the gaseous effluent 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 I 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, Preparationof RadiologicalEffluent Technical Specificationsfor Nuclear Power Plants,and NUREG-0473. This guidance was modified, as appropriate, to reflect the SAFSTOR decommissioning licensing basis contained in the HBPP SAFSTOR Decommissioning Plan, the Environmental Report submitted as Attachment 6 to the HBPP SAFSTOR licensing amendment request and NUREG- 1166, HBPP FinalEnvironmentalStatement.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE ii The ODCM contains the requirements for the REMP. This program consists of monitoring stations and sampling programs based on the SAFSTOR Decommissioning Plan and the Environmental Report which established baseline conditions for soil, biota and sediments. The REMP also includes requirements to participate in an interlaboratory comparison program.

Part II 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 effluents released from HBPP during the SAFSTOR period. Part II also contains the methodology used to determine effluent monitor alarm/trip setpoints which assure that releases of radioactive materials remain within specified concentrations.

The ODCM also contains the Process Control Program (PCP) for solid radioactive wastes, administrative controls regarding the content of the Annual Radiological Environmental Monitoring Program Report, administrative controls regarding the content of the Annual Radioactive Effluent Release Report, and administrative controls regarding major changes to radioactive waste treatment systems.

The ODCM shall become effective after review by the Plant Staff Review Committee and approval by the Plant Manager. 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 required level of radioactive' effluent control 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 Effluent 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE iii TABLE OF CONTENTS PART I - SPECIFICATIONS Section Title Page 1.0 DEFINITIONS I-1 2.0 SPECIFICATIONS 1-8 2.1 Radioactive Liquid Effluent Monitoring Instrumentation 1-8 2.2 Radioactive Gaseous Effluent Monitoring Instrumentation I-11 2.3 Liquid Effluent - Concentration 1-14 2.4 Liquid Effluent - Dose 1-17 2.5 Liquid Waste Treatment I-18 2.6 Gaseous Effluents - Dose Rate 1-19 Deleted 2.8 Gaseous Effluents: Dose - Tritium and Radionuclides in Particulate Form 1-23 2.9 Solid Radioactive Waste 1-24 2.10 Total Dose 1-25 2.11 REMP Monitoring Program 1-26 2.12 REMP Interlaboratory Comparison Program 1-39 2.13 Radioactive Waste Inventory 1-40

.3.0 SPECIFICATION BASES 1-41 3.1 Radioactive Liquid Effluent Monitoring Instrumentation Basis 1-41 3.2 Radioactive Gaseous Effluent Monitoring Instrumentation Basis 1-41 3.3 Liquid Effluent Concentration Basis 1-41 3.4 Liquid Effluent Dose Basis 1-42 3.5 Liquid Waste Treatment Basis 1-42 3.6 Gaseous Effluents Dose Rate Basis 1-42 3.7 Deleted 1-43 3.8 Gaseous Effluents: Tritium and Radionuclides in Particulate Form Dose Basis 1-44 3.9 Solid Radioactive Waste Basis 1-45 3.10 Total Dose Basis 1-45 3.11 REMP Monitoring Program Basis 1-45 3.12 REMP Interlaboratory Comparison Program Basis 1-46, 3.13 Radioactive Waste Inventory Basis 1-46

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE iv PART I - SPECIFICATIONS - (Continued)

Section Title Page 4.0 ADMINISTRATIVE CONTROLS 1-47 4.1 Annual Radiological Environmental Monitoring Report '-47 4.2 Annual Radioactive Effluent Release Report 1-52 4.3 Special Reports 1-53 4.4 Major Changes to Radioactive Waste Treatment Systems 1-53 4.5 Process Control Program Changes 1-54 PART II - CALCULATIONAL METHODS AND PARAMETERS Section Title Page 1.0 EFFLUENT MONITOR SETPOINT CALCULATIONS II-1 1.1 Liquid Effluent Monitors 1.2 Gaseous Effluent Monitor II-4 2.40 LIQUID EFFLUENT DOSE CALCULATIONS II-6 2.1 Month (31 Day Period) II-6 2.2 Calendar Quarter II-6 2.3 Calendar Year II-6 2.4 Liquid Effluent Dose Calculation Methodology II-6 3.40 LIQUID WASTE TREATMENT 11-11 3.1 Treatment Requirements 11-11 3.2 Treatment Capabilities 11-11 4.40 GASEOUS EFFLUENT DOSE CALCULATIONS 11-14 4.1 Dose Rate 11-14 4.2 Deleted 11-14 4.3 Dose - Tritium and Radionuclides in Particulate Form 11-17

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 415 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE v PART II - CALCULATIONAL METHODS AND PARAMETERS - (Continued)

Section Title Page 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE 11-35 5.1 Whole Body Dose 11-35 5.2 Skin Dose 11-35 5.3 Dose to Other Organs 11-36 5.4 Dose to the Thyroid 11-36 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING 11-37 SOLIDIFICATION 7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED 11-38 IN HIGH INTEGRITY CONTAINERS 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED 11-39 RESINS AND OTHER WET WASTES 9.0 PROGRAM CHANGES 11-40 10.0 COMMITMENTS 11-40 11.0 PROCEDURE OWNER 11-40 App. A SAFSTOR BASELINE CONDITIONS A-i App. B BASES FOR ATMOSPHERIC DISPERSION AND DEPOSITION VALUES B-1 App. C Deleted C-1 I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE vi LIST OF TABLES - PART I Table Title' Page 1-1 Frequency Notation 1-6 2-1 Radioactive Liquid Effluent Monitoring Instrumentation 1-9 2-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 1-10 2-3 Radioactive Gaseous Effluent Monitoring Instrumentation 1-12 2-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance 1-13 Requirements 2-5 Radioactive Liquid Waste Sampling and Analysis Program 1-15 2-6 Radioactive Gaseous Waste Sampling-and Analysis Program 1-20 2-7 HBPP Radiological Environmental Monitoring Program . 1-28 2-8 Reporting Levels for Radioactivity Concentrations In Environmental Samples 1-30 2-9 Detection Capabilities for Environmental Sample Analysis Lower Limits Of 1-31 Detection (LLD) 2-10 Distances and Directions To Environmental Monitoring Stations 1-33 4-1 Radiological Environmental Monitoring Report Annual Summary - Example 1-49

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE vii LIST OF TABLES - PART II Table Title Page 1-1 Liquid Effluent Monitor Alarm Setpoints 11-3 2-1 Ingestion Dose Factors for Adult Age Group 11-8 2-2 Ingestion Dose Factors for Teen Age Group 11-9 2-3 Ingestion Dose Factors for Child Age Group 11-9 2-4 Bioaccumulation Factors for Saltwater Environment 11-10 2-5 Average Individual Food's Consumption for Various Age Groups 11-10 2-6 Maximum Individual Foods Consumption for Various Age Groups 11-10 4-1 Inhalation Dose Factors for Adult Age Group 11-29 4-2 Inhalation Dose Factors for Teen Age Group 11-29 4-3 Inhalation Dose Factors for Child Age Group 11-30 4-4 Inhalation Dose Factors for Infant Age Group 11-30 4-5 External Dose Factors for Standing on Contaminated Ground 11-31 4-6 Average Individual Foods Consumption for Various Age Groups II-31 4-7 Maximum Individual Foods Consumption for Various Age Groups II-31 4-8 Ingestion Dose Factors for Adult Age Group 11-32 4-9 Ingestion Dose Factors for Teen Age Group II-32 4-10 Ingestion Dose Factors for Child Age Group II-33 4-11 Ingestion Dose Factors for Infant Age Group II-33 4-12 Stable Element Transfer Data For Cow-Milk Path II-34 4-13 Stable Element Transfer Data For Cow-Meat Path II-34

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE viii LIST OF FIGURES - PART I Figure Title Page 1-1 Site Boundary '-7 2-1 HBPP Onsite TLD Locations 1-34 2-2 HBPP Onsite Monitoring Well Locations 1-35 2-3 HBPP Offsite Sampling Locations 1-36 2-4 HBPP Offsite Sampling Locations (Continued) 1-37 2-5 HBPP Offsite Sampling Locations (Continued) 1-38

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NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 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 BASELINE COMPARISON A BASELINE COMPARISON shall be a comparison of cumulative radioactivity releases for a stated period with the baseline radioactivity release conditions established by the ENVIRONMENTAL REPORT.

1.3 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 encompas's the entire channel including the 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.

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

1.5 CHANNEL FUNCTIONAL TEST

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

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

NUCLEAR POWER GENERATION DEPARTMENT. SECTION ODCM VOLUME '4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION" 15 MANUAL 6 PAGE 1-2 1.6 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.

1.7 FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1-1.

1.8 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 complet'ed or implemented in accordance with specified requirements.

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

1.10 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.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-3 1.12 OFFSITE DOSE CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANUAL contains the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program.

The ODCM also contains the Radioactive Effluent Controls and Radiological Environmental Monitoring Program and descriptions of the information that should be included in the Annual Radiological Environmental Monitoring Report and the Annual Radioactive Effluent Release Report. The ODCM also contains the Process Control Program (PCP) for solid radioactive wastes.

1.13 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, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment, that are required for the system, subsystem, train, component or

-device to perform its function(s), are also capable of performing their related support function(s).

1.14 PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION '15 MANUAL' PAGE 1-4 1.16 RESTRICTED AREA The RESTRICTED AREA is defined by 10CFR20.1003. The physical location(s) of the RESTRICTED AREA shall be defined in plant procedures.

1.17 SITE BOUNDARY The SITE BOUNDARY shall be the boundary of the UNRESTRICTED AREA used in the offsite dose calculations for gaseous and liquid effluents. The SITE BOUNDARY is shown in Figure 1-1. Ingress and egress through the SITE BOUNDARY are controlled by the Company.

1.18 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).

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

1.20 UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area located beyond the boundary of the restricted area controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials and within, at, or beyond the SITE BOUNDARY.

1.21 URANIUM FUEL CYCLE As defined in 40 CFR Part 190.02(b), "URANIUM FUEL CYCLE means the operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public use utilizing nuclear energy, but excludes mining operations, operations at waste disposal 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."

NUCLEAR POWER GENERATION DEPARTMENT -SECTION ODCM VOLUME .4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-5 1.22 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce 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 particulates from the gaseous exhaust stream prior to release to the environment.

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

Vent, used in system names, does not imply a VENTING process.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-6 Table 1-1 FREQUENCY NOTATION Notation Frequency 'Extension Period D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . None W At least once per 7 days. 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months M At least once per 31 days. 7 days Q At least once per 92 days. 22 days SA At least once per 184 days. 45 days A At least once per 365 days. 91 days P Completed prior to each release.

N.A. Not applicable.

'The extension period for a frequency of a week or longeris 25% with a maximum tolerance of 325% for three consecutive periods.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-7 Figure 1-1 SITE BOUNDARY

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-8.

2.0 SPECIFICATIONS 2.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITIONS 2.1.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 2.3 are not exceeded.

APPLICABILITY: At all times ACTION:

a. With di radioactive liquid effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required above, without delay suspend the release of radioactive liquid 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 liquid effluent monitoring instrumentation channels inoperable, take the ACTION shown in Table 2-1. For the instrumentation covered by items 1 and 2 of the table, exert best efforts to return the inoperable instrument(s) to OPERABLE status within 30 days. If the affected instrument(s) cannot be returned to OPERABLE status within 30 days, provide information on the reasons for inoperability and lack of timely corrective action in the next Radioactive Effluent Release Report.

SURVEILLANCE REQUIREMENTS 2.1.2 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 2-2.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-9 Table 2-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument OPERABLE ACTION

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Process Water Monitor 1 21

2. Flow Rate Measurement Devices,

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

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

2.3.1, and

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

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

Otherwise, suspend releases of-radioactive materials via this pathway.

I-

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL -! PAGE 1-10 Table 2-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL Instrument CHECK CHECK CALIBRATION TEST

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

a. Process Water Monitor D Q A Q(1)(2)

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 10-6 micro-Ci/ml of Cs- 137, the cause will be investigated and remedial measures taken to reduce the background reading.

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

a. Instrument indicates measured levels above the alarm setpoint.

b. Circuit failure.

c. Instrument indicates a downscale failure.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOROFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-11 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITIONS 2.2.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 2-3 shall be OPERABLE with their'alarm/trip setpoints set to ensure that the limits of specification 2.6 are not exceeded.

APPLICABILITY: Whefiever the ventilation system is in operation.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip 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 instrumentation 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 2.2.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CALIBRATION operations at the frequencies shown in Table 2-4.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-12 Table 2-3 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument OPERABLE ACTION

1. Stack Gas Monitoring System

a. Noble Gas Activity Monitor* N.A.

b. Iodine Sampler* N.A.

c. Particulate Sampler 1 23,25

d. Effluent System Flow Rate Monitor 1 26

e. Sampler Flow Rate Monitor*"* 1 Table Notation ACTION 23 The particulate sampler 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 2.2.1 .b.

ACTION 24 Deleted ACTION 25 With the number of channels OPERABLE less than that required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided samples are continuously collected as required in Table 2-6.

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

Not included in the stack gas monitoring system.

- Loss of sampler flow would result in alarm and failure of the particulate sampler.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-13 Table 2-4 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE FUNCTIONAL Instrument CHECK CHECK CALIBRATION TEST

1. Stack Gas Monitoring System

a. Noble Gas Activity Monitor* *

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

b. Iodine Sampler* N.A. N.A. N.A. L N.A.

c. Particulate Sampler N/A N.A. N.A. N.A.

d. Effluent System Flow W N.A. A N.A.

Rate Monitor

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

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

a. Instrument indicates measured levels above the alarm setpoint.

b. Instrument indicates a downscale failure. **

Although this sampler is normally required for nuclear plant monitoring, it is not required or included in the HBPP stack gas monitoring system due to the long decay time since operation.

- Although this is a normal requirement of the CHANNEL FUNCTIONAL TEST for operating plants, no downscale failure indication is provided on this instrument at HBPP, and downscale failure indication is not required for the monitor to be OPERABLE.

Although this instrument is normally required for nuclear plant monitoring, it is not required or included in the HBPP stack gas monitoring system as all fuel (noble gas source) has been transferred to the Independent Spent Fuel Storage Installation (ISFSI).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-14 2.3 LIQUID EFFLUENT - CONCENTRATION LIMITING CONDITIONS 2.3.1 The instantaneous concentration of radioactive material released beyond the SITE BOUNDARY shall be less than or equal to 10 times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2.

APPLICABILITY: At all times.

ACTION:

With the instantaneous concentration of radioactive 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 REQUIREMENTS 2.3.2 Radioactive liquid wastes shall be sampled and analyzed in accordance with the sampling and analysis program of Table 2-5.

2.3.3 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 2.3.1.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-15 Table 2-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit Sampling Analysis Type of Activity of Detection Liquid Release Type Frequency Frequency Analysis (LLD)

( Ci/ml)a A. Batch Waste Release Tanksc P P Principal Gamma 5 x 10-7

1. Treated Waste Hold Tank(2) Each Batch Each Batch Emitters'

2. Waste Receiver Tanks(3) P M H-3 1 x IO0 Each Batch Compositeb Gross Alpha 1 x 10.7 P Q Sr-90 5 x 108 Each Batch Compositeb B. Plant Continuous Releasesd D W Principal Gamma 5 x 10i7

1. Caisson Sump Grab Sample Compositeb Emitterse D M H-3 1 x 10-5 Grab Sample Compositeb Gross Alpha 1 x 10"7 D Q Sr-90 5 x 108 Grab Sample Compositeb Table Notation a The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD =

(E) (V) (2.22 x 106) (e-A*t) Y Where:

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-16 Table 2-5 (Continued)

Table Notation (Continued)

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

V is the sample size (in units of mass or volume),

2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

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

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

The LLD is defined as an a pErori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

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

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

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

The principal gamma emitters for which the LLD specification applies exclusively are Co-60 and Cs-137. This does not mean that only these nuclides are to be detected and reported.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL - PAGE 1-17 2.4. LIQUID EFFLUENT - DOSE LIMITING CONDITIONS 2.4.1 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:

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

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

APPLICABILITY: At all times.

ACTION:

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

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

b. Corrective action taken to reduce the release of 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 mrem total body and less than or equal to 10 mrem to any organ during the calendar year.

SURVEILLANCE REQUIREMENTS 2.4.2 At least once per 31 days, perform a dose calculation for the current calendar quarter and the current calendar year, OR perform a BASELINE COMPARISON for liquid effluent radioactivity released.

to date for the current calendar quarter and current calendar year. IF the comparison indicates that the activity released to date exceeds the Environmental Report baseline annual release, THEN a dose calculation shall be performed for the current calendar quarter and the current calendar year.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-18 2.5 LIQUID WASTE TREATMENT LIMITING CONDITIONS 2.5.1 The LIQUID RADWASTE TREATMENT SYSTEM shall be used, as appropriate, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to liquid effluents discharged to Humboldt Bay would exceed the action levels of 0.06 mrem whole body or 0.2 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

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

a. Identification of inoperable equipment and the reasons for inoperability.

b. Actions taken to restore the inoperable equipment to OPERABLE status.

c. Actions taken to prevent recurrence.

SURVEILLANCE REQUIREMENTS 2.5.2 Before approving any release, perform a BASELINE COMPARISON for liquid effluent radioactivity released (or projected to be released) during the 31 day period prior to and including the projected release. IF the comparison indicates that the activity released will exceed the Environmental Report baseline monthly release, THEN a dose calculation shall be performed for comparison with Specification 2.5.1.

OR Before approving any release, a dose calculation shall be performed for comparison with Specification 2.5.1.

OR The LIQUID RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in liquid wastes prior to their discharge.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-19 2.6 GASEOUS EFFLUENTS - DOSE RATE LIMITING CONDITIONS 2.6.1 The dose rate at or beyond the SITE BOUNDARY, due to radioactive materials released in gaseous effluents, shall be limited as follows:

a. Tritium and radioactive particulates with half-lives of greater than 8 days:

less than or equal to 1500 mrem/year to any organ.

APPLICABILITY: At all times.

ACTION:

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

SURVEILLANCE REQUIREMENTS 2.6.2 Stack monitoring is not required for noble gases because the spent fuel (noble gas source term) has been transferred to the ISFSI.

2.6.3 The dose rate limit for Tritium in gaseous effluents is not likely to be exceeded, as explained in BASES section 3.6. Tritium monitoring is not required in gaseous effluents.

2.6.4 Radioactive particulates, with half-lives of greater than 8 days, in gaseous effluents released to the environment shall be sampled and analyzed in accordance with the sampling and analysis program of Table 2-6, and their concentrations shall be compared with the limits of 10CFR20, Appendix B, Table 2, Column 1. IF their concentrations exceed those limits, the calculational methods in Part II of the ODCM shall be used to determine whether or not the limits of Specification 2.6.1 have been exceeded. The actual sample period shall be used to determine the dose rate during the sample period.

Table 2-6 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit Sampling Analysis Type of Activity of Detection Gaseous Release Type Frequency Frequency Analysis (LLD)

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-20

(,tCi/ml)a Plant Stack Continuous d Mc Principal Gamma I x 10"ll Particulate Emitters' Sample Continuousd M Gross Alpha I x 10 Composite Particulate Sample Continuousd Q Sr-90 1 x 10"11 Composite Particulate Sample Table Notation a The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" ,signal.

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

LLD =

(E) (V) (2.22 x 106) (e-XAt) y Where:

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

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

-j

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE. SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-21 Table 2-6 (Continued)

Table Notation (Continued)

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

V is the sample size (in units of mass or volume),

3 2.22 x 106 is the number of disintegrations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),

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

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

The LLD is defined as an a prio (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

b Deleted.

CSamples shall be changed at least once per 31 days (7 day extension permitted).

d The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with the Specifications 2.6, and,2.8.

The principal gamma emitters for which the LLD specification applies exclusively are Co-60 and Cs-137 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are not detected for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the' required dose calculations. I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE- 1-22 2.7 Deleted

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-23 2.8 GASEOUS EFFLUENTS: DOSE - TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITIONS 2.8.1 The dose to a MEMBER OF THE PUBLIC from the release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents released beyond the SITE BOUNDARY shall be limited as follows:

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

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

APPLICABILITY: At all times.

ACTION:

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

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

b. Corrective action taken to reduce the release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents 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 REQUIREMENTS 2.8.2 At least once per 31 days, perform a dose calculation for the current calendar quarter and the current calendar year, for the' release of radioactive materials in particulate form with half-lives greater than 8 days, OR perform a BASELINE. COMPARISON for gaseous effluent radioactivity (particulate form) released to date for the current calendar quarter and current calendar year. IF the comparison indicates that the activity released to date exceeds the Environmental Report baseline annual release, THEN a dose, calculation shall be performed for the current calendar quarter and the current calendar year.

As explained in Specification Bases section 3.8, neither routine surveillance nor dose calculations are required for Tritium in gaseous effluents.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-24 2.9 SOLID RADIOACTIVE WASTE LIMITING CONDITIONS 2.9.1 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.

APPLICABILITY: At all times.

ACTION:

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

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

NUCLEAR POWER; GENERATION DEPARTMENT SECTION. ` ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL. PAGE 1-25 2.10 TOTAL DOSE LIMITING CONDITIONS 2.10.1 The calendar year dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 2.4.1 .a, 2.4.1 .b, 2.8.1 .a, or 2.8.1 .b, calculations should be made, which include direct radiation contributions from Unit No. 3, 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 Administrative Control 4.3, a

'Special Report that defines the corrective action to be taken to reduce subsequent releases-to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations' of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s)

.exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190.

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

SURVEILLANCE REQUIREMENTS 2.10.2 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, and 2.8.1.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 15 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-26 2.11 REMP MONITORING PROGRAM LIMITING CONDITIONS 2.11.1 A radiological environmental monitoring program shall be provided to monitor the radiation and radionuclides in the environs of the facility. The program shall be conducted as specified in Table 2-7.

APPLICABILITY: At all times.

ACTION:

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

b. With the level of radioactivity, resulting from plant effluents, in an environmental sampling medium exceeding the reporting levels of Table 2-8 when averaged over any calendar quarter, prepare and submit to the Commission, within 30 days of obtaining analytical results from the affected sampling period, a Special Report pursuant to Administrative Control 4.3, which includes an evaluation of any release conditions, environmental factors or other aspects which caused the limits of Table 2-8 to be exceeded. When more than one of the radionuclides in Table 2-8 are detected in the sampling medium, this report shall be submitted if:

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-27 2.11 REMP MONITORING PROGRAM - Continued When radionuclides other than those in Table 2-8 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is greater than or equal to the calendar year limits of Specifications 2.4 and 2.8. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental ,Monitoring Program Report.

SURVEILLANCE REQUIREMENTS 2.11.2 The radiological environmental monitoring samples shall be collected pursuant to Table 2-7 from the "Quality Related" 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 DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL REVISION 15 PAGE 1-28 Table 2-7 HBPP RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PROGRAM DESCRIPTION PROGRAM BASIS Exposure Pathway Number of Samples Sampling and Collection Type of Analysis ODCM State of PG&E/HBPP and/or Sample and Locations(a) Frequency Specs California Elective (QR) (NQR) (NQR)

AIRBORNE I offsite location Continuo'us sampler operation with Gross beta radioactivity 2

following X(5) X sample collection at least once per filter change( )

7 days(l) Gamma isotopic(c) analysis on quarterly composite (by station)(2)

DIRECT RADIATION(b) 16 onsite stations, at or within the TLDs exchanged quarterly(l) Gamma exposure(3) X SITE BOUNDARY fenceline, with TLDs I offsite control station with TLD TLDs exchanged quarterly(l) Gamma exposure(3 ) X 4 offsite stations, representing a TLDs exchanged quarterly(1 ) Gamma exposure(3) X X( 5) gradient downwind in the prevailing wind direction, with TLDs 23 offsite stations with TLDs TLDs exchanged quarterly(l) Gamma exposure(3) X WATERBORNE Surface Water Discharge canal effluent Continuous sampler operation with Gamma isotopic(c) and Tritium X X sample collection weeklyM) analysis of weekly sample(2)

Dip samples if sampler Sample submitted to the State inoperable(1) Department of Health Services monthly(t)

Groundwater 5 groundwater spent fuel pool Quarterly Tritium and 2

gamma isotopic(c) X monitoring wells analysis( )

Alpha and Beta Analysis(2) X 2

Sediment 3 locations located in Humboldt Quarterly(4) Gamma isotopic(c) analysis( ) X Bay Algae 3 stations located in Humboldt Quarterly, subject to availability(4) Gamma isotopic(c) analysis(2) X Bay

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL REVISION 15 PAGE 1-29 Table 2-7 (Continued)

PROGRAM DESCRIPTION PROGRAM BASIS Exposure Pathway Number of Samples Sampling and Collection Type of Analysis ODCM State of PG&E/HBPP and/or Sample and Locations(a) Frequency Specs California Elective (QR) (NQR) (NQR)

-INGESTION 1

Milk Pedrotti Dairy Annually0 ) Gamma isotopic(c) analysis(2) X Holgerson Dairy Annuallyt') Gamma isotopic(c) analysis(2 ) X Fish and Invertebrates I sample of fish from Station 55 Quarterly, subject to availability(4) Gamma isotopic(c) analysis(2) X I sample of clams from Station 59 Quarterly, subject to availability(4) Gamma isotopic(c) analysis( 2 ) X I sample of oysters from Station Quarterly, subject to availability(4) Gamma isotopic(c) analysis(2) X 65 TERRESTRIAL Soil 2 locations, one near the plant and Quarterly(4 ) Gamma isotopic(') analysis(2 ) X one from a control location Table Notations -

QR - Quality Related (')Performed by HBPP (3)Performed by DCPP (5)Performed by Humboldt Co. Health Dept.

NQR - Non-Quality Related (2)Performed by Offsite Laboratory (4)Performed by Humboldt State University Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the quality-related sampling schedule shall be documented in the Annual Radiological Environmental Monitoring Program Report. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances, suitable 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 Annual Radioactive Effluent 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 samples for that pathway and justifying the section of the new location(s) for obtaining samples. Note: This reporting requirement applies only to the quaiity-related portion of the REMP.

(b) At least 4 additional TLDs are deployed, one in each cardinal direction along the ISFSI fenceline, when fuel is in storage (c) Gamma isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the facility.

NUCLEAR POWER GENERATION DEPARTMENT SECTION' ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-30 Table 2-8 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Analysis Water (pCi/L)

H-3 20,000*

Co-60 300 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-31 Table 2-9 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS(a) (b)

LOWER LIMITS OF DETECTION (LLD)(c)

Airborne Food Water Particulate Fish Milk Products Sediment Analysis (pCi/L) (pCi/m 3) (pCi/kg, wet) (pCi/L) (pCi/kg, wet) (pCi/kg, dry)

Gross Beta 4 0.01 H-3 2 0 0 0 (d)

Co-60 15 130 Cs-137 18 0.06 150 18 80 180 Table Notations (a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Monitoring Program Report.

(b) 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.

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

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

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

Where:

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

Sb = 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 DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-32 Table 2-9 (Continued)

Table Notations (Continued)

E = the counting efficiency (as counts per transformation)

V the sample size (in units of mass or volume) 2.22 = the number of transformations per minute per pico-Curie Y = the fractional radiochemical yield (when applicable)

= the radioactive decay constant for the particular radionuclide At = the elapsed time between sample collection (or end of the sample collection period) and time of counting The value of Sb used in the calculation of the LLD for a detection system will be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background will include the typical contributions of other radionuclides normally present in the samples (e.g., potassium 40 in milk samples).

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

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

(d) For surface water samples, a value of 3000 pCi/L may be used.

NUCLEAR. POWER GENERATION DEPARTMENT SECTION 'ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15

MANUAL PAGE 1-33

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

.AO King Salmon Picnic Area W 270 0.3

2 A 180 Dinsmore Drive, Fortuna SSE 158 9.4.

3 AD Humboldt Hill Road at Bret Harte Lane SSE 158 0.9 4 A Wood and K Street, Eureka NNE 42 4.0 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 1.1 12 A Bell Hill Road and Sauters SSW 195 0.7

14 A South Bay School Parking Lot S 180 0.4 1,6 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 A 6418 Elk River Road ESE 112 2.0

19. A 5399 Noe Avenue NE 45 1.9 21 AO PG&E Well 2, HH Road 6.5 ESE 128 22 A Station B -: 14th Street NNE 23 4.0 24 A Pole at 7 th and L Street NNE 32 5.0

25 "" A Irving Drive, Humboldt Hill SSE 175 1.3 27- A 6700 Berta Road ESE 125 1.9 28 A 7200 Berta Road SSE 142 '2.1 29 A Vista'Road, Humboldt Hill SSE 148 1.5 31 A King Salmon Road East of Freeway SSE, 170 0:4 32 A Loma Road at Tip Top Club SSW 185 0.5 34 A King Salmon Road and RR Track, SSW 185 0.3 36 A Plant Entrance Road WSW 230 0*2 45 A Humboldt Substation (T17) ENE 61 5.9 48 0'. Holgerson Dairy S 180 5.1 55 0. HBPP Outfall Canal, NNW 338/ 0.1 56 0 *1000 ft North of Outfall Canal Discharge NE 45 0.2 57 "O 1000 ft South of Outfall Canal Discharge W 270 0.2 59 0 Hookton Channel SW 225 0.8

65. 0 Coast Oyster Company NNE 23 4.6

A ISFSI Fenceline ',

At least 4 additional TLDs are used, one in each direction, at the ISFSI Fenceline, when fuel is in storage.

Table Notations Code: A Dosimetry Station 0] Air Particulate Station 0 Biological Station Note: *Quality Related Station

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION VO N 4 MANUAL REVISION 1:5 PAGE 1-34 Figure 2-1 HBPP ONSITE TLD LOCATIONS

77 HIGHWAY 101

*At least 4 additional TLDs are used, one in each direction, at the ISFSI Fenceline, when fuel is in storage

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM

/

TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE f-35 Figure 2-2 HBPP ONSITE MONITORING WELL LOCATIONS

\

MW-1 0 Fl ~nqpyz~~~

IL

. S.

m RAILROAD 4I[Y1 WT-I r S SEPARATOR INTAKE STRUCTURE r LEGEND (Tide monitoring

Monitoring Well Location station]

-.*- Apparent Groundwater Flow Direction

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-36 Figure 2-3 HBPP OFFSITE SAMPLING LOCATIONS

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-37 Figure 2-4 HBPP OFFSITE SAMPLING LOCATIONS (CONTINUED)

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-38 Figure 2-5 HBPP OFFSITE SAMPLING LOCATIONS (CONTINUED)

Loleta Fortuna

NUCLEAR POWER GENERATION DEPARTMENT,, SECTION. ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-39 2.12 REMP INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITIONS.

2.12.1 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program.

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 Program Report.

SURVEILLANCE REQUIREMENTS 2.12.2 A summary of the results obtained from this program shall be included in the Annual Radiological Environmental Monitoring Program Report pursuant to Administrative Control 4.1.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-40 2.13 RADIOACTIVE WASTE INVENTORY LIMITING CONDITIONS 2.13.1 Liquid Radioactive Waste In Outdoor Tanks The radiological inventory of wastes in outdoor tanks that are not capable of retaining or treating tank overflows shall not exceed 0.25 Ci.

APPLICABILITY: At all times.

ACTION:

When the inventory exceeds the conditions as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Monitoring Program Report.

2.13.2 Solid Radioactive Waste The radiological inventory of wastes within the solid radioactive waste system shall not exceed 1000 Ci.

APPLICABILITY: Atiall times.

ACTION:

When the inventory exceeds the conditions as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Monitoring Program Report.

SURVEILLANCE REQUIREMENTS 2.13.3 A review of the estimated radioactive waste inventory shall be performed on a semi-annual basis.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-41 3.0 SPECIFICATION BASES 3.1 Radioactive Liquid Effluent Monitoring Instrumentation Basis The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated 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 Effluent Monitoring Instrumentation Basis The radioactive gaseous effluent instrumentation is provided to monitor the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents from the plant stack. The alarm setpoints for these instruments are calculated in 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 MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY of 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. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix Ato 10 CFR Part 50.

3.3 Liquid Effluent 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 effluent concentration limits specified in 10 CFR Part 20. The specification provides operational flexibility for releasing liquid effluents 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 MEMBER OF THE PUBIC and (2) the limits of 10 CFR 20.1302 to the population. This specification does not affect the requirement to comply with the annual limitations of 10 CFR 20.1301 (a).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-42 3.4 Liquid Effluent Dose Basis This specification is provided to implement the requirements of Sections II.A. 111-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 ACTIONstatement 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 MANUAL (ODCM) implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

Compliance with this Specification has been established on a licensing basis by the SAFSTOR Environmental Report and NUREG-1 166, "Final Environmental Statement for Decommissioning Humboldt Bay Power Plant." These reports have demonstrated that routine releases of radioactive materials in effluents during SAFSTOR will not cause the Specification to be exceeded. As long as routine releases do not exceed the baseline quantities evaluated in these reports, no further dose calculation is necessary.

3.5 Liquid Waste Treatment Basis The requirement that these systems be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as reasonably achievable" (ALARA). This specification implements the requirements of 10 CFR Part 50.3 6a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the. liquid radwaste treatment system were selected as one quarter of the dose design objectives (on a monthly basis). set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents (3 mremryr; 10 mrem/yr to any organ).

3.6 Gaseous Effluents Dose Rate Basis This specification provides reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA either within or outside the SITE BOUNDARY 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 II, Column 1. The specification provides operational flexibility for releasing gaseous effluents to satisfy the Section II.A and II.C design objectives of Appendix I to 10 CFR 50.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-43 For a MEMBER OF THE PUBLIC who may at times be within the SITE BOUNDARY, the period of occupancy (which is bounded by the maximumoccupational period while working in Units 1 or 2) will be sufficiently low to compensate for the reduced atmospheric dispersion of gaseous effluents relative to that for the SITE BOUNDARY.

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

The only tritium source term is the spent fuel pool water, which evaporates and is released from the stack as moisture in the air. The spent fuel pool water has a Tritium concentration below lx10-4 micro-Curies/ml, and air at 100 'F, saturated with moisture, can not hold more than 5x10 5 grams of moisture per cc. Therefore, it is unlikely that the Tritium concentration in the gaseous effluent could exceed 5xl0. 9 micro-Curies/cc. This is well below the 10CFR20 Effluent Concentration Limit of lxl0.7 micro-Curies/cc, which corresponds to a dose of 50 mremlyear, so it is not necessary to monitor for Tritium in the plant stack effluent stream.

3.7 Stack monitoring is not required for noble gases because the spent fuel (noble gas source term) has been transferred to the ISFSI.

3.8 Gaseous Effluents: 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 I to assure that the releases of radioactive materials in gaseous effluent will be kept "as low as is reasonably achievable" (ALARA). The calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

Compliance with this Specification has been established on a licensing basis by the SAFSTOR Environmental Report and NUREG- 1166, "Final Environmental Statement for Decommissioning Humboldt Bay Power Plant." These reports have demonstrated that

p NUCLEAR POWER GENERATION DEPARTMENT, SECTION ODCM VOLUME REVISION 415 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-44 routine release of Tritium and radioactiye materials in particulate form (with half-lives greater than 8 days) in gaseous effluents during SAFSTOR will not cause the Specification to be exceeded, As long as routine releases do not exceed the baseline quantities evaluated in these reports, no further dose calculation is necessary. Also, the ventilation system has since been modified to provide a full flow HEPA filtration system, significantly reducing routine particulate stack releases.

The only tritium source term is the spent fuel pool water, which evaporates and is released from the stack as moisture in the air. The spent fuel pool water has a Tritium concentration below 1x10. 4 micro-Curies/ml, and an evaporation rate less than 50 gallons per day, so the routine Tritium release rate is below 7 milli-Curies/year. Using this value, the equations in section 4.3.9 through 4.3.13 calculate a maximum annual dose of 1.08 x 10.5 milli-remryear, so it is not necessary to calculate doses for Tritium in the plant stack effluent stream.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL' PAGE 1-45 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' Report 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 Report, 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 Part 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 provisions of 40 CFR part 190.11 and 10 CFR Part 20.2203a4, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 2.3, 2.4, 2.6, 2.7 and 2.8. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

3.11 REMP Monitoring Program Basis The quality-related portion of the REMP satisfies the requirements in 10 CFR Parts 20, 50, and 72.44(d) that radiological environmental monitoring programs be established to provide data on measurable levels of radiation and radioactive materials in the site environs. It is required to provide assurance that the baseline conditions established by the Environmental Report are not deteriorating and 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-46 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 Assurance for RadiologicalMonitoringPrograms(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 baseline 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.

3.12 REMP Interlaboratory Comparison Program Basis The requirement for p'articipation 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 are reasonably valid.

3.13 Radioactive Waste Inventory Basis The requirements for limits on the accumulation of liquid radioactive waste in outdoor tanks and of solid radioactive waste were transferred from the license Technical Specifications.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-47 4.0 ADMINISTRATIVE CONTROLS 4.1 Annual Radiological Environmental Monitoring Report A report on the SAFSTOR Radiological Environmental Monitoring Program shall be prepared annually in accordance with the NRC Branch Technical Position and submitted to the NRC by May 1 of each year.

The Annual Radiological Environmental Monitoring Report shall include:

a. Summaries, interpretations, and an analysis of trends of the results of the quality related Radiological Environmental Monitoring Program activities for the report period. The material provided shall be consistent with the objectives outlined in the ODCM, and in 10CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

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

c. The results of analysis of quality related environmental samples and of quality related environmental radiation measurements taken during the period pursuant to the locations specified in Table 2-7 summarized and tabulated in the format of Table 4-1, Radiological Environmental Monitoring Program Report Annual Summary, or equivalent. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in the next annual report.

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

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

f. The results of licensee 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-48

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

i. A discussion of all analyses in which the LLD required by Table 2-9 was not achievable.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL REVISION 15 PAGE, 1-49 Table 4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

- EXAMPLE Name of Facility Humboldt Bay Power Plant Unit 3 Docket No. 50-133, OL-DPR-7<

Location of Facility Humboldt County, California Reporting Period January 1 - December 31, 1997 (County, State)

Medium or Type and Total All Indicator Location with Highest Annual Control Locations Mean Locations Number of Pathway Sampled Number of Lower Limit Mean, Name,' Mean, Mean, (Fraction) Nonroutine

[Unit of Measurement] Analyses of Detectiona (Fraction) Distance and (Fraction) & [Range, b Reported Performed (LLD) & [Range] bIb Direction & [Range. Maueet Measurements AIRBORNE Particulates Not Required N/A N/A N/A N/A Not Required N/A DIRECT RADIATION

[mR/quarter] Direct radiation 3 13.6 +/- 0.1 Station T7 15.4 +/- 0.2 12.7 +/- 0.3 0 (64) (64/64) (4/4) (4/4)

[11.8- 1.7.5] [13.8- 17.5] [12.5- 12.9]

WATERBORNE Surface Water Gamma isotopic Co-60: 15 <MDA N/A N/A Not Required 2 (Discharge canal effluent) (54) Cs-137:.18 (0/54)

[pCi/l] " [N/A] --.-

Tritium (54) 500 <MDA N/A N/A Not Required 2.

(0/54)

[N/A]

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL REVISION 15 PAGE 1-50 TABLE 4-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

All Indicator Location with Highest Annual Control Medium or Type and Total Locations Mean Locations Number of Pathway Sampled Number of Lower Limit Mean, Name, Mean, Mean, (Fraction) Nonroutine

[Unit of Measurement] Analyses of Detectiona (Fraction) Distance and (Fraction) & [Range] b Reported Performed (LLD) & [Range, b Direction & [Rangel b Measurements WATERBORNE (continued)

Groundwater Gross Alpha 37 6 Monitoring Well 7+/-6 N/A 2 (Monitoring wells) (22) (1/22) No. 2 (1/4) (0/4)

[pCi/1] [7---7] [7- 7] [N/A]

Gross Beta 4 8+/-2 Monitoring Well 10 +/- 3 10 +/- 3 2 (22) (9/22) No. 11 (3/6) (3/6)

[7- 5j--------- [7- 15]

Gamma isotopic Co-60: 15 <MDA N/A N/A N/A 2 (22) - Cs-137: 18 (0/20) (0/4)

[N/A] [N/Aj Tritium 500 (15/22) 461 +/- 64 Monitoring Well 484 +/- 94 444 +/- 88 2 (22) 200 (7/22)C (7/22) No. 1 (3/5) (4/5)

[299- 601] [409- 589] . [2 99- 60 11 DrinkingWater Not Required -- N/A N/A N/A N/A N/A ------ Not Required -- N/A Sediment Not Rectuired 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 invertebrates Not Required N/A N/A N/A N/A Not Required N/A TERRESTRIAL Soil Not Required N/A N/A N/A N/A Not Required N/A

NUCLEAR POWER GENERATION DEPARTMENT SECTION VOLUME ODCM 4

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL REVISION 15 PAGE 1-51 TABLE 4-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

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

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 as 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 MDA, minimum detectable concentration, as the detection capability for a given instrument, procedure and type of sample.) The actual MDA for these analyses was at or below the LLD.

b The mean and the range are based on detectable measurements only. The fraction of detectable measurements at specified locations is indicated in parentheses; e.g., (10/12) means that 10 out of 12 samples contained detectable activity. The range of detected results is indicated in brackets; e.g., [23-34].

C Tritium samples taken 10/24/97 and 11/18/97 were analyzed to a lower than normal LLD of 200 pCiI.

Not Required - not required by the HBPP Offsite Dose Calculation Manual. Baseline environmental conditions for this parameter were established in the Environmental Report as referenced by the SAFSTOR Decommissioning Plan.

N/A - Not applicable Note: The example data are based on the 1997 monitoring results and are provided for illustrative purposes only.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-52 4.2 Annual Radioactive Effluent Release Report This report shall be submitted prior to April 1 of each year. The following information shall be included:

a. A summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant as outlined in Regulatory Guide 1.21, Measuring, Evaluating,and Reporting Radioactivity in Solid Wastes andReleases ofRadioactive Materials in Liquid and GaseousEffluents from Light-Water-CooledNuclear Power Plants, (Rev. 1, 1974) with data summarized on a quarterly basis following the format of Appendix B thereof. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10CFR 50.36a and 10CFR Part 50, Appendix I, Section IV.B.I.

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

1. Container Volume

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

3. Principal Radionuclides (specified as measured or estimated)

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

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

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

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

e. A summary description of changes made to:

1. Process Control Program (PCP)

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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 1-53 4.3 Special Reports The originals of Special Reports shall be submitted to the Document Control Desk with a copy sent to the Regional Administrator, NRC Region IV, within the time period specified for each report. These reports shall be submitted covering the activities identified below to the requirements of the applicable Specification.

a. Radioactive Effluents - Specifications 2.4, 2.5, 2.8 and 2. 10.

b. Radiological Environmental Monitoring - Specification 2.11.

4.4 Major Changes to Radioactive Waste Treatment Systems

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

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, support 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 differ from those previously estimated in the Environmental Report, 6.. An estimate of the exposure to plant personnel as a result of the change, and

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

I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM, VOLUMIE 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 415 MANUAL PAGE 1-54 4.5 Process Control Program Changes

a. Changes to the Process Control Program (PCP) shall be documented and records of reviews performed shall be retained as required for the duration of SAFSTOR.

b. The following information shall be available for review:

1. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s) and,

2. A determination that the change will maintain the overall conformance of the solidified waste product to existingrequirements of Federal, State, or other applicable regulations.

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

c. The change shall become effective after review and acceptance by the PSRC and the approval of the Plant Manager.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE I1-1 PART II - CALCULATIONAL METHODS AND PARAMETERS 1.0 EFFLUENT MONITOR SETPOINT CALCULATIONS 1.1 LIQUID EFFLUENT MONITORS Specification 2.1 requires that the Radioactive Liquid Effluent Monitor (RLEM) and the caisson sump monitor be set to alarm to ensure that the limits of Specification 2.3 are not exceeded (the instantaneous concentration of radioactive 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).

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

A = F,+F2)

F x 10 x (ECLc) x K x,0.85 + B (1-1) where:

A = The alarm setpoint, counts per minute, of the RLEM or the caisson sump monitor.

F1 = Flow rate past the RLEM.

F2 = Flow rate past the caisson sump monitor.

F3 = Flow rate of the effluent canal into Humboldt Bay (Fl + F2 +

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

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

Baseline calibration of the RLEM (on 02/13/07) found this factor to be within +/-15% of 2.94 x 108 cpm per micro-Ci/ml.

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

B = The monitor background reading (prior~toany discharge) in counts per minute.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-2 ECLc Composite Effluent Concentration Limit (ECL) for the mix of radionuclides (micro-Ci/ml).

10 = Factor of 10 allowed above 10 CFR 20 Appendix B values for operational flexibility.

1.1.2 The composite ECL for the mix of radionuclides is calculated as follows:

ECLC - _ i (1-2)

SECL1 Zi ECLi where:

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

Ci = Concentration of radionuclide "i" in the mixture.

fi = Fraction of radionuclide "i" inthe mixture.

1.1.3 Table 2-2 of Specification 2.1 requires that if a background reading exceeds the equivalent of 5 x 10-6 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 (Bmax, cpm) is:

Bmax = K x (5 x 10 6)cpm (1-3) 1.1.4 The most conservative background limit is calculated as if the calibration factor was 2.50 x 108 cpm per micro-Ci/ml (-15% tolerance). This background limit would be 1,250 cpm. It is plant policy to use a background limit (slightly lower) at 1,200 cpm to ensure that this limit is satisfied. Note that if the background setting exceeds 1,200 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 F1 = 0 and F 3 = F 2).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-3 1.1.6 If the Specification 2.3 alarm setting is calculated for Cs-137, -15% tolerance, no dilution and for zero background, the alarm setting would be 2,500 cpm. Because the actual mixture may have a limit that is lower than that of Cs-137, and may also provide a reduced detector response, it is plant policy to maintain the alarm setting at or below 2,400 cpm, and to run at least one circulator during discharges, to ensure that this limit is satisfied. Refer to section 1.1.7 for the administrative (lower) alarm settings.

1.1.7 For routine liquid radwaste batch discharges, it is plant policy to set the Radioactive Liquid Effluent Monitor (RLEM) alarm no higher than necessary in order to provide protection against inadvertent releases. With at least one circulator operating, the alarm should be set according to the following table, and in any case, no higher than 25,000 cpm. The table is based approximately on the sum of twice the typical background2 and ,130% of the predicted countrate for the batch3 .

Table 1-1 Liquid Effluent Monitor Alarm Setpoints Undiluted Diluted Predicted RLEM Alarm CS-137 Cs-137 Reading (Net cpm) Setting Concentration Concentration (cpm)

(micro Ci/ml) (micro-Ci/ml) 1.5E-05 5.9E-08 Up to 2,538 5000 1.8E-05 7.2E-08 2,538 up to 3,308 6000 1, 2.1E-05 8.6E-08 3,308 up to 4,077 7000 2.5E-05 9.9E-08 4,077 up to 4,846 8000 2.8E-05 1.1E-07 4,846 up to 5,615 9000 3.2E-05 1.3E-07 5,615 up to 6,385 10,000 4.9E-05 1.9E-07 6,385 up to 10,231 15,000 6.6E-05 2.6E-07 10,231 up to 14,077 20,000 8.3E-05 3.3E-07 14,077 up to 17,923 25,000 2This table is based on a nominal background of 850 cpm. As of 2/13/07, the background reading is about 680 cpm. The extra 25% provides an allowance related to the uncertainty of reading the background.

3 See section 2.4 of TBD-206. The 30% tolerance is for a combination of analytical and RLEM uncertainties and a 10% margin between the ratemeter and chart recorder.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-4 1.2 GASEOUS EFFLUENT MONITOR Deleted

NUCLEAR POWER GENERATION DEPARTMENT SECTION- ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION, REVISION 15 MANUAL. PAGE 11-5 2.0 LIQUID EFFLUENT DOSE CALCULATIONS 2.1 MONTH (31 DAY PERIOD)

The calculation methodology for a,31 day period (a "month") is the same as for the calendar year calculations provided by section 2.4, except that the resulting yalue for D (dose commitment annual rate, mremlyear) must be divided by 12 to convert it to a monthly dose commitment, mrem/month. A factor of 12 is used (instead of the exact ratio of 365.25/31), for simplicity.

2.2 CALENDAR QU'ARTER The methodology for calendar quarter calculations is, the same as for the calendar year calculations provided by section 2.4, except that the resulting value. for D (dose commitment annual rate, mrem/year) must be divided by4 to convert it to a quarterly dose commitment, mrem/quartern 2.3 CALENDAR YEAR The methodology for calendar year calculations is provided by section 2.4.

2.4 LIQUID. EFFLUENT DOSE CALCULATION METHODOLOGY The equations specified in this section for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide. 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

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:

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME. 415 REVISION TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-6 D Y C x DF x {(BFishji X UFish)+(Binvj x fi , (2-1) i=1 where:

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

Ci The average diluted effluent concentration, pico-Curie/liter, for radionuclide, i. This will be estimated by dividing the total activity of the nuclide discharged during the period, pico-Curies, by the total circulating water discharge flow during the period, liters. If Gross Alpha radioactivity is determined to be in the discharge, Pu-241 will be considered to be present at 7.5 times the amount of detected Gross Alpha radioactivity.

Note that the resulting dose commitment is the annual dose rate (mrem per year) for a time frame with this average concentration. Doses (NOT dose rates) for periods shorter than a year must be proportionately reduced.

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

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

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

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

Ulnv Usage factor of invertebrates, 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 not considered to consume either fish or other seafood, and exposure to this age group need therefore not be calculated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME .4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-7 Table 2-1 Ingestion Dose Factors for Adult Age Group (mremlpico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 1I and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI 7 1.05 x 107 7 1.05 x 10-7 H-3 No Data 1.05 x 10-7 1.05 x 10.7 1.05 x 10-Co-60 No Data 2.14 x 1076 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 5 5 Cs-137 7.97 x 10- 5 1.09 x 10-4 7.14 x 10- 5 3.70 x 10- 1.23 x 10- 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 7

Pu-241 1.57 x 10- 7.45 x 10- 7 3.32 x 10- 1.53 x 10-6 No Data 1.40 x 10-6 Gross oc 7.55 x 10- 4 7.05 x 10-4 5.41 x 10- 5 4.07 x 10- 4 No Data 7.81 x 10-5 Table 2-2 Ingestion Dose Factors for Teen Age Group' (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table EL- 12 and from NUREG-4013 Organ Nuclide. Bone -Liver Total Body Kidney Lung GI-LLI 1.06 x10- 7 1.06 x 10-7 1.06 x 10-7 H-3 No Data 1.06X 1'0-7 1.06 x10-7 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 0-4 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 i0-8 No Data 3.69 x 10-10 No Data No Data 1.13 x 10-4 1.71 x'10- 6 No Data* 1.48x 10-6 Pu-241 1.75 x 10-5 8.40 x 10- 3.69 x 10-7 Gross cc 7.98 x 10- 4 7.53 x 10-4 5.75 x 10-5 4.31 x 10-4 No Data 8.28 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-8 Table 2-3 Ingestion Dose Factors for Child Age Group (mrem/pico-Curie ingested)

Selected Nuiclide.* from Rernilatorv Gulide 1 _109 Tahie F-1 3 and from NITREfl-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 2.03 x 10- 7 2.03 x10- 7 2.03 x 10-7 2.03 x 10- 7 2.03 x 10-7 Co-60 No Data 5.29 1x 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 Pu-241 3.87 x 10-5 1.58 x 10-6 8.04 x 10- 7 2.96 x 10-6 No Data 1.44 x 10-6 Gross at 1.36 x 10- 3 1.17 x 10-3 1.02 x 10- 4 6.23 x 104 No Data 8.03 x 10-5 Table 2-4 Bioaccumulation Factors for Saltwater Environment (pCi/kg per pCi/liter)

Selected Nuclides from Regulatory Guide 1.109, Table A- I and from NUREG-4013 Element Fish Invertebrate H 9.0x 10- 1 9.3 x 10-1 Co 1.0 x 102 1.0 x 103 Sr 2.0 2.0 x 101 Cs 4.0 x 101 2.5 x 101 Y 2.5 x 10 1, 1.0 x 103 Pu 3.0 2.0x 102 Grossc** 2.5x10 1 1.0x10 3

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION' REVISION 15 MANUAL PAGE 11-9 Table 2-5 Average Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

From Regulatory Guide, 1.109, Table E-4 Other Seafood Fruitsý and Age Group Fish (Invertebrates) Vegetables Milk Meat Adult 6.9 1.0 190 110 95 Teen' 5.2 0.75 240 200 59 Child 2.2 0.33 200 170 37 Infant 0 0 0 0 0 Table 2-6 Maximum Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

From Regulatory Guide 1.109, Table E-5 Other Seafood Fruits and Age Group Fish (Invertebrates) Vegetables Milk Meat Adult 21 5.0 520 310 *110 Teen 16 3.8 630 400 65 Child 6.9 1.7 520 330 41 Infant 0 0 0 330 0

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-10 3.0 LIQUID WASTE TREATMENT 3.1 TREATMENT REQUIREMENTS 3.1.1 ODCM Specification 2.5 Specification 2.5 requires that liquid radwaste shall be treated, as required, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to liquid effluents discharged to UNRESTRICTED AREAS would exceed 0.06 mrem whole body or 0.2 mrem to any organ.

3.1.2 NPDES Waste Discharge Requirement NPDES Permit No. CA0005622, issued by the California Regional Water Quality Control Board - North Coast Region, requires that the discharge of liquid wastes "shall not cause bottom deposits in the receiving waters." The permit also identifies Discharge Serial No. 001 E (liquid low level radioactive waste) that indicates that the waste may be treated prior to discharge. The permit does not mandate treatment.

3.2 TREATMENT CAPABILITIES

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

a. Turbine Building Drain Tank The TBDT, turbine building 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE I1-11

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

d. Radwaste Building Sump The radwaste building sump tank, with a capacity of 250 gallons, is located beneath the radwaste building 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 waste receiver tanks.

3.2.2 Liquid Waste Treatment System The liquid waste treatment system processes, stores and provides for disposal of radioactively contaminated wastes and other liquid wastes that are potentially radioactively contaminated. -These wastes are first collected by the radwaste collection system and are then pumped to the radwaste building on the 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:

waste receiver tanks (3)

radwaste demineralizer

resin disposal tank

concentrated waste tanks (2)

waste hold tanks (2)

radwaste filters (2)

a. Waste Receiver and Waste Hold Tanks 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 tanks are for storing treated wastes for retreatment or disposal. The tanks are located in an external section of the radwaste building, but are within the prefabricated steel radwaste enclosure.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE - 11-12

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

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

Demineralization is generally not an appropriate method to treat high TDS liquids, but selective ion-exchange media may be used to reduce the concentration of specific radioactive ions in high DTS liquids.

c. 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.

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

e. 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.

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION' ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE ".11-13 4.0 GASEOUS EFFLUENT DOSE CALCULATIONS 4.1 DOSE RATE 4.1.1 Deleted As explained in Specification Bases 3.7, Noble Gases are not required to be monitored, and the corresponding dose rate need not be calculated.

4.1.2 Tritium and Radioactive Particulates There are no short-lived radioactive particulates in the effluent, so radioactive decay can be neglected. Meteorological parameters are assumed to be constant, and applied for the most conservative location. Therefore, the radioactive particulates dose rate calculation methodology is the same as the radioactive particulates dose calculation methodology. Refer to sections 4.3.3 through'4.3.8 for the appropriate equations.

As explained in Specification Bases 3.6, Tritium isnfiot required to be monitored, and the corresponding dose rate need not be calculated. Nevertheless, if such a calculation is required, refer to sections 4:3.9 through 4.3.13 for the appropriate equations.

4.2 Deleted 4.3 DOSE - TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM 4.3.1 Calendar Quarter The methodology for calendar quarter calculations is the same as for the calendar year calculations provided by section 4.3.3, and'discussed in section 4.3.2, with the exception that the resulting values for D. (annual dose commitment, mremlyear) must-be divided by 4 to convert them to quarterly dose commitment, mrem/quarter.

4.3.2 Calendar Year The methods for calculating the dose due to release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE., SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-14 Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.

The equations provided for determining the doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

4.3.3 Particulate Organ Dose Calculation Summation Methodology, 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) Individual inhalation 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.

The releases of radioactive 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.

Radioactive decay may result in the dose from Transuranic radionuclides becoming significant. If Gross Alpha radioactivity is determined to be released, Pu-241 will be considered to be present at 7.5 times the amount of detected Gross Alpha radioactivity. The annual dose commitment will be calculated for any organ of an individual age group as follows:

n D [Qi x (PRnh,i + RGP,i + RMeat,i + R1Miki + Rveg,i)] (4-3) 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 of mrem/year per pico-Curie/sec.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITEIDOSE CALCULATION REVISION 15 MANUAL PAGE 11-15 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 than 1 x 10-6 mrem/year (for direct radiation exposure from material deposited on the ground).

4.3.4 Particulate Inhalation Pathway Dose Calculation Methodology Rnh, i= (z/Q) x BRa x .DFi,a (4-3a) where:

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

S= 1.0 x 105 seconds/cubic meter for releases from the 50 foot stack.

Refer to Appendix B.

6.59 x 10-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B.

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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-16 4.3.5 Particulate Ground Plane Pathway Dose Calculation Methodology RGPJi = (D/Q) x SF x DF, x K x W (4-3b) where:

K unit conversion constant, 8760 hr/yr.

DFi Theground plane dose conversion factor for radionuclide, i, in mremnhr per pCi/m2 from Table 4-5. No values are provided for Transuranic radionuclides, as their dose contribution to this pathway is negligible.

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

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

3.0 x 108 inverse square meters for releases from the 50 foot stack. Refer to Appendix B.

5.39 x 10-6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B.

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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-17 4.3.6 Particulate Grass-Cow-Milk Pathway Dose Calculation Methodology RMilk, = (D/Q) x QF x U x F x (4-3c) where:

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

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

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

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

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

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

3.0 x 10-8 inverse square meters for releases from the 50 foot stack. Refer Appendix B.

3.29 x 10-6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B.

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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-18 4.3.7 Particulate Grass-Cow-Meat Pathway Dose CalculationMethodology Rmeatj (D/Q) x (QF x U. x Ffx DFia (4-3d) where:

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

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 is 0.7 kg/m2 per Regulatory Guide 1.109, Table E- 15.

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

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

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

- 3.0 x 10-8 inverse square meters for releases from the 50 foot stack. Refer to Appendix B.

3.29 x 10.6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B.

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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION 'ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL- PAGE 11-19 4.3.8 Particulate Vegetation Pathway Dose Calculation Methodology (UT x DFi, a x W)

Rveg,, = (D/Q) x U (4-3e) where:

UT The total consumption rate of fruits 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 of inverse square meters.

3.0 x 108 inverse square meters for releases from the 50 foot stack. Refer to Appendix B.

3.29 x 10-6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B.

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.

Y The agricultural productivity by unit area of pasture. This parameter is 0.7 kg/m 2 per Regulatory Guide 1.109, Table E-15.

Note: this equation probably 'overestimates exposures, since it assumes that all of the deposition on a plant remains on the plant, while the Regulatory Guide allows a factor of 0.25. Also, the quantities assumed consumed include grain (none is grown in the vicinity of the plant), as well as vegetables and fruit grown in other areas (imported to Humboldt county).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM I

VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-20 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 = QH3 X (RJnh, H3 + RGP, H3 + RMeat, H3 + RMilk, H3 + RVeg, H3) (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 0.00 13 mrem/year.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-21 4.3.10 Tritium Inhalation Pathway Dose Calculation Methodology R/hH3 Q x BRa x DFH3,a (4-4a) where:

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

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

Refer to Appendix B.

6.59 x 10-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B.

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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-22 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:

RN%, H3 x (o.75 x 0.5 X QF x Ua x F m x DFa (4-4b) 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.

DFa - 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.

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

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

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

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

Refer to Appendix B.

3.29 x 10-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B.

NUCLEAR POWER GENERATION DEPARTMENT SECTION - ODCM TITLE SAFSTOR OFFSITE DOSE 'CALCULATION VOLUME REVISION 415 MANUAL PAGE 11-23 4.3.12 Tritium Grass-Cow-Meat Pathway Dose Calculation Methodology RMeat, H3 Q) :X~ 0.75 xO0.5~ x QF x Ua x FM x DFa (4-4 c)

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

FM = The fraction of the animal's intake of H-3 which appears in a.

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

0.75 = The fraction oftotal feed that is water.

0.5 = The ratio of specific activity of the feed grass to the atmospheric

' water.

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

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

= 1.0 X 10- seconds/cubic meter for releases from the 50 foot stack.

Refer to Appendix B.

3.29 x 10-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-24 4.3.13 Tritium Vegetation Pathway Dose Calculation Methodology The concentration of tritium is based on the airborne concentration rather than the deposition:

Rveg, H3 = X (0I.75 x x UT x DFa (4-4d) where:

UT The total consumption rate of fruits and vegetables, kilogram/year. This parameter is given in Tables 4-6.and 4-7.

H = Absolute humidity of the atmosphere, 0.008 gm/m 3 per Regulatory Guide 1.109.

0.75 = The fraction of total feed that is water.

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

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

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

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

Refer to Appendix B.

= 3.29 x 10-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-25 Table 4-1 Inhalation Dose Factors for Adult Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-7 and from NUREG-4013 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 io-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 10-6 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 Pu-241 3.42 x 10-2 8.69 x 10- 3 1.29 x 10- 3 5.93 x 10- 3 1.52 x 10- 4 8.65 x 10-7 Gross cc 1.68 1.13 7.75 x 10-2 5.04 x 10-1 1.82 x 10-1 4.84 x 10-5 Table 4-2 Inhalation Dose Factors for Teen Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-8 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI 7 1.59 x 10-7 H-3 No Data 1.59 x 0-7 1.59 x 10-7 1.59x 10-7 1.59 x 10-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 1 0 -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 Pu-241 3.74 x 10-2 9.56 x 10- 3 1.40 x 10-3 6.47 x 10-3' 2.60 x 10- 4 9.17 x 10-7 Gross cc 1.77 1.20 8.05 x 10-2 5.32 x 10-1 3.12 x 10-1 5.13 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-26 Table 4-3 Inhalation Dose Factors for Child Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-9 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 Co-60 No Data 3.55 x 10-6 6.12 X 10-6 No Data 1.91 x 10- 3 2.60 x 10-5 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 5

Cs-137 2.45 x 10- 4 2.23 x 10- 4 3.47 x 10-5 7.63 x 10- 2.81 x 10- 9.78 x 10-7 Y-90 1.1,1 x 10-6 No Data 2.99 x 10-8 No Data 7.07 x 10-5 7.24 x 10-5 1.75 x 10-2 2.93 x 10- 3 1.10 x 10-2 5.06 x 10-4 8.90 x 10-7 Pu-241 7.94 x 10-2 Gross cc 2.97 1.84 1.28 x 10-1 7.63 x 10-1 6.08 x 10-1 4.98 x 10-5 Table 4-4 Inhalation Dose Factors for Infant Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E- 10 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 4.62 x 10-7 4.62 x 10-7 4.62 x 10-7 4.62 x 10-7 4.62 x 10-7 Co-60 No Data 5.73 x 10-6 8.41 x 10-6 No Data 3.22 x 10- 3 2.28 x 10-5 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 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 110-6 No Data 6.30 x 10-8 No Data 1.92 x 10- 4 7.43 x 10-5 Pu-241 8.43 x 10-2 1.85 x 10-2 3.11 x 10-3 1.15 x 10-2 7.62 x 10- 4 8.97 x 10-7 Gross cc 3.15 1.95 1.34 x 10-1 7.94 x 10-1 9.03 x 10-1 5.02 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION R REVISION 15 MANUAL PAGE 11-27 Table 4-5 External Dose Factors for Standing on Contaminated Ground (mrem/hour per pico-Curie/square meter)

Selected Nuclides from Regulatory Guide 1.109, Table E-6 Total Nuclide Skin Body H-3 0 0 Co-60 2.00 x 10-8 1.70 x 10-8 Sr-90 2.60 x 10-12 2.20 x 10-12 Cs-1371, 4.90 x 10-9 4.20 x 10-9 Y-90 2.60 x 10- 1 2 2.20 x 10- 12 Values are not provided for Transuranic radionuclides, as their dose contribution to this pathway is negligible.

Table 4-6 Average Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

From Regulatory Guide 1.1 09. Table E-4 Other Seafood Fruits and Age Group Fish (Invertebrates) Vegetables MiIk Meat Adult 6.9 1.0 190 110 90 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)

From Regulatory Guide 1.109, Table E-5 Other Seafood Fruits and Age Group Fish (Invertebrates) Vegetables Milk Meat Adult 21 5.0 520 310 110 Teen 16 3.8 630 400 65 Child 6.9 1.7 520 330 41 Infant 0 0 0 330 0

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-28 Table 4-8 Ingestion Dose Factors for Adult Age Group (mrem/pico-Curie ingested)

Selected Nuiclides from Re~iulatorv Guide 1 109 Tahle F,-i1 and from NI JREGl-401 3 Organ Nuclide Bone Liver Total Body, Kidney Lung GI-LLI 7

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 I0 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 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 Pu-241 1.57 x 10- 5 7.45 x 10-47 3.32 x 10- 7 1.53 x 10-6 No Data 1.40 x 10-6 Gross ca 7.55 x 1 0 -4 7.05 x 10-4 5.41 x 10- 5 4.07 x 10-4 No Data 7.81 x 10-5 Table 4-9 Ingestion Dose Factors for Teen Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 12 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI 7 7 7 1.06 x 10-7 1.06 x 10-7 H-3 No Data 1.06 x 10- 1.06 x 10- 1.06 x 10-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 Pu-241 1.75 x 10- 5 8.40 x 10-7 3.69.x 10-7 1.71 x 10-6 No Data 1.48 x 10-6 Gross a 7.98 x 10-4 7.53 x l0-4 5.75 x10-5 4.31 x 10-4 No Data 8.28 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL' PAGE 11-29 Table 4-10 Ingestion Dose Factors for Child Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 13 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney' Lung GI-LLI H-3 No Data 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 1.56 x 10- 5 No Data No Data 2.93 x 10-5 Co-60 No Data 5.29 x 10-6 Sr-90 1.70 x 10-2 No Data 4.31 x 10- 3 No Data No Data 2.29 x 10-4 4 5 4 5 Cs-137 3.27x 10- 4 3.13 x 10- 4.62 x 10- 1.02 x 10- 3.67x 10- 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 Pu-241 3.87 x 10-5* 1.58 x 10-6 8.04 x i0-7 2.96 x 10-6 No Data 1.44 x 10-6 Gross cc 1.36 x i103 1.17 x 10-3 1.02 x 10-4 6.23 x 104 No Data 8.03 x 10- 5 Table 4-11 Ingestion Dose Factors for Infant Agd"Group (mrem/pico-Curie ingested)'

Selected Nuclides from Regulatory Guide 1.109, Table E- 14 and from NUREG-4013 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 1.08 x 10-5 2.55 x 10- 5 No Data No Data 2.57 x 10-5 Sr-90 1.85 x10-2 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 Pu-241 4.25 x 10- 1.76 x 10- 6 8.82 x 10-7 3.17 x 10-6 No Data 1.45x 10-6 Gross c 1.46 x 10-3 1.27 x 10-3 1.09 x 10- 4 6.55 x 104 ' No Data' 8.10 x 10-5

NUCLEAR. POWER GENERATION DEPARTMENT -SECTION ODCM VOLUME4 REVISION 15 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 11-30 Table 4-12 Stable Element Transfer Data For Cow-Milk Pathway (days/liter)

,Selected Nuclides from Re*,ulatorv Guide 1.109. Table E- 1 and from NI JREG-40 13 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 Pu 5.0 x 10-6 Gross a 5.0 x 10-6 Table 4-13 Stable Elemient Transfer Data For Cow-Meat Pathway (days/kilo-gram)

Selected Nuclides from Regulatory Guide 1.109, Table E- 1 and from NUREG-4013 Element Ff H 1.2 x 10-2 Co 1.3x 10-2 Sr 6.0x 10-4 Cs 4.0 x 10-3 Y 4.6x 10-3 Pu 2.0 x 10-4 Gross a 2.0 x 10-4

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL, PAGE 11-31' 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE 5.1 WHOLE BODY.DOSE Specification 2.10 limits the whole body dose equivalent from the Uranium fuel to no more than 25 mrem/year. The whole body dose is determined.by summing the calculated doses from the following:

a. Deleted

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 individual at the site boundary. The only portion of the site boundary where there issignificant direct 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 hour7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months s/year, given by Table E-5 ofRegulatory 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:

a. Deleted

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-32 5.3 DOSE TO OTHER ORGANS Specification 2.10 limits the dose to any organ (thyroid excepted) to less than or equal to 25 mrem!year. The dose to any individual other than skin organ is determined by summing the calculated doses from the following:

a. Deleted

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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION - ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-33 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING SOLIDIFICATION 6.1 SCOPE 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 land burial, or else be packaged in a high integrity container in accordance with Section 7.0.

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

6.2.1 Contract vendor solidification service may be utilized. The contract vendor solidification service may consist of solidification by the contractor or supply of materials, procedures and process control program (PCP) for HBPP solidification.

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 curing times.

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

processed to ensure solidification. The procedure should also include the actions 'to be taken if the test specimen fails to solidify.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-34 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 stabilized by packaging in dewatered form in a high-integrity container which meets burial ground and regulatory requirements, or else be solidified in accordance with Section 6.0.

7.2 PROGRAM ELEMENTS For land burial 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 approved or have a current Certificate of Compliance prior to acceptance for use by HBPP. This shall include provision by the vendor to HBPP of documentation reflecting this authorization.

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM

' VOLUME,4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE, 11-35 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 shipped for land burial which contain a total specific activity less than the burial ground criteria for solidification, and which does not exceed the concentration limits for Class A waste as defined in 10 CFR 61.

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION REVISION 15 MANUAL PAGE 11-36 9.0 PROGRAM CHANGES 9.1 PURPOSE OF THE OFFSITE DOSE CALCULATION MANUAL 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 manual is 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 Effluent 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.

9.3 HBPP does not intend to modify or reduce the environmental monitoring requirements as specified in the ODCM during the periods of SAFSTOR and decommissioning activities.

This applies to those environmental samples and analysis identified in Table 2-7 as either quality or non-quality samples. [CTS-291]

10.0 COMMITMENTS The following commitment is implemented by this procedure. The section number that implements to commitment is noted parenthetically.

CTS-291 (Section II, 9.3)

CTS-352 (Section I, Table 2-4) 11.0 PROCEDURE OWNER Radiation Protection Manager

ODCM APPENDIX A Revision 15 Page A-1 APPENDIX A SAFSTOR BASELINE CONDITIONS

ODCM APPENDIX A Revision 15 Page A-2 1.0 LIQUID AND GASEOUS EFFLUENTS 1.1 LIQUID EFFLUENTS Baseline levels of radioactive materials contained in liquid effluents during the SAFSTOR period were established in the Environmental Report submitted as Attachment 6 to the SAFSTOR license amendment request. These values are presented for cumulative annual release and average monthly discharge in Table A-1.

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

Table A-1 Baseline Liquid Effluent 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 Strontium-90 3.28E-4 2.73E-5 Table A-2 Baseline Gaseous Effluent Activity Type of Activity Annual Release Monthly Average Release (Curies) (Curies)

Tritium <4.OE-2 <3.3E-3 Particulate Gamma Emitters (total) 3.16E-4 2.63E-5 Strontium-90 3.38E-6 2.82E-7

ODCM APPENDIX B Revision 15 Page B-I 2

APPENDIX B BASES FOR ATMOSPHERIC DISPERSION AND DEPOSITION VALUES f\

ODCM APPENDIX B Revision 15 Page B-2 1.0 BASIS FOR DISPERSION/DEPOSITION VALUES - 50' STACK 1.1 The instantaneous atmospheric dispersion factor (X/Q) is taken from meteorological parameter calculations performed to evaluate reducing the height of the Unit 3 stack. The calculation report is number N238C, Revision 0, titled "Determine Effect of Humboldt Bay Power Plant Unit 3 Stack Reconfiguration on Downwind Effluent Concentrations".

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) of the calculation (N238C) provides "1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months " values for the instantaneous X/Q for the 50' stack for various stack flow rates, based on an EPA model named "ISCST". The instantaneous X/Q value used in the ODCM (6.52 x 10-4) is based on a stack flow of 25,000 cfm.

1.2 The annual average atmospheric dispersion factor (X/Q) is taken from meteorological parameter calculations performed to evaluate reducing the height of the Unit 3 stack. The calculation report is number N238C, Revision 0, titled "Determine Effect of Humboldt Bay Power Plant Unit 3 Stack Reconfiguration on Downwind Effluent Concentrations".

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) of the calculation (N238C) provides annual maximum values for X/Q for the 50' stack for various stack flow rates, based on an NRC model named "XOQDOQ". The annual average X/Q value used in the ODCM (1.00 x 105) is based on a stack flow of 25,000 cfm.

1.3 The annual average atmospheric deposition factor (D/Q) is taken from meteorological parameter calculations performed to evaluate reducing the height of the Unit 3 stack. The calculation report is number N238C, Revision 0; titled "Determine Effect of Humboldt Bay Power Plant Unit 3 Stack Reconfiguration on Downwind Effluent Concentrations".

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) of the calculation (N238C) provides annual maximum values for D/Q for the 50' stack for various stack flow rates, based on an NRC model named "XOQDOQ". The annual average D/Q value used in the ODCM (3.00 x 10-8) is based on a stack flow of 25,000 cfm.

2.0 BASIS FOR DISPERSION/DEPOSITION VALUES - INCIDENTAL RELEASE PATHS 2.1 The atmospheric dispersion factor (X/Q) for incidental releases is 6.59 x 10-3 seconds/cubic meter, calculated as described below 2.1.1 This factor is based on the atmospheric models of Regulatory Guide 1.145, Atmospheric DispersionModelsfor PotentialAccident Consequence Assessments at Nuclear Power Plants. These models are intended to estimate meteorological dispersion for "real time" conditions (i.e., hourly), rather than "annual average" conditions. The applicable guidance is section 1.3.1 (Releases Through Vents or Other Building Penetrations), as it applies to all releases from points lower than 2.5 times the height of adjacent structures. This calculation generally follows the guidance for the use of equations 1, 2 and 3 of Regulatory Guide 1.145.

ODCM APPENDIX B Revision 15 Page B-3 2.1.2 The assumed distance from the emission point to the potential receptor for this calculation is 150 meters. This is the approximate distance to publicly accessible areas from the structure with the most significant potential for airborne radioactivity (i.e. from the center of the Refueling Building to the trail at the edge of the bluff).

2.1.3 The meteorological conditions assumed for this calculation are for stable "fumigation" conditions (Pasquill stability class G), with a wind speed of 1 meters/second.

2.1.4 The applicable equations from Reg. Guide 1.145 are as follows:

X/Q (1 UIO(rrCT~a,

1. +A/2)

X/Q =(2)

X/Q- UlortXC Zycr (3) where:

U1 0 = wind speed at 10 meters above grade, equal to 1 meter/second.

Cry lateral plume spread, equal to 4.33 meters for Pasquill Class G at a distance of 150 meters.

oYz = vertical plume spread, equal to 1.86 meters for Pasquill Class G at a distance of 150 meters.

A vertical cross-sectional area of structures, equal to 375 meters2 , based on the Refueling Building dimensions (about 36 feet high, about 112 feet long).

XY = lateral plume spread (including meander and building wake), meters, equal to 6ay (for distances less than 800 meters, wind speeds below 2 meters/second, and stability class G).

2.1.5 With these values, the results for equations 1, 2, and 3 are as follows:

X/Q = 4.70 x 10-3 seconds/meter 3 (1)

ODCM APPENDIX B Revision 15 Page B-4 X/Q = 1.32 x 102 seconds/meter 3 (2)

X/Q = 6.59 x 10-3 seconds/meter 3 (3)

Per the Reg. Guide, the higher value of equations 1 and 2 is to be compared with the value for equation 3, and the lower value of that comparison should be used.

with this logic, the resulting value for X/Q is 6.59 x 10-3 seconds/meter3.

2.2 The atmospheric deposition factor (D/Q) for incidental releases is 5.39 x 10-6 meter-2 for the Particulate Ground Plane Pathway, and is 3.29 x 10-6 meter 2 for all other deposition related pathways. The factors are calculated as described below 2.2.1 These factors are based on the atmospheric models of Regulatory Guide 1.111, Methodsfor EstimatingAtmospheric Transportand Dispersionof Gaseous Effluents in Routine Releasesfrom Light-water-cooledReactors. The applicable guidance is section C.3.b (Dry Deposition), and Figure 6 (Relative Deposition for Ground-level Releases). To determine the atmospheric deposition across a downwind sector, the value from Figure 6 is to be multiplied by the fraction of the release transported into the sector, and divided by the sector cross-wind arc length at the distance being considered. For this calculation, the deposited contamination will be assumed to be evenly distributed across the width of the plume, rather than across an arbitrary angular sector.

2.2.2 Two factors are necessary because the nearest location (along the bay) is not a credible location for farming. For the purposes of estimating offsite doses from incidental releases, the nearest "farm" will be assumed to be beyond the railroad tracks, Southeast of the plant.

2.2.3 For the Particulate Ground Plane Pathway, the assumed distance from the emission point to the potential receptor for this calculation is 150 meters. This is the approximate distance to publicly accessible areas from the structure with the most significant potential for airborne radioactivity (i.e. from the center of the Refueling Building to the trail at the edge of the bluff). At this distance, Figure 6 provides a Relative Deposition Rate value of 1.4 x 10- meter-1 . The plume width assumed for this calculation is the same as was used in equation 3 of section 2.1.4 (above), so that the plume width is approximately 6oy. For Yy equal to 4.33 meters (Pasquill Class G at a distance of 150 meters), D/Q is (1.4 x 10-4 meterl)/

(6 x 4.33 meter) = 5.39 x 10-6 meter-2.

2.2.4 For the pathways involving farming or ranching, the assumed distance from the emission point to the potential receptor for this calculation is 220 meters. This is the approximate distance to publicly accessible "grazing" areas from the structure with the most significant potential for airborne radioactivity (i.e. from the center of the Refueling Building to the other side of the railroad). At this distance,

ODCM APPENDIX B Revision 15 Page B-5 Figure 6 provides a Relative Deposition Rate value of 1.2 x 10 -4meter-'. The plume width assumed -for this calculation is the same as was used in equation 3 of section 2.1.4 (above), with the plume width of approximately 6ory., but at a greater distance. For ay equal to 6.07 meters (Pasquill Class G at a distance of 220 meters), D/Q is (1.2 x 10-4 meter')/ (6 x 6.07 meter) = 3.29 x 10-6 meter-2 .

ODCM APPENDIX C Revision 15 Page C-1 APPENDIX C Deleted

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