Text

ODCM, Revision 22, Volume 4, Safstor/Decommissioning Offsite Dose Calculation Manual.
	ML13088A172
Person / Time
Site:	Humboldt Bay
Issue date:	07/20/2012
From:	; Pacific Gas & Electric Co
To:	; NRC/FSME
References
	HBL-13-004 ODCM, Rev 22
	Download: ML13088A172 (112)
	v • d • e

Nuclear Power Generation SECTION ODCM VOLUME 4 Humboldt Bay DATE I22-REVISION EFFEC -. ;

Power Plant PAGE TITLE APPROVED BY

& SAFSTOR/DECOMMISSIONING OFFSITE DOSE7

( o, D5tECTORJPLANT MANAGER'//[lATE CALCULATION MANUAL HB NUCLEAR (Procedure Classification - Quality Related)

INTRODUCTION The SAFSTOR/DECOMMISSIONING 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 Radiological Effluent 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 89-0 1, Implementation of Programmatic Controlsfor RadiologicalEffluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of ProceduralDetails ofRETS to the Offsite Dose CalculationManual or to the ProcessControl 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 I. 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 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 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-1 166, HBPPFinal EnvironmentalStatement.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION 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 aninterlaboratory 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 decommissioning 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION 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 1-15 2,4 Liquid Effluent - Dose 1-19 2.5 Liquid Waste Treatment 1-20 2.6 Gaseous Effluents - Dose Rate 1-21 2.7 Deleted 1-24 2.8 Gaseous Effluents: Dose - Tritium and Radionuclides in Particulate Form 1-25 2.9 Solid Radioactive Waste 1-26 2.10 Total Dose 1-27 2.11 REMP Monitoring Program 1-28 2.12 REMP Interlaboratory Comparison Program 1-41 2.13 Radioactive Waste Inventory 1-42 3.0 SPECIFICATION BASES 1-43 3.1 Radioactive Liquid Effluent Monitoring Instrumentation Basis 1-43 3.2 Radioactive Gaseous Effluent Monitoring Instrumentation Basis 1-43 3.3 Liquid Effluent Concentration Basis 1-43 3.4 Liquid Effluent Dose Basis 1-44 3.5 Liquid Waste Treatment Basis 1-44 3.6 Gaseous Effluents Dose Rate Basis 1-44 3.7 Deleted 1-45 3.8 Gaseous Effluents: Tritium and Radionuclides in Particulate Form Dose Basis 1-45 3.9 Solid Radioactive Waste Basis 1-46 3.10 Total Dose Basis 1-46 3.11 REMP Monitoring Program Basis 1-47 3.12 REMP Interlaboratory Comparison Program Basis 1-48 3.13 Radioactive Waste Inventory Basis 1-48

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE iv PART I - SPECIFICATIONS - (Continued)

Section Title Page 4.0 ADMINISTRATIVE CONTROLS 1-48 4.1 Annual Radiological Environmental Monitoring Report 1-48 4.2 Annual Radioactive Effluent Release Report 1-53 4.3 Special Reports 1-54 4.4 Major Changes to Radioactive Waste Treatment Systems 1-54 4.5 Process Control Program Changes 1-55 PART I1 - CALCULATIONAL METHODS AND PARAMETERS Section Title Page 1.0 EFFLUENT MONITOR SETPOINT CALCULATIONS 11-6 1.1 Liquid Effluent Monitors II-1 1.2 Gaseous Effluent Monitor II-6

2. 0 LIQUID EFFLUENT DOSE CALCULATIONS II-12 2.1 Month (31 Day Period) 11-12 2.2 Calendar Quarter 11-12 2.3 Calendar Year 11-12 2.4 Liquid Effluent Dose Calculation Methodology 11-12

3. 0 LIQUID WASTE TREATMENT 11-18 3.1 Treatment Requirements 11-18 3.2 Treatment Capabilities 11-18 4.'0 GASEOUS EFFLUENT DOSE CALCULATIONS 11-20 4.1 Dose Rate 11-20 4.2 Deleted 11-20 4.3 Dose - Tritium and Radionuclides in Particulate Form 11-20

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/IECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE v PART II - CALCULATIONAL METHODS AND PARAMETERS - (Continued)

Section Title Page 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE 11-38 5.1 Whole Body Dose 11-38 5.2 Skin Dose 11-38 5.3 Dose to Other Organs 11-39 5.4 Dose to the Thyroid 11-39 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE 11-40 REQUIRING SOLIDIFICATION 7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE 11-41 PACKAGED IN HIGH INTEGRITY CONTAINERS 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED 11-42 RESINS AND OTHER WET WASTES 9.0 PROGRAM CHANGES 11-43 10.0 COMMITMENTS 11-43 11.0 PROCEDURE OWNER 11-43

12.0 REFERENCES

11-43 App. A SAFSTOR BASELINE CONDITIONS A-I App. B BASES FOR ATMOSPHERIC DISPERSION AND DEPOSITION VALUES B-1 App. C Deleted C-1

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE vi LIST OF TABLES - PART I Table Title Pae 1-1 Frequency Notation 1-6 2-1 Radioactive Liquid Effluent Monitoring Instrumentation 1-9 2-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance 1-10 Requirements 2-3 Radioactive Gaseous Effluent Monitoring Instrumentation 1-13 2-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance 1-14 Requirements 2-5 Radioactive Liquid Waste Sampling and Analysis Program 1-16 2-6 Radioactive Gaseous Waste Sampling and Analysis Program 1-22 2-7 HBPP Radiological Environmental Monitoring Program 1-30 2-8 Reporting Levels for Radioactivity Concentrations In Environmental Samples 1-32 2-9 Detection Capabilities for Environmental Sample Analysis Lower Limits Of 1-33 Detection (LLD) 2-10 Distances and Directions To Environmental Monitoring Stations 1-35 4-1 Radiological Environmental Monitoring Report Annual Summary - Example 1-50

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 422 SAFSTORIDECOMMISSIONING OFFSITE REVISION TITLE DOSE CALCULATION MANUAL PAGE vii LIST OF TABLES - PART II "Table Title 2-1 Ingestion Dose Factors for Adult Age Group 11-15 2-2 Ingestion Dose Factors for Teen Age Group 11-15 2-3 Ingestion Dose Factors for Child Age Group 11-16 2-4 Bioaccumulation Factors for Saltwater Environment 11-16 2-5 Average Individual Foods Consumption for Various Age Groups 11-17 2-6 Maximum Individual Foods Consumption for Various Age Groups 11-17 4-1 Inhalation Dose Factors for Adult Age Group 11-32 4-2 Inhalation Dose Factors for Teen Age Group 11-32 4-3 Inhalation Dose Factors for Child Age Group 11-33 4-4 Inhalation Dose Factors for Infant Age Group 11-33 4-5 External Dose Factors for Standing on Contaminated Ground 11-34 4-6 Average Individual Foods Consumption for Various Age Groups 11-34 4-7 Maximum Individual Foods Consumption for Various Age Groups 11-34 4-8 Ingestion Dose Factors for Adult Age Group 11-35 4-9 Ingestion Dose Factors for Teen Age Group 11-35 4-10 Ingestion Dose Factors for Child Age Group 11-36 4-11 Ingestion Dose Factors for Infant Age Group 11-36 4-12 Stable Element Transfer Data For Cow-Milk Path 11-37 4-13 Stable Element Transfer Data For Cow-Meat Path 11-37

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE viii LIST OF FIGURES - PART I Limir Title Page 1-1 Site Boundary 1-7 2-1 HBPP Onsite TLD Locations 1-36 2-2 HBPP Onsite Monitoring Well Locations 1-37 2-3 HBPP Offsite Sampling Locations - Humboldt Hill 1-38 2-4 HBPP Offsite Sampling Locations - Eureka 1-39 2-5 HBPP Offsite Sampling Locations - Fortuna 1-40

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 422 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE I-1 PART I - SPECIFICATIONS 1.0 DEFINITIONS 1.1 ACTION ACTION shall be that part of a control that prescribes remedial measures required under designated conditions.

1.2 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 encompass 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 422 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL 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.

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 completed 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 SAFSTORJDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-3 1.12OFFSITE 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).

1.14PROCESS 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, 6 1, 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 T'ITLE SAFSTORJDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-5 1.22 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed t'0 reduce radioactive material in particulate form "ineffluents by passing "

ventilation or vent exhaust gases through charcoal absorbers 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 422 SAFSTOR/DECOMMISSIONING OFFSITE REVISION TITLE DOSE CALCULATION MANUAL PAGE 1-6 Table 1-1 FREQUENCY NOTATION Notation' Fre~iuencv '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 longer is 25% with a maximum tolerance of 325% for three consecutive periods.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-7 Figure 1-1 SITE BOUNDARY

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION 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 a 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 422 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-9 Table 2-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument OPERABLE AC TION

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Process Water Monitor I 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 for principal gamma emitters in accordance with Table 2-5 in Specification 2.3. 1, and

b. An INDEPENDENT VERIFICATION of instantaneous concentration calculations is performed, and

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

Otherwise, suspend releases of radioactive materials via this pathway.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 422 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION 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

Daily during normal work days only 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 422 SAFSTOR/DECOMMISSIONING OFFSITE REVISION TITLE DOSE CALCULATION MANUAL PAGE I-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: Whenever the ventilation system is in operation.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required above, suspend work that could result in 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.

The Continuous Alpha Monitor may be secured to replace the filter paper cassette without declaring the Stack Particulate Airborne Monitoring System inoperable.

The Particulate Sampler may be secured to replace the filter without declaring the Stack Particulate Airborne Monitoring System inoperable.

Performing the Quarterly SPAM Calibration STP requires declaring the Stack Particulate Airborne Monitoring System INOPERABLE.

The NRC classifies effluent monitoring as either liquid or gaseous. With the removal of the spent fuel to the ISFSI, the remaining significant radioactive source at Humboldt Bay Power Plant is particulate carrying alpha contamination, potentially released to the environment through the "gaseous" air flow of the ventilation system, in unlikely events including the failure of the HEPA filters. There is no significant gaseous activity, per se. Thus PG&E's "gaseous effluent monitoring instrumentation" is actually particulate alpha monitoring, and is commonly referred to as the Stack Particulate Alpha Monitor or Stack Particulate Alpha Monitoring System or SPAMS in most PG&E documentation.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-12 SURVEILLANCE REQUIREMENTS 2.2.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 2-4.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-13 Table 2-3 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument OPERABLE ACTION

1. Stack Particulate Airborne Monitoring System (SPAMS)

a. Continuous Alpha Monitor 2 1 A,C

b. Particulate Sampler I B

c. Effluent System Flow Rate Monitor I C,D

d. Continuous Alpha Monitor Flow Rate Monitor I A

e. Particulate Sampler Flow Rate Monitor* I A,

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

Table Notation ACTION A The continuous alpha monitor may be taken out of service for calibration or maintenance, but shall be returned to. service as soon as practicable within the 30 day period allowed by ACTION 2.2.1 .b. If the continuous alpha monitor becomes inoperable, place any current work in a safe condition, then suspend work that could result in the release of radioactive gaseous effluents monitored by the affected channel.

ACTION B 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. If the particulate sampler becomes inoperable, secure ventilation, place any current work in a safe condition, then suspend work that could result in the release of radioactive gaseous effluents monitored by the affected channel.

ACTION C 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.

2 The Humboldt Bay Power Plant's SPAMS consists of a shrouded sample nozzle assembly and sample line designed to meet ANSI N13.1-1999, requirements, feeding an MGP ABPM201S skid, where the sample stream is split, and feeds both a continuous near real time particulate alpha monitor, referred to as the Continuous Alpha Monitor, and a Particulate Sampler, referred to as such.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 422 SAFSTOR/DECOMMISSIONING OFFSITE REVISION TITLE DOSE CALCULATION MANUAL PAGE 1-14 ACTION D 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.

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. Continuous Alpha Monitor D* Q Q Q.

b. Particulate Sampler W N.A. N.A. N.A.

c. Effluent System Flow D* N.A. A N.A.

Rate Monitor

d. Continuous Alpha D* N.A. Q Q Monitor Flow Rate Monitor

e. Particulate Sampler Flow D* N.A. Q Q Rate Monitor

Daily during normal work days only 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.

c. Loss of sample flowrate

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-15 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 SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-16 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)

(jaCi/mlf A. Batch Waste Release Tanksc P P Principal Gamma 5 x 10"

1. Treated Waste Hold Tank(2) Each Batch Each Batch Emitterse

2. Waste Receiver Tanks(3) P M H-3 I x 10.5 Each Batch Compositeb Gross Alpha 1 x 10.

P Q Sr-90 5 x 10.8 Each Batch Compositeb P Q Ni-63 1 x 10-6 Each Batch Compositeb B. Plant Continuous Releasesd D W Principal Gamma 5 x 10-'

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 10-r Grab Sample Composite _

D Q Ni-63 1 x 10.6 Grab Sample Composite b Table Notation The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD = 4.66 sb (E)(V) (2.22 X106) (e-IA) y

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-17 Table 2-5 (Continued)

Table Notation (Continued)

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),

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 priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

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. Alternately, one or more individual samples may be separately analyzed and "composited" mathematically.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM REVISION 22 TITLE SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-18 Table 2-5 (Continued)

Table Notation (Continued)

C 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.

d 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.

f Analysis specific to Sr-90 may be replaced by analysis for total radioactive Strontium.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-19 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 or equal to 3 mrem to the total body and to less than or equal to 10 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

With the calculated dose from the release of radioactive materials in liquid 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 422 SAFSTOR/DECOMMISSIONING OFFSITE REVISION TITLE DOSE CALCULATION MANUAL PAGE 1-20 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 422 SAFSTOR/DECOMMISSIONING OFFSITE REVISION TITLE DOSE CALCULATION MANUAL PAGE 1-21 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 I0CFR20, 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 422 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-22 Table 2-6 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM 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 = 4.66sb (E)(v) (2.22 x 10o) K- t )Y E

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-23 Table 2-6 (Continued)

Table Notation (Continued)

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),

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 p!rori (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.

Samples 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/IDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-24 Table 2-6 (Continued)

Table Notation (Continued)

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.

f The LLD equation noted above does not apply to alpha spectroscopy instruments such as are used in the stack alpha continuous monitor.

g Analysis specific to Sr-90 may be replaced by analysis for total radioactive Strontium.

2.7 Deleted

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-25 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-26 As explained in Specification Bases section 3.8, neither routine surveillance nor dose calculations are required for Tritium in gaseous effluents.

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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-27 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 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-28 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 (I) concentration (2) r . 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 422 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-29 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 22 PAGE 1-30 Table 2-7 HBPP RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PROGRAM DESCRIPTION PROGRAM BASIS Exposure Pathway Number of Samples Sampling and Collection Frequency Type of Analysis ODCM State of and/or Sample and Locations(s) Specs California (QR) (NQR)

AIRBORNE 5 onsite locations, I offsite Continuous sampler operation with Gross alpha and gross beta radioactivity X location sample collection at least once per 7 following filter change dayst1 ) Gamma isotopic(c) analysis on quarterly composite (by station) t1 DIRECT RADIATION(b) 16 onsite stations, at or within the TLDs exchanged quarterly( Gamma exposure(3) X SITE BOUNDARY fence line, with TLDs I offsite control station with TLD TLDs exchanged quarterly(i) Gamma exposure(3) X 4 offsite stations with TLDs TLDs exchanged quarterly(O) Gamma exposuret3 ) X WATERBORNE Surface Water Discharge canal effluent Continuous sampler operation with Gamma isotopic(c), Strontium-90 and Tritium X X sample collection weeklyti) analysis of weekly sample Dip samples if sampler inoperablet') Sample submitted to the State Department of Health Services monthly(l)

Groundwater 12 groundwater spent fuel pool Quarterly Tritium, Strontium-90, Americium-241 and X monitoring wells gamma isotopic(c) analysis Alpha and Beta Analysis X

NUCLEAR POWER GENERATION DEPARTMENT SECTION VOLUME ODCM 4

TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL REVISION 22 PAGE 1-31 Table 2-7 (Continued)

PROGRAM DESCRIPTION PROGRAM BASIS Exposure Pathway Number of Samples Sampling and Collection Frequency Type of Analysis ODCM State of and/or Sample and Locations(A) Specs California (QR) (NQR)

INGESTION Milk Pedrotti Dairy Annuallyti) Strontium-90 and Gamma isotopic(c) X analysis"2 )

Holgerson Dairy Annually(l) Strontium-90 and Gamma isotopic(') X analysis(2 )

TERRESTRIAL None None N/A N/A Table Notations t

QR - Quality Related ()Performed by HBPP )Performed by DCPP NQR - Non-Quality Related (2)Performed by Offsite Laboratory 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 quality-related portion of the REMP.

(b) At least 4 additional TLDs are deployed, one in each cardinal direction along the ISFSI fence line, when fuel is in storage

(') 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 RVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-32 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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-33 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) (pCiFL) (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 E x V x 2.22 x Y x exp(-Xt)

Where:

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

Su = 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-34 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 ariori (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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-35 Table 2-10 DISTANCES AND DIRECTIONS TO ENVIRONMENTAL MONITORING STATIONS Radial;Direction Radial Distafnce Station By from Plant No. Code Station Name Sector Degrees (Miles)

I A King Salmon Picnic Area W 270 0.3 2 A 180 Dinsmore Drive, Fortuna SSE 158 9.4 3 "1 Humboldt Hill Road at Bret Harte Lane SSE 158 0.9 14 A South Bay School Parking Lot S 180 0.4 16 0 Elk River Road/PG&E Gas Reg/Pedrotti Dairy ENE 72 1.4 17 A Control Set at Humboldt Substation, Eureka NEE 61 5.8 25 A Irving Drive, Humboldt Hill SSE 175 1.3 48 0 Holgerson Dairy S 180 5.1

A ISFSI Fence line *

At least 4 additional TLDs are used, one in each direction, at the ISFSI Fence line, when fuel is instorage Table Notations Code: A Dosimetry Station 0 Air Particulate Station 0 Biological Station

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-36 Figure 2-1 HBPP ONSITE TLD LOCATIONS Location GPS Coordinates (NAD83/NAVD88 CA. Zone 1) HBPP (caled north)

Number Easting Northing el. East North TI 5949161.06 2160822.11 10.78 4873.87 9168.63 T2 5948804.52 2160710.72 11.56 4513.84 9268.18 T3 5948609.45 2161061.84 41.77 4540.12 9M68.91 T4 5948778.72 2161269.91 43.66 4795.13 9752.07 T5 5949002.39 2161368.44 38.19 5036.50 9713.72 T6 5949159.22 2161437.55 36.30 5205.77 9686.84 T7 5949280.02 2161494.61 32.04 5338.22 9689.36 T8 5949511.99 2161608.36 12.96 5594.82 9639.33 T9 5949651.46 2161588.47 11.79 5701.27 9547.04 T1O 5949912.89 2161633.96 11.17 5945.65 9443.64 T1l 5950011.77 2161297.55 14.18 5846.48 9107.30 T12 5950019.25 2160858.44 11.25 5614.86 8734.19 T13 5949841.53 2160718.03 9.79 5389.40 8712.46 T14 5949583.98 2160684.24 10.46 5154.63 8823.60 T15 5949448.88 2160600.96 10.34 4995.96 8826.81 T16 5949352.82 2160667.18 10.80 4951.10 8934.52 T18 5948867.24 2161239.36 43.47 4852.98 0678.44 T19 5948796.71 2161242.74 42.84 4795.52 9719.50 T20 5948747.14 2161191.68 44.14 4726.20 9703.44 T21 5948834.52 2161182.89 45.71 4799.39 9644.52

1 1 iwi it 41ý1 ý NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 422 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-37 Figure 2-2 HBPP ONSITE MONITORING WELL LOCATIONS Monitoring GPS Coordinates (NAD83/NAVD88 CA. Zone 1) HBPP (called north)

Well Easting Northing el. East North MW-1 5949428.45 2161020.10 11.35 5205.88 9190.17 MW-2 5949393.32 2161169.01 37.36 5257.03 9334.29 MW-4 5949470.92 2161159.02 11.41 5316.85 9283.91 MW-6 5949423.12 2161223.94 10.99 5311.84 9364.38 MW-11 5949588.32 2161053.64 12.04 5358.42 9131.73 RCW-CS-1 5949309.92 2161136.20 10.82 5169.16 9351.96 RCW-CS-2 5949446.86 2161208.52 10.87 5323.44 9338.56 RCW-CS-3 5949504.15 2161122.50 11.22 5324.99 9235.21 RCW-CS-4 5949448.47 2160980.19 11.17 5201.08 9145.77 RCW-CS-5 5949545.79 2160969.31 11.19 5276.99 9083.90 RCW-SFP-1 5949395.97 2161268.83 26.41 5313.34 9416.78 I RCW-SPF-2 1 5949204.48 2161235.37 32.63 5134.27 9492.39

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 422 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-38 Figure 2-3 HBPP OFFSITE SAMPLING LOCATIONS - HUMBOLDT HILL GPS Coordinates (NAD83/NAVD88 CA. Zone 1) Decimal Degrees Station Easting Northing el. Latitude Longitude 1 5948026.52 2161183.79 11.38 40.74156 -124.21903 3 5951260.28 2155706.11 234.94 40.72676 -124.20274 14 5949876.83 2158864.39 18.65 40.73533 -124.20802 25 5950247.30 2154214.18 229.22 40.72260 -124.20626

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 EVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-39 Figure 2-4 HBPP OFFSITE SAMPLING LOCATIONS - EUREKA J.""v.*4M I I 14'I O1 cod'kal.

I I L

I GPS Coordinates (NAD83.NAVD88 CA. Zone 1) Decimal Degrees Station I Eastim I Norig el. Latitude Longitude 17 5976549.55 2175490.19 164.85 40.78276 -124.11324

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-40 Figure 2-5 HBPP OFFSITE SAMPLING LOCATIONS - FORTUNA GPS Coordinates (NAD83INAVD88 CA. Zone 1) Decimal Degrees Station Easting I Northing I el. Latitude Longitude 2 5962583.86 2105797.82 35.53 40.59057 -124.15746

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-41 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.

IL~uj1UIhU~~I ~ -

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-42 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: 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.

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 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-43 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 A to 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 1, 10 CFR 50, to a MEMBER OF THE PUBIC and (2) the limits of 10 CFR 20.1302 to the population. This specification does not affect the requirement to comply with the annual limitations of 10 CFR 20.1301(a).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM TITLE SAFSTOR/DECOMMISSIONING OFFSITE IREVISION VOLUME 4 22 DOSE CALCULATION MANUAL PAGE 1-44 3.4 Liquid Effluent Dose Basis This specification is provided to implement the requirements of Sections II.A. III-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 ACTION statement provides the required operating flexibility and at that same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable" (ALARA). The dose calculations in the OFFSITE DOSE CALCULATION 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 decommissioning 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 50.36a, 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 1, 10 CFR Part 50, for liquid effluents (3 mrem/yr; 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 CER 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 IL.A and II.C design objectives of Appendix I to 10 CFR 50.

- ]LIJ - 4JE1k NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 422 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-45 For a MEMBER OF THE PUBLIC who may at times be within the SITE BOUNDARY, the period of occupancy (which is bounded by the maximum occupational period while working in Units I 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 1x10. 4 micro-Curies/ml, and air at 100 OF, 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 5x]0.9 micro-Curies/cc. This is well below the IOCFR20 Effluent Concentration Limit of 1xl0-7 micro-Curies/cc, which corresponds to a dose of 50 mrem/year, so it is not necessary to monitor for Tritium in the plant stack effluent stream.

3.7 Deleted 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 22 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-46 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 release of Tritium and radioactive materials ini particulate form (with half-lives greater than 8 days) in gaseous effluents during decommissioning 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 lx104 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-rem/year, so it is not necessary to calculate doses for Tritium in the plant stack effluent stream.

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.iO 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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-47 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 decommissioning and to document any increased nuclide concentrations and/or radiation levels resulting from accidents during decommissioning.

The non quality-related portion of the HBPP REMP fulfills commitments for environmental monitoring made to the state of California and conducts additional environmental monitoring which PG&E/HBPP has elected to continue from the REMP which was being implemented prior to approval of the SAFSTOR Decommissioning Plan.

Normally, non quality-related environmental monitoring (including sample collection and analysis) is conducted in accordance with the programmatic controls established for the quality-related environmental monitoring; however, this monitoring is not subject to the program requirements for radiological environmental monitoring contained in the NRC Radiological Assessment Branch's Branch Technical Position which was issued as Generic Letter 79-65 nor is it subject to the HBPP Decommissioning Quality Assurance Program requirements including adherence to Regulatory Guide 4.15, Quality Assurancefor RadiologicalMonitoringPrograms (Normal Operations)--EffluentStreams 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.

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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 422 RVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-48 3.12 REMP Interlaboratory Comparison Program Basis The requirement for participation in an Interlaboratory Comparison Program igsprovided 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.

4.0 ADMINISTRATIVE CONTROLS 4.1 Annual Radiological Environmental Monitoring Report A report on the Decommissioning Radiological Environmental Monitoring Program shall be prepared annually in accordance with the NRC Branch Technical Position and submitted to the NRC by May I 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 1,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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-49

d. A summary description of the Decommissioning 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.

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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-50 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 I - 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 Measurementj Analyses of Detectiona (Fraction) Distance and (Fraction) & [Rangei b Reported Performed (LLD) & [Rangel b Direction & [Rangel b Measurements AIRBORNE Particulates Not Required N/A N/A N/A N/A Not Required N/A DIRECT RADIATION

[mRlquarter] 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- 17.5] [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)

[pCil] /Aj.L..

[p..............................--- ---- ---- N/A) - - - - - -- - - -- - - - -- - - - - - - -- - - - - - - -

Tritium (54) 500 <MDA N/A N/A Not Required 2 (0/54)

[N/A_

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-51 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] t Analyses of Detection (Fraction) b Distance and (Fraction) & [Rangel b Reported Performed (LLD) & [Rangel b Direction & IRangel b Measurements WATERBORNE (continued)

Groundwater Gross Alpha 3 7+/-6 Monitoring Well 7 +/-6 N/A 2 (Monitoring wells) (22) (1/22) No. 2 (1/4) (0/4)

[pCi/I]- -------------------------------- U.---[7J _..--7------------------11.---

[N/A]

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

....- L7.--j ......... [-7:_5.L ...............

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

.... .. _NAj .... . . . . . . . . . . . . . N../.A L/A.....-----------------------------~

Tritium 500 (15/22) 461 +/- 64 Monitoring Well 484 +/- 94 444 +/-88 2 (22) 200 (7/22)c (7/22) No. 1 (3/5) ..

...................... .............................. [2.9E- 6_0J.. I4.9:. . . .L_. (4/5) 6j 9:A.. . . .. . .

p-.q-- ------------ I--J!Q: M9 ------- V? 1 -- --------

DrinkingWater Not Required ....... N/A .......... N/A ----...... N/A -----...... N/A -------- Not Reguired N/A Sediment Not Reqguired N/A N/A N/A N/A Not Reguired 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 ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-52 TABLE 4-I (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 1I/18/97 were analyzed to a lower than normal LLD of 200 pCi/I.

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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-53 4.2 Annual Radioactive Effluent Release Report This report shall be submitted prior to April I of each year. The following information shall be included: I I I

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 and Releases of Radioactive Materials in Liquid and Gaseous Effluentsfrom 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 IOCFR 50.36a and IOCFR 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 422 SAFSTOR/DECOMMISSIONING OFFSITE REVISION TITLE DOSE CALCULATION MANUAL PAGE 1-54 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 Maior Changes to Radioactive Waste Treatment Systems

a. Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid) shall be reported to the NRC in the Annual Radioactive 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 1-55 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 Decommissioning.

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 existing requirements 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 SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE IH-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). This requirement is defined by the following relationship.

Ci Canal l0x'ECLi 1 where:

Ci-.anai = The concentration of isotope" i" in the canal discharge point to Humboldt Bay.

ECL1 = Effluent Concentration Limit for radionuclide" i "from 10 CFR 20, Appendix B, Table 2, Column, 2 (pCi/ml) 1.1.1 The above relationship can be further defined for the concentration in the discharge effluent by taking into account the dilution factor provided in the discharge canal with dilution provided by the daily tidal exchange in the canal with Humboldt Bay.

This tidal dilution has been estimated by using a two-dimensional numerical analysis, depth-averaged, finite element hydrodynamic model, and validated by comparison of calculated and measured sample radioactivity concentrations (See reference 12.1).

Ci Cicanri = DFc.ana, (1-2)

NUCLEAR POWER GENERATION DEPARTMENT* SECTION ODCM VOLUME 422 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 where:

CQ = The concentration of radionuclide "i" in the effluent discharge.

DFcanai = Dilution factor provided in the Discharge Canal from tidal flow (taken as 50 for a liquid radioactive waste (LRW) tank batch release if the release commences approximately with low tide, there is an interval of at least four days between batch releases, and the release completes within the low-to-high part of the tide cycle. Use a dilution factor of 1 for batches which do not begin within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months before low tide or which do not end before one hour after the next high tide. Use a dilution factor of 1 for continuous discharge applications, or "batches" that exceed the tide times above).

For any batch release larger than a LRW tank (but within the same tide/time constraints), the dilution factor must be reduced by the ratio of the larger batch volume divided by the nominal LRW tank volume (e.g, a 75,000 gallon batch has a dilution factor of 5).

The preceding factors are based on minimal high tides and a conservative evaluation of outfall canal volume. Alternate values may be determined based on the ratio of actual canal volume (for a specific high tide) divided by the actual batch volume. The determination must be conservative (i.e., overestimate the batch volume and underestimate the canal volume) rather than nominal.

1.1 .2 The alarm setpoint (countrate) for each monitor is calculated from the relationships to canal concentration described in equations 1-1 and 1-2 above.

AR < [(DFc,,. 1)X (10)x (ECLc)x (KMix (0.85)]+ B (1-3) where:

AR = The alarm setpoint, counts per minute, of the RLEM.

KMiX = Calibration factor for the monitor, for the mixture, with units of cpm per micro-Ci/ml, further defined below.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-3 B = The monitor background reading (prior to any discharge) in counts per minute.

"ECLc = Composite Effluent Concentration Limit (ECL) for the mix of radionuclides (micro-Ci/ml) defined in detail below.

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

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

ECLc = i (1-4)

Ci f, ECLi iECLi 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" in the mixture.

1.1.4 The composite detector response for the mix of radionuclides is calculated as follows:

Z(Cq)(K1 )

KMi= (Ki)(f) - (1-5)

Ki for other nuclides is defined relative to Kcs-137 by the considering the summation of the other nuclides relative gamma detection efficiency, as follows:

Ki = (KCs- 137 { E(Rj'g)(Yjig) gamma (RCs -137,662 keV )(Y662 keV J

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

= Calibration factor for the monitor, for the nuclide "i", with units of cpm per micro-Ci/ml. " I

- Cs-137 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 pCi/ml for Cs-137.

Ri, g - Detector response factor for gamma "g" of nuclide "i", see figure below.

Yi, g - Gamma abundance for gamma "g" of nuclide "i".

RCs-13 7 , 662 keV - Detector response factor for 662 keV gamma of Cs- 137, 162 counts per gamma/ml (also equal to cpm per dpm/ml).

Yes-137,662 keV - Gamma abundance for 662 keV gamma of Cs-137 (85.1%).

Process Pvbnior Sensfitvty 240 220 I

~E2 Z E180 14-4--

12.

100 .

0 0.5 1 1.5 2 Gamma Energy (MeV)

For Co-60, the two gammas at 1172 and 1332 keV (each approx. 100 %

abundance) have a gamma response of about 138 and 132, so Kco.-60 is approximately 1.96. Similarly, KK-40 is about 0.1, and Ki for non-gamma nuclides is equal to zero.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 422 TITLE SAFSTOR/IDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE I1-5 1.1.5 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 (Bm., cpm) is:

Bmax = K x (5 x 10"6 )cpm (1-3)

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/IDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-6 1.1.6 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.

1.1.7 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 canal tidal dilution factor and no other liquid radwaste discharge is in progress.

1.1.8 Since the value of DFcanal in Equation 1-3 varies between batch and continuous modes of release, multiple concurrent discharges are not permitted.

1.1.9 If the Specification 2.3 alarm setting is calculated (per equation 1-3 above) for Cs-137 ECLc, -15% tolerance, no dilution, no factor of 10 for operational flexibility and for zero background the net 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 an administrative alarm setting. Refer to section 1.1.10 for the administrative (lower) alarm settings.

1.1.10For 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. The alarm should be set to the highest likely net reading, but no greater than that determined by equation 1-3 above. The highest likely net reading is based approximately on the sum of twice the typical background 2 and 130% of the predicted countrate for the batch3 .

2This is based on a nominal background of 500 cpm. As of 1/08/10, the background reading is about 473 cpm.

3 The 30% tolerance is for a combination of analytical and RLEM uncertainties and expected monitor fluctuations.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 422 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-7 1.2 GASEOUS EFFLUENT MONITOR 1.2.1 Equation C-4 of Regulatory Guide 1.109 demonstrates how to calculate dose from inhalation: I I The annual dose associatedwith inhalationof all radionuclides,to organj of an individual in age group a, is then:

Dja(r,0) = Ra Y-xi(r,0)DFAija where Dja is the annualdose rate to organj of an individual in age group a Ra is the breathingratefor age group a xi(r,0) is the annual averageground-level 3 concentrationof nuclide i in air in sector 0 at distance r, in pCi/mr DFAija is the dosefactorfornuclide i to organj of age group a To calculate xi(r,0) the annual average ground-level concentration of nuclide i in air in sector 0 at distance r, in pCi/in 3 the equation must be rearranged to:

Dja(r,0)/( DFAija Ra) = xi(r,O)

Assuming that:

Americium-241 is the primary nuclide The maximally exposed group is the Teen based on breathing rates and DFAija The DFAija to the bone of a Teen from Am-241 is 1.77 mrem/pCi The DFAija are taken from: NRC NUREG-4013, "LADTAP-II Technical Reference and User Guide" The Teen breathing rate is 8000 m3/year The release happens at a release rate of 30,000 cfm

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE IH-8 Therefore the ground-level concentration of Am-241 in air in sector 0 at distance r, in pCi/m 3 that will produce a dose rate of 1500 mrem/year to the bone of a Teen is:

3 (1500 mrem/year) / (1."77 mremn/pCi) / (8000 ni /year) = 1.06E--1 pCi/ m3 3

1.06E-I pCi/ m =

(1.06E-1 pCi/m3) / (1E6 pCi /j.Ci) / (1E6 mI/m 3) = 1.06E-13 IiCi/ml 1.2.2 Quantity of radioactive material released Equation C-3 of Regulatory Guide 1.109 demonstrates how to calculate the quantity of material that must be release to produce a given airborne concentration:

The annualaverage airborneconcentration of radionuclidei at the location (r, 0) with respect to the releasepoint may be determined as xi(r,O) = 3.17 x 104 Qi(X/Q)D(rO) where xi(r,0) is the annual averageground-level concentrationof nuclide i in air 3

in sector 0 at distance r, in pCi/mr 3.17 x 104 is the number ofpCi/Ci divided by the number of sec/yr (x/Q) (r,0) is the annual average atmosphere dispersionfactor, in sec/Im 3.

Qi is the release rate of nuclide I to the atmosphere, in Ci/yr A value of 7.3E-6 sec/mi3 was used for the annual average atmosphere dispersion factor at the site boundary (X/Q)D(r,0). This is based on a release rate of 30,000 cfm.

This value is obtained from Calculation N-238C, Rev. 0, "DeterminingEffect of HBPP Unit 3 Stack Reconfigurationon Downwind Effluent Concentrations".

To determine the release rate that will result in an average ground-level concentration the above equation must be rearranged to:

Qi = xi(r,0) / (3.17 x 10 4(X/Q)D(r,0))

Therefore the release rate of Am-241 required to equal the annual average ground-level concentration at the site boundary calculated above is:

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-9 1.06E-1 pCi/m3 / ((3.17E4 (pCi/Ci) (sec/yr)) * (7.3E-6 sec/m 3)) =

4.61E-1 Ci/yror4.61E5 uCi/yr 1.2.3 Transmission Fraction Particulate depositional losses will occur in the external transport sample lines that connect the MGP ABPM20I S alpha beta particulate monitor to the stack. These depositional losses have been calculated to determine a conservative correction factor.

Based on this calculation the transmission fraction at 30,000 scfm is 80.4%. The inverse of the transmission fraction, 1.24, defines a correction coefficient which can be applied to the release rate and public dose rate calculations of the stack monitor.

1.2.4 Stack Concentration The stack concentration that would result in a release rate of 4.61 E-1 Ci/yr is equal to:

Total release (Curies/year) / Release rate (cc/year)

The average annual stack flow rate is 30,000 cfm This results in a total volume of4.47E14 cc/yr This is based on (30,000 ft3/min

525,600 minutes/yr

28,317 cc/ft3).

(4.61E-1 Ci

1E6 uCi/Ci) / (4.47E14 cc/yr) = 1.03 E-9 uCi/cc Correcting for the transmission fraction this is equal to:

1.03 E-9 uCi/cc

0.804 = 8.28E-10 uCi/cc Therefore an indicated stack concentration of 8.28 E-1 0 itCi/cc at 30,000 cfm for one calendar year would result in a dose of 1500 mrem to a member of the public at the site boundary.

Two times the release rate is equal tol.66E-9 jiCi/cc.

Two hundred times the release rate is equal to 1.66E-7 pCi/cc.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 422 RVJSION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-10 1.2.5 Relationship to EPA PAG To compare the release rates"calculated above the following assumptions were made:

Am-241 dose conversion factor in rem / cm"3 uCi hr, from EPA 400 = 5.3E8 A value of3.71E-4 sec/m 3 was used for the atmosphere dispersion factor (X/Q)D(rO). This value is obtained from Safstor ODCM Appendix B "Bases for Atmospheric Dispersionand Deposition Values".

Assuming that an unplanned release occurs at two times the ODCM release rate for one hour the total activity released is equal to:

1.03 E-9 lICi/cc

2 = 2.06E-9 pCi/cc 2.06E-9 pCi/cc

30,000 ft3/min

28,317 cc/ft3

60 min = 1.05E2 ptCi (1.05E2 [tCi) * (5.3E8 rem / cm"3 uCi hr) * (3.71E-4 sec/mi3) / (1E6 cm 3/m 3)/

(3600 sec/hour) = 5.74E-3 rem This is much less than the EPA PAG of 1 Rem Assuming that an unplanned release occurs at two hundred times the ODCM release rate for 15 minutes the total activity released is equal to:

1.03 E-9 [iCi/cc

200 = 2.06E-7 [iCi/cc 2.06E-7 pCi/cc

30,000 ft3/min

28,317 cc/ft3

15 min = 2.62E3 pCi This results in a dose of:

(2.62E3 [LCi) * (5.3E8 rem / cm"3 uCi hr) * (3.71E-4 sec/mr3) / (1E6 cm 3/m 3)/

(3600 sec/hour) =

1.43E-1 rem This is much less than the EPA PAG of I Rem 1.2.6 Relationship to IOCFR20 Appendix B Table 2 Effluent Concentration limits

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE IH-11 The 10CFR20 Appendix B Table 2 Effluent Concentration limit for Am-241 is 2E-14 RCi/ml.

The average annual grounid-level concentration in'air (xi) in pCi/m 3 is equal to:

x = (3.17E4 (pCi/Ci)/ (see/year))

Q * (X/Q)

Where Q is equal to the quantity of radioactive material released in a year in Curies/year 3

ODCM average annual X/Q = 7.3E-6 sec/ m If xi= 2E-14 uCi/mI then:

Q = (2E-14 uCi/ml

IE6 mI/m 3

1E6 pCi/uCi) / ((3.17E4 (pCi/Ci)/

(sec/yr)*(7.3E-6 sec/ mi3 ))

Q = 8.64E-2 Ci/yr The average annual stack volume based on the ODCM is 4.47E14 cc/yr.

This is based on (30,000 cfm

525,600 minutes/yr

28,317 cc/cfm).

Therefore, the stack concentration required to result in a fence-line concentration of 2E-14 uCi/mI is:

(8.64-2 Ci/yr

IE6 uCi/Ci) / (4.47E14 cc/yr

I cc/ml) = 1.93 E-10 uCi/mI Correcting for the transmission fraction this is equal to 1.93 E-10 uCi/mI

0.804 = 1.55E-10 uCi/ml 1.2.7 SPAM Conversion Factor from Effluent Concentration to p.Ci/day The release rate in ýtCi/day = stack concentration in jLCi/cc

30,000 ft3/min

1440 minutes/day

28317 cC/ ft3

transmission factor of 1.24 The release rate in p.Ci/day = stack concentration in liCi/cc

1.52E1 2 ý.Ci/day 1.2.8 Conversion Factor from pCi/day to % of NUE An NUE is equal to a release rate of 3000 mrem/year

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-12

%NUE = (Offsite dose rate / NUE threshold)

100

%NUE = ((Conversion Factor-* Release Rate) / NUE threshold)

100

%NUE = ((Conversion Factor

100) / NUE threshold)

Release Rate 3

The Conversion Factor is equal to (1.77E6 mrem/uCi) * (7.3E-6 sec/i ) * (8000 m3/year) / (8.64E4 sec/day)

This is equal to 1.20 mrem/year per pCi/day 1.2.9 Results The 10CFR20 Appendix B Table 2 Effluent Concentration limit for Am-241 is 2E-14 pCi/ml. The SPAM indication that would result in a fence-line concentration of 2E-14 uCi/ml is 1.55 E-10 uCi/ml. This is approximately equal to 10% of an NUE.

This value is used as the high alarm setpoint for the SPAM.

A NUE is equal to two times the ODCM release rate limit and this is equal to a SPAM indication of 1.66 E-9 pCi/cc. This value is used as the high high alarm setpoint for the SPAM.

Two hundred times the ODCM release rate is equal to a SPAM indication of 1.66 E-7 pCi/cc.

Assuming that an unplanned release occurs at two times the ODCM release rate for one hour the offsite dose corresponding to an NUE would be 5.74E-3 rem (5.74 mrem) which is much less than the EPA PAG.

Assuming that an unplanned release occurs at two hundred times the ODCM release rate for fifteen minutes the offsite dose corresponding to an Alert would be 1.43E-1 rem (143 mrem) which is much less than the EPA PAG.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 422 REVISION TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-13 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 value for D (dose commitment annual rate, mrem/year) 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 QUARTER 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 by 4 to convert it to a quarterly dose commitment, mrem/quarter.

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.

Equation (2) of Regulatory Guide 1.109 provides for the use of a site specific mixing ratio (i.e. reciprocal of the dilution factor) that describes the near term and near field mixing of the tidal flow from the Discharge Canal into Humboldt Bay. A two-dimensional numerical analysis, depth-averaged, finite element hydrodynamic model (reference 12.1) was developed by CH2MHILL and used to estimate the dispersion of the canal discharge in the Bay. The analysis indicated that an additional dilution factor of 80 for batch release applications or a dilution factor of 20 for continuous release applications can conservatively be used to describe the Bay dilution. A factor of 20 will be applied in this calculation to address any combination of release modes.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-14 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:

D - [c- .Bay diluled X DF x {(BFish. i X UOish) + (Bi.j.i x Umnv)}] (2-1) i=l where:

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

Ci-Bay diluted = The average diluted Bay 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 tidal flow during the period in liters, and by the Bay dilution factor of 20. The total annual tidal flow for the outfall canal is 2.47E9 Liters. If Gross Alpha radioactivity is determined to be in the discharge, Pu-241 will be considered to be present at 3.25 times the amount of detected Gross Alpha radioactivity. Note that the resulting dose commitment is theannual dose rate (torem 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORLDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-15 Uinv 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.

Dose calculations after the transition from circulating water flow to tidal dilution flow can be performed using the new methodology for the current month, quarter, or year. This new methodology does not change the dose results significantly from the previous methodology, and therefore will not adversely affect the previous comparison with regulatory limits.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-16 Table 2-1 Ingestion Dose Factors for Adult Age Group (mreim/pico-Curie ingested)

Selected Nuclides from NUREG-4013 (LADTAP II input values)

Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 5.99 x 10-8 5.99 x 10-8 5.99 x 10-8 5.99 x 10-8 5.99 x 10-8 Co-60 No Data 2.14 x 10-6 4.72 x 10-6 No Data No Data 4.02 x 10-5 Ni-63 1.30 x 10 -4 9.01 x 10-6 4.36 x 10-6 No Data No Data 1.88 x 10-6 Sr-90 8.71 x 10- 3 No Data 1.75 x 10 -4 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 1 0 -4 Pu-241 1.57 x 10-5 7.45 x 10-7 3.32 x 10-7 1.53 x 10- 6 No Data 1.40 x 10-6 Am-241 7.55 x 10-4 7.05 x 10- 4 5.41 x 10-5 4.07 x 10-4 No Data 7.42 x 10-5 Gross a 7.55 x 10-4 7.05 x 10-4 5A4 x 10-5 4.07 x 10-4 No Data 7.42 x 10-5 Table 2-2 Ingestion Dose Factors for Teen Age Group (mrem/pico-Curie ingested)

Selected Nuclides from NUREG-4013 (LADTAP II input values)

Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI 8 6.04 x 10-8 6.04 x 10-8 H-3 No Data 6.04 x 10-8 6.04 x 10-8 6.04 x i0-6 No Data No Data 3.66 x 10-5 Co-60 No Data 2.81 x 10-6 6.33 x 10-Ni-63 1.77 x 10-4 1.25 x 10- 5 6.00 x 10- 6 No Data No Data 1.99 x 10-6 Sr-90 1.02 x 10-2 No Data 2.04 x 10- 4 No Data No Data 2.33 x 10-4 Cs-137 1.12x 10-4 1.49x 10 4 5.19x 10-5 5.07x 10-5 1.97x 10-5 2.12x 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 Am-241 7.98 x 10-4 7.53 x 10-4 5.75 x 10-5 4.31 x 10-4 No Data 7.87 x 10-5 Gross a 7.98 x 10-4 7.53 x 10-4 5.75 x 10-5 4.31 x 10- 4 No Data 7.87 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-17 Table 2-3 Ingestion Dose Factors for Child Age Group I (mrem/pico-Curie ingested)

Selected Nuclides from NUREG-4013 (ladTAP II input values)

Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.16x l0-7 1.16x 10-7 1.16x 10-7 1.16x 10-7 1.16x 10-7 Co-60 No Data 5.29 x 10-6 1.56 x 10- 5 No Data No Data 2.93 x 10-5 Ni-63 5.38 x 10-4 2.88 x 10-5 1.83 x 10-5 No Data No Data 1.94 x 10-6 Sr-90 2.56 x 10-2 No Data 5.15 x 10-4 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 Am-241 1.36 x 10-3 1.17 x 10-3 1.02 x 10-4 6.23 x 10-4 No Data 7.64 x 10-5 Gross ct 1.36 x 10-3 1.17 x 10-3 1.02 x 104 6.23 x 10-4 No Data 7.64 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-1 and from NUREG-4013 Element I Fish I Invertebrate H 9.0x 10"1 9.3 x 10-1 Co 1.0 x 102 1.0 x 103 Ni 1.0 x 102 2.5 x 102 Sr 2.0 2.0 x 101 Cs 4.0 x 10 1 2.5 x 10 1 Y 2.5x 10 1 1.0x 103 Pu 3.0 2.0 x 102 Am 2.5 x 101 1.0 x 103 Gross cc 2.5 x 10 1 1.0 x 103

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-18 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 Fruits and Age Group Fish (Invertebrates) Vegetables Milk Meat Adult 6.9 1.0 190 110 95 Teen 5.2 0.75 240 200 59 Child 2.2 0.33 200 170 37 Infant 0 0 0 0 0 Table 2-6 Maximum Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-19 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. 001E (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 SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-20

c. Reactor Caisson Sump The reactor caisson sump and its associated reactor caisson sump pump is 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 are as specified in the DSAR, Section 2.2.2.3.

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 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-21 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 is not 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, orem/year) 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-22 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 3.25 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:

D n=--[Qi x (RPh,i + Rrp.i + Rmeat,i + RMilki + Rveg.i)] .(4-3) i=1L 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 422 SAFSTOR/DECOMMISSIONING OFFSITE REVISION TITLE DOSE CALCULATION MANUAL PAGE 11-23 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,j = " The dose factor for th& 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 Decommissioning releases range from about 0.002 mremlyear (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 R,,i = (X/Q) x BIb x DFi,. (4-3a) 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.

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 SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-24 4.3.5 Particulate Ground Plane Pathway Dose Calculation Methodology R = (D/Q) x SF x DF, x K x W (4-3b) where:

K unit conversion constant, 8760 hr/yr.

DFi The ground plane dose conversion factor for radionuclide, i, in mremlhr per pCi/m 2 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-1 5 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 10.8 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 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-25 4.3.6 Particulate Grass-Cow-Milk Pathway Dose Calculation Methodology Rmilk, (D/Q) x (QF x Ua x Fm x DFiaX Wx (4-3c)

Y 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, or4-11.

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 422 REVISION TITLE SAFSTOR/IDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE H-26 4.3.7 Particulate Grass-Cow-Meat Pathway Dose Calculation Methodology (Qi:x U, x Ff x DR, x W RMeat,. (D/Q) x .I (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.

DR.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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-27 4.3.8 Particulate Vegetation Pathway Dose Calculation Methodology Rveg.i = (D/Q) x (UT x DFY.a_) - (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 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.

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 VOLUME 4 TITLE SAFSTORJDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-28 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 (Rnmýl.H3 + RGP, H3 + RMet, H3 + RMI~k, H3 + RvegH3) (4-4) where:

D = Annual dose commitment, mrem/year.

QH 3 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.

R-Meat, H3 The dose factor for the grass-cow-meat pathway for H-3, mrem/year per pico-Curie/sec.

Rmilk,I-H3 = The dose factor for the grass-cow-milk pathway for H-3, mrem/year per pico-Curie/sec.

RVeg,H3 = The dose factor for the vegetation consumption pathway, mrem/year per pico-Curie/sec.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-29 4.3.10 Tritium Inhalation Pathway Dose Calculation Methodology Rlflh, = (HQ3 x BR3 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 SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-30 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:

v /Q) (0.75 x 0.5)

Rmkk. = I x/ x Qx U. x Fm 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 10s 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-31 4.3.12 Tritium Grass-Cow-Meat Pathway Dose Calculation Methodology RMeat, H3 = xQ) (0.75x 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.

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.

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 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 VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-32 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 x Ur 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/im 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.

=/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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-33 Table 4-1 Inhalation Dose Factors for Adult Age Group (mr'em/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 10-7 1.58 x 10-7 1.58 x 10-7 Co-60 No Data 1.44 x 10-6 1.85 x 10-6 No Data 7.46 x 10-4 3.56 x 10-5 Sr-90 1.24 x 10-2 No Data 7.62 x 10 -4 No Data 1.20 x 10-3 9.02 x 10-5 Cs-137 5.98 x 10-5 7.76 x 10- 5 5.35 x 10- 5 2.78 x 10-5 9.40 x 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 1.29x 10-31 5.93 x 10-3 1.52 x 10 -4 8.65 x 10-7 Gross a 1.68 1.13 7.75x 10-2 5.04x 10-1 1.82x 10-1 4.84x 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 I Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 Co-60 No Data 1.89 x 10- 6 2.48 x -6 No Data 1.09 x 10- 3 3.24x 0-5 Sr-90 1.35 x 10-2 No Data 8.35 x 10- 4 No Data 2.06 x 10-3 9.56 x 10-5 Cs-137 8.38 x 10-5 1.06 x 10 -4 3.89x 10-5 3.80x 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 1 0 -4 9.17 x 10-7 Gross a 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 SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-34 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 Or an Nuclide Bone Liver Total Body Kidney Lung GI-LLI 7 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 H-3 No Data 3.04 x 10-Co-60 No Data 3.55 x 10-6 6.12 x 10-6 No Data 1.91 x 10-3 2.60 x 10-5 Sr-90 2.73 x 10-2 No Data 1.74 x 10- 3 No Data 3.99 x 10-3 9.28 x 10-5 Cs-137 2.45 x 10-4 2.23 x 1 0-4 3.47 x 10-5 7.63 x 10-5 2.81 x 10-5 9.78 x 10-7 Y-90 1.11 x 10-6 No Data 2.99 x 10-8 No Data 7.07 x 10- 5 7.24 x 10-5 Pu-241 7.94 x 10-2 1.75 x 10-2 2.93 x 10 3 1.10 x 10-2 5.06 x 1 0 - 8.90 x 10-7 Gross cx 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 -3 2.28 x 10 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 1 0 -4 5.09 x 10-5 9.53 x 10-7 Y-90 2.35 x 10-6 No Data 6.30 x 10-8 No Data 1.92 x 10- 4 7.43 x 10-5 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 1 0 -4 8.97 x 10-7 Gross (x 3.15 1.95 1.34 x 10-1 7.94 x 10-1 9.03 x 10-1 5.02 x 10-

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-35 Table 4-5 External Dose Factors for Standing on Contaminated Ground (mrem/hciur per pico-Curie/square meter)

Selected Nuclides from Regulatory Guide 1.109, Table E-6 Total Nuclide Skin Body 1--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-137 4.90 x 10-9 4.20 x 10-9 Y-90 2.60 x 10" 12 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 1.109, Table E-4 Other Seafood Fruits and Age Group Fish (Invertebrates) Vegetables Milk Meat Adult 6.9 1.0 190 110 95 Teen 5.2 0.75 240 200 59 Child 2.2 0.33 200 170 37 Infant 0 0 0 0 0 Table 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-36 Table 4-8 Ingestion Dose Factors for Adult Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 11 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 Co-60 No Data 2.14 x 10-6 4.72 x 10-6 No Data No Data 4.02 x 10-5 Sr-90 7.58 x 10- 3 No Data 1.86 x 10-3 No Data No Data 2.19 x 10 -4 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- 7 3.32 x 10-7 1.53 x 10-6 No Data 1.40 x 10-6 Gross cx 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 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 H-3 No Data 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 Co-60 No Data 2.81 x 10-6 6.33 x 10-6 No Data No Data 3.66 x 10-5 Sr-90 8.30 x 10-3 No Data 2.05 x 10- 3 No Data No Data 2.33 x 10-4 Cs-137 1.12 x 10 -4 1.49x 10 -4 5.19 x 10- 5 5.07 x 10-5 1.97x 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 (x 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-37 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 Co-60 No Data 5.29 x 10-6 1.56 x 10-5 No Data No Data 2.93 x 10-5 Sr-90 1.70 x 10-2 No Data 4.31 x 10-3 No Data No Data 2.29 x 10-4 Cs-137 3.27 x 1 0 -4 3.13 x 10 -4 4.62 x 10-5 1.02 x 1 0 -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 1 0 -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 a 1.36 x 10-3 1.17 x 10-3 1.02 x 1 0 -4 6.23 x 10- 4 No Data 8.03 x 10-5 Table 4-1 1 Ingestion Dose Factors for Infant Age 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-LL1 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 x 10-2 No Data 4.71 x 10-3 No Data No Data 2.31 x 10-4 Cs-137 5.22 x 10-4 6.11 x 10-4 4.33 x 10-5 1.64 x 10-4 6.64 x 10-5 1.91 x 10-6 Y-90 8.69 x 10-8 No Data 2.33 x 10- 9 No Data No Data 1.20 x 10 -4 Pu-241 4.25 x 10-5 1.76 x 10-6 8.82 x 10-7 3.17 x 10-6 No Data 1.45 x 10-6 Grossa 1.46x 10-3 1.27x 10-3 1.09x 1 0 -4 6.55x 10 -4 No Data 8.10x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-38 Table 4-12 Stable Element Transfer Data For Cow-Milk Pathway I (days/liter) I Selected Nuclides from Regulatory Guide 1.109, Table E-I and from NUREG-4013 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 ox 5.0 x 10-6 Table 4-13 Stable Element Transfer Data For Cow-Meat Pathway (days/kilo-gram)

Selected Nuclides from Regulatory Guide 1.109, Table E-I and from NUREG-4013 Element Ff H 1.2 x 10-2 Co 1.3 x 10-2 Sr 6.0 x 10-4 Cs 4.0 x i0-3 Y 4.6 x 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-39 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 is significant 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 of Regulatory Guide 1.109, as the maximum time for shoreline recreation for the Teen age group.

5.2 SK[N 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 SAFSTORIDECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-40 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-41 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 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-42 7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED IN HIGH INTEGRITY CONTAINERS 7.1 SCOPEI 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/DECOMMISSIONING OFFSITE REVISION 22 DOSE CALCULATION MANUAL PAGE 11-43 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 REVISION 22 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-44 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 Decommissioning 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 is allowed to modify or reduce environmental requirements in the ODCM provided HBPP considers the modification or reduction from a technical and decommissioning perspective. [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

12.0 REFERENCES

12.1 TBD-208, "Outfall Canal Effluent Dilution Factors".

ODCM APPENDIX A Revision 22 Page A-I APPENDIX A SAFSTOR BASELINE CONDITIONS

ODCM APPENDIX B Revision 22 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 I (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 104) 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 10-5) 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 DispersionModels for 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 22 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) 1 X/Q = U~o(37O'a*) (2)

X/Q (3) where:

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

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

C =z 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 meters 2 , based on the Refueling Building dimensions (about 36 feet high, about 112 feet long).

Fy = lateral plume spread (including meander and building wake), meters, equal to 6uy (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 (l)

ODCM APPENDIX B Revision 22 Page B-5 Figure 6 provides a Relative Deposition Rate value of 1.2 x 104 meter-'. 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 6ay., 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 22 Page C-I APPENDIX C Deleted J1

HBL-13-004, ODCM, Revision 22, Volume 4, Safstor/Decommissioning Offsite Dose Calculation Manual.

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