Text

Annual Radioactive Effluent Release Report for 2015
	ML16089A201
Person / Time
Site:	Humboldt Bay
Issue date:	03/24/2016
From:	Sharp L; Pacific Gas & Electric Co
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
	HBL-16-003
	Download: ML16089A201 (121)
	v • d • e

Loren D. Sharp Humboldt Bay Power Plant Pacific Gas and Director & Plant Manager 1000 King Salmon Avenue Electric Company"' Humboldt Bay' Nuclear Eureka, CA 95503 707.444.0819 Internal: 375.0819 Fax: 707.444.0871 March 24, 2016 PG&E Letter HBL-16-003 U.S. Nuclear Regulatory Commission 10 CFR 50, Appendix I ATTN: Document Control Desk 10 CFR 50.36 .

Washington, DC 20555-0001 Docket No. 50-133, OL-DPR-7 Humboldt Bay Power Plant Unit 3 Annual Radioactive Effluent Release Report for 2015

Dear Commissioners and Staff:

Enclosure 1 contains the Humboldt Bay Power Plant Unit 3 "Annual Radioactive Effluent Release Report," covering the period January 1 through December 31, 2015. This report is required by Appendix B, Section 6.3 of the Humboldt Bay Quality Assurance Plan, and by Section 4.2 of the "SAFSTOR/Decommissioning Offsite Dose Calculation Manual."

Enclosure 2 contains Revjsion 26 to the "SAFSTOR/Decommissioning* Offsite Dose Calculation Manual" as required by Specification Section 4.2 of the "SAFSTOR/Decommissioning Offsite Dose Calculation Manual."

There are no new.or revised regulatory commitments (as defined in NEI 99-04) in this letter.

If you have any questions concerning this information, please do not hesitate to contact Mr.Hossein Hamzehee at (805) 545-4720 .

.Sincerely, Loren D. Sharp cc: Marc L. Dapas, NRG Region IV Administrator John B. Hickman, NRG Project Manager HBPP Humboldt Distribution

. Enclosures

Enclosure 1 PG&E Letter HBL-16-003 PACIFIC GAS AND ELECTRIC COMPANY

HUMBOLDT BAY POWER PLANT DOCKET NO. 50-133, LICENSE NO. DPR-7 HUMBOLDT BAY POWER PLANT UNIT 3 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January 1 through December 31, 2015 /

Enclosure 1 PG&E Letter HBL-16-003 TABLE OF CONTENTS INTRODUCTION ......... .. ..... ...... ....... .... ....... ....... ...... .. ........... ..... .. ... ............. .. ... ........... .. 1 I. .

SUPPLEMENTAL INFORMATION .......................... :........................................... . 2 II. GASEOUS AND LIQUID EFFLUENTS................................................................ 6

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Table 1 - Gaseous Effluents - Summation of All Releases .. :......................... ,...... 7 Table 2A- Gaseous Effluents - Elevated Release - Nuclides Released ............... 8 Table 2B - Gaseous Effluents - Ground-Level Releases-. Nuclides Released ...... 9 Table 3 - Liquid Effluents - Summation of All Releases ............................... _....... 10 Table 4 - Liquid Effluents - Nuclides Released ................................................... 11 Ill. SOUD RADIOACTIVE WASTE.......................................................................... 12 Table 5 - Solid Waste and Irradiated Fuel Shipments ................ :.......................... 12 IV. RADIOLOGICAL IMPACT ON MAN ......... ..... .. .............. ... ... ....... ..... .... .......... ..... 16 T~ble 6 - Radiation Dose for Maximally Exposed Individuals .............................. 17 V. . CHANGES TO THE OFFSITE DOSE CALCULATIC?N MANUAL (O_DCM) ......... 18 ODCM Revision 26 (Effective 10/20/2015) ................................... :...................... 18 VI. CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)......................... 22

. VII. CHANGES TO.RADIOACTIVE WASTE TREATMENT SYSTEMS ..................... 22 VIII. INOPERABLE EFFLUENT MONITORING INSTRUMENTATION ...................... 23

. IX. ERRATA ....................................................................................................~......... 24 1

Enclosure 1 PG&E Letter HBL-16-003 INTRODUCTION This report summarizes gaseous and liquid radioactive effluent releases from Humboldt Bay Power Plant (HBPP) Unit 3 for the four quarters of 2015. The report includes calculated potential radiation doses from these radioactive effluents and a comparison with t~e numerical guidelines of 10 CFR 50, Appendix I, as well as a summary of shipments of solid radioactive waste. The concentrations of plant effluent releases during the reporting

. period were well below Offsite Dose Calculation Manu~I (ODCM) limits.

The HBPP Main Plant Stack, a ground level release path, and SPAMS, the real time stack effluent monitor, were shut down on 10/14/2015 and permanently removed from service to*

facilitate partial demolition of the Reactor Building. Modular HEPA units continue to be monitored as a potential gaseous effluent pathway.

The information' is reported as required by Appendix 8, Section 8.3 of the Humboldt Bay Quality Assurance Plan and Section 4.2 of the ODCM, and it is presented in the general format of Regulatory Guide 1.21, Appendix B (except for the topics identified below).

Meteorology The meteorological data logging system was removed from service in 1967 so the information specified by Regulatory Guide 1.21 .is not available. Previous HBPP Annual Radioactive Effluent Release Reports summarized the cumulative joint frequency

. distribution of wind speed, direction, and atmospheric stability for the period April 1962 through June 1967, when the meteorological data logging system was in service.

Short-lived Nuclides, Iodine and Noble Gasses

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The Unit was last operated on 7/2/1976. Due to the long decay time since operation, shdrt-

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. lived radionuclides are neither expected nor reported. This includes Iodines and noble gases other than Kr-85. During 2008, all of the spent nuclear*fuel was transferred from the Spent Fuel Pool to the Independent Spent Fuel Storage Installation (ISFSI), so there is now no source term for Kr-85.

Air Particulate Filter Composites - Sr-90 Air particulate sample filters are composited quarterly and analyzed off-site for Sr-90 ..

i Air Particulate Filter Composites - Am-241 Air particulate sample filters are composited quarterly and analyzed off-site for Am-241.

Air Particulate Filter Composites-- Gross Alpha /

Each weekly sample filter is individually 9ou.nted for gross alpha activity, rather than analyzing a *monthly composite of the filters, as described in Regulatory Guide 1.21.

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Enclosure 1 PG&E Letter HBL-16-003 Gaseous Effluents - Tritium Tritium releases during plant operation were less than detection levels. Because the plant was permanently shut down in 1976, current tritiurri release levels are less than the release levels that occurred during plant operations. Therefore, no tritium samples were collected during this reporting period. Since the fuel has been relocated to the ISFSI and the Spent Fuel Pool water is below the drinking water standard, no significant tritium can be released by the gaseous mode. The Spent Fuel Pool water was removed from the site in 2015, by shipment for disposal.

Liquid Effluents The last batch discharge of radioactive liquid effluent occurred on 12/11/2013. Subsequent radioactive liquid effluent batches were transported to US Ecology for offsite disposal under the 10 CFR 20.2002 exemption. These shipments, volumes, and activity totals are included in Table 5 of this report.

Average Energy, Calculation$ for the average energy of gaseous releases of fission and activation gases are not required for HBPP.

I. SUPPLEMENTAL INFORMATION

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A. Regulatory Limits

(

1. Gaseous Effluents

a. Noble Gas Release Rate Limit Noble gases are no lor:iger an issue since the spent nuclear fuel has been relocated to the l,SFSI. *

(

b. Iodine Release Rate Limit Due to the.long decay time since the Unit was shutdown, the license does not define an iodine relea~e rate limit.

Enclosure 1 PG&E Letter HBL-16-003

c. Particulate Release Rate Limit The radioactive particulate release rate limit is based on concentration i limits from 10 CFR 20, divided by an annual average dispersion factor for the sector with the least favorable atmospheric dispersion. If the total release for a pe.riod is determined to be a "less than" value, the limits are based on analytical results obtained in November, 2005, for which the mixture was determined to be 84% Cs-137, 11 % Co-60 and 5% Sr-90.

The applicable annual average dispersion factors for plant stack and for incidental releases are 1.0E-5 and 6.59E-3 seconds per cubic meter, respectively. When both plant stack and incidental releases occur, the "percent of applicable limit" in Table 1 is the sum of the values for "percent of applicable limit" for each of the release paths.

2. Liquid Effluents

a. Concentration Limit Concentration limits for liquid effluent radioactivity released to Humboldt Bay are taken from 10 CFR 20.

B. Maximum Permissible Concentrations

1. Gaseous Effluents Maximum Permissible Concentrations for gaseous effluents are taken from 10 CFR 20, Appendix 8, Table 2, Column 1.

2. Liquid Effluents Maximum Permissible Concentrations for liquid effluents are taken from 10 CFR 20, Appendix 8, Table 2, Column 2.

C. Measurements and Approximations of Total Radioactivity

1. Gaseous Effluents - Elevated Release The original plant stack (an elevated release point) was removed in 1998 and replaced with a roof-level discharge point that referred to as the plant stack, but considered a ground level release point. Ventilation and system vents were routed to this release point or to modular HEPA ventilation units. Following plant stack shutdown (10/14/2015), incidental ventilation used ground level modular HEPA ventilation units. Therefore, elevated releases did not occur at HBPP during 2015.

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Enclosure 1 PG&E Letter HBL-16-003

2. Gaseous Effluents - Ground-level Release

a. Fission and Activation Gases Fission and activation gases are no longer an *issue since the spent fuel has been relocated to the ISFSI.

b. Iodines Due to the long decay time since operation (shutdown 7/2/1976), no detectable releases .of radioactive Iodines can be expected. Therefore, .

neither the Technical Specifications nor the ODCM require that these radionuclides be monitored.

c. Particulates A continuous monitor equipped.with an alpha spectrometer, with its response calibrated for Am-241, rnonitored the alpha particulate activity

.released from the stack. This monitor was installed in December of 2009.

This monitoring ended 10/14/2015 when the plant stack was shutdown.

Radioactive particulates released from the plant stack were monitored by continuous sample collection on particulate filters. Filter papers were removed _from th.e stack sampling system weekly, and analyzed for the concentration of gamma-emitting nuclides using c:m intrinsic germanium detector. All statistically significant gamma peaks are identified,.

Radioactive particulates released from modular HEPA ventilation units are monitored by continuous sample collection on particulate filters. Filter papers cire removed from modular ventilation system weekly,. and are analyzed for the concentration of gamma-emitting nu elides using an intrinsic germanium detector. All statistically significant gamma peaks are identified.

After decaying at least seven days, the filters are analyzed for gross alpha radio?ctivity using a scintillation counter, Filters are composited and analyzed quarterly for Strontium-90 (the only fadioactive Strontium present) and Americium-241 by alpha spectroscopy.

The estimated error of the reported particulate release values is based on uncertainty in sample flow rate, stack flow rate, modular HEPA unit flow rate, detector calibri:ltion, and typical sample counting statistics.

The Minimum Detectable Activity (MDA) for all particulate filter samples was less than the applicable Lower Limit of Detection (LLD) presented in the ODCM.

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Enclosure 1 PG&E Letter HBL-16-003 Individual sample release results are assigned to calendar quarters as of the termination of the sample period. Composite sample release results are assigned to the applicable calendar quarter. The release activity is sample' concentration multiplied by sample duration and nominal release flow rate (30,500 cfm for the stack or 2,000 cfm for modular HEPA units).

The Main Plant Stack, a ground level release path, and SPAMs, the real time effluent monitor, were shut down on 10114/2015 and permanently removed from service to facilitate partial demolition of the Reactor Building. Incidental releases from modular HEPA units continue to be monitored as a potential gaseous effluent pathway.

3. Liquid Effluents

a. Batch Releases There were no batch liquid effluent releases during this report period.

b. Continuous Releases*

There were no _continuous liquid effluent releases during this report period.

D. Batch Release Statistics

, 1. Liquid

a. Number of batch releases..................................................... 0

b. Total time period for batch releases ....................... :............ NIA

c. Maximum time period for' a batch release .. :....................... NIA

d. Average time period for a batch release ......: ..................... NIA

e. Minimum time period for a batch release ............................ NIA

2. Gaseous

a.

Number of batch releases.................................................... O

b. Total time period for batch releases ................................... NIA

c. Maximum time period for a batch release: .................... ~ .... NIA

-d. Average time period for a batch release ............................ NIA

e. Minimum time period for a batch release........................... NIA 5

Enclosure 1 PG&E Letter H8L-16-003 E. Abnormal Release Statistics

1. Liquid

a.

Number of abnormal releases.............................................. 0

b. Total activity released....................................................... N/A

2. Gaseous

a. Number of abnormal releases ........................................ :...... 0

b.

Total activity released....................................................... N/A II. GASEOUS AND LIQUID EFFLUENTS A. Gaseous Effluents

. Table 1 summarizes the total quantities of radioactive gaseous effluents released.

Section A of Table 1 ha~ been omitted as Fission & Activation Gases are neither.

expected or measured. Table 2A is for reporting the quantities of each of these nuclides determined to be released from an elevated release point (there are none). Table 28 presents the quantities of each of the nuclides determined to be released by the stack or other routes (i.e., ground level release points). Section 1 of Tables 2A and 28 is omitted as Krypton-85 is neither expected nor measured .

. There were no "Batch Mode" gas_eous releases.

8. Liquid Effluents Table 3 summarizes the total quantities of radioactive liquid effluents. Table 4 presents the quantities of each of the nuclides determined to be released.

There were no batch liquid effluent releases during this report period.

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Enclosure 1 PG&E Letter HBL-16-003 TABLE 1 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES First Second Third Fourth Est. Total Units Quarter Quarter Quarter Quarter Error,%

B. Particulates (

1. Totalrelease Ci <4.58E-06 <1.26E-05 <4.67E-06 <2.24E-07 3.60E+1 I

2. Average release rate µCi/sec <5.83E-07 <1.61 E-06 <5.16E-07 <2.85E-08

3. Percent of applicable limit % <4.40E-05 <5.89E-03 <2.46E-04 <1.52E-04

4. Applicable limit µCi/cc 1.24E-10 1.24E-10 1.24E-10 1.24E-10

5. Gross alpha radioactivity Ci <3.01 E-07 <3.68E-07 <3.09E-07 <8.91E-09 Table Notes:

The < symbol used in this table means that a majority of the measurements contributing to the result were less than the Minimum Detectable Activity (MDA) for the analyses. Data for individual nuclides combines detected and non-detected results as if all values were detected. The < symbol is applied if less than 50% of the combined value is made up of detected results. When combining detected and non-detected results for different nuclides (e.g. activity totals of multiple nuclides), values with the<

symbol are ignored (i.e. treated as zero). When combining non-detected results for different nuclides (e.g. activity totals of multiple nuclides, when none were detected), all values with the < symbol are used.

If the total release for a period is determined to be a "less than" value, the limits are based on analytical results obtained in November, 2005, the mixture was determine.d to be 84% Cs-137, 11 % Co-60 and 5%

Sr-90.

The "percent of applicable limit" in Table 1 is the sum pf the values for "percent of applicable limit" for each of the release paths identified below:

First Second Third Fourth Units Quarter Quarter Quarter Quarter Stack Release Path % <4.09E-06 <4.02E-06 <3.79E-06 N/A Incidental Release Path % <3.99E-05 <5.89E-03 <2.42E-04 <1.52E-04 7

Enclosure 1 PG&E Letter HBL-16-003

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TABLE 2A GASEOUS EFFLUENTS-ELEVATED RELEASE-PARTICULATES CONTINUOUS MODE - NUCLIDES RELEASED Continuous Mode First Second Third Fourth Nudides Released. Unit Quarter Quarter Quarter Quarter

Particulates Cobalt-60 Ci NIA NIA NIA NIA Strontium-90 Ci NIA NIA NIA NIA Cesium-137 Ci NIA NIA NIA NIA Am-241 Ci NIA '\ NIA ' NIA NIA

ci NIA NIA NIA NIA Total for period Table Notes:

NIA - There were no elevated gaseous effluents during the report period.

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Enclosure 1 PG&E Letter HBL-16-003 TABLE 28 GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES NUCLIDES RELEASED Continuous Mode First Second Third Fourth Nuclides Released Unit Quarter Quarter Quarter Quarter

2. Particulates Cobalt-60 Ci <2.22E-06 <2.35E-06 <2.25E-06 <1.0?E-07 Strontium-90 Ci <2.89E-07 <3.65E-07 <2.84E-07 <1.23E-08 Cesium-137 Ci <2.06E-06 <9.90E-06 <2.11E-06 <1.04E-07 Americium-241 Ci <1.38E-08 <1.13E-08 <1.68E-08 <1.43E-10 Total for period Ci <4.58E-06 <1.26E-05 <4.67E-06 <2.24E-07 Table Notes:

The < symbol used in this table means that a majority of the measurements contributing to the result were less than the' Minimum Detectable Activity (MDA) for the analyses. Data for individual nuclides combines detected and non-detected results as if all values were detected, but the <

symbol is applied if less than 50% of the combined value is made up of detected results. When combining 'detected and non-detected results for different nuclides (e.g. activity totals of multiple miclides), values with the < symbol are ignored (i.e. treated as zero). When combining non-detected results for different nuclides (e.g. activity totals of multiple nuclides, when none were detected), all values with the < symbol are used.

Five environmental air samplers placed onsite are changed out weekly and analyzed for gross alpha and gross beta. Quarterly each station's samples are composit~d and analyzed for principal gamma emitters. The second quarter's east fence composite identified Co-60 activity of 5.24E-15

µCi/cc. This composite was re-analyzed by an offsite lab and the results verified no contamination, significant addition of personnel dose, or regulatory limits were reached. This occurrence was captured and documented in the site Corrective Action Program for tracking. See SAPN 1407196.

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\ Enclosure 1 PG&E Letter HBL-16-003 TABLE 3 LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES First Second Third Fourth Est. Total Units Quarter Quarter Quarter Quarter Error,%

A. Fission & Activation Products

1. Total release (not including tritium, gases, alpha) Ci N/A N/A N/A N/A NIA

2. Average diluted concentration µCi/ml N/A N/A N/A N/A

3. Percent of applicable limit % N/A N/A N/A N/A

4. Applicable limit µCi/ml N/A N/A N/A N/A B. Tritium

1. Total release Ci N/A N/A N/A N/A N/A I

2. Average diluted concentration µCi/ml N/A N/A N/A N/A

3. Percent of applicable limit % N/A N/A N/A N/A

4. Applicable limit µCi/ml N/A N/A N/A NIA C. Gross Alpha Radioactivity I1. Total release Ci N/A N/A N/A N/A N/A D. Volume of waste released (prior to dilution) Liters N/A N/A N/A N/A N/A IE. Volume of dilution water Liters N/A N/A I N/A N/A N/A Table Notes:

There were no batch liquid effluent releases during this report period.

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Enclosure 1 PG&E Letter HBL-16-003 TABLE 4 LIQUID EFFLUENTS - NUCLIDES 'RELEASED Batch Mode First Second Third .Fourth Nuclides Released Unit Quarter Quarter Quarter Quarter Strontium-90 Ci NIA NIA NIA NIA Cesium-137 Ci NIA NIA NIA NIA Cobalt-60 Ci NIA NIA NIA NIA Americium-241 Ci NIA NIA NIA NIA Nickel-63 Ci NIA NIA NIA NIA Tritium Ci NIA NIA NIA NIA Total for period

Ci NIA NIA NIA NIA Table Notes:

There were no batch liquid effluent releases during this report period.

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Enclosure 1 PG&E Letter HBL-16-003 Ill. SOLID RADIOACTIVE,WASTE Table 5 summarizes the disposal of sol.id r.adioactive waste during the report period.

Note: Processed Waste shipments sent for vendor processing are not considered **

disposed waste until aftet,waste is processed and shipped for disposal. At the time of this report, some Processed Waste shipments had not yet been processed and sent to disposal. Therefore, final data for Table 5 part 1.d ~'Other (Processed Waste)" and Table 5 part 3 "Solid Wasfe Disposition" will be resubmitted with next year's 2015 Radioactive Effluent Release Report.

TABLE 5 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. Solid Waste Shipped Offsite For Burial Or Disposal 12 Month . Estimated.Total Error,

1. Type of Waste Unit Period %

a. Spent resins, filter sludges, . There were no Spent resins, filter sludges, evaporator evaporator bottoms? etc. : bottoms, etc. shipments during this reporting period.

b. Dry compressible waste,

Cubic Meter .* 9.56E+03 1.00E1 contaminated equipment, etc. Ci 1.99E+01 5.60E1 c.. Irradiated components,

There were no Irradiated components, control rods, etc.

control rods, etc. shipments during this reporting period.

d. Other (Processed Waste) Cubic Meter 1.26E+02 1.00E1 Ci 1.54E+OO 5.60E1 12

Enclosure 1 PG&E Letter HBL-16-003 TABLE 5-Continued

2. Estimate of major nuclide composition (by type of Unit Nuclide 12 Month Period waste)

a. Spent resins, filter sludges, There were no spent resins, filter sludges, evaporator evaporator bottoms, etc. ' bottoms, etc. shipments during this reporting period.

b. Dry compressible waste, % H-3 1.65E-01 contaminated equipment, etc, % C-14 1.44E-01

% Fe-55 2.84E-01

% Co-60 1.21 E+01

% , Ni-59 7.39E-01

% Ni-63 8.17E+01

% Sr-90 4.43E:.01

% Nb-94 . fr.54E-03

% Tc-99 9.15E..:03

% 1-129 1.38E-02

% Cs-137 3.62E+OO

% Eu-152 7.46E-05

% Eu-154 6.97E-03 U-233 S.04E-03 I

% U-234 5.05E-03

% U-235, 5.93E-06

% U-238 9.73E-03

% Pu-238 4.02E-02

% Pu-239 2.47E-02

% Pu-240 2.-14E-02

' % Pu-241 4.75E-01

% Pu-242 8.33E-04

% Am-241 2.13E-01 0

% Cm-243 7.82E-03

% Cm-244 7.27E-03

c. Irradiated components, Th.ere were no irradiated components, control rods, etc.

control rods, etc. shipments durin'g this reporting period.

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Enclosure 1 PG&E Letter HBL-16-003 TABLE 5 - Continued

d. Other (processed waste), % H-3 4.43E+OO

% C-14 1.61 E-01

% Fe-55 8.18E-01

% Co-60 1.26E+01

% Ni-59 5.25E-01

% Ni-63 7.17E+01

% Sr-90 1.58E+OO I

% Nb-94 8.22E-04

% Tc-99 4.50E-02

% 1-129 1.88E-02

% Cs-137 5.19E+QO

% Eu-154 -* 1.91E-02

% U-233 6.74E-03

% U-234 6.79E-03

% U-235 2.21E-05

% U-238 1.13E-02

% Pu-238 1.78E-01

(

% Pu-239

1.37E-01

% Pu-240 1.36E-01

% Pu-241 2.23E+OO

% Pu-242 1.10E-03

% Am-241 6.67E-01

% Cm-242 6.?0E-06

% Cm-243 5.61E-02

% Cm-244 5.37E-02 14

Enclosure 1 PG&E Letter HBL-16-003 TABLE 5 ~Continued

3. Solid Waste Disposition Number of Mode of Shipments Transportation Destination Truck -

Hittman.TOPS 127 Energy Solutions, LLC Inc, Interstate Venture*s 833 (See Truck-US Ecology Table Notes) NCF/Savage 0 Truck - Hittman wcs

8. Irradiated Fuel Shipments

1. Irradiated Fuel Disposition Number of Mode of Shipm~nts Transportation Destination None N/A N/A Table Notes:

HBPP no longer performs batch liquid effluent discharges. The FIXS system was used to reduce liquid batch radioactivity to achieve the necessary concentration limits for liquids being transported to US Ecology for offsite disposal under the 10 CFR 20.2002 exemption. Shipments, volumes and activity totals for those liquid shipments are included in sections A.1.b, A.2.b and 3 of the table.

FILTERED ION EXCHANGE SYSTEM (FIXS) system was removed from service and dismantled during the summer of 2015 to facilitate RFB partial demolition and is no longer in service.

833 shipments (including 27 liquid shipments) were made to US Ecology under a 10 CFR 20.2002 exemption. These shipments included 5.11 E-02 Curies of Cs-137 and 1.71 E-03 Curies of Co-60 (to which the liquid shipments contributed 4.58E-05 Curies of Cs-137 and 4.06E-04 Curies of Co-60).

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Enclosure 1 PG&E Letter HBL-16-003 IV. RADIOLOGICAL IMPACT ON MAN A comparison of calculated doses from various paths has shown that the ~ffsite doses are primarily due to direct radiation. Maximum doses to individuals (for the maximally exposed organs and age groups) are summarized in Table .6. Dose~ from Noble Gases are not reported, as noble gas releases were neither expected nor. measured.

There are no airborne or liquid dose pathways from the adjacent Independent Spent Fuel Storage Installation (ISFSI), and the direct radiation measurement locations for Humboldt Bay Power Plant include the contribution from the ISFSI. Therefore, these doses comply with 40 CFR 190 as there are no other uranium fuel cycle facilities within 8 km of the Humboldt Bay Power Plant and ISFSI.

A. Doses to the average individual in the population, based on the guidance of

. Regulatory Guide 1.109, from all receiving-water-related pathways were not ca°lculated for 2015, because there were no batch liquid effluent releases during this report period. The last batch liquid effluent discharge occurred on December 11, 2013.

With no batch liquid effluent discharge, doses continue to be well below the 10 CFR 50, Appendix I numerical guidelines for limiting effluents as low as is reasonably achievable (ALARA)* (3 mrem/yr to the total body and 10 mrem/yr to any organ).

B. Total body doses to the average individual in the population from gaseous effluents to a distance of 50 miles from the site are not calculated, but this dose is less than the total body dose to an average individual present at the maximally exposed location. For an average individual at the maximally exposed location, the total body dose (determined with the same dispersion* and deposition parameters as used to calculate maximum exposure) was not explicitly calculated as there were no significant detected releases. Performing the calculation with the observed "less than" values for.releases produced a*result less than 0.02 mrem/yr.

This maximum calculated dose is well below the 10 CFR 50, Appendix I numerical ALARA guidelines (10 mrem/yr for gamma radiation and 20 mrad/yr for beta radiation from noble gases and 15 mrem/yr to any organ from tritium and radionuclides in particulate form).

C. Total body doses (to the average individual iri unrestricted areas from direct radiation from the facility) are based on TLD results of stations at the site boundary, using the shoreline occupancy factors given in Regulatory Guide 1.109 for the highest average potential individual (Teen age group). For this group, direct radiation would resultr'in an exposure of 0.022mrem/yr.-

This maximum potential dose is well below the 10 CFR 20.1302(b)(2)(ii) limit of 50 mrem/yr from external sources necessary to demonstrate compliance with the 10 CFR 20.1301 dose limit for individual members of the public.

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Enclosure 1 PG&E Letter HBL-16-003 TABLE 6 RADIATION DOSE FOR MAXIMALLY EXPOSED INDIVIDUALS Dose, milli-rem

(

First Second Third Fourth Annual Dose Source Quarter Quarter Quarter Quarter Total Liquid Effluents I Water-related Pathways (1) 11------+------+-------1------11--------4 Airborne Effluents Particulates (2)

IDirect Radiation (3) <0.01 <0.01 <0.01 <0.01 0.02 Notes 1.. Maximum total body and organ doses to individuals in unrestricted areas from receivirig-water-related exposure pathways were not calculated as there were no batch liquid effluent releases during this report period. The last batch liquid effluent discharge occurred on 12/11/2013.

2. Maximum total body and organ doses to individuals in unrestricted areas from airborne-particulate-related exposure pathways would be calculated from the average concentrations of airborne particulate 'releases detected during the report period, following the applicable portions of Regulatory Guide 1.109 and NUREG-4013. However, a majority of stack releases and incidental releases for all four quarters of 2015 were "not detected", resulting in a total activity considered "not detected", for which no dose is calculated. The plant stack was shut down and removed from service on 10/14/2015 to facilitate decommissioning.

3. Total body doses (to the mc;iximum individual in the population) are based on TLD results of stations at the site boundary, using the shoreline occupancy factors of Regulatory Guide 1.109 for the maximum potential individual (Teen age group).

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Enclosure 1 PG&E Letter HBL-16-003 V. CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM)

As decommissioning proceeds at HBPP, system changes or removal may require changes to the ODCM. During 2015 there was one ODCM revision: Revision 26.

The ODCM changes facilitated the removal of the Main Plant Stack, which was permanently shut down on 10/14/2015. Subsequently, small intermittent incidental releases will continue through portable monitored modular HEPA units. The changes to the ODCM reflect the elimination of various systems and controls on the Plant Stack effluent pathway and the end of associated sampling.

The specific changes are as follows:

ODCM Revision 26 (Effective 10/1/2015) lntroduGtion Eliminated the gaseous effluent monitor set point calculation and limited gaseous effluent to Modular HEPA Units. With the Main Plant Stack is no longer in use, SPAMS set point calculation methods are no longer relevant or included in the ODCM.

Table of Contents Deleted Gaseous Effluent Monitoring Instrumentation, the associated Tables for*

surveillance frequency, the associated Basis, the set point calculation methods, Process Control requirements for greater than Class A wastes, and repaginated to reflect other changes* throughout the document. The specifications, surveillances, and set point calculation methodology associated with the Plant Stack effluent monitoring system are no longer relevant and are being deleted from the ODCM. The deletion of the Process Control Program requirements is based on the decommissioning status at HBPP, no greater than Class A wastes are expected.

ODCM Part J Section 1.3, 1.4, 1.5 Deleted definitions for channel calibration, channel check, and channel functional test.

These definitions relate to the SPAMS no longer in use.

Section 1.14 Edited definition of ODCM to eliminated reference to Alarm Trip Set points. There are no longer any alarm or trip set points associated with effluent monitoring.

Section 1.16 Changed term burial ground to terminology appropriate for the disposal site(s).

Section 1.17, 1.20, 1.21,, 1.26 18

Enclosure 1 PG&E Letter HBL-16-003 Deleted Purge, Purging, Solidification, Source Checks, and Venting. The terms Venting, Purge or Purging do not appear elsewhere in the ODCM. HBPP does not solidify wastes.

The term Source Check was only used in surveillance for the Plant Stack instrumentation I

~

being deleted and is not used elsewhere in the ODCM. Corrected numbering (section 1.26 became 1.25).

Section 2.2.1, 2.2.2, Table 2-3 and Table 2-4 Deleted Limiting Conditions and Surveillance Requirements for effluent monitoring instrumentation along with the associated Tables detailing the surveillance requirements, action statements, and periodicity. These surveillance requirements were associated with the Plant Stack effluent monitoring instrumentation that is no longer in use or required.

Section 2.3.2 Deleted language about liquid releases to Humboldt Bay as no longer relevant.

Section 2.6.1.a, 2.6.2 and 2.6.3 Removed Tritium from 2.6.1.a.* Moved annotation regarding noble gas monitoring and l .

tritium monitoring to Bases Section 3.5. Noble gas and tritium monitoring are no longer required based on a lack of source term.

Table 2-6 I '

Eliminated Plant Stack sampling requirements from T~ble 2-6. Deleted footnotes "c" and "f'. Edited footnote "a" and "b" to provide additional guidance on expectations when LLD is not achieved and/or short term. use of Modular r,IEPA Ventilation. Plant stack sampling eliminated. Table content was rearranged to fit on three pages. Deleted footnote "c" that was a morthly sample frequency no longer applicable. Footnote "f" was related to the SPAMS system is no longer in service. Clarified action if LLD is not achieved. Clarified the sample frequency based on short term use of the Modular HEPA Ventilation .

Section 2.8.1 Deleted tritium, from limiting conditions. Tritium is no longer a source term contributing significantly to dose.

Section 2.8.2 The Plant Stack is no longer in service, removed reference to Plant Stack.

Sections 2.9.1 and 2.9.2 Revise terminology for land disposal.

Section 3.1 Edited Bases to reflect that gaseous effluent sampling and analysis is limited to Modular HEPA Units. The Plant Stack is no longer in service.

19

Enclosure 1 PG&E Letter HBL-16-003 Section 3.2 Moved explanation from surveillance section in Specification 2.3. Better fits in BASES section. GWTS processing discussion in Appendix A also referenced.

Section 3.5 Added comment that doses will include effluents from Plant stack for the operating period in 2015. Edited bases to reflect a change in source term for tritium. Stack operation and monitoring ceased operating in 2015. Spent fuel Pool is empty and no longer represents a source term of concern for tritium.

Section 3.6 Changed: "Stack. to "Gaseous effluent" monitoring. Stack no longer in service.

Section 3.8 Edited bases to reflect a change in tritium source term. Spent fuel Pool is empty and no longer represents a source term of concern for tritium.

Section 3.9 Clarified that the function of spec~ication 2.9 is to meet a free standing liquid criteria.

Clarified working to reflect the actual objective of the spe.cification.

Section 4.1 Editorial correction to 4.1.h-"no" to "not". Reformatted Table 4.1. No change to content Section 4.2.a Clarified wording regarding reporting liquid waste generation. Liquids shipped to a

.regulated landfill are reported annually~

Section 4.2.d Added explanation regarding inoperability of the Plant Stack would no longer be reflected in the Annual Radioactive Effluent Relea~e Report following 2015. The surveillance and reportability still applies for the 2015 reporting year while the Plant' Stack was in operation.

After removal from the ODCM the reportability requirement no longer applies.

ODCM Part II Section 1.0 & 1.2

Deleted "and Stack Monitor Set point". With the Plant Stack not an effluent pathway the effluent monitor set point is no longer calculated.

20

Enclosure 1 PG&E Letter HBL-16-003 Section 1.2.2 Changed default X/Q to 6.59E-3 sec/m3 which corresponds to incidental release path atmosphere dispersion factor and deleted reference to Calculation N-238 as no longer relevant. The old X/q was based on the 50ft stack.with continuous flow and was an average annual X/Q. The revised X/Q is for incidental release paths and based on R.G.

1.145 guidance.

Section 1.2.3 Deleted section text but preserved numeric format to avoid cross-reference errors. There is no inline monitoring of sample point for the Modular HEPA Units, so this section is not relevant because there is no line loss in sample lines (transmission*fraction).

Section 1.2.4 Deleted references to stack, added Modular HEPA Unit and recalculated using the new XIQ and 2000 cfm HEPA flow rate. These changes follow the same calculation methodology previously used with the new default parameters: X/Q=6.59E-3 and Modular HEPA Unit flow rate =2000 cfm.

Section 1.2.5 through Section 1.2.8 Continuation of the previous calculation that previously existed using the revised X/Q and flow rate. Deleted the transmission fraction correction related to sample line loss on SPAMS.

Section 1.2.9 Same calculation that previously existed using the revised HEPA Unit flow rate. Removed language regarding 10% _NUE and SPAMS since it is not relevant to a SPAMS ~et point with SPAMS no longer in *use.

Section 4.3.9, 4.3.10, 4.3.11, 4.3.12, & 4.3.13 Added note that the tritium source term is no longer evaluated once the spent fuel pool is empty. Once the spent fuel pool is empty there is no longer a credible source term for tritium *exposure in effluents and is no longer evaluated.

Section 5.1.c & 5.2:b Deleted references to Stack and tritium as a source term.

Section 5.3 Deleted reference to Stack.

Section 6.0

~

Deleted Section 6.0. HBPP no longer expects wastes exceeding a specific activity that is 21

Enclosure 1 PG.&E Letter HBL-16-003 unacceptable to disposal site without solidification of exceeding Class A as defined in 10CFR61.

Section 7.0 Deleted Section 7.0. HBPP no longer expects wastes exceeding a specific activity that is unacceptable to disposal site without solidification of exceeding Class A as defined in 10CFR61. HBPP no longer anticipates disposal of wastes requiring stabilization in a HIC.

Section 8.0 Added controls for use of vendor services, onsite gross dewatering, and waste acceptance criteria of disposal sites(s). New controls provide additional processing options and clarification of waste streams encountered during decommissioning.

Section 10.0 Editorial correction in text: "to" changed to "the". Deleted CTS-352 (Section I, Table 2-4).

The commitment tracking system no longer identifies CTS-352 in the CTS database. It is likely that it affected SPAMS surveillance periodicity in Table 2-4, corrected in an earlier revision. The stack monitoring system (SPAMS) is no longer operable.

Appendix A Removed references to liquid waste processing capability and Limiting Conditions 2.3.1.

The spent fuel pool is empty and the processing capability for liquids is minimal. Retained sufficient information* to convey that an effluent pathway exists for disposal of liquids not acceptable for release thought GWTS and GWTS is available to handle ground water intrusion.

Added "typ-ically" to statement regarding use of Ground Water Treatment System. The change. accurately reflects the GWTS is the normal pathway unless some contaminant prevents its use and requires altern*ative handling.

VI. CHANGES TO THE PROCESS CO_NTROL PRO_GRAM (PCP)

During the report period, the ODCM section describing the Process Control Program was revised. Based on the status of decommissioning, there is no anticipated need for disposal of waste exceeding Class A as defined by 10 CFR 61, to the corresponding sections of the ODCM (for waste solidification or high integrity containers) were deleted. The ODCM section for waste not exceeding Class A was revised to cover any appropriate processing (e.g.-, dewatering bed-type resin, dewatering of wet waste, etc.).

VII. CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS HBPP no longer performs bafoh liquid effluent discharges. The FILTERED ION EXCHANGE SYSTEM (FIXS) system was used to reduce liquid batch radioactivity to 22 .

Enclosure 1 PG&E Letter HBL--16-003 achieve the necessary concentration limits for liquids being transported to US Ecology for offsite disposal under the 10 CFR 20.2002 exemption. Shipments, volumes and activity totals for those liquid shipments are included in sections A.1.b, A.2.b and 3 of Table 5. The FIXS system was removed from service and dismantled ,during the summer of 2015 to facilitate RFB partial demolition and is no longer in service.

VIII. INOPERABLE EFFLUENT MONITORING INSTRUMENTATION Liquid Effluent Monitoring Effective 12/23/2013 HBPP no longer uses outfall canal dilution for liquid effluents.

There were no batch liquid effluent releases during this report period.

Airborne Effluent Monitoring Instrumentation Stack Particulate Airborne Monitoring System (SPAMS)

There were no unplanned SPAMS out of service periods d1,1ring the operable report period. SPAMS was removed from service on 10/14/2015 to facilitate demolition of the Main Plant Exhaust Fan, Plant Stack, *and for partial demolition of the Refueling

Building.

The Main Plant Exhaust Fan tripped at.1614 on 12/27/2014 due to loss of station power. The decision was made to leave the exhaust fan out of service, and the Air

. Handling Unit was secured 12/29/2014. SPAMS was left in operation, with restart of plant ventilation to be 1/4/2015 after the holiday break. The effluent sample filter was collected as usual this date and on 1/6/2015. SAPN logged to annotate both 2014 and 2015 efflL,Jent reports accordingly, and to adjust the net discharge (if activity is detected) to the actual operating hours of the ventilation system. No activity dete~ted.

See SAPN 1398659.

Main Plant Exhaust Fan and SPAMS reported as powered off for the last time at 1010 on 10/14/2015. See Task detail 0006 of SAPN 1401945. Modular HEPA air sample units are used to monitor air effluents that are not monitored by SPAMS.

A bearing failed on one of the two weekly air samplers running in the LRW building in the week prior to 2/17/2015 causing an abnormal totalizer volume as compared to the adjacent sampler which was running simultaneo'usly. The sample will not be counted due to questionable volume through the sample. See SAPN 1401080.

A power glitch in the Liquid Radwaste (LRW) caused a Modular HEPA to trip offline on 3/6/2015. The H-2000 effluent monitor continued to run but did not retain runtime information in memory. A second LRW H-2000 effluent monitor did retain the information and the adjusted runtime will be used for analysis of both air samplers.

See SAPN 1402023. .

A scheduled power outage took out the LRW Building on 6-15-2015. One of the two effluent monitors did not retain the totalizer volume. See SAPN 1406114.

23

Enclosure 1 PG&E Letter HBL-16-003 09/08/2015 Stack Gas Flow instrument was out of service for Service Air outage. The Stack Fan and the Stack sampler remained in operation during this period. See SAPN 1401945 Task detail 0005.

On 10/3/2015 a Modular HEPA was operated for only 65 mins and did not gather sufficient volume to reach site gross alpha and Am-241 concentration limits.

Concentrations of the air filter was identified as <4.19E-14 µCi/cc, compared to the site gross alpha concentration limit of 2.0E-14 µCi/cc. The Am-241 concentration of the air filter was determined to be <3.34E-13 µCi/cc, compared to* the 10CFR20 appendix B concentration limit of 2.0E-14 µCi/cc. Probable cause is most likely due to the short runtime. See SAPN 1412350.

IX. ERRATA 2014 Annual Radioactive Effluent Release Report Errata:

At the time of submitting the 2014 report, a waste processor had not completed processing the waste to be sent for disposal. The following update to Table 5 now includes the corrected processed waste volume and' isotopic percentages for the year.

TABLE 5 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A Solid Waste

.. Shipped Offsite For Burial Or Disposal 12 Month Estimated Total Error,

1. Type of Waste Unit Period %

d. Other (Processed Waste) Cubic Meter 4.96E+01 1.00E+01 Ci 1.60E+OO 5.60E+01 24

Enclosure 1 PG&E Letter HBL-16-003 TABLE 5 - Continued

2. Estimate of major nuclide composition (by type of waste) Unit Nuclide 12 Month Period

d. Other (processed waste), % H-3 5.18E-01

% C-14 9.29E-03

% Fe-55 2.64E-01 '

% Co-60 1.53E+OO

% Ni-59 5.16E-03

% Ni-63 3,22E+01

% Sr-90 5.87E+OO

% Tc-99 2.49E-02

% 1-129 4.78E-03

% Cs-137 5.80E+01

% U-233 1.30E-03

% U-234 9.28E-04

)

% U-238 3.40E-03

% Pu-238 8.97E-02

% Pu-239 7.37E-02

% Pu-240 7.34E-02

% Pu-241 1.03E+OO

% Pu-242 3.04E-03

% Am-241 2.41 E-01

% Cm-243 2.42E-02

% Cm-244 2.42E-02

3. Solid Waste Disposition Number of Mode of Shipments Transportation Destination

' 4 Truck Perm a Fix 25

Enclosure 1 PG&E Letter HBL-16-003 2014 ERRATA - Continued Addition to inoperable effluent monitoring instrumentation for 2014 effluent report.

VIII. INOPERABLE EFFLUENT MONITORING INSTRUMENTATION.

Airborne Effluent Monitoring Instrumentation The Main Plant Exhaust Fan tripped at 1614 on 12/27/2014 due to loss of station power. The decision was'made to leave the exhaust fan out of service and the Air Handling Unit was secured 12/29/2014. SPAMS was left in operation with restart of plant ventilation planned for 1/4/2015 after the holiday break. The effluent sample filter will be collected as usual this date and on 1/6/2015.

Annotate both 2014 and 2015 effluent reports accordingly. Adjust the net.

discharge (if activity is detected) to the actual operating hours of the ventilation system. No activity detected. See SAPN 1398659 & 1401945.

\ ,*

26

Enclosure 2 PG&E Letter HBL-16-003 HUMBOLDT BAY POWER PLANT UNIT 3 SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL REVISION 26 INCLUDING CHANGES MADE DURING 2015

SECTION ODCM Nuclear Power Generation VOLUME 4 Humboldt Bay REVISION '26 EFFEC DATE 10-20-15 Power Plant PAGE i APPROVED BY TITLE ORIGINAL SIGNED 10-20-15 SAFSTOR/DECOMMISSIONING OFFSITE DOSE DIRECTOR/PLANT MANAGER I. DATE CALCULATION MANUAL HB NUCLEAR (Procedure Classification - Quality Related)

INTRODUCTION The SAFSTORIDECOMMISSIONING.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 monitbring. 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-.01, Implementation ofProgrammatic Controls for Radiological Effluent Technical Specifications, in the Administrative Controls. Sedion of the Technical Specifications and the Relocation ofProcedural Details of RETS to the Ojfsite Dose Calculation Manual or to the Process Control Program, issued by the NRC in January, 1989.

Implementation of the LAR revised the instantaneous liquid concentration limits based on "old" 10 CFR 20 maximum permissible concentrations (MP Cs) to 10 times the "new" 10 CFR 20, Appendix B, Table 2, Column 2 effluent concentration limits (ECLs) and replaced the gaseous effluent

. instantaneous concentration limits at the site boundary with annual dose rate limits equating to the doses associated with the annual average concentrations of '.'old" 10 CFR 20, Appendix B, Table II, Column 1. The LAR also established limits for doses to members of the public from radiological effluents based on the as low as reasonably achievable (ALARA) design objectives of 'IO 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 ofNUREG-0133, Preparation ofRadiological Effluent Technical Specifications for Nuclear Power Plants, and NUREG-04 73. 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-1166, HBP P Final Environmental Statement. r

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 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 w~ich established baseline conditions for soil, biota and sediments. The REMP also includes requirements to participate in an interlaboratory comparison program. As of December 31, 2013, HBPP ceased liquid radioactive effluent discharges via the discharge canal to Humboldt Bay. The scope of the REMP and interlaboratory comparison program are the dosimeters and air samples required to evaluate the .direct radiation and gaseous effluents from HBPP ., -

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 10/rsreleased from HBPP during the decommissioning period. Part II of the ODCM contains the .

calculational methods for gaseous and liquid effl~ents to preserve site specific data although the gaseous effluent pathway is limited to Modular HEP A Units on a selected basis and the liquid discharge pathway has been terminated.

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 tre_atm.ent 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 ofreviews 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 ca~culations. ,

Changes shall be submitted to the NRC in therforni 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 tothe 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 26 DOSE CALCULATION MANUAL PAGE iii TABLE OF CONTENTS PART I~ SPECIFICATIONS Section Title Page

1.0 DEFINITIONS I-1 2.0 SPECIFICATIONS I-7 1p 2.1 2.2*

Deleted Deleted I.:.7 I-8 2.3 Liquid Effluent - Concentration I-9 2.4 Deleted I-9 2.5 Deleted I-9 2.6 Gaseous Effluents - Dose Rate I-10 2.7 Deleted I-13 2.8 Gaseous Effluents: Dose - Radionuclides in Particulate Form I-14 2.9 Solid Radioactive Waste I-15 2.10 Total Dose I-16 2.11 REMP Monitoring Program I-17

' 2.12 REMP Interlaboratory Comparison Program I I-26 2.13 Radioactive Waste Inventory I-27 3.0 *SPECIFICATION BASES I-28 1p 3.1 Deleted

)

I-28 3.2 Deleted I-28 3.3 Deleted I-28 3.4 Deleted I-28

.3.5 Gaseous Effluents Dose Rate Basis I-28 3.6 Deleted I-29 3.7 Deleted I-29 3.8 Gaseous Effluents: Tritium and Radionuclides in Particulate Form Dose Basis I-29 3.9 Solid Radioactive Waste Basis I-30 3.10 Total Dose Basis I-30 3.11 . REMP Monitoring Program Basis I-30

-3.12 REMP Interlaboratory Comparison Program Basis I-31 3.13 Radioactive Waste Inventory Basis I-31 I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE iv PART I- SPECIFICATIONS- (Continued)

Section Title . Page 4.0 ADMINISTRATIVE CONTROLS I-31 4.1 Annual Radiological Environmental Monitoring Report I-31 4.2 Annual Radioactive Effluent Release Report *I-35 4:3 Special Reports I-37 A.4 Major Changes to .Radioactive Waste Treatment Systems I-37 4.5 Process Control Program Changes , I-38 PART II- CALCULATIONAL METHODS AND PARAMETERS

. Section Title Page' 10F:1.0 UNRESTRICTED AREA EFFLUENT CONCENTRATIONS II-1 lOr 1.1 1.2 Liquid Effluent Unrestricted Area Concentrations Unrestricted Area Gaseous Effluent Concentrations II-1 II-2 2.0 LIQUID EFFLUENT DOSE CALCULATIONS II-9 2.1 Deleted II-9 2.2 Deleted II-9

2.3 Deleted II-9 2.4 Liquid Effluent Dose Calculation Methodology II-9 3.0 LIQUID EFFLUENT.TREATMENT II-14 3.1 Treatment Requirements - Deleted II-14 3.2 Treatment Capabilities - Deleted II-14 4.0 GASEOUS EFFLUENT DOS~ CALCULATIONS I II-15 4.1 Dose Rate II-15 4.2 Deleted II-15 4.3 Dose - Tritium and Radionuclides in Particulate Form II-15

NUCLEAR POWER GENERATION DEPARTMENT SECTION* ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE v PART II - CALCULATIONAL METHODS AND PARAMETERS - (Continued)

Section Title Page 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE II-33 5.1 Whole Body Dose II-33 5.2 Skin Dose II-33 5.3 Dose to Other Organs II-34 5.4 Dose to the Thyroid II-34 .

10/15 6.0 Deleted Il-35 I

10/15

.0 Deleted II-35 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED II-35 RESINS AND OTHER.WET WASTES 9.0 PROGRAM CHANGES II-37 10.0 COMMITMENTS II-37 11.0 RESPONSIBLE ORGANIZATION II-37

\

12.0 REFERENCES

, II-37 App.A SAFSTOR BASELINE CONDITIONS A-1 App.B BASES FOR ATMOSPHERIC DISPERSION AND DEPOSITION VALUES B-1 App.C Deleted C-1 I

. \,

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM

voLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE vi LIST OF TABLES - PART I I

1-1 Frequency Notation I-6 2-1 Deleted I-7 lOrlS 2-2 Deleted I-7 2-3

Deleted I-8

. 2-4 Deleted J-8 2-5 Deleted I-9 2-6 Radioactive Gaseous Waste Sampling and Analysis Program I-11 2-7 HBPP Radiological Environmental Monitoring Program I-18 2-8 Deleted I-19 2-9 Detection Capabilities for Environmental Sample Analysis Lower Limits Of I-19 Detection (LLD) *

2-10 Distances and Directions To Environmental Monitoring Stations

I-21 4-1 Radiological Environmental Monitoring Report Annual Summary - Example I-33 LIST OF TABLES - PART II 2-1 Ingestion Dose Factors for Adult Age Group II-12 2-2 Ingestion Dose Factors for Teen Age Group II-12 2-3 Ingestion Dose Factors for Child Age Group II-13 2-4 Bidaccumulation Factors for Saltwater Environment II-13 2-5 Average Individual Foods Consumption for Various Age Groups II-14

_) 2-6 Maximum Individual Foods Consumption for Various Age Groups II-14 4-1 ' Inhalation Dose Factors for Adult Age Group II-27 4-2 Ii:ihalation Dose Factors for Teen Age Group II-27 4-3 Inhalation Dose Factors for Child Age Group II-28 4-4 Inhalation Dose Factors for Infant Age Group II-28 4-5 External Dose Factors for Standing on Contaminated Ground II-29 4-6 Average Individual Foods Consumption for Various Age Groups II-29 4-7 Maxiillum Individual Foods Consumption for Various Age Groups II-29 4-8 Ingestion Dose Factors for Adult Age Group II-30 4-9 Ingestion Dose Factors for Teen Age Group II-30 4-10 Ingestion Dose Factors for Child Age Group* II-31 4-11 Ingestion Dose Factors for Infant Age Group II-31 4-12 Stable Element Transfer Data For Cow-Milk Path II-32 4-13 Stable Element Transfer Data For Cow-Meat Path II-32

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE vii

~,

LIST OF FIGURES - PART I Figure Title Page 1-1 Site Boundary 1-6 2-1 HBPP Onsite TLD Locations 1-22 2-2 Deleted 1-22 2-3 HBPP Offsite Sampling Locations ~ Hllinboldt Hill . 1-23 2-4 HBPP Offsite Sampling Locations - Eureka 1-24 2-5 HBPP Offsite Sampling Locations-Fortuna 1-25 I

.. \

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-1 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. .

10/15 1.3 Deleted 1.4 Deleted 1.5 Deleted 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 Deleted

. 1.8 FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance

Requirements shall correspond to the intervals defined in Table 1-1.

1.9 Deleted 1.10 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL

PAGE 1-2 1.11 INSTANTANEOUS CONCENTRATION INSTANTANEOUS CONCENTRATION is the concentration averaged over one hour of radioactive materials in effluents.

1.12 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.

l.13MODULARHEPA VENTILATIONUNIT MODULAR HEPA VENTILATION UNIT consists of HEPA filter trains dischargedto the environment and ~ampled in accordaµce with ANSI/RPS Nl3.l-1999.

1.14 OFFSITE DOSE.CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANUAL contains the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, 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 iriformation that should be included in the Annual Radiological Environmental Monitoring Report and the Allnual Radioactive Effluent Release Report. The ODCM also contains the Process Control Program (PCP) for solid radioactive wastes.

l.150PERABLE-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).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-3 1.16PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM-(PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be *accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, disposal site(s) requirements, and other requirements governing the disposal of solid radioactive waste.

1.17 Deleted 1.18 RESTRICTED AREA The RESTRICTED AREA is defined by 10CFR20.1003. The physical location(s) of the RESTRICTED AREA shall be defined in plant procedures.

l.19SITE 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 SITEBOUNDARY are controlled by the Company.

10/15 1.20 Deleted 1.21 Deleted 1.22 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. I 1.23 URANIUM FUEL CYCLE As defined in 40 CFR Part 190.02(b), "URAl':JIUM 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 operat~ons, and the reuse of recovered non-uranium special nuclear and by-product materials from the cycle."

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM 1

VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE . REVISION 26 DOSE CALCULATION MANUAL PAGE 1-4

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1.24 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce radioactive material. in particulate form in effluents by passing ventilation or vent eXhaust gases through charcoal absorbers and/or HEPA filters for the purpo.se of removing particulates from the gaseous exhaust stream prior to release to the environment.

'F 1.25 Deleted

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NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-5 Table 1-1 FREQUENCY NOTATION 1

Notation \

Frequency Extension Period D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . None w At least once per 7 days. 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months M At least once per 31 _days. 7 days Q At least once per 92 days. 22 days SA At least once per 184 days. 45.days A At least once per 365 days. 91 days p Completed prior to each release.

N.A. Not applicable.

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1 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 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-6 Figure 1-1 SITE BOUNDARY

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-7 2.0 SPECIFICATIONS 2.1 Deleted; Table 2 Deleted; Table 2.2 - Deleted

NUCLEAR POWERGENERATION DEPARTMENT SECTION

ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE* 1-8

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2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION 1 LIMITING CONDITIONS 10/15 I 2.2.1 Deleted - plant stack is no longer in operation.

SURVEILLANCE REQUIREMENTS 10/15 2:'22 Deleted I

Table 2 Deleted Table 2 Deleted

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NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-9 2.3 LIQUID EFFLUENT - CONCENTRATION LIMITING CONDITIONS 2.3.l 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 rele~sed beyond the SITE BOUNDARY to within the above limits.

SURVEILLANCE REQUIREMENTS-

!Or Deleted (See BASES Section 3.2 and Appendix A) r Table 2-5 (Deleted) 2.4 LIQUID EFFLUENT - DOSE Deleted - No longer applicable 2.5 Deleted- No longer applicable

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NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-10 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:

10/15 I a. 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:

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

to within the above limit(s).

SURVEILLANCE REQUIREMENTS 10/15 2.6.2

Deleted (see BASES section 3.5) 2.6.3 Deleted (see BASES section 3.5)

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

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NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-11 Table 2-6 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit Sampling Analysis

Type of Activity of Detection Gaseous Release Type Frequency Frequency Analysis (LLD)

[!0;1fil ,., (µCi/ml)a I Modular. HEP A Ventilation Discharge Contiriuousb,ct wb Principal Gamma 1 x 10-11 Mixing Box Emitterse I")

Particulate Sample

. Continuous 0 ' 0 wo I Gross Alpha 1x10-Ll Mixing Box Particulate Sample Continuousb,a wo Gross Beta,. 6.7 x 10-J:l '

Mixing Box Particulate Sample Continuousb,a Q Sr-90g 1 x 10-11 Composite of Mixing Box Particulate Samples Continuousb,a,n Q Am-241 1x10-Ll Composite of Mixing Box 1

Particulate

f Samples ...

Continuous 0

'a, 1

Q Am-241 1 x 10-14 Composite of Mixing Box I Particulate Samples

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NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-12 Table 2-6 (Continued)

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.

F'or a_ particular measurement systemi(which may include radiochemical separation):

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),

Eis the countiIJ.g 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),

/.... is the radioactive decay constant for the particular radionuclide, and

.Mis the elapsed time between midpoint of s_ample collection and time of counting (for plant effluents, not environmental samples).

Typical values ofE, V, Y, and .M shall be used in the calculation.

The LLD is defined as an£! priori (before the fact) limit representing the capability of a 10/1s measurement system and not as an£! posteriori (after the fact) limit for a particular measurement. NOTE: The LLDs are achievable with a reasonable count time assuming adequate effluent volume and sample volume. If the LLD is not achieved, initiate a condition report to document that the LLD was not achieved and indicate a probable cause (short runtime, equipment malfunction, etc.). RP Supervision will determine if additional calculations should be performed per Surveillance 2.6.4.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-13 Table 2-6 (Continued)

Table Notation (Continued) 10/15 b

Samples shall be changed at least once per 7 days (3 day extension permitted), assuming effluent pathway is in continuous use (typically> 40 hrs per week). Sarriples may be collected more frequently for short duration use of a. Modular HEPA Ventilation Unit.

c Deleted d

The ratio of the saniple 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.

e 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.

Deleted based on SPAMS no longer in service.

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

h When release volume is less than or equal to 3.26 X 10 11 ml (e.g., l.15Et7 cubic feet).

i When release volume exceeds 3 .26 X 10 11 ml (e.g., l .15E+7 cubic feet).

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2.7 Deleted

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NUCLEAR POWER GENERATION. DEPARTMENT *SECTION ODCM VOLUME .4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-14 10/i51 r 2.8 GASEOUS EFFLUENTS: DOSE - RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITIONS 1ofl 2.8.1 The dose to a MEMBER OF THE PUBLIC from the release of 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 nirem 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 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 radioactive materials in particulate form with half-lives. greater thad 8 days in gaseous effluents during the remainder of the current calendar quarter and during the reinainder 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 COMPARIS ON 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.

OR Perform a dose assessment, if weekly sampling indicates the effluent from modular HEPA units exceed 0. f uCi of alpha emitters or Sr-90. The assessment of alpha and beta may be performed with appropriate compensation for naturally occurring nuclide.s.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL , PAGE 1-15

\ As explaip.ed in SpecificationBases 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 disposal site(s) requi,rements.

J' APPLICABILITY: At all times.

ACTION:

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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, spentresins) meet the shipping, disposal site(s) requirements with regard to dewate~ing and off site vendor processes. \ ' '

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NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-16 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

'mr~m).

APPLICABILITY: At all times.

ACTlON:

With the calculated doses from the release of radioactive materials in gaseous effluents exceeding twice the limits of Specification 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. b.e 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 ifthe 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 gaseous effluents shall be calculated in accordance with dose calculation methodology provided for Specification 2.8.1.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-17 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 arid 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. A Special Report pursuant to Administrative Control 4.3, 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 Specification 2.8. Prepare and submit to the Commission within 30 days of obtaining analytical results :from the affected sampling period which includes an evaluation of release conditions, environmental factors or other aspects which caused the dose limits to be exceeded. 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, 2-3, 2-4 and 2-5 and shall be analyzed pursuant to the requirements of Tables 2-7 and 2-9.

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT

.VOLUME 4 REVISION 26 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-18 Table 2-7 HBPP RADIOLOGICAL ENVIRONMENTAL MONITORINq PROGRAM PROGRAM DESCRIPTION PROGRAM BASIS Exposure Pathway Number of Samples

Sampling and Collection Frequency Type of Analysis ODCM Specs (QR) and/or Samole and Locations<*>

AIRBORNE 5 onsite locations, 1 offsite Contiriuous sampler operation with Gross alpha and gross beta radioactivity x location sample collection at least once per 7 following filter change days< 1l Gamma isotopic(c) analysis on quarterly composite (by station)

DffiECT RADIATION(bJ Minimum of 8 onsite stations, at TLDs exchanged quarterly(!) Gamma exposure<3l x or within the SITE BOUNDARY fence line~ with TLDs 1 offsite control station with TLD TLDs exchanged quarterly(!) Gamma exposure<3l x 4 offsite stations with TLDs TLDs exchanged quai1er!v< 1l Gamma exoosure<3l x WATERBORNE None NIA NIA INGESTION None NIA NIA TERRESTRIAL None NIA NIA*

Table Notations QR - Quality Related <1lPerformed by HBPP <3lperformed by a NVLAP accredited processor (a)

Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, se!!Sonal 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 Iv):onitoring 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 locatfon(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 (c) Gamina isotopic analysis means the identification and quantification of gamma emitting radionuclides that may be attributable to the effluents from the facility.

(d)

Sampling may be suspended when no effluent discharge has occurred in the previous 10 days. This provision limits sampling inactive discharge points.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4-TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1~19 Table 2-8 (Deleted)

Table2-9 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANAL YSIS(a) (b)

LOWER LIMITS OF DETECTION (LLD)(c)

Airborne Particulate Analysis (pCi/m3) _

Gross* Beta -0.01 H-3 Co-60 Cs-137 0.06 Table Notations (a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Monitoring Program Report.

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

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

concluding that a blank observation represents a "real" signal.

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

LLD = 4.66Sb Ex V x 2.22 x Y x exp(-A-t)

Where:

LLD = the "a priori" lower limit of detection as defined above (as pCi per unit mass or volume) the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (a.s counts per minute)

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE I-20 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)

A, ;:::: the radioactive decay constant for the particular radionuclide

.M = 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.

Analyses will be performed in such a manner that the stated LLDs 'Yill 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 ofE, V, Y and t should be used in the calculation. It should be recognized that the LLD is defined as a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-21 Table 2-10 DISTANCES AND DIRECTIONS TO ENVIRONMENTAL MONITORING STATIONS Radial Direction Radial Distance Station By from Plant No. Code Station Name Sector Degrees (Miles) 1 A King Salmon Picnic Area w 270 0.3 2 A 180 Dinsmore Drive, Fortuna SSE 158 9.4 3 D Humboldt Hill Road at Bret Harte Lane SSE 158 0.9 14 A South Bay School Parking Lot s 180 0.4 17 A Control Set at Humboldt Substation, Eureka NEE 61 5.8 25 A Irving Drive, Humboldt Hill SSE 175 1.3

A ISFSI Fence line * *

At least 4 additional TLDs are used, one in each direction, at the ISFSI Fence line, when fuel is in storage Table Notations Code: A

Dosimetry Station D Air Particulate Station r

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE I-22 Figure 2-1 HBPP Onsite TLD Locations Monitoring locations T7, TIO, Tl I, Tl3, Tl6, T2, T3, and T5 generally represent REMP Site Boundary direct exposure monitoring locations in the 8 primary compass points beginning with T-7 to representing no.rth and moving clockwise. Monitoring locations T19, T18, T21 and T20 are at the perimeter of the ISFSI.

. Figure 2 Deleted

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-23 Figure 2-3 HBPP OFFSITE SAMPLING LOCATIONS - HUMBOLDT HILL

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GPS Coordinates (NAD83/NAVD88 CA. Zone 1) Decimal Degrees Station Eastinq Northinq el. Latitude Lonqitude 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 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-24 Figure 2-4 HBPP OFFSITE SAMPLING LOCATIONS - EUREKA Jn<l;;li*

6.1.lfd Eureka ?:.:-rw.:\*i Myf-let. '

Cutten C 0 ,I ::. I C.~k';

i!cb"1 fWi. , !lllo<I fortun.J

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GPS Coordinates (NAD83/NAVD88 CA. Zone 1) Decimal Deqrees Station EastinQ I NorthinQ I el. Latitude I Lonaitude 17 5976549.55 I 2175490.19 I 164.85 40.78276 I -124.11324

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-25 Figure 2-5 HBPP OFFSITE SAMPLING LOCATIONS - FORTUNA

=Eureka !:Jrui'f" fy'yrdet0'1.~

ifkfi.l"°';'.I S.ayview Cu tten Pi n e ~il is c " l l

I N dccOr li **: .. *: 1 Wilfow*oc' I

<mi,,..l>L.- 1 I

i GPS Coordinates (NAD83/NAVD88 CA. Zone 1) Decimal Deqrees Station Easting I Northinq I el. Latitude I Longitude 2 5962583.86 I 2105797.82 I 35.53 40.59057 I -124.15746

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4.

TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-26 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 Progr~.

APPLICABILITY: At all titnes.

ACTION:

With analyses not being perf9rmed as required above, report the corrective actions taken to prevent a recurrence to the Commission In the Annual Radiological Environmental Monitoring Program Report.

SURVEILLANCE REQUIREMENT.S 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. *I 1

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-27 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 Radiologi'cal Environmental Monitoring Program Report.

2.13 .2 Deleted SURVEILLANCE REQUIREMENTS 2.13 .3 An inventory of the estimated liquid radioactive waste in outdoor tanks inventory shall be maintained to verify the 0.25 Ci limit is not exceeded.

OR Provide overflow protection.

OR Use process knowledge of typical concentration and tank volume to verify that the 0.25 Ci is not exceeded.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-28 3.0 SPECIFICATION BASES 3.1 Radioactive Gaseous Effluent Monitoring Instrumentation Basis Deleted-The plant stack ceased operation in 2015. Monitoring gaseous ef'fluent is limited

to sampling and analysis of Modular HEPA Units. _,

3.2 Liquid Effluent Concentration Basis 10/15 Deleted - Liquid effluents are no longer discharged to Humboldt Bay. Effective December 31, 2013, discharge of processed radioactive liquid effluents to Humboldt Bay was terminated. Any remaining or incidental radioactive liquid in concentrations exceeding 10 times 10 CFR 20, Appendix B, Table 2 Column 2 are manifested for disposal at a licensed disposal site. Sampling and manifesting requirements are consistent with the requirements of the receiving facility not subject to ODCM methodology.

3.3 Liquid Effluent Dose Basis Deleted - Liquid effluents are no longer discharged to Humboldt Bay.

3 .4 Liquid Effluent Treatment Basis Deleted - Liquid effluents are no longer discharged to Humboldt Bay.

3 S 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 obj~ctives of Appendix I to 10 CFR 50. The annual dose rate limits are the doses associated with the annual average concentrations of"old" 10 CFR 20, Appertdix B, Table II, Column 1. The specification provides operational flexibility for releasing gaseous effluents*to satisfy the Section II.A and Il.C design objectives of Appendix I to 10 CFR 50.

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 1 or 2) will be sufficiently low to compensate for the reduced atmospheric dispersion of gaseous effluents relative to that for the SITE BOUNDARY. The specified '

release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem/year to the total body or to less than or equal to 3000 mrem/year to the skin. This specification does not affect the requirement to comply with the annual limitations of 10 CFR 20.130l(a).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-29 10/1sl Stack operation and monitoring ceased operation in 2015, so the reporting period for 2015 includes the dose contribution from the plant stack prior to ceasing operation. Modular HEP A Ventilation Units continue to be sampled as a gaseous effluent pathway.

Noble gas monitoring is not required because the spent fuel (noble gas source term) has been transferred to the ISFSI. Tritium monitoring is not required in gaseous effluents because the tritium source term was the spent fuel pool water which is now empty.

Residual water in various plant drains and sumps contain low levels of tritium (generally at or below the drinking water standard (2E-5 uCi/ml or 20,000 pCi/L) and does not require monitoring as ~ gaseous plant effluent.

3.6 Deleted Gaseous effluent moriitoring is not required for noble gases because the spent fuel (nobie gas source term) has been transferred to the ISFSI.

3.7 Deleted 3.8 Gaseous Effluents: Tritium and Radionuclides in Particulate Form Dose Basis This specification is provid~d to implement the requirements of Sections Il.C, 111.,A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section Il.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 Ill.A of Appendix I that conformance with the guides of Appendix I be* shown by calculation;:tl procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

The basis for the dose calculation threshold of 0.1 uCi alpha emission or Sr-90 in a week assumes a continuous ground level release of l.65E-13 uCi/sec and an X/Q of 6.59E-3 sec/m3

The limiting inhalation dose is to a teen age member of the public at the site boundary at approximately 0.3 mrem/wk (15 mrem/yr) to the bone from alpha emitters.

Compliance with this Specification has been established on a licensing basis by the SAFSTOR Environmental Report and NUREG-1166, "Final Environmental Statement for .

Decommissioning Humboldt Bay Power Plant." These reports have demonstrated that routine release of Tritium and radioactive materials in 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-30 The previously evaluated tritium source term was the spent fuel pool water, which is now empty. Residual water in various plant drains and sumps contain low levels of tritium (at or below the drinking water standard (2E-5 uCi/ml or 20,000 pCi/L) and does not require monitoring as a gaseous plant effluent.

3.9 Solid Radioactive Waste Basis This Specification ensures that radioactive wastes that are transported from the site shall meet the disposal site(s) licensee and/or waste acckptance criteria for free standing liquids 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.lOTotal 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 190.11and10 CFR Part 20.2203a4, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 2.3, 2.4, 2.6, 2.7 and 2.8. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engag~d 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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-31 Environmental Report baseline environmental conditions by conducting onsite and offsite environmental monitoring to evaluate routine conditions dutjng decommissioning and to document any increased nuclide concentrations and/or radiatfon levels resulting from ac_cidents during decommissioning.

'Fhe 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. HBPP no longer includes water, milk, fish, food products, or sediment in its routine REMP sampling program. Sampling and analysis in

suppo.rt of the License Termination Plan is independent of the ODCM requirements.

3.12 REMP Interlaboratory Comparison Program Basis The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the meas~rements 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.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 1 of each year.

The Annual Radiological Environmental Monitoring Report shall include:

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE 'REVISION 26 DOSE CALCULATION MANUAL PAGE 1-32

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 amiual report.

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. _

10F1 h. Deleted - water s&mples are not collected as a part of the REMP.

'_)

, I L A discussion of all analyses in which the LLD required by Table 2-9 was not achievable. '

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 PAGE 1-33 DOSE CALCULATION MANUAL Table 4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

- EXAMPLE N~me of Facility Humboldt Bay Power Plant Unit 3 Docket No. 50-133, OL-DPR-7 Location of Facility Humboldt County, California Reporting Period January 1 - December 31, 1997 (County, State)

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

[Unit of Measurement] Analyses of Detection" (Fraction) Distance and (Fraction) & [Range] b Reported Performed (LLD) & rRangel b Direction & rRangelh Measurements AIRBORNE -

Particulates Not Required NIA NIA NIA NIA Not Required NIA DIRECT RADIATION.

[mR/quarter] Direct radiation 3 13.6 +/- 0.1 Station T7 15.4 +/- 0.2 12.7 +/- 0.3 0 (64) (64164) .. (414) (414)

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

WATERBORNE NIA

---~-l!~~f!~~ -~~!~!- -- ------------ ---~C!! _~~gl!!t:.e..<! _-- ------------------ NIA NIA - NIA NIA

__ :~~!-~e..q!li!~~---- -------------------

NIA ------------------

NIA --------------------- --------------------

, NIA NIA

---~<!! B-~gi:!t:.e..<! _-- ------------------ --- ~~!-~e..q!li!~~---- -------------------

Groundwater NIA

. NIA ------------------

NIA ---------------------

NIA --------------------

__ P!!?!<:!l!.~!V_a!~-----~--- _ -~~t_l3:~q_u_!r_e~ __ NIA __ !'J"_o! ;!l_eg~~*~<! __ NIA NIA

---~<!! B-~gl!!t:.e..<! _-- ------------------

- - - - -NIA - - - - -------------

NIA -------------

NIA_ ------------

NIA Sediment

--------------------- --------------------- ---~~!-~e..q!li!~~- --- -------------------

Algae Not Required NIA NIA NIA NIA - Not Required NIA INGESTION Milk ~ -~~t_l3:~q_u_!r_e~ __ NIA . NIA NIA NIA - - !'J"_O! ;!l~g~i_!~<!_ - - NIA Fish and invertebrates Not Required ------------

NIA ------------

NIA -----NIA.- --"" - -------------

/ NIA Not Required ------------

NIA TERRESTRIAL Soil Not Required NIA NIA NIA NIA Not Required NIA

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 REVISION 26 TITLE SAFSTOR/DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-34 TABLE 4-1 (Continued)

)

RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

a The LLD is defined as the smallest concentration ofradioactive 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 observatio:o. 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 fL posteriori (after the fact) limit for a particular measurement. (Current literatirre 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.

h 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].

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

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.

NIA- 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 26 DOSE CALCULATION MANUAL PAGE 1-35 4.2 Annual Radioactive Effluent Release Report This report shall be submitted prior to April 1 of each year. The following information shall be included: "

a. A summary of the quantities of radioactive liquid and gaseous,effluents and solid waste releas~d from the plant as outlined in Regulatory Guide '1.21, Measuring, Evaluatip,g, and Reporting Radioactivity in Solid Wastes and Releases ofRadioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear 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 10115 outlined in the ODCM and in conformance with 10CFR 50.36a and 10CFR Part 50, Appendix I, Section N.B.I. Beginning in the reporting year 2014, liquid effluents shipped for processfo.g or disposal at a regulated disposal site are included in the annual report.

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

1. ContainerVplume

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

3. Principal Radionuclides (specified as measured h

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 CJ.Ction for 10/15 any radioactive gaseous monitoring instrumentation inoperable for greater tha.Ii 30 days in accordance With Specification 2.2. Beginning the reporting year 2015, following cessation of the plant stack operation, the effluent monitoring instrumentation associated with Specification 2.2 ceased operation. Inoperability and lack of timely corrective action is only applicable to the period of plant stack operation. Anomalies associated with monitoring effluent from Modular HEP A Ventilation systems will-be reported.

e. A summary des~ription of changes made to:

1. Process Control Program (PCP)

2. Radioactive Waste Treatment Systems

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-36 I

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.

I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITL,E SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 .

DOSE CALCULATION MANUAL PAGE 1-37 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 2.8 and.2.10. (

_}

b. Radiological Environmental Monitoring - Specification 2.11.

)

4.4 Major Changes to Radioactive Waste Treatment Systems **.

a. Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid) shall be reported to the NRC in the Annual Radioactive Effluent Release Report for the period in which the evaluation was reviewed. The changes shall be reviewed

and concurred with by the Plant Staff Review Committee and approved by the Plant Manager.

b. The following information *shall be available for review:

1. A summary of the evaluation that led to the determination that the change could be

made in accordance with 10 CFR 50.5Q,

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 syst~ms,

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 Envitpnniental 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 iii 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 tP.at the change was reviewed and approved in accordance with plant procedures.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORJDECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 1-38 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 plarit 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 S.ECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-1 PART II - CALCULATIONAL METHODS AND PARAMETERS 10~-

11 1.0 UNRESTRICTED AREA EFFLUENT CONCENTRATIONS 1.1 LIQUID EFFLUENT UNRESTRICTED AREA CONCENTRATIONS Specification 2.3.l requires that the Radioactive Liquid Effluent Sample concentrations (RLES) are calculated 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.

L c. canal .

=;1 I, (1-1) i lOxECLi where:

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

ECLr = Effluent Concentration Limit for radionuclide " i " from ~ 0 CFR 20, Appendix B, Table 2, Column, 2 (µCi/ml) 1.1.1 If the outfall location is not at the :furthermost portion of the canal from the .

entrance to the Bay_the concentration of the isotope Ci-Canal is equalto the

. concentration being discharged at the outfall.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-2

~o~

1.2 UNRESTRICTED AREA GASEOUS EFFLUENT CONCENTRATIONS 1.2.1 Equation C-4 of Regulatory Guide 1.109 demonstrates how to calculate* dose from inhalation:

The annual dose associated with inhalation ofall radionuclides, to organ j ofan individual in age group a, is then:

Dja(r,9) = *Ra LXi(r,S)DFAija where Dja is the annual dose rate to organ j ofan individual in age group

./

a Ra is the breathing rate for age group a

\

xi(r,e) is the annual average ground-level concentration of nuclide i in air in sector e at distance r; in pCi!m3

DFAija is the dose factor for nuclide .i to organ j ofage group a To calculate Xi(r,9) the annual average ground-level concentration of nuclide i in air in sector eat distance r, in pCi/m3 the equation must be rearranged to:

Dja(r,9)/( DFAijaRa) = Xi(r,9)

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 b~me of a Teen from Am-241is1.77 mrem/pCi The DFAija are taken from: NRC NUREG/CR-4013, "LADTAP-11 Technical Reference and User Guide" The Teen breathing rate is 8000 m 3/year

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-3 Therefore the ground-level concentration of Am-241 in air in sector 8 at distance r, in pCi/m3 that will produc*e a dose rate of 1500 mrem/year to. the bone of a Teen is:

(1500 mrem/year) I (1.77 mrem/pCi) / (8000 m/year) = l.06E-1 pCi/ m 3 3 l.06E-lpCi/ m3 =

(l.06E-1 pCi/m3) I (1E6 pCi /µCi) I (1E6 ml/m3) = l.06E-13 µCi/ml 1 l .2.2 Quantity of radioactive material released Equation C-3 of Regulatory Guide L 109 demonstrates how to calculate the quap.tity of material that must, be .released to produce a given airborne concentration:

The annual average airborne concentration ofradionuclide i at the location (r, fJ) wit~ respect to the release point may be determined as xi(r,8) = 3 .17 x 104 Qi(x/Q)0 (r,e) where Xi(r,8) is the annual average ground-level concentration ofnuclide i in air in sector 8 at distance r, in pCi!m3 4

3.17 x 10 is the* number ofpCi/Ci divided by the number ofsec/yr (x/Q)0 (r,8) is the annual average atmosphere dispersion factor, in sec/m3.

Qi is the release rate of nuclide I to the atmosphere, i!J Ci/yr 10/15 A value of 6.59E-3 sec/m3 was used for the incidental release path atmosphere dispersion factor at the site boundary (x/Q)0 (r,8) for releases from Modular HEP A Units. This is based on a release rate *of2000 cfm. (Ref: Safstor ODCM, Appendix B, 2.0) This factor is based on the atmospheric models of Regulatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants.

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

Qi= xi(r,8) I (3.17 x 104(x/Q)1?(r,8))

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-4 Therefore the Modular HEPA Unit release rate ofAm-241 required to equal the inCidental ground-level concentration at the'. site boundai:y calculated above is:

3 l.06E-1 pCi/m I ((3.17E4 (pCi/Ci)/{sec/yr)) * (6.59E-3 sec/ill3)) =

5.07E-4 Ci/yr or 5.07E2 uCi/yr 1.2.3 Transmission Fraction 10/15

.I Deleted - no on line monitoring provided.

10/15 1.2.4. Effluent Concentration ,. I The Modular HEPA Unit concentration that would result in a release rate of 5.07E-4 Ci/yr is equal' to:

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

Th~ average annual Modular HEPA Unit flow rate is 2,000.cfm *'

This results iri.a total vdlum~ of 2.98El3 cc/yr This is b*ased on (2000 ft3/min

525,600 minutes/yr

28,3 I 7* c~/ft3 ) .

(5.07E-4 ~i >t< 1E6 µCi/Ci) I (2.98El3 cc/yr)= l.70E-11 µCi/cc I Therefore an indicated Modular HEPA concentration of l.70E-'l l µCi/cc at 2000 cfm for one calendar year would resuit in adose of 1500 mrem to a member of the public at the site boundary. * *

Two times the indicated release rate is equal to3.4EM11 µCi/cc.

Two hundred times the indicated.release rate is equal to 3.4E:-9 µCi/cc ..

1.2.5 Relationship to EPA P AG To compare the release rates calculated above the following assumptions were m~: ;

Am-241 dose conversion factor in rem I cm-3 µci hr, from EPA 400 = 5.3E8 10/15 '

Sirice no credit is taken for an elevated release point or an annual average x/Q

the same atmospheric dispersion factor is used_ in the calculations below.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE . SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-5 Assuming that an unplanned release occurs at two times the ODCM release rate for one hour the total activity released is equal to: *'

10/15 3 .4E-11 µCi/cc

2000 ft3/min

28,317 cc/ft3

60 min = l.16E-1 µCi (l.16E-1 µCi)* (5.3E8 rem I cm-3 uCi hr)* (6.59E-3 sec/m3) I (1E6 cm3/m3) I (3600 sec/hour)= l.13E-4 rem This is much less than the EPA PAG of 1 Rem Assuming that an unplann~d release occurs at two hundred times the ODCM. release rate for 15 minutes the total activity released is equal to: **

3.4E-9 µCi/cc* 2000 ft3/min

28,317 cc/ft3

15 min= 2.89EO µCi This results in a dose of:

(2.89EO µCi)* (5.3E8 ~em./ om-3 uCi hr)* (6.59E-3 sec/m3) I (1E6.cm3/m3) I (3600 sec/hour)=* * "

2.80E-3rem This is much less than the EPA PAG of 1 Rem.

(

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-6 1.2.6 Relationship to 10CFR20 Appendix B Table 2 Effluent Concentration limits The 10CFR20 Appendix B Table 2 Effluent Concentration limit for Am-241is2E-14 µCi/ml. ,

The average annual ground-level concentration in air (xi) in pCi/m3 is equal to:

l xi=* (3.l 7E4 (pCi/Ci)/ (sec/year))* Q * (J:/Q)

Where Q is equal to the quantity of radioactive material released in a ye~ in Curies/year 10/15 ODCM-Modular HEPA Unit incidental releaseJ:/Q = 6.59E-3 sec/ m 3

_If xi= 2~-14 µCi/ml then:

Q,,;, (2E-14 µCi/ml* 1E6 ml/m3

1E6 pCi/µCi) I ((3.17E4 (pCi/Ci)/

(sec/yr)*(6.59E-3 sec/ m 3)) .

Q = 9.57E-5 Ci/yr The average .annual Modular HEPA U:riit volume based on the ODCM is 2.98E13 cc/yr.

This is based on (2000 cfm

525,600 minutes/yr

28,317 cc/cfm).

Therefore, the Modular HEPA Unit effluent concentration required to result in a fence . .line concentration of 2E-14 µCi/ml is:

(9.57E-5 Ci/yr* .1E6 µCi/Ci) I (2.9S:?13 cc/yr* 1 cc/ml)= 3.2E-12 µCi/ml 1.2. 7 Conversion Factor from Effluent Concentration to µCi/day 10/15 The release rate in µCi/day= Modular HEPA Unit concentration in µCi/cc >I< 2000 ft3/min

1440 minutes/day

28317 cc/ ft3

The release rate in µCi/day= Modular HEPA Unit concentration in µCi/cc

8.16E10 cc/day 1.2.8 Conversion Factor from µCi/day to % of NUB An NUB is equal to a release rate of.3000 mrem/year

%NUB = (Offsite dose rate I NUB threshold)

100

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE / 11-7

%NUE = ((Conversion Factor

Release Rate) I NUE threshold)

100

%NUE =((Conversion Factor* 100) I NUE threshold)* Release Rate 10/15 The Conversion Factor is equal to (l.77E6 ~em/µCi) * (6.59E-3 sec/ m 3) * (8000 m3/year) I (8.64E4 sec/day)

This is equal tol.08E3 mrem/year per µCi/day 1.2.9 Results 10/15 The 10CFR20 Appendix B '.fable 2 Effluent Concentration limit for Am-241is2E-14 µCi/ml. The Modular HEPA Unit effluent concentration that would result in a fence-line concentration of2E-14 µCi/ml is 3.2E-12 µCi/ml.

3.2E-12 uCi/ml

8.16E10 cc/day* lml/cc

l.08E3 mrem-day/uCi-yr = 4.70E2 mrem/yr.

470 mrem/yr I 8760 hr/yr= 5.365E-2*mrem/hr 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 l.07E-4 rem (0.107 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 2.675E-3 rem (2.7 mrem) which is much less than the EPA.PAG.

Note that Am-241 is used in the example calculations and is expected to be limiting.

Other alpha emitting isotopes such as Pu-238, Pu-239/240 and Cm-243/244 are evident in the contamination at HBPP. Since the Effluent Concentration Limits (ECLs), Derived Air Concentration (DAC)values and organ Dose Conversion Factors (DCFs) are similar, the Am-241 values may be assumed to be gross alpha with appropriate compensation for naturaily occurring isotopes.

Other radionuclides (Co-60, Sr-90, Cs-137, etc.) are important in determining actual offsite dose and in demonstrating compliance with the ECL using the sum of the fractions rule. The example calculations are used similarly for each isotope in the mix with their respective ECL, DCF and exposure pathway (inhalation, ingestion, and submersion).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-8

~ofill.

Although not relevant to the hypothetical offsite dose calculation in the ECL and NUE analysis above, assumed effluent concentrations are approximately 1DAC,2 DAC, and 200 DAC for Am-241 at the point ofrelease. Airborne radioactivity .

control measures to control worker dose, also limits the potential off(ite dos~.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIDECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-9 2.0 . LIQUID EFFLUENT DOSE CALCULATIONS 2.1 MONTH (31 DAY PERIOD) Deleted 2.2 CALENDAR QUARTER - Deleted 2.3 CALENDAR YEAR - Deleted 2.4 LIQUID EFFLUEN~ DOSE CALCULATION METHODOLOGY

_l

As of December 3 i, 2013, HBPP has ceased liquid radioactive effluent discharges via the discharge canal to Humboldt Bay. Any.remaining processed liquid radioactive waste is transported offsite for larid disposal at ail authorized disposal facility. The foliowing ,

calculation methodology is preserved as a part of the ODCM for ease of reference to site specific parameters in the event of an accidental release of liquid radioactive effluent. No

recurring liquid effluent dose calculations are expected for the remainder of decommissioning; The equations specified in this section for calculating the doses due to the actual release rates of radioactive m'aterials 'in liquid effluents are consistent with the methodology provided in.

Regulatory Guide 1.109, "Calculation of Annual Doses fo Man from Routine Releases of Reactor Effluent~ for the Purpose of Evaluating Compliance with 10 CFR Part 50; Appendix

. I," Revision 1, October 1977 arid Regulatory Guide Ll 13, "Estimating Aquatic.Dispersion of Effluents from.Accidental and Routine Reactor Releases for the Purpose oflmplementing 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 analysi~,

depth-averaged, finite element hydrodynamic model (reference J 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.

Since the intake canal contains a larger volume of water, use of the above dilution factors for effluent releases to the intake canal provides a simplified, conservative methodology for calculating annual dose from effluent releases to ihe intake canal.

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:

NUCL.EAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORJJ)ECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-10 n

D =~]Ci - BaydilutedX DFx {(BFish,i X UFish)+ (Brnv,i X Urnv)}] (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. If the outfall to the canal is at the furthest most portion of the canal from the entrance to the Bay, this will be estimated by calculating the total a~tivity released (e.g. effluent cpncentration Ci effiue~t in pCi/L times the discharge volume Vo in Liters) then'd~viding the total activity of the nuclide discharged during the period, pico,. Curies, by the dilution*volume (e.g. total discharged volume Vo plus total tidal flow V~o during the period in liters), and by the Bay dilution factor of 20. The total annual tidal flow for the outfall canal is 2.47E+9 Lite.rs/year (e.g.,

6.77E+6 Liters/day). If Gross Alpha radioactivity is determined to be in the effluent, Pu-241 will be considered to be present at3.25 times the amount of detected Gross Alpha radioactivity. Note that the resulting dose commitment is the .annual dose rate (mrem per year) for a time frame with this average concentration. Doses (NOT dose rates) for periods shorter than a year must be proportionately reduced.

Ci - EffluentX VD Ci - Baydiluted = - : - - - - - - (2-2)

(VD +Vm}x20 If the outfall. is not located in the furthest most portion of the canal from the entrance to the Bay, no credit for tidal dilution. of the canal wili be .

ti:iken and the diluted Bay concentration will be calculated u~ing the following equation.

Ci - Effluent

. Ci - Bay diluted=-:--- (2-3) 20 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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOl\'JMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-11 Brnv,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) offish, kilogram/year, for the age group and individual (average or maximum) in question. This factor is derived from Table 2-5 or 2-6.

Urnv = 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 can be performed using the above methodology for the current month, quarter, or year.

t'

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-12

\

Table 2-1 ,,

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

. Selected Nuclides from NUREG/CR-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 L3o x io-4 9.01x10-6 4.36 x 10-6 No Data NoData

1.88 x 10-6 Sr-90 8.7i x 10-3 No Data 1.75 x 10-4 No Data NoData

2.19 x 10-4 Cs-137 7.97 x 10-5 l.09 x 10-4 7.14 x 10-5 3.70 x 10-5 1.23 x 10-5' 2.11x10-6 Y-90 9.62x10"'"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 Am-241 7.55x10-4 7.05 x 10-4 5.41 x 10-5 4.07 x 10A

No Data 7.42 x 10-5 Gross a 7.55*x io-4 7.05 ~ 10-4 5.41 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/CR-4013 (LADTAP II input values)

Organ Nuclide Bone Liver Total Body _Kidney Lung GI-LLI H-3 No Data 6.04 x 10-8 . 6.04 x 10-8 6.04 x 10-8 6.04 x 10-8 6.04 x 10-8 Co-60 No Data 2.81x10-6 6.33 x 10-6 No Data No Data 3.66 x 10-5 Ni-63 1.77 x 10-4 1.25 x 10-5 6.00 x 10-6 No Data No Data l.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 1.12 x 10-4 1.49 x 10-4 5*_19 x 10-5 5.07 x 10-5 1.97 x 10-5 2.12 x 10-6.

Y-90 1.37 x 10-8 No Data 3.69x10-10 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 4.31x10-4 No Data 7.87 x 10-5 Gross a 7.98 x 10-4 7.53 x 10-4 5.75 x 10-5 . 4.31x10:-4 .No Data 7.87 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-13.

Table 2-3

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

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

Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.16 ~ 10-7 1.16 x 10-7 1.16 x 10-7 1.16 x 10-7 1.16 x 10-7 Co-60 No D~ta 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 4.11x10-8 No Data l.lOx 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 a l.3frx 10-3 1.17 x 10-3 1.02 x 10-4 6.23 x 10-4 No Data 7.64 x 10-5 Table 2-4 Bioaccumulation Factors for Saltwater Environment

. (pCifkg per pCi/liter)

Selected Nuclides from Regulatory Guide 1.109, Table A-1 and from NUREG/CR-4013 Element Fish Invertebrate H 9.0 x 10-l 9.3 x 10-l 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 101 2.5 x 101 y 2.5 x 101 1.0 x 103 Pu 3.0 2.0 x 102 Am 2.5 x*lQl 1.0 x 103 Gross a 2.5 x 101 1.0 x 103

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-14 Table 2-5 Average Individual Foods Consumption for Various Age Gi-oups (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 *o 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 *o 0 0 330 0 3.0* LIQUID EFFLUENT TREATMENT 3.1 TREATMENT REQUIREMENTS 3.1.1 Deleted 3.1.2 Deleted 3.2 Deleted

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NUCLEAR POWER GENERATION DEPARTMENT SECTI0N *ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL -*

PAGE 11-15 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 paiiiculates dose calculation methodology. Refer to sections 4.3.3 through 4.3.8 for t~e appropriate equations.

As explained in Specification Bases 3.5, 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 .DOS.E-TRITIUM ANDRADIONUCLIDES IN PARTICULATE FORM 4.3.l 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, mrem/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 CPR Part 50, Appendix I," Revisio:d 1, October 1977 and Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-16 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 dep~ndent on the existing radionuclide pathways to inan, 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 mari, 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 subs~9.uent exposure of m:an.

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 fili .

individual age group as follows:

n D= ~]QiX (Rrnh,i +RGP,i + RMeat,i+ RMilk,i + Rveg,i)]. (4-3) i=l when~:

D =

Annual dose comniitment, mrem/year.

Qi The average release rate of the nuclide in question, pico-Curies/second.

Rrnh, i = The dose factor for the inhalation pathway for the radionuclide, i, iri. 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.

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-17 RVeg,i = The dose factor for the pathway of deposition on vegetation for the radionuclide, i, in units 9f mrem/year per pico-Curie/sec.

In general, the calculations for these pathw~ys give results that represent trivial radiation exposure.* The values calculated for typical anticipated Decommissioning releases range from about 0.002 mrem/year (fruit/vegetable consumption pathway) to less than 1 x 10-6 mrem/year (for direct radiation exposure from material deposited on the ground) ..

4.3.4 Pru:ticulate Inhalation Pathway Dose Calculation Methodology Rnm,i = (x/Q) x BRa x DFi,a (4-3a) where:

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

= l.0 x 10,-5 seconds/cubic meter for releases from the 50 foot stack.

. Refer to Appendix B, 1.2. . .

= 6.59 x J 0-3 seconds per cubic meter for releases other than from .

the 50 foot stack. Refer to Appendix B, 2.1.

BR a The breathing rate ofthereceptor age group (a), cubic meters per year. The values to .be used are 1400, 3 700, 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, inrem/pico-Curie, for the receptor age group, a, for the radionuclide, i. The dose factors are given in Tables 4-1, 4-2, 4-3, and 4-4. * ,

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL C, PAGE Il-18 4.3.5 Particulate Ground Plane Pathway Dose.C::alculation Methodology R.GP,i = (D/Q) x SF x DE 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

mrem/hr per pCi/m2 from.Table 4-5. No values are provided for Transuranic radionuclides, as their dose contribution to this pathway is negligible.

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

. Guide 1.109 suggests values of0.7 fortJ;ie maximum individual.

D/Q = The atmospheric deposi~ion factor, with units of inverse square

.meters.

3.0 x 10-s inverse square meters for releases from the 50 foot stack. Refer to Appendix B, 1.3.

5d9 x 10-6 inverse square meters for releases other than from th~

50 foot stack. Refer to Appendix B, 2.2.

w Weathering factc:>r. This is the reciprocal of the weathering time constant given irlRegulatory 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 TITL.E SAFSTOR/DECOMMISSIONING OFFSITE REVISION* 26 DOSE CALCULATION MANUAL PAGE 11-19 4.3.6 Particulate Grass-Cow-Milk Pathway Dose Calculation Methodology RMiik.i ~ (D/Q) x ( QF x Ua x F; x DFi,a x W) (4-3c) where:

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

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

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

)

DFi,a = The ingestion dose factor. for radionuclide, i, for the 'receptor in

age group (a), in units of mreiplpico-Curie, from Tables 4-8, 4-9, 4-10, or 4-11.

Fm The fraction of the cow's intake of a nuclide which appears in a is*

liter of milk, with units of days/liter. This parameter 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 ~eleases from the 50 foot stack. Refer Appendix B, 1.3. .

= 3 .29 x 1o- iriverse square meters for r~leases other than from the 6

O 50 foot stack. Refer to Appendix B, 2.2.

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 thi~ equation, *W has the value of 1. 74 x 106 seconds.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORJDECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-20 4.3.7 Particulate Grass-Cow-Meat Pathway Dose Calculation Methodology RMeat,i = (D/Q) x ( QF x Ua x F~x DFi,a x W) (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 (Juide 1.109, Table E-15.,

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

Ff = The fraction of the animal's int~e 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 '1 o-8 inverse square meters for releases froin the 50 foot stack. Refer to Appendix B, 1.3.

6 3 .29 x 10- inverse square meters f9r releases other than from the 50 foot stack. Refer to Appendix B, 2.2.

w = Weathering factor. This is the reciprocal of the. weathering time constant given in Regulatory Guide l.109, for a 14 day removal half-life. In this .equation, W has the value of 1.74 x 106 seconds.

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NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORJDECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-21 4.3.8 Particulate Vegetation Pathway Dose Calculation Methodology (4-3e) where:

UT The total consumption rate of fruits and vegetables, kilograrp/year. This parameter is determined with the default values from Regulatory Guide 1.109, as reproduced in Tables 4-6 and4-7.

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, 1.3.

= 3 .29 x 10-6 inverse square meters for releases other than from the 50 foot stack. Refer to Appendix B, 2.2.

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 eql;lation, 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/m2 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 8:11d fruit grown in other areas (import~d to Humboldt county).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-22 4.3.9 Tritium Organ Dose Calculation Methodology The annual dose commitment may be calculated for any organ of an individual age group as follows: -

D = Qm X {Rwi,m + RGP,~. + RMeat,H3 ~ RMilk,ID + Rveg,m) * (4-4) where:

D Annual dose* commitment, mrem/year.

"* QH3

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

Rlnh,H3 = The dose factor for the inhalation pathway for H-3, mrem/year per pico-Curie/sec. *

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

RMilk,H3 = The dose factor for the grass-cow-milk pathway for H~3,

mrem/year per pico-Curie/sec. .

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

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

as-sumption, the calculated maximum ~ffsite exposure is 0.0013 mrem/year: One~

the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-23 4.3.lOTritium Inhalation Pathway Dose Calculation Methodology

Rmh,m = (Yo) x BRa x DFm,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, 1.2.

6.59 x 10~3 seconds per cubic meter for releases other than from

. the 50 foot stack. Refer to Appendix B, 2.1.

BR a = 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 ofmrem/pico-Curie by Tables 4-1, 4-2, 4-3, and 4-4.

Once the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ,

TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE Il-24 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: **

D. ..

..NV!Jlk, H3

= (x/1 IQ)' x (0.75 H

X 0.5) x QF x Ua 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, mrein/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.

I 0.75 = The fraction of total feed that is water.

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

H = Absolute humidity of the atmosphere, 0.008 kilograms/cubic meter, 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, 1.2.

= 6.59 x 10-3 seconds per cubic meter for releases other than from ..

the 50 foot stack. Refer to Appendix B, 2~ 1.

Once the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE Il-25 4.3.12 Tritium Grass-Cow-Meat Pathway Dose Calculation Methodology (4-4 c)

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

= The cow's vegetation consumption rate; SO kg/day per Regulatory Ouide 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 ofH-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 t4e 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, 1.2. . .

= 6.59 x 10*3 seconds per cubic II1eter for releases other than from the 50 foot stack. Refer to Appendix B, 2.1.

10/15 Once the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL _PAGE 11-26 4.3.13 Tritium Vegetation Pathway I)ose Calculation Methodology The concentration of tritium is based on the airborne concentration rather than the deposition:

Rveg,H3 = (%) X( 0.

75

~ 05

) X UT X DFa (4-4d) where:

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

H = Absolute humidity of the atmosphere, 0.008 gm/m3 per Regulatory Guide I . I 09.

0.75 == The fraction of total feed that is water.

0.5 = The ratio of specific activity ofH-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, froin Tables 4-8 through 4-11.

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

I .O x I 0-5 seconds/cubic met~r for releases from the 50 foot stack.

Refer to Appendix B, I .2.

6.59 x I 0-3 seconds per cubic meter for releases other than from the 50 foot stack. Refer to Appendix B, 2.1.

10/15 Once the spent fuel pool is emptied, this source term and exposure pathway is no longer evaluated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM.

VOLUME "4 TITLE *SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-27' Table 4-1 Inhalation Dose Factors for Adult Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-7 and from ;NUREG/CR-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 l.58 x 10-7 Co-60 'No Data L44 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 2.61x10-7 No Data 7.01x10-9 No Data 2.12 x 10-5 6.32 x 10-5

-Pu-241 3.42 x 10-2 8.69 x 10-3 1.29 x 10-3 5.93 x 10-3 1.52 x 10-4 8.65 x 10-7 Gross a 1.68 1.13 7.75 x 10-2 5.04 x 10-l 1.82 x 10-l 4.84 x 10-5 Table 4-2 Inhalation Dose Factors for Teen Age Group *

(1nrem/pico-Curie inhaled) . "

Selected Nuclides from Regulatory Guide 1.109, Table E-8 and from NUREG/CR-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 Co-60 No Data 1.89 x 10-6 2.48 x 10-6 No Data 1.09 x 10-3 3.24 x 10-5 Sr-90 1.35 x 10-2 No Data 8.35 x 10-4 No Data 2.06x10-J 9.56 x 10-5 Cs-137 8.38 x 10-5 1.06 x 10-4 3.89 x 10-5 3.80 x 10-5 1.51x10-5 1.06 x 10~6

r. 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-?

Pu-241 3.74 x 10-2 9.56 x 10-3 1.40 x 10-3 6.47 x 10-3 2.60 x 10-4 9.17 x 10-7 Gross a 1.77 1.20 8.05 x 10-2 5.32 x 10-l 3.12 ';( 10-l 5.13 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-28 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/CR-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 3.04 x 10-7 3.04x 10-7 Co-60 No Data 3.55 x 10-6 6.12 x 10~6 No Data 1.91x10-3 2.60 x 10-5 Sr-90 2.73 x 10-2 No Data 1.74 x 10*-3 No Data 3.99 x 10-3 9.28 x 10-5 Cs-137 2.45 x 10-4 2.23 x 10-4 3.47 x 10-5 7.63 x 10-5 2.81x10-5 9.78x 10-7 Y-90 1.11x10-6 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 l.75x 10-2 2.93 x 10-3 1.10 x 10-2 5.06 x 10-4 8.90 x 10-7 Gross a 2.97 1.84 1.28 x 10-l 7.63 x 10-l 6.08 x 10-l 4.98 X: 10-5 Table4-4 Inhalation Dose Factors for Infant Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-10 and from NUREG/CR-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-T 4.62 x 10-7 4.62 x 10-7

'Co-60 No Data 5.73 x 10-6 8.41 x 10-6 No Data 3.22 x 10-3 2.28 x 10-5 Sr-90 2.92 x 10-2 No Data 1.85 x 10-3 No Data 8.03 x 10-3 9.36 x10-5 Cs-137 3.92 x 10-4 4.37 x 10-4 3.25 x 10-? 1.23 x 10-4 5.09 x 10-5 9.53 x 10-7 Y-90 2.35 x ro-6 No Data 6.30 x 10-8 No Data 1.92 x 10-4 7.43 x 10-5 Pu-241 8.43 x 10-2 1.85 x 10-2 3.11 x 10-3 1.15 x 10-2 7.62 x 10-4 8.97 x 10-7 Gross a 3.15 1.95 1.34 x 10-l 7.94 x 10-l 9.03 x 10-l 5.02 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL . PAGE 11-29 Table 4-5 External Dose Factors for Standing on Contaminated Ground (mrem/hour per pico-Curie/square meter)

Selected Nuclides from Regulatory Guide 1.109, Table E-6 Total Nuclide Skin Body H-3 0 0 Co-60 2.00 x 10-8 1.70 x 10-8 Sr-90 2.60 x 10-12 2.20 x 10-12 Cs-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 26 DOSE CALCULATION MANUAL PAGE 11-30 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/CR-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 1.05x10-7 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 7.58x10-3 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 2.11x10-6 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 x10-5 7.45 x 10-7 '3.32 x 10-7 1.53 x 10-6 .No Data 1.40 x 10-6 Gross a . 7.55x10'"4 7.05 x 10-4 5.41 x 10-5 4~07 x 10-4 No Data . 7.81x10-5 Table 4-9 Ingestion Dose Factors for Teen Age Group (nirem/pico-Curie fugested)

Selected Nuclides from Regulatory Guide 1.109, Table E-12 and from NUREG/CR-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI

. H-3 No Data l.06x io-7 1.06x10-7 1.06 x 10-7 1.06 x 10-7 . 1.06x10-7 Co-60 No Data 2.81x10-6 6.33 x 10-:6 No Data No Data 3.66 x 10-5 Sr-90 8.30 x 10-3 No Data 2.05 x 10-3 No Data No Data 2.33 x 10-4 Cs-137 1.12 x 10-4 1.49 x 10-4 5.19 x 10.;5 5.07 x 10-5 1.97 x 10-5 2.12 x 10-6 Y-90 1.37 x 10-8 No Data 3.69 x 10-10 No Data No Data

1.13 x 10-4 Pu-241 1.75 x 10-5 8.40 x 10-7 .3.69 x 10-7 1.71 x 10-6 No Data 1.48 x 10-6 Gross a 7.98 x 10-4 7.53 x 10-4 5.75 x 10-5 4.31x10-4 No Data 8.28x10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-31 Table 4-10 Ingestion Dose Factors for Child Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-i3 and from NUREG/CR-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 2.03x10-7 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 4.31x10-3 No Data No Data 2.29 x 10-4 Cs-137 3.27x10-4 3.13 x 10-4 4.62 x 10-5 1.02 x 10-4 3.67x 10-5 1.96 x 10-6 Y-90 4.11 x 10-8 No Data 1.10 x 10-9 No Data No Data 1.17 x 10-4 Pu-241 3.87 x 10-5 1.58 x 10-6 8.04 x 10-7 2.96 x 10-6 No Data 1.44 x 10-6 Gross a 1.36 x 10-3 l.17x 10-3 1.02 x 10-4 6.23 x 10-4 No Data 8.03 x 10-5 Table 4-11 Ingestion Dose Factors for Infant Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E-14 and from NUREG/CR-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 3.08 x 10-7 3.08 x 10-7 3.08x 10-7 3.08 x 10-7 3.08 x 10-7 C9-60 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 4.71x10-3 No Data No Data 2.31x10-4 Cs-137 5.22 x 10-4 6.11x10-4 4.33 x 10-5 1.64 x 10-4 6.64 x 10-5 1.91x10-6 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 8.82x10-7 3.17 x 10-6 No Da,ta 1.45 x 10-6 Gross a 1.46 x 10-3 1.27 x 10-3 1.09 x 10-4 . 6.55 x 10-4 No Data 8.10 x 10-5

)

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-32 Table 4-12 Stable Element Transfer Data For Cow-Milk Pathway

. (days/liter)

Selected Nuclides from Regulatory Guide 1.109, Table E-1 and from NUREG/CR-4013 I

Element Fm H 1.0 x 10-2 Co 1,0 x 10-3 Sr 8.0 x 10-4 Cs 1.2 x 10-2 y 1.0 x 10-5

pu 5.0 x 10-6 Gross a 5.0 x 10-6 Table 4-13 Stable Element Transfer Data For <:;ow-Meat Pathway (days/kilo-gram)

Selected Nuclides from Regulatory Guide 1.109, Table E-1 and from NUREG/CR-4013 Elem~nt Ff H 1.2 x 10-2 Co 1.3 x 10-2 Sr 6.0 x 10-4 Cs . 4.0 x 10-3 y 4.6 x 10-3 Pu 2.0x10-4 I j

Gross a 2.0 x 10-4

~

NU'cLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORJDECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE il-33 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 slimming the calculated doses from the, following:

a. Deleted 10/15

b. Modular HEPA Ventilation Particulate releases, using equation (4-3).

c. 'Deleted.. Tritium is no longer a gaseous effluent source form.

d. Liquid releases, No longer applicable.

To\this calculated exposure is added,potential direct radiation exposure to an individual at the site* boundary. The only portion of the site bound~y where there is significant direct

radiation is near the radw:aste facilities at the TPG&E] North edge of the site. Due to the possibility that an individual at the ~horeline (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 rac,liation 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 shor~line recreation for the .Teen age group.

5.2 SKIN DOSE Specification 2.10 limits the dose to any organ (thyroid excepted) to less than or equal to

25 mrem/year. The dose to the ,skin is determined by summing the calculated doses from the following:

a. Deleted 10/15

b. Modular HEPA Ventilation releases, using equation (4-3). Tritium is no longer a gaseous effluent source term. '

c. Liq1:1id releases, No longer appl~cable.

d. The potential direct radiation exposure to an individual at the site boundary based on /

TLD stations, as determined in Section 5 .1 above.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-34 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 10~

II b. Modular HEP A Ventilation releases, using equation (4-3 ).

c. Liquid releases, No longer applicable.

d. The potential direct radiation exposure to an individual at the site boundary based 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 GE:NERATION DEPARTMENT SECTION ODCM VOLUME* 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-35 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING SOLIDIFICATION ' .

10/15 Deleted - Based on the status of decommissioning, HBPP no longer anticipates wastes exceeding a specific activity that is unacceptable to disposal site without solidification or exceeding Class A as defined in 10 CFR 61.

7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED IN IDGH INTEGRITY CONTAINERS 10/15 HBPP no longer anticipates wastes exceeding a specific activity that is unacceptable to disposal site without solidification or exceeding Class A as defined in 10 CFR 61. HBPP no longer anticipates disposal of wastes requiring stabilization in a High Integrity Container (HIC).

8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED RESINS AND OTHERWET WASTES ,

8.1 SCOPE 10/15 This section pertains to bead-type spent radioactive demineralizer resin, filters and other wet wastes shipped for land burial which contain a total specific activity less than the disposal site(s) *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 6l.56 or those of the disposal site(s) (whichever is more restrictive) for freestanding, noncorrosive liquid.

8.2.2 For bead resins, the precediiig 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.

10/15 8.2.4 Contract vendor solidification or dewatering services are utilized in accordance with PG&E approved supplier list and procurement procedures.

8.2.5 Vendor services may be conducted off site in accordance with their facility license and procedures. Vendor services include written confirmation of acceptable.

disposal waste form.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE. 11-36.

\

[10112]

8.2~6 Gross dewatering of resins and filters niay be performed onsite to achieve transport requirements in preparation for additional p_rocessing to a final waste form by offsite vendor services.

8.2.7 On site activities, such as managing wet soils from decommissioning excavations and process water shall be performed utilizing approved procedures or work instructions to ensure compliance with transportation regulations, disposal facility license requirements and/or waste acceptance criteria.

~

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4

.TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 26 DOSE CALCULATION MANUAL PAGE 11-37 9.0 . PROGRAM CHANGES 9.1 PURPOSE OF THE OFFSITE DOSE CALCULATION MANUAL The Offsite Dose Calculation Manual was developed to support the jmplementation of the

' Radiological Effluent Technical Specifications required by 10 CFR 50, Appendix I, and 10 CFR 50.36. The purpose of the mariual is to provide the NRCwith 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 recogi?ized that changes to the ODCM may be required dliring the Decommissioning period. All changes shall be reviewed and approved by the PSRC and the Plant Manager prior to .implementation. Tue 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 dqcumented .and retained for the duration of the possession only license.

9.3 HBPP is allowed to modify or reduce* envirorimental requirements'in the ODCM provided HBPP considers the modification or reduction from a technical and decommissioning perspective: [CTS-291]

10.0 COMMITMENTS 10115 The following commitment is implemented by this procedure. The section number that implements the commitment is noted parenthetically.

10/15 CTS-291 (Section II, 9.3) 11.0 RESPONSIBLE ORGANIZATION Radiation Protection Manager

12.0 REFERENCES

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

ODCM APPENDIX A Revision 26 PageA-1 APPENDIX A SAFSTOR BASELINE CONDITIONS

ODCM APPENDIX A Revision 26 PageA-2 1.0 LIQUID AND GASEOUS EFFLUENTS 1.1 LIQUID EFFLUENTS Baseline levels of radioactive materials contained in liquid effluents during the SAFSTOR period were established in the Environmental Report submitted as Attachment 6 to the SAFSTOR license amendment request. These values are presented for cumulative annual release and average monthly discharge in Table A-1. As of December 31, 2013, HBPP T

ceased processed liquid effluent to the discharge canal and processed liquid effluent will be transported for disposal at a regulated disposal site. Storm water and groundwater associated with excavations and groundwater inleakage to structures during decommissioning will typically be treated and released using the Ground Water Treatment System in accordance with the Stomi Water Pollution Prevention Plan (SWPP) and the associated NPDES permit. The GWTS is an Active Treatment System (ATS) is designed to remove suspended solids in order to meet release criteria of the SWPP. The system will be limited to treating water containing soluble radionuclides less than 10 times the "new"_

10 CFR 20, Appendix B, Table 2, Column 2 effluent concentration limits (ECLs) in order to ensure c*oncentrations at the Site Boundary are maintained less than limiting condition 2.3.l.

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

Table A-1 Baseline Liquid Effluent Activity Type of Activity Annual Release Monthly Average Release (Curies) (Curies)

Tritium 8.60E-2 7.17E-3 Principal Gamma Emitters (total) l.85E-1 l.54E-2 Strontium-90 3.28E-4 2.73E-5

ODCM APPENDIX A Revision 26 PageA-3 TableA-2 Baseline Gaseous Effluent Activity Type of Activity Annual Release Monthly Average Release (Curies) (Curies)

Tritium <4.0E-2 <3.3E-3 Particulate Gamma Emitters (total) 3.16E-4 2.63E-5 Strontium-90 3.38E-6 2.82E-7 TableA-3 below reflects the Gaseous Effluent Activity as a representation of the state of decommissioning during the calendar year 2013 relative to the Baseline above.

TableA-3 2013 Gaseous Effluent Activity Type of Activity Annual Release Monthly Average Release (Curies) (Curies)

Particulate Gamma.Emitters (total) <l.5E-5 <l.3E-6 Strontium-90 <lE-6 <lE-7 Particulate Alpha Emitters (total) <lE-6 <lE-7 Table A-3 data is sunimarized from the 2013 Annual ~ffluent Release Report and are listed as_ less than values because sampling results were the composite of LLD values. Tritium is no longer monitored due to a lack of significant source term.

ODCM APPENDIXB Revision 26 Page B-1 APPENDIXB BASES FOR ATMOSPHERIC DISPERSION AND DEPOSITION VALUES

ODCM APPENDIXB Revision 26 PageB-2 1.0 BASIS FOR DISPERSION/DEPOSITION VALUES - 50' STACK 1.1 The instantaneous atmospheric dispersion factor (X/Q) is taken from meteorological parameter calculations performed to evaluate reducing the height of the Unit 3 stack. The calculation report is number N238C, Revision 0, titled "Determine Effect of Humboldt Bay Power Plant Unit 3 Stack Reconfiguration on Downwind Effluent Concentrations".

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) of the calculation (N238C) provides "1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months " values for the instantaneous X/Q for the 50' stack for various stack flow rates, based on an EPA model named "ISCST". The instantaneous X/Q value used in the ODCM (6.52 x 10*4) is based on a stack flow of25,000 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 of25,000 cfin.

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 flowof25,000 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 Reglilatory Guide 1.145, Atmospheric Dispersion Models for Potential Accident 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 appllcable guidance is section 1.3.l (Releases Through Vents or Other Building Penet~ations); 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 APPENDIXB Revision 26 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:

1 X/Q = (1)

Uro( ncr y<:> z + A/2)

I XIQ = (2)

. U10(3ncrycrz) 1 XIQ = (3)

U10nl.:y cr~

where:

U10 wind speed at 10 meters above grade, equal to 1 meter/second.

cry = lateral pluine spread, equal to 4.33 meters for Pasquill Class G at a distance of 150 meters.

crz 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 3 75 meters2; based on the Refueling Building dimensions (about 36 feet high, about 112 feet long).

l:Y = lateral plume spread (including meander and building wake), meters, equal to 6cry (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/meter3 (1)

ODCM APPENDIXB Revision 26 PageB-4 X!Q = 1.32 x 10*-2 seconds/meter3 (2)

X!Q 6.59 x 10-3 second~/meter3 (3)

Per the Reg. Guide, the higher value of equations 1 and 2 is to be compared with the value for equation 3, and the lower value of that comparison should be used, with this logic, the resulting value for X/Q is 6.59 x 10-3 seconds/meter3 .

2.2 The atmospheric deposition factor (D/Q) for incidental releases is 5.39 x 10-6 meter-2 for the Particulate Ground Plane Pathway, and. is 3.29 x 10-6 meter-2 for all other deposition related pathways. The factors are calculated as described below -

2.2.1 These factors are based on the atmospheric models of Regulatory Gujde 1.111, Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-water-cooled Reactors. The applicable guidance is section C.3.b (Dry Deposition), and Figure 6 (Relative Deposition for Ground-level Releases). To determine the atmospheric deposition across a downwind sector, the value from Figure 6 is to be multiplied by the fraction of the release transported into the sector, and divided by the sector cross-wind arc length at the distance being ~onsidered. For this calculation, the deposited contamination will be assumed to be evenly distributed across the width of the plume, rather than across an arbitrary angular sector.

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

2.2.3 For the Particulate Ground Plane Pathway, the assumed distance from the emission point to the potential receptor for this calculation is 150 meters. This is the approximate distance to publicly accessible areas from the structure with the most significant potential for airborne radioactivity (i.e. from the center of the Refueling Building to the trail at the edge of the bluff). At this distance, Figure 6 provides a Relative Deposition Rate value of 1.4 x 10-4 mete{ 1. The plume width assumed for this calculation is the same as was used in equation 3 of section 2.1 .4 (above), so that the plume width is approximately 6cry. For cry equal to 4.33 meters (Pasquill Class Gata distance of 150 meters), D/Q is (1.4 x 10-4 mete{ 1)/

(6 x 4.33 meter)= 5.39 x 10-6 mete{2

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

ODCM APPENDIXB Revision 26 PageB-5 Figure 6 provides a Relative Deposition Rate value of 1.2 x 10-4 mete{ 1* 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 6cry., but at a greater distance. For cry equal to 6.07 meters (Pasquill Class Gata distance of 220 meters), D/Q is (1.2 x 104 mete{ 1)/ (6 x 6.07 meter)= 3.29 x 10-6 meter-2.

ODCM APPENDIXC Revision 26 Page C-1 APPENDIXC Deleted

Loren D. Sharp Humboldt Bay Power Plant Pacific Gas and Director & Plant Manager 1000 King Salmon Avenue Electric Company"' Humboldt Bay' Nuclear Eureka, CA 95503 707.444.0819 Internal: 375.0819 Fax: 707.444.0871 March 24, 2016 PG&E Letter HBL-16-003 U.S. Nuclear Regulatory Commission 10 CFR 50, Appendix I ATTN: Document Control Desk 10 CFR 50.36 .

Washington, DC 20555-0001 Docket No. 50-133, OL-DPR-7 Humboldt Bay Power Plant Unit 3 Annual Radioactive Effluent Release Report for 2015

Dear Commissioners and Staff:

Enclosure 1 contains the Humboldt Bay Power Plant Unit 3 "Annual Radioactive Effluent Release Report," covering the period January 1 through December 31, 2015. This report is required by Appendix B, Section 6.3 of the Humboldt Bay Quality Assurance Plan, and by Section 4.2 of the "SAFSTOR/Decommissioning Offsite Dose Calculation Manual."

Enclosure 2 contains Revjsion 26 to the "SAFSTOR/Decommissioning* Offsite Dose Calculation Manual" as required by Specification Section 4.2 of the "SAFSTOR/Decommissioning Offsite Dose Calculation Manual."

There are no new.or revised regulatory commitments (as defined in NEI 99-04) in this letter.

If you have any questions concerning this information, please do not hesitate to contact Mr.Hossein Hamzehee at (805) 545-4720 .

.Sincerely, Loren D. Sharp cc: Marc L. Dapas, NRG Region IV Administrator John B. Hickman, NRG Project Manager HBPP Humboldt Distribution

. Enclosures

Enclosure 1 PG&E Letter HBL-16-003 PACIFIC GAS AND ELECTRIC COMPANY

HUMBOLDT BAY POWER PLANT DOCKET NO. 50-133, LICENSE NO. DPR-7 HUMBOLDT BAY POWER PLANT UNIT 3 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January 1 through December 31, 2015 /