Text

Annual Radioactive Effluent Release Report for 2009
	ML101020171
Person / Time
Site:	Humboldt Bay
Issue date:	03/30/2010
From:	Roller P; Pacific Gas & Electric Co
To:	; Document Control Desk, NRC/FSME
References
	HBL-10-011, OL-DPR-07, PG&E Letter HBL-10-011
	Download: ML101020171 (135)
	v • d • e

Pacific Gas and Electric Company 1000 King Salmon Avenue Humboldt Bay PowerPlant Eureka, CA 95503 PaulJ. Roller 707/444-0700 Director andPlantManagerHumboldt Bay Nuclear March 30, 2010 PG&E Letter HBL-1 0-011 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-133, OL-DPR-7 Humboldt Bay Power Plant Unit 3 Annual Radioactive Effluent Release Report for 2009

Dear Commissioners and Staff:

contains the Humboldt Bay Power Plant Unit 3 "Annual Radioactive' Effluent Release Report," covering the period January 1 through December 31, 2009. This report is required by the Humboldt Bay Quality Assurance Plan, Attachment 4.1, Section 3.7.3. contains Revision 16 to the "SAFSTOR Offsite Dose Calculation Manual" as required by Specification Section 4.2 of the "SAFSTOR Decommissioning Offsite Dose Calculation Manual."

There are no new regulatory commitments made in this letter.,

Sincerely, Paul J. Roller cc: Elmo E. Collins, Jr.

L. Joe Davis John B. Hickman PG FossilGen HBPP Humboldt Distribution Enclosures

_ _ _ __ F 5i4'-

Enclosure 1 PG&E Letter HBL-10-011 HUMBOLDT BAY POWER PLANT UNIT 3 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January I through December 31, 2009

PACIFIC GAS AND ELECTRIC COMPANY HUMBOLDT BAY POWER PLANT DOCKET NO. 50-133, LICENSE NO. DPR-7 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT JANUARY 1, 2009 THROUGH DECEMBER 31, 2009 TABLE OF CONTENTS INTRODUCTION ................................................... 2 I. SUPPLEM ENTAL INFO RMATION ................................................................... 3 I1. GASEOUS AND LIQUID EFFLUENTS ............................................................. 7 Table 1 - Gaseous Effluents - Summation of All Releases ................................ 8 Table 2A - Gaseous Effluents - Elevated Release - Nuclides Released ............. 9 Table 2B - Gaseous Effluents - Ground-Level Releases - Nuclides Released ... 10 Table 3 - Liquid Effluents - Summation of All Releases .................................... 11 Table 4 - Liquid Effluents - Nuclides Released ............................................... 12 Ill. SO LID RA DIOACTIV E W ASTE ........................................................................ 13 Table 5 - Solid W aste/and Irradiated Fuel Shipments ...................................... 14 IV. RADIOLOGICAL IMPACT ON MAN ................................................................. 16 Table 6 - Radiation Dose for Maximally Exposed Individuals .......................... 17 V. CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM) ......... 18 VI. CHANGES TO THE PROCESS CONTROL PROGRAM (PCP) ...................... 18 VII. CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS .................. 19 VIII. INOPERABLE EFFLUENT MONITORING INSTRUMENTATION ............... I...... 19 1

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 INTRODUCTION This report summarizes gaseous and liquid radioactive effluent releases from Humboldt Bay Power Plant (HBPP) Unit 3 for the four quarters of 2009. The report includes calculated potential radiation doses from these radioactive effluents and a comparison with the 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 Manual (ODCM) limits.

During 2008 all of the spent nuclear fuel was transferred from the Spent Fuel Pool to the Independent Spent Fuel Storage Installation (ISFSI). Therefore, there was no source term for noble gases for the entire year.

The information is reported as required by the Humboldt Bay Quality Assurance Plan, .1, Section 3.7.3 and the ODCM, Section 4.2, 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, Appendix B, Section F, is not available.

Previous Humboldt Bay Power Plant 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 The Unit was last operated on July 2, 1976. Due to the long decay time since operation, short-lived radionuclides are neither expected nor reported. This includes lodines and noble gases. Kr-85 is no longer an issue since the spent fuel has been relocated to the ISFSI.

Air Particulate Filter Composites - Sr-90 Air particulate sample filters are combined for approximately monthly intervals and analyzed off-site for Sr-90.

Air Particulate Filter Composites - Am-241 Air particulate sample filters are combined for approximately monthly intervals and analyzed off-site for Am-241.

Air Particulate Filter Composites - Gross Alpha Each weekly sample filter is individually counted for gross alpha activity, rather than analyzing a monthly composite of the filters, as described in Regulatory Guide 1.21.

2

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 Gaseous Effluents - Tritium Tritium releases during plant operation were less than detection levels. Since the plant was permanently shutdown in 1976, current tritium release levels are less than the release levels that occurred during plant operations. Therefore, no tritium samples were collected during this reporting period.

Liquid Effluents - Sr-90 Batch releases are analyzed individually offsite for Sr-90, rather than analyzed as a quarterly composite as described in Regulatory Guide 1.21.

Liquid Effluents - Ni-63 Batch releases are analyzed individually offsite for Ni-63, rather than analyzed as a quarterly composite as described in Regulatory Guide 1.21.

Average Energy For HBPP, calculations for the average energy of gaseous releases of fission and activation gases are not required to be performed or reported.

I. SUPPLEMENTAL INFORMATION A. Regulatory Limits

1. Gaseous Effluents

a. Noble Gas Release Rate Limit Noble gases are no longer an issue since the spent nuclear fuel was relocated to the ISFSI in 2008.

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

c. Particulate Release Rate Limit The radioactive particulate release rate limit is based on concentration limits from 10 CFR 20, divided by an annual average dispersion factor for the sector with the least favorable atmospheric dispersion. The applicable annual average dispersion factors for elevated releases and for ground-level releases are 1.OE-5 and 6.59E-3 seconds per cubic meter, respectively.

When both elevated and ground-level releases occur, the "percent of applicable limit" in Table 1 is the sum of the values for "percent of 3

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 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 B, Table 2, Column 1.

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

C. Measurements and Approximations of Total Radioactivity

1. Gaseous Effluents - Elevated Release

a. Fission and Activation Gases Fission and activation gases are no longer an issue since the spent fuel was relocated to the ISFSI in 2008.

b. lodines Due to the long decay time since operation (shutdown July 2, 1976), no detectable releases of radioactive lodines 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, monitors the particulate activity released from the stack. This monitor was installed in December of 2009; however, initial operational issues delayed its use until the first quarter of 2010. The 2010 report will provide additional information regarding the stack monitor.

4

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 Radioactive particulates released from the plant stack are monitored by continuous sample collection on particulate filters. Filter papers are removed from the stack sampling system weekly, and are analyzed for the concentration of gamma-emitting nuclides (intrinsic germanium detector).

All statistically significant gamma peaks are identified.

After decaying at least 7 days, the filters are analyzed for gross alpha radioactivity (scintillation counter).

Filters are composited monthly and analyzed monthly for Strontium-90 (the only radioactive Strontium present) and Americium-241 by alpha spectroscopy. The monthly composite results are averaged together to produce the quarterly composite result.

The estimated error of the reported particulate release values is based on uncertainty in sample flow rate, stack flow rate, detector calibration, and typical sample counting statistics.

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

Samples are assigned to calendar quarters as of the termination of the sample period. The amount of activity reported for a calendar quarter is the activity for the combined sample time, multiplied by the ratio of the length of the calendar quarter to the sample period.

2. Gaseous Effluents - Ground-level Release

a. Fission and Activation Gases All ventilation and system vents were routed to the Unit 3 stack during the report period. Refer to the discussion for elevated releases.

b. lodines All ventilation and system vents were routed to the Unit 3 stack during the report period. Refer to the discussion for elevated releases.

c. Particulates All ventilation and system vents were routed to the Unit 3 stack during the report period. Refer to the discussion for elevated releases.

3. Liquid Effluents 5

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009

a. Batch Releases Water from contaminated plant systems was collected, filtered, treated with Cesium-specific ion-exchange media, and analyzed before discharge (on a batch basis) through the liquid radwaste process monitor. Analysis of weekly composite samples from the plant effluent canal did not detect any additional release of radioactive liquids during the report period.

Samples of liquid waste batches were analyzed for the concentration of gamma-emitting nuclides (intrinsic germanium detector). All statistically important peaks were identified. Additionally, all batches were analyzed for radioactive strontium (Sr-90), gross alpha, Ni-63 and tritium.

The error of the reported release values is estimated based on uncertainty in sample volume, batch volume, detector calibration, and typical sample counting statistics.

The MDA for all batch samples was less than the applicable LLD presented in the ODCM.

b. Continuous Releases There were no continuous liquid effluent releases during this report period.

D. Batch Release Statistics

1. Liquid

a. Num ber of batch releases .................................................. 5

b. Total time period for batch releases ................. 8.02E2 minutes

c. Maximum time period for a batch release ....... 1.64E2 minutes

d. Average time period for a batch release ......... 1.60E2 minutes

e. Minimum time period for a batch release ....... 1.55E2 minutes

2. Gaseous

a. Num ber of batch releases .................................................. 0

b. Total time period for batch releases ................................... N/A

c. Maximum time period for a batch release .......................... N/A 6

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009

d. Average time period for a batch release ............................ N/A

e. Minimum time period for a batch release ........................... N/A E. Abnormal Release Statistics

1. Liquid,

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

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

2. Gaseous )

a. Num ber of abnorm al releases .............................................. 0

b. Total activity released ........................... ..................... N/A I1. GASEOUS AND LIQUID EFFLUENTS A. Gaseous Effluents Table 1 summarizes the total quantities of radioactive gaseous effluents.

Table 2A presents the quantities of each of the nuclides determined to be released from the stack (elevated release point). Table 2B presents the quantities of each of the nuclides determined to be released by other routes (ground level release points).

B. 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.

The quantity of radionuclides released in 2009 is similar to 2008, but higher than in previous years prior to 2006. The increase in the activity was due to the 2006 breach in the resin transfer line from the spent fuel pool demineralizer to the resin disposal tank. This breach resulted in resin being spilled into the offgas tunnel and rainwater in-leakage to the tunnel transporting the contamination into the liquid radioactive waste system. There was no unmonitored release of radioactivity from this incident.

The higher quantities of radionuclides discharged did not exceed the requirements of the ODCM.

7

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 TABLE 1 GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES Units First Second Third Fourth Est. Total Particulates I Quarter I Quarter I Quarter I Quarter I Error. %

1. Total release Ci <2.04E-06 <2.67E-06 <2.14E-06 <2.04E-061 3.60E1

2. Average release rate tICi/sec <2.60E-07 <3.40E-071 <2.73E-07 <2.74E-07

3. Percent of applicable limit % <2.88E-06 <3.77E-06 <3.03E-06 <3.05E-06

4. Applicable limit 4Ci/cc 9.01E-11 9.01E-11 9.01 E-11 9.01E-11

5. Gross alpha radioactivity Ci <7.24E-8 <6.38E-08 <7.84E-08 9.89E-08 Note: 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 determined to be 84% Cs-137, 11% Co-60 and 5% Sr-90.

8

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 TABLE 2A GASEOUS EFFLUENTS - ELEVATED RELEASE - NUCLIDES RELEASED Continuous Mode Nuclides Released Unit First Second Third Fourth Quarter Quarter Quarter Quarter Particulates Cobalt-60 Ci <1.01E-06 <9.36E-07 <9.30E-07 <8.88E-07 Strontium-90 Ci <2.25E-07 <8.71 E-07 <3.73E-07 <3.40E-07 Cesium-137 Ci <8.02E-07 <8.50E-07 <8.39E-07 <7.58E-07 Am-241 Ci <3.70E-09 <4.64E-09 <4.18E-09 4.99E-09 Total for period Ci <2.04E-06 <2.66E-06 <2.15E-06 <2.04E-06 Note: 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 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.

9

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 TABLE 2B GASEOUS EFFLUENTS - GROUND-LEVEL RELEASES NUCLIDES RELEASED Continuous Mode Nuclides Released Unit First Second Third Fourth Quarter Quarter[ Quarter Quarter Particulates Cobalt-60 Ci N/A N/A N/A N/A Strontium-90 Ci N/A N/A N/A N/A Cesium-137 Ci N/A N/A N/A N/A Americium-241 Ci N/A N/A N/A N/A Total for period Ci N/A N/A N/A N/A Note: N/A - There were no ground level gaseous effluents during the report period.

10

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 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 3.43E-4 7.26E-05 0 9.29E-5 1.OOE1

2. Average diluted p.Ci/ml 1.36E-11 3.01E-12 0 4.03E-12 concentration

3. Percent of applicable limit % 2.06E-03 3.75E-4 0 5.38E-4

4. Applicable limit PCi/ml 6.60E-7 8.02E-07 0 7.50E-7 B. Tritium

1. Total release Ci 2.57E-3 2.88E-5 0 1.45E-4 1.50E1

2. Average diluted pCi/ml 1.02E-10 1.91E-12 0 6.31E-12 concentration

3. Percent of applicable limit %

_1.02E-5 1.19E-7 0 6.31 E-7

4. Applicable limit p.Ci/ml 1.00E-03 1.00E-03 0 1.00E-03 C. Gross Alpha Radioactivity

1. Total release Ci 1.19E-6 F9.93E-7 0 1.06E-7 1.005I D. Volume of waste released Liters 7.62E+04 2.48E+04 0 2.54+04 3.00E0 (prior to dilution)

E. Volume of dilution water Liters 2.52E+10 2.41E+10 N/A 2.30E+10 1.50E1 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 nuclides), values with the < symbol are ignored (i.e. treated as zero).

I1

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 TABLE 4 LIQUID EFFLUENTS - NUCLIDES RELEASED Batch Mode Nuclides Released Unit First Second Third Fourth Quarter Quarter Quarter Quarter Strontium-90 Ci 2.12E-4 2.34E-5 0 5.07E-5 Cesium-1 37 Ci 9.50E-05 4.37E-5 0 2.22E-5 Cobalt-60 Ci <1.70E-6 <4.74E-7 0 6.92E-7 Americium-241 Ci <8.49E-6 <3.43E-6 0 <1.82E-6 Nickel-63 Ci 3.55E-5 5.49E-6 0 1.93E-5 Tritium Ci 2.57E-3 2.88E-5 0 1.45E-4 Alpha Emitters Ci 1.19E-6 9.93E-7 0 1.06E-7 Total for period Ci 2.91E-3 1.02E-04 0 2.38E-4 Continuous Mode Nuclides Released Unit First Second Third Fourth Quarter Quarter Quarter Quarter Strontium-90 Ci N/A N/A N/A N/A Cesium-1 37 Ci N/A N/A N/A N/A Cobalt-60 Ci N/A N/A N/A N/A Americium-241 Ci N/A N/A N/A N/A Total for period Ci N/A N/A N/A N/A 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 nuclides), values with the < symbol are ignored (i.e. treated as zero).

12

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 III. SOLID RADIOACTIVE WASTE Table 5 summarizes the disposal of solid radioactive waste made during 2009.

13

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 TABLE 5 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. Solid Waste Shipped Offsite For Burial Or Disposal

1. Type of Waste Unit 12 Month Estimated Total Error, Period %

a. Spent resins, filter sludges, Cubic Meter 0 NA evaporator bottoms, etc. Ci 0 NA

b. Dry compressible waste, Cubic Meter 252.8 1.00E1 contaminated equipment, etc. Ci 0.101 5.60E1

c. Irradiated components, Cubic Meter 0 NA control rods, etc. Ci 0 NA

d. Other (Processed Waste) Cubic Meter 0 NA Ci 0 NA

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

a. Spent resins, filter sludges, % NA NA.

evaporator bottoms, etc.

b. Dry compressible waste, % H-3 1.02E-2 contaminated equipment, etc. % C-14 5.12E-5

% Fe-55 1.91 E-3

.% Co-60 8.75E-3

% Ni-59 8.93E-5

% Ni-63 2.25E-2

% Sr-90 9.84E-4

% Tc-99 2.08E-4

% 1-129 1.69E-4

% Cs-137 5.17E-2

% U-233 1.07E-4

% U-238 1.03E-4

% Pu-238 1.94E-4

% Pu-239 2.35E-4

% Pu-240 2.35E-4

% Pu-241 2.74E-3

% Am-241 1.08E-3

% Cm-243 2.04E-4

% Cm-244 2.04E-4

c. Irradiated components, % NA NA control rods, etc. I I 14

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 TABLE 5 - Continued SOLID WASTE AND IRRADIATED FUEL SHIPMENTS B. Irradiated Fuel Shipments

1. Irradiated Fuel Disposition Number of Mode of Destination Shipments Transportation None N/A, N/A 15

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 IV. RADIOLOGICAL IMPACT ON MAN A comparison of calculated doses from various paths has shown that the offsite doses are primarily due to direct radiation and to the consumption of aquatic foods. Maximum doses to individuals (for the maximally exposed organs and age groups) are summarized in Table 6. 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.

A. Doses to the average individual in the population from all receiving-water-related pathways were calculated for detected releases, based on the guidance of Regulatory Guide 1.109. The highest results were less than 0.01 mrem/yr (total body) for the Adult age group, and less than 0.030 mrem/yr for the bone of the Adult age group.

These doses are 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 (calculated with the same dispersion and deposition parameters as were used to calculate maximum exposure) was less than 0.001 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 in 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 result in an exposure of 0.01mrem/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.

16

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 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 Water-related Pathways (1) <0.01 (5) <0.01 (5) <0.01 (5)7 <0.01 (5) <0.01 (5)]

<0.01 (6) <0.01 (6) <0.01 (6) <0.01 (6) <0.01 (6)

Airborne Effluents Particulates (2) 0.00 (7) 0.00 (7) 0.00 (7) 1.60E-05 (8) 1.60E-05 (8) 0.00 (7) 0.00 (7) 0.00 (7) 3.54E-04 (9) 3.54E-04 (9)

Direct Radiation (4) <0.01 <0.01 <0.01 <0.01 <0.01 Notes

1. Maximum total body and organ doses to individuals in unrestricted areas from receiving-water-related exposure pathways were calculated from the average concentrations of liquid releases detected during the report period, following the applicable portions of Regulatory Guide 1.109 and NUREG-4013.

2. Maximum total body and organ doses to individuals in unrestricted areas from airborne-particulate-related exposure pathways were 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.

3. Total body and skin doses to potentially exposed individuals located at the point of maximum offsite ground-level concentrations of radioactive gaseous effluents calculated because there were detected releases of radioactive noble gases.

4. Total body doses (to the maximum 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).

5. Total body (Adult age group).

6. Bone (Adult age group).

7. For stack releases for the first three quarters of 2009, a majority of the results were "not detected", resulting in a total activity considered "not detected," for which no dose is calculated.

8. A small amount of Am-241 was detected by alpha spectroscopy on the November monthly composite. Total body (Teen age group) value is a calculated Am-241 exposure to the maximum exposed age group.

9. A small amount of Am-241 was detected by alpha spectroscopy on the November monthly composite. Bone (Teen age group) value is a calculated Am-241 exposure to the maximum exposed age group.

17

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 V. CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL (ODCM)

The ODCM was revised once during 2009. The changes maintained the level of radioactive effluent control and dose commitment required by regulation, and did not adversely affect the accuracy or reliability of effluent, dose or setpoint calculations.

Revision 16 to the ODCM became effective on December 5, 2009. The changes in this revision included:

The Title was changed to the SAFSTOR/Decommissioning Offsite Dose Calculation Manual.

Table 2-2 was updated to include the new Stack Particulate Airborne Monitoring System.

- Table 2-4 was updated to include the Stack Particulate Airborne Monitoring System.

- Table 2-6 was updated to include the Stack Particulate Airborne Monitoring System.

- Table 2-7 was updated to include changes to the REMP Program. These changes include:

adding the Cross-contamination Plan Air Samplers to the REMP Program, deleting the 24 non-quality related offsite TLD stations, changing the offsite control location to quality related, Sr-90 analysis was added to the weekly canal sample, 7 new groundwater wells were added, Sr-90, Am-241 were added to the groundwater well analyses, the groundwater well alpha/beta analysis was made quality related, algae sampling was eliminated, milk sampling was made quality related and Sr-90 analysis was added to mil samples..

- Table 2-10 was updated to include changes to the REMP Program.

- Part II Section 1.2 - The section was changed to include the alarm setpoint calculation for the Stack Particulate Airborne Monitoring System.

Table 2-8 was updated to include changes to the REMP Program.

Table 2-10 was updated to include changes to the REMP Program.

Part II Section 1.2 - The section was changed to include the alarm setpoint calculation for the Stack Particulate Airborne Monitoring System.

VI. CHANGES TO THE PROCESS CONTROL PROGRAM (PCP)

There were no changes to the Process Control Program during the report period.

18

HUMBOLDT BAY POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2009 VII. CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS There were no changes to the Radioactive Waste Treatment Systems during the report period.

VIII. INOPERABLE EFFLUENT MONITORING INSTRUMENTATION The Radioactive Liquid Effluent Monitoring System was declared inoperable from March 4, 2009, through May 26, 2009, and July 24, 2009, through September 28, 2009. There were no planned or unplanned radioactive liquid discharges during these periods of inoperability.

The cause of the inoperability was determined to be communication errors between the computers in the system. The communication errors resulted in spurious alarms.

The system was declared operational again after thorough testing determined that any communication alarms would result in a conservative failure of the system. HBPP is working with the manufacturer of the system to establish a permanent resolution to this issue.

New Stack Monitor A continuous monitor equipped with an alpha spectrometer, with its response calibrated for Am-241, to monitor the particulate activity released from the stack was installed in December of 2009. On December 16, 2009, initial operational issues resulted in the particulate channel being inoperable until identified during the first quarter of 2010 surveillance testing. There were no detectable radioactive discharges during 2009 for the period of inoperability. The 2010 report will provide additional information regarding the new stack monitor inoperability and corrective actions taken.

19

Enclosure 2 PG&E Letter HBL-10-011 HUMBOLDT BAY POWER PLANT UNIT 3 SAFSTORIDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL REVISION 16 INCLUDING CHANGES MADE DURING 2009

U p Nuclear Power Generation SECTION ODCM VOLUME 4 Humboldt Bay REVISION 16 EFFEC DATE 12-5-09 Power Plant PAGE i TITLE APPROVED BY SAFSTOR/SAFSTOR7 ORIGINAL SIGNED 11-12-09 O..SIT.DECOMMISSIONING OFFSIT-EE DOE- DIRECTORIPLANT MANAGER / DATE 1-3HB NUCLEAR (Procedure Classification - Quality Related)

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

Part I - Specifications and Part II - Calculational Methods and Parameters.

Part I contains the specifications for liquid and gaseous radiological effluents (RETS) developed in accordance with NUREG-0473, Draft RadiologicalEffluent Technical Specifications - BWR, by License Amendment Request (LAR) 96-02 and the radiological environmental monitoring program (REMP). Both the RETS and the REMP were relocated from the Technical Specifications by LAR 96-02 in accordance with the provisions of Generic Letter 89-01, Implementation of Programmatic Controlsfor RadiologicalEffluent Technical Specifications in the Administrative ControlsSection of the Technical Specifications and the Relocation of ProceduralDetails ofRETS to the Offsite Dose CalculationManual or to the Process Control Program,issued by the NRC in January, 1989.

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

/

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTOR OFFSITEDECOMMISSIONING REVISION 16 OFFSITE PAGE ii DOSE CALCULATION -MANUAL The ODCM contains the requirements for the REMP. This program consists of monitoring stations and sampling programs based on the SAFSTOR Decommissioning Plan and the Environmental Report which established baseline conditions for soil, biota and sediments. The REMP also includes requirements to participate in an interlaboratory comparison program.

Part II of the ODCM contains the calculational methods developed, following the above guidance, to be used in determining the dose to members of the public resulting from routine radioactive effluents released from HBPP during the SAFSTOR decommissioningperiod. Part II also contains the methodology used to determine effluent monitor alarm/trip setpoints which assure that releases of radioactive materials remain within specified concentrations.

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTOR OFFSITEDECOMMISSIONING REVISION 16 OFFSITE PAGE iii DOSE CALCULATION -MANUAL TABLE OF CONTENTS PART I - SPECIFICATIONS Section Title Page 1.0 DEFINITIONS I-1 2.0 SPECIFICATIONS 1-8 2.1 Radioactive Liquid Effluent Monitoring Instrumentation 1-8 2.2 Radioactive Gaseous Effluent Monitoring Instrumentation I-11 2.3 Liquid Effluent - Concentration 1-14 2.4 Liquid Effluent - Dose 1-17 2.5 Liquid Waste Treatment 1-18 2.6 Gaseous Effluents - Dose Rate 1-19 Deleted 2.8 Gaseous Effluents: Dose - Tritium and Radionuclides in Particulate Form 1-23' 2.9 Solid Radioactive Waste 1-24 2.10 Total Dose 1-25 2.11 REMP Monitoring Program 1-26 2.12 REMP Interlaboratory Comparison Program 1-39 2.13 Radioactive Waste Inventory 1-40 3.0 SPECIFICATION BASES 1-41 3.1 Radioactive Liquid Effluent Monitoring Instrumentation Basis 1-41 3.2 Radioactive Gaseous Effluent Monitoring Instrumentation Basis 1-41 3.3 Liquid Effluent Concentration Basis 1-41 3.4 Liquid Effluent Dose Basis 1-42 3.5 Liquid Waste Treatment Basis 1-42 3.6 Gaseous Effluents Dose Rate Basis 1-42 3.7 Deleted 1-43 3.8 Gaseous Effluents: Tritium and Radionuclides in Particulate Form Dose Basis 1-44 3.9 Solid Radioactive Waste Basis 1-45 3.10 Total Dose Basis 1-45 3.11 REMP Monitoring Program Basis 1-45 3.12 REMP Interlaboratory Comparison Program Basis 1-46 3.13 Radioactive Waste Inventory Basis 1-46

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTORISAFSTOR.FFS.T.EDECOMMISSIONING REVISION 16 OFFSITE PAGE iv DOSE CALCULATION -MANUAL PART I - SPECIFICATIONS - (Continued)

Section Title Page 4.0 ADMINISTRATIVE CONTROLS 1-47 4.1 Annual Radiological Environmental Monitoring Report 1-47 4.2 Annual Radioactive Effluent Release Report 1-52 4.3 Special Reports 1-53 4.4 Major Changes to Radioactive Waste Treatment Systems 1-53 4.5 Process Control Program Changes 1-54 PART II - CALCULATIONAL METHODS AND PARAMETERS Section Title Page 1.0 EFFLUENT MONITOR SETPOINT CALCULATIONS 11-1 1.1 Liquid Effluent Monitors 11-1 1.2 Gaseous Effluent Monitor II-4 2.(0 LIQUID EFFLUENT DOSE CALCULATIONS II-6 2.1 Month (31 Day Period) II-6 2.2 Calendar Quarter 11-6 2.3 Calendar Year 11-6 2.4 Liquid Effluent Dose Calculation Methodology 11-11 3.0 LIQUID WASTE TREATMENT 3.1 Treatment Requirements II"-11 IH-1 1 3.2 Treatment Capabilities

4. 0 GASEOUS EFFLUENT DOSE CALCULATIONS II-14 4.1 Dose Rate 11-14 4.2 Deleted 4.3 Dose - Tritium and Radionuclides in Particulate Form 11-17

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFSTOR OFFSITEDECOMMISSIONING REVISION 16 OFFSITE PAGE v DOSE CALCULATION -MANUAL PART II - CALCULATIONAL METHODS AND PARAMETERS - (Continued)

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTOR OFFSITEDECOMMISSIONING REVISION 16 OFFSITE PAGE vi DOSE CALCULATION -MANUAL LIST OF TABLES - PART I Table Title Page 1-1 Frequency Notation I-6 2-1 Radioactive Liquid Effluent Monitoring Instrumentation 1-9 2-2 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 1-10 2-3 Radioactive Gaseous Effluent Monitoring Instrumentation 1-12 2-4 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance 1-13 Requirements 2-5 Radioactive Liquid Waste Sampling and Analysis Program 1-15 2-6 Radioactive Gaseous Waste Sampling and Analysis Program 1-20 2-7 HBPP Radiological Environmental Monitoring Program 1-28 2-8 Reporting Levels for Radioactivity Concentrations In Environmental Samples 1-30 2-9 Detection Capabilities for Environmental Sample Analysis Lower Limits Of 1-31 Detection (LLD) 2-10 Distances and Directions To Environmental Monitoring Stations 1-33 4-1 Radiological Environmental Monitoring Report Annual Summary - Example 1-49

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTOR OFFSITEDECOMMISSIONING REVISION 16 OFFSITE PAGE vii DOSE CALCULATION -MANUAL LIST OF TABLES - PART II Table Title Page 1-1 Liquid Effluent Monitor Alarm Setpoints 11-3 2-1 Ingestion Dose Factors for Adult Age Group 11-8 2-2 Ingestion Dose Factors for Teen Age Group 11-9 2-3 Ingestion Dose Factors for Child Age Group 11-9 2-4 Bioaccumulation Factors for Saltwater Environment 11-10 2-5. Average Individual Foods Consumption for Various Age Groups 11-10 2-6 Maximum Individual Foods Consumption for Various Age Groups, 11-10 4-1 Inhalation Dose Factors for Adult Age Group 11-29 4-2 Inhalation Dose Factors for Teen Age Group 11-29 4-3 Inhalation Dose Factors for Child Age Group 11-30 4-4 Inhalation Dose Factors for Infant Age Group 11-30 4-5 External Dose Factors for Standing on Contaminated Ground 11-31 4-6 Average Individual Foods Consumption for Vari6us Age Groups 11-31 4-7 Maximum Individual Foods Consumption for Various Age Groups 11-31 4-8 Ingestion Dose Factors for Adult Age Group 11-32 4-9 Ingestion Dose Factors for Teen Age Group 11-32 4-10 Ingestion Dose Factors for Child Age Group 11-33 4-11 Ingestion Dose Factors for Infant Age Group 11-33 4-12 Stable Element Transfer Data For Cow-Milk Path 11-34 4-13 Stable Element Transfer Data For Cow-Meat Path 11-34

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTORISAFST*0R OFF....DECOMMISSIONING REVISION 16 OFFSITE PAGE viii DOSE CALCULATION -MANUAL LIST OF FIGURES - PART I Figure Title Page 1-1 Site Boundary 1-7 2-1 HBPP Onsite TLD Locations 1-34 2-2 HBPP Onsite Monitoring Well Locations 1-35 2-3 HBPP Offsite Sampling Locations 1-36 2-4 HBPP Offsite Sampling Locations (Continued) 1-37 2-5 HBPP Offsite Sampling Locations (Continued) 1-38

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE I-1 PART I - SPECIFICATIONS 1.0 DEFINITIONS 1.1 ACTION ACTION shall be that part of a control that prescribes remedial measures required under designated conditions.

1.2 BASELINE COMPARISON A BASELINE COMPARISON shall be a comparison of cumulative radioactivity releases for a stated period with the baseline radioactivity release conditions established by the ENVIRONMENTAL REPORT.

1.3 CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. CHANNEL CALIBRATION may be performed.

by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.

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

1.5 CHANNEL FUNCTIONAL TEST

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAfSTOR.DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-2 1.6 ENVIRONMENTAL REPORT Submitted as Attachment 6 to the SAFSTOR license amendment request, the ENVIRONMENTAL REPORT established baseline radiological environmental conditions for soil, biota and sediments. In acor-dane with the NRC. approved SAFSTOR DecommnissioningPlan, these baseline. conditions will only need to be r-eestablished pior to IDECON if a significant release during SAFSTOR occurfs as the result of an accident 1.7 FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1-1.

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAF-STOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-3 1.12 OFFSITE DOSE CALCULATION MANUAL The OFFSITE DOSE CALCULATION MANUAL contains the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm Trip Setpoints, and in the conduct of the Radiological Environmental Monitoring Program.

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

1.13 OPERABLE - OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment, that are required for the system, subsystem, train, component or device to perform its function(s), are also capable of performing their related support function(s).

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

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

m

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-4 1.16 RESTRICTED AREA The RESTRICTED AREA is defined by 10CFR20.1003. The physical location(s) of the RESTRICTED AREA shall be defined in plant procedures.

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

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4

'ITLE SAFSTOR/SAF-S-TORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-5 1.22 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing particulates from the gaseous exhaust stream prior to release to the environment.

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-6 Table 1-1 FREQUENCY NOTATION Notation Frequency 1Extension 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.

j 1The 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 rITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-7 Figure 1-1 SITE BOUNDARY

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-8 2.0 SPECIFICATIONS 2.1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITIONS 2.1.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 2-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Specification 2.3 are not exceeded.

APPLICABILITY: At all times ACTION:

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 416 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-9 Table 2-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument OPERABLE A( TION

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Process Water Monitor 1 21

2. Flow Rate Measurement Devices

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

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

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

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

Otherwise, suspend releases of radioactive materials via this pathway.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR.DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-10 Table 2-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL Instrument CHECK CHECK CALIBRATION TEST

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

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

2. Flow Rate Measurement Devices

a. None Table Notation (1) Alarm functions and background readings shall be checked weekly. If a background reading exceeds the equivalent of 5 x 10- 6 micro-Ci/ml of Cs- 137, the cause will be investigated and remedial measures taken to reduce the background reading.

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

a. Instrument indicates measured levels above the alarm setpoint.

b. Circuit failure.

c. Instrument indicates a downscale failure.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-11 2.2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION1 LIMITING CONDITIONS 2.2.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 2-3 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of specification 2.6 are not exceeded.

APPLICABILITY: Whenever the ventilation system is in operation.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required above, withou, delay-suspend work that could result in the release of radioactive gaseous effluents monitored by the affected channel, or change the setpoint so that it is acceptably conservative, or declare the channel inoperable.

b. -b---With one or more radioactive gaseous effluent monitoring instrumentation channels inoperable, take the ACTION shown in Table 2-3.

For the instrumentation covered, exert best efforts to return the inoperable instrument(s) to OPERABLE status within 30 days. If the affected instrument(s) cannot be returned to OPERABLE status within 30 days, provide information on the reasons for inoperability and lack of timely corrective action in the next Radioactive Effluent Release Report.

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

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

Performing the Quarterly SPAM Calibration STP requires declaring the Stack.

Particulate Airborne Monitoring System INOPERABLE.

The NRC classifies effluent monitoring as either liquid or gaseous. With the removal of~the spent fuel to the ISFSI, the remaining significant radioactive sourcc at Humboldt Bay Power Plant is particulate canxing alpha contamination, potentially released to the environment through the "gaseous" airflow of the ventilation svstem, in unlikely events including the failure of the HEIPA filters, T[here is no significant gaseous activity.

per se: Thus PG&E's "gaseeous effluent monitoring instrumentation' is actually particulate alpha monitoring, and is commonly referred to as the Stack Particulate Alpha Monitor or Stack Particuilite Alpha Monitoring System or SPAMS in most PG&E documentation,

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-12 IT-he NRC Classifies e4tlh mim*rI" her lid s. With4he-remwa1-ethe-&pet fue! to ihe 1SFSl-he-r-minini-iniat araioaetive source at Hanbldl a.. '....r ,,ult .ar1yin olant-ia a...t 1 tw.am+,H., ra.. tentially released to the nv-on"m--throoat--

t4eu"Hrfo ~e ,Itinwtm in unlikel'g events ineludinmo rhe4ioG-44hc

" 141PA~ Alturs. Th1were isoa-4i~-if4 ali~t-iea~ts-actwt SURVEILLANCE REQUIREMENTS 2.2.2 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST CHANNEL CH4ECK and CALIBRATION operations at the frequencies shown in Table 2-4.

________ -- I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 416 TITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-13 Table 2-3 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument OPERABLE ACTION

1. Stack Gas-Particulate Airborne Monitoring System (SPAMS)

a. Noble Gas ActivityContinuous Alpha Monitor!*2 N-,A I A,C

b. Particulate Sampler iodine-.. v.pe B

c. Paeticulate Sampler Effluent System Flow Rate Monitor 1 23C_-2-D

d. Continuous Alpha Monitor Flow Rate Monitor Efflulent -26A System Flo w Rate Monitor *

e. Particulate Sampler Flow Rate Monitorý* ~~lj 11- A, Rate Menitefr**

Table Notation I ACTION A The pa;4ie atesamplercontinuous alpha monitor may be taken out of service for calibration or maintenance, but shall be returned to service as soon as practicable within the 30 day period allowed by ACTION 2.2.l.b. If the continuous alpha monitor is inoperable suspend work that could result in the release of radioactive gaseous effluents monitored by the affected channel.

ACTION B The narticulate samnler may be taken out of service for calibration or maintenance.

but shall be returned to service as soon as practicable within the 30 day period allowed bv ACTION 2.2.1 .b. If the particulate samnler is inoperable secure ventilation and suspend work that could result in the release of radioactive gaseous effluents monitored by the affected channel.

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 IITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-14 ACTION D With the number of channels OPERABLE less than that required by the Minimum Channels OPERABLE requirement, the effluent system default flow rate may be used for effluent calculations.

XT- -'-- - -- l-fNtj. titeL1ttitu ifi ttii- g.iZ I- utiiiuiittriii tjy te Loss of sampler flow would result in alarm and failure of the particulate sampler.

Table 2-4 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE FUNCTIONAL Instrument CHECK CHECK CALIBRATION TEST

1. Stack Gas Monitoring System

a. Noble Gas Activity Monitor*. Continuous N.AD N-.A.Q N-7A-. N A-Q.

Alpha Monitor

b. Particulate Samplerledine- WN=A,. N.A. N.A. 4.A.

Sampler-*

c. Par-tiulate Sampler DNýA N.A.N-.A-. AN-.A N.

Effluent System Flow Rate Monitor

d. Effluent System Flow DW N.A.N-.A= QA Q Rate MenitefrContinuous Alpha Monitor Flow Rate Monitor

e. Particulate Sampler Flow QD N.A. N-A.Q N4 Q Rate Monitor Table Notation (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the alarm setpoint.

b. Instrument indicates a downscale failure.-*

c. Loss of sample flowrate

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

-L 1,1 1 1

11 ý , riaoll : ra A -fl- r "llp I P - V -+ -,--:+

12111a Q Q:Qs=n AV I

C n1r-MQ Ckr-: 41" + - -- + 6a A Phr

-3

I- T 1FT-) T's Tlb

--- l -- - -..-.. . .. . .-

-.-.. -I I I ........

& 1,1 1 ,1 "

1 I\ItOuH94 tn1S 1' omlr~uir-ement t oitilhe [44IUIA+L 444+NU-t-H +4N icrOneratnna ft f p -hty1xs xxill rrs31'rn arsrtTJTr32+~fn.. in v,- ý -- 1.11

+fl o 1111. Lfhtfl"rzrj. n- s. ý V+ tm Lt 5

f.;.ill.rp nt.'n i' nt r*ni,;r'*ll.nfr th4 .mrnit.irto he .PE. APT F.

1

I ! . .. . .. . 1

- 4UAlhIiq-h thvir'tiinn i' nor-Mnllv rni-i14rp'd tor-nh14Icer- nlnnt mnnitpving' it i'q not mriii~irerl or

I 1 I I TTrinn .

1 11 - n a Q QC2 C "PC mlck:n 9:=_ ncx 9:3 Q 6AM tic P I IQ CXQQ Q'2113=6a 2.3 LIQUID EFFLUENT - CONCENTRATION LIMITING CONDITIONS 2.3.1 The instantaneous concentration of radioactive material released beyond the SITE BOUNDARY shall be less than or equal to 10 times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2.

APPLICABILITY: At all times.

ACTION:

With the instantaneous concentration of radioactive materials released beyond the SITE BOUNDARY exceeding the above limits, without delay restore the concentration of radioactive materials being released beyond the SITE BOUNDARY to within the above limits.

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

2.3.3 The results of the radioactivity analyses shall be used with the calculational methods in Part II of the ODCM to assure that the concentrations of radioactive

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 416 SAFSTOR/SAF.STOR.DECOMMISSIONING OFFSITE TITLE DOSE CALCULATION MANUAL PAGE 1-16 material released' to Humboldt Bay are maintained within the limits of Specification 2.3.1.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME REVISION 416 tITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-17 Table 2-5 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM Minimum Lower Limit Sampling Analysis Type of Activity of Detection Liquid Release Type Frequency Frequency Analysis (LLD)

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

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

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

1. Caisson Sump Grab Sample Compositeb Emitterse D M H-3 1 x 10- 5 Grab Sample Compositeb Gross Alpha 1 x 107 D Q Sr-90 5 x 108 Grab Sample Compositeb Table Notation The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 sb LLD =

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTORISAFST 'TO-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-18 Table 2-5 (Continued)

Table Notation (Continued)

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

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

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

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

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

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

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

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

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

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

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

I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-19 2.4 LIQUID EFFLUENT - DOSE LIMITING CONDITIONS 2.4.1 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released beyond the SITE BOUNDARY shall be limited as follows:

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

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

APPLICABILITY: At all times.

ACTION:

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

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

b. Corrective action taken to reduce the release of radioactive materials in liquid effluents during the remainder of the current calendar quarter and >during the remainder of the current calendar year so that the dose or dose commitment to a MEMBER OF THE PUBLIC from this source is less than or equal to 3 mrem total body and less than or equal to 10 mrem to any organ during the calendar year.

SURVEILLANCE REQUIREMENTS 2.4.2 At least once per 31 days, perform a dose calculation for the current calendar quarter and the current calendar year, OR perform a BASELINE COMPARISON for liquid effluent radioactivity released to date for the current calendar quarter and current calendar year. IF the comparison indicates that the activity released to date exceeds the Environmental Report baseline annual release, THEN a dose calculation shall be performed for the current calendar quarter and the current calendar year.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-20 2.5 LIQUID WASTE TREATMENT LIMITING CONDITIONS 2.5.1 The LIQUID RADWASTE TREATMENT SYSTEM shall be used, as appropriate, to reduce radioactive materials in liquid wastes prior to their discharge, when projected monthly doses due to liquid effluents discharged to Humboldt Bay would exceed the action levels of 0.06 mrem whole body or 0.2 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

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

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

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

c. Actions taken to prevent recurrence.

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

OR Before approving any release, a dose calculation shall be performed for.

comparison with Specification 2.5.1.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-21 2.6 GASEOUS EFFLUENTS - DOSE RATE LIMITING CONDITIONS 2.6.1 The dose rate at or beyond the SITE BOUNDARY, due to radioactive materials released in gaseous effluents, shall be limited as follows:

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

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

APPLICABILITY: At all times.

ACTION:

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

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

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

2.6.4 Radioactive particulates, with half-lives of greater than 8 days, in gaseous effluents released to the environment shall be sampled and analyzed in accordance with the sampling and analysis program of Table 2-6, and their concentrations shall be compared with the limits of 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 hive been exceeded. The actual sample period shall be used to determine the dose rate during the sample period.

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

(pCi/ml)a Plant Stack Continuous d' W__W Principal Gamma I x 10-11 Particulate Emitterse Sample Continuous MW Gross Alpha I x 1011 Particulate Sample Continuous. W Gross Beta 6.7 x 10-12 Particulate Sample Continuousd Q Sr-90 I x 10-11 Composite Particulate Sample Continuousd O Am-241 I X 10-14 Composite Particulate Sample Continuousd Continuous Gross Alpha 1 x 10-12 f Alpha Particulate Monitor Table Notation a The LLD* is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

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

4.66 sb LLD =

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-23 Table Notation (Continued)

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

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

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

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

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

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

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

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

b Deleted.

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-24

( _____________________________________

2.7 Deleted 2.8 GASEOUS EFFLUENTS: DOSE - TRITIUM AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITIONS 2.8.1 The dose to a MEMBER OF THE PUBLIC from the release of tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents released beyond the SITE BOUNDARY shall be limited as follows:

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

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

APPLICABILITY: At all times.

ACTION:

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-25 As explained in Specification Bases section 3.8, neither routine surveillance nor dose calculations are required for Tritium in gaseous effluents.

2.9 SOLID RADIOACTIVE WASTE LIMITING CONDITIONS 2.9.1 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements.

APPLICABILITY: At all times.

ACTION:

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-26 2.10 TOTAL DOSE LIMITING CONDITIONS 2.10.1 The calendar year dose or dose commitment to any MEMBER OF THE PUBLIC, due to releases of radioactivity and radiation, from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the total body or any organ (except the thyroid, which shall be limited to less than or equal to 75 mrem).

APPLICABILITY: At all times.

ACTION:

With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specification 2.4.1 .a, 2.4.1 .b, 2.8.1 .a, or 2.8.1.b, calculations should be made, which- include direct radiation contributions from Unit No. 3, to determine whether the above limits of Specification 2.10 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Administrative Control 4.3, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR Part 20.2203, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190.

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

SURVEILLANCE REQUIREMENTS 2.10.2 DOSE CALCULATIONS - Annual dose contributions from liquid and gaseous effluents shall be calculated in accordance with dose calculation methodology provided for Specifications 2.4.1, and 2.8.1.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-27 2.11 REMP MONITORING PROGRAM LIMITING CONDITIONS 2.11.1 A radiological environmental monitoring program shall be provided to monitor the radiation and radionuclides in the environs of the facility. The program shall be conducted as specified in Table 2-7.

APPLICABILITY: At all times.

ACTION:

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

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

concentration (1) concentration (2)

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFS-TOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-28 2.11 REMP MONITORING PROGRAM - Continued When radionuclides other than those in Table 2-8 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is greater than or equal to the calendar year limits of Specifications 2.4 and 2.8. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Monitoring Program Report.

SURVEILLANCE REQUIREMENTS 2.11.2 The radiological environmental monitoring samples shall be collected pursuant to Table 2-7 from the "Quality Related" locations given in Tables 2-7 and 2-10 and Figures 2-1, 2-2, 2-3, 2-4 and 2-5 and shall be analyzed pursuant to the requirements of Tables 2-7 and 2-9.

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 SAFSTOR OFFSITE DOSE CALCULATION MANUAL REVISION 16 TITLE PAGE 1-29 Table 2-7 HBPP RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PROGRAM DESCRIPTION PROGRAM BASIS Exposure Pathway Number of Samples Sampling and Collection Type of Analysis ODCM State of PG&E/HBPP and/or Sample and Locations(a) Frequency Specs California Elective (QR) (NQR) (NQR)

Continuous sampler operation with Gross alpha and gross beta X X(5)

AIRBORNE 5 onsite locations. I offsite location sample collection at least once per radioactivity following filter 7 days1) change(2)

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

DIRECT RADIATION(b) 16 onsite stations, at or within the TLDs exchanged quarterly(t) Gamma exposure (3) X SITE BOUNDARY fenceline, with TLDs I offsite control station with TLD TLDs exchanged quarterlyt1 ) Gamma exposure(3) X X t X X(5) 4 offsite stations with TLDs-- TLDs exchanged quarterly( ) Gamma exposure(3) rereenting a gradient downwl~ind w,,'ith TLDI 23 i th sta .. ...... hn qo....... ............. _......X WATERBORNE Surface Water Discharge canal effluent Continuous sampler operation with Gamma.isotopic(c). Strontium-90 X X 1

sample collection weeklyM) and Tritium 2

analysis of weekly Dip samples if sampler sample( )

1 inoperable( ) Sample submitted to the State Department of Health Services monthly(1)

Groundwater 12 groundwater'spent fuel pool Quarterly Tritium-, Strontium-90. X monitoring wells Americium-241 and2 gamma isotopic(c) analysis( ) 2 Alpha and Beta Analysis( ) X Sediment 3 locations located in Humboldt Quarterly(4) Gamma isotopic(c) analysis(2) X Bay

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 REVISION 16 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL PAGE 1-30 PROGRAM DESCRIPTION PROGRAM BASIS Exposure Pathway Number of Samples Sampling and Collection Type of Analysis ODCM State of PG&E/HBPP and/or Sample and Locations(a) Frequency Specs California Elective (QR) (NQR) (NQR) 4 kc~p *alaysisf')

Gamem)a I I I [ I [

kA~gae 3 stations located in Huambldt Oualterlv. subaect to availffabi v ý Bay I 11. 11ý

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 TITLE SAFSTOR OFFSITE DOSE CALCULATION MANUAL REVISION 16 PAGE 1-31 Table 2-7 (Continued)

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

_QR) (NQR) (NQR)

INGESTION Milk Pedrotti Dairy AnnuallyM) Strontium-90 and Gamma X X isotopic(') analysis (2Gafm*-

isotpiec~ ý_

Holgerson Dairy AnnuallyM) Strontium-90 and Gamma X X isotopic(c) analysis(2,GaMMa-isetEopie -~anfl 1S`2 Fish and Invertebrates I sample of fish from Station 55 Quarterly, subject to availability(4) Gamma isotopic(c) analysis 2z) X I sample of clams from Station 59 Quarterly, subject to availability(4) Gamma isotopic(c) analysis 2z) X I sample of oysters from Station Quarterly, subject to availability(4) Gamma isotopic(c) analysis( 2) X 65 TERRESTRIAL Soil 2 locations, one near the plant and Quarterly(4) Gamma isotopic(c) analysis(2) X one from a control location Table Notations QR - Quality Related (')Performed by HBPP (3)Performed by DCPP (5)Performed by Humboldt Co. Health Dept.

NQR - Non-Quality Related (2)Performed by Offsite Laboratory (4)Performed by Humboldt State University (a) Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous.conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the quality-related sampling schedule shall be documented in the Annual Radiological Environmental Monitoring Program Report. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances, suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the REMIP, and submitted in the next Annual Radioactive Effluent Release Report, including a revised figure(s) and table for the REMP reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples for that pathway and justifying the section of the new location(s) for obtaining samples. Note: This reporting requirement applies only to the quality-related portion of the REMP.

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

1' NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFST4ORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-32 Table 2-8 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES Analysis Water (pCi/L)

H-3 20,000*

Co-60 300 Cs-137 50 1-* l

1

A 1 *1r*1 *

Jl

= xx,-+,ar- catn + a av ý

& I F ~. 1QQ Q 0

'Ur-I1'i i Yt11ULe. 11IIi[toH1utwfittgvtut be-used,

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTORISAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-33 Table 2-9 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS(a) (b)

LOWER LIMITS OF DETECTION (LLD)(c)

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

Gross Beta 4 0.01 H-3 2000(d)

Co-60 15 130 Cs-137 18 0.06 150 18 80 1.80 Table Notations (a) This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Monitoring Program Report.

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

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

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

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

Where:

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-34 Table 2-9 (Continued)

Table Notations (Continued)

/

E = the counting efficiency (as counts per transformation)

V the sample size (in units of mass or volume) 2.22 the number of transformations per minute per pico-Curie Y the fractional radiochemical yield (when applicable) k =the radioactive decay constant for the particular radionuclide At = the elapsed time between sample collection (or end of the sample collection period) and time of counting The value of Sb used in the calculation of the LLD for a detection system will be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma ray spectrometry, the background will include the typical contributions of other radionuclides.

normally present in the samples (e.g., potassium 40 in milk samples).

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

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

(d)

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-35 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) 2* 1 AO King Salmon Picnic Area W 270 0.3

-t2 A 180 Dinsmore Drive, Fortuna SSE 158 9.4 3 AE] Humboldt Hill Road at Bret Harte Lane SSE 158 0.9

-4 A Wood and K Street, Eureka NNE -42 4-0 5 0 Redwood Avenue, Arcata NE 45 12.3

-6 A Table Bluff and Clough Rea ---480 5-5

---47 A College of the Redwoods Is --- 8 26

--- 8 A Humboldt Hill Road near TV Station S9E -170 -48

-9 A 2376 Harbor View Drive &SE 15 4-6 0 A B Street, Fields Landin gSSW -200 4-

-14 A WVhittier Court & irving, Humboldt Hil~ 144S _ i5- 144

-42 A Bell Hill Road and Sauters &gW 195 0-7 "14 A South Bay School Parking Lot S 180 0.4 16 AO Elk River Road/PG&E Gas Reg/Pedrotti Dairy ENE 72 1.4

-14 A Bassford Road at Grauer's Lane E -90 2,0

---4 A 61 12 Elk River Road ESE --1p4ý 9 A 5399 Noe Avenue NE -454

-A PG&E- ' eIlI2,H4H Road E"S -- t28 04

-22 A Station B 14th Street NNE -23 4-0

-- 24 A .eat. I a-d- L Street NNE 32 5-A

-i25 A Irving Drive, Humboldt Hill SSE' 175 1.3

-q-- A 6700 Berta Road E-gS -25 4-9

-2' A 7200 Berta Road ggS -- 42 21-4

-29 A Vista Road, Humboldt Hill SgE 48 4-5

-4 A King Salmon Road East of F-reeway SS<E -14 0,

-42 A Loma Road at Tip Top Club &S-W 85- 04

-4 A King Salmon Road and RR Traci 88W -185- 0,4

-436 A Plant Entrance Road wgw --n0-2 230 5 -A Humboldt Substation (T-17) ENEB -- 6 5-.9 48 o Holgerson Dairy S 180 5.1 55 0 HBPP Outfall Canal NNW 338 0.1 56 0 1000 ft North of Outfall Canal Discharge NE 45 0.2 57 0 1000 ft South of Outfall Canal Discharge W 270 0.2 59 0 Hookton Channel SW 225 0.8 65 0 Coast Oyster Company NNE 23 4.6

A ISFSI Fenceline *

At least 4 additional TLDs are used, one in each direction, at the ISFSI Fenceline, when fuel is in storage Table Notations Code: A Dosimetry Station l , Air Particulate Station 0 Biological Station N.t.. *p'-lit. Related Station

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-36 Figure 2-1 HBPP ONSITE TLD LOCATIONS *

At least 4 additional TLDs are used, one in each direction, at the lSFSI Fenceline, when fuel is in storage I-

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFS..O.DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-37 Figure 2-2 HBPP ONSITE MONITORING WELL LOCATIONS p

MW-1 .

/1 tifl~NftJ% \'K4 -

t__ -__

SU 0iL

iiJ&-1arr-n RAILROAD

\-U. YYMIEr-S SEPARATOR INTAKE STRUCTURE LEGEND (Tide monitoring *Monitoring Well Location station)

-.*.- Apparent Groundwater Flow Direction

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-38 Figure 2-3 HBPP OFFSITE SAMPLING LOCATIONS U

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-39 Figure 2-4 HBPP OFFSITE SAMPLING LOCATIONS (CONTINUED) fr

,-(650 A.

(22 A

.. . P, ,*.

101. "

X- *;* v-7, LL*2_

, ~ . r ' 21. Mi ariuae tto

1 iloialS4to

'"*" *::I:SEQUOIA'*

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFST.OR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-40 Figure 2-5 HBPP OFFSITE SAMPLING LOCATIONS (CONTINUED)

Loleta Fortuna Arcata Eureka I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-41 2.12 REMP INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITIONS 2.12.1 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program.

APPLICABILITY: At all times.

ACTION:

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAF-TSOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-42 2.13 RADIOACTIVE WASTE INVENTORY LIMITING CONDITIONS 2.13.1 Liquid Radioactive Waste In Outdoor Tanks The radiological inventory of wastes in outdoor tanks that are not capable of.

retaining or treating tank overflows shall not exceed 0.25 Ci.

APPLICABILITY: At all times.

ACTION:

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

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

APPLICABILITY: At all times.

ACTION:

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-43 3.0 SPECIFICATION BASES 3.1 Radioactive Liquid Effluent Monitoring Instrumentation Basis The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The alarm/trip setpoints for these instruments shall be calculated in accordance with Part II of the ODCM to ensure that the alarm/trip will occur prior to exceeding 10 times the effluent concentration limits of 10 CFR Part 20 for releases to Humboldt Bay. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.

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

3.3 Liquid Effluent Concentration Basis This specification is provided to ensure that the instantaneous concentration of radioactive materials released in liquid waste effluents beyond the SITE BOUNDARY will be less than 10 times the effluent concentration limits specified in 10 CFR Part 20. The specification provides operational flexibility for releasing liquid effluents in concentrations to follow the Section II.A and II.C design objectives of Appendix I to 10 CFR 50. This limitation provides reasonable assurance that the levels of radioactive materials released to Humboldt Bay will result in exposures within (1) the Section II.A design objectives of Appendix I, 10 CFR 50, to a MEMBER OF THE PUBIC and (2) the limits of 10 CFR 20.1302 to the population. This specification does not affect the requirement to comply with the annual limitations of 10 CFR 20.1301 (a).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-44 3.4 Liquid Effluent Dose Basis This specification is provided to implement the requirements of Sections II.A. II-A and IV.A of Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A of Appendix I. The ACTION statement provides the required operating flexibility and at that same time implements the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as lowas is reasonably achievable" (ALARA). The dose calculations in the OFFSITE DOSE CALCULATION MANUAL (ODCM) implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

Compliance with this Specification has been established on a licensing basis by the SAFSTOR Environmental Report and NUREG-1 166, "Final Environmental Statement for Decommissioning Humboldt Bay Power Plant." These reports have demonstrated that routine releases of radioactive materials in effluents during SAFSTOR decommisssioning will not cause the Specification to be exceeded. As long as routine releases do not exceed

the baseline quantities evaluated in these reports, no further dose calculation is necessary.

3.5 Liquid Waste Treatment Basis The requirement that these systems be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as reasonably achievable" (ALARA). This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the liquid radwaste treatment system were selected as one quarter of the dose design objectives (on a monthly basis) set forth in Section II.A of Appendix I, 10 CFR Part 50, for liquid effluents (3 mrem/yr; 10 mrem/yr to any organ).

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 JITLE SAFSTOR/SAISTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-45 For a MEMBER OF THE PUBLIC who may at times be within the SITE BOUNDARY, the period of occupancy (which is bounded by the maximum occupational period while working in Units 1 or 2) will be sufficiently low to compensate for the reduced atmospheric dispersion of gaseous effluents relative to that for the SITE BOUNDARY.

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

The only tritium source term is the spent fuel pool water, which evaporates and is released from the stack as moisture in the air. The spent fuel pool water has a Tritium concentration below lx i0 4 micro-Curies/ml, and air at 100 'F, saturated with moisture, can not hold more than 5x10-5 grams of moisture per cc. Therefore, it is unlikely that the Tritium concentration in the gaseous effluent could exceed 5x10-9 micro-Curies/cc. This is well below the 10CFR20 Effluent Concentration Limit of 1x 10 7 micro-Curies/cc, which corresponds to a dose of 50 mrem/year, so it is not necessary to monitor for Tritium in the plant stack effluent stream.

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

3.8 Gaseous Effluents: Tritium and Radionuclides in Particulate Form Dose Basis This specification is provided to implement the requirements of Sections II.C, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluent will be kept "as low as is reasonably achievable" (ALARA). The calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated.

Compliance with this Specification has been established on a licensing basis by the SAFSTOR Environmental Report and NUREG-1 166, "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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16.

DOSE CALCULATION MANUAL PAGE 1-46 greater than 8 days) in gaseous effluents during SAFSTOR decommissioning will not cause the Specification to be exceeded. As long as routine releases do not exceed the baseline quantities evaluated in these reports, no further dose calculation is necessary.

Also, the ventilation system has since been modified to provide a full flow HEPA filtration system, significantly-reducing routine particulate stack releases.

The only tritium source term is the spent fuel pool water, which evaporates and is released from the stack as moisture in the air. The spent fuel pool water has a Tritium concentration below 1x10-4 micro-Curies/ml, and an evaporation rate less than 50 gallons per day, so the routine Tritium release rate is below 7 milli-Curies/year. Using this value, the equations in section 4.3.9 through 4.3.13 calculate a maximum annual dose of 1.08 x 10-5 milli-rem/year, so it is not necessary to calculate doses for Tritium in the plant stack effluent stream.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 16 rITLE SAFSTOR/SAFS-TO R-DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 1-47 3.9 Solid Radioactive Waste Basis This Specification ensures that radioactive wastes that are transported from the site shall meet the solidification requirements specified by the burial ground licensee of the respective states to which the radioactive material will be shipped. It also implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3.10 Total Dose Basis This specification is provided to meet the dose limitations of 40 CFR Part 190 that have now been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR part 190.11 and 10 CFR Part 20.2203a4, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190 and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Specifications 2.3, 2.4, 2.6, 2.7 and 2.8. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

3.11 REMP Monitoring Program Basis The quality-related portion of the REMP satisfies the requirements in 10 CFR Parts 20, 50, and 72.44(d) that radiological environmental monitoring programs be established to provide data on measurable levels of radiation and radioactive materials in the site environs. It is required to provide assurance that the baseline conditions established by the Environmental Report are not deteriorating and it supplements the SAFSTOR Environmental Report baseline environmental conditions by conducting onsite and offsite environmental monitoring to evaluate routine conditions during SAFSTOR decommissioning and to document any increased nuclide concentrations and/or radiation levels resulting from accidents during SA2FSTORdecommissioning.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-48 The non quality-related portion of the HBPP REMP fulfills commitments for environmental monitoring made to the state of California and conducts additional environmental monitoring which PG&E/HBPP has elected to continue from the REMP which was being implemented prior to approval of the SAFSTOR Decommissioning Plan.

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

The SAFSTOR Environmental Report, submitted to the NRC as Attachment 6 to the SAFSTOR license amendment request, established baseline conditions for soil, biota and sediments. in accordancewith the NRC appro.ved SAFSTOR Decommissioning lan-,

these baseline conditions will only need to be reestablished prior-to DECON if a significant release durfing SAFSTOR occurs as the result of an accident.

The LLD's required by Table 2-9 are considered optimum for routine environmental measurements in industrial laboratories. The LLD's for drinking water meet the requirements of 40 CFR 141.

3.12 REMP Interlaboratory Comparison Program Basis The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-49 4.0 ADMINISTRATIVE CONTROLS 4.1 Annual Radiological Environmental Monitoring Report A report on the SAESTOR 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.

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

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

d. A summary description of the SAF.ST.R.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.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-50

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

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

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 SAFSTORIDECOMMISSIONING OFFSITE REVISION 16 TITLE PAGE 1-51 DOSE CALCULATION MANUAL Table 4-1 RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

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

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

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

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

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

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

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

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

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

[N/A]

SECTION ODCM VO NUME4 NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 REVISION 16 TITLE SAFSTOR/DECOMMISSIONING OFFSITE PAGE 1-52 DOSE CALCULATION MANUAL TABLE 4-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

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

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

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

[pCi/l]------------- --------- --- -------- -------- F-LZ]---------------- [N/A]

Gross Beta 48 +/-2 Monitoring Well 16+/-3 10 +/- 3 2 (22) (9/22) No. 11 (3/6) (3/6)

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

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

A-. [j---------------------------------------NA]

Tritium 500 (15/22) 461 +/- 64 Monitoring Well 484 +/- 94 444 +/- 88 2 (22) 200 (7/22) c (7/22) No. 1 (3/5) (4/5)

..................... 409- 589] [299-601]-__

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

SECTION ODCM NUCLEAR POWER GENERATION DEPARTMENT VOLUME 4 TITLE SAFSTOR/DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-53 TABLE 4-1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING REPORT ANNUAL

SUMMARY

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

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

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

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.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-54 4.2 Annual Radioactive Effluent Release Report This report shall be submitted prior to April 1 of each year. The following information shall be included:

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

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

1. Container Volume

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

3. Principal Radionuclides (specified as measured or estimated)

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

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

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

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

e. A summary description of changes made to:

1. Process Control Program (PCP)

2. Radioactive Waste Treatment Systems

f. A complete, legible copy of the entire ODCM if any change to the ODCM was made during the reporting period. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-55 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, NRCRegion IV, within the time period specified for each report. These reports shall rbe submitted covering the activities identified below to the requirements of the applicable Specification.

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

b. Radiological Environmental Monitoring - Specification 2.11.

4.4 Major Changes to Radioactive Waste Treatment Systems

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

b. The following information shall be available for review:

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

2. Sufficient information to totally support the reason for the change,

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

4. A evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously estimated in the Environmental Report submitted to the NRC as Attachment 6 to the SAFSTOR license amendment request,

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 1-56 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 SAFSTORDecommissioning.

b. The following information shall be available for review:

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 riTLE SAFSTOR/SAFST.OR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE II-1 PART II - CALCULATIONAL METHODS AND PARAMETERS 1.0 EFFLUENT MONITOR SETPOINT CALCULATIONS 1.1 LIQUID EFFLUENT MONITORS Specification 2.1 requires that the Radioactive Liquid Effluent Monitor (RLEM) and the caisson sump monitor be set to alarm to ensure that the limits of Specification 2.3 are not exceeded (the instantaneous concentration of radioactive material released to UNRESTRICTED AREAS shall be less than or equal to 10 times the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2).

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

A = Xl0 ECLj xKxB +/-F3 (1-1) where:

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

F1 Flow rate past the RLEM.

F2 Flow rate past the caisson sump monitor.

F3 Flow rate of the effluent canal into Humboldt Bay (F 1 + F2 +

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

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 16 FITLE SAFSTOR/SAFSTIOR-DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-2 ECLc = Composite Effluent Concentration Limit (ECL) for the mix of radionuclides (micro-Ci/ml).

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

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

jCi ECLC i Y fi (1-2)

CiECLi iECLi where:

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

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

fi = Fraction of radionuclide "i" in the mixture.

1.1.3 Table 2-2 of Specification 2.1 requires that if a background reading exceeds the equivalent of 5 x 10-6 micro-Ci/ml of Cs-137, the cause will be investigated and remedial measures taken to reduce the background reading. Therefore, the maximum background allowable (Bmax, cpm) is:

Bmax = K x (5 x 106)cpm (1-3) 1.1.4 The most conservative background limit is calculated as if the calibration factor was 2.50 x 108 cpm per micro-Ci/ml (-15% tolerance). This background limit would be 1,250 cpm. It is plant policy to use a background limit (slightly lower) at 1,200 cpm to ensure that this limit is satisfied. Note that if the background setting exceeds 1,200 cpm, the monitor should be declared INOPERABLE until the background has been reduced.

1.1.5 For continuous direct caisson sump discharges, the monitor should be set to alarm at or below 7.5 times the Cs-137 ECL from 10 CFR 20, Appendix B, Table 2, column 2 (75 percent of the Specification 2.3 limit for Cs- 137), assuming no circulating water pump flow and that no liquid radwaste discharge is in progress (i.e., Equation 1-1 is solved with F1 = 0 and F 3 = F 2).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAfSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-3 1.1.6 If the Specification 2.3 alarm setting is calculated for Cs-137, -15% tolerance, no dilution and for zero background, the alarm setting would be 2,500 cpm. Because the actual mixture may have a limit that is lower than that of Cs-137, and may also provide a reduced detector' response, it is plant policy to maintain the alarm setting at or below 2,400 cpm, and to run at least one circulator during discharges, to ensure that this limit is satisfied. Refer to section 1.1.7 for the administrative (lower) alarm settings.

1.1.7 For routine liquid radwaste batch discharges, it is plant policy to set the Radioactive Liquid Effluent Monitor (RLEM) alarm no higher than necessary in order to provide protection against inadvertent releases. With at least one circulator operating, the alarm should be set according to the following table, and in any case, no higher than 25,000 cpm. The table is based approximately on the sum of twice the typical background 2 and 130% of the predicted countrate for the batch 3 .

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

(micro-Ci/ml) (micro-Ci/ml) 1.5E-05 5.9E-08 Up to 2,538 5000 1.8E-05 7.2E-08 2,538 up to 3,308 6000 2.1E-05 8.6E-08 3,308 up to 4,077 7000 2.5E-05 9.9E-08 4,077 up to 4,846 8000 2.8E-05 1.1E-07 4,846 up to 5,615 9000 3.2E-05 1.3E-07 5,615 up to 6,385 10,000 4.9E-05 1.9E-07 6,385 up to 10,231 15,000 6.6E-05 2.6E-07 10,231 up to 14,077 20,000 8.3E-05 3.3E-07 14,077 up to 17,923 25,000 t

2This table is based on a nominal background of 850 cpm. As of 2/13/07, the background reading is about 680 cpm. The extra 25% provides an allowance related to the uncertainty of reading the background.

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

I

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAF.ST.O.DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-4 1.2 GASEOUS EFFLUENT MONITOR 1.2.1 Equation C-4 of Regulatory Guide 1.109 demonstrates how to calculate dose from inhalation:

The annual dose associatedwith inhalationof all radionuclides,to organ j of an individual in age group a, is then.

D-(rO) = Rax(r,0)DFAi 1 where DQ is the annual dose rate to organ i of an individual in age group a Ra is the breathingrate for age group a xi(r,O) is the annualaverage ground-level concentrationof nuclide i in air in sector 0 at 3

distance r, in pCi/mi DFAijis the dose factor for nuclide i to organ i of age group a To calculate x_(r,0) the annual average ground-level concentration of nuclide i in air in sector 0 at distance r, in pCi/m 3 the equation must be rearranged to:

Dij(r,O)/( DFAija RI xi r(

Assuming that:

Americium-241 is the primary nuclide The maximally exposed group is the Teen based on breathing rates and DFAija The DFAiia to the bone of a Teen from Am-241 is 1.77 mrem/pCi The DFAiia are taken from: NRC NUREG-4013, "LADTAP-JJ Technical Reference and User Guide" The Teen breathing rate is 8000 m 3/year The release happens at a release rate of 30,000 cfm

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-5 3

Therefore the ground-level concentration of Am-241 in air in sector 0 at distance r. in DCi/m that will produce a dose rate of 1500 mrem/year to the bone of a Teen is:

(1500 mrem/vear) / (1.77 mrem/ICi) / (8000 m /vear) 3 1.06E-1 oCi/ m 3 33 1.06E-lpCi/m (1.06E-1 pCi/m3) / (1E6 pCi /[jCi) / (1E6 ml/mr3 ) = 1.06E-13 gCi/ml 1.2.1.1 0uantitv of radioactive material released Equation C-3 of Regulatory Guide 1.109 demonstrates how to calculate the quantity of material that must be release to produce a given airborne concentration:

The annual averaze airborneconcentrationof radionuclide i at the location (r, 0) with resnect to the release noint may be determinedas xi(r.0) = 3.17 x 10 Qi(Y/Q)D (r.0) where xi(r,0) is the annual3average ground-level concentrationof nuclide i in air in sector 0 at distance r, in pCi/rn 3.17 x 104 is the number ofpCi/Ci divided by the number of sec/yr (7/o)D(r.0) is the annual average atmosvhere dispersion factor. in sec/im 3.

0O is the release rate of nuclide Ito the atmosphere, in Ciyvr A value of 7.3E-6 sec/mr3 was used for the annual average atmosphere dispersion factor at the site boundary (,Y/O)D(r.0). This is based on a release rate of 30.000 cfm. This value is obtained from Calculation N-238C.I Rev. 0.I "Determining" Effect of HBPP Unit 3 Stack obtained from Calculation N-238C Rev. 0 "DetermininpFffectofHBP Unit3Stack Reconf7zuration on Downwind Effluent Concentrations".

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 lTLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-6 1.06E-1 pCi/m3 / ((3.17E4 (pCi/Ci) (sec/yr)) * (7.3E-6 sec/m 3))

4.61E-l Ci/yr or 4.61E5 uCi/yr 1.2.1.2 Transmission Fraction Particulate depositional losses will occur in the external transport sample lines that connect the MGP ABPM201 S alpha beta particulate monitor to the stack. These depositional losses have been calculated to determine a conservative correction factor.

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

1.2.1.3 Stack Concentration The stack concentration that would result in a release rate of4.61E-1 Ci/y is equal to:

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

The average annual stack flow rate is 30,000 cfm This results in a total volume of 4.47E14 cc/yr This is based on (30,000 ft3/min

525,600 minutes/yr

28,317 cc/ft3).

(4.61E-1 Ci

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

1.03 E-9 uCi/cc

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-7 1.2.1.4 Relationship to EPA PAG To compare the release rates calculated above the following assumptions were made:

Am-241 dose conversion factor in rem / cm 3 uCi hr, from EPA 400 = 5.3E8 A value of 3.7 1E-4 sec/mi3 was used for the atmosphere dispersion factor (X/Q)D(r,0). This value is obtained from Safstor ODCM Appendix B "Bases for Atmospheric Dispersionand Deposition Values".

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

1.03 E-9 I.Ci/cc

2 2.06E-9 gCi/cc 2.06E-9 [!Ci/cc

30,000 ft3/min

28,317 cc/ft3

60 min = 1.05E2 [ACi (1.05E2 tCi) * (5.3E8 rem / cm" uCi hr) * (3.71E-4 sec/m 3 ) / (1E6 cm 3 /m3 ) / (3600 sec/hour)

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

1.03 E-9 VCi/cc

200 = 2.06E-7 [!Ci/cc 2.06E-7 IACi/cc

30,000 ft3/min

28,317 cc/ft3

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

(2.62E3 vCi) * (5.3E8 rem / cm-3 uCi hr) * (3.7 1E-4 sec/m 3) / (1E6 cm 3/m3 ) / (3600 sec/hour) 1.43E-,1 rem This is much less than the EPA PAG of 1 Rem 6.5 Relationship to 10CFR20 Appendix B Table 2 Effluent Concentration limits The 10CFR20 Appendix B Table 2 Effluent Concentration limit for Am-241 is 2E-14 LiCi/ml.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-8 The average annual ground-level concentration in air (xi) in pCi/mr3 is equal to:

x_- (3.17E4 (pCi/Ci)/ (sec/year))

Q *(x/o)

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

ODCM average annual X10 = 7.3E-6 sec/ m If x_= 2E- 14 uCi/ml then:

o = (2E-14 uCi/ml

1E6 ml/m 3* 1E6 pCi/uCi) / ((3.17E4 (pCi/Ci)/ (sec/yr)*(7.3E-6 sec/

mi) m3 0 = 8.64E-2 Ci/yr The average annual stack volume based on the ODCM is 4.47E14 cc/yr.

This is based on (30,000 cfrn

525,600 minutes/yr

28,317 cc/cfm).

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

(8.64-2 Ci/yr

1E6 uCi/Ci) / (4.47E14 cc/yr

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

0.804 = 1.55E-10 uCi/ml 6.6 SPAM Conversion Factor from Effluent Concentration to LtCi/day The release rate in LLCi/day stack concentration in LCi/cc

30,000 ft3/mnin

1440 minutes/day

28317 cc/ ft3

transmission factor of 1.24 The release rate in gCi/day = stack concentration in VCi/cc

1.52E12 iCi/day 1.2.1.5 Conversion Factor from gCi/day to % of NUE An NUE is equal to a release rate of 3000 mrem/year

%NUE = (Offsite dose rate / NUE threshold)

100

%NUE = ((Conversion Factor

Release Rate) / NUE threshold)

100

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4' ITLE SAFSTOR1SAf.ST.R-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-9

%NUE = ((Conversion Factor

100) / NUE threshold)

Release Rate The Conversion Factor is equal to (1.77E6 mrem/uCi) * (7.3E-6 see/ m3 ) * (8000 m 3/year) /

(8.64E4 sec/day)

This is equal tol.20 mrem/vear per aCi/dav 1.2.1.6 RESULTS The 10CFR20 Appendix B Table 2 Effluent Concentration limit for Am-241 is 2E- 14 LLCi/ml. The SPAM indication that would result in a fence-line concentration of 2E-14 uCi/ml is 1.55 E-10 uCi/ml. This is approximately equal to 10% of an NUE. This value is used as the high alarm setpoint for the SPAM.

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

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

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

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

Deleted

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-10 2.0 LIQUID EFFLUENT DOSE CALCULATIONS 2.1 MONTH (31 DAY PERIOD)

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

2.2 CALENDAR QUARTER The methodology for calendar quarter calculations is the same as for the calendar year calculations provided by section 2.4, except that the resulting value for D (dose.

commitment annual rate, mrem/year) must be divided by 4 to convert it to a quarterly dose commitment, mrem/quarter.

2.3 CALENDAR YEAR The methodology for calendar year calculations is provided by section 2.4.

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR.DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE I-11 D = Y-[Ci x DF x {(BFish, i X UFish)-+-(Binvi X Ulnv)}] (2-1) i=1 --

where:

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

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

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

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

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

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

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

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

The total exposure to an organ (or whole body) is found from the summation of the contributions of each of the individual nuclides calculated. Note that the infant age group is not considered to consume either fish or other seafood, and exposure to this age group need therefore not be calculated.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFST:OR~DECOMMISSION1NG OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-12 Table 2-1 Ingestion Dose Factors for Adult Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 11 and from NUREG-4013

_ _Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.05 x 10- 7 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 Co-60 No Data 2.14 x 10-6 4.72 x 10-6 No Data No Data 4.02 x 10-5 Sr-90 7.58 x 10- 3 No Data 1.86 x 10-3 No Data No Data 2.19 x 10-4 Cs-137 7.97 x 10-5 1.09 x 10- 4 7.14 x 10- 5 3.70 x 10-5 1.23 x 10-5 2.11 x 10-6 Y-90 9.62 x 10- 9 No Data 2.58 x 10-10 No Data No Data 1.02 x 10-4 Pu-241 1.57 x 10-5 7.45 x 10-7 3.32 x 10-7 1.53 x 10-6 No Data 1.40 x 10-6 Gross ca 7.55 x 10-4 7.05 x 1 0 -4 5.41 x 10-5 4.07 x 10- 4 No Data 7.81 x 10-5 Table 2-2 Ingestion Dose Factors for Teen Age Group (mremlpico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 12 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.06 x 10- 7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 Co-60 No Data 2.81 x 10-6 6.33 x 10-6 No Data No Data 3.66 x 10-5 Sr-90 8.30 x 10-3 No Data 2.05 x 10-3 No Data No Data 2.33 x 10-4 Cs-137 1.12 x 10-4 1.49x 10-4 5.19 x 10-5 5.07 x 10-5 1.97x 10-5 2.12 x 10-6 Y-90 1.37 x 10-8 No Data 3.69 x 10-10 No Data No Data 1.13 x 10-4 Pu-241 1.75 x 10-5 8.40 x 10-7 3.69 x 10-7 1.71 x 10-6 No Data 1.48 x 10-6 Gross cc 7.98 x 10-4 7.53 x 10-4 5.75 x 10-5 4.31 x 10-4 No Data 8.28 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAF-STORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-13 Table 2-3 Ingestion Dose Factors for Child Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 13 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI 7 2.03 x 10-7 H-3 No Data 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 2.03 x 10-Co-60 No Data 5.29 x 10-6 1.56 x 10- 5 No Data No Data 2.93 x 10-5 Sr-90 1.70 x 10-2 No Data 4.31 x 10-3 No Data No Data 2.29 x 10-4 Cs-137 3.27 x 10-4 3.13 x 10-4 4.62 x 10- 5 1.02 x 10-4 3.67 x 10-5 1.96 x 10-6 Y-90 4.11 x 10-8 No Data 1.10x 10-9 No Data No Data 1.17x 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 4 8.03 x 10-5 Gross ox 1.36 x 10- 3 1.17 x 10-3 1.02 x i0- 6.23 x 10-4 No Data Table 2-4 Bioaccumulation Factors for Saltwater Environment (pCi/kg per pCi/liter)

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 iTLE SAFSTOR/SA......-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-14 Table 2-5 Average Individual Foods Consumption for Various Age Groups (kilo-gram/year or liters/year)

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

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

m-NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-15 3.0 LIQUID WASTE TREATMENT 3.1 TREATMENT REQUIREMENTS 3.1.1 ODCM Specification 2.5 Specification 2.5 requires that liquid radwaste shall be treated, as required,, to reduce radioactive materials in liquid, wastes prior to their discharge, when projected monthly doses due to liquid effluents discharged to UNRESTRICTED AREAS would exceed 0.06 mrem whole body or 0.2 mrem to any organ.

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

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

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

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

/

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAF.ST..R-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-16

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

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

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

waste receiver tanks (3)

" radwaste demineralizer

resin disposal tank

" concentrated waste tanks (2)

waste hold tanks (2)

" radwaste filters (2)

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-17

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

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

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

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

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORIS RDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-18 4.0 GASEOUS EFFLUENT DOSE CALCULATIONS 4.1 DOSE RATE 4.1.1 Deleted As explained in Specification Bases 3.7, Noble Gases are not required to be monitored, and the corresponding dose rate need not be calculated.

4.1.2 Tritium and Radioactive Particulates There are no short-lived radioactive particulates in the effluent, so radioactive decay can be neglected. Meteorological parameters are assumed to be constant, and applied for the most conservative location. Therefore, the radioactive particulates dose rate calculation methodology is the same as the radioactive particulates dose calculation methodology. Refer to sections 43.3 through 4.3.8 for the appropriate equations.

As explained in Specification Bases 3.6, Tritium is not required to be monitored, and the corresponding dose rate need not be calculated. Nevertheless, if such a calculation is required, refer to sections 4.3.9 through 4.3.13 for the appropriate equations.

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFSTORPDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-19 Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977.

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

4.3.3 Particulate Organ Dose Calculation Summation Methodology The release rate specifications for radioactive particulates with half-life greater than eight days are dependent on the existing radionuclide pathways to man, in areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these calculations were: 1) Individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leaf vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

The releases of radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents will be essentially limited to Cs-137, Co-60, and Sr-90.

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

D = [Qi X (Rnh, i + RGP, i + RMeat, i + RMilk, i + Rveg,i)] (4-3) where:

D = Annual dose commitment, mrem/year.

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-20 RMeat, i = The dose factor for the grass-cow-meat pathway for the radionuclide, i, in units of mrem/year per pico-Curie/sec.

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

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

In general, the calculations for these pathways give results that represent trivial radiation exposure. The values calculated for typical anticipated SAFSTOR 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 Particulate Inhalation Pathway Dose Calculation Methodology Rinh, = (X/Q) x BRa x DFi,a (4-3a) where:

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

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

Refer to Appendix B.

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFST.ORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-21 4.3.5 Particulate Ground Plane Pathway Dose Calculation Methodology RGP, i = (D/Q) x SF x DFi x K x W (4-3b) where:

K = unit conversion constant, 8760 hr/yr.

DFi - The ground plane dose conversion factor for radionuclide, i, in mrem/hr per pCi/m 2 from Table 4-5. No values are provided for Transuranic radionuclides, as their dose contribution to this pathway is negligible,.

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

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

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

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

W Weathering factor. This is the reciprocal of the weathering time constant given in Regulatory Guide 1.1.09, 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 ITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-22 4.3.6 Particulate Grass-Cow-Milk Pathway Dose Calculation Methodology Rmilki = (D/Q) x QF X Ua x F. x DFi'. xW 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/mn per Regulatory Guide 1.109, Table E- 15.

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

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

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

3.0 x 10.8 inverse square meters for releases from the 50 foot stack. Refer Appendix B.

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-23 4.3.7 Particulate Grass-Cow-Meat Pathway Dose Calculation Methodology QF X Ua x Ff x DFi, a X W)

RMeat, i (D/Q) x j (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/m 2 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 mremipCi, from Tables 4-8, 4-9, or 4-10, as appropriate. Note that this path is not considered to apply to the infant age group.

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

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 REVISION 16 tITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE DOSE CALCULATION MANUAL PAGE 11-24 4.3.8 Particulate Vegetation Pathway Dose Calculation Methodology Rveg, i = (D/Q) x (UT x DFia x (4-3e) where:

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

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

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

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

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

Y The agricultural productivity by unit area of pasture. This parameter is 0.7 kg/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 and fruit grown in other areas (imported to Humboldt county).

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTORISAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-25 4.3.9 Tritium Organ Dose Calculation Methodology The annual dose commitment may be calculated for any organ of an individual age group as follows:

D = QH3 X (RInh, H3 + RGP, H3 + RMeat, H3 + RMiIk, H3 + Rveg, H3) (4-4) where:

D - Annual dose commitment, mrem/year.

QH3 - The average release rate of H-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, mremlyear per pico-Curie/sec.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-26 4.3.10 Tritium Inhalation Pathway Dose Calculation Methodology Rlnh,3 = (/Q) x BRa x DFH3,a (4-4a) where:

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

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

Refer to Appendix B.

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

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

DFH3, a = The organ (or total body) inhalation dose factor for the receptor age group, a, for H-3. This is given in units of mremlpico-Curie by Tables 4-1, 4-2, 4-3, and 4-4.

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SA.ST.O.DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-27 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:

R(y*) (.57 5 05.

RMilk, H3 Q) x*. Ix 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, mrem/pico-Curie, from Tables 4-8, 4-9, 4-10, and 4-11.

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

0.75 = The fraction of total feed that is water.

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

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

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

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

Refer to Appendix B.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAf-STOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-28 4.3.12 Tritium Grass-Cow-Meat Pathway Dose Calculation Methodology RmeatH3 (Q)J (0. 7 5 H H 0x QF x Ua x FM x DFa (4-4 c)

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

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

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

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

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

0.75 The fraction of total feed that is water.

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

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

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

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

Refer to Appendix B.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE HI-29 4.3.13 Tritium Vegetation Pathway Dose Calculation Methodology The concentration of tritium is based on the airborne concentration rather than the deposition:

Rve (QH3 Q) x*f075 SH x 0.51 X UT X DFa (4-4d) where:

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

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

0.75 = The fraction of total feed that is water.

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

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

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

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

Refer to Appendix B.

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE II-30 Table 4-1 Inhalation Dose Factors for Adult Age Group (mrem/pico-Curie inhaled).

Selected Nuclides from Regulatory Guide 1.109, Table E-7 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.58 x 10-7 1.58 x 10-7 1.58 x 10-7 1.58 x 10-7 1.58 x 10-7 Co-60 No Data 1.44 x 10-6 .1.85 x 10-6 No Data 7.46 x 1 0 -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 1075 5.35 x 10-5 2.78 x 10-5 9.40 x 10-6 1.05 x 10-6 Y-90 2.61 x 10- 7 No Data 7.01 x 10- 9 No Data 2.12 x 10-5 6.32 x 10-5 Pu-241 3.42 x 10-2 8.69 x 10-3 1.29 x 10- 3 5.93 x 10-3 1.52 x 10-4 8.65 x 10- 7 Gross cc 1.68 1.13 7.75 x 10-2 5.04 x 10-1 1.82 x 10-1 4.84 x 10-5 Table 4-2 Inhalation Dose Factors for Teen Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-8 and from NUREG-4013 Organ _ _

Nuclide' Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 1.59 x 10-7 Co-60 No Data 1.89 x 10-6 2.48 x 10-6 No Data 1.09 x 10- 3 3.24 x 10-5 Sr-90 1.35 x 10-2 No Data 8.35 x 10- 4 No Data 2.06 x 10- 3 9.56 x 10-5 Cs-137 8.38 x 10-5 1.06 x 10-4 3.89 x 10-5 3.80 x 10-5 1.51 x 10-5 1.06 x 10-6 Y-90 3.73 x 10- 7 No Data 1.00 x 10-8 No Data 3.66 x 10-5 6.99 x 10-5 Pu-241 3.74 x 10-2 9.56 x 10- 3 1.40 x 10-3 6.47 x 10- 3 2.60 x 1 0 -4 9.17 x 10-7 Gross a 1.77 1.20 8.05 x 10-2 5.32 x 10-1 3.12 x 10-1 5.13 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 TITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-31 Table 4-3 Inhalation Dose Factors for Child Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E-9 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI 7 7 10- 7 3.04 x 10-7 H-3 No Data 3.04 x 10- 3.04 x 10- 3.04 x 10-7 3.04 x Co-60 No Data 3.55 x 10-6 6.12 x 10-6 No Data 1.91 x 10-3 2.60 x 10-5 Sr-90 2.73 x 10-2 No Data 1.74 x 10- 3 No Data 3.99 x 10-3 9.28 x 10-5 Cs-137 2.45 x 10- 4 2.23 x 10- 4 3.47 x 10- 5 7.63 x 10- 5 2.81 x 10- 5 9.78 x 10-7 Y-90 1.11 x 10-6 No Data 2.99 x 10-8 No Data 7.07 x 10-5 7.24 x 10-5 Pu-241 7.94 x 10-2 1.75 x 10-2 2.93 x 10-3 1.10 x 10-2 5.06 x 10-4 8.90 x 10-7 Gross cc 2.97 1.84 1.28 x 10-1 -7.63 x 10-1 6.08 x 10-1 4.98 x 10-5 Table 4-4 InhalationDose Factors for Infant Age Group (mrem/pico-Curie inhaled)

Selected Nuclides from Regulatory Guide 1.109, Table E- 10 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI 7 4.62 x 10-7 4.62 x 10-7 4.62 x 10-7 4.62 x 10-7 H-3 No Data 4.62 x 10-Co-60 No Data 5.73 x 10-6 8.41 x 10-6 No Data 3.22 x 10-3 2.28 x 10-5 Sr-90 2.92 x 10-2 No Data 1.85 x 10-3 No Data 8.03 x 10-3 9.36 x 10-5 Cs-137 3.92 x ji-4 4.37 x 10- 4 3.25 x 10-5 1.23 x 10- 4 5.09 x 10-5 9.53 x 10-7 Y-90 2.35 x 10- 6 No Data 6.30 x 10-8 No Data 1.92 x 10- 4 7.43 x 10-5 Pu-241 8.43 x 10-2 1.85 x 10-2 3.11 x 10- 3 1.15 x 10-2 7.62 x 10- 4 8.97 x 10-7 Gross cc 3.15 1.95 134 x 10-1 7.94 x 10-1 9.03 x 10-1 5.02 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFS-TOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-32 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 10-1 2 2.20 x 10- 12 Values are not provided for Transuranic radionuclides, as their dose contribution to this pathway is negligible.

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

From Regulatory Guide 1.109, 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 Groupsj (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/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-33 Table 4-8 Ingestion Dose Factors for Adult Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 11 and from NUREG-4013 Organ -

Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 1.05 x 10- 7 1.05 x 10- 7 1.05 x 10-7 1.05 x 10-7 1.05 x 10-7 Co-60 No Data 2.14 x 10-6 4.72 x 10-6 No Data No Data 4.02 x 10-5 Sr-90 7.58 x 10-3 No Data 1.86 x 10-3 No Data No Data 2.19 x 10-4 Cs-137 7.97 x 10-5 1.09 x 10 -4 7.14 x 10-5 3.70 x 10-5 1.23 x 10-5 2.11 x 10-6 Y-90 9.62 x 10-9 No Data 2.58 X 10-10 No Data No Data 1.02 x 10-4 Pu-241 1.57 x 10-5 7.45 x 10-7 3.32 x 10-7 1.53 x 10-6 No Data 1.40 x 10-6 Gross cc 7.55 x 10- 4 7.05 x 10- 4 5.41 x 10-5 4.07 x 10- 4 No Data 7.81 x i0-5 Table 4-9 Ingestion Dose Factors for Teen Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 12 and from NUREG-4013 Organ Nuclide Bone Liver Total Body. Kidney Lung GI-LLI H-3 No Data 1.06 x 10- 7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 1.06 x 10-7 Co-60 No Data 2.81 x 10-6 6.33 x 10-6 No Data No Data 3.66 x 10-5 Sr-90 8.30 x 10-3 No Data 2.05 x 10-3 No Data No Data 2.33 x 10-4 Cs-137 1.12 x 10-4 1.49 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 cc 7.98 x 10-4 7.53 x 10- 4 5.75 x 10-5 4.31 x 10-4 No Data 8.28 x 10-5

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTORISAFSTORADECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-34 Table 4-10 Ingestion Dose Factors for Child Age Group (mrem/pico-Curie ingested)

Selected Nuclides from Regulatory Guide 1.109, Table E- 13 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 2.03 x 10- 7 2.03 x 10- 7 2.03 x 10-7 2.03 x 10-7 2.03 x 10-7 Co-60 No Data 5.29 x 10-6 1.56 x .10- 5 No Data No Data 2.93 x 10-5 Sr-90 1.70 x 10-2 No Data 4.31 x 10- 3 No Data No Data 2.29 x 10-4 Cs-137 3.27 x 10-4 3.13 x 10-4 4.62 x 10-5 1.02 x 10-4 3.67 x 10-5 1.96 x 10-6 Y-90 4.11 x 10-8 No Data 1.10 x 10-9 No Data No Data 1.17 x 10-4 Pu-241 3.87 x 10-5 1.58 x 10-6 8.04 x 10-7 2.96 x 10-6 No Data 1.44 x 10-6 Gross oc 1.36 x 10-3 1.17 x 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-i14 and from NUREG-4013 Organ Nuclide Bone Liver Total Body Kidney Lung GI-LLI H-3 No Data 3.08 x 10-7 3.08 x 10-7 3.08 x 10-7 3.08 x 10-7 3.08 x 10-7 Co-60 No Data 1.08 x 10- 5 2.55 x 10- 5 No Data No Data 2.57 x 10-5 Sr-90 1.85 x 10-2 No Data 4.71 x 10-3 No Data No Data 2.31 x 10- 4 Cs-137 5.22 x 10-4 6.11 x 10- 4 4.33 x 10- 5 1.64 x 1 0 -4 6.64 x 10-5 1.91 x 10-6 Y-90 8.69 x 10-8 No Data 2.33 x 10- 9 No Data No Data 1.20 x 10-4 Pu-241 4.25 x 10-5 1.76 x 10-6 8.82 x 10-7 3.17 x 10-6 No Data 1.45 x 10-6 Gross cc 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 a

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM I VOLUME 4 TITLE SAFSTOR/SAFSTORDECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-35 Table 4-12 Stable Element Transfer Data For Cow-Milk Pathway (days/liter)

Selected Nuclides from Re~ulatorv Guide 1.109. Table E-1 and from NUREG-4013 Element Fm H 1.0 x 10-2 Co 1.0 x 10-3 Sr 8.0 x 10-4 Cs 1.2 x 10-2 Y 1.0 x 10-5 Pu 5.0 x 10-6 Gross cc 5.0 x 10-6 Table 4-13 Stable Element Transfer Data For Cow-Meat Pathway (days/kilo-gram)

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-36 5.0 URANIUM FUEL CYCLE CUMULATIVE DOSE 5.1 WHOLE BODY DOSE Specification 2.10 limits the whole body dose equivalent from the Uranium fuel to no more than 25 mrem/year. The whole body dose is determined by summing the calculated doses from the following:

a. Deleted

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

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

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

To this calculated exposure is added potential direct radiation exposure to an individual at the site boundary. The only portion of the site boundary where there is significant direct radiation is near the radwaste facilities at the [PG&E] North edge of the site. Due to the possibility that an individual at the shoreline (fishing, bird watching, etc.) may use the path at the brow of the cliff for access, the TLD stations along the path are used to estimate an annual radiation exposure. The time period used for this estimate is 67 hour7.75463e-4 days 0.0186 hours 1.107804e-4 weeks 2.54935e-5 months s/year, given by Table E-5 of Regulatory Guide 1.109, as the maximum time for shoreline recreation for the Teen age group.

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

a. Deleted

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 FITLE SAFSTOR/SAFSTIOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-37 5.3 DOSE TO OTHER ORGANS Specification 2.10 limits the dose to any organ (thyroid excepted) to less than or equal to 25 mrem/year. The dose to any individual other than skin organ is determined by summing the calculated doses from the following:

a. Deleted

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

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-38 6.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE REQUIRING SOLIDIFICATION 6.1 SCOPE This section pertains to radioactive waste containing a total specific activity which exceeds the burial ground criteria for solidification, or which exceeds the concentration limits for Class A waste as defined in 10 CFR 61. These wastes must be stabilized by solidification and contain no freestanding liquids prior to shipment offsite for land burial,. or else be packaged in a high integrity container in accordance with Section 7.0.

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

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

6.2.2 This vendor service shall include transmittal to HBPP of copies of their solidification procedure and PCP prior to performing the solidification.

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 ITLE SAFSTOR/SAFST4OR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-39 7.0 PROCESS CONTROL PROGRAM FOR RADIOACTIVE WASTE PACKAGED IN HIGH INTEGRITY CONTAINERS 7.1 SCOPE This section pertains to radioactive waste containing specific activity which exceeds the burial ground criteria for solidification, or which exceeds the concentration limits for Class A wasfe as defined in 10 CFR 61. These wastes must be stabilized by packaging in dewatered form in a high-integrity container which meets burial ground and regulatory requirements, or else be solidified in accordance with Section 6.0.

7.2 PROGRAM ELEMENTS For land burial disposal of radioactive waste requiring a high-integrity container, HBPP shall implement the following steps:

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

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 tITLE SAFSTOR/SAFST4OR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-40 8.0 PROCESS CONTROL PROGRAM FOR LOW ACTIVITY DEWATERED RESINS AND OTHER WET WASTES 8.1 SCOPE This section pertains to bead-type spent radioactive demineralizer resin and other wet wastes shipped for land burial which contain a total specific activity less than the burial ground criteria for solidification, and which, does not exceed the concentration limits for Class A waste as defined in 10 CFR 61.

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

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

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

NUCLEAR POWER GENERATION DEPARTMENT SECTION ODCM VOLUME 4 rITLE SAFSTOR/SAFSTOR-DECOMMISSIONING OFFSITE REVISION 16 DOSE CALCULATION MANUAL PAGE 11-41 9.0 PROGRAM CHANGES 9.1 PURPOSE OF THE OFFSITE DOSE CALCULATION MANUAL The Offsite Dose Calculation Manual was developed to support the implementation of the Radiological Effluent Technical Specifications required by 10 CFR 50, Appendix I, and 10 CFR 50.36. The purpose of the manual is to provide the NRC with sufficient information relative to effluent monitor setpoint calculations, effluent related dose calculations, and environmental monitoring to demonstrate compliance with radiological effluent controls.

9.2 CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL It is recognized that changes to the ODCM may be required during the SAPFSTOR Decommissioning period. All changes shall be reviewed and approved by the PSRC and the Plant Manager prior to implementation. The NRC shall be informed of all changes to the ODCM by providing a description of the change(s) in the first Annual Radioactive Effluent Release Report following the date the change became effective. Records of the reviews performed on change to the ODCM should be documented and retained for the duration of the possession only license.

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

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

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

CTS-291 (Section II, 9.3)

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

ODCM APPENDIX A Revision 16 Page A-I APPENDIX A SAFSTOR BASELINE CONDITIONS rýl

ODCM APPENDIX A Revision 16 Page A-2 1.0 LIQUID AND GASEOUS EFFLUENTS 1.1 LIQUID EFFLUENTS Baseline levels of radioactive materials contained in liquid effluents during the SAFSTOR period were established in the Environmental Report submitted as Attachment 6 to the SAFSTOR license amendment request. These values are presented for cumulative annual release and average monthly discharge in Table A-1.

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

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

Tritium 8.6E-2 7.2E-3 Principal Gamma Emitters (total) 1.85E-1 1.54E-2 Strontium-90 3.28E-4 2.73E-5 Table A-2 Baseline Gaseous Effluent Activity Type of Activity Annual Release Monthly Average Release (Curies) (Curies)

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

ODCM APPENDIX B Revision 16 Page B-1 APPENDIX B BASES FOR ATMOSPHERIC DISPERSION AND DEPOSITION VALUES

ODCM APPENDIX B Revision 16 Page B-2 1.0 BASIS FOR DISPERSION/DEPOSITION VALUES - 50' STACK 1.1 The instantaneous atmospheric dispersion factor (X/Q) is taken from meteorological parameter calculations performed to evaluate reducing the height of the Unit 3 stack. The calculation report is number N238C, Revision 0, titled "Determine Effect of Humboldt Bay Power Plant Unit 3 Stack Reconfiguration on Downwind Effluent Concentrations".

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

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

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) of the calculation (N238C) provides annual maximum values for X/Q for the 50' stack for various stack flow rates, based on an NRC model named "XOQDOQ". The annual average X/Q value used in the' ODCM (1.00 x 10-5) is based on a stack flow of 25,000 cfm.

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

This calculation is microfilmed (with calculations N238A & N238B), at microfilm reel/frame location (RLOC) 07175-4939 thru 5359. Table 1 (frame number 5140) of the calculation (N238C) provides annual maximum values for D/Q for the 50' stack for various stack flow rates, based on an NRC model named "XOQDOQ". The annual average D/Q value used in the ODCM (3.00 x 108) is based on a stack flow of 25,000 cfm.

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

ODCM APPENDIX B Revision 16 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)

U10(7cyyy,+ A/2)

X/Q 1 (2) 1 X/Q (3)

U1071 Zy oz where:

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

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

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

A vertical cross-sectional area of structures, equal to 375 meters 2 , based on the Refueling Building dimensions (about 36 feet high, about 112 feet long).

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

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

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

ODCM APPENDIX B Revision 16 Page B-4 X/Q = 1.32 x 10-2 seconds/meter 3 , (2)

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

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

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

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

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

2.2.3 For the Particulate Ground Plane Pathway, the assumed distance from the emission point to the potential receptor for this calculation is 150 meters. This is the approximate distance to publicly accessible areas from the structure with the most significant potential for airborne radioactivity (i.e. from the center of the Refueling Building to the trail at the'edge of the bluff). At this distance, Figure 6 provides a Relative Deposition Rate value of 1.4 x 10-4 meter 1 . The plume width assumed for this calculation is the same as was used in equation 3 of section 2.1.4 (above), so that the plume width is approximately 6cy. For ay equal to 4.33 meters (Pasquill Class G at a distance of 150 meters), D/Q is (1.4 x 1 0 -4 meter-l)

(6 x 4.33 meter) = 5.39 x 10-6 meter2.

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

ODCM APPENDIX B Revision 16 Page B-5 Figure 6 provides a Relative Deposition Rate value of 1.2 x 10-4 meter 1 . The plume width assumed for this calculation is the same as was used in6 equation 3 of section 2.1.4 (above), with the plume width of approximately 67y., but at a greater distance. For Ty equal to 6.07 meters (Pasquill Class G at a distance of 220 meters), D/Q is (1.2 x 1 0 4 meter')/ (6 x 6.07 meter) = 3.29 x 10-6 meter-2 .

ODCM APPENDIX C Revision 16 Page C-I APPENDIX C Deleted

HBL-10-011, Annual Radioactive Effluent Release Report for 2009

Contents

Text

Dear Commissioners and Staff:

SUMMARY

SUMMARY

SUMMARY

Navigation menu

HBL-10-011, Annual Radioactive Effluent Release Report for 2009

Text

Dear Commissioners and Staff:

SUMMARY

SUMMARY

SUMMARY

Navigation menu

Search