Text

2010 Annual Radioactive Effluent Release Report, in Accordance with 10 CFR 50.36a(a)(2) and Section 5.6.3
	ML11133A245
Person / Time
Site:	Diablo Canyon
Issue date:	04/28/2011
From:	Becker J; Pacific Gas & Electric Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	DCL-11-049
	Download: ML11133A245 (285)
	v • d • e

Pacific Gas and Electric Company' James R. Becker Diablo Canyon Power Plant Site Vice President Mail Code 104/5/601 P 0. Box 56 Avila Beach, CA 93424 805.545.3462 Internal: 691.3462 April 28, 2011 Fax: 805.545.6445 PG&E Letter DCL-1 1-049 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 2010 Annual Radioactive Effluent Release Report

Dear Commissioners and Staff:

Pacific Gas and Electric Company is submitting the enclosed 2010 Annual Radioactive Effluent Release Report in accordance with 10 CFR 50.36a(a)(2) and Section 5.6.3 of the Diablo Canyon Power Plant Technical Specifications. N The report describes the quantities of radioactive gaseous and liquid effluents released from the plant, and the solid radioactive waste shipments made during the period of January 1 through December 31, 2010.

One compact disk is enclosed with the report. The disk contains meteorological data. If you have any questions, please contact Jeff Gardner at (805) 545-4385.

swh/64037107 Enclosure cc: Diablo Distribution cc/enc: Edgar D. Bailey, DHS Roger W. Briggs, Executive Officer, CRWQCB Elmo E. Collins, Regional Administrator, Region IV William A. Nestel, INPO Michael S. Peck, NRC Senior Resident Inspector Gregory Thomas, MD, San Luis Obispo County Health Officer Jim Polickoski, NRR Project Manager Alan B. Wang, NRR Project Manager A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Cattaway e Comanche Peak ° Diablo Canyon

Palo Verde

South Texas Project e Wolf Creek

DIABLO CANYON POWER PLANT 2010 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT Iktl&l I January 1 - December 31, 2010

Enclosure PG&E Letter DCL-1 1-049 DIABLO CANYON POWER PLANT Annual Radioactive Effluent Release Report January 1, 2009 Through December 31, 2010 Table of Contents Introduction 1 I. Supplemental Information 2 II. Major Changes to Liquid, Gaseous, and Solid Radwaste Systems 9 Ill. Changes to the Offsite Dose Calculation Manual 9 IV. Land Use Census 11 V. Gaseous and Liquid Effluents 11 Table 1- Gaseous Effluents - Summation of All Releases 12 Table 2- Gaseous Effluents - Ground-Level Releases 15 Table 3- Gaseous Effluents - Lower Limits of Detection 19 Table 4- Gaseous Effluents - Summation of All Releases 21 Table 5- Gaseous Effluents - Nuclides Releases 24 Table 6- Liquid Effluents - Lower Limit of Detection (LLD) 30 VI. Solid Waste Shipments 33 VII. Radiation Dose due to Gaseous and Liquid Effluents 35 Table 7 - Radiation Dose Due to the Release of Radioactive Liquid 40 Effluents Table 8 - Radiation Dose Due to the Release of Radioactive Gaseous 41 Effluents Table 9 - Percent of Technical Specification Limits for Radioactive Liquid 44 Effluents Table 10 - Percent of Technical Specification Limits for Radioactive 45 Gaseous Effluents Table 11 - Onsite Dose to Members of the Public 47 VII. Meteorological Data 52 i

Enclosure PG&E Letter DCL-1 1-049 DIABLO CANYON POWER PLANT Annual Radioactive Effluent Release Report January 1, 2010 Through December 31, 2010 Attachments

1. Diablo Canyon Power Plant Program Directive, CY 2, "Radiological Monitoring and Controls Program," Revision 7

2. Diablo Canyon Power Plant Interdepartmental Administrative Procedure (IDAP),

CY2.ID1, "Radioactive Effluent Controls Program," Revision 11

3. IDAP RPI.ID11, "Environmental Radiological Monitoring Procedure," Revision 10

4. Diablo Canyon Power Plant Chemical Analysis Procedure, CAP A-8, "Off-Site Dose Calculations," Revision 34

5. Diablo Canyon Power Plant Chemical Analysis Procedure, CAP A-8, "Off-Site Dose Calculations," Revision 35

6. Diablo Canyon Power Plant Administrative Procedure, RP2.DC2, "Radwaste Solidification Process Control Program," Revision 15

7. 2010 Land Use Census ii

Enclosure PG&E Letter DCL-1 1-049 DIABLO CANYON POWER PLANT 2010 Annual Radioactive Effluent Release Report Introduction The 2010 Annual Radioactive Effluent Release Report summarizes gaseous and liquid effluent releases from Diablo Canyon Power Plant's (DCPP) Units 1 and 2. The report includes the dose due to release of radioactive liquid and gaseous effluents and summarizes solid radwaste shipments. The report contains information required by Units 1 and 2 Technical Specification (TS) 5.6.3 and is presented in the general format of 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 Nuclear Power Plants," Appendix B, "Effluent and Waste Disposal Report."

Procedure revisions, which implement the Off-Site Dose Calculation Manual, and one compact disk containing meteorological data, are attached.

In all cases, the plant effluent releases were well below TS limits for the report period.

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Enclosure PG&E Letter DCL-1 1-049 Supplemental Information A. Regulatory Limits

1. Gaseous Effluents

a. Noble Gas Dose Rate Limit The dose rate in areas at or beyond the site boundary due to radioactive noble gases released in gaseous effluents is limited to less than or equal to 500 millirem (mR) per year to the total body and less than or equal to 3000 mR per year to the skin. (Radioactive Effluent Controls Program [RECP],

Attachment 6)

b. Particulate and Iodine Dose Rate Limit The dose rate in areas at or beyond the site boundary due to iodine-131, iodine-1 33, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents, is limited to less than or equal to 1500 mR per year to any organ. (RECP Attachment 6)

c. Noble Gas Dose Limit The air dose due to noble gases released in gaseous effluents from each reactor unit to areas at or beyond the site boundary is limited to the following:

Radiation Type Calendar Quarter Calendar Year Limit RECP Limit Attachment 7 Attachment 7 Gamma 5 millirad 10 millirad Beta 10 millirad 20 millirad

d. Particulate and Iodine Dose Limit The dose to an individual from iodine-1 31, iodine-1 33, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released from each reactor unit to areas at or beyond the site boundary is limited to less than or equal to 7.5 mR to any organ in any calendar quarter and less than or equal to 15 mR to any organ during a calendar year. (RECP Attachment 8) 2

Enclosure PG&E Letter DCL-1 1-049

2. Liquid Effluents

a. Concentration The concentration of radioactive material released from the site is limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration is limited to 2 x 10-4 microcuries/milliliter (gCi/ml) total activity.

(RECP Attachment 3)

b. Dose The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released from each reactor unit to areas at or beyond the site boundary is limited to the following:

Dose Type Calendar Quarter Limit Calendar Year Limit RECP Attachment 4 RECP Attachment 4 Total Body 1.5 millirem 3 millirem Any Organ 5 millirem 10 millirem B. Maximum Permissible Concentrations

1. Gaseous Effluents Maximum permissible concentrations are not used for determining allowable release rates for gaseous effluents at DCPP.

2. Liquid Effluents The concentrations listed in 10 CFR 20, Appendix B, Table 2, Column 2, for radionuclides other than dissolved or entrained noble gases are used for determining the allowable release concentration at the point of discharge from the site for liquid effluents. For dissolved or entrained noble gases, the allowable release concentration at the point of discharge is limited to 2 x 10.4 [tCi/ml total activity for liquid effluents.

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Enclosure PG&E Letter DCL-1 1-049 C. Measurements and Approximations of Total Radioactivity

1. Gaseous Effluents

a. Fission and Activation Gases A pair of off-line monitors equipped with beta scintillator detectors monitors the gaseous radioactivity released from the plant vent. The monitor readings are correlated to isotopic concentration based on laboratory isotopic analysis of grab samples using a germanium detector.

For plant vent noble gas releases, grab sample results are used to quantify releases. The individual batch release data are used to quantify the radioactivity discharged from the gas decay tanks and containment.

A noble gas grab sample is obtained and analyzed at least weekly. The isotopic mixture is assumed to remain constant between grab sample analyses.

Containment purges, gas decay tank releases, and air ejector discharges are released via the plant vent.

The gaseous radioactivity released from the steam generator blowdown tank vent is measured by analyzing liquid or steam condensate grab samples with a germanium detector.

A factor R, a ratio of unit masses between water flashing to steam and water entering the tank, is used to calculate the activity. The isotopic concentrations are assumed to remain constant between grab samples.

Other potential pathways for releasing gaseous radioactivity are periodically monitored by collecting grab samples and analyzing these samples with a germanium detector system.

b. lodines Radioiodines released from the plant vent are monitored by continuous sample collection on silver zeolite cartridges.

The cartridges are changed at least weekly and analyzed with a germanium detector. The radioiodine releases are averaged over the period of cartridge sample collection.

Other potential pathways for releasing radioiodines are periodically monitored by collecting samples using charcoal 4

Enclosure PG&E Letter DCL-1 1-049 or silver zeolite cartridges and analyzing with a germanium detector.

Radioactive materials in particulate form released from the plant vent are monitored by continuous sample collection on particulate filters. The filters are changed at least weekly and analyzed with a germanium detector. The particulate radioactivity is averaged over the period of particulate filter sample collection. Each filter is analyzed for alpha emitters using an internal proportional counter. Plant vent particulate filters collected during a quarter are used for the composite analysis for strontium-89 and -90, which is counted on an internal proportional counter after chemical separation.

Other potential pathways for releasing radioactive particulate are periodically monitored by collecting samples using particulate filters and analyzing these filters with a germanium detector.

c. Tritium Tritium released from the plant vent is monitored by passing a measured volume of plant vent sample through a water column and determining the tritium increase in the water. An aliquot of the water is counted in a liquid scintillation spectrometer. The minimum routine sample frequency for tritium is weekly. The tritium concentration is assumed to remain constant between samples.

d. Estimations of Overall Error Sources of error considered for batch release are:

1) calibration source; 2) calibration counting; 3) sampling;

4) sample counting; and 5) gas decay tank pressure gauge/containment exhaust fan flow rate.

Sources of error for continuous release are: 1) calibration source; 2) calibration counting; 3) sampling; 4) sample counting; 5) process monitor (RE-14) reading (fission gases only); and 6) plant vent exhaust fan flow rate.

2 Total error ( 2 2 1 + a2 + + ... a+ i) 1/2 Where a* = error associated with each component.

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Enclosure PG&E Letter DCL-1 1-049

2. Liquid Effluents

a. Batch Releases Each tank of liquid radwaste is analyzed for principal gamma emitters using a germanium detector prior to release. A monthly prerelease analysis includes dissolved and entrained gases. Volume proportional monthly and quarterly composites are prepared from aliquots of each tank volume discharged. The monthly composite is analyzed for tritium using a liquid scintillation spectrometer and gross alpha radioactivity using an internal proportional counter. The quarterly composite is analyzed for iron-55 using a liquid scintillation spectrometer and for strontium-89 and -90 using an internal proportional detector following chemical separations. The monthly composite for discharges from the auxiliary building are also analyzed for nickel-63, uranium-233/234/235/236/238 and plutonium-238/239/240/241/242.

b. Continuous Releases For the continuous liquid releases of the steam generator blowdown tank and turbine building sump oily water separator, daily grab samples are collected and aliquots are proportioned for weekly, monthly, and quarterly composites.

The oily water separator weekly composite is analyzed for principal gamma emitters using a germanium detector. The steam generator blowdown tank weekly composite is analyzed for principal gamma emitters and iodine-131.

The steam generator blowdown tank monthly composite is analyzed for tritium using a liquid scintillation spectrometer and for gross alpha using an internal proportional counter.

The steam generator blowdown tank quarterly composite is analyzed for iron-55 using a liquid scintillation spectrometer and for strontium-89 and strontium-90 using an internal proportional counter following chemical separation. The results for each of the composites are averaged over the period of the composite.

In addition, one grab sample of the steam generator blowdown tank is analyzed monthly for dissolved and entrained gases using a germanium detector. The results of 6

Enclosure PG&E Letter DCL-1 1-049 this analysis are assumed to remain constant over the period of one month.

A grab sample of the steam generator blowdown is collected at least weekly and analyzed for gamma activity using a germanium detector. This analysis is used to monitor activity, however, is not used in effluent calculations unless a significant change is detected.

Note on dilution volume:

Tables 4A, 4B and 4C, "Liquid Effluents - Summation of All Releases," Item F., lists the, "Volume of circulating saltwater used during release periods," in liters. This value is calculated by multiplying the discharge duration by the circulating water flow rate. The values listed in the Tables are the summation of the circulating water discharge volume calculated for each individual batch and continuous discharge period. Therefore, in the case where two or more simultaneous discharges into the same circulating water are occurring, the calculated volume of circulating water is duplicated, and therefore the sum of the dilution volumes for the batch releases and continuous releases are greater than the actual dilution volume since each discharge incorporates the circulating discharge flow rate in its own dose calculation.

c. Estimation of Overall Error Sources of error considered are: 1) calibration source error;

2) calibration counting error; 3) sampling error; 4) sample counting error; and 5) volume of waste release error.

These sources of error are independent; therefore the total error is calculated according to the following formula:

2 2 Total error = (a 2

1 + G2a + C2 3 + ... a- i) 1/2 Where a* = error associated with each component.

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Enclosure PG&E Letter DCL-1 1-049 D. Batch Releases

1. Liquid

a. Number of batch releases ........................................... 608

b. Total time period for batch releases ........................ 3630 hours0.042 days 1.008 hours 0.006 weeks 0.00138 months

c. Maximum time period for a batch release ................. 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months

d. Average time period for a batch release ..................... 5.97 hours0.00112 days 0.0269 hours 1.603836e-4 weeks 3.69085e-5 months

e. Minimum time period for a batch release .................... 0.33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months

f. Average saltwater flow during batch releases ............. 5.72E+05 gallons per minute (gpm)

2. Gaseous

a. Number of batch releases ....................................... 124

b. Total time period for batch releases ........................... 772 hours0.00894 days 0.214 hours 0.00128 weeks 2.93746e-4 months

c. Maximum time period for a batch release .................. 72.00 hours0 days 0 hours 0 weeks 0 months

d. Average time period for a batch release .................. 6.22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months

e. Minimum time period for a batch release ................. 0.43 hours4.976852e-4 days 0.0119 hours 7.109788e-5 weeks 1.63615e-5 months E. Abnormal Release (Gaseous and Liquid)

No abnormal releases occurred during the reporting period.

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Enclosure PG&E Letter DCL-1 1-049 Ih. Major Changes to Liquid, Gaseous and Solid Radwaste Treatment System There were no major changes to liquid, gaseous, and solid radwaste treatment systems during the reporting period.

III. Changes to The Offsite Dose Calculation Manual (ODCM)

The Diablo Canyon ODCM is made up of the following procedures:

Nuclear Power Generation Program Directive Procedure, CY2, "Radiological Monitoring and Controls Program" (RMCP)

" Nuclear Power Generation Interdepartmental Administrative Procedure (IDAP), CY2.ID1, "Radioactive Effluent Controls Program" (RECP)

IDAP RP1.ID1 1, "Environmental Radiological Monitoring Procedure (ERMP)

" Diablo Canyon Power Plant Chemical Analysis Procedure, CAP A-8, "Offsite Dose Calculation Procedure" (ODCP)

Changes made to these procedures during the reporting period are described below. A copy of each revision made during the reporting period is included as an attachment.

If no changes were made to a particular procedure during the reporting period, the most recent revision is included as an attachment in order to provide a complete, current copy of the ODCM used during the reporting period.

Also included is a copy of Diablo Canyon Power Plant Administrative Procedure, RP2.DC2, "Radwaste Solidification Process Control Program," (PCP). No changes were made to this procedure in 2010.

A. Changes to the RMCP

1. The editorial correction reported in the 2009 report was permanently incorporated as revision 7.

2. This incorporation was completed on 1/21/10. See Attachment 1.

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Enclosure PG&E Letter DCL-1 1-049 B. Changes to the RECP

1. Appendix 6, Table 6, "Radioactive Gaseous Waste Sampling and Analysis Program," contains superscript (2). This superscript requires that, in addition to routine monthly grab sampling for noble gas, that a grab sample also be collected following:

i. startup ii. shutdown iii. THERMAL POWER change exceeding 15%.

The terms "startup" and "shutdown" were not previously defined. This revision of the procedure now includes a definition (Section 3) to make interpretation of these terms more consistent.

2. An industry review was performed concerning the superscript (2) discussed above. Most PWRs do not require the compensatory action in the superscript unless:

i. Analysis shows that the DOSE EQUIVALENT 1-131 concentration in the reactor coolant has increased more than a factor of 3; or ii. The noble gas monitor shows that effluent activity has increased more than a factor of 3.

These two conditions were included in the procedure.

3. Revision 11 was approved by the Station Director on 12/14/09, and implemented on 1/20/10. See Attachment 2.

C. Changes to ERMP

1. Removed a requirement to report groundwater protection initiative (GPI) related spills in the annual radiological environmental operating report (AREOR). This information will now be reported in the annual radioactive effluents release report (ARERR).

2. Air sampler availability reporting is now included in the AREOR.

3. Added requirements to report GPI monitoring results in the AREOR.

4. Added reference to air sampling stations 1S1 and 8S2.

5. Added groundwater well and sample point 8S3.

6. Changed to outer ring station maximum distance to 14 km to capture location 6D1.

7. Revision 10 was approved by the Station Director, and implemented, on 11/08/10. See Attachment 3.

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Enclosure PG&E Letter DCL-1 1-049 D. Changes to the ODCP

1. This procedure was revised twice during 2010.

2. Revision 34:

i. An editorial correction was implemented to correct a typographical error for a conversion factor:

"1.14x105" was changed to "1.14E+05."

ii. The editorial correction was implemented on 5/21/10. See attachment 4

3. Revision 35:

i. A routine update of meteorological dispersion (X/Q) and deposition (D/Q) values was made to Table 10.2.

ii. Revision 35 was approved by the station director on 10/13/10, and implemented on 10/22/10. See attachment 5 E. No changes were made to the PCP

1. See attachment 6 IV. Land Use Census Changes to the Land Use Census Program are included as Attachment 7.

V. Gaseous and Liquid Effluents Tables 1 through 3 describe gaseous effluents. Tables 4 through 6 describe liquid effluents.

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Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE IA GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES Units UIts First Quarter 1 Second IQuarter Est.Total IIError % I A. Fission and activation gases

1. Total release Ci 2.38E-2 2.52E-2 24% i

2. Average release rate for period ýiCi/sec 3.06E-3 3.21 E-3

3. Percent of technical specification limit 1 % 2.81 E-5 2.87E-5 B. lodines

1. Total iodine-131 Ci MDA MDA 24%

2. Average release rate for period jiCi/sec MDA MDA

3. Percent of technical specification limit1 % MDA MDA C. Particulates

1. Particulates with half-lives >8 days Ci MDA MDA 24%

2. Average release rate for period pCi/sec MDA MDA

3. Percent of technical specification limit1 % MDA MDA

4. Gross alpha radioactivity Ci MDA MDA D. Tritium,

1. Total release Ci 3.26E+1 2.23E+1 13 9

2. Average release rate for period ý.iCi/sec 4.19E+Q 2.84E+O

3. Percent of technical specification limit 1 % 9.30E-6 6.30E-6 MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

RECP 6.1.6.1 Limit 12

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 1B GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES I I Third Fourth [EI otai Units Quarter __Quarter Error %

A. Fission and activation gases

1. Total release Ci 3.38E-2 1.18E+0 24%

2. Average release rate for period p.Ci/sec 4.25E-3 1.48E-1

3. Percent of technical specification limit1 % 3.87E-5 1.12E-3 B. lodines

1. Total iodine-131 Ci MDA MDA 24%

2. Average release rate for period pCi/sec MDA MDA

3. Percent of technical specification limit' % MDA MDA C. Particulates

1. Particulates with half-lives >8 days Ci MDA 1.07E-5 24%

2. Average release rate for period gCi/sec MDA 1.34E-6

3. Percent of technical specification limit1 % MDA 5.17E-7

4. Gross alpha radioactivity Ci 2.28E-7 4.83E-8 D. Tritium

1. Total release Ci 2.19E+1 7.63E+1 13%

2. Average release rate for period ptCi/sec 2.76E+0 9.60E+0

3. Percent of technical specification limit % 6.12E-6 2.13E-5 MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

1RECP 6.1.6.1 Limit 13

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 1C GASEOUS EFFLUENTS - SUMMATION OF ALL RELEASES FUnits Annual Totalj A. Fission and activation gases

1. Total release Ci 1.26E+O

2. Average release rate for period piCi/sec 4.OOE-2

3. Percent of technical specification limit' % 3.06E-4 B. lodines

1. Total iodine-131 Ci MDA

2. Average release rate for period [tCi/sec MDA

3. Percent of technical specification limit 1 % MDA C. Particulates

1. Particulates with half-lives >8 days Ci 1.07E-5

2. Average release rate for period p.Ci/sec 3.39E-7

3. Percent of technical specification limit 1 % 1.30E-7

4. Gross alpha radioactivity Ci 2.76E-7 D. Tritium

1. Total release Ci 1.53E+2

2. Average release rate for period giCi/sec 4.86E+0

3. Percent of technical specification limit' % 1.08E-5 1RECP 6.1.6.1 Limit 14

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 2A GASEOUS EFFLUENTS - GROUND LEVEL RELEASES

1. Fission gases argon-41 Ci MDA 2.38E-2 MDA 2.45E-2 krypton-85 Ci MDA MDA MDA MDA krypton-85m Ci MDA MDA MDA MDA krypton-87 Ci MDA MDA MDA MDA krypton-88 Ci MDA MDA MDA MDA xenon-1 31 m Ci MDA MDA MDA MDA xenon-133 Ci MDA MDA MDA 6.99E-4 xenon-1 33m Ci MDA MDA MDA MDA xenon-1 35 Ci MDA MDA MDA MDA xenon-135m Ci MDA MDA MDA MDA xenon-1 38 Ci MDA MDA MDA MDA TOTAL FOR PERIOD Ci MDA 2.38E-2 MDA 2.52E-2

2. lodines iodine-131 Ci MDA MDA iodine-133 Ci MDA MDA iodine-1 35 Ci MDA MDA TOTAL FOR PERIOD Ci MDA MDA MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

15

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 2A (Continued)

GASEOUS EFFLUENTS - GROUND LEVEL RELEASES

3. Particulates barium-1 40 Ci MDA MDA cesium-134 Ci MDA MDA cesium-1 37 Ci MDA MDA cerium-141 Ci MDA MDA cerium-144 Ci MDA MDA chromium-51 Ci MDA MDA cobalt-57 Ci MDA MDA cobalt-58 Ci MDA MDA cobalt-60 Ci MDA MDA iron-59 Ci MDA MDA lanthanum-140 Ci MDA MDA manganese-54 Ci MDA MDA molybdenum-99 Ci MDA MDA ruthenium-1 03 Ci MDA MDA strontium-89 Ci MDA MDA strontium-90 Ci MDA MDA zinc-65 Ci MDA MDA zirconium-95 Ci MDA MDA TOTAL FOR PERIOD Ci MDA MDA MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume).

This note applies to all tables.

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Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 2B GASEOUS EFFLUENTS - GROUND LEVEL RELEASES

1. Fission gases argon-41 Ci MDA 3.34E-2 MDA 9.61 E-1 krypton-85 Ci MDA MDA MDA 5.86E-2 krypton-85m Ci MDA MDA MDA MDA krypton-87 Ci MDA MDA MDA MDA krypton-88 Ci MDA MDA MDA MDA xenon-1 31 m Ci MDA MDA MDA 2.27E-4 xenon-133 Ci MDA 3.48E-4 MDA 1.59E-1 xenon-133m Ci MDA MDA MDA MDA xenon-135 Ci MDA MDA MDA MDA xenon-135m Ci MDA MDA MDA MDA xenon-138 Ci MDA MDA MDA MDA TOTAL FOR PERIOD Ci MDA 3.38E-2 MDA 1.18E+0

2. lodines iodine-1 31 Ci MDA MDA iodine-1 33 Ci MDA MDA iodine-135 Ci MDA MDA TOTAL FOR PERIOD Ci MDA MDA MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

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Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 2B (Continued)

GASEOUS EFFLUENTS - GROUND LEVEL RELEASES

3. Particulates barium-140 Ci MDA MDA cesium-134 Ci MDA MDA cesium-1 37 Ci MDA MDA cerium-141 Ci MDA MDA cerium-144 Ci MDA MDA chromium-51 Ci MDA MDA cobalt-57 Ci MDA MDA cobalt-58 Ci MDA 1.07E-5 cobalt-60 Ci MDA MDA iron-59 Ci MDA MDA lanthanum-140 Ci MDA MDA manganese-54 Ci MDA MDA molybdenum-99 Ci MDA MDA ruthenium-103 Ci MDA MDA strontium-89 Ci MDA MDA strontium-90 Ci MDA MDA zinc-65 Ci MDA MDA zirconium-95 Ci MDA MDA TOTAL FOR PERIOD Ci MDA 1.07E-5 MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume).

This note applies to all tables.

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Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 3 GASEOUS EFFLUENTS - LOWER LIMITS OF DETECTION

1. Fission gases argon-41 vCi/ml 1.38E-8 1.38E-8 1.38E-8 krypton-85 viCi/ml 2.90E-6 2.90E-6 2.90E-6 krypton-85m viCi/ml 8.34E-8 8.34E-8 8.34E-8 krypton-87 vCi/ml 3.56E-8 3.56E-8 3.56E-8 krypton-88 ýtCi/ml 3.46E-8 3.46E-8 3.46E-8 xenon-131m ptCi/ml 2.73E-7 2.73E-7 2.73E-7 xenon-133 vCi/mI 1.84E-8 1.84E-8 1.84E-8 xenon-1 33m pCi/mI 4.63E-8 4.63E-8 4.63E-8 xenon-135 p.Ci/mi 7.72E-9 7.72E-9 7.72E-9 xenon-135m ý.Ci/ml 7.28E-7 7.28E-7 7.28E-7 xenon-138 piCi/ml 2.50E-7 2.50E-7 2.50E-7

2. Tritium Ihydrogen-3 ICi/mI 4.95E-9 I 4.95E-9 N/A

3. lodines iodine-131 JpCi/ml 6.86E-13 N/A iodine-i 33 pCi/ml 1.28E-12 N/A iodine-135 pCi/ml 3.58E- 11 N/A 19

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 3 (Continued)

GASEOUS EFFLUENTS - LOWER LIMITS OF DETECTION Nuclide Units Continuous Mode

4. Particulates barium-140 PCi/ml 2.92E-12 cesium-134 pCi/ml 4.45E-13 cesium-1 37 pCi/ml 5.33E-13 cerium-141 ýICi/ml 4.26E-1 3 cerium-144 pCi/ml 2.26E-12 chromium-51 pCi/ml 3.14E-12 cobalt-57 pCi/ml 2.50E-13 cobalt-58 pCi/ml 4.82E-1 3 cobalt-60 p.Ci/ml 6.50E-13 iron-59 pCi/ml 1.13E-12 lanthanum-140 PLCi/ml 1.02E-12 manganese-54 gCi/mI 7.21 E-13 molybdenum-99 PCi/ml 3.27E-13 ruthenium-1 03 p.Ci/ml 4.49E-13 strontium-89 PCi/ml 4.88E-1 6 strontium-90 pCi/ml 4.33E-15 zinco65 PCi/ml 1.10E-12 zirconium-95 ltCi/ml 8.91 E-13 gross alpha PCi/ml 1.84E-15 20

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 4A LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES its First Second Est.Total Quarter Quarter Error %

A. Fission and activation products

1. Total release (not including tritium, gases, alpha) Ci 6.52E-3 2.40E-3 24%

2. Average diluted concentration during period ýiCi/ml 2.66E-12 8.99E-13

3. Percent of applicable limit 1 % 2.23E-5 7.98E-6 B. Tritium

1. Total release Ci 3.74E+2 1.25E+2 13%

2. Average diluted concentration during period ýiCi/ml 1.53E-7 4.67E-8

3. Percent of applicable limit 1 % 1.53E-2 4.67E-3 C. Dissolved and entrained gasses

1. Total release Ci 8.30E-5 2.83E-5 24%

2. Average diluted concentration during period pCi/ml 3.39E-14 1.06E-14

3. Percent of applicable limit 1 % 1.69E-8 5.30E-9 D. Gross Alpha 1.Total release Ci MDA MDA 61%

E. PVolume of waste release (prior to dilution)

I liters I 7.70E+7 I 8.16E+7 I 5% I F. IVolume of circulating saltwater used during Irelease periods I liters I 2.45E+12I 2.67E+12I 7% 1 MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

1RECP 6.1.6.1 Limit 21

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 4B LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES I ts Third Fourth Est.Total Units Quarter Quarter Error %

A. Fission and activation products

1. Total release (not including tritium, gases, alpha) Ci 5.85E-3 3.38E-2 _24%

2. Average diluted concentration during period g.Ci/ml 2.07E-12 1.44E-1 1

3. Percent of applicable limit 1 % 1.35E-5 1.05E-4 B. Tritium

1. Total release Ci 7.74E+2 9.03E+2 13%

2. Average diluted concentration during period gCi/ml 2.73E-7 3.86E-7

3. Percent of applicable limit1 % 2.73E-2 3.86E-2 C. Dissolved and entrained gasses

1. Total release Ci 4.74E-4 2.16E-3 24%

2. Average diluted concentration during period gCi/ml 1.67E-13 9.24E-13

3. Percent of applicable limit 1 % 8.37E-8 4.62E-7 D. Gross Alpha

1. Total release Ci MDA MDA 61%

E. IVolume of waste release (prior to dilution) 5% 1 I titers I 8.50E+7 I 8.04E+7 I F. IVolume of circulating saltwater used during Irelease periods I liters I 2.83E+12I 2.34E+12I 7% 1 MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

RECP 6.1.6.1 Limit 22

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 4C LIQUID EFFLUENTS - SUMMATION OF ALL RELEASES

[ Units Annual Total A. Fission and activation products

1. Total release (not including tritium, gases, alpha) Ci 4.86E-02

2. Average diluted concentration during period 4Ci/ml 4.72E-12

3. Percent of applicable limit 1 % 3.50E-5 B. Tritium

1. Total release Ci 2.18E+3

2. Average diluted concentration during period liCi/ml 2.11 E-7

3. Percent of applicable limit' % 2.11E-2 C. Dissolved and entrained gasses

1. Total release Ci 2.75E-3

2. Average diluted concentration during period ýtCi/ml 2.67E-13

3. Percent of applicable limit1 % 1.33E-7 D. Gross Alpha

1. Total release Ci MDA E. PVolume of waste release (prior to dilution)

I liters I 3.24E+8 I

JVolume of circulating saltwater used during liters 1.03E+13 F.

release periods I I I MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume).

This note applies to all tables.

1 RECP 6.1.6.1 Limit 23

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 5A LIQUID EFFLUENTS - NUCLIDES RELEASED First Quarter Second Quarter Continuous Continuous BthMd Nuclides Released Units o Batch Mode Batch Mode Mode Mode antimony-122 Ci MDA MDA MDA MDA antimony-124 Ci MDA MDA MDA MDA antimony-125 Ci MDA 6.04E-4 MDA 2.18E-5 barium-1 39 Ci MDA MDA MDA MDA barium-140 Ci MDA MDA MDA MDA beryllium-7 Ci MDA MDA MDA MDA bromine-82 Ci MDA MDA MDA MDA cerium-141 Ci MDA MDA MDA MDA cerium-143 Ci MDA MDA MDA MDA cerium-144 Ci MDA MDA MDA MDA cesium-134 Ci MDA MDA MDA MDA cesium-1 36 Ci MDA MDA MDA MDA cesium-137 Ci MDA 4.23E-6 MDA 2.47E-6 chromium-51 Ci MDA MDA MDA MDA cobalt-57 Ci MDA 1.16E-5 MDA MDA cobalt-58 Ci MDA 1.53E-3 MDA 2.35E-4 cobalt-60 Ci MDA 1.21 E-3 MDA 5.35E-4 iron-55 Ci MDA 2.44E-3 MDA 7.05E-4 iron-59 Ci MDA 1.60E-5 MDA MDA lanthanum-140 Ci MDA MDA MDA MDA MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

24

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 5A (CONTINUED)

LIQUID EFFLUENTS - NUCLIDES RELEASED First Quarter Second Quarter Continuous Cniuu Nuclides Released Units Batch Mode Continuous Batch Mode Mode Mode manganese-54 Ci MDA 2.52E-5 MDA 5.05E-7 molybdenum-99 Ci MDA MDA MDA MDA nickel-63 Ci MDA 5.96E-4 MDA 8.83E-4 niobium-95 Ci MDA 2.75E-5 MDA MDA niobium-97 Ci MDA MDA MDA MDA neodymium-147 Ci MDA MDA MDA MDA plutonium-238 Ci MDA MDA MDA MDA plutonium-239/240 Ci MDA MDA MDA MDA plutonium-241 Ci MDA MDA MDA MDA plutonium-242 Ci MDA MDA MDA MDA ruthenium-103 Ci MDA MDA MDA MDA silver-11 Oin Ci MDA MDA MDA MDA sodium-24 Ci MDA MDA MDA MDA strontium-89 Ci MDA MDA MDA MDA strontium-90 Ci MDA MDA MDA MDA strontium-91 Ci MDA MDA MDA MDA tellurium-125m Ci MDA MDA MDA MDA tellurium-1 29 Ci MDA MDA MDA MDA tellurium-129m Ci MDA MDA MDA MDA MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

25

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 5A (CONTINUED)

LIQUID EFFLUENTS - NUCLIDES RELEASED First Quarter Second Quarter Continuous Continuous BthMd Nuclides Released Units o Batch Mode Batch Mode Mode Mode tellurium-132 Ci MDA MDA MDA MDA tin-1 13 Ci MDA MDA MDA MDA tin-1l17m Ci MDA MDA MDA MDA tungsten-1 87 Ci MDA MDA MDA MDA uranium-233/234 Ci MDA MDA MDA MDA uranium-235/236 Ci MDA MDA MDA MDA uranium-238 Ci MDA MDA MDA MDA zinc-65 Ci MDA 4.67E-5 MDA 1.89E-5 zirconium-95 Ci MDA 9.24E-6 MDA MDA iodine-131 Ci MDA MDA MDA MDA iodine-1 33 Ci MDA MDA MDA MDA iodine-1 35 Ci MDA MDA MDA MDA TOTAL FOR PERIOD Ci MDA 6.52E-3 MDA 2.40E-3 DISSOLVED AND ENTRAINED GASES xenon-1 33 Ci MDA 8.30E-5 MDA 2.83E-5 xenon-1 33m Ci MDA MDA MDA MDA xenon-135 Ci MDA MDA MDA MDA krypton-85 Ci MDA MDA MDA MDA TOTAL FOR PERIOD Ci MDA 8.30E-5 MDA 2.83E-5 MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

26

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT

ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 5B LIQUID EFFLUENTS - NUCLIDES RELEASED 9

Third Quarter Fourth Quarter Continuous Cniuu Nuclides Released Units Batch Mode Continuous Batch Mode Mode Mode antimony-1 22 Ci MDA MDA MDA MDA antimony-124 Ci MDA MDA MDA MDA antimony-1 25 Ci MDA 6.61 E-5 MDA 7.48E-4 barium-1 39 Ci MDA MDA MDA MDA barium-140 Ci MDA MDA MDA MDA beryllium-7 Ci MDA MDA MDA MDA bromine-82 Ci MDA MDA MDA MDA cerium-1 41 Ci MDA MDA MDA MDA cerium-143 Ci MDA MDA MDA MDA cerium-144 Ci MDA MDA MDA MDA cesium-1 34 Ci MDA 6.21E-6 MDA 1.33E-4 cesium-136 Ci MDA MDA MDA MDA cesium-137 Ci MDA 2.70E-5 MDA 3.83E-4 chromium-51 Ci MDA MDA MDA 2.16E-3 cobalt-57 Ci MDA 1.58E-6 MDA 5.18E-6 cobalt-58 Ci MDA 2.01 E-4 MDA 3.06E-3 cobalt-60 Ci MDA 8.40E-4 MDA 4.19E-3 iron-55 Ci MDA 1.10E-3 MDA 2.75E-3 iron-59 Ci MDA MDA MDA 2.91E-5 lanthanum-140 Ci MDA MDA MDA MDA MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

27

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 5B (CONTINUED)

LIQUID EFFLUENTS - NUCLIDES RELEASED Third Quarter Fourth Quarter Continuous BotthnMode Nuclides Released Units Batch Mode Batch Mode Mode Mode manganese-54 Ci MDA 9.48E-8 MDA 1.96E-5 molybdenum-99 Ci MDA 1.42E-6 MDA MDA nickel-63 Ci MDA 3.56E-3 MDA 2.02E-2 niobium-95 Ci MDA 4.55E-7 MDA 8.98E-5 niobium-97 Ci MDA MDA MDA MDA neodymium-147 Ci MDA MDA MDA MDA plutonium-238 Ci MDA MDA MDA MDA plutonium-239/240 Ci MDA MDA MDA MDA plutonium-241 Ci MDA MDA MDA MDA plutonium-242 Ci MDA MDA MDA MDA ruthenium-103 Ci MDA MDA MDA MDA silver-1 I0m Ci MDA MDA MDA MDA sodium-24 Ci MDA MDA MDA MDA strontium-89 Ci MDA MDA MDA MDA strontium-90 Ci MDA MDA MDA 5.50E-5 strontium-91 Ci MDA MDA MDA MDA tellurium-125m Ci MDA MDA MDA MDA tellurium-129 Ci MDA MDA MDA MDA tellurium-129m Ci MDA MDA MDA MDA MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This note applies to all tables.

28

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 5B (CONTINUED)

LIQUID EFFLUENTS - NUCLIDES RELEASED Third Quarter Fourth Quarter Continuous Cniuu Nuclides Released Units Batch Mode Continuous Batch Mode Mode Mode tellurium-132 Ci MDA MDA MDA MDA tin-1 13 Ci MDA MDA MDA MDA tin-1 17m Ci MDA MDA MDA MDA tungsten-187 Ci MDA MDA MDA MDA uranium-233/234 Ci MDA MDA MDA MDA uranium-235/236 Ci MDA MDA MDA MDA uranium-238 Ci MDA MDA MDA MDA zinc-65 Ci MDA 4.83E-5 MDA 2.85E-5 zirconium-95 Ci MDA 3.93E-7 MDA 1.74E-5 iodine-131 Ci MDA MDA MDA MDA iodine-133 Ci MDA MDA MDA MDA iodine-135 Ci MDA MDA MDA MDA TOTAL FOR PERIOD Ci MDA 5.85E-3 MDA 3.38E-2 DISSOLVED AND ENTRAINED GASES xenon-133 Ci MDA 4.74E-4 MDA 2.13E-3 xenon-133m Ci MDA MDA MDA MDA xenon-135 Ci MDA MDA MDA 2.97E-5 krypton-85 Ci MDA MDA MDA MDA TOTAL FOR PERIOD Ci MDA 4.74E-4 MDA 2.16E-3 MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit mass or volume). This noteapplies to all tables.

29

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT NUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 6 LIQUID EFFLUENTS - LOWER LIMITS OF DETECTION Nuclide Units LLD antimony-122 gCi/ml 1.47E-7 antimony-1 24 gCi/ml 8.96E-8 antimony-125 [tCi/ml 2.84E-7 barium-140 gCi/ml 4.71 E-7 beryllium-7 pCi/ml 9.98E-7 bromine-82 gCi/ml 1.87E-7 cerium-141 gCi/ml 1.17E-7 cerium- 143 p*Ci/ml 3.71 E-7 cerium-144 [1Ci/ml 5.75E-7 cesium-1 34 gCi/mI 9.81 E-8 cesium-1 36 pCi/mI 1.33E-8 cesium-1 37 gCi/ml 1.27E-8 chromium-51 gCi/ml 8.12E-7 cobalt-57 gCi/ml 7.92E-8 cobalt-58 pCi/ml 1.15E+6 cobalt-60 ýICi/ml 1.31 E-7 iron-55 [LCi/ml 9.06E-7 iron-59 11Ci/ml 2.69E-7 lanthanum-140 pCi/ml 2.53E-7 manganese-54 gCi/ml 1.40E-7 molybdenum-99 pCi/ml 8.20E-8 nickel-63 pCi/ml 8.61 E-8 niobium-95 p.Ci/ml 1.16E-7 30

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 6 (CONTINUED)

LIQUID EFFLUENTS - LOWER LIMITS OF DETECTION Nuclide Units LLD neodymium-147 pLci/ml 2.64E-7 plutonium-238 !ICi/ml 3.54E-8 plutonium-239/240 ýLCi/ml 3.06E-8 plutonium-241 p.Ci/ml 3.12E-8 plutonium-242 ýLCi/ml 1.91 E-8 ruthenium- 103 jiCi/ml 1.01 E-7 silver- 11Dm jiCi/ml 9.98E-8 sodium-24 piCi/ml 5.47E-7 strontium-89 ACi/ml 4.61 E-8 strontium-90 pCi/ml 4.11 E-8 strontium-91 pICi/ml 9.40E-7 tellurium-125m p.Ci/ml 2.27E-5 tellurium-129m paCi/ml 3.75E-6 tellurium- 132 pCi/ml 8.91 E-8 tin-1 13 pCi/ml 1.58E-7 tin-1i17m jiCi/ml 7.82E-8 tungsten-1 87 4iCi/ml 6.08E-7 uranium-233/234 pCi/ml 4.91 E-8 uranium-235/236 pC i/ml 4.42E-8 uranium-238 p.Ci/ml 4.13E-8 zinc-65 Aci/ml 2.56E-7 zirconium-95 pLCi/ml 2.34E-7 gross alpha pCi/ml 9.92E-8 hydrogen-3 gCi/ml 4.89E-6 31

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 6 (CONTINUED)

LIQUID EFFLUENTS - LOWER LIMITS OF DETECTION Nuclide Units LLD iodine-1 31 gCi/ml 9.43E-8 iodine-133 piCi/ml 2.1OE-7 iodine-135 ýiCi/ml 5.80E-6 xenon-1 33 pCi/ml 3.64E-7 xenon-1 33m 11Ci/ml 8.58E-7 xenon-135 piCi/ml 4.09E-7 krypton-85 11Ci/ml 2.66E-5 32

Enclosure PG&E Letter DCL-1 1-049 VI. Solid Radwaste Shipments Solid Waste and Irradiated Fuel Shipment A. Solid Waste Shipped Off-site for Burial or Disposal (Not irradiated fuel)

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

a. Spent Resins, Filter Sludges, m3 5.65E+00 Evaporator Bottoms, etc. Ci 1.05E+01 9.OOE+0

b. Dry Compressible Waste, m3 4.74 E+00 Contaminated Equipment, etc. Ci 1.07E+00 9.OOE+0

c. Irradiated Components, Control m3 8.83E-01 Rods, etc. Ci 1.68E+00 0.OOE+0

d. Other m3 0.OOE+0 Ci O.OOE+0 O.OOE+O

2. Estimate of Major Nuclide Composition (by type of waste)

a. H-3 % 5.07E+01 Ni-63 % 2.29E+01 Fe-55 % 1.51E+01 Co-60 % 8.63E+00 Zn-65 % 3.91 E-01

b. Fe-55 % 3.80E+01 Co-58 % 2.44E+01 Co-60 % 2.21 E+01 Ni-63 % 8.1OE+00 Zn-65 % 2.31 E+00

c. I Not Applicable I % I N/A I

d. I Not Applicable I I N/A I 33

Enclosure PG&E Letter DCL-1 1-049 Solid Waste and Irradiated Fuel Shipment (Continued)

3. Solid Waste Disposition Number of Shipments Mode of Destination Transportation 4 Truck Clive, UT 2 Rail Clive, UT

4. Supplemental Information Required by former TS 6.9.1.6 Solidification Type of Container Number of 10 CFR 61 Shipping Type Agent Containers Waste Class None IPi 1 A IP2 - LSAII None IPi 9 A IP1 - LSA B. Irradiated Fuel Shipments (Disposition) 34

Enclosure PG&E Letter DCL-1 1-049 VII. Radiation Dose Due to Gaseous and Liquid Effluents Radiation Doses A. Radiation Doses from Radioactive Liquid Effluents The radiation dose contributions due to releases of radioactive liquid effluents to the total body and each individual organ for the maximum exposed adult have been calculated in accordance with the methodology in the ODCP. Dose contributions listed in Table 7 show conformance to RECP Attachment 4.

B. Radiation Doses from Radioactive Gaseous Effluents

1. The radiation dose contributions due to radioactive gaseous effluents at the site boundary for the land sectors have been calculated in accordance with the calculation methodology in the ODCP. Each unit's dose contribution has been calculated separately. The latest five-year historical average meteorology conditions were used in these calculations. In addition to the site boundary doses, the dose to an individual (critical receptor) due to radioiodines, tritium, and particulates released in gaseous effluents with half-lives greater than eight days is determined in accordance with the methodology in the ODCP based on the methodology described in NUREG-0133. Dose contributions listed in Table 8, which represents the maximum dose for age groups, organs, and geographic locations for the report period, show conformance to RECP Attachments 6, 7 and 8.

2. The radiation dose contribution from locations (2), other than the plant vent, are based upon calculation only (no direct sampling), and reported separately here.

a. Radiation Doses from Chemistry Laboratory Radioactive Gaseous Effluents -

Closest Site Boundary (800m)

i. The radiation dose due to the primary chemistry laboratory radioactive gaseous effluents for the report period is evaluated to be:

1. Total gamma air dose = 4.17E-06 mrad

2. Total beta air dose = 1.53E-06 mrad

b. Radiation Doses from Post Accident Sampling System Radioactive Gaseous Effluents - Closest Site Boundary (800m)

i. The radiation doses due to post accident sampling system radioactive gaseous effluents for the report period is evaluated to be:

1. Total gamma air dose = 4.95E-07 mrad

2. Total beta air dose = 1.80E-07 mrad 35

Enclosure PG&E Letter DCL-1 1-049

3. Radiation Dose from Radioactive Gaseous Effluents to Individual Due to Consumption of Grazing animals on Property Surrounding the Site.

The Land Use Census identified that during 2010, ranchers in the area around the plant slaughtered goat, sheep, deer, wild pig and cattle for personal consumption. As part of the DCPP Radiological Environmental Monitoring Program (REMP), samples of cow, goat and sheep meat were analyzed for radioactivity. Results of those analyses are available in the 2010 REMP report.

Based upon the isotopes (other than C-14) discharged in gaseous form during 2010, the maximum calculated dose due to these identified meat pathways is 6.53E-03 mrem.

4. Atmospheric Release Of Carbon-14

a. Historically, C-14 has not been considered a "principle radionuclide" in terms of its concentration in radioactive effluents and its dose consequence to offsite receptors. Naturally occurring C-14 is prevalent in the environment. In addition, nuclear weapons testing in the 1950s and 1960s significantly increased the amount of C-14 in the atmosphere. Therefore, power plants typically did not analyze for, nor report information for C-14 in annual reports.

b. In June, 2009, the NRC revised its guidance in Regulatory Guide (RG) 1.21, "Measuring, Evaluating and Reporting Radioactivity In Solid Wastes And Releases Of Radioactive Materials In Liquid And Gaseous Effluents From Light-Water-Cooled Nuclear Power Plants", Revision 2. This documents states that analytical methods for determining C-14 have improved. Coincidentally, the radioactive effluents from commercial power plants have decreased to the point that C-14 is likely to be a principle radionuclide in gaseous effluents. Also, because the dose contribution of C-14 from liquid radioactive waste is much less than that contributed by gaseous radioactive waste, evaluation of C-14 in liquid radioactive waste is not required.

c. As a results, the NRC states that licensees should evaluate whether C-14 is a principal radionuclide for gaseous releases from their facility.

d. RG 1.21 goes on to explain, in part:

i. The quantity of C-14 discharged can be estimated by sample measurements or by use of a normalized C-14 source term and scaling factors based on power generation, or estimated by use of the GALE code from NUREG-0017.

ii. If sampling is performed, the sampling frequency may be adjusted to that interval that allows adequate measurement and reporting of effluents.

iii. If estimating C-14 based on scaling factors and fission rates, a precise and detailed evaluation of C-14 is not necessary. It is not necessary to calculate uncertainties for C-14 or to include C-14 uncertainty in any subsequent calculation of overall uncertainty.

36

Enclosure PG&E Letter DCL-1 1-049

e. In order to aid the nuclear power industry in estimating C-14 production, the Electric Power Research Institute (EPRI) developed 2010 Technical Report 1021106, "Estimation of Carbon-14 in Nuclear Power Plant Gaseous Effluents," December 2010. This is the guidance Diablo Canyon will use to estimate the production and release quantities of C-14.

f. Carbon 14 calculated production, discharge parameters and resulting dose are reported here, separately from tables 1, 2, 8, 10 and 11 Calculated C-14 production, per Unit 12.0 Ci/EFPY(1 )

2010 Unit capacity factors U1 = 0.89 U2 = 1.00 Fraction release of produced C-14 0.98 Fraction of C-14 chemical form assumed Organic = 0.70 Inorganic = 0.30 Curies C-14 released to atmosphere U2 = 10.5 Critical receptor dose(2) [Adult (bone)], mrem 3.66E-01

("Effective Full Power Year (2)Due to ingestion of meat (cattle) assumed to graze up to the site boundary in the NW sector

g. Summary:

i. The quantity of C-14 released to the atmosphere for 2010 was Unit 1 = 10.5 curies Unit 2 = 11.8 curies ii. The resulting critical receptor dose is:

Adult (bone), due to ingestion of meat = 3.66E-01 mrem.

37

Enclosure PG&E Letter DCL-1 1-049 C. Radiation Doses from Direct Radiation (Line-of-Sight Plus Sky-Shine) -

Closest Site Boundary

1. For the report period, the radiation dose from the following areas is evaluated to be 1.98E-01 mR:

radioactive waste containers outside of plant buildings

the storage of contaminated tools and equipment inside plant buildings

Old Steam Generator Storage Facility, which was loaded with Unit 2 old steam generators for the first time in 2008. In 2009, the Unit 1 old steam generators and the Unit 2 old reactor head were loaded into the facility.

In 2010, the U1 old reactor head was loaded into the facility

Independent Spent Fuel Storage Installation (ISFSI).

The Diablo Canyon ISFSI received its first of eight casks of spent fuel starting on June 23, 2009. The eighth and final cask for 2009 was received by ISFSI in August. Eight additional casks were received during 2010.

The nearest resident to the site is approximately 1.5 miles away. However, doses from direct radiation are calculated at the site boundary. The occupancy time at the site boundary is assumed to be 2,080 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months , based upon a 40-hour work week and 52 weeks per year.

D. 40 CFR 190 Considerations The release of radioactivity in liquid and gaseous effluents resulted in doses that are small percentages of the TS limits as shown in Tables 9 and 10.

In addition, the direct radiation from various sources, including the ISFSI, resulted in doses that are a small percentage of 40 CFR 190 limits.

The total dose from liquid and gaseous effluents, and direct radiation, shows conformance to 40 CFR 190.

E. Radiation Doses from Radioactive Liquid and Gaseous Effluents to Members of the Public Due To Their Activities Inside The Site Boundary

1. Liquid Effluents The radiation dose to members of the public within the site boundary due to the release of radioactive liquid effluents is negligible because the discharge piping for liquid radwaste is mostly imbedded in concrete, located in remote or inaccessible areas, or is underground. In addition, the quantity of radioactivity released was very low.

2. Gaseous Effluents The radiation dose to members of the public within the site boundary due to the release of radioactive gaseous effluents are listed in Table 11.

38

Enclosure PG&E Letter DCL-1 1-049 G. Sub-surface water radioactive contamination Recent industry events have identified equipment leaks containing low levels of radioactivity, resulting in contaminated ground water and storm water radioactivity concentrations that may leave the plant sites. Suspected plant equipment leaks that could result in such an event at Diablo Canyon are documented in the corrective action program (CAP). Analysis of samples of sub-surface water at Diablo Canyon has indicated the presence of tritium. At this time, we have no reason to point to plant system leakage as the source for this tritium.

Sampling and analysis of the Auxiliary building roof drains, Auxiliary building drywell and the containment structure observation wells have identified detectable tritium activity. This tritium is most likely coming from the rain wash-out of gaseous tritium exiting the plant vent system. This tritium is being accounted for in the plant vent release, and rain water tritium is being accounted for per plant approved procedures.

This practice will continue until such time that new industry guidance is provided to account for this pathway in a different manner.

Refer to the 2010 REMP report for the data that has been evaluated to date for the sampling locations.

H. NEI 07-07 Voluntary Communication for sub-surface water Section 2, "Communication," Objective 2.4, "Annual Reporting," of NEI 07-07 lists the acceptance criteria for annual reporting of "ground water" information, as follows:

1. Acceptance Criterion 2.4.b.i:

Reporting of on-site ground water sample results that are taken in support of the GPI but are not part of the REMP program are reported in the ARERR required by 10 CFR 50.36a(a)(2).

For 2010, there were no on-site ground water results for samples taken in support of the GPI that are not part of the REMP program.

2. Acceptance Criterion 2.4.b.ii:

Reporting of on-site ground water "sample results for those longterm monitoring sample points that are not included in REMP are reported in the ARERR."

For 2010, there were no on-site ground water sample results for longterm monitoring sample points not included in the REMP.

3. Acceptance Criterion 2.4.c.i Voluntary communications of the description of all spills or leaks, made per Objective 2.2 acceptance criterion, shall be included in the ARERR.

For 2010, no voluntary communications were made per Objective 2.2.

4. Acceptance Criterion 2.4.c.ii:

Voluntary communications of all on-site or off-site ground water sample results that exceeded the REMP reporting thresholds as described in the ODCM that were communicated per Objective 2.2 acceptance criterion b shall be included in either the ARERR and/or in the ARREOR.

For 2010, no voluntary communications were made per Objective 2.2.

39

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 7 RADIATION DOSE DUE TO THE RELEASE OF RADIOACTIVE LIQUID EFFLUENTS millirem First Quarter Second Quarter Third Quarter Fourth Quarter Annual Total Total Body 5.16E-05 1.74E-05 7.89E-05 1.69E-04 3.17E-04 Bone 8.50E-05 4.23E-05 1.23E-04 8.39E-04 1.09E-03 Liver 9.24E-05 3.OOE-05 1.03E-04 2.49E-04 4.74E-04 Thyroid 3.11 E-05 1.02E-05 6.36E-05 1.07E-04 2.12E-04 Kidney 3.92E-05 1,35E-05 7.21E-05 1.13E-04 2.38E-04 Lung 5.67E-05 1.73E-05 7.47E-05 1.48E-04 2.96E-04 G.I. LLI 1.20E-04 3.13E-05 9.98E-05 3.70E-04 6.21E-04 40

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 8A RADIATION DOSE 1 DUE TO THE RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS (UNIT 1)

I FirstDose Quarter I

Second Quarter Dosc Third Quarter Dose I Fourth Quarter Dose I Annual Total Dose S ite Boundary NobieGa [ I T I I Gamma Air Dose mrad 1.60E-5 1.65E-5 2.27E-5 1.40E-3 1.46E-3 Beta Air Dose mrad 5.65E-6 5.81 E-6 8.05E-6 5.18E-4 5.38E-4 First Quarter o [Second Quarter Third Quarterf Fourth Quarter Annual Total Nearest Residence-NNW

_ _ _ _Dose j Dose Dose j Dose Dose I IPImI I3 Critical Receptor (Highest Organ) I mrem 2.18E-4 1.24E-4 1.27E-4 9.49E-4 1.42E-3 1

Second Quarter [2AThird Quarter

[Nearest First Quarter Fourth Quarter Annual Total

- ESE "

Vegetable Garden I CriticalReceptor T (Highest Organ) mrem I 1.66E-4 I 9.26E-5 1.00E-4 7.47E-4 1.11E-3 41

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 8B RADIATION DOSE1 DUE TO THE RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS (UNIT 2)

Site Boundary F First Quarter FrtIDose Second Quarter D

Third Quarter Dose I Fourth Quarter Dose I Annual Total

ýDose Noble Gas 2 66 - 2.2 E-Gamma Air Dose mrad 1.90E-5 1.98E-5 2.66E-5 2.29E-5 8.84E-5 Beta Air Dose mrad 6.71 E-6 7.07E-6 9.40E-6 2.73E-5 5.05E-5 First Quarter [Second Quarter [Third Quarterf Fourth Quarter Annual Total FrtjDose D Dose Dose 77 Dose Nearest Residence-NNW Critical Receptor (Highest Organ) mrem 2.54E-4 1.99E-4 1.91 E-4 1.65E-4 8.09E-4 First Quarter Second Quarter Third Quarter Fourth Quarter Annual Total Dose I Dose Dose Dose Dose Nearest Vegetable Garden - ESE

,P Tm2'4 Critical Receptor (Highest Organ) mrem 1.90E-4 1.58E-4 1.51 E-4 1.30E-4 6.29E-4 42

Enclosure PG&E Letter DCL-1 1-049 Notes for Tables 8A and 8B

1. This represents the maximum dose of age groups, organs, and geographic locations for the quarter and the year.

2. Radioiodines, radioactive material in particulate form, and radionuclides other than noble gases with half-lives greater than eight days.

3. The inhalation and ground plane pathways are included for this location.

4. The inhalation, ground plane and vegetable pathways are included for this location. An occupancy factor of 0.5 was used for the inhalation and ground plane pathways. The teen age group had the highest calculated dose for this location.

43

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 9 PERCENT OF TECHNICAL SPECIFICATION LIMITS' FOR RADIOACTIVE LIQUID EFFLUENTS Percent Y I First Quarter Second Quarter Third Quarter Fourth Quarter Annual Total Total Body 3.44E-03 1.16E-03 5.26E-03 1.13E-02 1.06E-02 Bone 1.70E-03 8.46E-04 2.46E-03 1.68E-02 1.09E-02 Liver 1.85E-03 5.99E-04 2.07E-03 4.97E-03 4.74E-03 Thyroid 6.22E-04 2.05E-04 1.27E-03 2.14E-03 2.12E-03 Kidney 7.85E-04 2.71 E-04 1.44E-03 2.27E-03 2.38E-03 Lung 1.13E-03 3.47E-04 1.49E-03 2.95E-03 2.96E-03 G.I. LLI 2.41 E-03 6.25E-04 2.OOE-03 7.40E-03 6.21 E-03 NOTE:

1RECP 6.1.4.1 44

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE1 OA PERCENT OF TECHNICAL SPECIFICATION LIMITS 1 FOR RADIOACTIVE GASEOUS EFFLUENTS (UNIT 1)

J r ISite Boundary Noble Gas Gamma Air Dose mrad 1

First Quarter % Second Quarter Third Quarter % Fourth Quarter of TS Limit 3.20E-4 I

% of TS Limit 3.29E-4 I

of TS Limit 4.53E-4 I

T of TS Limit 2.81 E-2

%IAnnual Total %

I I

of TS Limit 1.46E-2 Beta Air Dose mrad 5.65E-5 5.81E-5 8.05E-5 5.18E-3 2.69E-3 I First Quarter % Second Quarter Third Quarter % IFourth Quarter %° Annual Total %

of TS Limit % of TS Limit of TS Limit lof TS Limit I of TS Limit Nearest Residence - NNW Critical Receptor (Highest Organ) I mrem I 2.90E-3 I 1.65E-3 I 1.69E-3 I 1.27E-2 I 9.45E-3 1 First Quarter % Second Quarter Third Quarter% IFourth Quarter %I Annual Total %

of TS Limit %of TS Limit ofTSLimit Iof TS Limit of TS Limit Nearest Vegetable Garden - ESE I 1p T Critical Receptor (Highest Organ) Imrem I 2.21 E-3 1.24E-3 1.34E-3 9.96E-3 7.37E-3 NOTE:

1RECP 6.1.6.1, 6.1.7.1 and 6.1.8.1 Page 45

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 10B PERCENT OF TECHNICAL SPECIFICATION LIMITS' FOR RADIOACTIVE GASEOUS EFFLUENTS (UNIT 2)

First Quarter % Second Quarter Third Quarter % Fourth Quarter % Annual Total %

of TS Limit % of TS LimitI of TS Limit of TS Limit of TS Limit Site Boundary Noble Gas Gamma Air Dose mrad 3.80E-4 3.95E-4 5.33E-4 4.59E-4 8.84E-4 Beta Air Dose mrad 6.71 E-5 7.07E-5 9.40E-5 2.73E-4 2.52E-4 First Quarter % Second Quarter Third Quarter % ourth Quarter % Annual Total %

of TS Limit % of TS Limit of TS Limit I of TS Limit of TS Limit Nearest Residence-NNW IPITIIIII Critical Receptor (Highest Organ) ,

mrem 3.39E-3 2.66E-3 l 2.55E-3 I 2.20E-3 l 5.40E-3 First Quarter % Second Quarter Third Quarter % IFourth Quarter %I Annual Total %

l INearest Vegetable Garden - ESE I P T (ESE) of TS Limit % of TS Limit of TS Limit of TS Limit of TS Limit Critical Receptor (Highest Organ) I mrem I 2.53E-3 I 2.OE-3 I 2.02E-3 I 1.74E-3 I 4.19E-3 NOTE:

iRECP 6.1.6.1, 6.1.7.1 and 6.1.8.1 46

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 11A RADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS FIRST QUARTER, 2010 ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

External Dose I Internal Dose Noble Gas lodines, Particulates, and Tritium Exposure Exposure Exposure Time Location Closest Dist. Exoure Whole Body Skin Ground Plane Inhalation Specific Activity (Sectors) (meters) (Hours)

Police at Shooting Range SE 700 52.0 1.35E-6 1.96E-6 0.00E+0 1.34E-4 Tour Participants (a) Simulator Bldg. S 310 1.00 1.76E-8 2.57E-8 0.00E+0 1.75E-6 (b) Bio Lab SSE 460 1.50 2.65E-8 3.86E-8 0.00E+0 2.63E-6 (c) Overlook E 210 0.25 1.23E-8 1.79E-8 0.00E+0 1.22E-6 American Indians NW 200 24.0 4.87E-6 7.11E-6 0.00E+0 4.84E-4 at Burial Grounds NNW 200 24.0 3.41 E-6 4.98E-6 0.00E+0 3.39E-4 Ranch Hands driving NW 250 0.25 3.40E-8 4.96E-8 0.OOE+0 3.38E-6 cattle around site NNW 350 0.25 1.30E-8 1.90E-8 0.OOE+0 1.29E-6 N 320 0.25 8.65E-9 1.26E-8 0.OOE+0 8.59E-7 NNE 450 0.25 3.38E-9 4.94E-9 0.OOE+0 3.36E-7 NE 630 0.25 1.72E-9 2.51E-9 0.OOE+0 1.71E-7 NOTE: All doses are in mrem.

47

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 11 B RADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS SECOND QUARTER, 2010 ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

External Dose Internal Dose Noble Gas lodines, Particulates, and Tritium Exposure Exposure Exposure Time Specific Activity Location Closest Dist. (Hours) Whole Body Skin Ground Plane Inhalation (Sectors) (meters)

Police at Shooting Range SE 700 52.0 1.39E-6 2.03E-6 0.OOE+0 9.15E-5 Tour Participants (a) Simulator Bldg. S 310 1.00 1.82E-8 2.66E-8 0.OOE+0 1.20E-6 (b) Bio Lab SSE 460 1.50 2.74E-8 3.99E-8 0.00E+0 1.80E-6 (c) Overlook E 210 0.25 1.27E-8 1.85E-8 0.00E+0 8.35E-7 American Indians NW 200 24.0 5.04E-6 7.35E-6 0.OOE+0 3.31E-4 at Burial Grounds NNW 200 24.0 3.52E-6 5.15E-6 0.00E+0 2.32E-4 Ranch Hands driving NW 250 0.25 3.52E-8 5.13E-8 0.OOE+0 2.31 E-6 cattle around site NNW 350 0.25 1.35E-8 1.97E-8 0.OOE+0 8.86E-7 N 320 0.25 8.94E-9 1.31 E-8 0.00E+0 5.88E-7 NNE 450 0.25 3.50E-9 5.11E-9 0.OOE+0 2.30E-7 NE 630 0.25 1.78E-9 2.60E-9 0.00E+0 1.17E-7 NOTE: All doses are in mrem.

48.

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 11C RADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS THIRD QUARTER, 2010 ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

External Dose Internal Dose Noble Gas lodines, Particulates, and Tritium Exposure Exposure Exposure Time Specific Activity Location Closest Dist. (Hours) Whole Body Skin Ground Plane Inhalation (Sectors) (meters) (

Police at Shooting Range SE 700 52.0 1.89E-6 2.76E-6 0.00E+0 8.99E-5 Tour Participants (a) Simulator Bldg. 5 310 1.00 2.48E-8 3.63E-8 0.00E+0 1.18E-6 (b) Bio Lab SSE 460 1.50 3.73E-8 5.44E-8 0.OOE+0 1.77E-6 (c) Overlook E 210 0.25 1.73E-8 2.52E-8 0.OOE+0 8.21 E-7 American Indians NW 200 24.0 6.86E-6 1.00E-5. 0.OOE+O 3.26E-4 at Burial Grounds NNW 200 24.0 4.80E-6 7.OOE-6 0.00E+0 2.28E-4 Ranch Hands driving NW 250 0.25 4.79E-8 6.99E-8 0.OOE+O 2.27E-6 cattle around site NNW 350 0.25 1.83E-8 2.68E-8 0.OOE+0 8.70E-7 N 320 0.25 1.22E-8 1.78E-8 0.OOE+0 5.78E-7 NNE 450 0.25 4.76E-9 6.95E-9 0.OOE+0 2.26E-7 NE 630 0.25 2.42E-9 3.54E-9 0.OOE+0 1.15E-7 NOTE: All doses are in mrem.

49

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 11D RADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS FOURTH QUARTER, 2010 ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

External Dose Internal Dose Noble Gas lodines, Particulates, and Tritium Exposure Exposure Exposure Time Specific Activity Location Closest Dist (Hours) Whole Body Skin Ground Plane Inhalation (Sectors) (meters)

Police at Shooting Range SE 700 52.0 5.47E-5 8.06E-5 2.46E-7 3.13E-4 Tour Participants (a) Simulator Bldg. S 310 1.00 7.18E-7 1.06E-6 1.71 E-9 4.10E-6 (b) Bio Lab SSE 460 1.50 1.08E-6 1.59E-6 3.88E-9 6.16E-6 (c) Overlook E 210 0.25 5.OOE-7 7.36E-7 5.57E-10 2.86E-6 American Indians NW 200 24.0 1.98E-4 2.92E-4 2.54E-7 1.13E-3 at Burial Grounds NNW 200 24.0 1.39E-4 2.04E-4 1.45E-7 7.93E-4 Ranch Hands driving NW 250 0.25 1.38E-6 2.04E-6 1.86E-9 7.91 E-6 cattle around site NNW 350 0.25 5.30E-7 7.81 E-7 6.29E-10 3.03E-6 N 320 0.25 3.52E-7 5.18E-7 3.17E-10 2.01E-6 NNE 450 0.25 1.38E-7 2.03E-7 1.36E-10 7.87E-7 NE 630 0.25 7.OOE-8 1.03E-7 8.39E-11 4.OOE-7 NOTE: All doses are in mrem.

50

Enclosure DCL-1 1-049 DIABLO CANYON POWER PLANT ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 2010 TABLE 11E RADIATION DOSE DUE TO RELEASE OF RADIOACTIVE GASEOUS EFFLUENTS ANNUAL TOTAL, 2010 ON-SITE DOSE TO MEMBERS OF THE PUBLIC (SPECIAL INTEREST GROUPS)

External Dose Internal Dose Noble Gas lodines, Particulates, and Tritium Exposure Exposure Exposure Time Specific Activity Location Closest DistE (Hours) Whole Body Skin Ground Plane Inhalation (Sectors) (meters)

Police at Shooting Range SE 700 208.0 5.94E-5 8.74E-5 2.46E-7 6.28E-4 Tour Participants (a) Simulator Bldg. S 310 4.00 7.78E-7 1.15E-6 1.71 E-9 8.24E-6 (b) Bio Lab SSE 460 6.00 1.17E-6 1.72E-6 3.88E-9 1.24E-5 (c) Overlook E 210 1.00 5.42E-7 7.98E-7 5.57E-10 5.73E-6 American Indians NW 200 96.0 2.15E-4 3.17E-4 2.54E-7 2.27E-3 at Burial Grounds NNW 200 96.0 1.50E-4 2.21 E-4 1.45E-7 1.59E-3 Ranch Hands driving NW 250 1 1.50E-6 2.21 E-6 1.86E-9 1.59E-5 cattle around site NNW 350 1 5.75E-7 8.46E-7 6.29E-10 6.08E-6 N 320 1 3.82E-7 5.62E-7 3.17E-10 4.04E-6 NNE 450 1 1.49E-7 2.20E-7 1.36E-10 1.58E-6 NE 630 1 7.59E-8 1.12E-7 8.39E- 11 8.03E-7 NOTE: All doses are in mrem.

51

Enclosure PG&E Letter DCL-1 1-049 VIII. Meteorological Data Meteorological Data The hour-by-hour listing of wind speed, wind direction, atmospheric stability and precipitation is contained on compact disc with this submittal.

Compact Diskette Information Special Instructions: The CD-R media are read-only, 700MB compact diskettes.

52

Attachment 1 PG&E Letter DCL-1 1-049 Attachment 1 Diablo Canyon Power Plant Program Directive, CY2, "Radiological Monitoring and Controls Program," Revision 7

- - ISSUED FOR USE BY: DATE: EXPIRES:

DIABLO CANYON POWER PLANT CY2 PROGRAM DIRECTIVE Rev. 7 Page 1 of 12 Radiological Monitoring and Controls Program 01/25/10 Effective Date QUALITY RELATED Table of Contents

1. PRO G RAM O VERVIEW ............................................................................ 1

2. A P P LIC A B ILITY .......................................................................................... 2

3. D E FIN IT IO NS ............................................................................................ 3

4. PROGRAM OBJECTIVES AND REQUIREMENTS .................................... 3

5. R ES PO NS IBILIT IES ................................................................................... 7

6. KEY IMPLEMENTING DOCUMENTS ......................................................... 8

7. CLOSELY RELATED PROGRAMS .......................................................... 8

8. R EC O R D S ................................................................................................. 9

9. R EFE R EN C ES ......................................................................................... 9

10. GRADED QA REQUIREMENTS FOR RADIOLOGICAL MONITORING AND CONTROLS PROGRAM ................................................................. 11

1. PROGRAM OVERVIEW 1.1 It is the policy of nuclear generation that the release of radioactive materials to the environment be in compliance with Federal regulations and be "As Low As Reasonably Achievable" (ALARA). The overall objectives are to protect the health and safety of the public from undue radiation exposure and to minimize the amount of radioactive effluents resulting from the operation of the plant.

1.2 This PD defines the overall policies and general requirements related to the Radiological Monitoring and Controls Program (RMCP). This includes the Radiological Environmental Monitoring Program (REMP), and the Radioactive Effluent Controls Program (RECP).

CY2u3rO7.DOC 0121.1235

Radiological Monitoring and Controls Program CY2 R7 Page 2 of 12 1.3 The scope of this PD is focused on the control of releases of radioactive material to the environment, and minimizing radiological impact on the general public. Radiation protection of plant workers and visitors within the restricted area of the plant is within the scope of RP1, "Radiation Protection."

1.4 Figure 1 illustrates the hierarchy of procedures associated with this PD.

Figure 1: CY2 Hierarchy of Procedures CY2 Radiological Monitoring and Controls Program IDAPs Radioactive Effluent Control Program Environmental Radiological Monitoring Procedure DLAPs Department Specific Administrative Controls 1.5 This document was converted; therefore, revision bars are not included.

2. APPLICABILITY This PD is applicable to all persons involved in radioactive effluent control, monitoring, and management activities. This includes all nuclear generation personnel, personnel matrixed to nuclear generation from other company organizations, personnel in other company organizations that are engaged in activities in support of nuclear generation, and contractor personnel that are working under nuclear generation supervision.

CY2u3ro7.DOC 0121.1235

Radiological Monitoring and Controls Program CY2 R7 Page 3 of 12

3. DEFINITIONS 3.1 ALARA (acronym for "as low as reasonably achievable"): A term that means making every reasonable effort to maintain exposures to radiation as far below the dose limits of 10 CFR 20 as is practical consistent with the purpose for which the licensed activity is undertaken, taking into account the state of technology, the economics of improvements in relation to state of technology, the economics of improvements in relation to benefits to the public health and safety, and in relation to utilization of nuclear energy and licensed materials in the public interest. The specific objectives of achieving ALARA effluents are based on those described in 10 CFR 50, Appendix I.

3.2 Environmental Radiological Monitoring Procedure (ERMP): Contains a description of sample locations, types of sample locations, methods and frequency of analysis, and reporting requirements.

3.3 Offsite Dose Calculation Procedure (ODCP): Contains the methodology and parameters Osed in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints.

3.4 Radiological Monitoring and Controls Program (RMCP): Contains the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specifications 5.5.1 and 5.5.4 and descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Technical Specifications 5.6.2 and 5.6.3.

4. PROGRAM OBJECTIVES AND REQUIREMENTS 4.1 Program Objectives:

The nuclear generation radiological monitoring and controls program is established to meet the following objectives:

4.1.1 Ensure that systems, methods, and controls are established to meet applicable regulatory requirements and objectives for release of radioactive effluents.

Liquid and gaseous radioactive waste processing systems provide the means for controlling radioactive releases. It is also important to establish administrative controls with clear delineation of responsibilities to ensure that monitoring, measurement, and release activities are properly sequenced, authorized, and controlled.

4.2 Program Requirements The basic requirement for the radiological monitoring and controls program shall be to maintain radioactive releases to the unrestricted areas surrounding the plant in conformance with applicable Federal regulations and ALARA. The following sections provide additional requirements for various elements of the program.

CY2u3rO7.DOC 0121.1235

Radiological Monitoring and Controls Program CY2 R7 Page 4 of 12 4.2.1 Changes to the RMCP (including ODCP, ERMP and RECP) shall be processed in accordance with the requirements of the plant Technical Specification Section 5.5.1.

4.2.2 Radiological Environmental Monitoring Program

a. A Radiological Environmental Monitoring Program (REMP) shall be established and maintained to comply with the plant Technical Specification 5.5.1, Radiological Environmental Monitoring Program requirements. The program shall be provided to monitor the radiation and radionuclides in the environs of the plant, and shall address the following:

1. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the Environmental Radiological Monitoring Procedure (ERMP),

2. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required by the results of this census, and

3. Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in the environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

4.2.3 Radioactive Effluent Control Program

a. Monitoring requirements shall be established and maintained for all major and potentially significant paths for release of radioactive material during normal plant operation, including anticipated operational occurrences, to comply with Regulatory Guide 1.21, Revision 1, June 1974, requirements.

b. Procedures shall be established and maintained to define the methods and requirements for control of liquid and gaseous radioactive discharges within the limits of the plant Technical Specification Section 5.5.4. These procedures shall address the following:

1. Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance requirements and setpoint determination in accordance with methodology in the Offsite Dose Calculation Procedure, (ODCP).

2. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas conforming to 10 CFR Part 20, Appendix B, Table 2, Column 2.

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Radiological Monitoring and Controls Program CY2 R7 Page 5 of 12

3. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCP.

4. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to Appendix I to 10 CFR Part 50.

5. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCP at least every 31 days.

6. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50.

7. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

a) For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin.

b) For Iodine-1 31, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/yr to any organ.

8. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to Appendix I to 10 CFR Part 50.

9. Limitations on the annual and quarterly doses to members of the public from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary conforming to Appendix I to 10 CFR Part 50.

10. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.

11. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program Surveillance Frequency.

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Radiological Monitoring and Controls Program CY2 R7 Page 6 of 12

c. Sampling and analysis methods associated with effluent monitoring activities shall be controlled in accordance with a department level administrative procedure that controls material and equipment used for analysis for the chemistry and radiochemistry programs.

d. Systems that are known pathways for radioactive releases shall be explicitly addressed. Periodic sampling of systems with the potential of becoming radioactively contaminated should also be addressed.

e. An onsite meteorological program shall be established and maintained in accordance with the requirements of Regulatory Guide 1.23, February 1972, to provide sufficient data for the performance of dose assessments.

f. The collection and processing of technical data required to support the Annual Radioactive Effluent Release Report and non-routine reports to the NRC to comply with the plant Technical Specification 5.6.3 shall be defined as part of this program. The processing of these reports shall be performed in accordance with X11, "Regulatory Interface."

4.2.4 Offsite Dose Calculation Procedures

a. Offsite Dose Calculation Procedures (ODCP) shall be established and maintained to define and control the methods for determining offsite doses.

NRC Regulatory Guide 1.109, Revision 1, October 1977, as well as its interpretation through NUREG 0133, should be used as guidance for establishing acceptable methods. These procedures shall address the following:

1. Methods for determining monitoring instrumentation alarm setpoints are addressed in accordance with a Department-Level Administrative Procedure (DLAP) under CY2.

2. Methods for determining effluent concentrations.

3. Methods for calculating doses to persons in unrestricted areas surrounding the plant from all exposure pathways.

b. Changes to the ODCP shall be processed in accordance with the requirements of Technical Specification 5.5.1.

4.2.5 Environmental Radiological Monitoring Procedure

a. An Environmental Radiological Monitoring Procedure (ERMP) shall be established and shall contain a description of sample locations, types of sample locations, methods and frequency of analysis, and reporting requirements.

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Radiological Monitoring and Controls Program CY2 R7 Page 7 of 12 4.2.6 Radwaste Treatment Systems

a. Radwaste treatment systems shall be provided to control the processing and release of radioactive materials in gaseous and liquid effluent in compliance with Technical Specification requirements. The design of these systems shall be controlled in accordance with CF3, "Design Control," and the requirements of Regulatory Guide 1.143, October 1979.

b. Approval of changes to the radwaste treatment systems shall be processed in accordance with the requirements of CF4, "Modification Control."

4.2.7 Quality Assurance Requirements In addition to requirements specified in earlier sections and those requirements utilizing procedures in Section 6.2, the control program shall be subject to the quality assurance requirements specified in CY1, "Chemistry/Radiochemistry."

4.3 Support by the Company Departments Departments outside of nuclear generation may be called upon to support nuclear generation activities associated with the Radioactive Monitoring Controls Program. The contract or agreement between nuclear generation and other departments shall ensure the support is performed in accordance with the requirements of this PD.

For example: Meteorological services may perform annual meteorological data reviews and calculate dispersion and deposition factors for use the radioactive effluents control program.

5. RESPONSIBILITIES 5.1 The Chief Nuclear Officer is responsible for establishing the policy and general requirements for the Radiological Monitoring and Controls Program, for providing management support and guidance for the program's implementation, and ensuring compliance with all regulatory requirements is maintained. The chief nuclear officer is also responsible for ensuring that support from reporting departments is provided for the Radiological Monitoring and Controls Program.

5.2 The Station Director is responsible for the overall development, implementation, and maintenance of the Radiological Monitoring and Controls Program in accordance with the requirements of this PD.

5.3 The senior director operations services is responsible for the direct implementation of the Radiological Monitoring and Controls Program with the exception of the design of radwaste treatment and effluent monitoring systems.

5.4 The senior director engineering services is responsible for maintaining the design bases for installed plant radwaste treatment and effluent monitoring systems, structures, and components and providing technical support to the plant for the operation and maintenance of these systems.

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Radiological Monitoring and Controls Program CY2 R7 Page 8 of 12 5.5 The quality verification director is responsible for auditing the Radiological Monitoring and Controls Program as outlined in Section 10.3.10.

5.6 The maintenance services director is responsible for maintaining the radiation monitoring systems and the hardware and software for the Rad Effluent program.

6. KEY IMPLEMENTING DOCUMENTS 6.1 Inter-Departmental Administrative Procedures (IDAPs)

Inter-Department Administrative Procedures shall be developed to address the following aspects of the Radiological Monitoring and Controls Program:

6.1.1 An IDAP shall be developed to define the requirements and responsibilities associated with the Radioactive Effluent Control Program.

6.1.2 An IDAP shall be developed to define the requirements and responsibilities associated with the Environmental Radiological Monitoring Procedure.

6.2 Department-Level Administrative Procedures (DLAPs)

Departments responsible for performing activities related to the Radioactive Effluent Control program shall develop DLAPs as appropriate to control program activities.

7. CLOSELY RELATED PROGRAMS 7.1 Interfaces This section describes each of the principal interfaces and boundaries between this Program Directive and other management processes.

7.1.1 AD10, "Records" "Records" provides for the retention of Radiological Monitoring and Controls Program records.

7.1.2 CF3, "Design Control" "Design Control" addresses the implementation of design activities for installed radwaste treatment and effluent monitoring systems in accordance with the requirements of NRC Regulatory Guide 1.143.

7.1.3 CF4, "Modification Control" "Modification Control" addresses the implementation of modification activities for installed effluent monitoring systems.

7.1.4 CY1, "Chemistry/Radiochemistry" "Chemistry/Radiochemistry" addresses the methods for chemistry/radiochemistry sampling and analysis of liquid and gaseous radioactive effluents in support of this PD.

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Radiological Monitoring and Controls Program CY2 R7 Page 9 of 12 7.1.5 CY2.ID1, "Radioactive Effluent Controls Program" "Radioactive Effluent Controls Program" contains the general program requirements to ensure the requirements of 10 CFR Part 20 and 10 CFR Part 50, Appendix I, are met.

7.1.6 OM7, "Problem Resolution" "Problem Resolution" addresses deficiencies identified during the implementation of the radioactive effluent control program. OM7 also addresses evaluating nonconformances for reportability in accordance with Technical Specifications.

7.1.7 TQ1, "Personnel Training and Qualification" "Personnel Training and Qualification" identifies training and qualification requirements for personnel.

7.1.8 X11, "Regulatory Interface" "Regulatory Interface" addresses the process for required reporting and communication with outside agencies.

8. RECORDS None

9. REFERENCES 9.1 Diablo Canyon Nuclear Power Plant Facility Operating Licenses (Unit 1, Unit 2) 9.2 QA Commitments:

9.2.1 FSAR Chapter 17.2 9.2.2 Regulatory Guide 1.33 9.3 Regulatory Guides:

9.3.1 Guide 1.109, Revision 1, October 1977, "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" 9.3.2 Guide 1.143, October 1979, "Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants" 9.3.3 Guide 1.21, Revision 1, June 1974, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluent from Light-Water-Cooled Nuclear Power Plants" CY2u3rO7.DOC 0121.1235

Radiological Monitoring and Controls Program CY2 R7 Page 10 of 12 9.3.4 Guide 1.23, February 1972, "Onsite Meteorological Programs" 9.3.5 Guide 4.1, Revision 1, April 1975, "Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants" 9.3.6 Guide 4.15, Revision 1, February 1979, "Quality Assurance For Radiological Monitoring Programs (Normal Operations) - Effluent Streams and the Environment" 9.4 Title 10, Code of Federal Regulations:

9.4.1 Part 20, "Standards for Protection Against Radiation" 9.4.2 Part 50, Appendix I 9.4.3 Part 50, Appendix A, GDC 60, 64 9.4.4 Part 50.36a, "Technical Specifications on Effluents from Nuclear Power Reactors" 9.5 Title 40, Code of Federal Regulations, "Environmental Radiation Protection Standards for Nuclear Power Operations" CY2u3r07.DOC 0121.1235

Radiological Monitoring and Controls Program CY2 R7 Page 11 of 12

10. GRADED QA REQUIREMENTS FOR RADIOLOGICAL MONITORING AND CONTROLS PROGRAM The basis for these Graded QA requirements is to comply with the regulations of 10 CFR 20, 10 CFR 50, 40 CFR 190, the Technical Specifications and Regulatory Guides 1.21, and 4.15.

10.1 Graded Items Radioactive Effluent monitoring instruments are classified as Category 2 or Category 3 items per Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident." These instruments are used for detection and assessment of releases and possibly detection of containment breach with accomplishment of mitigation of the breach. These items fall under Graded QA requirements.

10.2 Graded Activities 10.2.1 Installed radiation monitors required per Technical Specification 3.3.3 for monitoring radioactive effluents during plant operations or accidents shall be calibrated at prescribed intervals.

10.2.2 Sampling and analysis of liquid and gaseous effluents shall be performed in accordance with CY1, "Chemistry/Radiochemistry."

10.2.3 Calculations, computer programs, and procedures for evaluating the dose associated with radioactive effluents shall be performed in accordance with approved quality related procedures.

10.3 Graded Requirements 10.3.1 Effluent releases shall be maintained ALARA and shall be performed in accordance with the requirements of this Program Directive to limit the concentrations, doses and doserates as specified in DCPP Technical Specification 5.5.4, NRC regulations 10 CFR 20, 10 CFR 50 Appendix I, and EPA regulation 40 CFR 190.

10.3.2 The Annual Radiological Environmental Operating Report shall be developed in accordance with Technical Specification 5.6.2.

10.3.3 The Annual Radioactive Effluent Release Report shall be developed in accordance with Technical Specification 5.6.3.

10.3.4 Records that support and document the Radiological Monitoring and Controls Program shall be controlled in accordance with AD10, "Records."

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Radiological Monitoring and Controls Program CY2 R7 Page 12 of 12 10.3.5 Personnel involved in direct implementation of chemistry/radiochemistry, operations, or radiation protection activities in support of the Radiological Monitoring and Controls Program are qualified in accordance with the requirements of TQ1, "Personnel Training and Qualification." In addition, personnel involved in direct implementation of activities in support of the Radiological Environmental Monitoring Program are qualified in accordance with the requirements of an interdepartmental administrative procedure for Environmental Radiological Monitoring (ERMP).

10.3.6 Notifications and reports to and correspondence with regulatory agencies shall be done in accordance with X11, "Regulatory Interface."

10.3.7 Written plans, procedures and instructions for implementing Radiological Monitoring and Controls Program shall be prepared, processed, and controlled in accordance with AD1, "Administrative Control Program."

10.3.8 Procurement of quality-related equipment or services shall be in accordance with written procedures. Applicable regulatory requirements, design bases, and any other requirements necessary to assure adequate quality shall be included in or invoked by reference in documents for procurement of items or services. Test or acceptance requirements and documentation to be submitted by the supplier shall be identified in the procurement documents. Receipt inspection requirements, if required, shall be identified in the procurement documents.

10.3.9 Deficiencies identified during implementation of this program shall be documented and controlled in accordance with OM7, "Problem Resolution."

10.3.10 FSAR Chapter 17.18, "Audits," prescribes the audit frequency for various portions of the RMCP:

a. The performance of activities required by the quality assurance program for the Radioactive Effluents Control Program shall be audited at least once per 24 months unless specified otherwise.

b. The Radiological Environmental Monitoring Program, implementing procedures, and program results shall be audited at least once per 24 months.

c. The Offsite Dose Calculation Procedure and its implementing procedures shall be audited at least once per 24 months.

CY2u3r07DOC 0121.1235

Attachment 2 PG&E Letter DCL-1 1-049 Attachment 2 Diablo Canyon Power Plant Interdepartmental Administrative Procedure, CY2.ID1, "Radioactive Effluent Controls Program," Revision 11

- - ISSUED FOR USE BY: DATE: EXPIRES:____

DIABLO CANYON POWER PLANT CY2.ID1 INTERDEPARTMENTAL ADMINISTRATIVE PROCEDURE Rev. 11 Page 1 of 37 Radioactive Effluent Controls Program 01/20/10 Effective Date QUALITY RELATED Table of Contents

1. S CO P E ..................................................................................................... .. 1

2. D IS C US S IO N ............................................................................................... 2

3. D E FIN IT IO N S ............................................................................................ 2

4. RESPONSIBILITIES ................................................................................... 3

5. IN ST R UCT IO NS ....................................................................................... 3 5.1 Administrative Requirements ...................................................................... 3 5.2 Reporting Requirements ............................................................................ 4 5.3 Revisions to the RECP ............................................................................... 7 5.4 Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment S ystem s ................................................................................................... .. 7

6. R E C O R D S ................................................................................................. 8

7. R E FE R EN C ES ............................................................................................ 8 ATTACHMENTS:

1. Radioactive Liquid Effluent Monitoring Instrumentation Operational R equirem ents ........................................................................................... . 10

2. Radioactive Gaseous Effluent Monitoring Instrumentation Operational R equirem ents ............................................................................................ 14

3. Liquid Effluents - CONCENTRATION Operational Requirements ............. 18

4. Liquid Effluents - Dose Operational Requirements ................................... 22

5. Liquid Radwaste Treatment System Dose Operational Requirements ..... 23

6. Gaseous Effluents - Dose Rate Operational Requirements ....................... 24

7. Dose - Noble Gases Operational Requirements ....................................... 28

8. Iodine 131, Iodine 133, Tritium, and Radioactive Material in Particulate Fo rm ....................................................................................................... . . 29

9. Gaseous Radwaste Treatment System Operational Requirements ........... 31

10. Total Dose Operational Requirements ...................................................... 32

11. Radiological Environmental Monitoring Operational Requirements ........... 34

12. Land Use Census Operational Requirements ........................................... 35

13. High Alarm Setpoints - FB & CR Ventilation Systems Actuation Instrum entation ........................................................................................ 36

1. SCOPE 1.1 This procedure contains the general program requirements of the Radioactive Effluent Controls Program. This program ensures that the requirements of 10 CFR Part 20 and 10 CFR Part 50 Appendix I are met.

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2. DISCUSSION 2.1 This procedure provides the general requirements for Radioactive Effluent Controls Program in accordance with the Technical Specifications and the implementation Generic Letter 89-01, "Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of RETS to the Off-Site Dose Calculation Manual or to the Process Control Program."

2.2 The following Technical Specification definitions are applicable: T.S. Section 5.5.1 2.2.1 The Off-site Dose Calculation Manual (ODCM) shall contain the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of radiological environmental monitoring program; and 2.2.2 The ODCM shall contain the radioactive effluent controls and radiological environmental monitoring activities, and the description of the information that should be included in the Annual Radiological Environmental Operating, and the Radioactive Effluent Release Reports required by Technical Specification 5.6.2 and 5.6.3.

2.2.3 The Diablo Canyon ODCM is made up of the following procedures:

CAP A-8, "Off-site Dose Calculation Procedure"

CY2.1D1, "Radioactive Effluent Controls Program"

RP1.ID11, "Environmental Radiological Monitoring Procedure"

CY2, "Radiological Monitoring and Controls Program" Changes to each of these procedures shall be processed in accordance with the requirements of Technical Specification Section 5.5.1.

2.3 The specific methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, is contained in CAP A-8, "Off-Site Dose Calculations Procedure (ODCP)." As such, CAP A-8 is incorporated in this procedure by reference. Therefore, the requirements for revisions to this procedure also apply to CAP A-8.

3. DEFINITIONS

3.1 Shutdown

Taking the reactor sub-critical.

3.2 Startup

Taking the reactor critical.

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4. RESPONSIBILITIES 4.1 The chemistry manager is responsible for:

4.1.1 Implementation of the Off-Site Dose Calculation Procedure in a manner that meets regulatory requirements and preparing the Annual Radiological Effluent Release Report.

4.1.2 Providing direction to the operations staff in the processing of radioactive waste streams.

4.1.3 Ensuring that a comparison of the Annual Radioactive Effluent Release Report and the Annual Radiological Environmental Operating Report is performed.

4.1.4 Ensuring that dose commitment increases due to the Land Use Census in accordance with Attachment 12 are determined and communicated promptly to radiation protection.

4.2 The radiation protection manager is responsible for:

4.2.1 Ensuring the performance of the annual land use census and that the results are provided to chemistry so that chemistry can establish the dose requirements of Attachment 12.

4.2.2 Ensuring that the results of the annual Land Use Census are included in the Annual Radiological Environmental Operating Report.

4.2.3 Ensuring that changes to the Environmental Radiological Monitoring Procedure are provided to chemistry for inclusion in the Annual Radiological Effluent Release Report.

4.2.4 Ensuring preparation, review and approval of the Nonroutine Radiological Environmental Operating Report when required by Attachment 11.

5. INSTRUCTIONS 5.1 Administrative Requirements 5.1.1 It is the intent of the Radioactive Effluent Control Program to monitor all significant release pathways from the power plant. A pathway is considered significant if a conservative evaluation yields an additional dose increment equal to or more than 10 percent of the total from all pathways considered in this program.

5.1.2 At times, minor release pathways occur from the power plant due to plant maintenance activities. For example, leaving Containment when the equipment hatch or door is open. These minor release pathways are usually negligible from a radioactive effluents control program release and dose perspective, and are well within the margin of error of the significant monitored pathway and dose models.

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Radioactive Effluent Controls Program CY2.ID1 R11 Page 4 of 37 5.1.3 Attachments 1-12 of this procedure contains the operational requirements of the Radioactive Effluent Controls Program.

5.1.4 The operational requirements are implemented by Equipment Control Guidelines (reference OP1 .DC16), CAP A-8, and XI1.ID2.

a. The Equipment Control Guidelines implement those requirements that are related to equipment and have specific allowed outage times or operator actions.

b. CAP A-8 includes the methodology and parameters used in the calculation of off-site doses resulting from radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints.

c. XIl.ID2 implements the reporting requirements.

5.2 Reporting Requirements 5.2.1 Annual Radioactive Effluent Release Report

a. Report Schedule

1. Annual Radioactive Effluent Release Reports covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year, in accordance with 10 CFR 50.36a.

b. The Annual Radioactive Effluent Release Reports shall include:

1. A summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit 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-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories; class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement, urea formaldehyde).

2. A list and description of unplanned releases as defined in ODCP from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period; CY2!ID1 u3rl 1.DOC 1209.0951

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3. Changes to the OCDM.

a) Pursuant to Technical Specification 5.5.1, changes to the following procedures made during the reporting period shall be included as described below:

NOTE: An FSAR update may be used in lieu of the ARERR for communicating changes to the NRC, regarding the PCP.

1) RP2.DC2, "Radwaste Solidification Process Control Program (PCP)"

2) CY2.ID1, "Radioactive Effluent Controls Program (RECP)"

3) CY2, "Radiological Monitoring and Controls Program (RMCP)"

4) RP1.ID11, "Environmental Radiological Monitoring Program (ERMP)"

5) CAP A-8, "Off-Site Dose Calculations (ODC)"

b) If a change is made to any of these procedures, include a legible copy of each procedure in the report. This provides a complete copy of the ODC in the report.

c) If multiple changes to any one procedure are made during the reporting period, include a copy of each revision.

d) Each procedure change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed.

e) The date the change was implemented shall be indicated (e.g., the first page of the procedure should indicate the implementation date).

4. A listing of new locations for dose calculations and/or environmental monitoring identified by the Land Use Census pursuant to Attachment 12.

5. An explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in attachments 1 and 2.

6. Description of the events leading to liquid holdup tanks or gas storage tanks exceeding the limits of ECG 19.1 or ECG 24.3.

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c. A discussion of major changes to the Radwaste Treatment Systems (liquid, gaseous and solid). The discussion of each change shall contain:

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 detailed information to totally support the reason for the change without benefit of additional or supplemental information.

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

4. An 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 predicted in the License application and amendments thereto.

5. An evaluation of the change which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto.

6. A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be make.

7. An estimate of the exposure to plant operating personnel as a result of the change.

8. Otherwise the above information may be submitted as part of the annual FSAR update.

d. In addition, the Annual Radioactive Effluent Release Report shall also include:

1. An annual summary of hourly meteorological data collected over the previous year. This annual summary may be either:

a) In the form of an hour-by-hour listing on magnetic tape/hard disk or other media of wind speed, wind direction, atmospheric stability, and precipitation (if measured).

b) In the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.

c) The licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

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2. An assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year.

3. An assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (see FSAR Figure 2.1-2) during the report period.

4. All assumptions used in making these assessments, i.e., specific activity, exposure time and location.

a) The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses.

b) The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFF-SITE DOSE CALCULATIONS (ODC).

5. An assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, "Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

e. A single submittal may be made for a multiple unit plant.

1. The submittal should combine those sections that are common to all units at the plant.

2. For units with separate radwaste system, the submittal shall specify the releases of radioactive material from each unit.

5.3 Revisions to the RECP 5.3.1 The requirements for revision to the RECP also apply to CAP A-8.

5.3.2 The requirements are provided in Technical Specification 5.5.1.

5.4 Maior Changes to Liquid, Gaseous, and Solid Radwaste Treatment Systems 5.4.1 Major changes to the liquid, gaseous, and solid radwaste treatment systems shall become effective upon review and acceptance provided the change could be made in accordance with 10 CFR 50.59.

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6. RECORDS 6.1 Data Sheets and records will be maintained in the Records Management System (RMS) in accordance with CY1 .DC1, "Analytical Data Processing Responsibilities."

7. REFERENCES 7.1 CAP A-8, "Off-site Dose Calculations (ODC)"

7.2 CAP A-5, "Liquid Radwaste Discharge Management" 7.3 CAP A-6, "Gaseous Radwaste Discharge Management" 7.4 CF4.ID1, "Modification Request and Authorization" 7.5 CF6.ID1, "Setpoint Control Program" 7.6 CY2, "Radiological Monitoring and Controls Program" 7.7 OP1 .DC1 6, "Control of Plant Equipment Not Required by the Technical Specifications" 7.8 RP1 .ID1 1, "Environmental Radiological Monitoring Procedure" 7.9 XI1.ID2, "Regulatory Reporting Requirements and Reporting Process" 7.10 Action Requests/Notifications 7.10.1 A0581581, "Track LAR Revising Fuel Handling Requirements per TSTF-51" 7.10.2 A0619600, "Rad Effluent Sampling of Ni-63" 7.10.3 A0660441, "ODCM Change Identification Discrepancies" 7.10.4 A0702694, "Evaluate Detection Limit Differences for ECG and Procedures" 7.11 License Amendment 67/66, January 22, 1992 7.12 License Amendment 184/186, January 3, 2006 7.13 License Amendment Request 93-04 7.14 10 CFR 20.1302 7.15 10 CFR 50.36a 7.16 10 CFR 50 Appendix I 7.17 40 CFR 190 CY2!ID1 u3rl 1.DOC 1209.0951

Radioactive Effluent Controls Program CY2.1D1 Rll Page 9 of 37 7.18 Regulatory Guide 1.21, Revision 1, June 1974 7.19 Regulatory Guide 1.109, Revision 1, October 1977 7.20 QA Commitments:

7.20.1 FSAR Chapter 17.2 7.20.2 Reg Guide 4.15 CY2!IDl u3rl 1.DOC 1209.0951

CY2.1D1 R11 Page 10 of 37 Radioactive Liquid Effluent Monitoring Instrumentation Operational Requirements Attachment 1: Page 1 of 4

1. Radioactive Liquid Effluent Monitoring Instrumentation (also covered by ECG 39.3)

a. Commitment for Operation

1) The radioactive liquid effluent monitoring instrumentation channels shown in Table 1 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Attachment 3 are not exceeded. The Alarm/Trip Setpoints of these channels shall be determined in accordance with the methodology and parameters in the OFF-SITE DOSE CALCULATIONS (ODC).

2) Applicability: At all times.

3) Action:

a) With a radioactive liquid effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by Attachment 3, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable.

b) With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 1.

Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Annual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

b. Surveillance Requirements

1) 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 at the frequencies shown in Table 2.

2) At least one saltwater pump shall be determined operating and providing dilution to the discharge structure at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months whenever dilution is required to meet the limits of Attachment 3.

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CY2.1D1 R11 Page 11 of 37 Radioactive Liquid Effluent Monitoring Instrumentation Operational Requirements Attachment 1: Page 2 of 4 Table 1: Radioactive Liquid Effluent Monitoring Instrumentation Minimum Channels Instrument Operable Action

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Line (RM-18)# 1 1

b. Steam Generator Blowdown Tank (RM-23) 1 2

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line (FIT-243)# 1 4

b. Steam Generator Blowdown Effluent Lines (FR-53) 1 4

c. Oily Water Separator Effluent Line (FR-251)# 1 4

3. Radioactivity Monitor Not Providing Automatic Termination of Release Oily Water Separator Effluent Line (RM-3)# 1 3

This Radioactive Liquid Effluent Monitoring Instrumentation is common to both units.

Action Statements ACTION 1 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 14 days provided that prior to initiating a release:

a. At least two independent samples are analyzed in accordance with Attachment 3.

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valvings.

Otherwise, suspend release of radioactive effluents Via this pathway.

ACTION 2 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are analyzed for radioactivit (beta or gamma) at a lower limit of microcuries/m .4 detection of no more than 10

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the secondary coolant is greater than 0.01 microcuries/gram DOSE EQUIVALENT 1-131, or

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the secondary coolant is less than or equal to 0.01 microcuries/gram DOSE EQUIVALENT 1-131.

ACTION 3 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for radioactivity (beta or gamma) at a lower limit of detection of no more than 10-7 microcuries/ml or transfer the oily water separator effluent to the Liquid Radwaste Treatment System.Ref 7.10.4 ACTION 4 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump performance curves may be used to estimate flow.

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CY2.ID1 R11 Page 12 of 37 Radioactive Liquid Effluent Monitoring Instrumentation Operational Requirements Attachment 1: Page 3 of 4 Table 2: Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Channel Channel Source Channel Functional Instrument Check Check Calibration Test 1 Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluent Line (RM-18) D P R(3) Q(1)

b. Steam Generator Blowdown Tank (RM-23) D M R(3) 0(1)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line (FIT-243) D(4) N.A. R Q

b. Steam Generator Blowdown Effluent Line (FR-53) D(4) N.A. R 0

c. Oily Water Separator Effluent Line (FR-251) Daily(4) N.A. R Q

3. Radioactivity Monitor Not Providing Automatic Termination of Release

a. Oily Water Separator Effluent Line (RM-3) D M R(3) Q(2)

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and Control Room alarm annunciation occurs if any of the following conditions exits:

a. Instrument indicates measured levels above the Alarm/Trip Setpoint (isolation and alarm), or

b. Relay control circuit failure (isolation only), or

c. Instrument indicates a downscale failure (alarm only), or

d. Instrument controls not set in operate mode (alarm only).

(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that Control Room alarm annunciation occurs if any of the following conditions exist:

a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Circuit failure, or

c. Instrument indicates a downscale failure, or

d. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK for FR-251 shall be made once per calendar day,* and for FIT-243 and FR-53 shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are made.

(5) Frequency Notation Notation Frequency D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Daily At lease once per calendar day*

M At least once per 31 days Q At least once per 92 days R At least once per 18 months P Completed prior to each release N.A. Not Applicable The frequency "once per calendar day" could result in two successive channel checks nearly 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months apart over a two day period. This frequency is different from and should not be confused with the frequency notation "D" (at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) defined in Technical Specifications.

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CY2.1D1 R11 Page 13 of 37 Radioactive Liquid Effluent Monitoring Instrumentation Operational Requirements Attachment 1: Page 4 of 4 Bases: Radioactive Liquid Effluent Monitoring Instrumentation 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 and adjusted in accordance with the methodology and parameters in the ODCP to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. 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.

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CY2.ID1 R11 Page 14 of 37 Radioactive Gaseous Effluent Monitoring Instrumentation Operational Requirements Attachment 2: Page 1 of 4

1. Radioactive Gaseous Effluent Monitoring Instrumentation (also covered by ECG 39.4)

a. Commitment for Operation

1) The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of Attachment 6 is not exceeded. The Alarm/Trip Setpoints of these channels meeting Attachment 6 shall be determined and adjusted in accordance with the methodology and parameters in the ODCP.

2) Apolicability: As shown in Table 3.

3) Action:

a) With a radioactive gaseous effluent monitoring instrumentation channel Alarm/Trip Setpoint less conservative than required by the above Commitment, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable.

b) With the number of OPERABLE radioactive gaseous effluent monitoring instrumentation channels less than the Minimum Channels OPERABLE, take the ACTION shown in Table 3. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION or explain in the next Annual Radioactive Effluent Release Report why this inoperability was not corrected within the time specified.

b. Surveillance Requirements

1) 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 at the frequencies shown in Table 4.

CY2!IDl u3rl 1.DOC 1209.0951

CY2.1D1 R11 Page 15 of 37 Radioactive Gaseous Effluent Monitoring Instrumentation Operational Requirements Attachment 2: Page 2 of 4 Table 3: Radioactive Gaseous Effluent Monitoring Instrumentation Minimum Instrument Channel Operable Applicability Action

1. Gaseous Radwaste System Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of Release (RM-22) 1 5

2. Plant Vent system

a. Noble Gas Activity Monitor Providing Alarm 1 7 (RM-14 or RM-14R)

b. Iodine Sampler 1

9

c. Particulate Sampler 1

9

d. Flow Rate Monitor (FR-12) 1

6

e. Iodine Sampler Flow Rate Monitor 1

6

3. Containment Purge System Noble Gas Activity Monitor - Providing Alarm and 2(1) 8 Automatic Termination of Release (RM-44A or 44B)

(1) 2 channels required in MODES 1, 2, 3 and 4. Only 1 channel required during movement of recently irradiated fuel assemblies within containment.

At all times.

- MODES 1-4; also MODE 6 during CORE ALTERATIONS or movement of irradiated fuel within containment.

ACTION 5 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 6 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 7 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 8 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend containment PURGING of radioactive effluents via this pathway.

ACTION 9 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Commitment Table 6.

NOTE FOR ACTION 9: To respond to the low flow alarm, determine that a simple fix cannot be made and that an auxiliary sampler is needed. Move the sampler in, hook up and verify operation, a maximum of two hours is considered a reasonable time. Over two hours should be considered as exceeding the time limitation of the commitment for operation (ECG 39.4).

CY201D1u3r 11DOC 1209.0951

CY2.ID1 Rll Page 16 of 37 Radioactive Gaseous Effluent Monitoring Instrumentation Operational Requirements Attachment 2: Page 3 of 4 Table 4: Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Channel Modes for Which Channel Source Channel Functional Surveillance Instrument Check Check Calibration Test Is Required I Gaseous Radwaste System Noble Gas Activity Monitor - Providing Alarm and Automatic Termination of P P R(3) Q(1)

Release (RM-22)

2. Plant Vent System

a. Noble Gas Activity Monitor Providing D M R(3) Q(2)

Alarm (RM-14 or RM-14R)

b. Iodine Sampler W(4) N.A. N.A. N.A. *

c. Particulate Sampler W(4) N.A. N.A. N.A.

d. Flow Rate Monitor (FR-12) D N.A. R Q *

e. Iodine Sampler Flow Rate Monitor D N.A. R Q *

3. Containment Purge System Noble Gas Activity Monitor - Providing D P R(3) Q0(1)

Alarm and Automatic Termination of Release (RM-44A or 44B)

At all times.

MODES 1-4; also MODE 6 during CORE ALTERATIONS or movement of irradiated fuel within containment.

(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm/Trip Setpoint (isolation and alarm), or

b. Instrument indicates a downscale failure (alarm only), or

c. Instrument controls not set in operate mode (alarm only).

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

a. Instrument indicates measured levels above the Alarm Setpoint, or

b. Circuit failure, or

c. Instrument indicates a downscale failure, or

d. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) The CHANNEL CHECK shall consist of verifying that the iodine cartridge and particulate filter are installed in the sample holders.

(5) Frequency Notation Notation Frequency D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months W At least once per 7 days M At least once per 31 days Q At least once per 92 days R At least once per 18 months P Completed prior to each release N.A. Not Applicable CY2IID1u3r11.DOC 1209.0951

CY2.1D1 R11 Page 17 of 37 Radioactive Gaseous Effluent Monitoring Instrumentation Operational Requirements Attachment 2: Page 4 of 4 Bases: Radioactive Gaseous Effluent Monitoring Instrumentation The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable,

.the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCP to ensure that the alarm/trip will occur prior to exceeding the limits of NUREG 0133. 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. The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of Attachment 7 shall be such that concentrations as low as 1 x 10-5 pCi/ml are measurable.

CY2!ID1u3rl 1.DOC 1209.0951

CY2.1D1 R11 Page 18 of 37 Liquid Effluents - CONCENTRATION Operational Requirements Attachment 3: Page I of 4

1. Liquid Effluents- CONCENTRATION

a. Commitment for Operation

1) The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see FSAR Figure 2.1-2) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 microcurie/ml total activity.

2) Applicability: At all times.

3) Action:

With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits.

b. Surveillance Requirements

1) Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 5.

2) The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCP to assure that the concentrations at the point of release are maintained within the limits of Attachment 3.

CY2!ID1 u3rl 1.DOC 1209.0951

CY2.1D1 R11 Page 19 of 37 Liquid Effluents - CONCENTRATION Operational Requirements Attachment 3: Page 2 of 4 Table 5: Radioactive Liquid Waste Sampling and Analysis Program Minimum Lower Limit of Sampling Analysis Detection (LLD)

Liquid Release Type Frequency Frequency Type Of Activity Analysis (PCi/ml)(1 )

1. Batch Waste Release P P Principal Gamma Emitters(6) 5x1 0.7 Tanks(4) Each Batch Each Batch 1-131 lx10 8 P M Dissolved and Entrained lxi 0s One Batch/M Gases (Gamma emitters)

P M H-3 1x10 5 Each Batch Composite(2) Gross Alpha 1 x10.7 P Q Sr-89, Sr-90 5x10-8 Each Batch Composite(2) Fe-55 lxi 0-6 Pu-238, Pu-239, Pu-240, 5x1 0.8 Pu-241, Pu-242 U-233, U-234, U-235, U-236, 5x10.8 U-238 Ni-63 lx10-6

2. Continuous Releases(5) D W Principal Gamma Emitters(6) 5x10.7 Steam Generator Grab Sample Composite(3 ) 1-131 1x1 Blowdown Tank 1-131 lxl__6 M M Dissolved and Entrained lxi 0s Grab Sample Gases (Gamma emitters)

D M H-3 1x10.5 Grab Sample Composite(3) Gross Alpha 1xl10.7 I

D Q Sr-89, Sr-90 5x10-8 Grab Sample Composite(3) Fe-55 1X,0 6

3. Continuous Releases(5) D W Principal Gamma Emitters(6) 5x10-7 Oily Water Separator Grab Sample Composite(3)

Effluent Table Notations - next page CY2!ID1u3rl 1.DOC 1209.0951

CY2.1D1 R11 Page 20 of 37 Liquid Effluents - CONCENTRATION Operational Requirements Attachment 3: Page 3 of 4 Table Notations (1) The LLD is defined, for the 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% probability with only 5%

probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD = 4. 6 6 sb 6

E

V

2.22x10

Y

exp(-XAt)

Where:

LLD = the "a priori" lower limit of detection (microcuries 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 (counts per minute),

E = the counting efficiency (counts per disintegration),

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

2.22 x 106 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec 1 ), and At = the elapsed time between the midpoint of sample collection and the time of counting (sec).

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

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posterior (after the fact) limit for a particular measurement.

(2) 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.

(3) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be composited in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

(4) 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, by a method described in the ODCP, to assure representative sampling.

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

(6) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-1 34, Cs-1 37, and Ce-1 41. Ce-1 44 shall also be measured but with an LLD of 5xi 0-6. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report.

If two or more redundant sample and analysis results of the same discharge pathway indicate slight differences in low level isotopic mixture, the analysis resulting in higher calculated dose should be used. The isotopes chosen to represent the discharge shall be reported.

(7) Frequency Notation:

Notation Frequency D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

P Completed prior to each release.

CY2!lID u3rl i .DOC 1209.0951

CY2.1D1 R11 Page 21 of 37 Liquid Effluents - CONCENTRATION Operational Requirements Attachment 3: Page 4 of 4 Bases: Liquid Effluents - Concentration This Regulatory Commitment is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR 20.1301(e) to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-1 35 is the controlling radioisotope and its Effluent Concentration Limit (ECL) in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.

This Regulatory Commitment applies to the release of radioactive materials in liquid effluents from all units at the site.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the Lower Limits Of Detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in Currie, L.A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

CY2!ID1u3rl 1.DOC 1209.0951

CY2.ID1 R11 Page 22 of 37 Liquid Effluents - Dose Operational Requirements Attachment 4: Page 1 of 1

1. Liquid Effluents - Dose

a. Commitment for Operation

1) The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see FSAR Figure 2.1-2) shall be limited to the following:

a) During any calendar quarter to less than or equal to 1.5 mrem to the whole body and to 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 whole body and to less than or equal to 10 mrem to any organ.

2) Applicability: At all times.

3) Action:

a) 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, pursuant to 10 CFR 50.4, a Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. Surveillance Requirements

1) Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCP at least once per 31 days.

Bases: Liquid Effluents - Dose This Regulatory Commitment is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix 1, 10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.A 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 material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The dose calculation methodology and parameters in the ODCP 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. The equations specified in the ODCP 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.

This Regulatory Commitment applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

CY2!lDlu3rl 1.DOC 1209.0951

CY2.1D1 Rll Page 23 of 37 Liquid Radwaste Treatment System Dose Operational Requirements Attachment 5: Page I of 1

1. Liquid Radwaste Treatment System

a. Commitment for Operation NOTE: The Liquid Radwaste Treatment System is common to both units.

1) The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see FSAR Figure 2.1-2) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period.

2) Applicability: At all times.

3) Action:

a) With any portion of the Liquid Radwaste Treatment System not in operation and with radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to 10 CFR 50.4, a Special Report which includes the following information:

(1) Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability, (2) Action(s) taken to restore the inoperable equipment to OPERABLE status, and (3) Summary description of action(s) taken to prevent a recurrence.

b. Surveillance Requirements

1) Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days, in accordance with the methodology and parameters in the ODCP when Liquid Radwaste Treatment Systems are not being fully utilized.

2) The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting attachments 3 and 4.

Bases: Liquid Radwaste Treatment System The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR 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 specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix 1,10 CFR Part 50, for liquid effluents.

This Regulatory Commitment applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.

CY2!lDlu3rl 1.DOC 1209.0951

CY2.ID1 R11 Page 24 of 37 Gaseous Effluents - Dose Rate Operational Requirements Attachment 6: Page 1 of 4

1. Gaseous Effluents - Dose Rate

a. Commitment for Operation

1) The dose rate due to radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see FSAR Figure 2.1-2) shall be limited to the following:

a) For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mrem/yr to the skin.

b) For Iodine-1 31, for Iodine-1 33, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.

2) Applicability: At all times.

3) Action:

With the dose rate(s) exceeding the above limits, immediately decrease the release rate to within the above limit(s).

b. Surveillance Requirements

1) The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and procedures of the ODCP.

2) The dose rate due to Iodine-1 31, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and procedures of the ODCP by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 6.

CY2!IDlu3rl ,DOC 1209.0951

CY2.1D1 R11 Page 25 of 37 Gaseous Effluents - Dose Rate Operational Requirements Attachment 6: Page 2 of 4 Table 6: Radioactive Gaseous Waste Sampling and Analysis Program Lower Limit of Sampling Minimum Analysis Detection (LLD)

Gaseous Release Type Frequency Frequency Type of Activity Analysis (pCi/ml)( 1 )

1. Waste Gas Decay Tank P P Principal Gamma lx, 0-4 Each Tank Grab Each Tank Emitters(7)

Sample (noble gases)

2. Containment Purge P P Principal Gamma lx1i0 4 Each Purge(2) Each Purge(2) Emitters(7)

Grab Sample (noble gases) 1-131, 1-133 1x10 9 Principal Gamma lxli 0 9 Emitters (particulates)

H-3 1x10.6

3. Plant Vent M(2) M(2) Principal Gamma lx1i0 4 Grab Sample Emitters(7)

(noble gases)

W(3) (5) W H-3 1 x1 0"1 Grab Sample

4. All Release Types as listed Continuous(6 ) W(4) 1-131 lx1012 in 1., 2., 3., above, at the plant vent Charcoal 1-133 1x10"1° Sample 11 Continuous(6) W(4) Principal Gamma 1x10 Particulate Sample Emitters(7)

Continuous(6) M Gross Alpha lxI011 Composite Particulate Sample Continuous(6) Q Sr-89, Sr-90 1xl 011 Composite Particulate Sample

5. Steam Generator Blowdown M(8) M(8) Principal Gamma lxi 1.4 Tank Vent Emitters(7) (noble gases)

Table Notations - next page CY2!ID1u3rl 1.DOC 1209.0951

CY2.1D1 R11 Page 26 of 37 Gaseous Effluents - Dose Rate Operational Requirements Attachment 6: Page 3 of 4 Table Notations (1) 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% probability with only 5%

probability of falsely concluding that a blank observation represents a "real" signal.

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

LLD = 4.66sb E* V* 2.22x10 6

Y

exp(-X~t)

Where:

LLD = the "a priori" lower limit of detection (microcuries 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 (counts per minute),

E = the counting efficiency (counts per disintegration),

V 6= the sample size (units of mass or volume),

2.22 x 10 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec-'), and At = the elapsed time between the midpoint of sample collection and the time of counting (sec).

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

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posterior (after the fact) limit for a particular measurement.

(2) Sampling and analyses shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a 1-hour period, when either:

a. Analysis shows that the DOSE EQUIVALENT 1-131 concentration in the reactor coolant has increased more than a factor of 3.

b. The noble gas monitor shows that effluent activity has increased more than a factor of 3.

(3) Tritium grab samples shall be taken a least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

(4) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing or after removal from sampler. Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLD's may be increased by a factor of 10.

This requirement only applies when either:

a. Analysis shows that the DOSE EQUIVALENT 1-131 concentration in the reactor coolant has increased more than a factor of 3.

b. The noble gas monitor shows that effluent activity has increased more than a factor of 3.

(5) Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

(6) 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 attachments 6, 7, and 8.

(7) The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, 1-131, Cs-1 34, Cs-137, Ce-141, and Ce-144 in Iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent Release Report.

If two or more redundant sample and analysis results of the same discharge pathway indicate slight differences in low level isotopic mixture, the analysis resulting in higher calculated dose should be used. The isotopes chosen to represent the discharge shall be reported (8) Grab samples shall be taken and analyzed at least once per 31 days whenever there is flow through the steam generator blowdown tank. Releases of radioiodines shall be estimated based on secondary coolant concentration and partitioning factors during releases or shall be measured.

(9) Frequency Notation Notation Frequency W At least once per 7 days M At least once per 31 days Q At least once per 92 days P Completed prior to each release CY2!lDlu3rl I.DOC 1209.0951

CY2.ID1 Rll Page 27 of 37 Gaseous Effluents - Dose Rate Operational Requirements Attachment 6: Page 4 of 4 Bases: Gaseous Effluents - Dose Rate This Regulatory Commitment is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of NUREG 0133 to UNRESTRICTED AREAS. The annual dose limits are the doses to be associated with the concentrations of 10 CFR Part 20, Appendix B, Table 2, Column 1. These limits provide 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, to average concentrations exceeding the limits to be specified in Appendix B, Table 2 of 10 CFR Part 20 (10 CFR Part 20.1302(c)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCP. The specified release rate limits of NUREG 0133 restrict, presently, 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 mrems/year to the whole body or to less than or equal to 3000 mrems/year to the skin.

These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/year.

This Regulatory Commitment applies to the release of radioactive materials in gaseous effluents from all units at the site.

The required detection capabilities for radioactive material in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in Currie, L.A., "Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements," NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revised annually).

CY20ID1u3rl 1 .DOC 1209.0951

CY2.1D1 R11 Page 28 of 37 Dose - Noble Gases Operational Requirements Attachment 7: Page 1 of 1

1. Dose - Noble Gases

a. Commitment for Operation

1) The air dose due to noble gases released in gaseous effluents, from each unit, to areas at or beyond the SITE BOUNDARY (see FSAR Figure 2.1-2) shall be limited to the following:

a) During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and b) During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

2) Applicability: At all times.

3) Action:

a) With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to 10 CFR 50.4, a Special Report that identifies the cause(s) for exceeding the limit(s), defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. Surveillance Requirements

1) Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCP at least once per 31 days.

Bases: Dose - Noble Gases This Regulatory Commitment is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I. The ACTION statements provide the required operation flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." 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. The dose calculation methodology and parameters established in the ODCP for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, "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 Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. The ODCP equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

This Regulatory Commitment applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are proportioned among the units sharing the system.

CY2!ID1u3rl 1.DOC 1209.0951

CY2.1D1 R11 Page 29 of 37 Iodine 131, Iodine 133, Tritium, and Radioactive Material in Particulate Form Attachment 8: Page 1 of 2

1. Dose - Iodine-131, Iodine-i133, Tritium, and Radioactive Material in Particulate Form

a. Commitment for Operation

1) The dose to a MEMBER OF THE PUBLIC from Iodine-1 31, Iodine-1 33, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 2.1-2) shall be limited to the following:

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.

2) Applicability: At all times.

3) Action:

a) With the calculated dose from the release of Iodine-1 31, Iodine-1 33, tritium, and radionuclides 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, pursuant to 10 CFR 50.4, a Special Report that identifies the cause(s) for exceeding the limit(s), defines the corrective actions that have been taken to reduce the releases and the proposed actions to be taken to assure that subsequent releases will be in compliance With the above limits.

b. Surveillance Requirements

1) Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-1 31, Iodine-1 33, tritium, and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCP at least once per 31 days.

CY2!ID1u3rl 1.DOC 1209.0951

CY2.1D1 R11 Page 30 of 37 Iodine 131, Iodine 133, Tritium, and Radioactive Material in Particulate Form Attachment 8: Page 2 of 2 Bases: Dose - Iodine-1 31, Iodine-1 33, Tritium, and Radioactive Material in Particulate Form This Regulatory Commitment is provided to implement the requirements of Sections II.C, Ill.A, and IV.A of Appendix 1,10 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 effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCP 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. The ODCP calculational methodology and parameters for calculating the doses due to the actual 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 Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,: Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate specifications for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of the calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy 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.

This Regulatory Commitment applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

CY2!IDlu3rl IDOC 1209.0951

CY2.1D1 R11 Page 31 of 37 Gaseous Radwaste Treatment System Operational Requirements Attachment 9: Page 1 of 1

1. Gaseous Radwaste Treatment System

a. Commitment for Operation

1) The GASEOUS RADWASTE SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see FSAR Figure 2.1-2), would exceed 0.2 mrad to air from gamma radiation or 0.4 mrad to air from beta radiation or 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

2) Applicability: At all times.

3) Action:

a) With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to 10 CFR 50.4, a Special Report that includes the following information:

(1) Identification of the inoperable equipment or subsystems and the reason for inoperability, (2) Action(s) taken to restore the inoperable equipment to OPERABLE status, and (3) Summary description of action(s) taken to prevent a recurrence.

b. Surveillance Requirements

1) Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days, in accordance with the methodology and parameters in the ODCP when Gaseous Radwaste Treatment Systems are not being fully utilized.

2) The installed VENTILATION EXHAUST TREATMENT SYSTEM and GASEOUS RADWASTE SYSTEM shall be considered OPERABLE by meeting attachments 6, 7, or 8.

Bases: Gaseous Radwaste Treatment System The OPERABILITY of the GASEOUS RADWASTE SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections 1I.B and II.C of Appendix 1,10 CFR Part 50, for gaseous effluents.

This Regulatory Commitment applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.

CY2!lDlu3rl .0DOC1209.0951

CY2.1D1 Rll Page 32 of 37 Total Dose Operational Requirements Attachment 10: Page 1 of 2

1. Total Dose

a. Commitment for Operation

1) The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.

2) Applicability: At all times.

3) Action:

a) With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of attachments 4, 7, or 8, calculations shall be made including direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of Attachment 10 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to 10 CFR 50.4, 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 20.2203(a), shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathway 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 granted until staff action on the request is complete.

b. Surveillance Requirements

1) Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with attachments 4, 7, and 8, and in accordance with the methodology and parameters in the ODCP.

2) Cumulative dose contributions from direct radiation from the units and from outside storage tanks shall be determined in accordance with the methodology and parameters in the ODCP. This requirement is applicable only under conditions set forth in ACTION a) of Attachment 10.

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CY2.ID1 R11 Page 33 of 37 Total Dose Operational Requirements Attachment 10: Page 2 of 2 Bases: Total Dose This Regulatory Commitment is provided to meet the dose limitations of 40 CFR Part 190 that have 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 due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units and from outside storage tanks are kept small. 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 contribution 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 190.11 and 10 CFR 20.2203(a), 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 attachments 3 and 6. 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.

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CY2.ID1 R11 Page 34 of 37 Radiological Environmental Monitoring Operational Requirements Attachment 11: Page I of I

1. Radiological Environmental Monitoring

a. Commitment for Operation

1) The Radiological Environmental Monitoring Program shall be conducted as specified in RP1 .ID11, "Environmental Radiological Monitoring Procedure."

2) Applicability: At all times.

3) Action:

a) With the confirmed level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the "Reporting Levels for Nonroutine Operating Reports" in RP1 .ID11 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days from the end of the quarter, pursuant to 10 CFR 50.4, a Nonroutine Radiological Environmental Operating Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of attachments 4, 7, or 8,. A confirmatory reanalysis of the original, a duplicate, or a new sample may be desirable, as appropriate. The I

results of the confirmatory analysis shall be completed at the earliest time consistent with the analysis, but in any case within 30 days. When more than one of the radionuclides from "Reporting Levels for Nonroutine Operating Reports" in RP1 .ID11 are detected in the sampling medium, this report shall be submitted if:

concentraton + con centration(2) .. 10 When radionuclides other than those in the "Reporting Levels for Nonroutine Operating Reports" in RP1 .ID11 are detected and are the result of plant effluents, a Nonroutine Radiological Environmental Operating Report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar year limits of attachments 4, 7, or 8,. This report shall include an evaluation of any release conditions, environmental factors, I

or other aspects necessary to explain the anomalous result.

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CY2.1D1 R11 Page 35 of 37 Land Use Census Operational Requirements Attachment 12: Page 1 of 1

1. LAND USE CENSUS

a. A Land Use Census shall be conducted as specified in RP1.ID11, "Environmental Radiological Monitoring Procedure."

1) Applicability: At all times.

2) Action:

a) With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Attachment 8, identify the new location(s) in the next Annual Radioactive Effluent Release Report.

b) With a Land Use Census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Attachment 11, add the new location(s) within 30 days to the Radiological Environmental Monitoring Program given in the ERMP. The sampling location(s),

excluding the control station location, having the lowest calculated dose or dose commitment(s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ERMP including a revised figure(s) and table(s) for the ERMP reflecting the new location(s) with information supporting the change in sampling locations.

CY2!ID1 u3rl 1.D0C 1209.0951

CY2.1D1 R11 Page 36 of 37 High Alarm Setpoints - FB & CR Ventilation Systems Actuation Instrumentation Attachment 13: Page 1 of 2

1. Fuel Building Ventilation System (FBVS) Instrumentation, RE-58 and RE-59

a. RE-58 Nominal Setpoint < 75 mr/hr

1) Bases a) Fuel Handling Accident in Fuel Handling Building (1) The basis for the RE-58 high alarm setpoint is to initiate actions to mitigate offside dose consequences from air borne releases resulting from a fuel handling accident in the Spent Fuel Pool area. Routing ventilation exhaust from the Spent Fuel Pool area through the charcoal filter, thus stripping halogens (principally iodine isotopes) mitigates off-site dose consequences.

The rerouting of the ventilation is accomplished automatically upon receipt of a RE-58 high alarm. Receipt of the high alarm also signals personnel to evacuate the area. PG&E performed a calculation (RA-90-1 -0 "High and Alert Alarm Setpoint for RE-58") to base the high alarm setpoint of RE-58 on the airborne radioactivity concentration in the fuel Handling Building for the FSAR Update Expected Case accident release during a fuel handling accident. The Expected Case Accident consequence presented in the FSAR Update is a less severe, but more probable accident than the FSAR Update Design Basis Case fuel handling accident. This resulted in a more conservative (lower) setpoint than that for the Design Basis Accident Case. This calculation analyzed the detector sensitivity to the various release nuclides as presented in the FSAR Update.

(2) The high alarm setpoint is set to a value more consistent with the Expected Case Accident dose rates which eliminates spurious ESF actuation while limiting the off-site consequences due to this accident. A < 75mr/hr setpoint equates to a site boundary whole body dose of 1.46mr for the duration of the accident, which has been evaluated as being acceptable.

b) Inadvertent Criticality in the Spent Fuel Pool (1) The high density Spent Fuel Pool racks were redesigned to assure that a Ke, equal to or less than 0.95 is maintained with the racks fully loaded with fuel of the highest anticipated reactivity in each of two regions, and flooded with unborated water at a temperature corresponding to the highest reactivity.

PG&E submitted PG&E Letter No. DCL-85-30, "Re-racking of Spent Fuel Pools," on September 19, 1985. The results of the criticality analysis for normal and abnormal operations were evaluated in this report. LAR 85-13 (PG&E Letter No. DCL-85-333) was submitted on October 30, 1985, and summarized the results of the Spent Fuel Pool re-racking report. Postulated events that could potentially involve accidental criticality were examined and it was concluded that the limiting value for criticality (Keff of 0.95) would not be exceeded. Therefore, an inadvertent criticality in the Spent Fuel Pool is not considered a credible accident and an evaluation of the effect of raising the actuation setpoint on RE-58 was not required.

(2) Radiation monitor RE-59, which monitors the new fuel storage vault area, also provides indication of inadvertent criticality and changes in dose rate for radiation protection purposes.

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CY2.ID1 R11 Page 37 of 37 High Alarm Setpoints - FB & CR Ventilation Systems Actuation Instrumentation Attachment 13: Page 2 of 2

b. RE-59 Nominal Setpoint - 15mr/hr

1) Bases a) In DCPP FSAR update in Chapter 12, Table 12.1-1, "Plant Zone Classifications,"

the Fuel Handling Building areas in which RE-59 is located is classified as Zone III, "Controlled Assess Requiring Short-Term Occupancy" as indicated in Design Criteria Memorandum DCM-T20.

b) The Zone III design maximum dose rate is < 15mr/hr. If the radiation flux reaches this value, the high alarm will be actuated and the ventilation mode will change as indicated above with the RE-58 high alarm actuation.

2. Control Room Ventilation System (CRVS) Instrumentation, RS-25A and RS-26A

a. RS-25A/RS-26A Nominal Setpoint-< 2 mr/hr

1) Bases a) Per calculations DV-1-23 and DV-2-23 attachment 2:

(1) The applicable NRC requirement for the radiation exposure (dose) to personnel in the Main Control Room is 10 CFR 20.105 "Permissible Level of Radiation in Unrestricted Areas" (superseded). Section b (1) of 10 CFR 20.105 limits the radiation dose for unrestricted areas to 2mr/hr and Section b (2) limits the dose to 100mr is seven consecutive days. If an operator works in the control room for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months per week (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per day and 4 days per week, a normal operator work week) during a 2mr/hr dose rate, the operator will receive a 96mr dose. This is less than the 100mr/week dose limit of 10 CFR 20.105 (superseded). Thus, a setpoint of < 2 mr/hr has been established for the Control Room Air Inlet Radiation Monitors. If the radiation flux reaches this value at any of the detectors, a change in the ventilation to Mode 4 will be initiated.

(2) 10 CFR 20.105 has been superseded and any changes to this setpoint will require a basis change to reflect the requirements of 10 CFR 50 Appendix A Criterion 19.

3. All of the setpoints are controlled by the setpoint control program CF6.1D1 and require a design change vehicle (request per CF4.1D1) to change. Actual field setpoints are set more conservative to account for instrument errors.

CY2!ID1 u3rl 1.DOC 1209.0951

Attachment 3 PG&E Letter DCL-1 1-049 Attachment 3 Diablo Canyon Power Plant Interdepartmental Administrative Procedure, RPI.ID11, "Environmental Radiological Monitoring Procedure," Revision 10

- - ISSUED FOR USE BY: DATE: EXPIRES:_***

DIABLO CANYON POWER PLANT RP1.ID11 INTERDEPARTMENTAL ADMINISTRATIVE PROCEDURE Rev. 10 Page 1 of 27 Environmental Radiological Monitoring Procedure 11/08/10 Effective Date QUALITY RELATED Table of Contents

1. S C O P E ..................................................................................................... . . 1

2. D IS C US S IO N ............................................................................................... 2

3. D E FIN IT IO N S ............................................................................................ 2

4. RESPO NS IBILITIES ................................................................................... 3

5. INST R UCT IO NS ........................................................................................ 5

6. R EC O R D S ................................................................................................ 10

7. R E FE R EN C ES ......................................................................................... 10

8. Radiological Environmental Monitoring Program ....................................... 11

9. Detection Capabilities / Lower Limits of Detection (LLD) ........................... 17

10. AREOR REMP Program Summary Matrix ................................................ 19

11. Reporting (Notification) Levels for REMP .................................................. 20

12. Distances and Directions to REMP Monitoring Stations ............................ 21

13. Summary of Cross-Check Program with State of California ...................... 24 ATTACHMENTS:

1. DCPP Onsite ERMP Stations .................................................................... 25

2. DCPP Offsite ERMP Stations .................................................................... 26

3. DCPP Onsite ERMP Stations Satellite Map .............................................. 27

1. SCOPE 1.1 This procedure constitutes the Environmental Radiological Monitoring Procedure (ERMP) and implements the requirements of the Radiological Environmental Monitoring Program (REMP) for the plant per Technical Specification 5.5.1, 5.6.2, and the ISFSI Environmental Report, Section 6.2. This procedure is part of the Offsite Dose Calculation Manual (ODCM).

1.2 This document was converted; therefore, revision bars are not included.

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Environmental Radiological Monitoring Procedure RP1.ID11 R10 Page 2 of 27

2. DISCUSSION 2.1 The purpose of a Radiological Environmental Monitoring Program is to provide a basis for evaluating concentrations of radioactive materials and radiation levels in the environment from radiological releases once a reactor is operational.

2.2 This procedure describes the supplemental and minimum required program, describing sample locations, types of sample locations, methods and frequency of analysis, reporting requirements, performance of land use census, and participation in an interlaboratory comparison program.

2.3 This procedure also contains the state of California Department of Public Health (CDPH) cross-check program.

2.4 The program described by this procedure provides measurements of radiation and of radioactive materials in those exposure pathways and for those nuclides that lead to the highest potential radiation exposures of members of the public resulting from plant operation.

This monitoring program implements section IV.B.2, IV.B.3 and IV.C of Appendix I to 10 CFR 50 and supplements the Radioactive Effluent Control Program. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Radiological Environmental Monitoring, revision 1, November, 1979. Due to DCPP site characteristics, some program requirements may vary from the Branch Technical Position on REMP (BTP 1979).

2.5 The Diablo Canyon ODCM is made up of the following procedures:

2.5.1 CY2, "Radiological Monitoring and Controls Program" 2.5.2 CY2.1D1, "Radioactive Effluent Controls Program" 2.5.3 RP1.ID11, "Environmental Radiological Monitoring Procedure" 2.5.4 CAP A-8, "Off-site Dose Calculation Procedure"

3. DEFINITIONS 3.1 Broadleaf Vegetation: The leafy portion of plants such as lettuce, cabbage, spinach, or other similar leafy plants, typically used for human consumption.

3.2 Supplemental Sample: Samples that should be collected when available, but when they are not available, they are not subject to the substitution requirements of Table 1, Note 1. These types of samples are specifically designated as "supplemental" in Table 1.

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Environmental Radiological Monitoring Procedure RP1.ID11 R10 Page 3 of 27

4. RESPONSIBILITIES 4.1 Radiation protection manager is responsible for ensuring implementation of this procedure, including approving the Annual Radiological Environmental Operating Report (AREOR) prior to its submittal to the NRC. This should be a signature page at the front of the AREOR.

4.2 REMP engineer is responsible for the following:

4.2.1 Identifying and evaluating the significant pathways of radiological impact to man and biota (e.g. food vectors, recreational use, water use) and subsequent updating of the REMP.

4.2.2 Ensuring the performance of the annual land use census and that the results are provided to chemistry so that chemistry can establish the dose requirements of CY2.1D1," Radioactive Effluent Controls Program."

4.2.3 Preparation of the Annual Radiological Environmental Operating Report (AREOR).

4.2.4 Ensuring that changes to the ERMP are provided to chemistry for inclusion in the Annual Radiological Effluent Release Report.

4.2.5 Ensuring the required REMP sampling and monitoring are performed.

4.2.6 Ensuring that the environmental TLD program is maintained.

4.2.7 Implementing and communicating contracts with the environmental vendor lab (REMP).

4.2.8 Ensuring that the personnel responsible for the management and for the implementation of the REMP receive training on the changes to the REMP at least annually.

a. Such training consists of topics related to the changes in the plant's REMP procedures, industry events, any changes in technology that pertain to REMP sampling techniques or to the analysis of REMP samples and the nature and goals of the quality assurance program.

b. Proficiency of personnel who perform activities affecting the quality of the REMP can be maintained by retraining, reexamining, recertifying, or by periodic performance reviews as appropriate.

c. Initial training is provided on an as needed basis to new personnel responsible for quality related REMP activities.

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Environmental Radiological Monitoring Procedure RP1.ID11 R10 Page 4 of 27 4.3 REMP environmental lab vendor is responsible for the following:

4.3.1 Ensuring analysis is performed for the samples as required by the ERMP.

4.3.2 Ensuring that participation is maintained in an interlaboratory comparison program sufficient to satisfy step 5.3.1 of this procedure.

4.3.3 Ensuring that REMP sample results exceeding the criteria of RP1 .ID1 1, "Environmental Radiological Monitoring Procedure" Table 4, are communicated promptly to DCPP as per the contract.

4.3.4 Ensuring communication with DCPP per the contract. Examples include:

Sample results not meeting the contract "a-priori" LLDs

" Exceeding confract notification levels

Problems with lab REMP sample analyses

" Problems with sample shipments

Interlaboratory comparison program issues 4.3.5 Ensuring that appropriate procedures are established and maintained for sample handling, sample analysis and all associated laboratory equipment.

4.3.6 Ensuring qualifications and training of vendor lab personnel.

4.3.7 Ensuring contract compliance.

4.4 DCPP regulatory services department is responsible for ensuring that reports required by the ERMP are submitted to the appropriate recipients per Xl1 .I02, "Regulatory Reporting Requirements and Reporting Process."

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Environmental Radiological Monitoring Procedure RP1.ID11 R10 Page 5 of 27

5. INSTRUCTIONS 5.1 Sampling and Monitoring NOTE: Sampling procedures for the REMP pathways and OSGSF sump monitoring are controlled in the DCPP plant manual, Volume 7 "Radiation Protection-Radiation Control Procedures," "RCP EM" series of procedures.

5.1.1 Environmental samples shall be collected and analyzed according to Table 1 at the locations shown in Table 5 and Figure 1, (Attachment 1) and Figure 2 (Attachment 2).

a. Analytical techniques used shall be such that the detection capabilities in Table 2 are routinely achieved.

b. The sampling frequencies specified in Table 1 are allowed an extension of 25 percent (NUREG 1301, Page 16, Section 4.0.2).

5.1.2 The Old Steam Generator Storage Facility (OSGSF) inspection sumps shall be monitored quarterly to ensure there is no standing water in the sumps. If water is found: (Ref A0719469)

a. Initiate the corrective action process.

b. Perform isotopic analysis for plant related isotopes.

c. Disposition the water per plant protocols.

5.2 Land Use Census (LUC) 5.2.1 The land use census satisfies the requirements of section IV.B.3 of Appendix I to 10 CFR 50.

a. Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via broadleaf vegetation will be identified and monitored as necessary.

b. The size of the garden is the minimum to produce the quantity of 26 kg/year of broadleaf assumed in Regulatory Guide 1.109 for consumption by a child.

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Environmental Radiological Monitoring Procedure RP1.ID11 R10 Page 6 of 27 5.2.2 A land use census shall be conducted at least once per year, during the growing season between February 15 and December 1. The LUC will determine the locations in each of the 16 meteorological sectors within a distance of 8 km (5 mi) for:Ref A0565194

a. The nearest milk animal.

J

b. The nearest residence.

c. The nearest garden of greater than 50 square meters (500 sq. ft.) producing broadleaf vegetation.

NOTE: Broadleaf vegetation sampling may be performed at the site boundary in a sector with the highest D/Q in lieu of the garden census portion of the land use census.

5.2.3 The land use census shall be conducted using that information which will provide the best results, which typically consists of discussions with landowners/tenants in conjunction with an aerial survey. Local agricultural authorities may also be consulted.

5.2.4 If the land use census identifies a location(s) that yields a calculated dose or dose commitment greater than the values currently being calculated per CY2.1D1, the new location(s) shall be identified in the next Annual Radioactive Effluent Release Report.

5.2.5 If the land use census identifies a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained per CY2.1D1, add the new location(s) within 30 days to the radiological environmental monitoring program given in this ERMP.

a. The sampling location(s), excluding the control station location, having the lowest calculated dose or dose commitment(s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted.

b. Documentation for a change in the ERMP shall be submitted in the next Annual Radioactive Effluent Release Report including a revised figure(s) and table(s) for the ERMP reflecting the new location(s) with information supporting the change in sampling locations.

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Environmental Radiological Monitoring Procedure RP1.ID11 RIO Page 7 of 27 5.3 Cross-Check Programs 5.3.1 Participation shall be maintained in an interlaboratory comparison program. This participation should include each of the determinations (sample medium-radionuclide combination) as described in Table 1 to ensure independent checks on the precision and accuracy of the measurements of radioactive materials in the ERMP samples.

5.3.2 Participation shall be maintained in a split-sampling program with the State of California Department of Public Health (CDPH) as required by CDPH.

a. The program shall consist of radiological analyses of duplicate (split) samples or replicate analyses of the same sample by both the REMP environmental lab vendor and CDPH.

b. The sample results from the REMP environmental lab vendor are available to CDPH in the DCPP Annual Radiological Environmental Operating Report (AREOR) or as requested by CDPH.

c. The sample types, analyses performed, and frequencies of collection are given in Table 6.

5.4 Annual Radiological Environmental Operating Report (AREOR) 5.4.1 A report on the radiological environmental monitoring program shall be prepared annually and submitted to the NRC prior to May 1 of the following year per DCPP Tech Spec 5.6.2.

5.4.2 The Annual Radiological Environmental Operating Report shall include:

a. Summaries, interpretations, and an analysis of trends of the results of the radiological environmental monitoring program activities for the report period.

b. A comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports.

c. An assessment of the observed impacts of the plant operation on the environment.

d. The results of the land use census.

e. The results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in tables and figures in this ERMP as well as summarized and tabulated results of these analyses and measurements in the format of Table 3 or equivalent.

f. A summary description of the radiological environmental monitoring program.

g. Legible maps covering all sampling locations keyed to a table giving distances and directions from the centerline of Unit One Reactor.BeeSAPN 5035 6855 RP1!1D11u3r10.DOC 1030.1655

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h. The results of licensee or REMP analysis lab vendor participation in the interlaboratory comparison program and the corrective action taken ifthe specified program is not being performed as required.

i. The reason for not conducting the radiological environmental monitoring program as required, and discussion of all deviations from the sampling schedule of Table 1, including plans for preventing a recurrence.

j. A discussion of environmental sample measurements that exceed the reporting levels of Table 4, but are not the result of plant effluents (i.e.,

demonstrated by comparison with a control station or with preoperational data).

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

I. Signature approval of the AREOR by the DCPP RP manager.

m. Results of the Old Steam Generator Storage Facility (OSGSF) quarterly sump monitoring for standing water in vault sumps. Report plant related isotopic activity detected and disposition of water. Ref A071 9469

n. Combined percent availability of REMP airsamplers. This is done by reporting the percentage of actual runtime (as compared to available runtime) 3 during the year for all REMP airsamplers as a combined percentage.RefT 1 39

o. Routine groundwater radiological monitoring as performed for the industry NEI Groundwater Protection Initiative 07-07 (GPI) shall be reported in the AREOR.

p. All on-site or off-site groundwater sample results that exceeded the REMP reporting thresholds in Table 4 that were communicated per NEI Groundwater Protection Initiative 07-07 (Objective 2.2) shall be included in the AREOR 5.4.3 The Annual Radiological Environmental Operating Report (AREOR) shall be distributed to the following agencies:Ref A0619597 NOTE: The May 1st deadline for submittal applies to the NRC only.

" Nuclear Regulatory Commission (NRC)

" Chief, Radiological Health Branch, CA Dept of Public Health (CDPH)

Executive Officer, Regional Water Quality Control Board, Central Coast

Executive Officer, San Luis Obispo County Air Pollution Control District

San Luis Obispo County Health Officer (Environmental Health Dept)

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Environmental Radiological Monitoring Procedure RP1.ID11 R10 Page 9 of 27 5.5 Nonroutine Reports 5.5.1 Supplementary Report

a. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

1. The missing data shall be submitted as soon as possible in a supplementary report.

5.5.2 Nonroutine Radiological Environmental Operating Report

a. If a measured radionuclide concentration resulting from plant effluents in an environmental sampling medium, averaged over any calendar quarter sampling period, exceeds the reporting level given in Table 4, or when radionuclides other than those in Table 4 are detected and are the result of plant effluents where the potential annual dose to a member of the public from all radionuclides is equal to or greater than the calendar year limits of Appendix I of 10 CFR 50, a nonroutine radiological environmental operating report shall be prepared per CY2.ID1.

1. This report is not required ifthe 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 operating report.

5.6 ERMP Chancqes 5.6.1 Changes to this procedure shall be processed per the requirements of the plant Technical Specification Section 5.5.1.

5.6.2 Notification of Sample Unavailability NOTE: The currently approved ERMP allows for substitution of milk and broadleaf vegetation sampling with additional air sampling in two sectors.

a. If milk or vegetation sampling is not being performed as required by Table 1, identify specific locations for obtaining replacement samples and add them within 30 days to the radiological environmental monitoring program given in the ERMP.

1. The specific locations from which samples were unavailable may then be deleted from the monitoring program.

2. Submit in the next Annual Radioactive Effluent Release Report documentation for a change in the ERMP including a revised figure(s) and table for the ERMP reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location(s) for obtaining samples.

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6. RECORDS 6.1 Performance of the land use census shall be documented, including as a minimum the names of persons contacted and dates of contact. This documentation does not need to appear in the required reports, but should be entered into the records management system for the duration of the plant license following the guidance of AD1 0.ID1 "Storage and Control of Quality Assurance Records".

6.2 Changes to the ERMP shall be retained in the records management system for the duration of the plant operating license following the guidance of AD10.1D1 "Storage and Control of Quality Assurance Records".

6.3 REMP training shall be documented and recorded per plant protocols. Training records shall also be placed in the Records Management System (RMS) following the guidance of AD1 0.1D1 "Storage and Control of Quality Assurance Records".

7. REFERENCES 7.1 RMS RLOC 001157-1162 & 001157-1163, "State Acceptance Letter" 7.2 Corrective Actions:

7.2.1 SAPN 50032742, "REMP audit dates and numbers, important document locations 7.3 DCPP ISFSI, License SNM-251 1, Environmental Report, Docket 72-26 7.4 PCD T02694, T04341, T31239, T35263, T36656 RPI!ID1lu3rlO.DOC 1030.1655

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8. Radiological Environmental Monitoring Program Table 1: Radiological Environmental Monitoring Program Matrix Exposure Pathway Number of Representative Sampling Collection Required or and/or Sample Type Samples and Sample Locations1 Stations Frequency Type of Analysis Supplemental 1 Direct Thirty-one routine monitoring Radiation stations containing thermo luminescent dosimeters (TLDs) such that at least two (2) phosphors are present at each station, placed as follows:

An inner ring of stations, one in OS1,0S2, Quarterly Gamma Dose Required each terrestrial meterorological WN1, 1S1, sector in the general area of the 2S1, 3S1, 4S1, SITE BOUNDARY; 5S1, 5S3, 6S1, 7S1,8S1,8S2, 9S1, and MT1 An outer ring of stations, one in 0B1, 1A1, 1C1, Quarterly Gamma Dose Required each terrestrial meterorological 2D1,3D1,4C1, sector in the 2.5 to 14 km range 5C1,6D1, and from the site; and 7C1 One or two areas to serve as 4D1,5F1 Quarterly Gamma Dose Required control stations; and The balance of the stations to be 5F3, 7D1, 7D2, Quarterly Gamma Dose Required placed in special interest areas 7F1, and 7G2 such as population centers, nearby residences, or schools.

A minimum of four stations IS1, IS2, IS3, Quarterly Gamma Dose Required around the ISFSI IS4, IS5, IS6, IS7,IS8

2. Airborne Samples from > 4 stations: 0S2, 8S1, Continuous 1-131 analysis Required Radioiodine MT1, 7D1 sampler operation Three samples from close to the with sample three SITE BOUNDARY 5F1 (control) collection weekly, locations (0S2, 8S1, & MT1) in or more different sectors. frequently if One sample from the vicinity of a required by dust community having the highest loading.

calculated annual average ground level D/Q (7D1).

Iffood products are unavailable, 1S1 & 8S2 Continuous 1-131 analysis Required additional air sampling will be sampler operation done in the NNW (station 1S1) with sample and SE (Station 8S2) sectors. collection weekly, or more frequently if required by dust loading.

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Exposure Pathway Number of Representative Sampling Collection Required or and/or Sample Type Samples and Sample Locations1 Stations Frequency Type of Analysis Supplemental

3. Airborne Samples from > 4 stations: 0S2, 8S1, Continuous Weekly gross Required Particulate MT1, 7D1 sampler operation beta radioactivity Three samples from close to the with sample analysis following three SITE BOUNDARY 5F1 (control) collection weekly, filter change 3 .

locations (0S2, 8S1, & MT1) in or more Quarterly gamma4 different sectors. frequently if isotopic analysis One sample from the vicinity of a required by dust of composite community having the highest loading, consisting of calculated annual average approx 12 filters ground level D/Q (7D1). (by location).

If food products are unavailable, 1S1 & 8S2 Continuous Weekly gross Required additional air sampling will be sampler operation beta radioactivity done in the NNW (station 1S1) with sample analysis following and SE (Station 8S2) sectors. collection weekly, filter change 3 .

or more Quarterly gamma4 frequently if isotopic analysis required by dust of composite loading, consisting of approx 12 filters (by location).

4. Waterborne

a. Surface One sample from the plant OUT, DCM, Monthly Gamma isotopic 4 Required Ocean Outfall, Diablo Cove, and an and 7C2 (grab sample) and tritium Water area not influenced by plant analysis.

discharge.

One sample from the plant OUT, DCM, Quarterly Gross Beta, Supplemental Outfall, Diablo Cove, and an and 7C2 (grab sample) Total Sr, area not influenced by plant Fe-55, and Ni-63 discharge.

b. Drinking One sample from the plant DW1 and 5S2; Monthly Gamma isotopic4 , Required Water drinking water, one sample from OEL (control) (grab sample) 1-131, and tritium Diablo Creek (upstream of plant), analysis.

and one control sample.

One sample from the plant DW1 and 5S2; Quarterly Gross Beta, Supplemental drinking water, one sample from OEL (control) (grab sample) Total Sr, Diablo Creek (upstream of plant), Fe-55, and Ni-63 and one control sample.

One sample from Diablo Creek WN2 and 1A2 Quarterly Gamma isotopic 4 , Supplemental (downstream of plant) and one (grab sample) tritium, 1-131, sample from Blanchard Spring gross beta, Total Sr, Fe-55, and Ni-63 RP1!1Dllu3rl0.DOC 1030.1655

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Exposure Pathway Number of Representative Sampling Collection Required or and/or Sample Type Samples and Sample Locations 1 Stations Frequency Type of Analysis Supplemental

c. Groundwater One sample from wells located OWl, OW2, Quarterly Gamma isotopic 4 , Supplemental under the plant power block, and DY1 (grab sample, tritium, gross when available) beta, Total Sr, Fe-55, and Ni-63 One sample from a well located WW2, 8S3 Quarterly Gamma isotopic 4 , Supplemental outside the plant power block (grab sample, tritium, gross (control sample). when available) beta, Total Sr, Fe-55, and Ni-63

d. Sediment One sample of offshore ocean DCM and 7C2 Annual Gamma isotopic 4 Required sediment from Diablo Cove and (grab sample)

Rattlesnake Canyon.

One sample of offshore ocean DCM and 7C2 Annual Total Sr, Fe-55, Supplemental sediment from Diablo Cove and (grab sample) and Ni-63 Rattlesnake Canyon.

One sample from each of five AVA, MDO, Semi- Annual Gamma isotopic 4 , Supplemental local recreational beaches. PMO, CYA, (grab sample) Total Sr, Fe-55, and CBA and Ni-63

e. Marine Flora One sample of kelp DCM, PON, Quarterly Gamma isotopic 4 Supplemental POS, and 7C2 (when available)

One sample of intertidal algae DCM and 7C2 Quarterly Gamma isotopic 4 Supplemental (when available)

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Exposure Pathway Number of Representative Sampling Collection Required or and/or Sample Type Samples and Sample Locations1 Stations Frequency Type of Analysis Supplemental

5. Ingestion

a. Milk Samples from milking animals in 5F2 Semimonthly Gamma isotopic 4 Supplemental three locations within 5 km when animals are and 1-131 distance having the highest dose on pasture; analysis.

potential. If there are none, then monthly at other one sample from milking animals times.

in each of three areas between 5 to 8 km distance where doses are calculated to be greater than 1 mrem per year. One sample from milking animals at a control location 15 to 30 km distant and in the least prevalent wind direction.

NOTE: The sample (5F2) should be taken monthly even if there are no indicator samples available.

b. Fish and One sample of rock fish (family DCM and 7C2 Quarterly Gamma isotopic 4 Required Invertebrates Sebastes) and one sample of (grab sample) analysis on edible perch (family Embiotocidae) portions of each sample.

One sample of rock fish PON and POS Quarterly Gamma isotopic 4 Supplemental (family Sebastes) and (grab sample) analysis on edible one sample of perch portions of each (family Embiotocidae) sample.

One sample of mussel DCM and 7C2 Quarterly Gamma isotopic 4 Required (family Mytilus) (grab sample) analysis on edible portions of each sample.

One sample of mussel PON Annual (Ao584392) Gamma isotopic 4 Supplemental (family Mytilus) (grab sample) analysis on edible portions of each sample.

One sample of mussel POS Quarterly Gamma isotopic 4 Supplemental (family Mytilus) (grab sample) analysis on edible portions of each sample.

One sample of locally harvested 7D3 or 2F1 Quarterly Gamma isotopic 4 Supplemental market fish. (should (grab sample) analysis on edible alternate portions of each between sample.

locations)

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Exposure Pathway Number of Representative Sampling Collection Required or and/or Sample Type Samples and Sample Locations 1 Stations Frequency Type of Analysis Supplemental

c. Broadleaf Three samples of broadleaf Monthly Gamma isotopic 4 Required Vegetation 5 vegetation grown nearest off-site (when available) analysis (that (see notation locations of highest calculated includes 1-131) on #5) annual average ground level D/Q edible portion.

IF milk sampling is not performed.

One sample of each of the Monthly Gamma isotopic 4 Required similar broadleaf vegetation (when available) analysis (that (see notation grown 15 to 30 km distant in the includes 1-131) on #5) least prevalent wind direction IF edible portion.

milk sampling is not performed.

d. Vegetative Crops One sample of broadleaf 5F2, 7C1, and Monthly Gamma isotopic 4 Supplemental vegetation or vegetables or fruit 7G1 (when available) analysis on edible portion.

One sample of broadleaf 3C1, 6C1 Quarterly Gamma isotopic 4 Supplemental vegetation or vegetables or fruit. (as provided by analysis on edible land owner) portion.

e. Meat One sample of each species BCM, BGM, Quarterly Gamma isotopic 4 Supplemental (cow, goat, sheep, deer, or pig) BSM, JDM, (when available analysis, and of edible meat portion JPM, ACM, and provided by Total Sr on edible slaughtered for personal ADM, APM land owners portion.

consumption (not mass market). within 8 km of plant site)

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Table 1 Notes:

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 sampling schedule shall be documented in the Annual Radiological Environmental Operating 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 Radiological Environmental Monitoring Program, and submitted in the next Annual Radioactive Effluent Release Report, including a revised figure(s) and table for the ERMP 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.

2 For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor. There are normally three calcium sulfate phosphors in an environmental TLD BADGE. Film badges shall not be used as dosimeters for measuring direct radiation.

3 Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

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

5 If broadleaf vegetation food products are unavailable, additional air sampling as specified in Table 1, Parts 2 & 3 will be done in the NNW (Station 1Si) and SE (Station 8S2) sectors.

6 The Branch Technical Position (Nov 79) states, "Any location from which milk can no longer be obtained may be dropped from the surveillance program after notifying the NRC in writing that they are no longer obtainable at that location". Although the milk sampling performed at 5F2 is outside the 5-mile radius and is supplemental to the REMP, this notification should take place if 5F2 milk sampling ceases.

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9. Detection Capabilities / Lower Limits of Detection (LLD)

Table 2: Detection Capabilities / Lower Limits of Detection (LLD) Matrix Airborne Water Particulate or Fish Milk Food Products Sediment Analysis (pCi/L) Gases (pCi/m 3) (pCi/kg, wet) (pCi/L) (pCi/kg, wet) (pCi/kg, dry)

Gross beta 4 0.01 H-3 400 Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-Nb-95 15 Total Sr 1 1 500 2,000 1-131

1 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15 1 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 Operating Report.

Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be per the recommendations of Regulatory Guide 4.13, Revision 1, July 1977.

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.

If no drinking water pathway exists, a value of 15 pCi/L may be used.

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Table 2 Notes:

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

LLD = 4 .6 6 sb E x V x 2.22 x Y x exp(-Xt)

Where:

LLD = the "a priori" the 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)

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 picocurie Y = the fractional radiochemical yield (when applicable)

X = the radioactive decay constant for the particular radionuclide t = 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 Environmental Radiological Operating 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 prior (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.

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10. AREOR REMP Program Summary Matrix Table 3: Environmental Radiological Monitoring Program Summary Matrix Name of Facility Docket No.

Location of Facility Reporting Period (County, State)

Location with Highest Annual Mean Medium or Type and Total All Indicator Control Pathway Number of Lower Limit of Name, Locations Locations Number of Sampled (Unit of Analyses Detection (a) Distance and Mean (b) Mean (b) Mean (b) Reportable Measurement) Performed (LLD) Direction Range (b) Range (b) Range(b) Occurrences (a) Unless indicated the LLDs specified in Table 2 were met.

(b) Mean and the range based upon detectable measurements only. 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.

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11. Reporting (Notification) Levels for REMP Table 4: Reporting (Notification) Levels for Radioactivity Concentration in Environmental Samples Matrix Airborne Water Particulate or Fish Food Products Analysis (pCi/L) Gases (pCi/m 3) (pCi/kg, wet) Milk (pCi/L) (pCi/kg, wet)

H-3

20,000 Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Sr-89 20 Sr/Y-90 8 Zr-Nb-95 400 1-131 ** 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-1 40 200 300

For drinking water samples. This is the 40 CFR 141 value. If no drinking water pathway exists, a value of 30,000 pCi/L may be used.

- If no drinking water pathway exists, a value of 20 pCi/L may be used.

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12. Distances and Directions to REMP Monitoring Stations Table 5: Distances and Directions to REMP Monitoring Stations Matrix Radial Direction Radial Distance from (True Heading) Plant Station Code* Station Name (Degrees) Km (Miles) 0S1 Exclusion Fence-Northwest Corner 320 0.16 (0.1) 0S2 North Gate 320 0.8 (0.5) 151 Wastewater Pond 330 0.64 (0.4) 2S1 Back Road-300 m North of Plant 0 0.32 (0.2) 3S1 Road NW of 230 kV Switchyard 23 0.64 (0.4) 4S1 Back Road Between Switchyards 43 0.8 (0.5) 5S1 500 kV Switchyard 58 0.64 (0.4) 5S2 Diablo Creek Weir 65 0.96 (0.6) 5S3 Microwave Tower Road 70 1.02 (0.7) 6S1 Microwave Tower 94 0.8 (0.5) 7S1 Overlook Road 112 0.48 (0.3) 8S1 Target Range 125 0.8 (0.5) 8S2 Southwest Site Boundary 128 1.76 (1.1) 8S3 DCSF96-1 (monitor well) 145 0.52 (0.33) 9S1 South Cove 167 0.64 (0.4)

MT1 Meteorological Tower 185 0.32 (0.2)

DCM Diablo Cove Marine 270 0.32 (0.2)

WN1 Northwest Guard Shack 290 0.32 (0.2)

WN2 Diablo Creek Outlet 283 0.25 (0.15) 1A1 Crowbar Canyon 327 2.56 (1.6) 1A2 Blanchard Spring 331 2.4 (1.5)

OB1 Point Buchon 325 5.76 (3.6) 1C1 Montana de Oro Campground 336 7.52 (4.7) 3C1 Ranch Vegetation 20 7.16 (4.5) 4C1 Clark Valley Gravel Pit 45 9.28 (5.8) 5C1 Junction Prefumo/See Canyon Roads 64 7.52 (4.7) 6C1 Household Garden 98 7.24 (4.5) 7C1 Pecho Creek Ruins (Mello Farm) 120 6.56 (4.1) 7C2 Rattlesnake Canyon 124 7.52 (4.7)

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Radial Direction Radial Distance from (True Heading) Plant Station Code* Station Name (Degrees) Km (Miles) 2D1 Sunnyside School 10 11.04 (6.9) 3D1 Clark Valley 24 9.92 (6.2) 4D1 Los Osos Valley Road 36 12.16 (7.6) 6D1 Junction See/Davis Canyon Roads 89 13.4 (8.3) 7D1 Avila Gate 118 10.56 (6.6) 7D2 Avila Beach 110 12.16 (7.6) 7D3 Avila Pier 120 11.0 (6.9) 2F1 Morro Bay (Commercial Landing) 0 17.44 (10.9) 5F1 SLO OEL 79 16.41 (10.2) 5F2 Cal Poly Farm 60 20.16 (12.6) 5F3 SLO County Health Department 70 20.32 (12.7) 7F1 Shell Beach 110 17.28 (10.8) 7G1 Arroyo Grande (Kawaoka Farm) 115 26.88 (16.8) 7G2 Oceano Substation 118 27.68 (17.3)

AVA Avila Beach (near pier) 109 11.75 (7.3)

CBA Cambria Moonstone Beach 330 45.86 (28.5)

CYA Cayucos Beach (near pier) 350 26.87 (16.7)

DY1 Drywell 115' 77 0.041 (0.026)

DW1 Drinking Water from Plant Potable Water System 161 0.59 (0.37)

IS1 - IS8 ISFSI 65 0.48 (0.3)

MDO Montana de Oro (Spooners Cove) 336 7.56 (4.7)

OWl Observation Well 01 336 0.07 (0.046)

OW2 Observation Well 02 157 0.07 (0.045)

OEL Offsite Emergency Lab 79 16.41 (10.2)

OUT Plant Outfall 270 0.32 (0.2)

PMO Pismo Beach (near pier) 113 20.76 (12.9)

PON Pacific Ocean North of Diablo Cove 305 2.4 (1.5)

POS Pacific Ocean South of Diablo Cove 180 0.64 (0.4)

WW2 Water Well 02 70 1.02 (0.63)

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Table 5 Notes:

" Station Code (XYZ):

X - First number (0-9) represents the radial sector in which the station is located:

0 - Northwest 5 - East-northeast 1 - North-northwest 6 - East 2 - North 7 - East-southeast 3 - North-northeast 8 - Southeast 4 - Northeast 9 - South-southeast Y - Letter (S, A-H) represents the distance from the plant:

S - On-site A 2 miles from plant (but off-site)

B 4 miles from plant C 6 miles from plant D 8 miles from plant E 10 miles from plant F 15 miles from plant G 20 miles from plant H - Greater than 20 miles from plant Z - Second number represents the station number within the zone.

" Station Codes exceptions:

The following stations do not follow the coding system:

S Diablo Cove Marine (DCM) 0 Montana de Oro - Spooners Cove (MDO)

S Meteorological Tower (MT1) 0 Pismo Beach (PMO)

S Northwest guard shack (WN1) 0 Cayucos Beach (CYA) 0 Diablo Creek outlet (WN2) 0 Cambria - Moonstone Beach (CBA)

S Pacific Ocean North (PON) S Blanchard Cow Meat (BCM)

S Pacific Ocean South (POS) 0 Blanchard Goat Meat (BGM)

S Offsite Emergency Lab (OEL) 0 Blanchard Sheep Meat (BSM)

S Plant outfall (OUT) 0 Johe Deer Meat (JDM)

S Drinking water (DW1) 0 Johe Pig Meat (JPM) 0 Water Well 02 (WW2) 0 Andre Cow Meat (ACM)

S Observation Well 01 (OW1) 0 Andre Deer Meat (ADM) 0 Observation Well 02 (OW2) 0 Andre Pig Meat (APM)

S Drywell 115 (DY1) S ISFSI TLDs (IS1 - IS8)

S Avila Beach (AVA)

RP1!Il1lu3rl0.DOC 1030.1655

Environmental Radiological Monitoring Procedure RP1.ID11 R10 Page 24 of 27

13. Summary of Cross-Check Program with State of California Table 6: Summary of Cross-Check Program with State of California Matrix Sample Type Station No. Frequency* Radioanalyses Milk (supplemental) 5F2 Monthly Gamma Isotopic (incl. 1-131 and K-40)

Drinking Water DW1 Monthly Gamma Isotopic (incl. 1-131), H-3 Outfall Water OUT Monthly Gamma Isotopic, H-3 Diablo Creek 5S2 Monthly Gamma Isotopic, H-3 Vegetative Crops 7G1 Quarterly Gamma Isotopic (supplemental)

Fish or Invertebrate DCM Quarterly Gamma Isotopic Air Particulates and 5F1"*, 7D1 Weekly & Gross Beta, 1-131, Radioiodine Quarterly Qtrly Gamma Isotopic (composite airborne particulate samples)

Sediment DCM Annually Gamma Isotopic Direct Radiation (TLD) MT1, 1A1, 1C1, 4D1, 5F3, Quarterly Gamma Exposure (mR) 5S1, 7D1, 7C1, 7F1, 8S2 Kelp (supplemental) DCM Quarterly Gamma Isotopic

When available

- The California State CDPH air sampler used for correlation with DCPP 5F1 is located at the SLO County Health Dept on Johnson Avenue (near DCPP TLD station 5F3)

RP1NID11u3rl0.DOC 1030.1655

RP1.ID11 RIO Page 25 of 27 DCPP Onsite ERMP Stations Attachment 1: Page 1 of 1 0

C')

w C,,

0L C) 0)

z 0

a.

Figure 1 RP1!IDllu3rlO.DOC 1030.1655

RP1.ID11 R10 Page 26 of 27 DCPP Offsite ERMP Stations Attachment 2: Page 1 of 1 0 5 miles I I I Units 1 and 2 Diablo Canyon off-site stations.

Figure 2 RP1!ID11u3r10.DOC 1030.1655

RP1.ID11 R10 Page 27 of 27 DCPP Onsite ERMP Stations Satellite Map Attachment 3: Page 1 of I Figure 3 RP1!lDllu3rlO.DOC 1030.1655

Attachment 4 PG&E Letter DCL-1 1-049 Attachment 4 Diablo Canyon Power Plant Chemical Analysis Procedure, CAP A-8, "Offsite Dose Calculations," Revision 34

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 NUCLEAR POWER GENERATION REVISION 34 DIABLO CANYON POWER PLANT PAGE 1 OF 63 CHEMICAL ANALYSIS PROCEDURE UNITS TITLE: Off-Site Dose Calculations 1AND2 INFO ONLY EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED TABLE OF CONTENTS SECTION PAGE SC O PE ............................................................................................................................................................... 2 DISCUSSION .................................................................................................................................................... 3 RESPONSIBILITIES ........................................................................................................................................ 3 PREREQUISITES ............................................................................................................................................. 3 PRECAUTIONS ................................................................................................................................................ 4 IN STRUCTIONS .............................................................................................................................................. 4 Liquid Effluents ............................................................................................................................................. 4 Liquid Effluents - Dose Calculation ...................................................................................................... 4 10 CFR 20, Appendix B, Table 2, Column 2, Effluent (liquid) Concentration Limit (ECL) Calculation 6 Liquid Effluent Radiation M onitor Set Point M ethodology ................................................................. 7 Dose Projection (for Liquid Effluents) ............................................................................................... 13 Liquid Limiting Flow Rate M ethodology - ECL Based ..................................................................... 15 Liquid Limiting Flow Rates - LLD Based .......................................................................................... 16 Gaseous Effluents ........................................................................................................................................ 18 Plant Vent Noble Gas Monitor - RE-14 HASP .............................................................................. 21 Containm ent Purge - RE-44 HASP ................................................................................................. 28 Dose To Critical Receptor Due To Radioiodines, Tritium and Particulates Released in Gaseous Effl u en ts ................................................................................................................................................... 34 40 CFR 190 Dose Calculations .................................................................................................................... 42 ACCEPTANCE CRITERIA ............................................................................................................................ 48 REFERENCES ................................................................................................................................................ 48 RECORDS ....................................................................................................................................................... 49 APPENDICES ................................................................................................................................................. 49 ATTACHM ENTS ............................................................................................................................................ 49 Table 6.1- Typical Liquid Effluent Discharge Pathway Allocation Factors 8 Table 6.2- Typical Gaseous Effluent Discharge Pathway Allocation Factors 20 Table 6.3- Expected On-Site Distance and Visitation Times for Members of the Public 46 v20_CAPA-8u3r34.DOC 08 0901.1100

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1. SCOPE This procedure describes the methodology for the following:

Effluent RECP or RECP or Tech Spec Surveillance Requirement Implements Type Tech Spec Liquids 6.1.1.1 (RECP) Determination of alarm/trip setpoints for RE- 18, 10 CFR 20 App. B 6.1.3.1 (RECP) RE-23, and RE-3 Table 2, Col. 2 Gases 6.1.2.1 (RECP) Determination of alarm/trip setpoints for RE-22, NUREG 0133 6.1.6.1 (RECP) RE-14, and RE-14R Liquids 6.1.3.2 (RECP) Prerelease analyses of effluents 10 CFR 20 App. B 6.1.3.3 (RECP) Table 2, Col. 2 Post release analysis of effluents Liquids 6.1.4.2 (RECP) Dose calculations 10 CFR 50 App. I Liquids 6.1.5.2 (RECP) Dose projections 10 CFR 50 App. I Gases 6.1.6.2 Dose Rate calculations, Noble Gases, Total Body and NUREG 0133 Skin Gases 6.1.6.3 Dose Rate calculations, lodines, Particulates and NUREG 0133 Radionuclides other than Noble Gases, per organ, per age group Gases 6.1.7.2 (RECP) Noble Gas Air Dose Calculations 10 CFR 50 App. I Gases 6.1.8.2 (RECP) lodines, Particulates, and Radionuclides other than 10 CFR 50 App. I Noble Gases Organ Dose Calculations per age group Gases 6.1.9.2 (RECP) Noble Gases, Iodines, Particulates, and Radionuclides 10 CFR 50 App. I other than Noble Gases, Dose Projection Liquids 6.1.10.2 (RECP) Cumulative Dose from: Liquids, Noble Gases, 40 CFR 190 and 4.4.2.b. 1 Iodines, Particulates, and Radionuclides other than Gases (RECP) Noble Gases per age group, per organ Direct 6.1.10.3 (RECP) Direct Radiation Dose Rate and Dose Calculations to 40 CFR 190 Radiation unrestricted areas due to plant and high radwaste storage sky-shine The calculational methodology for doses are based on models and data that make it unlikely to substantially underestimate the actual exposure of an individual through any of the appropriate pathways. Appendixes containing the values for the various parameters used in these expressions are also included.

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2. DISCUSSION 2.1 This procedure is used in support of the Radiological Monitoring and Controls Program (RMCP), and Radioactive Effluent Controls Program (RECP), and the portion that deals with routine radioactive liquid and gaseous releases to the unrestricted area. Limits are based on the dose commitment to a member of the general public related to the release of radionuclides through either direct or indirect exposure (e.g., submersion in a cloud of radioactive Noble Gases, radionuclides deposited on the ground, direct radiation from radionuclides stored on-site, inhalation of radionuclides or ingestion of radionuclides via a food pathway such as milk, meat, vegetable or fish, etc.).

2.2 The conduct of the Environmental Radiological Monitoring Procedure (ERMP) is found in RPI.ID11.

2.3 Changes to CAP A-8 shall be processed in accordance with the requirements of DCPP Technical Specification Section 5.5.1.

3. RESPONSIBILITIES 3.1 The manager, chemistry is the overseeing authority of responsibility for ensuring that the off-site dose calculational procedure (ODCP) meets all RECP and Tech Spec requirements with regards to calculated doses delivered by the plant to the unrestricted area surrounding the site.

3.2 The senior radiochemistry engineer assumes the overall responsibility for ensuring that this procedure's program is followed and implemented where appropriate, especially in regards to RECP or Tech Spec requirements.

3.3 The radiochemistry effluents engineer has the responsibility of correct and timely implementation of all the procedure's calculational methodology, where appropriate, for each radioactive effluent released. Furthermore this engineer is responsible for:

reviewing the results; cross (spot) checking the calculations; and maintaining an updated archive of post release calculated doses for annual report purposes.

3.4 The digital systems group assures that any supporting computer software is maintained current and compatible with the procedure's calculational methodology and that the computer hardware is maintained operable at all times.

3.5 The radiochemistry staff engineer provides an oversight of the effluents program's ODCP to: confirm compliance with RECP or Tech Specs; provide technical support; recommend or design improvements to the dose calculational methodology and the effluent program control; and investigate long-term planning toward effluent related activities and their associated dose calculations.

3.6 Responsibilities as described in CY1, " Chemistry and Radiochemistry," and CYl .DC1, "Analytical Data Processing Responsibilities," apply.

4. PREREQUISITES None v20_CAP_A-8u3r34.DOC 08 0901.1100

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5. PRECAUTIONS None

6. INSTRUCTIONS 6.1 Liquid Effluents 6.1.1 Liquid Effluents - Dose Calculation The dose contributions to the total body and each individual organ (bone, liver, thyroid, kidney, lung and GI-LLI) of the maximum exposed individual (adult) due to consumption of saltwater fish and saltw~ater invertebrate is calculated for all radionuclides identified in liquid effluents released to unrestricted areas using the following expression:

D. = Fe At A ioC ie-Ait (1)

Where:

Do The dose commitment to organ, o, in mrem.

Ft = Near field average dilution factor during the period of the release. It is defined as:

Flow F, =Waste (2)

Dilution Flow x Z Where:

Z Z is the site specific factor for the mixing effect of the discharge structure. Specifically, it is the credit taken for dilution which occurs between the discharge structure and the body of water which contaminates fish or invertebrates in the liquid ingestion pathway. For DCPP Z = 5.

At = The time period for the release in hours.

Aio = The site specific ingestion dose commitment factor to organ, o, due to radionuclide, i, in mrem/hr per pCi/ml as defined by Equation 3.

Ci = Concentration of radionuclide, i, in the undiluted liquid effluent, in pCi/ml.

i= Decay constant of radionuclide, i.

tm = Time interval between end of sampling and midpoint of release.

v20 CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 5 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 The site specific ingestion dose commitment factor, Aio, is defined as:

Aio = ko(UFBF, + UIBIi)DFi (3)

Where:

k. Units conversion factor of 1.14x 105 in units of pCi/pCi x ml/l x yr/hr.

UF = Saltwater fish consumption rate in kilograms of fish per year.

DCPP value for this parameter is 21 kg/yr and is taken from NUREG 0133, Section 4.3.1.

BFj = Saltwater bioaccumulation factor for nuclide, i, in fish flesh in units of pCi/Kg per pCi/l. Values for BFi are taken from Table A-1 of Reg. Guide 1.109, except uranium and plutonium, which were taken from NUREG/CR-4013.

U, = Saltwater invertebrate consumption rate in kilograms per years. DCPP value for this parameter is 5 kg/yr and is taken from NUREG 0133, Section 4.3.1.

BEi Saltwater bioaccumulation factor for nuclide, i, in invertebrate flesh in units of pCi/Kg per pCi/l. Values for BEi are taken from Table A-1 of Reg. Guide 1.109, except uranium and plutonium, which were taken from NUREG-4013.

DFi Adult ingestion dose conversion factor for nuclide, i, in mrem per pCi ingested, from Table E-1 1 of Regulatory Guide 1.109, with exceptions detailed below.

DFj exceptions: H-3, Br-82, Sb-124, Sb-125, Pu-238, Pu-239, Pu-240, Pu-241 and Pu-242, ingestion dose conversion factors are taken from EMP-155.

As-76, Sn-1 13, Sn-1 17m and Sb-122 ingestion dose conversion factors were calculated by ORNL using ICRP-2 methodology.

U-233, U-234, U-235, U-236 and U-238 ingestion dose conversion factors are taken from NUREG-0 172.

The site specific values for Ai, are listed in Appendix 10.1. When necessary, these factors were corrected for the ingrowth of daughter radionuclides following ingestion of the parent. All radionuclides treated in this manner are followed by a "+D." Reference NUREG-0172, "Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake," and A0619601.

Units 1 and 2 share a common liquid radwaste (LRW) treatment system. The effluent doses due to releases discharged via the common LRW are apportioned between the units with 50% credited to Unit 1 and 50% credited to Unit 2.

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a. The ECL for the identified mixture of radionuclides in the "jth," batch of liquids is calculated as follows:

i=1 ECLj - n i= (4) j=1 ECLi, Where:

ECLj = The unrestricted area total undiluted ECL for the "j1h,"

particular mixture of identified radionuclides, in pLCi/ml.

Cij = The concentration of radionuclide "i," in ýiCi/ml for the

,Th, mixture.

ECLij = The ECL in unrestricted area water for radionuclide "i,"

in general, in tCi/ml (from 10 CFR 20, Appendix B, Table 2, Column 2).

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b. The overall ECL for simultaneous discharges is given by Equation 5.

ECor tj~l (5) j=1 ECLj Where:

ECLoverall = The unrestricted area ECL for the current radionuclide mixture for concurrent "j"discharges (in ýtCi/ml).

Cj = The total activity concentration for the "jth" individual stream in ptCi/ml.

ECLj = The total ECL for the "jh,"individual mixture (or stream) determined as defined in Equation 4, in ptCi/ml.

(ID = The ratio of an individual discharge ,jth, pathway flowrate to the sum total of all individual undiluted pathway flowrates as defined by:

(Di = fj (6) 1zfj Where:

f = Undiluted effluent flowrate for pathway, "j".

6.1.3 Liquid Effluent Radiation Monitor Set Point Methodology

a. Introduction The DCPP radiological effluent controls program requires that the liquid effluent monitors be operable with their alarm/trip set points set to ensure that the effluent concentration limits of 10 CFR 20 are not exceeded.

The alarm/trip set point for the liquid effluent radiation monitors is derived from the concentration limit set forth in Appendix B, Table 2, Column 2 of 10 CFR 20.1001-2404.

The alarm/trip set points are applied at the unrestricted area boundary.

The set points take into account appropriate factors for dilution, dispersion, or decay of radioactive materials that may occur between the point of discharge and the unrestricted area boundary.

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b. Allocation and Safety Factors The limits of RECP 6.1.3.1 are site limits which require that the set point methodology must ensure simultaneous releases do not exceed the liquid effluent concentration limits of 10 CFR 20 in the unrestricted area. The DCPP High Alarm Set Point (HASP) methodology makes use of an Allocation Factor (AF) to limit the effluent concentrations from simultaneous liquid discharges. The Allocation Factors can be adjusted based upon operational requirements with the restriction that the sum of the Allocation Factors must be less than or equal to 1.

Typical Allocation Factors are shown.

Table 6.1 Typical Liquid Effluent Discharge Pathway Allocation Factors Discharge Pathway Rad Monitor Allocation Factor (AF)

Oily Water Separator RE-3 0.01 Liquid Radwaste System RE-18 0.90 Steam Generator Blow Down (Unit 1) RE-23 (Ul) 0.04 Steam Generator Blow Down (Unit 2) RE-23 (U2) 0.04 Miscellaneous none 0.01 An additional level of conservatisfn in the HASP methodology is implemented by the use of a Safety Factor (SF). The Safety Factor is defined as 0.9 and provides for a High Alarm Set Point at 90% of the 10 CFR 20 concentration limits.

c. Tritium Correction Factor As result of an aggressive liquid radwaste treatment program, the liquid effluents at DCPP typically contain very low levels of gamma emitters.

In order to reduce the over all volume of liquid waste discharged, DCPP also recycles waste water. This recycling results in higher tritium concentration in liquid effluents when compared with the low gamma emitter concentrations. As a result, standard HASP methodology results in very low set points. In some cases the calculated set points are barely above the monitor background.

The liquid HASP methodology used by DCPP uses a Tritium Correction Factor (TCF) which assumes a constant, but conservative tritium concentration in the liquid effluent. This results in an operationally reasonable set point while ensuring that the liquid effluent concentrations released to the unrestricted areas do not exceed the limits of 10 CFR 20.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 9 OF 63 TITLE: Off-Site Dose Calculations UNITS I AND 2 The Tritium Correction Factor is defined as shown in Equation 7.

TCF 2 I - CH3/ECLH3, F/U(7)

Where:

ECLH3 effluent concentration limit of tritium.

CH3 concentration of tritium in the release mix, pre-dilution (jiCi/ml).

F = conservative dilution flow rate (gpm).

f = conservative undiluted effluent flow rate (gpm).

The concentration of tritium, CH3, is conservatively estimated.

d. Liquid Effluent Radiation Monitor Set Point Calculations The High Alarm Set Point (HASP) are calculated to ensure that the liquid effluent concentration limits of 10 CFR 20 are not exceeded. The set points represent the maximum operational set point. The actual set point used by operations will be equal to or less than the actual value as determined by the HASP methodology described in this section.

1. Set Point Methodology for RE-3 HASP: Oily Water Separator Under normal conditions, the Oily Water Separator stream does not contain any radioactive material. Only in the event that there is primary to secondary leakage does this become a potential liquid effluent discharge point. In order to insure that no unplanned or unmonitored releases take place by way of the Oily Water Separator, RE-3 serves to monitor the discharge even when no activity has been identified in the effluent. When no significant primary to secondary leakage is taking place or when no activity has been identified in the Oily Water Separator, the High Alarm Set Point for RE-3 is calculated as shown in Equation 8.

HASPs_3 = 3 x BKGDRE_3 (8) v20_CAPA-8u3r34.DOC 08 0901.1100

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 10 OF 63 TITLE: Off-Site Dose Calculations UNITS I AND 2 In the event that primary to secondary leakage results in activity being detected in the Oily Water Separator, Equation 9 will be used to calculate a High Alarm Set Point value. The greater HASP value as determined by Equation 8 or Equation 9 will be used.

HASPs-3 =BKGDRE-3+(AFXSF)x -kC ZCiL x TCF (9) i#H3 Where:

HASPRE-3 = high alarm setpoint for RE-3 (cpm).

BKGDRE_3 = background reading for RE-3 (cpm).

(AF) = allocation factor for the oily water separator effluent system from Table 6.1.

(SF) = safety factor for RE-3 (0.9).

ky = monitor response factor (cpm/VCi/ml).

C= concentration of gamma emitting isotopes in the release mix, pre-dilution ([tCi/ml).

F = dilution flow rate (gpm).

f = undiluted effluent flow rate (gpm).

Ci = concentration of isotope "i," in the release mix, pre-dilution (ýtCi/ml).

ECLi = effluent concentration limit of isotope "i".

TCF = tritium correction factor as defined by Equation 7.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 11 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

2. Set Point Methodology for RE-18 HASP: Liquid Radwaste System.

The High Alarm Set Point for the RE- 18 Liquid Radwaste System liquid effluent radiation monitor is calculated as shown in Equation 10.

HASPR- 8 BKGDE 18 +(AF XSF) x E .y y

F/f Ci ECL1 1 xTCF (10)

Where:

HASPRE*1 8 high alarm setpoint for RE-18 (cpm).

BKGDRE-18 background reading for RE- 18 (cpm).

(AF) allocation factor for the liquid radwaste effluent system from Table 6.1.

(SF) = safety factor for RE-18 (0.9).

ky = monitor response factor (cpm/ntCi/ml).

C= concentration of gamma emitting isotopes in the release mix, pre-dilution (gCi/ml).

F = dilution flow rate (gpm).

f = undiluted effluent flow rate (gpm).

C, = concentration of isotope "i," in the release mix, pre-dilution (liCi/ml).

ECLi = effluent concentration limit of isotope 'T'.

TCF = tritium correction factor as defined by Equation 7.

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3. Set Point Methodology for RE-23 HASP: Steam Generator Blowdown Tank.

The High Alarm Set Point for the RE-23, Steam Generator Blowdown Tank liquid effluent radiation monitor, is calculated as shown in Equation 11.

HASPs_23 : BKGDR_23 + (AFXSF) x*krC, F/f x TCF E Ci/ECL(

LiH3 /

Where:

HASPRE-23 high alarm setpoint for RE-23 (cpm).

BKGDRE 23 background reading for RE-23 (cpm).

(AF) allocation factor for the steam generator blowdown effluent system for each unit from Table 6.1.

(SF) = safety factor for RE-23 (0.9).

ky = monitor response factor (cpm/ýCi/ml).

C= concentration of gamma emitting isotopes in the release mix, pre-dilution ([tCi/ml).

F = dilution flow rate (gpm).

f = undiluted effluent flow rate (gpm).

Ci = concentration of isotope "i," in the release mix, pre-dilution (jiCi/ml).

ECLi = effluent concentration limit of isotope "i".

TCF = tritium correction factor as defined by Equation 7.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 13 OF 63 TITLE: Off-Site Dose Calculations UNITS I AND 2 6.1.4 Dose Projection (for Liquid Effluents)

The projected dose contributions from each reactor unit due to liquid effluents for the current calendar month, quarter and current calendar year must be determined in accordance with the methodology and parameters in the ODCP at least once per 31 days.

The purpose of this is to determine if appropriate treatment of liquid radioactive materials in relation to maintaining releases "as low as reasonably achievable," is necessary.

The projected dose from each reactor unit is given by:

Dp = DP,U + IDP Con (12)

Where:

= Projected Dose.

- Projected dose attributed to reactor unit, U.

Dp,com = Projected dose common to both reactor units.

The 31-day projected dose is calculated by Equation 13.

Di =3 1xDA +d7 +dcB (T+t) (13)

Where:

D = Monthly Projected Dose.

OM = Previous Month's Actual Dose.

dCM = Current Month Actual Dose to date.

dcB = Projected Dose from Current Batch Release.

T = Number of days in the previous month.

t = Number of days into the present month.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 14 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Projected quarterly doses are determined by Equation 14.

DcQ=dcQ +(92_t)DP +dA +/-dc(B P A (T + t) (14)

Where:

DcQ = Projected dose for the current calendar quarter.

dc = Current quarter to date actual dose.

D= Previous quarter's actual dose.

dCB = Projected dose as a result of the current batch release.

T = Number of days in the previous quarter.

t = Number of days into the present quarter.

Projected yearly doses are determined by Equation 15.

Dcy~dcy +(366-t)- A (T+t)+d

+dA (15)

Where:

Dc = Projected dose for the current calendar year.

dc = Current year to date actual dose.

D= Previous year's actual dose.

dcB = Projected dose as a result of the current batch release.

T = Number of days in the previous year.

t = Number of days into the present year.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 15 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.1.5 Liquid Limiting Flow Rate Methodology - ECL Based The maximum effluent flow rate through monitors RE-3, RE-1 8, and RE-23 as well as for releases from the Condensate Demineralizer Regenerate waste tank or miscellaneous release points is established in order to provide further control over the effluent releases. The release rate limit is determined by the effluent concentration and the 10 CFR 20 Effluent Concentration Limits (ECLs) as shown in Equation 16.

f F(AFXSFXTCF)

Ci (16) i#H-3 ECLi Where:

f Maximum operational undiluted liquid radwaste effluent discharge flow rate (gpm).

F = Expected dilution flow rate (gpm).

AF = allocation factor for the liquid radwaste effluent source from Table 6.1.

SF = safety factor (0.9).

TCF = tritium correction factor as defined by Equation 7.

Ci concentration of isotopes "i" in the release mix, pre-dilution (jiCi/ml).

ECLi = effluent concentration limit of isotope "i" (ptCi/ml).

C.

When the term C- = 0 then the Limiting Flow Rate is calculated by:

i*H-3 ECLi f = F(AFXSFXTCF) (17)

Where the terms are as previously defined.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 16 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.1.6 Liquid Limiting Flow Rates - LLD Based When there is no primary to secondary leakage, the Oily Water Separator and various miscellaneous release points are assumed to be uncontaminated.

Furthermore, in order to establish practical operational flow rate limits for any sources when they are considered uncontaminated, Equation 18 is used. While no activity may be present, Equation 18 assumes a concentration equal to the Lower Limit of Detection for the nuclides listed in CY2. ID1, Appendix 6.1, Table 6.1.3-1.

f=F(AFXSF)

(18)

4.3 Where

f Maximum operational undiluted liquid radwaste effluent discharge flow rate (gpm).

F = Expected dilution flow rate (gpm).

AF allocation factor for the liquid radwaste effluent source from Table 6.1.

SF = safety factor (0.9).

4.3 = Total ECL fraction as given by:

LLDi ECLi Where:

LLDi = Lower limit of detection for isotope "i" from CY2.IDI, Appendix 6.1, Table 6.1.3-1. (ptCi/ml).

ECLi = effluent concentration limit of isotope "i" (jiCi/ml).

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- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 17 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.1.7 Unplanned Liquid Releases (Abnormal Releases)

An unplanned release is an unexpected and potentially unmonitored release to the environment due to operational error or equipment malfunctions.

a. Unmonitored unplanned releases shall have a report written by the Radiochemistry Effluents Engineer describing the event with a calculation, if possible, of the percent of Tech Spec release rate limit.

This will then be forwarded to PSRC for review. Describe these unplanned releases in the Annual Radioactive Effluent Release Report.

b. Monitored unplanned releases which exceed 1% of the RECP release rate limit will also have a report written describing the event and must be forwarded to the PSRC for review. Describe these unplanned releases in the Annual Radioactive Effluent Release Report.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 18 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2 Gaseous Effluents The only significant path for gaseous radioactive releases to the environment during normal operations is via the plant vent. This source is used for calculating dose rates and real-time doses to the unrestricted area due to noble gases, vaporous radioiodines and airborne radio-particulates. The plant vent also has redundant monitoring for these types of gaseous releases.

Other paths such as the steam generator blowdown tank vent, the chemistry lab fume hood, the main condenser Nash vacuum pump discharge, hot machine shop vent, etc., are considered miscellaneous release sources. These miscellaneous release sources are not continuously monitored but can have dose rates and dose calculated for their path to the unrestricted area.

6.2.1 Meteorological Methodology The equations for determining gaseous effluent concentration limits, high alarm setpoints, dose rates, and critical receptor doses make use of the historical average atmospheric conditions in accordance with methodologies of Regulatory Guides 1.109 and 1.111 and NUREGs 0133 and 0472. The historical average dispersion (x/Q) and deposition (D/Q) values are derived from the methodology of Regulatory Guide 1.111 as implemented by NUREG 2919 (computer code XOQDOQ). The DCPP dispersion and deposition values are based on the latest five years of meteorological data and are updated when the value of x/Q or D/Q changes by more than ten percent.

The present values are listed in Appendix 10.2.

Long-term releases are characterized as those that are generally continuous and stable in release rate, such as normal ventilation systems effluents. Doses due to long-term releases are modeled using historical annual average dispersion and deposition values in accordance with the guidance of Regulatory Guide 1.109, Regulatory Guide 1.111, NUREG 0133 and NUJREG 0472.

Short-term releases are defined as those which occur for a total of 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months or less in a calendar year but not more than 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months in any quarter. In accordance with NUREG 0133 and based upon an operational history that has demonstrated short term gaseous releases can be characterized as random in both time of day and duration, historical average atmospheric dispersion and deposition values are used to model doses due to short-term releases.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 19 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2.2 Gas Effluent Concentration Limits

a. Philosophy of Concentration Limits The radiological effluent controls restrict at all times the dose rate due to radioactive materials released in gaseous effluents from the site to areas at or beyond the site boundary for noble gases to less than or equal to 500 mrem/yr to the total body and 3000 mrem/yr to the skin. For iodine-131, iodine-1 33, tritium and for all radionuclides in particulate form with half-lives greater than 8 days, the dose rate is limited to less than or equal to 1500 mrem/yr to any organ.

These dose rate limits act to restrict at all times the instantaneous concentrations of radionuclides in gaseous effluents at the site boundary.

1. Allocation and Safety Factors The limits set forth by RECP 6.1.6.1 are site limits which require that the set point methodology must ensure simultaneous releases do not exceed the off-site dose rate limits set forth by RECP 6.1.6.1(a) and 6.1.6.1(b). The DCPP High Alarm Set Point methodology makes use of an Allocation Factor (AF) to limit the noble gas effluent dose rate from simultaneous atmospheric releases.

The Allocation Factors can be adjusted based upon operational requirements with the following restrictions:

The sum of the Allocation Factors for RE- 14 (plant vent noble gas monitor), the SGBD tank vents, and miscellaneous release points from both units must be less than or equal to 1.

The Allocation Factors for RE-22 (Waste Gas Decay Tanks) and RE-44 (Containment Purge) can also be adjusted based upon operational requirements with restriction that the sum of the Allocation Factors for RE-22 and RE-44 must be less than or equal to the Allocation Factor for RE-14.

The Allocation Factors for RE-24 (Plant Vent Iodine Monitor) and RE-28 (Plant Vent Particulate Monitor) are set equal to the Allocation Factor for RE-14.

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PACIEFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 20 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND2 Typical Allocation Factors are shown:

Table 6.2 Typical Gaseous Effluent Discharge Pathway Allocation Factors Discharge Pathway Rad Monitor Allocation Factor (AF)

Plant Vent - NG Monitor RE-14 0.48 Plant Vent Iodine Monitor RE-24 0.48 Plant Vent Part Monitor RE-28 0.48 Waste Gas Decay Tank RE-22 0.10 Containment Purge RE-44 0.38 SGBD Tank Vent 0.01 Miscellaneous 0.01 An additional level of conservatism in the HASP methodology is implemented by the use of a Safety Factor (SF). The Safety Factor is defined as 0.9 and provides for a High Alarm Set Point at 90% of the dose rate limits.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 21 OF 63 TITLE: Off-Site Dose Calculations UN I TS 1 AND 2

b. Gaseous Effluent Radiation Monitor Set Points

1. PLANT VENT NOBLE GAS MONITOR - RE-14 HASP The Plant Vent effluent stream is monitored by rad monitor RE- 14.

RE-14 provides alarm function only.

The High Alarm Set Point methodology for RE-14 is given by Equation 19, which is based upon the assumption that the total body dose rate limit is most limiting.

14M axCT(NG) = AF x SF x 500 (19) 472xFp, x(zQ)M x 294 Where:

4 1 MaxCT(NG)

= the maximum allowable total concentration of noble gas in the plant vent, in pCi/cc.

AF = the allocation factor for the plant vent for one unit from Table 6.2.

SF = a safety factor to ensure that dose rate limits of the radiological effluent controls are not exceeded (0.9).

500 = (mrem/yr) the site Total Body dose rate limit for instantaneous releases.

472 = the conversion constant to cc/sec from cfm.

Fpv = total flow rate in the plant vent, in cfm (maximum plant vent flow rate is 263,000 cfm).

(x/Q)Max = the maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological data base, from Appendix 10.2.

294 = the whole body dose factor (mrem/yr/ýtCi/m3 ) for Xe-133 as presented in Appendix 10.3, (for the plant vent HASP, the release is assumed to be all Xe- 133).

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2. PLANT VENT NOBLE GAS MONITOR - RE-14 SCALING In order to correlate the readings of RE-14 to noble gas concentration during periods between samplings, the concentration is scaled according to Equation 20.

CT =CPM x CS (20)

CPMS Where:

CPMT = RE-14 time weighted arithmetic mean (cpm).

CPMs = RE-14 gross count rate at the time of sampling (cpm).

Cs = Concentration of noble gas corresponding to CPMs, based upon noble gas grab sample ([tCi/cc).

CT = Scaled concentration of noble gas ([tCi/cc).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 23 OF 63 TITLE: Off-Site Dose Calculations UNITS I AND 2

3. PLANT VENT IODINE MONITOR - RE-24 The Plant Vent Iodine concentration is monitored by rad monitor RE-24. RE-24 provides alarm function only. The alarm setpoint methodology is based upon the assumption that RE-24 responds only to 1-131. The methodology also presumes a release mixture based upon the RCS source term.

The High Alarm Set Point methodology of RE-24 is given by Equation 21.

24 MOfxc (Iodine) = SF x AF x 1500 (21) 472xFp, x(-* u*"ZPiWf, i

Where:

24 MxCT(lodine) = the maximum allowable concentration of 1-13 1 in the plant vent.

AF = The allocation factor for the plant vent for one unit from Table 6.2.

SF = A safety factor to insure that the dose rate limits of the radiological effluent controls are not exceeded (0.9).

f131 = fraction of the total non-noble gas concentration that is due to 1-131. Defined as:

f =13cI- 13 1 (22)

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- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 24 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 1500 = (mrem/yr) the site organ dose rate limit for Iodine-131, for Iodine-133, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days.

472 = the conversion constant to cc/sec from cfm.

Fpv = total flowrate in the plant vent, in cfm (maximum plant vent flowrate is 263,000 cfm).

(z1Q)ML = maximum historical site boundary dispersion factor, .based on 5 year averages derived from the meteorological database, from Appendix 10.2.

Pw = Inhalation dose factor for nuclide 'T' (mrem/yr/jtCi/m3) for child age group for worst case organ, from Appendix 10.4.

Dose factors are based upon NUREG 0133 methodology. Inhalation dose conversion factors are taken from Reg. Guide 1.109, Rev 1, Table E-9, with the following exceptions: H-3, Sb-124 and Sb-125 inhalation dose conversion factors taken from NUREG/CR-4013.

f = fraction of total non-noble gas concentration (excluding tritium) that is due to nuclide, i, and defined as:

fi - ci (23) ci v20_CAP_A-8u3r34.DOC 08 0901.1100

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4. PLANT VENT PARTICULATE MONITOR - RE-28 The Plant Vent Particulate concentration is monitored by rad monitor RE-28. The alarm setpoint methodology is based upon the assumption of a 5% cross talk from the iodine channel. This is due to the retention of a small portion of iodine on the particulate filter.

A release mixture based upon the RCS source term is also assumed.

The High Alarm Set Point methodology for RE-28 is given by Equation 24.

28Marc (Particulates)= AF x SF x 0.05 x Ioe fIdines + Particulaes es 1500 (24)

X472 x Fpv x ()I L piwf Where:

28 MaxCT(Particulate) = Maximum allowable particulate concentration in the plant vent.

AF = The allocation factor for the plant vent for one unit from Table 6.2.

SF = A safety factor to insure that the dose rate limits of the radiological effluent controls are not exceeded (0.9).

0.05 = Fraction of total iodine activity retained on particulate filter.

faodines = Fraction of the total non-noble gas concentration that is due to iodines.

fparticulates = Fraction of the total non-noble gas concentration that is due to particulates.

1500 = (mrem/yr) the site organ dose rate limit for Iodine-131, for Iodine-133, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days.

472 = Conversion constant to cc/sec from cfm.

Fpv = Total flowrate in the plant vent, in cfm (maximum plant vent flowrate is 263,000 cfm).

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- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 26 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 (x Q)ML = Maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological database, from Appendix 10.2.

piw = Inhalation dose factor for nuclide "i" (mrem/yr/tCi/m3) for child age group for worst case organ, from Appendix 10.4. Dose factors are based upon NUREG 0133 methodology.

Inhalation dose conversion factors are taken from Reg. Guide 1.109, Rev 1, Table E-9, with the following exceptions: H-3, Sb-124 and Sb-125 inhalation dose conversion factors taken from NUREG/CR-4013.

fi Fraction of total non-noble gas concentration (excluding tritium) that is due to nuclide, i, as defined by Equation 23.

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5. WASTE GAS DECAY TANK MONITOR - RE-22 HASP Effluent releases from the Waste Gas Decay Tank are monitored by rad monitor RE-22. RE-22 provides alarm and automatic release termination functions.

The High Alarm Set Point methodology for RE-22 is given by Equation 25, which is based upon the assumption that the skin dose rate limit is most limiting.

22MaC (NG)= AFxSF 3000 (25) 4.72xFg, x(z---*)L xl.34x 103 Wherre:

22 MaXCT(NG

= the maximum allowable total concentration of noble gas in the gas decay tank exhaust line, in jtCi/cc.

AF = the allocation factor for the waste gas decay tank for one unit from Table 6.2.

SF = safety factor to ensure that dose rate limits of the radiological effluent controls are not exceeded (0.9).

3000 = (mrem/yr) the site Skin dose rate limit for instantaneous releases.

472 = the conversion constant to cc/sec from cfm.

Fgdt = maximum flow rate for the gas decay tank system (31 cfm).

(-I-Q)ML = the maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological data base from Appendix 10.2.

1.34x 103 = the skin dose factor for Kr-85 (mrem/yr/ tCi/m 3, from Appendix 10.3.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 28 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

6. CONTAINMENT PURGE - RE-44 HASP The Containment Purge is monitored by rad monitor RE-44. RE-44 provides alarm and automatic release termination functions.

The HASP for RE-44 must limit the noble gas dose rate for skin and total body exposure. In order to address this, two set points are calculated. One set point is calculated based upon limiting the total body dose rate and the other limits the skin dose rate. The more limiting set point is used. The High Alarm Set Point methodology for RE-44 is given by Equations 26 and 27.

a) Limiting Concentration Based on Total Body Dose 44 A/BM ax c(NG) AF x SF x 500 (26) 4 72 x F, x(x -Q)L x Z K fi 26 Where:

44 ABMaXCT(NG) = the maximum allowable total concentration of noble gas in the containment purge exhaust line, in [tCi/cc.

AF = the allocation factor for the containment purge for one unit from Table 6.2.

SF = safety factor to ensure that dose rate limits of radiological effluent controls are not exceeded (0.9).

500 = (mrem/yr) the site Total Body dose rate limit for instantaneous releases.

472 = the conversion constant to cc/sec from cfm.

Fct maximum flow rate in the containment purge system (maximum containment purge flowrate is 55,000 cfm).

(zQ)ML = the maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological data base averages, from Appendix 10.2.

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- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 29 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Ki= whole body dose factor (mrem/yr/jiCi/im3) for isotope "i." Dose factors are presented in Appendix 10.3.

fi the fraction of the concentration of the individual noble gas radionuclide, "i," in the total mix of noble gas effluents in the containment purge line.

b) Limiting Concentration Based on Skin Dose Skin dose should be calculated whenever an RCS sample contains a significant fraction of Kr-85.

The concentration limit calculated by this method should then be compared to the concentration limit calculated for total body dose. The smaller concentration should be selected as the limiting concentration for the HASP.

44 A/B M axc(NG) =AF x SF x 3000 (27) 472xF x+1.1M)f Where:

44 A/BMaXCT(NG) = the maximum allowable total concentration of noble gas in the plant vent in iiCi/cc.

AF = the allocation factor for the containment purge for one unit from Table 6.2.

SF = safety factor to ensure that dose rate limits of the radiological effluent controls are not exceeded (0.9).

3000 = (mrem/yr) the site skin dose rate limit for instantaneous releases.

472 = the conversion constant to cc/sec from cfm.

F, = total flow rate in the containment purge system, in cfm (maximum containment purge flow rate is 55,000 cfm).

(z/Q)M, = the maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological data base averages, from Appendix 10.2.

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- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 30 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Li the skin dose factor (mrem/yr per pCi/m 3) for isotope "i." Dose factors are presented in Appendix 10.3.

1.1 = Conversion factor mrem/mrad. Converts air dose to skin dose.

Mi = the gamma air dose factor (mrad/yr per ptCi/rn3) for isotope 'i." Dose factors are presented in Appendix 10.3.

f= the fraction of the concentration of the individual noble gas radionuclide, "i," in the total mix of noble gas effluents in the containment purge line.

c. Mode 6 Particulate activity.

The HASP calculation specified in this section based upon Noble Gas effluent limitations conservatively bounds the Tech. Spec. requirement for particulate activity in Mode 6. The FSAR expected case accident for Mode 6 is a containment fuel handling accident which does not include a particulate release. Therefore, the HASP for RM-44 in this section conservatively satisfies the Tech. Spec. (Ref.: 8.15).

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a. Total Body Noble Gas Dose Rate Methodology The dose rate to the total body due to immersion in a cloud of noble gases is given by:

(x-/Q)AL YKj <500

ý mrem/year (28)

Where:

Q, = The release rate of radionuclide i in units of pCi/sec.

All other terms are as previously defined.

b. Skin Dose Rate Methodology The dose rate to the skin due to immersion in a cloud of noble gases is given by:

(--Q)A-fVZ(Li +'1-.1Mi)Q.i <3000 mrem/year (29) i Where the terms are as previously defined.

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c. Radioiodine, Tritium and Particulate Dose Rate Methodology The dose rate to organ, o, due to radioiodines, tritium and particulates released in gaseous effluents is given by:

(x-/Q-)AL, Z Pi. Qi <1500 mntern/year (30)

Where:

Pi. = K'(BR)DFAio and K' = 106 pCi/ýLCi BR The breathing rate of the a child age group in m 3/yr.

The default value of 3700 m3/yr is taken from Table E-5 of Reg. Guide 1.109.

DFAj= The inhalation dose factor for organ o, for the child age group for radionuclide, i, from Table E-9 of Reg. Guide 1.109 in mrem/pCi, with the following exceptions:

H-3, Sb-124 and Sb-125 inhalation dose conversion factors taken from NUREG/CR4013.

All other terms are as previously defined.

Values for Pi. are listed in Appendix 10.6.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 33 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2.4 Noble Gas Air Dose Calculation Methodology

a. Gamma Air Dose The gamma air dose due to immersion in a cloud of noble gases is given by:

Day =3.l7xlo8(z-/Q)ym~i~i (31)

Where:

Da, = Gamma air dose in mrad.

3.17x10. 8 = Conversion constant yr/sec.

Mi = Gamma air dose factor for nuclide i, in mrad/yr per 11Ci/m 3. Values are listed in Appendix 10.3.

I Q = Total release of noble gas radionuclide, i, in gCi.

All other terms are as previously defined.

b. Beta Air Dose The beta air dose due to immersion in a cloud of noble gases is given by:

Dap = 3.17 x 10- (z-/)MX NiQi (32)

Where:

Da, = Beta air dose in mrad.

Ni = Beta air dose factor for nuclide i, in mrad/yr per ýiCi/m 3.

Values are listed in Appendix 10.3.

All other terms are as previously defined.

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- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 34 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2.5 Dose To Critical Receptor Due To Radioiodines, Tritium and Particulates Released in Gaseous Effluents

a. Calculation Methodology The dose to an individual (critical receptor) due to radioiodines, tritium and particulates released in gaseous effluents with half-lives greater than 8 days is determined based upon the methodology described in NUREG 0133. This methodology makes use of the maximum individual concept described in Regulatory Guide 1.109. The maximum individual is characterized as maximum with regard to food consumption, occupancy, and other usage parameters. This concept therefore models those individuals within the local population with habits representing reasonable deviations from the average. In all physiological and metabolic respects, the maximum individual is assumed to have those characteristics that represent the average for the age group of interest.

The concept of critical receptor is introduced as a further refinement of the maximum individual. The critical receptor is defined as that individual that receives the largest dose based upon the combination of dose pathways that have been shown to actually exist. The critical receptor concept is applied at that location where the combination of dispersion (x/Q), deposition (D/Q), existing pathways, occupancy time, receptor age group, and effluent source term indicates the maximum potential exposure. The inhalation and ground plane exposure pathways are considered to exist at all locations. The grass-cow-milk, grass-cow-meat, and vegetation pathways are considered based on their actual existence in the vicinity of the plant.

The dose pathways that have been shown to actually exist at DCPP are the ground plane, inhalation and the vegetation pathways. These dose pathways are reviewed yearly and updated based upon the annual land use census survey in order to insure that actual exposure to an individual will not be substantially underestimated.

The locations of the pathways and descriptions are listed in Appendix 10.2.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 35 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

b. Dose Calculation The dose contributions to the total body and each individual organ (bone, liver, thyroid, kidney, lung and GI-LLI) of the maximum exposed individual (Critical Receptor) due to radioactive gaseous effluent releases is calculated for all radionuclides identified in gaseous effluents released to unrestricted areas using the following expression:

Dapo = 3.17 x 10- 8 WCRZRipoQi (33)

Where:

Dapo = Dose to the critical receptor for age group a, pathway p, and organ o, in mrem.

WCR = Critical receptorz/Q for immersion, inhalation and all tritium pathways (seconds/m 3) from Appendix 10.2.

= Critical receptor D/Q for ground plane and all ingestion pathways (1/M 2) from Appendix 10.2.

Raipo = Site specific dose factor for age group a, radionuclide i, pathway p, and organ j (mrem/yr per jiCi/m3 for inhalation and tritium pathways - mrem/yr per [tCi/(sec in 2) for ground plane and ingestion pathways). These dose factors are listed in Appendix 10.6.

The site specific dose factors are calculated based upon NUREG 0133 methodology. All dose conversion factors are taken from Reg. Guide 1.109, Rev 1, Tables E6-E14, with the following exceptions: H-3, Sb-124 and Sb- 125 dose conversion factors taken from NUREG/CR-4013.

Qi = The total release of radionuclide i, in units of jiCi.

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The Percent Release Rate Limit (PRRL) for noble gas releases for each unit is calculated based upon the 500 mrem/yr whole body dose rate limit, and is given by Equation 34.

PRRL = (0.48X500mrem/ ix yr) 100% (34)

Where:

(z----), = The maximum site boundary dispersion factor based on 5 year averages from Appendix 10.2.

Ki = Whole body dose factor (mrem/yr per tCi/m 3) for isotope "i." Dose factors are presented in Appendix 10.3.

Qi = Total release rate of isotope "i" from all sources discharged through this release point in pCi/sec.

0.48 = Plant vent allocation factor for one unit from Table 6.2.

500 mrem/yr = Site noble gas dose rate limit.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 37 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 The Expected Reading (ER) is the anticipated monitor response based upon the known plant vent concentration and the monitor response factors. The Expected Readings for RE-22, RE-44, RE-14 are given by Equations 35, 36, and 37.

ER(RE- 2 ) = BKG(PE-2 2) + I (E2)iCR-2 (35)

Where:

ER(RE_22) = Expected reading on monitor RE-22 in ýiCi/cc.

BKG(RE-22) = Monitor background.

k(RE.22)i Noble gas monitor response factor for nuclide "i" for monitor RE-22.

C(RE-22)i = Concentration of nuclide "i" seen by RE-22.

ER(RE 44 ) -ý:BKG(PE44 +)CCSP(PE 44 ) Z k(pE-4) iCP-4 (36)

Where:

ERARE-44) = Expected reading on monitor RE-44 in ,iCi/cc.

BKG(RE44) = Monitor background.

CCSPCPE.44) = Conversion constant setpoint for monitor RE-44.

k(RE-44)i = Noble gas monitor response factor for nuclide "i" for monitor RE-44.

C(RE_44)i Concentration of nuclide "i" seen by monitor RE-44.

ER (PE- 4) = BKG(PEl 4 ) +CCSP(REi1) k 14 E )i (E14 i (37)

Where:

ER -14) = Expected reading on monitor RE-14 in pCi/cc.

BKG(RE-14) = Monitor background.

CCSP(REI 4) = Conversion constant setpoint for monitor RE-i14.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 38 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 k(RE-14)i Noble gas monitor response factor for nuclide "i" for monitor RE-14.

C(RE.14)i = Concentration of nuclide "i" seen by monitor RE-14.

Generally if the Expected Reading (ER) is greater than the existing HASP setting (an "administrative limit" as set by CY2.DC 1) then no release should be made until a calculation shows that the HASP (Admin Limit) can be raised so the release can be legally discharged. On the other hand should the ER be less than the existing HASP (Admin Limit), then the release can be discharged.

6.2.7 IPT - PRRL The Percent Release Rate Limit (PRRL) for radioiodines, tritium and particulates for each unit is calculated based upon the 1500 mrem/yr organ dose rate limit. The dose rate is calculated for the inhalation pathway to the child age group using the highest (worst case) organ dose factor for nuclide.

The Percent Release Rate Limit based on the worst case organ is given by Equation 38.

Wre: =(0.48X1 500mrem / yr)'x 100%(8 PRRL0, = xlO (38)

Where:

(zQ)M = The maximum site boundary dispersion factor based on 5 year averages from Appendix 10.2.

piw Inhalation dose factor for nuclide "i" (mrem/yr/[tCi/m 3) for child age group for worst case organ, from Appendix 10.4.

Dose factors are based upon NUREG 0133 methodology.

Inhalation dose conversion factors are taken from Reg. Guide 1.109, Rev 1, Table E-9, with the following exceptions: H-3, Sb-124 and Sb-125 inhalation dose conversion factors taken from NUJREG/CR-4013.

= Release rate of isotope "i" in pCi/sec.

0.48 = Plant vent location factor for one unit from Table 6.2.

1500 mrem/yr Site radioiodine, tritium and particulate dose rate limit.

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- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 39 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2.8 Alternate Dose Methodologies For purposes of routine gaseous effluent dose assessment, the methodology of NUREG 0133 (described in Section 6.2.5) will be used. However, DCPP may elect to utilize the dose methodologies of Regulatory Guide 1.109 or the GASPAR computer code for special purposes such as evaluation of potential new gaseous effluent dose pathways or critical receptors.

6.2.9 Gas Effluent Dose Projection The projected dose contributions from each reactor unit due to gaseous effluents for the current calendar month, quarter and current calendar year must be determined in accordance with the methodology and parameters in the ODCP at least once per 31 days.

The computer program, Radioactive Effluent Management System (REMS), is used for this projection. Therefore, by the first day of the year, the following tables in REMS need to be updated:

GRW dose receptor

GRW dose rate receptor

GRW external dose select

GRW external occupancy

GRW internal dose select

GRW internal occupancy The purpose of this is to determine if appropriate treatment of gaseous radioactive materials in relation to maintaining releases "as low as reasonably achievable," is necessary.

The projected dose from each reactor unit is given by:

Dp = DPU + IDPcom (39)

Where:

DP = Projected Dose.

Dp,U = Projected dose attributed to reactor unit, U.

Dp=coln Projected dose common to both reactor units.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 40 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 The 31 day projected dose is calculated by Equation 40.

DPM +dcm +dcB DAI=31x.

D A. (T.A. t) P(0 4(T)+t)

Where:

Dm Projected Dose for the next 31 day period.

D PM Previous Month's Actual Dose.

'4 d CM = Current Month Actual Dose to date.

dcB Projected Dose from Current Batch Release.

T = Number of days in the previous month.

t = Number of days into the present month.

Projected quarterly doses are determined by Equation 41.

P -A +(92-t)

DCQ=-rcQ +((T+t) A +t)

A Where:

= Projected dose for the current calendar quarter.

dCA = Current quarter to date actual dose.

= Previous quarter's actual dose.

dcs = Projected dose as a result of the current batch release.

T = Number of days in the previous quarter.

t = Number of days into the present quarter.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 41 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Projected yearly doses are determined by Equation 42.

(36tD7r +dcy +dB(

D +(366-t) (T+t) (42)

Where:

DP CY Projected dose for the current calendar year.

dA Current year to date actual dose.

DA Previous year's actual dose.

dPB = Projected dose as a result of the current batch release.

T = Number of days in the previous year.

t = Number of days into the present year.

6.2.10 Unplanned Gaseous Releases (Abnormal Releases)

a. An unplanned release is an unexpected and potentially unmonitored release to the environment due to operational error or equipment malfunctions.

1. Unmonitored unplanned releases shall have a report written by the Radiochemistry Effluents Engineer describing the event with a calculation, if possible, of the percent of RECP limit. This will then be forwarded to PSRC for review. Describe these unplanned releases in the Annual Radioactive Effluent Release Report.

2. Monitored unplanned releases which exceed 1% of the RECP limit will also have a report written describing the event and must be forwarded to the PSRC for review. Describe these unplanned releases in the Annual Radioactive Effluent Release Report. For purposes of classification only, unplanned release puffs through the plant vent may use one hour integrated resolution times.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 42 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.3 40 CFR 190 Dose Calculations 6.3.1 Pathway Calculation of total uranium fuel cycle dose for purposes of demonstrating compliance with 40 CFR 190 requires the contributions from liquid and gaseous effluent as well as direct radiation from selected outside storage tanks and storage buildings. The total uranium fuel cycle dose to any member of the public will be calculated by summing the following doses:

Direct Radiation Dose

Liquid Effluent Dose

Noble Gas Dose

Radioiodine, Tritium and Particulate Gaseous Effluent Dose v20 CAPA-8u3r34.DOC 08 0901.1100

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a. Direct Radiation Dose Routine determination of direct radiation dose from selected outside storage tanks and storage buildings may be made by direct survey measurements, derived from area TLD data, or calculated by shielding code.

The direct radiation dose will also take into account residence times near the site based upon land use census information.

The direct radiation determination using environmental TLD is given by equation 43.

Frad] 12 L -( xrb)

D s~b b-r rb b x D rxBxe 0 Sb (43 (43) where:

D'sb = the dose rate at the site boundary, in mrem D'ro = the dose rate from the dosimetry reading, in mrem radj = the distance from the point source to the dosimetry, in meter rsb = the distance from the point source to site boundary, in meter B = buildup factor

= 1+(u xrb x((/a +s

[La = total absorption coefficient

= 0.001x e(1.34 - (0.105 x (ln(10 x E) - 1.57)2 x (273°kTk)

ýts = total Compton scatter coefficient

= 0.001 x e (3.10 - (0.089 x (1n(10 x E) + 1.89) )) x (273ok/Tok) go = total attenuation coefficient

= ýIa + ýLls E = external effective average gamma energy per disintegration of the source (Mev)

Tok = average absolute temperature (Kelvin) v20_CAPA-8u3r34.DOC 08 0901.1100

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 44 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

b. Noble Gas Dose The noble gas skin dose and total body dose contributions to the total uranium fuel cycle dose to a member of the public will be determined as shown in Equations 44 and 45.

Noble Gas Total Body Dose= 3.17 x 108 (X/Q)R ZKLQi (44) i Noble Gas Skin Dose = 3.17 x I 0-8 (ZIQ)R I (Li + 1.1IMi )(_i (45)

Where:

3.17x10 8 Conversion constant yr/sec.

Qz/-)R = Maximum historical dispersion factor for receptor of interest, based on 5 year averages from Appendix 10.2.

Ki Whole body dose factor for nuclide i, in mrem/yr per ýtCi/m 3 . Values are listed in Table 6.3.

Li Skin dose factor for nuclide i, in mrem/yr per jiCi/m3 . Values are listed in Table 6.3.

1.1 Conversion factor mrem/mrad. Converts air dose to skin dose.

Mi Gamma air dose factor for nuclide i, in mrad/yr per pCi/m 3 . Values are listed in Appendix 10.3.

Q, Total release of noble gas radionuclide, "i", in gCi/sec.

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c. Liquid and Gaseous Effluent Dose The doses from liquid effluents and radioiodines, tritium and particulates in gaseous effluents will be determined by Equations 1 and 33, respectively.

For purposes of calculating the dose required by the radiological effluent controls, more realistic assumptions concerning the liquid and gaseous effluent dose pathways will be used, based upon the most recent land use census data as well as the latest environmental monitoring information.

These assumptions may include, but not be limited to: more realistic liquid dilution factors, location and age of actual individuals, site specific food pathway parameters, and documentation of true food consumption. Other assumptions may be used provided they can be substantiated by census or direct measurement.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 46 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.4 On-Site Dose to Members of the Public Members of the public are occasionally granted access within the site boundary, but only in the owner controlled area up to the protected area boundary. The most common public access activities are: tours to the simulator (training building) or Bio Lab, policemen using the shooting range (most frequent activity), cattle drives through to adjacent properties, and visits of American Indians to on-site burial grounds (closest to the plant).

Exposure to members of the public due to liquid releases while on-site is highly unlikely and therefore not addressed. Exposure due to gaseous releases and direct radiation are credible and therefore are considered.

The dose to members of the public during on-site activities will be primarily determined by the duration of the on-site visitation time and by the closest proximity to the plant.

For gaseous releases the doses are calculated using Equations 44, 45 and 33. The Ri's in Equation 33 consider only the inhalation and ground plane pathway and exclude the infant age group.

The X/Q and D/Q values are modified using logarithmic extrapolation from the site boundary to the on-site location of interest as shown in Equations 46 and 47.

1og[X/Q]~.sj 1e lg[X/Q]SB... - log[X/Q]°c.0 og,(dist.on - site) - log(dist.S.B.)](4 6 )

log(dzrst.S.B.) - jo~iso.

+ log[X/Q]s.i 1og[D/Q]s.B. - 1og[D/Q] 0 c. [log(dist on- site) - log(dist.S.B.)] (47) log(dist.S.B.)- log(distloc.)

+ 1og[D/Q]s B Based upon Regulatory Guide 1.111, these equations can be expected to provide reasonable dispersion and deposition estimates for distances as close as 200 meters.

Determination of direct radiation dose from the reactor units and from outside storage tanks may be made by direct survey measurements, derived from environmental TLD data, or calculated by shielding code.

A distance of 200 meters from the plant (both units) equidistant from the plant vent is arbitrarily selected as the closest perimeter for which on-site doses will be calculated.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE I PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 47 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 The activities of the members-of-the-public while on-site (described above), are at or beyond 200 meters. Table 6.3 details the types of on-site activities that members-of-the-public might be expected to participate in at DCPP. The sectors and closest distances in which they may visit as well as expected visitation duration are also shown (based on Security Section information).

Table 6.3 Expected On-Site Distimces and Visitation Times for Members of the Public SECTOR CLOSEST POINT AVERAGE EXPECTED ONSITE MEMBER OF OF APPROACH VISITATION OF THE PUBLIC VISITATION TO PLANT TIME PER YEAR Police at SE 700m 208 hours0.00241 days 0.0578 hours 3.439153e-4 weeks 7.9144e-5 months shooting range Tour Participants Simulator Bldg S (SE) 310m 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Overlook E 210m 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months American Indians NW 200m 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months at burial grounds NNW 200m 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months Ranch hands driving NW 250m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months cattle around site NNW 350m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months N 320m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months NNE 450m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months NE 630m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months v20_CAPA-8u3r34.DOC 08 0901.1100

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7. ACCEPTANCE CRITERIA 7.1 There is no quantitative acceptance for this procedure. If the task or analysis has been accomplished within the bounds of this procedure, it is considered acceptable.

8. REFERENCES 8.1 License Amendment 67/66, January 22, 1992.

8.2 Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, Regulatory Guide 1.109, Rev. 0, March 1976.

8.3 Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, Regulatory Guide 1.109, Rev. 1, October 1977.

8.4 Preparation of Radiological Effluent Tech Specs for Nuclear Power Plants, NUREG No. 0133, October 1978.

8.5 LADTAP II - Technical Reference and User Guide, NUREG/CR-4013.

8.6 Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel Cycle Standard 40 CFR 190, NUREG No. 0543, January 1980.

8.7 Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors, Regulatory Guide, 1.111, Rev. 1, July, 1977.

8.8 Radioactive Decay Data Tables, David C. Kocher. DOE/TIC-11026, 1981.

8.9 CAP A-6, "Gaseous Radwaste Discharge Management."

8.10 CAP A-5, "Liquid Radwaste Discharge Management."

8.11 CAP D-15, "Steam Generator Leak Rate Determination."

8.12 CAP D-19, "Correlation of Rad Monitors to Radioactivity."

8.13 CY2.DC 1, "Radiation Monitoring System High Alarm Setpoint Control Procedure."

8.14 CY2.IID1, "Radiological Effluent and Controls Program" (RECP) 8.15 "Setpoint Calculation for Containment Ventilation Exhaust Monitor,"

Calc # NSP-1&2-39-44, 10/92 and 11/92 and AR A0430610.

8.16 NUREG 2919, Computer Code XOQDOQ, Revision 2, September, 1982.

8.17 Meteorology Services Report Number 420DC.08.15, March 2008 (XOQDOQ data) 8.18 Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake, NUREG-0172, November 1977.

8.19 Include Tc-99M In ODCM and ARER Reports, Action Request A0619601.

8.20 Rad Effluent Sampling of Ni-63, Action Request A0619600.

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8.22 "DRADCALC.xls Computer Program Verification and Validation Report", Revision 0, June 1997.

8.23 CAP A-8, "Off-Site Dose Calculation", Revision 10 (direct radiation calculation).

9. RECORDS 9.1 Data Sheets and records will be maintained in the Records Management System (RMS) in accordance with CYl .DC1, "Analytical Data Processing Responsibilities."

10. APPENDICES 10.1 LRW Composite Dose Factors, Aio, For Adults At A Saltwater Site (mrem/hour per RCi/ml) organ "0" 10.2 Summary Of Land Use Census Evaluation 10.3 GRW Dose Factors For Noble Gases 10.4 Child Inhalation Pathway Dose Factors For Worst Case Organ 10.5 Ground Plane Dose Factors 10.6 GRW Dose Parameters For Radioiodines, Radioactive Particulates, and Any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW) 3 10.6.1 Infant Age Group, Inhalation Pathway Organ "0"(mrem/yr per [tCi/m )

R ilnhal 10.6.2 Child Age Group, Inhalation Pathway Organ "0"(mrem/yr per ýiCi/m 3)

. Tienhal 10.6.3 Teen Age Group, Inhalation Pathway Organ "0"(mrem/yr per ýICi/m 3) Rithal 10.6.4 Adult Age Group, Inhalation Pathway Organ "0" (mrem/yr per ýtCi/m3)

Rilnhal 10.6.5 Child Age Group, Vegetation Pathway Organ "0"(mrem/yr per [tCi/(sec in 2)) Ri'vegi 10.6.6 Teen Age Group, Vegetation Pathway Organ "0" (mrem/yr per .iCi/(sec M2 )) Rivegi 10.6.7 Adult Age Group, Vegetation Pathway Organ "0" (mrem/yr per ptCi/(sec M2 )) Rivegi

11. ATTACHMENTS 11.1 "Liquid Discharges (LRW) Monitored for Radioactivity," 10/04/00 11.2 "Gaseous Releases (GRW) Monitored for Radioactivity," 10/31/00 v20 CAPA-8u3r34.DOC 08 0901.1100

PSRUNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 50 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND2 Appendix: 10.1 LRW Composite Dose Factors', Ai,,

for Adults at a Saltwater Site (mrem/hour per ptCi/ml) organ "0" Nuclide Tot Body Thyroid Kidney Lung GI-LLI Bone Liver H--3 1.61E-01 1.61E-01 1.61E-01 1.61E-01 1.61E-01 0.OOE+00 1.61E-01 Nia-24 4.57E-01 4.57E-01 4.57E-01 4.57E-01 4.57E-01 4.57E-01 4.57E-01 Cr -51 5.58E+00 3.34E+00 1.23E+00 7.40E+00 1.40E+03 0.OOE+00 0.OOE+00 Mn-54 1.35E+03 O.OOE+00 2.1OE+03 0.OOE+00 2.16E+04 0.OOE+00 7.06E+03 Mn-56 3.15E+01 0.OOE+00 2.26E+02 0.OOE+00 5.67E+03 0.OOE+00 1.78E+02 Fe -55 8.23E+03 0.00E+00 0.OOE+00 1.97E+04 2.03E+04 5.11E+04 3.53E+04 Fe -59 7.27E+04 0.00E+00 O.OOE+00 5.30E+04 6.32E+05 8.06E+04 1.90E+05 Cc)-57 2.36E+02 0.00E+00 0.OOE+00 0.OOE+00 3.59E+03 0.OOE+00 1.42E+02 Cc)-58 1.35E+03 0.OOE+00 0.OOE+00 0.OOE+00 1.22E+04 0.OOE+00 6.03E+02 CN)-60 3.82E+03 0.OOE+00 0.OOE+00 0.OOE+00 3.25E+04 0.OOE+00 1.73E+03 Ni -63 1.67E+03 0.OOE+00 0.OOE+00 0.OOE+00 7.18E+02 4.96E+04 3.44E+03 Nii-65 1.20E+01 0.00E+00 0.OOE+00 0.00E+00 6.65E+02 2.02E+02 2.62E+01 Cu1-64 1.01E+02 O.OOE+00 5.40E+02 0.00E+00 1.83E+04 0.OOE+00 2.14E+02 Znn-65 2.32E+05 0.OOE+00 3.43E+05 0.OOE+00 3.23E+05 1.61E+05 5.13E+05 Zrn-69 4.56E+01 0.OOE+00 4.26E+02 0.00E+00 9.85E+01 3.43E+02 6.56E+02 Ass-76 4.42E+01 0.OOE+00 8.72E+01 0.0OE+00 0.00E+00 0.OOE+00 4.62E+0 1 Brr-82 4.07E+00 0.OOE+00 0.OOE+00 0.OOE+00 4.67E+00 0.OOE+00 0.OOE+00 Br--84 9.39E-02 0.OOE+00 0.OOE+00 0.OOE+00 7.37E-07 0.OOE+00 0.OOE+00 Rib-86 2.91E+02 0.OOE+00 O.OOE+00 0.OOE+00 1.23E+02 0.OOE+00 6.24E+02 Ri -88 9.49E-01 0.00E+00 O.OOE+00 0.OOE+00 2.47E-11 0.OOE+00 1.79E+00 RIb-89+D 8.34E-01 0.OOE+00 0.OOE+00 0.OOE+00 6.89E-14 0.OOE+00 1.19E+00 Sr -89+D 1.43E+02 0.OOE+00 0.OOE+00 0.OOE+00 8.OOE+02 4.99E+03 0.OOE+00 Sr-90+D 2.83E+03 0.OOE+00 0.00E+00 0.OOE+00 3.55E+03 1.41E+05 0.OOE+00 Sr-91+D 3.71E+00 0.OOE+00 0.0OE+00 0.OOE+00 4.37E+02 9.18E+01 0.OOE+00 Sr'-92+D 1.51E+00 0.OOE+00 0.OOE+00 0.OOE+00 6.90E+02 3.48E+01 0.OOE+00 y--90 1.63E-01 0.OOE+00 0.OOE+00 0.OOE+00 6.42E+04 6.06E+00 0.OOE+00 y--91m+D 2.22E-03 0.OOE+00 0.OOE+00 0.OOE+00 1.68E-01 5.73E-02 0.OOE+00 y--92 1.56E-02 0.OOE+00 0.OOE+00 0.OOE+00 9.32E+03 5.32E-01 0.OOE+00 Zr -95+D 3.46E+00 0.OOE+00 8.02E+00 0.OOE+00 1.62E+04 1.59E+01 5.11E+00 Zrr-97+D 8.13E-02 0.OOE+00 2.68E-01 O.OOE+00 5.5 1E+04 8.8 1E-01 1.78E-01 Nib-95 1.34E+02 0.OOE+00 2.46E+02 0.OOE+00 1.51E+06 4.47E+02 2.49E+02 Mo-99+D 2.43E+01 O.OOE+00 2.89E+02 0.OOE+00 2.96E+02 0.OOE+00 1.28E+02 Tc -101 1.88E-01 O.OOE+00 3.46E-01 9.8 1E-03 5.77E-14 1.33E-02 1.92E-02 Riu-103+D 4.60E+01 0.00E+00 4.07E+02 0.OOE+00 1.25E+04 1.07E+02 0.OOE+00 Riu-105+D 3.51E+00 0.OOE+00 1.15E+02 0.OOE+00 5.44E+03 8.89E+00 0.OOE+00 Ri:u-106+D 2.01E+02 0.OOE+00 3.06E+03 0.OOE+00 1.03E+05 1.59E+03 0.OOE+00 A g-110m+D 8.60E+02 0.OOE+00 2.85E+03 0.OOE+00 5.91E+05 1.56E+03 1.45E+03 Srn-113 3.53E+03 9.85E+02 0.OOE+00 0.OOE+00 0.00E+00 6.06E+04 1.66E+03 Sr -1 17m 8.76E+02 2.52E+02 0.OOE+00 0.OOE+00 0.OOE+00 3.02E+03 3.41E+02 St b-122 6.65E+00 3.09E-01 0.OOE+00 1.18E+01 0.OOE+00 2.19E+01 4.47E-01 Stb-124 1.09E+02 6.70E-01 O.OOE+00 2.15E+02 7.84E+03 2.76E+02 5.22E+00 Stb-125 4.20E+01 1.79E-01 0.OOE+00 1.36E+02 1.94E+03 1.77E+02 1.97E+00 v20_CAPA-8u3r34.DOC 08 0901.1100

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 51 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.1 (continued)

LRW Composite Dose Factors', Aio, for Adults at a Saltwater Site (mremihour per pCi/ml) organ "0" Nuclide Tot Body Thyroid Kidney Lung GI-LLI Bone Liver Te-125m 2.91E+01 6.51E+01 8.82E+02 0.OOE+00 8.66E+02 2.17E+02 7.86E+01 Te-129m+D 1.47E+02 3.20E+02 3.89E+03 O.OOE+00 4.69E+03 9.31E+02 3.47E+02 Te-129 6.19E-01 1.95E+00 1.07E+0I O.OOE+00 1.92E+00 2.54E+00 9.55E-01 Te-131m+D 5.71E+O1 1.08E+02 6.94E+02 O.OOE+00 6.80E+03 1.40E+02 6.85E+01 Te-131+D 5.03E-01 1.31E+00 6.99E+00 O.OOE+00 2.26E-01 1.59E+00 6.66E-01 Te-132+D 1.24E+02 1.46E+02 1.27E+03 O.OOE+00 6.24E+03 2.04E+02 1.32E+02 1-131+D 1.79E+02 1.02E+05 5.35E+02 O.OOE+00 8.23E+01 2.18E+02 3.12E+02 1-132 9.96E+00 9.96E+02 4.54E+01 O.OOE+00 5.35E+00 1.06E+01 2.85E+01 1-133+D 3.95E+01 1.90E+04 2.26E+02 0.OOE+00 1.16E+02 7.45E+01 1.30E+02 1-134 5.40E+00 2.62E+02 2.40E+01 O.OOE+00 1.32E-02 5.56E+00 1.51E+01 1-135+D 2.24E+01 4.01E+03 9.75E+01 0.OOE+00 6.87E+01 2.32E+01 6.08E+01 Cs-134 1.33E+04 0.OOE+00 5.27E+03 1.75E+03 2.85E+02 6.84E+03 1.63E+04 Cs-136 2.04E+03 0.OOE+00 1.57E+03 2.16E+02 3.21E+02 7.16E+02 2.83E+03 Cs-137+D 7.85E+03 0.OOE+00 4.07E+03 1.35E+03 2.32E+02 8.77E+03 1.20E+04 Cs-138 5.94E+00 0.OOE+00 8.81E+00 8.70E-01 5.12E-05 6.07E+00 1.20E+01 Ba-139 2.30E-01 0.OOE+00 5.23E-03 3.17E-03 1.39E+01 7.85E+00 5.59E-03 Ba-140+D 1.08E+02 0.OOE+00 7.02E-01 1.18E+00 3.38E+03 1.64E+03 2.06E+00 Ba-141+D 1.29E-01 O.OOE+00 2.68E-03 1.63E-03 1.80E-09 3.81E+00 2.88E-03 La-140 2.1OE-01 0.OOE+00 0.OOE+00 O.OOE+00 5.83E+04 1.57E+00 7.94E-01 La-142 9.13E-03 0.00E+00 O.OOE+00 O.OOE+00 2.68E+02 8.06E-02 3.67E-02 Ce-141 2.63E-01 O.OOE+00 1.08E+00 0.OOE+00 8.86E+03 3.43E+00 2.32E+00 Ce-143+D 4.94E-02 O.OOE+00 1.97E-01 0.00E+00 1.67E+04 6.04E-01 4.46E+02 Ce-144+D 9.59E+00 O.OOE+00 4.43E+01 O.OOE+00 6.04E+04 1.79E+02 7.47E+01 Pr-144 9.64E-04 O.OOE+00 4.44E-03 0.OOE+00 2.73E-09 1.90E-02 7.87E-03 Nd-147+D 2.74E-01 0.OOE+00 2.68E+00 0.OOE+00 2.20E+04 3.96E+00 4.58E+00 Pu-238 2.07E+03 O.OOE+00 8.87E+03 0.00E+00 8.85E+03 7.62E+04 9.66E+03 Pu-239 2.31E+03 O.OOE+00 9.83E+03 O.OOE+00 8.07E+03 8.79E+04 1.06E+04 Pu-240 2.31E+03 O.OOE+00 9.82E+03 O.OOE+00 8.23E+03 8.76E+04 1.05E+04 Pu-241+D 4.01E+O1 O.OOE+00 1.85E+02 O.OOE+00 1.70E+02 1.90E+03 9.03E+01 Pu-242 2.23E+03 0.OOE+00 9.46E+03 O.OOE+00 7.91E+03 8.13E+04 1.02E+04 U-233+D 1.56E+03 0.00E+00 6.02E+03 O.OOE+00 1.86E+03 2.58E+04 0.OOE+00 U-234 1.53E+03 0.OOE+00 5.90E+03 O.OOE+00 1.82E+03 2.48E+04 0.OOE+00 U-235+D 1.44E+03 0.OOE+00 5.54E+03 0.OOE+00 2.31E+03 2.37E+04 0.OOE+00 U-236 1.47E+03 0.00E+00 5.66E+03 O.OOE+00 1.71E+03 2.37E+04 O.OOE+00 U-238 D 1.35E+03 O.OOE+00 5.19E+03 O.OOE+00 1.63E+03 2.27E+04 O.OOE+00 W-187 2.68E+00 O.OOE+00 O.OOE+00 0.OOE+00 2.51E+03 9.16E+00 7.66E+00 Np-239 1.91E-03 0.OOE+00 1.08E-02 O.OOE+00 7.11E+02 3.53E-02 3.47E-03 Dose factors are based upon NUREG 0133 methodology.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 52 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.2 Summary of Land Use Census Evaluation Sector Receptor Distance X/Q D/Q Comments Description (miles)

N none no receptors within 5 miles NNE residence + garden 4.4 4.80E-8 9.50E-11 full time occupancy residence 3.3 7.30E-08 1.60E-10 trailer - limited use residence 3.2 7.70E-08 1.70E-10 cabin - limited use NE residence 4.9 3.30E-08 6.20E- 11 full time occupancy ENE residence 4.4 3.10E-08 7.10E-1 1 full time occupancy residence 5.0 2.50E-08 5.60E- 11 full time occupancy E residence 4.0 4.80E-08 1.30E-10 cabin - limited use residence 3.7 5.30E-08 1.50E-10 part time occupancy residence + garden 4.5 4.OOE-08 L.1OE-10 full time occupancy ESE oat hay and sugar 3.3 1.70E-07 L.1OE-09 field workers present only during the day -

peas critical receptor ground plane, inhalation, and vegetation ingestion dose assessed at this location SE none no receptors within 5 miles SSE none over water S none over water SSW none over water SW none over water WSW none over water W none over water WNW none over water NW highest site 0.5 5.OOE-6 1.80E-8 Gas effluent dose rates. PRRLs and boundary HASPs evaluated at this location.

dispersion value residence 1.2 1.10E-6 4.1OE-9 trailer - limited use residence 3.6 2.OOE-7 6.OOE-10 full time occupancy NNW residence 1.5 6.30E-07 2.OOE-09 full time occupancy (trailer) - critical receptor ground plane and inhalation dose assessed at this location Public campground 4.6 1.20E-7 2.80E-10 Ranger Station 4.6 1.20E-07 2.80E-10 Occupied during normal work hours v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE I PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE. 53 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND2 Appendix: 10.3 GRW Dose Factors for Noble Gases' Whole Body Gamma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor Ki Li M i Ni (mremi/yr per ýtCi/m3) (mrad/yr per ýtCi/m3) (mrad/yr per [tCi/m3)

Radionuclide (mrem/yr per ýtCi/m 3)

Kr-83m 7.56E-02 1.93E+01 2.88E+02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11 E+03 Xe-133m 2.5 1E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.1 1E+02 3.36E+03 7.39E+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.2 1E+03 4.75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03 From Table B-1 of Regulatory Guide 1.109 (Rev. 1, Oct. 1977) v20 CAP A-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 54 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.4 Child Inhalation Pathway Dose Factors for Worst Case Organ Radionuclide Piw H-3 6.40E+02 CR-51 1.70E+04 MN-54 1.58E+06 FE-59 1.27E+06 CO-58 1.11E+06 CO-60 7.07E+06 ZN-65 9.95E+05 RB-86 1.98E+05 SR-89 2.16E+06 Y-90 2.68E+05 SR-90 3.85E+07 ZR-95 2.23E+06 NB-95 6.14E+05 RU-103 6.62E+05 RU-106 1.43E+07 AG-110M 5.48E+06 SB-124 3.24E+06 SB-125 2.32E+06 TE-129M 1.76E+06 1-131 1.62E+07 1-133 3.85E+06 CS-134 1.01E+06 CS-136 1.71E+05 CS-137 9.07E+05 BA-140 1.74E+06 CE-141 5.44E+05 CE-144 1.20E+07 ND-147 3.28E+05 v20_CAPA-8u3r34.DOC 08 0901.1100

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 55 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.5 Ground Plane Dose Factors GRW Dose Parameters', Ri.Gp for Radioiodines, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW),

any Age Group, Ground Plane Pathway (mrem/yr per ptCi/(sec in 2 ))

Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI H-3 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 CR-51 4.65E+06 4.65E+06 4.65E+06 4.65E+06 4.65E+06 4.65E+06 4.65E+06 MN-54 1.38E+09 1.38E+09 1.38E+09 1.38E+09 1.38E+09 1.38E+09 1.38E+09 FE-59 2.73E+08 2.73E+08 2.73E+08 2.73E+08 2.73E+08 2.73E+08 2.73E+08 CO-58 3.80E+08 3.80E+08 3.80E+08 3.80E+08 3.80E+08 3.80E+08 3.80E+08 CO-60 2.15E+10 2.15E+10 2.15E+10 2.15E+10 2.15E+10 2.15E+10 2.15E+ 10 ZN-65 7.46E+08 7.46E+08 7.46E+08 7.46E+08 7.46E+08 7.46E+08 7.46E+08 RB-86 8.98E+06 8.98E+06 8.98E+06 8.98E+06 8.98E+06 8.98E+06 8.98E+06 SR-89 2.16E+04 2.16E+04 2.16E+04 2.16E+04 2.16E+04 2.16E+04 2.16E+04 Y-90 4.50E+03 4.50E+03 4.50E+03 4.50E+03 4.50E+03 4.50E+03 4.50E+03 SR-90 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 0.OOE+00 ZR-95 2.45E+08 2.45E+08 2.45E+08 2.45E+08 2.45E+08 2.45E+08 2.45E+08 NB-95 1.37E+08 1.37E+08 1.37E+08 1.37E+08 1.37E+08 1.37E+08 1.37E+08 RU-103 1.08E+08 1.08E+08 1.08E+08 1.08E+08 1.08E+08 1.08E+08 1.08E+08 RU-106 4.20E+08 4.20E+08 4.20E+08 4.20E+08 4.20E+08 4.20E+08 4.20E+08 AG-110M 3.45E+09 3.45E+09 3.45E+09 3.45E+09 3.45E+09 3.45E+09 3.45E+09 SB-124 5.99E+08 5.99E+08 5.99E+08 5.99E+08 5.99E+08 5.99E+08 5.99E+08 SB-125 2.34E+09 2.34E+09 2.34E+09 2.34E+09 2.34E+09 2.34E+09 2.34E+09 TE-129M 1.98E+07 1.98E+07 1.98E+07 1.98E+07 1.98E+07 1.98E+07 1.98E+07 1-131 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1-133 2.45E+06 2.45E+06 2.45E+06 2.45E+06 2.45E+06 2.45E+06 2.45E+06 CS-134 6.90E+09 6.90E+09 6.90E+09 6.90E+09 6.90E+09 6.90E+09 6.90E+09 CS-136 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.5 1E+08 CS-137 1.03E+10 1.03E+10 1.03E+10 1.03E+10 1.03E+10 1.03E+10 1.03E+10 BA-140 2.05E+07 2.05E+07 2.05E+07 2.05E+07 2.05E+07 2.05E+07 2.05E+07 CE-141 1.37E+07 1.37E+07 1.37E+07 1.37E+07 1.37E+07 1.37E+07 1.37E+07 CE-144 6.96E+07 6.96E+07 6.96E+07 6.96E+07 6.96E+07 6.96E+07 6.96E+07 ND-147 8.39E+06 8.39E+06 8.39E+06 8.39E+06 8.39E+06 8.39E+06 8.39E+06 Dose factors are based upon NUREG 0133 methodology.

v20_CAPA-1u3r34.DOC 08 0901.1100

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 56 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.6 GRW Dose Parameters for Radioiodines, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW) v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE I PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 57 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND2 Appendix: 10.6.1 GRW Dose Parameters' for Radioiodmies, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW), Infant Age Group, Inhalation Pathway Organ "0" (mrem/yr per iýCi/m 3 ) Rit,,hal Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI H-3 0.OOE+00 3.68E+02 3.68E+02 3.68E+02 3.68E+02 3.68E+02 3.68E+02 CR-51 0.OOE+00 0.OOE+00 8.95E+01 5.75E+01 1.32E+01 1.28E+04 3.57E+02 MN-54 0.OOE+00 2.53E+04 4.98E+03 0.OOE+00 4.98E+03 1.00E+06 7.06E+03 FE-59 1.36E+04 2.35E+04 9.48E+03 0.OOE+00 0.OOE+00 1.02E+06 2.48E+04 CO-58 0.OOE+00 1.22E+03 1.82E+03 0.OOE+00 0.OOE+00 7.77E+05 1.11E+04 CO-60 0.OOE+00 8.02E+03 1.18E+04 0.OOE+00 0.OOE+00 4.5 1E+06 3.19E+04 ZN-65 1.93E+04 6.26E+04 3.11E+04 0.OOE+00 3.25E+04 6.47E+05 5.14E+04 RB-86 0.OOE+00 1.90E+05 8.82E+04 0.OOE+00 0.OOE+00 0.OOE+00 3.04E+03 SR-89 3.98E+05 0.OOE+00 1.14E+04 0.OOE+00 0.OOE+00 2.03E+06 6.40E+04 Y-90 3.29E+03 0.OOE+00 8.82E+01 0.OOE+00 0.OOE+00 2.69E+05 1.04E+05 SR-90 1.55E+07 0.OOE+00 3.12E+05 0.OOE+00 0.OOE+00 1.12E+07 1.31E+05 ZR-95 1.15E+05 2.79E+04 2.03E+04 0.OOE+00 3.11E+04 1.75E+06 2.17E+04 NB-95 1.57E+04 6.43E+03 3.78E+03 0.OOE+00 4.72E+03 4.79E+05 1.27E+04 RU-103 2.02E+03 0.OOE+00 6.79E+02 0.OOE+00 4.24E+03 5.52E+05 1.61 E+04 RU-106 8.68E+04 0.OOE+00 1.09E+04 0.OOE+00 1.07E+05 1.16E+07 1.64E+05 AG-110M 9.98E+03 7.22E+03 5.OOE+03 0.00E+00 1.09E+04 3.67E+06 3.30E+04 SB-124 3.79E+04 5.56E+02 1.20E+04 1.0 1E+02 0.OOE+00 2.65E+06 5.91E+04 SB-125 5.17E+04 4.77E+02 1.09E+04 6.23E+01 0.OOE+00 1.64E+06 1.47E+04 TE-129M 1.41E+04 6.09E+03 2.23E+03 5.47E+03 3.18E+04 1.68E+06 6.90E+04 1-131 3.79E+04 4.44E+04 1.96E+04 1.48E+07 5.18E+04 0.OOE+00 1.06E+03 1-133 1.32E+04 1.92E+04 5.60E+03 3.56E+06 2.24E+04 0.OOE+00 2.16E+03 CS-134 3.96E+05 7.03E+05 7.45E+04 0.OOE+00 1.90E+05 7.97E+04 1.33E+03 CS-136 4.83E+04 1.35E+05 5.29E+04 0.OOE+00 5.64E+04 1.18E+04 1.43E+03 CS-137 5.49E+05 6.12E+05 4.55E+04 0.00E+00 1.72E+05 7.13E+04 1.33E+03 BA-140 5.60E+04 5.60E+01 2.90E+03 0.OOE+00 1.34E+01 1.60E+06 3.84E+04 CE-141 2.77E+04 1.67E+04 1.99E+03 0.OOE+00 5.25E+03 5.17E+05 2.16E+04 CE-144 3.19E+06 1.21E+06 1.76E+05 0.00E+00 5.38E+05 9.84E+06 1.48E+05 ND-147 7.94E+03 8.13E+03 5.OOE+02 0.OOE+00 3.15E+03 3.22E+05 3.12E+04 Dose factors are based upon NUREG 0133 methodology.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 58 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.6.2 GRW Dose Parameters1 for Radioiodines, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW), Child Age Group, Inhalation Pathway Organ "0" (mrem/yr per pCi/mr3) Rinhal Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI H-3 0.OOE+00 6.40E+02 6.40E+02 6.40E+02 6.40E+02 6.40E+02 6.40E+02 CR-51 O.OOE+00 O.00E+00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 MN-54 O.OOE+00 4.29E+04 9.5 1E+03 0.OOE+00 1.00E+04 1.58E+06 2.29E+04 FE-59 2.07E+04 3.34E+04 1.67E+04 0.OOE+00 0.OOE+00 1.27E+06 7.07E+04 CO-58 0.OOE+00 1.77E+03 3.16E+03 0.OOE+00 0.OOE+00 1.11E+06 3.44E+04 CO-60 0.OOE+00 1.31E+04 2.26E+04 0.OOE+00 0.OOE+00 7.07E+06 9.62E+04 ZN-65 4.26E+04 1.13E+05 7.03E+04 0.OOE+00 7.14E+04 9.95E+05 1.63E+04 RB-86 0.00E+00 1.98E+05 1.14E+05 0.OOE+00 0.OOE+00 0.OOE+00 7.99E+03 SR-89 5.99E+05 0.OOE+00 1.72E+04 0.OOE+00 0.OOE+00 2.16E+06 1.67E+05 Y-90 4.11E+03 O.OOE+00 1.11 E+02 0.OOE+00 0.OOE+00 2.62E+05 2.68E+05 SR-90 3.85E+07 0.OOE+00 7.66E+05 0.OOE+00 0.OOE+00 1.48E+07 3.43E+05 ZR-95 1.90E+05 4.18E+04 3.70E+04 0.OOE+00 5.96E+04 2.23E+06 6.11E+04 NB-95 2.35E+04 9.18E+03 6.55E+03 0.OOE+00 8.62E+03 6.14E+05 3.70E+04 RU-103 2.79E+03 0.OOE+00 1.07E+03 0.OOE+00 7.03E+03 6.62E+05 4.48E+04 RU-106 1.36E+05 0.OOE+00 1.69E+04 0.OOE+00 1.84E+05 1.43E+07 4.29E+05 AG-110M 1.69E+04 1.14E+04 9.14E+03 0.OOE+00 2.12E+04 5.48E+06 1.00E+05 SB-124 5.74E+04 7.40E+02 2.OOE+04 1.26E+02 0.OOE+00 3.24E+06 1.64E+05 SB-125 9.84E+04 7.59E+02 2.07E+04 9.1 OE+0 1 0.OOE+00 2.32E+06 4.03E+04 TE-129M 1.92E+04 6.85E+03 3.04E+03 6.33E+03 5.03E+04 1.76E+06 1.82E+05 1-131 4.81E+04 4.8 1E+04 2.73E+04 1.62E+07 7.88E+04 0.OOE+00 2.84E+03 1-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 0.OOE+00 5.48E+03 CS-134 6.5 1E+05 1.01E+06 2.25E+05 0.OOE+00 3.30E+05 1.21E+05 3.85E+03 CS-136 6.5 1E+04 1.71E+05 1.16E+05 0.00E+00 9.55E+04 1.45E+04 4.18E+03 CS-137 9.07E+05 8.25E+05 1.28E+05 0.OOE+00 2.82E+05 1.04E+05 3.62E+03 BA-140 7.40E+04 6.48E+01 4.33E+03 0.OOE+00 2.11E+01 1.74E+06 1.02E+05 CE-141 3.92E+04 1.95E+04 2.90E+03 0.OOE+00 8.55E+03 5.44E+05 5.66E+04 CE-144 6.77E+06 2.12E+06 3.61E+05 0.OOE+00 1.17E+06 1.20E+07 3.89E+05 ND-147 1.08E+04 8.73E+03 6.8 1E+02 0.OOE+00 4.8 1E+03 3.28E+05 8.21E+04 Dose factors are based upon NUREG 0133 methodology.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 59 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.6.3 I GRW Dose Parameters' for Radioiodines, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW), Teen Age Group, Inhalation Pathway Organ "0" (mrem/yr per ptCi/m 3) Rji.n,aI Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI H-3 O.00E+00 7.25E+02 7.25E+02 7.25E+02 7.25E+02 7.25E+02 7.25E+02 CR-51 O.OOE+00 0.OOE+00 1.35E+02 7.50E+01 3.07E+01 2.1OE+04 3.00E+03 MN-54 O.00E+00 5.11E+04 8.40E+03 0.00E+00 1.27E+04 1.98E+06 6.68E+04 FE-59 1.59E+04 3.70E+04 1.43E+04 0.OOE+00 0.OOE+00 1.53E+06 1.78E+05 CO-58 O.OOE+00 2.07E+03 2.78E+03 0.OOE+00 0.OOE+00 1.34E+06 9.52E+04 CO-60 O.OOE+00 1.51E+04 1.98E+04 0.OOE+00 0.OOE+00 8.72E+06 2.59E+05 ZN-65 3.86E+04 1.34E+05 6.24E+04 0.OOE+00 8.64E+04 1.24E+06 4.66E+04 RB-86 0.OOE+00 1.90E+05 8.40E+04 0.OOE+00 0.OOE+00 0.OOE+00 1.77E+04 SR-89 4.34E+05 0.OOE+00 1.25E+04 0.OOE+00 0.OOE+00 2.42E+06 3.7 1E+05 Y-90 2.98E+03 O.00E+00 8.OOE+01 0.OOE+00 0.OOE+00 2.93E+05 5.59E+05 SR-90 3.3 1E+07 O.OOE+00 6.66E+05 0.OOE+00 0.OOE+00 1.65E+07 7.65E+05 ZR-95 1.46E+05 4.58E+04 3.15E+04 0.0OE+00 6.74E+04 2.69E+06 1.49E+05 NB-95 1.86E+04 1.03E+04 5.66E+03 0.OOE+00 1.00E+04 7.5 1E+05 9.68E+04 RU-103 2.10E+03 0.00E+00 8.96E+02 0.OOE+00 7.43E+03 7.83E+05 1.09E+05 RU-106 9.84E+04 O.OOE+00 1.24E+04 0.OOE+00 1.90E+05 1.61E+07 9.60E+05 AG-110M 1.38E+04 1.31E+04 7.99E+03 0.OOE+00 2.50E+04 6.75E+06 2.73E+05 SB-124 4.30E+04 7.94E+02 1.68E+04 9.76E+01 0.OOE+00 3.85E+06 3.98E+05 SB-125 7.38E+04 8.08E+02 1.72E+04 7.04E+0 1 0.00E+00 2.74E+06 9.92E+04 TE-129M 1.39E+04 6.58E+03 2.25E+03 4.58E+03 5.19E+04 1.98E+06 4.05E+05 1-131 3.54E+04 4.91E+04 2.64E+04 1.46E+07 8.40E+04 0.OOE+00 6.49E+03 1-133 1.22E+04 2.05E+04 6.22E+03 2.92E+06 3.59E+04 0.OOE+00 1.03E+04 CS-134 5.02E+05 1.13E+06 5.49E+05 0.OOE+00 3.75E+05 1.46E+05 9.76E+03 CS-136 5.15E+04 1.94E+05 1.37E+05 0.OOE+00 1.1OE+05 1.78E+04 1.09E+04 CS-137 6.70E+05 8.48E+05 3.11E+05 0.OOE+00 3.04E+05 1.21E+05 8.48E+03 BA-140 5.47E+04 6.70E+01 3.52E+03 0.OOE+00 2.28E+01 2.03E+06 2.29E+05 CE-141 2.84E+04 1.90E+04 2.17E+03 0.00E+00 8.88E+03 6.14E+05 1.26E+05 CE-144 4.89E+06 2.02E+06 2.62E+05 0.OOE+00 1.21E+06 1.34E+07 8.64E+05 ND-147 7.86E+03 8.56E+03 5.13E+02 0.OOE+00 5.02E+03 3.72E+05 1.82E+05 Dose factors are based upon NUREG 0133 methodology.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 60 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.6.4 GRW Dose Parameters' for Radioiodines, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW), Adult Age Group, Inhalation Pathway Organ "0" (mrem/yr per jiCi/m 3) Rilnhal Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI H-3 0.00E+00 7.18E+02 7.18E+02 7.18E+02 7.18E+02 7.18E+02 7.18E+02 CR-51 0.OOE+00 0.OOE+00 1.00E+02 5.95E+01 2.28E+OI 1.44E+04 3.32E+03 MN-54 0.00E+00 3.96E+04 6.30E+03 O.OOE+00 9.84E+03 1.40E+06 7.74E+04 FE-59 1.18E+04 2.78E+04 1.06E+04 O.OOE+00 0.OOE+00 1.02E+06 1.88E+05 CO-58 0.OOE+00 1.58E+03 2.07E+03 0.OOE+00 0.00E+00 9.28E+05 1.06E+05 CO-60 0.OOE+00 1.15E+04 1.48E+04 O.OOE+00 0.OOE+00 5.97E+06 2.85E+05 ZN-65 3.24E+04 1.03E+05 4.66E+04 O.OOE+00 6.90E+04 8.64E+05 5.34E+04 RB-86 0.OOE+00 1.35E+05 5.90E+04 O.OOE+00 0.OOE+00 O.OOE+00 1.66E+04 SR-89 3.04E+05 O.OOE+00 8.72E+03 O.OOE+00 O.OOE+00 1.40E+06 3.50E+05 Y-90 2.09E+03 O.OOE+00 5.61E+01 0.OOE+00 O.OOE+00 1.70E+05 5.06E+05 SR-90 2.87E+07 O.OOE+00 5.77E+05 O.OOE+00 0.OOE+00 9.60E+06 7.22E+05 ZR-95 1.07E+05 3.44E+04 2.33E+04 O.OOE+00 5.42E+04 1.77E+06 1.50E+05 NB-95 1.41E+04 7.82E+03 4.21E+03 O.OOE+00 7.74E+03 5.05E+05 1.04E+05 RU-103 1.53E+03 O.OOE+00 6.58E+02 O.OOE+00 5.83E+03 5.05E+05 1.10E+05 RU-106 6.91E+04 O.OOE+00 8.72E+03 0.OOE+00 1.34E+05 9.36E+06 9.12E+05 AG-110M 1.08E+04 1.00E+04 5.94E+03 0.OOE+00 1.97E+04 4.63E+06 3.02E+05 SB-124 3.12E+04 5.89E+02 1.24E+04 7.55E+01 0.OOE+00 2.48E+06 4.06E+05 SB-125 5.34E+04 5.95E+02 1.26E+04 5.40E+01 0.OOE+00 1.74E+06 1.01E+05 TE-129M 9.76E+03 4.67E+03 1.58E+03 3.44E+03 3.66E+04 1.16E+06 3.83E+05 1-131 2.52E+04 3.58E+04 2.05E+04 1.19E+07 6.13E+04 O.OOE+00 6.28E+03 1-133 8.64E+03 1.48E+04 4.52E+03 2.15E+06 2.58E+04 O.OOE+00 8.88E+03 CS-134 3.73E+05 8.48E+05 7.28E+05 O.OOE+00 2.87E+05 9.76E+04 1.04E+04 CS-136 3.90E+04 1.46E+05 1.1OE+05 O.00E+00 8.56E+04 1.20E+04 1.17E+04 CS-137 4.78E+05 6.21E+05 4.28E+05 0.OOE+00 2.22E+05 7.52E+04 8.40E+03 BA-140 3.90E+04 4.90E+01 2.57E+03 0.OOE+00 1.67E+01 1.27E+06 2.18E+05 CE-141 1.99E+04 1.35E+04 1.53E+03 0.OOE+00 6.26E+03 3.62E+05 1.20E+05 CE-144 3.43E+06 1.43E+06 1.84E+05 O.OOE+00 8.48E+05 7.78E+06 8.16E+05 ND-147 5.27E+03 6.1OE+03 3.65E+02 0.OOE+00 3.56E+03 2.21E+05 1.73E+05 Dose factors are based upon NUREG 0133 methodology.

v20_CAPA-8u3r34.DOC 08 0901.1100

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 61 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.6.5 GRW Dose Parameters' for Radioiodines, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW), Child Age Group, Vegetation Pathway Organ "0" (mrem/yr per ýtCi/(sec M2 )) RiVegi Nuclide Bone Liver T Body Thvroid Kidnev Lun2 GI-LLI 2

H-3 0.OOE+00 2.29E+03 2.29E+03 2.29E+03 2.29E+03 2.29E+03 2.29E+03 CR-51 0.00E+00 0.OOE+00 1.17E+05 6.49E+04 1.77E+04 1.18E+05 6.20E+06 MN-54 0.00E+00 6.65E+08 1.77E+08 O.OOE+00 1.86E+08 0.OOE+00 5.58E+08 FE-59 3.97E+08 6.42E+08 3.20E+08 0.OOE+00 0.OOE+00 1.86E+08 6.69E+08 CO-58 0.OOE+00 6.45E+07 1.97E+08 0.OOE+00 0.OOE+00 0.OOE+00 3.76E+08 CO-60 0.OOE+00 3.78E+08 1.12E+09 0.OOE+00 0.OOE+00 0.OOE+00 2.1OE+09 ZN-65 8.12E+08 2.16E+09 1.35E+09 0.OOE+00 1.36E+09 0.OOE+00 3.80E+08 RB-86 0.OOE+00 4.54E+08 2.79E+08 O.OOE+60 0.OOE+00 0.OOE+00 2.92E+07 SR-89 3.59E+10 0.OOE+00 1.03E+09 0.OOE+00 0.OOE+00 0.OOE+00 1.39E+09 Y-90 2.3 1E+04 0.OOE+00 6.18E+02 0.OOE+00 0.OOE+00 0.OOE+00 6.57E+07 SR-90 1.87E+12 0.OOE+00 3.77E+ 10 0.OOE+00 0.OOE+00 0.OOE+00 1.67E+10 ZR-95 3.86E+06 8.50E+05 7.56E+05 0.OOE+00 1.22E+06 0.OOE+00 8.86E+08 NB-95 4.12E+05 1.61E+05 1.15E+05 0.OOE+00 1.51E+05 0.00E+00 2.97E+08 RU-103 1.53E+07 0.OOE+00 5.89E+06 0.OOE+00 3.86E+07 0.OOE+00 3.96E+08 RU-106 7.45E+08 0.OOE+00 9.30E+07 0.OOE+00 1.01E+09 0.OOE+00 1.16E+10 AG-1 10M 3.21E+07 2.17E+07 1.74E+07 0.OOE+00 4.04E+07 0.OOE+00 2.58E+09 SB-124 3.52E+08 4.57E+06 1.23E+08 7.78E+05 0.OOE+00 1.96E+08 2.20E+09 SB-125 4.99E+08 3.85E+06 1.05E+08 4.62E+05 0.OOE+00 2.78E+08 1.19E+09 TE-129M 8.40E+08 2.35E+08 1.30E+08 2.71E+08 2.47E+09 0.OOE+00 1.02E+09 1-131 1.43E+08 1.44E+08 8.17E+07 4.75E+ 10 2.36E+08 0.OOE+00 1.28E+07 1-133 3.52E+06 4.36E+06 1.65E+06 8.09E+08 7.26E+06 0.OOE+00 1.76E+06 CS-134 1.60E+10 2.63E+ 10 5.55E+09 0.OOE+00 8.16E+09 2.93E+09 1.42E+08 CS-136 8.18E+07 2.25E+08 1.46E+08 0.OOE+00 1.20E+08 1.79E+07 7.90E+06 CS-137 2.39E+10 2.29E+10 3.38E+09 0.00E+00 7.46E+09 2.68E+09 1.43E+08 BA-140 2.77E+08 2.43E+05 1.62E+07 0.OOE+00 7.90E+04 1.45E+05 1.40E+08 CE-141 6.55E+05 3.27E+05 4.85E+04 0.OOE+00 1.43E+05 0.0OE+00 4.08E+08 CE-144 1.27E+08 3.98E+07 6.78E+06 0.OOE+00 2.21E+07 0.OOE+00 1.04E+10 ND-147 7.27E+04 5.89E+04 4.56E+03 0.0OE+00 3.23E+04 0.OOE+00 9.33E+07 1 Dose factors are based upon NUREG 0133 methodology.

2 For Tritium the units of the dose parameters are mrem/yr per ptCi/m 3 for all pathways, and they must be multiplied by X/Q.

v20 CAP A-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 62 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.6.6 GRW Dose Parameters' for Radioiodines, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW), Teen Age Group, Vegetation Pathway Organ "0" (mrem/yr per pCi/(sec in 2)) Rivegi Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI H-3 2 0.OOE+00 1.47E+03 1.47E+03 1.47E+03 1.47E+03 1.47E+03 1.47E+03 CR-51 O.OOE+00 O.OOE+00 6.16E+04 3.42E+04 1.35E+04 8.79E+04 1.03E+07 MN-54 O.OOE+00 4.54E+08 9.01E+07 O.OOE+00 1.36E+08 O.OOE+00 9.32E+08 FE-59 1.79E+08 4.18E+08 1.61E+08 O.OOE+00 O.OOE+00 1.32E+08 9.89E+08 CO-58 O.OOE+00 4.37E+07 1.01E+08 O.OOE+00 0.OOE+00 0.OOE+00 6.02E+08 CO-60 O.OOE+00 2.49E+08 5.60E+08 0.OOE+00 0.OOE+00 O.OOE+00 3.24E+09 ZN-65 4.24E+08 1.47E+09 6.86E+08 O.OOE+00 9.41E+08 0.OOE+00 6.23E+08 RB-86 0.OOE+00 2.75E+08 1.29E+08 0.OOE+00 0.OOE+00 O.OOE+00 4.06E+07 SR-89 1.51E+10 0.OOE+00 4.33E+08 O.OOE+00 O.OOE+00 O.OOE+00 1.80E+09 Y-90 1.24E+04 O.OOE+00 3.35E+02 O.OOE+00 O.OOE+00 O.OOE+00 1.02E+08 SR-90 9.22E+11 0.OOE+00 1.84E+10 0.OOE+00 O.OOE+00 O.OOE+00 2.11 E+10 ZR-95 1.72E+06 5.44E+05 3.74E+05 0.OOE+00 7.99E+05 O.OOE+00 1.26E+09 NB-95 1.93E+05 1.07E+05 5.90E+04 0.OOE+00 1.04E+05 O.OOE+00 4.58E+08 RU-103 6.82E+06 O.OOE+00 2.91E+06 0.OOE+00 2.40E+07 O.OOE+00 5.69E+08 RU-106 3.09E+08 O.OOE+00 3.90E+07 O.OOE+00 5.97E+08 O.OOE+00 1.48E+10 AG-1 10M 1.52E+07 1.44E+07 8.73E+06 0.OOE+00 2.74E+07 0.OOE+00 4.03E+09 SB-124 1.55E+08 2.85E+06 6.03E+07 3.51E+05 O.OOE+00 1.35E+08 3.11E+09 SB-125 2.14E+08 2.34E+06 5.01E+07 2.05E+05 O.OOE+00 1.88E+08 1.67E+09 TE-129M 3.61E+08 1.34E+08 5.72E+07 1.17E+08 1.51E+09 O.OOE+00 1.36E+09 1-131 7.68E+07 1.08E+08 5.78E+07 3.14E+10 1.85E+08 0.OOE+00 2.13E+07 1-133 1.93E+06 3.28E+06 1.00E+06 4.58E+08 5.75E+06 O.OOE+00 2.48E+06 CS-134 7.1OE+09 1.67E+10 7.75E+09 0.OOE+00 5.31E+09 2.03E+09 2.08E+08 CS-136 4.35E+07 1.71E+08 1.15E+08 O.OOE+00 9.31E+07 1.47E+07 1.38E+07 CS-137 1.01E+10 1.35E+10 4.69E+09 O.OOE+00 4.59E+09 1.78E+09 1.92E+08 BA-140 1.38E+08 1.69E+05 8.90E+06 0.OOE+00 5.74E+04 1.14E+05 2.13E+08 CE-141 2.83E+05 1.89E+05 2.17E+04 O.OOE+00 8.89E+04 O.OOE+00 5.40E+08 CE-144 5.27E+07 2.18E+07 2.83E+06 O.OOE+00 1.30E+07 O.OOE+00 1.33E+10 ND-147 3.67E+04 4.OOE+04 2.39E+03 O.OOE+00 2.35E+04 O.OOE+00 1.44E+08

' Dose factors are based upon NUREG 0133 methodology.

3 2 For Tritium the units of the dose parameters are mrem/yr per flCi/m for all pathways, and they must be multiplied by X/Q.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 34 PAGE 63 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Appendix: 10.6.7 GRW Dose Parameters' for Radioiodines, Radioactive Particulates, and any Radionuclide Other Than Noble Gas (IPT), Gaseous Effluents (GRW), Adult Age Group, Vegetation Pathway Organ "o" (mrem/yr per [tCi/(sec M2)) Rjyegj Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI H-3 2 0.OOE+00 1.29E+03 1.29E+03 1.29E+03 1.29E+03 1.29E+03 1.29E+03 CR-51 0.OOE+00 O.OOE+00 4.64E+04 2.77E+04 1.02E+04 6.15E+04 1.17E+07 MN-54 0.OOE+00 3.13E+08 5.97E+07 O.OOE+00 9.31E+07 0.OOE+00 9.58E+08 FE-59 1.26E+08 2.96E+08 1.13E+08 0.OOE+00 O.OOE+00 8.27E+07 9.87E+08 CO-58 O.OOE+00 3.08E+07 6.90E+07 O.OOE+00 0.OOE+00 0.OOE+00 6.24E+08 CO-60 O.OOE+00 1.67E+08 3.69E+08 O.OOE+00 0.OOE+00 0.OOE+00 3.14E+09 ZN-65 3.17E+08 1.01E+09 4.56E+08 0.OOE+00 6.75E+08 0.OOE+00 6.36E+08 RB-86 0.OOE+00 2.20E+08 1.03E+08 0.OOE+00 0.OOE+00 0.OOE+00 4.34E+07 SR-89 9.95E+09 0.OOE+00 2.86E+08 O.OOE+00 O.OOE+00 0.OOE+00 1.60E+09 Y-90 1.33E+04 O.OOE+00 3.57E+02 O.OOE+00 0.OOE+00 0.OOE+00 1.41E+08 SR-90 6.95E+11 0.OOE+00 1.40E+ 10 O.OOE+00 O.OOE+00 0.OOE+00 1.75E+10 ZR-95 1.18E+06 3.77E+05 2.55E+05 O.OOE+00 5.92E+05 0.OOE+00 1.20E+09 NB-95 1.43E+05 7.95E+04 4.27E+04 O.OOE+00 7.86E+04 0.OOE+00 4.83E+08 RU-103 4.77E+06 O.OOE+00 2.05E+06 0.OOE+00 1.82E+07 O.OOE+00 5.57E+08 RU-106 1.93E+08 O.OOE+00 2.44E+07 0.OOE+00 3.72E+08 0.OOE+00 1.25E+10 AG-1 10M 1.05E+07 .9.75E+06 5.79E+06 0.OOE+00 1.92E+07 0.OOE+00 3.98E+09 S13-124 1.04E+08 1.96E+06 4.11E+07 2.52E+05 O.OOE+00 8.08E+07 2.95E+09 S13-125 1.37E+08 1.53E+06 3.25E+07 1.39E+05 O.OOE+00 1.05E+08 1.50E+09 TE-129M 2.51E+08 9.37E+07 3.97E+07 8.62E+07 1.05E+09 O.OOE+00 1.26E+09 1-131 8.07E+07 1.15E+08 6.62E+07 3.78E+10 1.98E+08 0.OOE+00 3.05E+07 1-133 2.08E+06 3.62E+06 1.1OE+06 5.32E+08 6.31E+06 O.OOE+00 3.25E+06 CS-134 4.67E+09 1.11E+10 9.08E+09 0.OOE+00 3.59E+09 1.19E+09 1.94E+08 CS-136 4.25E+07 1.68E+08 1.21E+08 0.OOE+00 9.33E+07 1.28E+07 1.90E+07 CS-137 6.36E+09 8.70E+09 5.70E+09 O.OOE+00 2.95E+09 9.81E+08 1.68E+08 BA-140 1.29E+08 1.61E+05 8.42E+06 O.OOE+00 5.49E+04 9.24E+04 2.65E+08 CE-141 1.97E+05 1.33E+05 1.51E+04 O.OOE+00 6.19E+04 O.OOE+00 5.09E+08 CE-144 3.29E+07 1.38E+07 1.77E+06 O.OOE+00 8.16E+06 0.OOE+00 1.11 E+10 ND-147 3.37E+04 3.90E+04 2.33E+03 O.OOE+00 2.28E+04 O.OOE+00 1.87E+08

' Dose factors are based upon NUREG 0133 methodology.

3 2 For Tritium the units of the dose parameters are mrem/yr per PCi/m for all pathways, and they must be multiplied by X/Q.

v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

10/04/00 Page 1 of 1 DIABLO CANYON POWER PLANT TITLE:

CAP A-8 ATTACHMENT 11.1 Liquid Discharges (LRW) Monitored for Radioactivity 1 2 AND LIQUID RADWASTE SYSTEM CONDENSATE DEMINERALIZER STEAM GENERATOR BLOWDOWN nnRAQqnq v20_CAPA-8u3r34.DOC 08 0901.1100

- - UNCONTROLLED PROCEDURE- DO NOT USE TO PERFORM WORK or ISSUE FOR USE ***

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Attachment 5 PG&E Letter DCL-1 1-049 Attachment 5 Diablo Canyon Power Plant Chemical Analysis Procedure, CAP A-8, "Offsite Dose Calculations," Revision 35

- - ISSUED FOR USE BY:_ DA TE: EXPIRES:____

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 NUCLEAR POWER GENERATION REVISION 35 DIABLO CANYON POWER PLANT PAGE 1 OF 63 CHEMICAL ANALYSIS PROCEDURE UNITS TITLE: Off-Site D ose Calculations 1 2 10/22/10 EFFECTIVE DATE PROCEDURE CLASSIFICATION: QUALITY RELATED LEVEL OF USE: REFERENCE TABLE OF CONTENTS SECTION PAGE SCOPE .............................................................................................................................................................. 2 D ISCU SSION .................................................................................................................................................... 3 RESPON SIBILITIES ........................................................................................................................................ 3 PREREQU ISITES ............................................................................................................................................. 3 PRECA U TION S ................................................................................................................................................ 4 IN STRU CTION S............................................................................................................................................... 4 Liquid Effluents ............................................................................................................................................. 4 Liquid Effl uents - D ose Calculation ...................................................................................................... 4 10 CFR 20, Appendix B, Table 2, Column 2, Effluent (liquid) Concentration Limit (ECL) Calculation.6 Liquid Effl uent Radiation M onitor Set Point M ethodology ................................................................. 7 D ose Projection (for Liquid Effl uents) ............................................................................................... 13 Liquid Lim iting Flow Rate M ethodology - ECL Based ...................................................................... 15 Liquid Lim iting Flow Rates - LLD Based ........................................................................................... 16 G aseous Effluents ........................................................................................................................................ 18 Plant V ent Noble G as M onitor - RE-14 HA SP ............................................................................... 21 Containm ent Purge - RE-44 HA SP .................................................................................................. 28 Dose To Critical Receptor Due To Radioiodines, Tritium and Particulates Released in Gaseous Effl uents ................................................................................................................................................... 34 40 CFR 190 D ose Calculations .................................................................................................................... 42 A CCEPTAN CE CRITERIA ............................................................................................................................ 48 REFEREN CES ................................................................................................................................................ 48 RE CORD S ....................................................................................................................................................... 49 APPEND ICES ................................................................................................................................................. 49 A TTA CHM ENT S ............................................................................................................................................ 49 Table 6.1- Typical Liquid Effluent Discharge Pathway Allocation Factors 8 Table 6.2- Typical Gaseous Effluent Discharge Pathway Allocation Factors 20 Table 6.3- Expected On-Site Distance and Visitation Times for Members of the Public 46 CAPA-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 2 OF 63 TITLE: Off-Site Dose Calculations UNITS 1AND2

1. SCOPE This procedure describes the methodology for the following:

Effluent RECP or RECP or Tech Spec Surveillance Requirement Implements Type Tech Spec Liquids 6.1.1.1 (RECP) Determination of alarm/trip setpoints for RE-18, 10 CFR 20 App. B 6.1.3.1 (RECP) RE-23, and RE-3 Table 2, Col. 2 Gases 6.1.2.1 (RECP) Determination of alarm/trip setpoints for RE-22, NUREG 0133 6.1.6.1 (RECP) RE-14, and RE-14R Liquids 6.1.3.2 (RECP) Prerelease analyses of effluents 10 CFR 20 App. B 6.1.3.3 (RECP) Table 2, Col. 2 Post release analysis of effluents Liquids 6.1.4.2 (RECP) Dose calculations 10 CFR 50 App. I Liquids 6.1.5.2 (RECP) Dose projections 10 CFR 50 App. I Gases 6.1.6.2 Dose Rate calculations, Noble Gases, Total Body and NUREG 0133 Skin Gases 6.1.6.3 Dose Rate calculations, lodines, Particulates and NUREG 0133 Radionuclides other than Noble Gases, per organ, per age group Gases 6.1.7.2 (RECP) Noble Gas Air Dose Calculations 10 CFR 50 App. I Gases 6.1.8.2 (RECP) lodines, Particulates, and Radionuclides other than 10 CFR 50 App. I Noble Gases Organ Dose Calculations per age group Gases 6.1.9.2 (RECP) Noble Gases, Iodines, Particulates, and Radionuclides 10 CFR 50 App. I other than Noble Gases, Dose Projection Liquids 6.1.10.2 (RECP) Cumulative Dose from: Liquids, Noble Gases, 40 CFR 190 and 4.4.2.b. 1 lodines, Particulates, and Radionuclides other than Gases (RECP) Noble Gases per age group, per organ Direct 6.1.10.3 (RECP) Direct Radiation Dose Rate and Dose Calculations to 40 CFR 190 Radiation unrestricted areas due to plant and high radwaste storage sky-shine The calculational methodology for doses are based on models and data that make it unlikely to substantially underestimate the actual exposure of an individual through any of the appropriate pathways. Appendixes containing the values for the various parameters used in these expressions are also included.

CAPA-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 3 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

2. DISCUSSION 2.1 This procedure is used in support of the Radiological Monitoring and Controls Program (RMCP), and Radioactive Effluent Controls Program (RECP), and the portion that deals with routine radioactive liquid and gaseous releases to the unrestricted area. Limits are based on the dose commitment to a member of the general public related to the release of radionuclides through either direct or indirect exposure (e.g., submersion in a cloud of radioactive Noble Gases, radionuclides deposited on the ground, direct radiation from radionuclides stored on-site, inhalation of radionuclides or ingestion of radionuclides via a food pathway such as milk, meat, vegetable or fish, etc.).

2.2 The conduct of the Environmental Radiological Monitoring Procedure (ERMP) is found in RPL.IDI1.

2.3 Changes to CAP A-8 shall be processed in accordance with the requirements of DCPP Technical Specification Section 5.5.1.

3. RESPONSIBILITIES 3.1 The manager, chemistry is the overseeing authority of responsibility for ensuring that the off-site dose calculational procedure (ODCP) meets all RECP and Tech Spec requirements with regards to calculated doses delivered by the plant to the unrestricted area surrounding the site.

3.2 The senior radiochemistry engineer assumes the overall responsibility for ensuring that this procedure's program is followed and implemented where appropriate, especially in regards to RECP or Tech Spec requirements.

3.3 The radiochemistry effluents engineer has the responsibility of correct and timely implementation of all the procedure's calculational methodology, where appropriate, for each radioactive effluent ieleased. Furthermore this engineer is responsible for:

reviewing the results; cross (spot) checking the calculations; and maintaining an updated archive of post release calculated doses for annual report purposes.

3.4 The digital systems group assures that any supporting computer software is maintained current and compatible with the procedure's calculational methodology and that the computer hardware is maintained operable at all times.

3.5 The radiochemistry staff engineer provides an oversight of the effluents program's ODCP to: confirm compliance with RECP or Tech Specs; provide technical support; recommend or design improvements to the dose calculational methodology and the effluent program control; and investigate long-term planning toward effluent related activities and their associated dose calculations.

3.6 Responsibilities as described in CYl, " Chemistry and Radiochemistry," and CY1.DC I, "Analytical Data Processing Responsibilities," apply.

4. PREREQUISITES None CAPA-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 4 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

5. PRECAUTIONS None

6. INSTRUCTIONS 6.1 Liquid Effluents 6.1.1 Liquid Effluents - Dose Calculation The dose contributions to the total body and each individual organ (bone, liver, thyroid, kidney, lung and GI-LLI) of the maximum exposed individual (adult) due to consumption of saltwater fish and saltwater invertebrate is calculated for all radionuclides identified in liquid effluents released to unrestricted areas using the following expression:

D,, = FtAtjAioCi e-"l (1)

Where:

Do, The dose commitment to organ, o, in mrem.

Near field average dilution factor during the period of the release. It is defined as:

t= Waste Flow (2)

Dilution Flow x Z Where:

Z Z is the site specific factor for the mixing effect of the discharge structure. Specifically, it is the credit taken for dilution which occurs between the discharge structure and the body of water which contaminates fish or invertebrates in the liquid ingestion pathway. For DCPP Z = 5.

At = The time period for the release in hours.

Ai= The site specific ingestion dose commitment factor to organ, o, due to radionuclide, i, in mrem/hr per pCi/ml as defined by Equation 3.

Ci Concentration of radionuclide, i, in the undiluted liquid effluent, in jiCi/ml.

X = Decay constant of radionuclide, i.

tm = Time interval between end of sampling and midpoint of release.

CAPA-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 5 OF 63 TITLE: Off-Site Dose Calculations UNITS I AND 2 The site specific ingestion dose commitment factor, Ajo, is defined as:

Aio =ko(UFBFI + U1 BI)DF 1 (3)

Where:

k. = Units conversion factor of 1.14E + 05 in units of pCi/pCi x ml/I x yr/hr.

UF = Saltwater fish consumption rate in kilograms of fish per year.

DCPP value for this parameter is 21 kg/yr and is taken from NUREG 0133, Section 4.3.1.

BFj = Saltwater bioaccumulation factor for nuclide, i, in fish flesh in units of pCi/Kg per pCi/i. Values for BFi are taken from Table A-1 of Reg. Guide 1.109, except uranium and plutonium, which were taken from NUREG/CR-4013.

UI Saltwater invertebrate consumption rate in kilograms per years. DCPP value for this parameter is 5 kg/yr and is taken from NUREG 0133, Section 4.3.1.

BEi = Saltwater bioaccumulation factor for nuclide, i, in invertebrate flesh in units of pCi/Kg per pCi/l. Values for BEi are taken from Table A-1 of Reg. Guide 1.109, except uranium and plutonium, which were taken from NUREG-4013.

DFi = Adult ingestion dose conversion factor for nuclide, i, in mrem per pCi ingested, from Table E-1 1 of Regulatory Guide 1.109, with exceptions detailed below.

DFi exceptions: H-3, Br-82, Sb-124, Sb-125, Pu-238, Pu-239, Pu-240, Pu-241 and Pu-242, ingestion dose conversion factors are taken from EMP-155.

As-76, Sn- 113, Sn- 117m and Sb-122 ingestion dose conversion factors were calculated by ORNL using ICRP-2 methodology.

U-233, U-234, U-235, U-236 and U-238 ingestion dose conversion factors are taken from NUREG-0 172.

The site specific values for Ai, are listed in Appendix 10.1. When necessary, these factors were corrected for the ingrowth of daughter radionuclides following ingestion of the parent. All radionuclides treated in this manner are followed by a "+D." Reference NUREG-0172, "Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake," and A0619601.

Units 1 and 2 share a common liquid radwaste (LRW) treatment system. The effluent doses due to releases discharged via the common LRW are apportioned between the units with 50% credited to Unit 1 and 50% credited to Unit 2.

CAPA-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 6 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.1.2 10 CFR 20, Appendix B, Table 2, Column 2, Effluent (liquid) Concentration Limit (ECL) Calculation

a. The ECL for the identified mixture of radionuclides in the "jh" batch of liquids is calculated as follows:

ny'~

ECLj n _C (4) j ECL..

Where:

ECLj = The unrestricted area total undiluted ECL for the "jth",

particular mixture of identified radionuclides, in jiCi/ml.

Cij = The concentration of radionuclide "i," in pCi/ml for the

,th, mixture.

ECLij = The ECL in unrestricted area water for radionuclide "i,"

in general, in tCi/ml (from 10 CFR 20, Appendix B, Table 2, Column 2).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 7 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

b. The overall ECL for simultaneous discharges is given by Equation 5.

n ECLoveratl j n (5)

I Where:

ECLover 11 = The unrestricted area ECL for the current radionuclide mixture for concurrent "j"discharges (in pCi/ml).

Cj = The total activity concentration for the "jth, individual stream in pCi/ml.

ECLj = The total ECL for the "jth" individual mixture (or stream) determined as defined in Equation 4, in pCi/ml.

(Dj = The ratio of an individual discharge "jth, pathway flowrate to the sum total of all individual undiluted pathway flowrates as defined by:

fi *j- *-, fj(6)

Where:

f = Undiluted effluent flowrate for pathway, "j".

6.1.3 Liquid Effluent Radiation Monitor Set Point Methodology

a. Introduction The DCPP radiological effluent controls program requires that the liquid effluent monitors be operable with their alarm/trip set points set to ensure that the effluent concentration limits of 10 CFR 20 are not exceeded.

The alarm/trip set point for the liquid effluent radiation monitors is derived from the concentration limit set forth in Appendix B, Table 2, Column 2 of 10 CFR 20.1001-2404.

The alarm/trip set points are applied at the unrestricted area boundary.

The set points take into account appropriate factors for dilution, dispersion, or decay of radioactive materials that may occur between the point of discharge and the unrestricted area boundary.

CAP A-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 8 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

b. Allocation and Safety Factors The limits of RECP 6.1.3.1 are site limits which require that the set point methodology must ensure simultaneous releases do not exceed the liquid effluent concentration limits of 10 CFR 20 in the unrestricted area. The DCPP High Alarm Set Point (HASP) methodology makes use of an Allocation Factor (AF) to limit the effluent concentrations from simultaneous liquid discharges. The Allocation Factors can be adjusted based upon operational requirements with the restriction that the sum of the Allocation Factors must be less than or equal to 1.

Typical Allocation Factors are shown.

Table 6.1 Typical Liquid Effluent Discharge Pathway Allocation Factors Discharge Pathway Rad Monitor Allocation Factor (AF)

Oily Water Separator RE-3 0.01 Liquid Radwaste System RE-18 0.90 Steam Generator Blow Down (Unit 1) RE-23 (Ul) 0.04 Steam Generator Blow Down (Unit 2) RE-23 (U2) 0.04 Miscellaneous none 0.01 An additional level of conservatism in the HASP methodology is implemented by the use of a Safety Factor (SF). The Safety Factor is defined as 0.9 and provides for a High Alarm Set Point at 90% of the 10 CFR 20 concentration limits.

c. Tritium Correction Factor As result of an aggressive liquid radwaste treatment program, the liquid effluents at DCPP typically contain very low levels of gamma emitters.

In order to reduce the over all volume of liquid waste disc'arged, DCPP also recycles waste water. This recycling results in higher tritium concentration in liquid effluents when compared with the low gamma emitter concentrations. As a result, standard HASP methodology results in very low set points. In some cases the calculated set points are barely above the monitor background.

The liquid HASP methodology used by DCPP uses a Tritium Correction Factor (TCF) which assumes a constant, but conservative tritium concentration in the liquid effluent. This results in an operationally reasonable set point while ensuring that the liquid effluent concentrations released to the unrestricted areas do not exceed the limits of 10 CFR 20.

CAP A-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 9 OF. 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 The Tritium Correction Factor is defined as shown in Equation 7.

TCF I- CH3 /ECLH3 (7)

Where:

ECLH3 = effluent concentration limit of tritium.

CH3 = concentration of tritium in the release mix, pre-dilution (pCi/ml).

F = conservative dilution flow rate (gpm).

f = conservative undiluted effluent flow rate (gpm).

The concentration of tritium, CH3, is conservatively estimated.

d. Liquid Effluent Radiation Monitor Set Point Calculations The High Alarm Set Point (HASP) are calculated to ensure that the liquid effluent concentration limits of 10 CFR 20 are not exceeded. The set points represent the maximum operational set point. The actual set point used by operations will be equal to or less than the actual value as determined by the HASP methodology described in this section.

1. Set Point Methodology for RE-3 HASP: Oily Water Separator Under normal conditions, the Oily Water Separator stream does not contain any radioactive material. Only in the event that there is primary to secondary leakage does this become a potential liquid effluent discharge point. In order to insure that no unplanned or unmonitored releases take place by way of the Oily Water Separator, RE-3 serves to monitor the discharge even when no activity has been identified in the effluent. When no significant primary to secondary leakage is taking place or when no activity has been identified in the Oily Water Separator, the High Alarm Set Point for RE-3 is calculated as shown in Equation 8.

HASPs_3 = 3 x BKGD _3 (8)

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 10 OF 63 TITLE: Off-Site Dose Calculations UNITS I AND 2 In the event that primary to secondary leakage results in activity being detected in the Oily Water Separator, Equation 9 will be used to calculate a High Alarm Set Point value. The greater HASP value as determined by Equation 8 or Equation 9 will be used.

HASPs 3 BKGD _ = _3 + (AFXSF)x-kC *~ ), y. F/f

-, "--"- xTCF xTC F (9)

Where:

HASPU- 3 = high alarm setpoint for RE-3 (cpm).

BKGDRE_3 = background reading for RE-3 (cpm).

(AF) = allocation factor for the oily water separator effluent system from Table 6.1.

(SF) = safety factor for RE-3 (0.9).

ky = monitor response factor (cpm/gCi/ml).

C= concentration of gamma emitting isotopes in the release mix, pre-dilution (jiCi/ml).

F = dilution flow rate (gpm).

f = undiluted effluent flow rate (gpm).

Ci = concentration of isotope "i," in the release mix, pre-dilution (ptCi/ml).

ECLi = effluent concentration limit of isotope "i".

TCF = tritium correction factor as defined by Equation 7.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 11 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

2. Set Point Methodology for RE-18 HASP: Liquid Radwaste System.

The High Alarm Set Point for the RE-1 8 Liquid Radwaste System liquid effluent radiation monitor is calculated as shown in Equation 10.

HASPRE 18 BKGDRE18 + (AFXSF)x kC F/f x TCF (10o)

Where:

HASPR- 18 = high alarm setpoint for RE-i18 (cpm).

BKGDR_ 18 = background reading for RE- 8 (cpm).

(AF) = allocation factor for the liquid radwaste effluent system from Table 6.1.

(SF) = safety factor for RE-18 (0.9).

ky = monitor response factor (cpn/gCi/ml).

CY = concentration of gamma emitting isotopes in the release mix, pre-dilution (p.Ci/ml).

F = dilution flow rate (gpm).

f = undiluted effluent flow rate (gpm).

Ci = concentration of isotope "i," in the release mix, pre-dilution (ýtCi/ml).

ECLi = effluent concentration limit of isotope 1T1.

TCF = tritium correction factor as defined by Equation 7.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 12 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

3. Set Point Methodology for RE-23 HASP: Steam Generator Blowdown Tank.

The High Alarm Set Point for the RE-23, Steam Generator Blowdown Tank liquid effluent radiation monitor, is calculated as shown in Equation 11.

HASPE-2 3 BKGDRE- 23 +(AFXSF)x ECi TCkC( 11) y cC1 ECUj Where:

HASPRE- 2 3 = high alarm setpoint for RE-23 (cpm).

BKGDRE 23 = background reading for RE-23 (cpm).

(AF) = allocation factor for the steam generator blowdown effluent system for each unit from Table 6.1.

(SF) = safety factor for RE-23 (0.9).

k7 = monitor response factor (cpmrdtCi/ml).

C= concentration of gamma emitting isotopes in the release mix, pre-dilution (p.Ci/ml).

F = dilution flow rate (gpm).

f = undiluted effluent flow rate (gpm).

Ci = concentration of isotope "i," in the release mix, pre-dilution (itCi/ml).

ECLj = effluent concentration limit of isotope "i".

TCF = tritium correction factor as defined by Equation 7.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 13 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.1.4 Dose Projection (for Liquid Effluents)

The projected dose contributions from each reactor unit due to liquid effluents for the current calendar month, quarter and current calendar year must be determined in accordance with the methodology and parameters in the ODCP at least once per 31 days.

The purpose of this is to determine if appropriate treatment of liquid radioactive materials in relation to maintaining releases "as low as reasonably achievable," is necessary.

The projected dose from each reactor unit is given by:

DP =DPU +TDp Com (12)

Where:

DP = Projected Dose.

DR = Projected dose attributed to reactor unit, U.

Dp,com = Projected dose common to both reactor units.

The 31-day projected dose is calculated by Equation 13.

DM 3l-Dpm +d~cM + dc D! =3Ix. (13)

(T+t)

Where:

D = Monthly Projected Dose.

D = Previous Month's Actual Dose.

dAM = Current Month Actual Dose to date.

dCB = Projected Dose from Current Batch Release.

T = Number of days in the previous month.

t = Number of days into the present month.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 14 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Projected quarterly doses are determined by Equation 14.

DCQ=dcQ +(92-t) +d +dB (14)

P A (T + t)(

Where:

DCQ Projected dose for the current calendar quarter.

dc° = Current quarter to date actual dose.

Do = Previous quarter's actual dose.

dcB = Projected dose as a result of the current batch release.

T = Number of days in the previous quarter.

t = Number of days into the present quarter.

Projected yearly doses are determined by Equation 15.

Dcy(dcy D~r~dr +(66_tDpr

+( AdA+ dac + dcBB (15)

P -A +36) (T +t)

Where:

Dcy = Projected dose for the current calendar year.

dc = Current year to date actual dose.

Dr = Previous year's actual dose.

dCB = Projected dose as a result of the current batch release.

T = Number of days in the previous year.

t = Number of days into the present year.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 15 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.1.5 Liquid Limiting Flow Rate Methodology - ECL Based The maximum effluent flow rate through monitors RE-3, RE-18, and RE-23 as well as for releases from the Condensate Demineralizer Regenerate waste tank or miscellaneous release points is established in order to provide further control over the effluent releases. The release rate limit is determined by the effluent concentration and the 10 CFR 20 Effluent Concentration Limits (ECLs) as shown in Equation 16.

f F(AFXSFXTCF) (16) i H-3 ECLi Where:

f = Maximum operational undiluted liquid radwaste effluent discharge flow rate (gpm).

F = Expected dilution flow rate (gpm).

AF = allocation factor for the liquid radwaste effluent source from Table 6.1.

SF safety factor (0.9).

TCF = tritium correction factor as defined by Equation 7.

Ci concentration of isotopes "i" in the release mix, pre-dilution (ViCi/ml).

ECLi = effluent concentration limit of isotope "i" (laCi/ml).

When the term - then the Limiting Flow Rate is calculated by:

=0 i#H-3 ECLj f = F(AFXSFXTCF) (17)

Where the terms are as previously defined.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 16 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.1.6 Liquid Limiting Flow Rates - LLD Based When there is no primary to secondary leakage, the Oily Water Separator and various miscellaneous release points are assumed to be uncontaminated.

Furthermore, in order to establish practical operational flow rate limits for any sources when they are considered uncontaminated, Equation 18 is used. While no activity may be present, Equation 18 assumes a concentration equal to the Lower Limit of Detection for the nuclides listed in CY2.ID1, Attachment 3, Table 5.

f- F(AFXSF) 4.3 (18)

Where:

f Maximum operational undiluted liquid radwaste effluent discharge flow rate (gpm).

F Expected dilution flow rate (gpm).

AF = allocation factor for the liquid radwaste effluent source from Table 6.1.

SF = safety factor (0.9).

4.3 = Total ECL fraction as given by:

~LLD, ECL, Where:

LLDi = Lower limit of detection for isotope "i" from CY2.ID1, Appendix 6.1, Table 6.1.3-1. (ýtCi/ml).

ECLi = effluent concentration limit of isotope "i" (ýtCi/ml).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 17 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.1.7 Unplanned Liquid Releases (Abnormal Releases)

An unplanned release is an unexpected and potentially unmonitored release to the environment due to operational error or equipment malfunctions.

a. Unmonitored unplanned releases shall have a report written by the Radiochemistry Effluents Engineer describing the event with a calculation, if possible, of the percent of Tech Spec release rate limit.

This will then be forwarded to PSRC for review. Describe these unplanned releases in the Annual Radioactive Effluent Release Report.

b. Monitored unplanned releases which exceed 1% of the RECP release rate limit will also have a report written describing the event and must be forwarded to the PSRC for review. Describe these unplanned releases in the Annual Radioactive Effluent Release Report.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 18 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2 Gaseous Effluents The only significant path for gaseous radioactive releases to the environment during normal operations is via the plant vent. This source is used for calculating dose rates and real-time doses to the unrestricted area due to noble gases, vaporous radioiodines and airborne radio-particulates. The plant vent also has redundant monitoring for these types of gaseous releases.

Other paths such as the steam generator blowdown tank vent, the chemistry lab fume hood, the main condenser Nash vacuum pump discharge, hot machine shop vent, etc., are considered miscellaneous release sources. These miscellaneous release sources are not continuously monitored but can have dose rates and dose calculated for their path to the unrestricted area.

6.2.1 Meteorological Methodology The equations for determining gaseous 'effluent concentration limits, high alarm setpoints, dose rates, and critical receptor doses make use of the historical average atmospheric conditions in accordance with methodologies of Regulatory Guides 1.109 and 1.111 and NUREGs 0133 and 0472. The historical average dispersion (X/Q) and deposition (D/Q) values are derived from the methodology of Regulatory Guide 1.111 as implemented by NUREG 2919 (computer code XOQDOQ). The DCPP dispersion and deposition values are based on the latest five years of meteorological data and are updated when the value of X/Q or D/Q changes by more than ten percent.

The present values are listed in Appendix 10.2.

Long-term releases are characterized as those that are generally continuous and stable in release rate, such as normal ventilation systems effluents. Doses due to long-term releases are modeled using historical annual average dispersion and deposition values in accordance with the guidance of Regulatory Guide 1.109, Regulatory Guide 1.111, NUREG 0133 and NUREG 0472.

Short-term releases are defined as those which occur for a total of 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months or less in a calendar year but not more than 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months in any quarter. In accordance with NUJREG 0133 and based upon an operational history that has demonstrated short term gaseous releases can be characterized as random in both time of day and duration, historical average atmospheric dispersion and deposition values are used to model doses due to short-term releases.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 19 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2.2 Gas Effluent Concentration Limits

a. Philosophy of Concentration Limits The radiological effluent controls restrict at all times the dose rate due to radioactive materials released in gaseous effluents from the site to areas at or beyond the site boundary for noble gases to less than or equal to 500 mrem/yr to the total body and 3000 mrem/yr to the skin. For iodine- 131, iodine- 133, tritium and for all radionuclides in particulate form with half-lives greater than 8 days, the dose rate is limited to less than or equal to 1500 mrem/yr to any organ.

These dose rate limits act to restrict at all times the instantaneous concentrations of radionuclides in gaseous effluents at the site boundary.

1. Allocation and Safety Factors The limits set forth by RECP 6.1.6.1 are site limits which require that the set point methodology must ensure simultaneous releases do not exceed the off-site dose rate limits set forth by RECP 6.1.6.1(a) and 6.1.6.1(b). The DCPP High Alarm Set Point methodology makes use of an Allocation Factor (AF) to limit the noble gas effluent dose rate from simultaneous atmospheric releases.

The Allocation Factors can be adjusted based upon operational requirements with the following restrictions:

The sum of the Allocation Factors for RE-14 (plant vent noble gas monitor), the SGBD tank vents, and miscellaneous release points from both units must be less than or equal to 1.

The Allocation Factors for RE-22 (Waste Gas Decay Tanks) and RE-44 (Containment Purge) can also be adjusted based upon operational requirements with restriction that the sum of the Allocation Factors for RE-22 and RE-44 must be less than or equal to the Allocation Factor for RE-14.

The Allocation Factors for RE-24 (Plant Vent Iodine Monitor) and RE-28 (Plant Vent Particulate Monitor) are set equal to the Allocation Factor for RE-14.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 20 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Typical Allocation Factors are shown:

Table 6.2 Typical Gaseous Effluent Discharge Pathway Allocation Factors Discharge Pathway Rad Monitor Allocation Factor (AF)

Plant Vent - NG Monitor RE-14 0.48 Plant Vent Iodine Monitor RE-24 0.48 Plant Vent Part Monitor RE-28 0.48 Waste Gas Decay Tank RE-22 0.10 Containment Purge RE-44 0.38 SGBD Tank Vent 0.01 Miscellaneous 0.01 An additional level of conservatism in the HASP methodology is implemented by the use of a Safety Factor (SF). The Safety Factor is defined as 0.9 and provides for a High Alarm Set Point at 90% of the dose rate limits.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 21 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

b. Gaseous Effluent Radiation Monitor Set Points

1. PLANT VENT NOBLE GAS MONITOR - RE-14 HASP The Plant Vent effluent stream is monitored by rad monitor RE-14.

RE- 14 provides alarm function only.

The High Alarm Set Point methodology for RE-14 is given by Equation 19, which is based upon the assumption that the total body dose rate limit is most limiting.

14M.CT(NG)= AFxSFx 500 (19) 472xF,, xý -Q)MLx294 Where:

4 1 MaXCT(NG) = the maximum allowable total concentration of noble gas in the plant vent, in ýiCi/cc.

AF = the allocation factor for the plant vent for one unit from Table 6.2.

SF = a safety factor to ensure that dose rate limits of the radiological effluent controls are not exceeded (0.9).

500 = (mrem/yr) the site Total Body dose rate limit for instantaneous releases.

472 = the conversion constant to cc/sec from cfm.

Fpv = total flow rate in the plant vent, in cfm (maximum plant vent flow rate is 263,000 cfm).

(z/-- )M = the maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological data base, from Appendix 10.2.

294 = the whole body dose factor (mrem/yr/[tCi/m 3) for Xe-1 33 as presented in Appendix 10.3, (for the plant vent HASP, the release is assumed to be all Xe- 133).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 22 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND2

2. PLANT VENT NOBLE GAS MONITOR - RE-14 SCALING In order to correlate the readings of RE- 14 to noble gas concentration during periods between samplings, the concentration is scaled according to Equation 20.

C = CPMT XC (20)

CPMS Where:

CPMT = RE-14 time weighted arithmetic mean (cpm).

CPMs = RE-14 gross count rate at the time of sampling (cpm).

Cs = Concentration of noble gas corresponding to CPMs, based upon noble gas grab sample (pCi/cc).

CT = Scaled concentration of noble gas (ptCi/cc).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 23 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

3. PLANT VENT IODINE MONITOR - RE-24 The Plant Vent Iodine concentration is monitored by rad monitor RE-24. RE-24 provides alarm function only. The alarm setpoint methodology is based upon the assumption that RE-24 responds only to I-131. The methodology also presumes a release mixture based upon the RCS source term.

The High Alarm Set Point methodology of RE-24 is given by Equation 21.

24 MAaCT (Iodine) = SF x AF x fI- 131 4721x F 0x(z/Q)M Pf (21) i Where:

24 MaxCT(iodine) = the maximum allowable concentration of 1-131 in the plant vent.

AF = The allocation factor for the plant vent for one unit from Table 6.2.

SF = A safety factor to insure that the dose rate limits of the radiological effluent controls are not exceeded (0.9).

fi-131 = fraction of the total non-noble gas concentration that is due to 1-131. Defined as:

C1_3 f,-,31 - .*-131 (22)

Ci CAP A-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 24 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 1500 = (mrem/yr) the site organ dose rate limit for Iodine-131, for Iodine-133, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days.

472 = the conversion constant to cc/sec from cfm.

Fpv = total flowrate in the plant vent, in cfm (maximum plant vent flowrate is 263,000 cfm).

(--Q-)Ma maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological database, from Appendix 10.2.

Piw Inhalation dose factor for nuclide "i" (mrem/yr/jiCi/m3) for child age group for worst case organ, from Appendix 10.4.

Dose factors are based upon NUREG 0133 methodology. Inhalation dose conversion factors are taken from Reg. Guide 1.109, Rev 1, Table E-9, with the following exceptions: H-3, Sb-124 and Sb-125 inhalation dose conversion factors taken from NUREG/CR-4013.

fi = fraction of total non-noble gas concentration (excluding tritium) that is due to nuclide, i, and defined as:

f Ci (23) zci CAPA-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 25 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

4. PLANT VENT PARTICULATE MONITOR - RE-28 The Plant Vent Particulate concentration is monitored by rad monitor RE-28. The alarm setpoint methodology is based upon the assumption of a 5% cross talk from the iodine channel. This is due to the retention of a small portion of iodine on the particulate filter.

A release mixture based upon the RCS source term is also assumed.

The High Alarm Set Point methodology for RE-28 is given by Equation 24.

28M*C articulates)= AF x SF x 0.05 x Ilodines Particulatesficates 1500 (24) 472 x Fp,, (z/Q)Mx ~fi Where:

28 MaxCT(Particulate) = Maximum allowable particulate concentration in the plant vent.

AF = The allocation factor for the plant vent for one unit from Table 6.2.

SF = A safety factor to insure that the dose rate limits of the radiological effluent controls are not exceeded (0.9).

0.05 = Fraction of total iodine activity retained on particulate filter.

flodines = Fraction of the total non-noble gas concentration that is due to iodines.

fParticulates = Fraction of the total non-noble gas concentration that is due to particulates.

1500 = (mrem/yr) the site organ dose rate limit for Iodine- 131, for Iodine- 133, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days.

472 = Conversion constant to cc/sec from cfm.

Fpv = Total flowrate in the plant vent, in cfm (maximum plant vent flowrate is 263,000 cfm).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 26 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

- Maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological database, from Appendix 10.2.

PiW Inhalation dose factor for nuclide "i" (mrem/yr/tCi/m3) for child age group for worst case organ, from Appendix 10.4. Dose factors are based upon NUREG 0133 methodology.

Inhalation dose conversion factors are taken from Reg. Guide 1.109, Rev 1, Table E-9, with the following exceptions: H-3, Sb-124 and Sb-125 inhalation dose conversion factors taken from NUREG/CR-4013.

= Fraction of total non-noble gas concentration (excluding tritium) that is due to nuclide, i, as defined by Equation 23.

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5. WASTE GAS DECAY TANK MONITOR - RE-22 HASP Effluent releases from the Waste Gas Decay Tank are monitored by rad monitor RE-22. RE-22 provides alarm and automatic release termination functions.

The High Alarm Set Point methodology for RE-22 is given by Equation 25, which is based upon the assumption that the skin dose rate limit is most limiting.

22 MaxC' (NG) = AF x SF 3000 X (25) 4[72xFgd xV-I)M xl .34 x 10 Wherre:

22 MaxCT(N(

3) = the maximum allowable total concentration of noble gas in the gas decay tank exhaust line, in pCi/cc.

AF = the allocation factor for the waste gas decay tank for one unit from Table 6.2.

SF = safety factor to ensure that dose rate limits of the radiological effluent controls are not exceeded (0.9).

3000 = (mrem/yr) the site Skin dose rate limit for instantaneous releases.

472 = the conversion constant to cc/sec from cfm.

Fgdt = maximum flow rate for the gas decay tank system (31 cfm).

1.4Q)L1 = the maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological data base from Appendix 10.2.

1.34x 103 = the skin dose factor for Kr-85 (mrem/yr/ lCi/m3, from Appendix 10.3.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 28 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

6. CONTAINMENT PURGE - RE-44 HASP The Containment Purge is monitored by rad monitor RE-44. RE-44 provides alarm and automatic release termination functions.

The HASP for RE-44 must limit the noble gas dose rate for skin and total body exposure. In order to address this, two set points are calculated. One set point is calculated based upon limiting the total body dose rate and the other limits the skin dose rate. The more limiting set point is used. The High Alarm Set Point methodology for RE-44 is given by Equations 26 and 27.

a) Limiting Concentration Based on Total Body Dose 44A/BMAc T(NG) = AF x SF x 500 (26) 4 72 xF, xýz-/c)L I ~Kj Where:

44 A/BMaxCT(NG) = the maximum allowable total concentration of noble gas in the containment purge exhaust line, in jiCi/cc.

AF = the allocation factor for the containment purge for one unit from Table 6.2.

SF = safety factor to ensure that dose rate limits of radiological effluent controls are not exceeded (0.9).

500 = (mrem/yr) the site Total Body dose rate limit for instantaneous releases.

472 = the conversion constant to cc/sec from cfm.

Fct = maximum flow rate in the containment purge system (maximum containment purge flowrate is 55,000 cfm).

(z-/-Q)M = the maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological data base averages, from Appendix 10.2.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 29 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Ki = whole body dose factor (mrem/yr/jiCi/m3) for isotope "i." Dose factors are presented in Appendix 10.3.

f= the fraction of the concentration of the individual noble gas radionuclide, "i," in the total mix of noble gas effluents in the containment purge line.

b) Limiting Concentration Based on Skin Dose Skin dose should be calculated whenever an RCS sample contains a significant fraction of Kr-85.

The concentration limit calculated by this method should then be compared to the concentration limit calculated for total body dose. The smaller concentration should be selected as the limiting concentration for the HASP.

44A'M*C(NG= A4AMC(N x F Ax472XFc, x(x__*)*t 3000 x (Li + 1.1IMi)fi (27)

Where:

44 A/BMaxCT(NG) = the maximum allowable total concentration of noble gas in the plant vent in p[Ci/cc.

AF = the allocation factor for the containment purge for one unit from Table 6.2.

SF = safety factor to ensure that dose rate limits of the radiological effluent controls are not exceeded (0.9).

3000 = (mrem/yr) the site skin dose rate limit for instantaneous releases.

472 = the conversion constant to cc/sec from cfm.

Fct = total flow rate in the containment purge system, in cfm (maximum containment purge flow rate is 55,000 cfm).

(z/Q)M = the maximum historical site boundary dispersion factor, based on 5 year averages derived from the meteorological data base averages, from Appendix 10.2.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 30 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Li the skin dose factor (mrem/yr per ptCi/m 3 ) for isotope "i." Dose factors are presented in Appendix 10.3.

1.1 = Conversion factor mrem/mrad. Converts air dose to skin dose.

Mi = the gamma air dose factor (mrad/yr per pCi/rn 3) for isotope 'i." Dose factors are presented in Appendix 10.3.

f the fraction of the concentration of the individual noble gas radionuclide, "i," in the total mix of noble gas effluents in the containment purge line.

c. Mode 6 Particulate activity.

The HASP calculation specified in this section based upon Noble Gas effluent limitations conservatively bounds the Tech. Spec. requirement for particulate activity in Mode 6. The FSAR expected case accident for Mode 6 is a containment fuel handling accident which does not include a particulate release. Therefore, the HASP for RM-44 in this section conservatively satisfies the Tech. Spec. (Ref.: 8.15).

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 31 OF 63 TITLE: Off-Site Dose Calculations UNITS I AND 2 6.2.3 Gaseous Dose Rate Calculation Methodology

a. Total Body Noble Gas Dose Rate Methodology The dose rate to the total body due to immersion in a cloud of noble gases is given by:

(z~-~)Ma Ki!i <500 mrnem/year (28)

Where:

i = The release rate of radionuclide i in units of jtCi/sec.

All other terms are as previously defined.

b. Skin Dose Rate Methodology The dose rate to the skin due to immersion in a cloud of noble gases is given by:

(z--/)j(Li +l.1Mi)M0iQ <3000mrem/year (29) i Where the terms are as previously defined.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 32 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND2

c. Radioiodine, Tritium and Particulate Dose Rate Methodology The dose rate to organ, o, due to radioiodines, tritium and particulates released in gaseous effluents is given by:

(z-Q-)M z Pi 0Qi <1500 mremlyear (30)

Where:

Pi. = K'(BR)DFAio and K' = 106 pCi/ptCi BR The breathing rate of the a child age group in m 3/yr.

The default value of 3700 m 3/yr is taken from Table E-5 of Reg. Guide 1.109.

DFAi= The inhalation dose factor for organ o, for the child age group for radionuclide, i, from Table E-9 of Reg. Guide 1.109 in mrem/pCi, with the following exceptions:

H-3, Sb-124 and Sb-125 inhalation dose conversion factors taken from NUREG/CR4013.

All other terms are as previously defined.

Values for Pi, are listed in Appendix 10.6.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 33 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND2 6.2.4 Noble Gas Air Dose Calculation Methodology

a. Gamma Air Dose The gamma air dose due to immersion in a cloud of noble gases is given by:

Day = 3.17 x 10-8 (z/Q-)M MQi (31) i Where:

Day = Gamma air dose in mrad.

3.17x 10-8 = Conversion constant yr/sec.

Mi = Gamma air dose factor for nuclide i, in mrad/yr per

ýtCi/m 3 . Values are listed in Appendix 10.3.

Q = Total release of noble gas radionuclide, i, in jiCi.

All other terms are as previously defined.

b. Beta Air Dose The beta air dose due to immersion in a cloud of noble gases is given by:

Dal = 3.17 x 10-8 (z )MZNiQi (32)

Where:

Da, = Beta air dose in mrad.

3.

Ni = Beta air dose factor for nuclide i, in mrad/yr per ýiCi/m Values are listed in Appendix 10.3.

All other terms are as previously defined.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 34 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2.5 Dose To Critical Receptor Due To Radioiodines, Tritium and Particulates Released in Gaseous Effluents

a. Calculation Methodology The dose to an individual (critical receptor) due to radioiodines, tritium and particulates released in gaseous effluents with half-lives greater than 8 days is determined based upon the methodology described in NUREG 0133. This methodology makes use of the maximum individual concept described in Regulatory Guide 1.109. The maximum individual is characterized as maximum with regard to food consumption, occupancy, and other usage parameters. This concept therefore models those individuals within the local population with habits representing reasonable deviations from the average. In all physiological and metabolic respects, the maximum individual is assumed to have those characteristics that represent the average for the age group of interest.

The concept of critical receptor is introduced as a further refinement of the maximum individual. The critical receptor is defined as that individual that receives the largest dose based upon the combination of dose pathways that have been shown to actually exist. The critical receptor concept is applied at that location where the combination of dispersion (X/Q), deposition (D/Q), existing pathways, occupancy time, receptor age group, and effluent source term indicates the maximum potential exposure. The inhalation and ground plane exposure pathways are considered to exist at all locations. The grass-cow-milk, grass-cow-meat, and vegetation pathways are considered based on their actual existence in the vicinity of the plant.

The dose pathways that have been shown to actually exist at DCPP are the ground plane, inhalation and the vegetation pathways. These dose pathways are reviewed yearly and updated based upon the annual land use census survey in order to insure that actual exposure to an individual will not be substantially underestimated.

The locations of the pathways and descriptions are listed in Appendix 10.2.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 35 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

b. Dose Calculation The dose contributions to the total body and each individual organ (bone, liver, thyroid, kidney, lung and GI-LLI) of the maximum exposed individual (Critical Receptor) due to radioactive gaseous effluent releases is calculated for all radionuclides identified in gaseous effluents released to unrestricted areas using the following expression:

Dapo =3.17xx0-8 WcRZRapoQi (33) i Where:

Dapo - Dose to the critical receptor for age group a, pathway p, and organ o, in mrem.

WCR = Critical receptor X/Q for immersion, inhalation and all tritium pathways (seconds/m 3) from Appendix 10.2.

= Critical receptor D/Q for ground plane and all ingestion pathways (1/mi2) from Appendix 10.2.

Raipo = Site specific dose factor for age group a, radionuclide i, pathway p, and organ j (mrem/yr per jiCi/m 3 for inhalation and tritium pathways - mrem/yr per ýtCi/(sec 2

in ) for ground plane and ingestion pathways). These dose factors are listed in Appendix 10.6.

The site specific dose factors are calculated based upon NUREG 0133 methodology. All dose conversion factors are taken from Reg. Guide 1.109, Rev 1, Tables E6-E14, with the following exceptions: H-3, Sb-124 and Sb-125 dose conversion factors taken from NUREG/CR-4013.

= The total release of radionuclide i, in units of jiCi.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 36 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2.6 Noble Gas Gaseous Radioactive Waste (GRW) Batch Release Percent Release Rate Limits (PRRLs) and Expected Reading (ER)

The Percent Release Rate Limit (PRRL) for noble gas releases for each unit is calculated based upon the 500 mrem/yr whole body dose rate limit, and is given by Equation 34.

PRRL =m (0.48)(500torero// yr) X100% (34)

Where:

L (x-Q)M- The maximum site boundary dispersion factor based on 5 year averages from Appendix 10.2.

3 Ki Whole body dose factor (mrem/yr per ýLCi/im ) for isotope "i." Dose factors are presented in Appendix 10.3.

Qi - Total release rate of isotope "i" from all sources discharged through this release point in [tCi/sec.

0.48 Plant vent allocation factor for one unit from Table 6.2.

500 mrem/yr = Site noble gas dose rate limit.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 37 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 The Expected Reading (ER) is the anticipated monitor response based upon the known plant vent concentration and the monitor response factors. The Expected Readings for RE-22, RE-44, RE-14 are given by Equations 35, 36, and 37.

ERR-2)":BKG(RE2 2 ) + k(E2)iCP-2 (35)

Where:

ERR- 22) = Expected reading on monitor RE-22 in pCi/cc.

BKG(U_22) = Monitor background.

(RE-22)i = Noble gas monitor response factor for nuclide "i" for monitor RE-22.

C(E-22)i = Concentration of nuclide "T' seen by RE-22.

ER(PE-44) - BKG(PE-44) + CCSP(PE-44) k(PE-44 ) iCP-4 (36)

Where:

ERýRE4a) = Expected reading on monitor RE-44 in pCi/cc.

BKG(u44) = Monitor background.

CCSP(RE4 4) = Conversion constant setpoint for monitor RE-44.

k(RE44)i = Noble gas monitor response factor for nuclide "i" for monitor RE-44.

R-44)i = Concentration of nuclide "i" seen by monitor RE-44.

ER(RE-14) ýBKG(PE-14) + CCS"'P(P~E-4) k(PE-14) iCR-4 (37)

Where:

ER(RE-14) = Expected reading on monitor RE- 14 in piCi/cc.

BKG(Rl1 4 ) = Monitor background.

CCSP(RE-14 ) = Conversion constant setpoint for monitor RE-14.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 38 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 N{REI4)i Noble gas monitor response factor for nuclide "i" for monitor RE-14.

CE-14)i = Concentration of nuclide "i" seen by monitor RE-14.

Generally if the Expected Reading (ER) is greater than the existing HASP setting (an "administrative limit" as set by CY2.DC 1) then no release should be made until a calculation shows that the HASP (Admin Limit) can be raised so the release can be legally discharged. On the other hand should the ER be less than the existing HASP (Admin Limit), then the release can be discharged.

6.2.7 IPT - PRRL The Percent Release Rate Limit (PRRL) for radioiodines, tritium and particulates for each unit is calculated based upon the 1500 mrem/yr organ dose rate limit. The dose rate is calculated for the inhalation pathway to the child age group using the highest (worst case) organ dose factor for nuclide.

The Percent Release Rate Limit based on the worst case organ is given by Equation 38.

PRRL = 0lO (38)

(0.48X1500mrem/ yr)x100 (

Where:

-= The maximum site boundary dispersion factor based on 5 year averages from Appendix 10.2.

Piw = Inhalation dose factor for nuclide "i" (mrem/yr/ýtCi/m 3) for child age group for worst case organ, from Appendix 10.4.

Dose factors are based upon NUREG 0133 methodology.

Inhalation dose conversion factors are taken from Reg. Guide 1.109, Rev 1, Table E-9, with the following exceptions: H-3, Sb-124 and Sb-125 inhalation dose conversion factors taken from NUREG/CR-4013.

= Release rate of isotope "i" in pCi/sec.

0.48 = Plant vent location factor for one unit from Table 6.2.

1500 mrem/yr = Site radioiodine, tritium and particulate dose rate limit.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 39 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.2.8 Alternate Dose Methodologies For purposes of routine gaseous effluent dose assessment, the methodology of NUREG 0133 (described in Section 6.2.5) will be used. However, DCPP may elect to utilize the dose methodologies of Regulatory Guide 1.109 or the GASPAR computer code for special purposes such as evaluation of potential new gaseous effluent dose pathways or critical receptors.

6.2.9 Gas Effluent Dose Projection The projected dose contributions from each reactor unit due to gaseous effluents for the current calendar month, quarter and current calendar year must be determined in accordance with the methodology and parameters in the ODCP at least once per 31 days.

The computer program, Radioactive Effluent Management System (REMS), is used for this projection. Therefore, by the first day of the year, the following tables in REMS need to be updated:

GRW dose receptor

GRW dose rate receptor

GRW external dose select

GRW external occupancy

GRW internal dose select

" GRW internal occupancy The purpose of this is to determine if appropriate treatment of gaseous radioactive materials in relation to maintaining releases "as low as reasonably achievable," is necessary.

The projected dose from each reactor unit is given by:

Dp = DpU + {Dpcm (39)

Where:

DP = Projected Dose.

DPU = Projected dose attributed to reactor unit, U.

Dp,com = Projected dose common to both reactor units.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 40 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 The 31 day projected dose is calculated by Equation 40.

D*: xDpm + dcm + dc (T+ t) (40)

Where:

D*f Projected Dose for the next 31 day period.

D'M Previous Month's Actual Dose.

dcM = Current Month Actual Dose to date.

dcB = Projected Dose from Current Batch Release.

T Number of days in the previous month.

t = Number of days into the present month.

Projected quarterly doses are determined by Equation 41.

C o +D +dcQ +dcB DcP =dAQ+ (92-t) A A-P(1 (T+t) (41)

Where:

= Projected dose for the current calendar quarter.

dA Current quarter to date actual dose.

DA = Previous quarter's actual dose.

d = Projected dose as a result of the current batch release.

T = Number of days in the previous quarter.

t = Number of days into the present quarter.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 41 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 Projected yearly doses are determined by Equation 42.

DCY

,-- t3 6+d36-t

=dD7

_ DAdA p +dcy + d CB p B (42)

P= 6A (T+t)

Where:

DPCY Projected dose for the current calendar year.

d cy

.4 Current year to date actual dose.

Dpy A Previous year's actual dose.

d PCB = Projected dose as a result of the current batch release.

T = Number of days in the previous year.

t = Number of days into the present year.

6.2.10 Unplanned Gaseous Releases (Abnormal Releases)

a. An unplanned release is an unexpected and potentially unmonitored release to the environment due to operational error or equipment malfunctions.

1. Unmonitored unplanned releases shall have a report written by the Radiochemistry Effluents Engineer describing the event with a calculation, if possible, of the percent of RECP limit. This will then be forwarded to PSRC for review. Describe these unplanned releases in the Annual Radioactive Effluent Release Report.

2. Monitored unplanned releases which exceed 1% of the RECP limit will also have a report written describing the event and must be forwarded to the PSRC for review. Describe these unplanned releases in the Annual Radioactive Effluent Release Report. For purposes of classification only, unplanned release puffs through the plant vent may use one hour integrated resolution times.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 42 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND2 6.3 40 CFR 190 Dose Calculations 6.3.1 Pathway Calculation of total uranium fuel cycle dose for purposes of demonstrating compliance with 40 CFR 190 requires the contributions from liquid and gaseous effluent as well as direct radiation from selected outside storage tanks and storage buildings. The total uranium fuel cycle dose to any member of the public will be calculated by summing the following doses:

Direct Radiation Dose

Liquid Effluent Dose

Noble Gas Dose

" Radioiodine, Tritium and Particulate Gaseous Effluent Dose CAPA-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 43 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.3.2 Methodology

a. Direct Radiation Dose Routine determination of direct radiation dose from selected outside storage tanks and storage buildings may be made by direct survey measurements, derived from area TLD data, or calculated by shielding code.

The direct radiation dose will also take into account residence times near the site based upon land use census information.

The direct radiation determination using environmental TLD is given by equation 43.

Fra 12 - xr )

DSb L ad]-j rsb xD o xBxe 0 sb (3 (43) where:

D'sb = the dose rate at the site boundary, in mrem D',0 = the dose rate from the dosimetry reading, in mrem radj = the distance from the point source to the dosimetry, in meter rsb = the distance from the point source to site boundary, in meter B = buildup factor

= l+(,u xb x((,Ua +Us)//u))

gta = total absorption coefficient

= 0.001 x e(. 34 - (0.105 x (n(10 xE) - 57)2 x(2730k/Tok) ts = total Compton scatter coefficient

= 0.001xe (3.10- (0.089 x (ln(10 x E) + 1.89) 2 x (273°k/Tok)

ýto = total attenuation coefficient

= ga + Is E = external effective average gamma energy per disintegration of the source (Mev)

T'k = average absolute temperature (Kelvin)

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 44 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

b. Noble Gas Dose The noble gas skin dose and total body dose contributions to the total uranium fuel cycle dose to a member of the public will be determined as shown in Equations 44 and 45.

Noble Gas Total Body Dose = 3.17 x lO8/Q)R I KiQ, (44)

Noble Gas Skin Dose = 3.17 x 10-V- -(Li + 1. 1M )i (45)

Where:

3.17xlO8 Conversion constant yr/sec.

(---)R = Maximum historical dispersion factor for receptor of interest, based on 5 year averages from Appendix 10.2.

Ki= Whole body dose factor for nuclide i, in mrern/yr per ptCi/m 3 . Values are listed in Table 6.3.

Li = Skin dose factor for nuclide i, in rnremi/yr per jCi/m3. Values are listed in Table 6.3.

1.1 = Conversion factor mrem/mrad. Converts air dose to skin dose.

Mi= Gamma air dose factor for nuclide i, in mrad/yr per pCi/m3 . Values are listed in Appendix 10.3.

S= Total release of noble gas radionuclide, "i", in pCi/sec.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 45 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2

c. Liquid and Gaseous Effluent Dose The doses from liquid effluents and radioiodines, tritium and particulates in gaseous effluents will be determined by Equations 1 and 33, respectively.

For purposes of calculating the dose required by the radiological effluent controls, more realistic assumptions concerning the liquid and gaseous effluent dose pathways will be used, based upon the most recent land use census data as well as the latest environmental monitoring information.

These assumptions may include, but not be limited to: more realistic liquid dilution factors, location and age of actual individuals, site specific food pathway parameters, and documentation of true food consumption. Other assumptions may be used provided they can be substantiated by census or direct measurement.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 46 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2 6.4 On-Site Dose to Members of the Public Members of the public are occasionally granted access within the site boundary, but only in the owner controlled area up to the protected area boundary. The most common public access activities are: tours to the simulator (training building) or Bio Lab, policemen using the shooting range (most frequent activity), cattle drives through to adjacent properties, and visits of American Indians to on-site burial grounds (closest to the plant).

Exposure to members of the public due to liquid releases while on-site is highly unlikely and therefore not addressed. Exposure due to gaseous releases and direct radiation are credible and therefore are considered.

The dose to members of the public during on-site activities will be primarily determined by the duration of the on-site visitation time and by the closest proximity to the plant.

For gaseous releases the doses are calculated using Equations 44, 45 and 33. The Ri's in Equation 33 consider only the inhalation and ground plane pathway and exclude the N

infant age group.

The X/Q and D/Q values are modified using logarithmic extrapolation from the site boundary to the on-site location of interest as shown in Equations 46 and 47.

logIX on -site 1 log[X/Q]S.B - 1og[X/Q] 10oc .[lg(diston- site) - log(dist.S.B.)](46) log(dist.S.B.) - log(distjloc.)[Odt"

+ 1og[X/Q]s..

1og[D/Q]~o 1 -e 1lg[D/Q]sff- 1lg[D/Q] 0 c' - site)- log(dist.S.B.)](47) log(dist.S.B.) - log(dist.lo.)1

+ 1og[D/Q]s...

Based upon Regulatory Guide 1.111, these equations can be expected to provide reasonable dispersion and deposition estimates for distances as close as 200 meters.

Determination of direct radiation dose from the reactor units and from outside storage tanks may be made by direct survey measurements, derived from environmental TLD data, or calculated by shielding code.

A distance of 200 meters from the plant (both units) equidistant from the plant vent is arbitrarily selected as the closest perimeter for which on-site doses will be calculated.

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PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 47 OF 63 TITLE: Off-Site Dose Calculations UNITS 1AND 2 The activities of the members-of-the-public while on-site (described above), are at or beyond 200 meters. Table 6.3 details the types of on-site activities that members-of-the-public might be expected to participate in at DCPP. The sectors and closest distances in which they may visit as well as expected visitation duration are also shown (based on Security Section information).

Table 6.3 Expected On-Site Distances and Visitation Times for Members of the Public SECTOR CLOSEST POINT AVERAGE EXPECTED ONSITE MEMBER OF OF APPROACH VISITATION OF THE PUBLIC VISITATION TO PLANT TIME PER YEAR Police at SE 700m 208 hours0.00241 days 0.0578 hours 3.439153e-4 weeks 7.9144e-5 months shooting range Tour Participants Simulator Bldg S (SE) 310m 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Overlook E 210m 1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months American Indians NW 200m 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months at burial grounds NNW 200m 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months Ranch hands driving NW 250m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months cattle around site NNW 350m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months N 320m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months NNE 450m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months NE 630m 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months CAPA-8u3r35.DOC 08 1005.1256

PACIFIC GAS AND ELECTRIC COMPANY NUMBER CAP A-8 DIABLO CANYON POWER PLANT REVISION 35 PAGE 48 OF 63 TITLE: Off-Site Dose Calculations UNITS 1 AND 2