· enclosure pg8 e letter dcl 99-055 diablocanyon power plant annual radioactive effluent release...

EnclosurePG&E Letter DCL 99-055

DIABLOCANYON POWER PLANT -—

1998 ANNUALRADIOACTIVEEFFLUENT RELEASE REPORT

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January 1 - December 31, 1998

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EnclosurePG8E Letter DCL 99-055

DIABLOCANYON POWER PLANT

ANNUALRADIOACTIVEEFFLUENT RELEASE REPORT

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JANUARY 1 - DECEMBER 31, 1998

EnclosurePG8 E Letter DCL 99-055


Annual Radioactive Effluent Release ReportJanuary 1, 1998 Through December 31, 1998

Table of Contents

Introduction

Supplemental Information 4

II.

III.

IV.

V.

VI.

VII.

VIII

IX.

XI.

XII

Major Changes to Liquid, Gaseous, and Solid Radwaste Systems ..........

Changes to the Radiological Monitoring and Controls Program (RMCP) ...

Changes to the Radiological Effluent Controls Program (RECP) ............

Changes to the Environmental Radiological Monitoring (ERMP) ...........

Changes to the Off-Site Dose Calculation Procedure (ODCP) .

Changes to the Process Control Program (PCP)

Land Use Census

Gaseous and Liquid Effluents

TABLE 1 - GASEOUS EFFLUENTS - SUMMATIONOF ALLRELEASES

TABLE2 - GASEOUS EFFLUENTS - GROUND-LEVELRELEASES

TABLE3 - GASEOUS EFFLUENTS - LOWER LIMITSOF DETECTION .

TABLE4 - LIQUIDEFFLUENTS - SUMMATIONOF ALLRELEASES

TABLE5 - LIQUIDEFFLUENTS - NUCLIDES RELEASED

TABLE6 - LIQUID EFFLUENTS - LOWER LIMITSOF DETECTION (LLD) ...

Solid Radwaste Shipments

Radiation Dose Due to Gaseous and Liquid Effluents

TABLE7 - RADIATIONDOSE DUE TO THE RELEASE OF RADIOACTIVELUQUID

EFFLUENTS

TABLE8 - RADIATIONDOSE DUE TO THE RELEASE OF RADIOACTIVEGASEOUS

EFFLUENTS

TABLE9 - PERCENT OF TECHNICALSPECIFICATION LIMITSFOR RADIOACTIVELIQUID

EFFLUENTS

TABLE 10 - PERCENT OF TECHNICALSPECIFICATION LIMITSFOR RADIOACTIVE

GASEOUS EFFLUENTSTABLE 11 - ON-SITE DOSE TO MEMBERS OF THE PUBLIC .

Meteorological Data .

10

10

10

10

10

12

121519212733

36

39

41

42

45

46

48

53

EnclosurePGSE I etter DCL 99-055


Annual Radioactive Effluent Release ReportJanuary 1, 1998 Through December 31, 1998

Attachments

1. Corrections to 1997 Annual Radioactive Effluent Release Report

2. Radiological Monitoring And Controls Program (RMCP), CY2 Revision 2A

3. Radioactive Effluent Control Procedure (RECP), IDAP CY2.ID1 Revision 3A

4. Off-Site Dose Calculation Procedure (ODCP), CAP A-8 Revision 22I

5. Process Control Program, RP2. DC2 Revision 4

6. Process Control Program, RP2. DC2 Revision 5

7. Mobile Service Operating Procedures for Low Level Radioactive Waste Processing,RP2. DC4 Revision 9

8. Land Use Census

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1998 Annual Radioactive Effluent Release Report

Introduction

The 1998 Annual Radioactive Effluent Release Report summarizes gaseous and liquideffluent releases from Diablo Canyon Power Plant's (DCPP) Units 1 and 2. The reportincludes the dose due to release of radioactive liquid and gaseous effluents-andsummarizes solid radwaste shipments. The report contains information required by Units 1

and 2 Technical Specification (TS) 6.9.1.6 and 6.14 and is presented in the general formatof Regulatory Guide 1.21, Appendix B. Procedure revisions which implement theRadiological Monitoring and Controls Program (RMCP), Radioactive Effluent ControlProcedure (RECP), Off-Site Dose Calculation Procedure (ODCP), and Process ControlProgram (PCP) and three diskettes containing meteorological data are attached.

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

-3-


I. 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 radioactivenoble gases released in gaseous effluents is limited to less than or equal to500 millirem (mR) per year to the total body and less than or equal to 3000mR per year to the skin. [Radioactive Effluent Control Procedure (RECP)[6.1.6.1.a].

b. Particulate and Iodine Dose Rate Limit

The dose rate in areas at or beyond the site boundary due to iodine-131,iodine-133, tritium, and all radionuclides in particulate form with half livesgreater than 8 days in gaseous effluents, is limited to less than or equal to1500 mR per year to any organ. (RECP 6.1.6.1.b).

c. Noble Gas Dose Limit

The air dose due to noble gases released in gaseous effluents from eachreactor unit to areas at or beyond the site boundary is limited to the following:

Radiation Type

GammaBeta

Calendar Quarter LimitRECP 6.1.7.1.a

5 millirad10 millirad

Calendar Year LimitRECP 6.1.7.1.b

10 millirad20 millirad

d. Particulate and iodine Dose Limit

The dose to an individual from iodine 131, iodine 133, tritium, and allradionuclides in particulate form with half lives greater than 8 days in gaseouseffluents released from each reactor unit to areas at or beyond the siteboundary is limited to less than or equal to 7.5 mR to any organ in anycalendar quarter and less than or equal to 15 mR to any organ during acalendar year. (RECP 6.1.8.1).


2. Liquid Effluents

a. Concentration

The concentration of radioactive material released from the site is limited tothe 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 10microcuries/milliliter (pCi/ml) total activity. -(RECP 6:1:3;1).

b. Dose

The dose or dose commitment to a member of the public from radioactivematerials in liquid effluents released from each reactor unit to areas at orbeyond the site boundary is limited to the following:

Dose Type

Total BodyAny Organ

Calendar QuarterLimitRECP 6.1.4.1

1.5 millirem5 millirem

Calendar Year LimitRECP 6.1.4.1

3 millirem10 millirem

B. Maximum Permissible Concentrations


Maximum permissible concentrations are not used for determining allowablerelease rates for gaseous effluents at DCPP.

2. Liquid Effluents

The concentrations listed in 10 CFR 20, Appendix B, Table 2, Column 2, forradionuclides other than dissolved or entrained noble gases are used fordetermining the allowable release concentration at the point of discharge from thesite for liquid effluents. For dissolved or entrained noble gases, the allowablerelease concentration at the point of discharge is limited to 2 x 10~ pCi/ml totalactivity for liquid effluents.

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C. Measurements and Approximations of Total Radioactivity


Fission and Activation Gases

The gaseous radioactivity released from the plant vent is monitored by a pairof off-line monitors equipped with beta scintillator detectors. The monitorreadings are correlated to isotopic concentration based on. laboratory isotopicanalysis of grab samples using a germanium detector.

When plant vent measurements, as indicated by the process monitors, arebelow the lower limitof detection, grab samples results are used to quantifyreleases. The individual batch release data are used to quantify theradioactivity discharged from the gas decay tanks and containment.

A noble gas grab sample is obtained and analyzed at least weekly. Theisotopic mixture is assumed to remain constant between grab sampleanalyses.

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

The gaseous radioactivity released from the steam generator blowdown tankvent is measured by analyzing liquid or steam condensate grab samples witha germanium detector. A factor R, a ratio of unit masses between waterflashing to steam and water entering the tank, is used to calculate the activity.The isotopic concentrations are assumed to remain constant between grabsamples.

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 continuoussample collection on silver zeolite cartridges. The cartridges are changed atleast weekly and analyzed with a germanium detector. The radioiodinereleases are averaged over the period of cartridge sample collection.

Other potential pathways for releasing radioiodines are periodically monitoredby collecting samples using charcoal cartridges and analyzing with agermanium detector.

-6-


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

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

Tritium

Tritium released from the plant vent is monitored by passing a measured'olumeof plant vent sample through a water column and determining the

tritium increase in the water. An aliquot of the water is counted in a liquidscintillation spectrometer. The minimum'routine sample frequency for tritiumis weekly. The tritium concentration is assumed to remain constant betweensamples.

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 decaytank 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.

'Total error = (@2< + a 2 + ac 3+ a

Where cz, = error associated with each component.

2. Liquid Effluents

Batch Releases

Each tank of liquid radwaste is analyzed for principal gamma emitters using agermanium detector prior to release. A monthly prerelease analysis includesdissolved and entrained gases. Volume proportional monthly and quarterlycomposites are prepared from aliquots of each tank released. The monthlycomposite is analyzed for tritium using a liquid scintillation spectrometer and

-7-


gross alpha radioactivity using an internal proportional counter. The quarterlycomposite is analyzed for iron-55 using a liquid scintillation spectrometer andfor strontium-89 and -90 using an internal proportional detector followingchemical separations.

b. Continuous releases

For the continuous liquid releases of the steam generator blowdown tank andturbine building sump oily water separator;-daily-grab samples are collectedand aliquots are proportioned for weekly, monthly, and quarterly composites.

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

The steam generator blowdown tank monthly composite is analyzed for tritiumusing a liquid scintillation spectrometer and for gross alpha using an internalproportional 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. Theresults for each of the composites are averaged over the period of thecomposite.

In addition, one grab sample of the steam generator blowdown tank isanalyzed monthly for dissolved and entrained gases using a germaniumdetector. The results of this analysis are assumed to remain constant overthe period of one month.

c. Estimation of Overall Error

Sources of error considered are: (1) calibration source error; (2) calibrationcounting error; (3) sampling error; (4) sample counting error;.and (5) volumeof waste release error.

These sources of error are independent; thus the total error is calculatedaccording to the following formula:

Total error = (a', + a', + a', + ... a',) '"

Where o; = error associated with each component

-8-'

g .il '


D. Batch Releases

1. Liquida. Number of batch releasesb. Total time period for batch releasesc. Maximum time period for a batch released. Average time period for a batch releasee. Minimum time period for a batch releasef. Average saltwater flow during batch releases

6471541.38 hours

48.00 hours2.38 hours0.03 hours

— - 1.63E+6 GPM

2. Gaseousa. Number of batch releasesb. Total time period for batch releasesc. Maximum time period for a batch released. Average time per'iod for a batch release,e. Minimum time period for a batch release

135926.82 hours

72.00 hours6.87 hours0.12 hours

E. Abnormal Release (Gaseous and Liquid

From December 24, 1998, through December 31, 1998, an unplanned, butmonitored, release of Unit 1 condensate occurred due to a leak on the CondensateDischarge Overboard Valve CND-1-128. During this event, 3.09E-2 curies of tritiumwere released. The percent of effluent concentration limitwas 5.11 E-4 percent ofthat specified in RMCP IDAP CY2.ID1, step 6.1.3.1.

F. Effluent Monitoring Instrumentation

On March 25, 1998, RE-18 (Liquid Effluent Monitor) alarmed. The alarm was mostlikely due to activated corrosion product accumulation in the piping dead legs sharedby the liquid radwaste system discharge and process pathways. The activatedcorrosion products were transported to the RE-18 monitor during the beginning ofdischarge causing a RCV-18 trip which terminated the release. The residue in theline was then flushed back to the liquid radwaste system after a sample was obtainedand analyzed. The percent of effluent concentration limitwas 1.65 E+0 percent ofthat specified in RMCP IDAP CY2.ID1, step 6.1.3.1.

II. Major Changes to Liquid, Gaseous and Solid Radwaste Treatment System

There were no major changes to liquid, gaseous, and solid radwaste treatment systemsduring the report period.

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III. Changes to the RMCP

CY2, "Radiological Monitoring And Controls Program," was revised during the reportperiod. Revision 2A made the following change.

Organizational name was changed from "Chemistry and Environmental Operations" to"Chemistry."

The change was reviewed by the Plant Staff Review Committee (PSRC) and approvedby the Plant Manager on March 26, 1998. (See Attachment 2).

IV. Changes to the RECP

CY2.ID1, "Radioactive Effluent Controls Program," was revised during the report period.Revision 3A made the following change.

Organizational name was changed from "Chemistry and Environmental Operations" to"Chemistry."

The change was reviewed by the PSRC and approved by the Plant Manager onMarch 24, 1998. (See Attachment 3).

V. Changes to the ERMP

There were no changes to RP1.ID11, "Environmental Radiological MonitoringProcedure," during the report period.

Vl. Changes to the ODCP

CAP A-8, "Off-Site Dose Calculations," was revised during the report period. Revision22 made the following changes.

A. The methodology for direct radiation determination using environmental TLDs wasadded.

B. Table 10.2 was updated with the most recent 5-year meteorology data.

These changes were reviewed by the PSRC and approved by the Plant Manager onApril 3, 1998. (See Attachment 4).

Vll. Changes to the Process Control Program

RP2.DC2, "Process Control Program," Revision 4, adopted a revised polymersolidification vendor data sheet.

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EnclosUrePG8E Letter DCL 99-055

March 13, 1998. (See Attachment 5).

RP2.DC2, Revision 5, adopted vendor procedure revisions to change the name of thevendor.

The change was reviewed by the PSRC and approved by the Plant Manager onOctober 2, 1998. (See Attachment 6).

RP2.DC4, "Mobile Service Operating Procedures For Low-level Radioactive WasteProcessing," Revision 9, adopted vendor procedure revisions to change the name ofthe vendor.

The change was reviewed by the PSRC and approved by the Plant Manager onFebruary 27, 1998. (See Attachment 7).

Vill. Land Use Census

Changes to the Land Use Census Program are included as Attachment 8.

-11-


IX. Gaseous and Liquid Effluents

Table 1a

Gaseous Effluents - Summation of All Releases

Units First SecondQuarter Quarter

A. Fission and activation gases

Est. TotalError, %

1. Total release

2. Average release rate for period

3. Percent of technical specification limit"

Ci

pCi/sec

6.38E+0

8.20E-1

1.03E-3

1.16E-1

1.48E-2

4.58E-5

5.98E+1

B. Iodines

1. Total iodine-131



Ci

pCi/sec

1.39E-4

1.78E-5

1.04E-4

MDA

MDA

MDA

2.35E+1

C. Particulate

1. Particulate with half-lives >8 days Ci 1.84E-3 MDA 2.35E+1



ijCi/sec 2.37E-4

9.37E-5

MDA

MDA

4. Gross alpha radioactivity Ci 9.73E-8 1.58E-7

D. Tritium

1. Total release



Ci 7.72E+1

pCi/sec 9.93E+0

2.30E-5

3.28E+1

4.17E+0

9.65E-6

1.30E+1

MDA= Less than the "a posteriori" minimum detectable activity (microcuries per unit mass orvolume). This note applies to all tables.

" RECP 6.1.6.1 Limit

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Table 1b


Units Third FourthQuarter Quarter

Est. TotalError%

A. Fission & activation gases

1. Total release



Ci

ijCi/sec

1.09E-1

1.37E-2

3.96E-5

4.49E+0

5.64E-1

2.10E-4

5.98E+1

B. lodines

1. Total iodine-131



Ci MDA

pCi/sec MDA

MDA

MDA

MDA

MDA

2.35E+1

C. Particulate

1. Particulate with half-lives >8 days



Ci MDA

pCi/sec MDA

MDA

MDA

MDA

MDA

2.35E+1

4. Gross alpha radioactivity Ci 2.98E-7 3.28E-7

D. Tritium

1. Total release



Ci

pCi/sec

3.80E+1 2.93E+1

4.78E+0 3.68E+0

1.10E-5 8.52E-6

1.30E+1


" RECP 6.1.6.1 Limit.

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Table 1c


A. Fission & activation gases

1. Total release

Units

Ci

AnnualTotal

1.11E+1



pCi/sec 3.52E-1

3.28E-4

B. lodines

1. Total iodine-131 Ci 1.39E-4



pCi/sec 4.40E-6

2.57E-5

C. Particulate




4. Gross alpha radioactivity

Ci

pCi/sec

Ci

1.84E-3

5.84E-5

2.31E-5

'.81E-7

D. Tritium

1. Total release



Ci

ijCi/sec

1.77E+2

5.62E+0

1.30E-5

" RECP 6.1.6.1 Limit.

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Table 2a

Gaseous Effluents - Ground-Level Releases

Nuclides Released Unit

First Quarter

Continuous BatchMode Mode

Second Quarter


1. Fission gases

argon-41

krypton-85

krypton-85m

krypton-87

krypton-88

xenon-1 31m

xenon-1 33

xenon-133m

xenon-135

xenon-1 35m

xenon-1 38

Total For Period

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

MDA

MDA

MDA

MDA

MDA

MDA

4.52E+0

MDA

MDA

MDA

MDA

4.52E+0

6.66E-1

6.72E-1

3.98E-6

MDA

MDA

MDA

5.18E-1

2.02E-3

1.09E-3

MDA

MDA

1.86E+0

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

3.73E-2

7.63E-2

MDA

MDA

MDA

MDA

2.34E-3

MDA

4.61E-5

MDA

MDA

1.16E-1

2. lodines

iodine-131

iodine-133

iodine-135

Total For Period

Ci

Ci

Ci

Ci

1.39E-4

MDA

MDA

1.39E-4

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.

-15-


Table 2a (Continued)


Continuous Mode

Nuclides Released Unit First Quarter Second Quarter

3. Particulate

barium-140

cesium-134

cesium-137

cerium-141

cerium-144

chromium-51

cobalt-57

cobalt-58

cobalt-60

iron-59

lanthanum-1 40

manganese-54

molybdenum-99

niobium-95

ruthenium-103

strontium-89

strontium-90

zinc-65

zirconium-95

Total For Period

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

MDA

MDA

MDA

MDA

MDA

2.33E-5

MDA

1.82E-3

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

1.84E-3

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA


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EnclosurePGLE Letter DCL 99-055

Table 2b


Third Quarter Fourth Quarter

Nuclides Released Unit Continuous Batch ModeMode


1 ~ Fission gases

argon-41

krypton-85

krypton-85m

krypton-87

krypton-88

xenon-131m

xenon-1 33

xenon-1 33m

xenon-135

xenon-135m

xenon-1 38

Total For Period

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

3.22E-2

6.00E-2

MDA

MDA

MDA

MDA

1.64E-2

MDA

1.65E-5

MDA

MDA

1.09E-1

MDA

MDA

MDA

MDA

MDA

MDA

4.36E+0

MDA

MDA

MDA

MDA

4.36E+0

2.58E-2

MDA

MDA

MDA

MDA

MDA

9.96E-2

MDA

3.71E-4

MDA

MDA

1.26E-1

2. lodines

iodine-1 31

iodine-133

iodine-135

Total for period

Ci

Ci

Ci

Ci

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA


-17-

EnclosurePGSE Letter DCL 99-055

Table 2b (Continued)


Nuclides Released Unit

Continuous Mode


3. Particulates

barium-140

cesium-134

cesium-137

cerium-141

cerium-144

chromium-51

cobalt-57

cobalt-58

cobalt-60

iron-59

lanthanum-140

manganese-54

molybdenum-99

niobium-95

ruthenium-103

strontium-89

strontium-90

zinc-65

zirconium-95

Total For Period

Ci

Ci

Ci

,Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

'MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA= Less than the "a posteriori" minimum detectable activity (microcuries per unit mass orvolume). This note applies to ail tables.

-18-


Table 3

Gaseous Effluents - Lower Limits of Detection

Batch Mode

Nuclide ~ Unit ContinuousMode

ContainmentPurge

Gas DecayTank

1 ~ Fission gases

argon-41

krypton-85

krypton-85m

krypton-87

krypton-88

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

2.05E-S

2.99E-6

6.91E-9

3.18E-8

2.80E-8

2.05E-8

2.99E-6

6.91E-9

3.18E-S

2.80E-S

6.91E-8

7.84E-6

1.54E-8

5.74E-S

5.54E-S

xenon-131m

xenon-133

xenon-133m

xenon-1 35

pCi/ml

pCi/ml

pCi/ml

pCi/ml

xenon-1 35m pCi/ml

2.59E-7

2.14E-8

6.84E-8

6.84E-9

1.41E-7

2.59E-7

2.14E-8

6.84E-8

6.84E-9

1.41E-7

5.32E-7

5.06E-S

1.48E-7

1.48E-S

4.79E-7

xenon-138 pCi/mi 4.94E-7 4.94E-7 1.11E-6

2. Tritium

, hydrogen-3 pCi/ml 4.74E-9 4.74E-9

3. lodines

iodine-131

iodine-1 33

iodine-1 35

pCi/ml

pCi/ml

pCi/ml

5.19E-13

6.00E-13

3.27E-12

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Table 3 (Continued)

Gaseous Effluents - Lower Limits of Detection

Nuclide Unit Continuous Mode

4. Particulate

barium-140

cerium-141

cerium-144

cesium-134

cesium-137

chromium-51

cobalt-57

cobalt-58

cobalt-60

iron-59

lanthanum-140

manganese-54

molybdenum-99

ruthenium-1 03

strontium-89

strontium-90

zinc-65

gross alpha

pCi/ml

pCi/ml

pCi /m

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/m

pCi/ml

1.93E-12

4.30E-13

1.44E-12

3.50E-13

4.70E-13

2.96E-12

2.06E-13

6.92E-13

1.03E-12

1.01E-12

6.71E-13

4:63E-13

1.95E-13

4.35E-13

4.70E-15

1.78E-15

1.30E-12

1.36E-15

-20-


Table 4a

Liquid Effluents - Summation of All Releases

Unit FirstQuarter

SecondQuarter

Est. TotalError, %

A. Fission and activation products

1. Total release (not includingtritium, gases, alpha)

2. Average diluted concentrationduring period

3. Percent of applicable limit"

Ci

pCi/ml

3.20E-1

1.43E-10

9.22E-4

1.14E-1

4.19E-11

2.58E-4

2.35E+1

B. Tritium

1. Total release



Ci

pCi/ml

5.16E+2

2.31E-7

2.31E-2

2.78E+2

1.02E-7

1.02E-2

1.30E+1

C. Dissolved and entrained gases

1. Total release



Ci

pCi/ml

6.69E-4

3.00E-13

1.50E-7

4.77E-6

1.75E-15

8.75E-10

2.35E+1

D. Gross alpha radioactivity

1. Total release Ci MDA MDA 6.10E+1


'RECP 6.1.6.1 Limit

-21-

EnclosurePGSE Letter DCL 99-055



Unit FirstQuarter

SecondQuarter

Est. TotalError, %

E. Volume of waste released(prior to dilution)

liters 7.74E+7 8.12E-7 5.0E+0

F. Volume of circulating saltwater litersused during the release

2.23E+12 2.73E+12 6.5E+0

-22-


Table 4b


Unit ThirdQuarter

FourthQuarter

Est. TotalError, %


1. Total release (not includingtritium, gases, alpha)



Ci

pCi/ml

2.72E-2

1.02E-11

8.93E-5

7.50E-2

2.83E-11

3.53E-4

2.35E+1

B. Tritium

1. Total release



Ci

pCi/ml

2.17E+2

8.14E-8

8.14E-3

2.74E+2

1.03E-7

1.03E-2

1.30E+1


1. Total release



Ci

pCi/ml

2.66E-5

1.00E-14

5.00E-9

7.20E-4

2.72E-13

1.36E-7

2.35E+1


1. Total release Ci MDA MDA 6.10E+1



-23-




Unit ThirdQuarter

FourthQuarter

Est. TotalError, %

E. Volume of waste released(prior to dilution)

liters 8.04E+7 8.27E+7 5.0E+0

F. Volume of circulating saltwater liters 2.66E+12used during release

2.65E+12 6.5E+0

i

-24-


Table 4c


Unit Annual Total


1. Total release (not including tritium,gases, alpha)



Ci

pCi/ml

'5.36E-1

5.22E-11

3.83E-4

B. Tritium

1. Total release

2. Average diluted concentrationduring period during period


Ci

pCi/ml

1.28E+3

1.25E-7

1.25E-2


1. Total release

2. Average diluted concentrationduring period during period


Ci

pCi/ml

1.42E-3

1.38E-13

6.92E-8


1. Total release Ci MDA



-25-


Table 4c (Continued)


Unit Annual Total

E. Volume of waste released (prior todilution)

liters 3.22E+8

F. Volume of circulating saltwater usedduring release

liters 1.03E+13

-26-


Table 5a

Liquid Effluents - Nuclides Released

First Quarter Second Quarter

Nuclides Released

antimony-1 22

antimony-124

antimony-1 25

barium-140

beryllium-7

bromine-82

cerium-141

cerium-143

cerium-144

cesium-134

cesium-136

cesium-1 37

cesium-1 38

chromium-51

cobalt-57

cobalt-58

cobalt-60

iron-55

iron-59

lanthanum-140

Unit

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

ContinuousModeMDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

2.20E-4

MDA

MDA

MDA

MDA

Batch Mode

1.63E-3

5.60E-3

6.16E-2

7.83E-7

MDA

MDA

9.30E-6

MDA

MDA

MDA

MDA

1.06E-3

3.17E-5

2.73E-2

4.82E-4

1.71E-1

1.87E-2

1.77E-2

2.28E-3

1.75E-4

ContinuousModeMDA

'DA'DA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

'DA

MDA

MDA

MDA

MDA

MDA

Batch Mode

MDA

1.39E-3

2.19E-2

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

8.70E-5

MDA

3.05E-3

2.25E-4

7.20E-2

6.05E-3

6.39E-3

1.93E-4

MDA


-27-





Nuclides Released

lanthanum-142

manganese-54

manganese-56

molybdenum-99

niobium-95

niodymium-147

rubidium-89

ruthenium-1 03

silver-110m

sodium-24

strontium-89

strontium-90

strontium-91

strontium-92

tellurium-129m

tellurium-1 31

tellurium-1 32

tin-113

Unit

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

ContinuousModeMDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA"

MDA'DA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

Batch Mode

MDA

1.86E-3

MDA

2.15E-4

3.27E-3

1.82E-5

MDA

1.44E-4

8.09E-5

5.76E-5

MDA

MDA

MDA

MDA

5.61E-4

MDA

1.42E-3

4.61E-4

ContinuousNlodeMDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

Batch Mode

MDA

6.54E-4

MDA

2.96E-5

9.91E-4

MDA

MDA

6.86E-6

3.97E-4

4.20E-6

MDA

MDA

MDA

MDA

MDA

MDA

MDA

5.37E-5


-28-





Nuclides Released

tin-117m

tungsten-1 87

zinc-65

zirconium-95

iodine-1 31

iodine-1 32

iodine-133

iodine-134

iodine-135

Total For Period

Unit

,Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

ContinuousModeMDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

2.20E-4

BatchMode

1.65E-4

MDA

MDA

1.95E-3

6.38E-4

1.24E-3

3.63E-4

2.33E-7

3.72E-5

3.20E-1

ContinuousNlodeMDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

BatchMode

3.31E-5

MDA

7.60E-5

4.56E-4

1.38E-4

MDA

5.08E-5

MDA

MDA

1.14E-1

Dissolved And Entrained Gases

xenon-133

xenon-1 33m

xenon-1 35

krypton-85

krypton-87

krypton-88

Total For Period

Ci

Ci

Ci

Ci

Ci

Ci

MDA

MDA

MDA

MDA

MDA

MDA

MDA

5.60E-4

MDA

1.10E-4

MDA

MDA

MDA

6.69E-4

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

4.77E-6

MDA

MDA

MDA

4.77E-6


-29-

0'I

I


Table 5b



Nuclides Released

antimony-122

antimony-124

Unit

Ci

Ci

ContinuousModeMDA

MDA

Batch Mode

MDA

ContinuousModeMDA

4.94E-5 'DABatch Nlode

MDA

MDA

antimony-1 25

barium-140

. beryllium-7

bromine-82

cerium-141

cerium-143

cerium-144

cesium-1 34

cesium-136

cesium-1 37

cesium-1 38

chromium-51

cobalt-57

cobalt-58

cobalt-60

iron-55

iron-59

lanthanum-1 40

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

5.82E-3

MDA

MDA

MDA

MDA

MDA

MDA

5.21E-6

MDA

1.46E-4

MDA

MDA

6.66E-5

1.11E-2

2.54E-3

6.24E-3

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

1.02E-2

MDA

MDA

MDA

MDA

MDA

MDA

7.67E-6

MDA

2.06E-4

MDA

MDA

1.35E-4

8.37E-3

1.59E-2

3.50E-2

MDA

MDA


- 30-





Nuclides Released

lanthanum-1 42

manganese-54

manganese-56

molybdenum-99

niobium-95

neodymium-147

rubidium-89

ruthenium-103

silver-110m

sodium-24

strontium-89

strontium-90

strontium-91

strontium-92

tellurium-129m

tellurium-1 31

tellurium-132~

tin-113

tin-117m

tungsten-187

zinc-65

Unit

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

Ci

ContinuousModeMDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

Batch Mode

1.17E-6

2.35E-4

MDA

2.65E-5

1.61E-4

MDA

MDA

MDA

6.16E-6

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

2.58E-6

MDA

MDA

5.51E-5

ContinuousModeMDA

- MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

MDA

Batch Mode

MDA

6.54E-4

MDA

1.17E-4

1.18E-4

MDA

MDA

MDA

1.18E-4

MDA

MDA

MDA

MDA

MDA

MDA

MDA

2.73E-6

1.64E-5

MDA

MDA

1.32E-3

-31-





Nuclides Released

zirconium-95

iodine-1 31

iodine-1 32

iodine-1 33

iodine-1 34

iodine-1 35

Total For Period

Unit

Ci

Ci

Ci

Ci

Ci

Ci

Ci

ContinuousModeMDA

MDA

MDA

MDA

MDA

MDA

MDA

Batch Mode

5.80E-5

5.08E-4

MDA

1.35E-4

MDA

MDA

2.72E-2

ContinuousModeMDA

MDA

MDA

MDA

MDA

MDA

MDA

Batch Mode

MDA

2.34E-3

MDA

4.65E-4

MDA

MDA

7.50E-2

Dissolved And Entrained Gases

xenon-133

xenon-1 33m

xenon-1 35

krypton-85

krypton-87

krypton-88

Total For Period

Ci

Ci

Ci

Ci

Ci

Ci

Ci

MDA

MDA

MDA

MDA

MDA

MDA

MDA

2.01E-5

MDA

6.51E-6

MDA

MDA

MDA

2.66E-5

MDA

MDA

MDA

MDA

MDA

MDA

MDA

7.10E-4

MDA

1.93E-5

MDA

MDA

MDA

7.20E-4


-32-


Nuclide

antimony-1 22

antimony-124

antimony-1 25

barium-140

beryllium-7

bromine-82

cerium-141

cerium-143

cerium-144

cesium-1 34

cesium-1 36

cesium-137

cesium-1 38

chromium-51

cobalt-57

cobalt-58

cobalt-60

iron-55

iron-59

lanthanum-140

lanthanum-142

manganese-54

manganese-56

molybdenum-99

Table 6

Liquid Effluents - Lower Limits of Detection (LLD)

Unit

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

LLD

5.71E-8

1.04E-7

8.55E-8

1.73E-7

3.34E-7

5.11E-8

3.73E-8

7.53E-8

1.89E-7

3.93E-8

5.14E-8

5.51E-8

3.00E-7

3.04E-7

2.28E-8

4.41E-8

5.80E-8

3.00E-7

9.66E-8

6.40E-8

1.19E-7

5.84E-8

3.52E-7

2.06E-8

-33-


niobium-95

Nuclide

Table 6 (Continued)


Unit

pCi/ml

LLD

4.86E-8

neodymium-147

rubidium-89

ruthenium-1 03

silver-110m

sodium-24

strontium-89

strontium-90

strontium-91

strontium-92

tellurium-1 29m

tellurium-1 31

tellurium-1 32

tin-113

tin-117m

tungsten-1 87

zinc-65

zirconium-95

gross alpha

hydrogen-3

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

1.20E-7

1.56E-6

4.13E-8

4.89E-S

7.56E-S

4.20E-S

1.50E-8

8.05E-8

8.41E-S

1.32E-6

1.65E-7

3.30E-8

4.52E-S

2.21E-S

1.17E-7

1.18E-7

7.05E-S

9.17E-8

4.47E-6

-34-


Table 6 (Continued)


iodine-131

iodine-1 32

iodine-1 33

iodine-1 34

iodine-1 35

xenon-133

Nuclide Unit

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

LLD

3.65E-8

5.78E-8

4.16E-8

8.60E-8

2.61E-7

, 1.01E-7

xenon-1 33m

xenon-135

krypton-85

krypton-87

krypton-88

pCi/ml

pCi/ml

pCi/ml

pCi/ml

pCi/ml

2.29E-7

2.60E-8

1.29E-5

9.64E-8

1.13E-7

-35-


X. Solid Radwaste Shipments

Solid Waste and Irradiated fuel Shipment

A. Solid Waste Shipped Offsite for Burial or Disposal (Not irradiated fuel)

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

a. Spent Resins, Filter Sludges,Evaporator Bottoms, etc.

b. Dry Compressible Waste,Contaminated Equipment, etc.

c. Irradiated Components, ControlRods, etc.

d. Other

m3Ci

m3Ci

m3Ci

m3Ci

9.20E+0

3.77E+2'.62E+1

5.01E+0

0.00E+00.00E+0

0.00E+00.00E+0

0.00E+04.92E+0

0.00E+01.00E+1

0.00E+00.00E+0

0.00E+00.00E+0

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

Fe-55

Co-60

Ni-63

Cs-137

Mn-54

43

30

17

b. Fe-55

Co-58

Co-60

Zr-95

Ni-63

Sb-125

H-3

C-14

Cr-51

41

10

Not Applicable N/A

Not Applicable N/A

-37-


Solid Waste and Irradiated Fuel Shipment (Continued)

3. Solid Waste Disposition

Number of Shipments

36

Mode ofTransportation

Truck

Truck

Rail

Destination

BURNWELL, SC

ENVIROCARE, UT

ENVIROCARE, UT

4. Supplemental Information Required by TS 6.9.1.6

Solidification Agent

CementMetalVESNoneNone

Type of Container

Strong TightType A, Metal HIC

Strong TightType A, Poly HIC

Strong Tight

No. ofContainers

97

10 CFR 61WasteClass

ASAU

AU

Shipping Type

LSA II

LSA II

LSA II

LSA II (LSA Cask)LSA II

B. Irradiated Fuel Shipments (Disposition)

Number of Shipments

None

Mode of Transportation

N/A

Destination

N/A

-38-


XI. 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 thetotal body and each individual organ for the maximum exposed adult have beencalculated in accordance with the methodology in the ODCP;-Dose contributionslisted in Table 7 show conformance to RECP 6.1.4.1.

B. Radiation doses from radioactive gaseous effluents

The radiation dose contributions due to radioactive gaseous effluents at the siteboundary for the land sectors have been calculated in accordance with thecalculation meth'odology in the ODCP. Each unit's dose contribution has beencalculated separately. The latest five year historical average meteorologyconditions were used in these calculations. In addition to the site boundary doses,the dose to an individual (critical receptor) due to radioiodines, tritium, andparticulates released in gaseous effluents with half-lives greater than,8 days isdetermined in accordance with the methodology in the ODCP based on themethodology described in NUREG 0133. Dose contributions listed in Table 8,which represents the maximum dose for age groups, organs, and geographiclocations for the report period, show conformance to RECP 6.1.6.1, 6.1.7.1, and6.1.8.1.

C. Radiation Doses from Direct Radiation (Line-of-Sight Plus Sky-Shine)-Closest Site Boundary (800 m)

For the report period, the radiation dose is evaluated to be 8.72E-2 mrem due tothe presence of radioactive waste containers outside of plant buildings and thestorage of contaminated tools and equipment inside plant buildings.

D. Radiation Doses from Chemistry Laboratory Radioactive Gaseous Effluents-Closest Site Boundary (800m)

The radiation doses due to chemistry laboratory radioactive gaseous effluents forthe report period is evaluated to be 2.31E-6 mrem.

E. Radiation Doses from Post-accident Sampling System Radioactive GaseousEffluents - Closest Site Boundary (800m)

The radiation doses due to post-accident sampling system radioactive gaseouseffluents for the report period is evaluated to be 2.29E-6 mrem.

- 39-


F. 40CFR190 Considerations

The release of radioactivity in liquid and gaseous effluents during 1998 resulted indoses that are small percentages of the TS limits as shown in Tables 9 and 10.This, coupled with the fact that there are no other uranium fuel cycle sources withineight kilometers of the DCPP, shows conformance to 40CFR190.

G. Radiation Doses from Radioactive Liquid And Gaseous Effluents to Membersof 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 therelease of radioactive liquid effluents is negligible because the discharge piping forliquid radwaste is mostly imbedded in concrete, located in remote or inaccessibleareas, or is underground. In addition, the quantity of radioactivity released was verylow during 1998.


The radiation dose to members of the public within the site boundary due to therelease of radioactive gaseous effluents are listed in Table 11 for the year of 1998.

-40-


Table 7

Radiation Dose Due to the Release of Radioactive Liquid Effluents

INillirem

Total Body

Bone

Liver

Thyroid

Kidney

Lung

G.l. LLI

FirstQuarter5.86E-4

9.05E-4

1.07E-3

1.21E-4

7.67E-5

4.54E-4

7.85E-3

SecondQuarter1.28E-4

1.85E-4

2.07E-4

2.97E-5

3.77E-5

9.39E-5

1.61E-3

ThirdQuarter6.47E-5

1.71E-4

1.54E-4

4.48E-5

2.84E-5

8.20E-5

3.38E-4

FourthQuarter3.80E-4

1.09E-3

1.07E-3

1A8E-4

2.59E-4

4.04E-4

1.11E-3

AnnualTotal

1.16E-3

2.36E-3

2.51E-3

3.43E-4

4.02E-4

1.03E-3

1.09E-2

-41-

nclosurePG8 E Lette L 99-055

Table Sa

Radiation Dose Due to the Release of Radioactive Gaseous Effluents (Unit 1)

Site Boundary

First Second Third Fourth AnnualQuarter Dose Quarter Dose Quarter Dose Quarter Dose Total Dose

Noble GasGamma Air DoseBeta Air Dose

mradmrad

3.87E-51.89E-4

4.16E-54.02E-5

2.82E-53.00E-5

2.18E-57.80E-6

1.30E-42.67E-4

Nearest Residence - NNNP


I P TCritical Receptor (Highest

Organ)mrem 3.55E-4 2.05E-4 2.34E-4 1.97E-4 9.90E-4

Nearest Vegetable Garden-ESE"


I

PT'ritical Receptor (HighestOrgan)

mrem 2.94E-4 1.70E-4 1.94E-4 1.63E-4 8.22E-4

-42-

—nclosurePG8 E Lette L 99-055

Table Sb

Radiation Dose Due to the Release of Radioactive Gaseous Effluents (Unit 2)

Site Boundary



mradmrad

1.30E-31.30E-3

1.83E-56.85E-6

2.45E-51.12E-5

2.90E-48.12E-4

1.64E-32.13E-3

Nearest Residence -NNW'irst Second Third Fourth Annual

Quarter Dose Quarter Dose Quarter Dose Quarter Dose Total Dose

I P TCritical Receptor (HighestOrgan)

mrem 1.50E-3 2.41E-4 2.81E-4 2.01E-4 2.22E-3

Nearest Vegetable Garden-ESE'irst Second Third Fourth Annual

Quarter Dose Quarter Dose Quarter Dose Quarter Dose Total Dose—

I

PT'ritical Receptor (HighestOrgan)

mrem 1.17E-3 2.00E-4 2.33E-4 1.67E-4 1.77E-3

-43-


Notes for Tables Sa And Sb

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

'. Radioiodines, radioactive material in particulate form and radionuclides other than noble gases withhalf-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 occupancyfactor of 0.5 was used for the inhalation and ground plane pathways. The teen age group had thehighest calculated dose for this location.

-44-


Table 9

Percent of Technical Specification Limits for Radioactive Liquid Effluents

ORGAN FirstQuarter

SecondQuarter

Percent

ThirdQuarter

FourthQuarter

AnnualTotal

Total Body 3.91E-2 8.55E-3 4.31E-3 2.54E-2 3.87E-2

Bone

Liver

Thyroid

Kidney

Lung

G.l. LLI

1.81E-2

2.15E-2

2.42E-3

1.53E-3

9.07E-3

1.57E-1

3.71E-3

4.13E-3

5.93E-4

7.54E-4

1.88E-3

3.23E-2

3.41E-3

3.08E-3

8.96E-4

5.67E-4

1.64E-3

6.76E-3

2.19E-2

2.15E-2

2.95E-3

5.18E-3

8.08E-3

2.22E-2

2.36E-2

2.51E-2

3A3E-3

4.02E-3

1.03E-2

1.09E-1

NOTE:

" RECP 6.1.4.1.

-45-

nclosurePG8 E Lett L 99-055

Table 10a

Percent of Technical Specification Limits'for Radioactive Gaseous Effluents (Unit 1)

Site Boundary

First SecondQuarter % of Quarter% of

TS Limit TS Limit

Third FourthQuarter% of Quarter% of

TS Limit TS Limit

AnnualTotal % ofTS Limit


mradmrad

7.74E-41.89E-3

8.32E-44.02E-4

5.63E-43.00E-4

4.37E-47.80E-5

1.30E-31.34E-3

Nearest Residence - NNW

First SecondQuarter % of Quarter% of

TS Limit TS Limit


TS Limit TS Limit


I PTCritical Receptor (Highest

Organ)mrem 4.73E-3 2.73E-3 3.12E-3 2.63E-3 6.60E-3

Nearest Vegetable Garden - ESE

FirstQuarter % of

TS Limit

Second ThirdQuarter% of Quarter% of

TS Limit TS Limit

FourthQuarter % of

TS Limit.


I PTCritical Receptor (Highest

Organ)mrem 3.92E-3 2.27E-3 2.59E-3 2.18E-3 5.48E-3

NOTE:'ECP 6.1.6.1, 6.1.7.1, and

6.1.8.1

-46-


Table 10b

Percent of Technical Specification Limits'forRadioactive Gaseous Effluents (Unit 2)

Site Boundary

First Second Third Fourth AnnualQuarter% of Quarter% of Quarter% of Quarter% of Total % of

TS Limit TS Limit TS Limit TS Limit TS Limit


mradmrad

2.61E-21.30E-2

3.67E-46.85E-5

4.89E-41.12E-4

5.80E-38.12E-3

1.64E-21.07E-2

Nearest Residence - NNW

First Second ThirdQuarter% of Quarter% of Quarter% of

TS Limit TS Limit TS Limit

FourthQuarter % of

TS Limit


I PTCritical Receptor (Highest Organ) mrem 2.00E-2 3.21E-3 3.74E-3 2.68E-3 1.48E-2

Nearest Vegetable Garden - ESE

First SecondQuarter% of Quarter% of

TS Limit TS Limit


TS Limit TS Limit


IPTCritical Receptor (Highest Organ) mrem 1.56E-2 2.67E-3 3.11E-3 2.23E-3 1.18E-2

NOTE:'ECP 6.1.6.1, 6.1.7.1, and6.1.8.1

-47-


Table 11a

Radiation Dose Due to Release of Radioactive Gaseous EffluentsFirst Quarter, 1998

On-Site Dose to Members of the Public(Special Interest Groups)

External Dose (mrem)

Noble Gas

Internal Dose(m rem)

lodines, Particulate, and Tritium

Specific ActivityExposureLocation(Sectors)

Exposure ExposureClosest TimeDistance (Hours)

Whole Body Skin Ground Plane Inhalation

Police at ShootingRangeTour Participants

(a) Simulator Bldg.

(b) Bio Lab

(c) OverlookAmerican Indiansat Burial GroundsRanch hands drivingcattle around site

SE

S

SSE

NWNNWNW

NNWN

NNENE

700m

310m

460m

210m200m200m250m350m320m450m630m

52

1.0

1.5

252424.25.25.25.25.25

4.73E-5

6.20E-7

9.30E-7

4.31E-71.71E-41.20E-41.19E-64.58E-73.04E-71.19E-76.05E-S

8.33E-5

'.09E-6

1.64E-6

7.60E-73.02E-42.11E-42.11E-68.06E-75.36E-72.10E-71.07E-7

4.19E-5

2.91E-7

6.61E-7

9.50E-S4.33E-52.47E-53.18E-71.07E-75.40E-S

— 2.33E-S1.43E-S

3.44E-4

4.51E-6

6.76E-6

3.14E-61.24E-38.71E-48.69E-63.33E-62.21E-68.65E-74.40E-7

-48-


Table 11b

Radiation Dose Due to Release of Radioactive Gaseous EffluentsSecond Quarter, 1998



Noble Gas

Internal Dose(mrem)

lodines, Particulate, and TritiumSpecific Activity


(a) Simulator Bldg.

(b) Bio Lab


ExposureLocation(Sectors)

SE

SSE

NWNNWNW

NNWN

NNENE

ExposureClosestDistance

700m

310m

460m

210m200m200m250m350m320m450m630m

ExposureTime

(Hours)52

1.0

1.5

252424.25.25.25.25.25

Whole Body

2.13E-6

2.79E-S

4.18E-S

1.94E-S7.70E-65.39E-65.37E-82.06E-81.37E-85.34E-92.72E-9

Skin

3.76E-6

4.93E-S

7.39E-S

3.43E-S1.36E-59.53E-69.50E-S3.64E-82.42E-S9.45E-94.81E-9

Ground Plane

0.00E+0

0.00E+0

0.00E+0

0.00E+00.00E+00.00E+00.00E+00.00E+00.00E+00.00E+00.00E+0

Inhalation

1.34E-4

1.76E-6

2.64E-6

1.23E-64.87E-43.41E-43.40E-61.30E-68.64E-73.38E-71.72E-7

-49-


Table 11c

Radiation Dose Due to Release of Radioactive Gaseous EffluentsThird Quarter, 1998



Noble Gas

Internal Dose, (mrem)



(a) Simulator Bldg.

(b) Bio Lab



SE

SSE

NWNNWNW

NNWN

NNENE


700m

310m

460m

210m200m200m250m350m320m450m630m

ExposureTime

(Hours)52

1.0

1.5

252424.25.25.25.25.25

Whole Body

1.01E-5

1.32E-7

1.98E-7

9.19E-S3.65E-52.55E-52.55E-79.75E-S6.47E-S2.53E-S1.29E-8

Skin

2.26E-5

2.97E-7

4.45E-7.

2.07E-78.20E-55.74E-55.72E-72.19E-71.46E-75.69E-S2.90E-S

Ground Plane

0.00E+0

0.00E+0

0.00E+0

0.00E+00.00E+00.00E+00.00E+00.00E+00.00E+00.00E+00.00E+0

Inhalation

1.56E-4

2.04E-6

3.06E-6

1 42E-65.64E-43.94E-43.93E-61.51E-61.00E-63.91E-71.99E-7

-50-

nclosurePG&E Lett L 99-055

Table 11d

Radiation Dose Due to Release of Radioactive Gaseous EffluentsFourth Quarter, 1998

On-Site Dose to Members of the Public(Special interest Groups)


Noble Gas

Internal Dose(mrem)



(a) Simulator Bldg.

(b) Bio Lab



SE

SSE

NWNNWNW

NNWN

NNENE


700m

310m

460m

210m200m200m250m350m320m450m630m

ExposureTime

(Hours)52

1.0

1.5 *

252424.25.25.25.25.25

Whole Body

9.86E-6

1.29E-7

1.94E-7

9.00E-S3.57E-52.50E-52.49E-79.55E-S6.34E-S2.48E-S1.26E-S

Skin

2.20E-5

2.88E-7

4.32E-7

2.00E-77.95E-55.57E-55.55E-72.13E-71.41E-75.52E-S2.81E-S

Ground Plane

0.00E+0

0.00E+0

0.00E+0

0.00E+00.00E+00.00E+00.00E+00.00E+00.00E+00.00E+00.00E+0

Inhalation

1.20E-4

1.57E-6

2.36E-6

1.10E-64.35E-43.04E-43.03E-61.16E-67.72E-73.02E-71.54E-7

-51-


Table 11e

Radiation Dose Due to Release of Radioactive Gaseous EffluentsAnnual Total 1998

On-Site Dose to INembers of the Public(Special Interest Groups)

External Dose (mrem) Internal Dose(mrem)

Noble Gas lodines, Particulate, and TritiumSpecific Activity


(a) Simulator Bldg.

(b) Bio Lab



SE

S

SSE

NWNNWNW

NNWN

NNENE


700m

310m

. 460m

210m200m200m250m350m320m450m630m

ExposureTime

(Hours)208

4.0

6.0

1.0

96961.01.01.01.01.0

Whole Body

6.93E-5

9.09E-7

1.36E-6

6.33E-72.51E-41.76E-41.75E-66.71E-74.46E-71.74E-78.87E-8

Skin

1.32E-4

1.73E-6

2.59E-6

1.20E-64.77E-43.34E-43.33E-61.27E-68.46E-73.31E-71.68E-7

Ground Plane

4.19E-5

2.91E-7

6.61E-7

9.50E-84.33E-52.47E-53.18E-71.07E-75.40E-82.33E-81.43E-8

Inhalation

7.54E-4

9.89E-6

1.48E-5

6.88E-62.73E-31.91E-31.91E-57.30E-64.85E-61.90E-69.64E-7

-52-


Meteorological Data

The hour-by hour listing of wind speed, wind direction, atmospheric stability, andprecipitation is contained on disks with this submittal in accordance with TS 6.9.1.6.

-54-

Attachment 1

PG8 E Letter DCL 99-055

Corrections to 1997 AnnualRadioactive Effluent Release Report

(Total Error, % for Fission and ActivationGases changed from 23.5E+1 to 2.35E+1)

Attachment 1

PG&E Letter DCL 99-055

IX. Gaseous and Liquid Effluents

Table 1a


A. Fission and activation gases

1. Total release

Units

Ci

FirstQuarter

4.40E-2

SecondQuarter

2.01E+0

Est. TotalError, %

2.35E+1



pCi/sec 5.66E-3

5.12E-5

2.55E-1

1.52E-3

B. lodines

1. Total iodine-131



Ci MDA

pCi/sec MDA

MDA

MDA

MDA

MDA

2.35E+1

C. Particulate





Ci

pCi/sec

Ci

MDA

MDA

MDA

4.05E-7

. 2.79E-5

3.55E-6

1.38E-6

1.40E-7

2.35E+1

D. Tritium

1. Total release Ci 2.71E+1 4.97E+1 1.30E+1



pCi/sec 3.48E+0

7.79E-6

6.32E+0

1.41E-5

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit massor volume). This note applies to all tables.


Attachment 1

PG8 E Letter DCL 99-055

Table 1b


A. Fission 8 activation gases

Units Third FourthQuarter Quarter

Est. TotalError%

1. Total release Ci 3.08E-2 1.54E-1 2.35E+1



pCi/sec 3.88E-3

3.39E-5

1.93E-2

5.10E-5

B. lodines

1. Total iodine-131



Ci

ljCi/sec

MDA

MDA

MDA

MDA

MDA

MDA

2.35E+1

C. Particulate





Ci

pCi/sec

Ci

MDA

MDA

MDA

3.75E-7

MDA

MDA

MDA

6.43E-8

2.35E+1

D. Tritium

1. Total release



Ci 2.47E+1 3.68E+1

pCi/sec 3.10E+0 4.64E+0

6.94E-6 1.04E-5

1.30E+1

MDA = Less than the "a posteriori" minimum detectable activity (microcuries per unit massor volume). This note applies to all tables." RECP 6.1.6.1 Limit.

Attachment 2PG8 E Letter DCL 99-055

Radiological Monitoring And Controls Program (RMCP)

(CY2 Revision 2A)

~ ~* ISSUED FOR USE BKPACIFIC GAS AND ELECTRIC COMPANYNUCLEAR POWER GENERATIONPROGRAM DIRECTIVE

DATE: EXPIRES:NUMBERREVISION 2A

PAGE 1 OF 12

TITLE: RADIOLOGICALMONITORINGAND CONTROLSPROGRAM

EFFECTIVE DATE

SPONSORING ORGANIZATION: CHEMISTRYCLASSIFICATION: QUALITYRELATED

SECTION

PROGRAM OVERVIEW.APPLICABILITY.................DEFINITIONS.

TABLEOF CONTENTS

PAGE

~ ~ ~ ~ ~ ~ ~ 3

I

PROGRAM OBJECTIVES AND RE UIREMENTS.Program ObjectivesProgram Requirements .

Radiological Environmental Monitoring Program ...........Radioactive Effluent Control Prograin . ~ ..Offsite Dose Calculation Procedures .

Environmental Radiological Monitoring Procedure .

Radwaste Treatment Systems........~.........~................~....Quality Assurance Requirements.

RESPONSIBILITIES .....KEY IMPLEMENTINGDOCUMENTS.

Inter-Departmental Administrative Procedures (IDAPs) .

Department-Level Administrative Procedures (DLAPs)CLOSELY RELATED PROGRAMS .

RECORDS.....APPENDICES .

ATTACHMENTS.

REFERENCES .

PONSOR ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~S

3

. .. ... 4...4

.. 10

..... 10

......... 10

PROGRAM OVERVIEW

It is the policy of Nuclear Power Generation (NPG) that the release of radioactive materials to theenvironment be in compliance with Federal regulations and be "As Low As ReasonablyAchievable" (ALARA). The overall objectives are to protect the health and safety of the publicfrom undue radiation exposure and to minimize the amount of radioactive effluents resulting fromthe operation of the Diablo Canyon Power Plant.

This PD defines the overall policies and general requirements related to the RadiologicalMonitoring and Controls Program (RMCP). This includes the Radiological EnvironmentalMonitoring Program, and the Radioactive Effluent Controls Program.

01317002.DOE 1A 1 1 0324.0631

PACIFIC GAS AND ELECTRIC COMPANYNUCLEAR POWER GENERATION


NUMBER CY2REVISION 2APAGE 2 OF 12

The scope of this PD is focused on the control of releases of radioactive material to theenvironment, and minimizing radiological impact on the general public. Radiation protection ofplant workers and visitors within the restricted area of the plant is within the scope of RP1,"Radiation Protection."

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

FIGURE 1

CY2Hierarchy of Procedures

CY2

Radiological Monitoring and Controls Program

Radioactive Effluent Control ProgramEnvironmental Radiological Monitoring Procedure

DLAPs

Department Specific Administrative Controls

01317002.DOE IA 2 0324.063 t

PACIFIC GAS AND ELECTRIC COMPA~NUCLEAR POWER GENERATION



2. APPLICABILITY

This PD is applicable to all persons involved in radioactive effluent control, monitoring, and

management activities. This includes all NPG personnel, personnel matrixed to NPG from otherPG&E organizations, personnel in other PG&E organizations that are engaged in activities in

support of NPG, and contractor personnel that are working under NPG supervision.

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 limitsof 10 CFR 20 as is practical consistent with the purpose for which the licensed activityis undertaken,'aking into account the state of technology, the economics ofimprovements in relation to state of technology, the economics of improvements inrelation to benefits to the public health and safety, and in relation to utilization ofnuclear energy and licensed materials in the public interest. The specific objectives ofachieving ALARAeffluents are based on those described in 10 CFR 50, Appendix I.

The Radiological Monitoring and Controls Program (RMCP) - Contains the RadioactiveEffluent Controls and Radiological Environmental Monitoring Programs required byTechnical Specification 6.8.4 and descriptions of the information that should be

included in the Annual Radiological Environmental Operating and Annual RadioactiveEffluent Release Reports required by Technical Specifications 6.9.1.5 and 6.9.1.6.

3.3 Offsite Dose Calculation Procedure (ODCP) - Contains the methodology and

parameters used in the calculation of offsite doses due to radioactive gaseous and liquideffluents and in the calcuhtion of gaseous and liquid effluent monitoring Alarm/TripSetpoints.

3.4 Environmental Radiological Monitoring Procedure (ERMP) - Contains a description ofsample locations, types of sample locations, methods and frequency of analysis, and

reporting requirements.

PROGRAM OBJECTIVES AND UIREMENTS

4.1 Program Objectives

The NPG 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 applicableregulatory 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 establishadministrative controls with clear delineation of responsibilities to ensure that

monitoring, measurement, and release activities are properly sequenced,

authorized, and controlled.

01317002.DOE 1A 3 I 0324.0631

PACIFIC GAS AND ELECTRIC COMPANYNUCLEARPOWER GENERATION



4.2 Program Requirements

The basic requirement for the Radiological Monitoring and Controls Program shall beto maintain radioactive releases to the unrestricted areas surrounding the plant inconformance with applicable Federal regulations and ALARA. The following

sections'rovide

additional requirements for various elements of the program.

4.2.1 Changes to the RMCP (including ODCP, ERMP and RECP) shall beprocessed in accordance with the requirements of DCPP TechnicalSpecification Section 6. 14.2.

4.2.2

4.2.3

Radiological Environmental Monitoring Program

a. A Radiological Environmental Monitoring Program (REMP) shall beestablished and maintained to comply with the DCPP TechnicalSpecification 6.8.4.h, Radiological Environmental Monitoring Programrequirements. The program shall be provided to monitor the radiationand radionuclides in the environs of the plant, and shall address thefollowing:

1. Monitoring, sampling, analysis, and reporting of radiation andradionuclides in the environment in accordance with themethodology and parameters in the Environmental RadiologicalMonitoring Procedure (ERMP),

2. A Land Use Census to ensure that changes in the use of areas atand beyond the SITE BOUNDARYare identified and thatmodifications to the monitoring program are made ifrequired bythe results of this census, and

3. Participation in an Interlaboratory Comparison Program to ensurethat independent checks on the precision and accuracy of themeasurements of radioactive materials in the environmentalsample matrices are performed as part of the quality assurance

program for environmental monitoring.

Radioactive Effluent Control Program

a. Monitoring requirements shall be established and maintained for allmajor and potentially significant paths for release of radioactivematerial during normal plant operation, including anticipatedoperational occurrences, to comply with Regulatory Guide 1.21,Revision 1, June 1974, requirements.

Ol317002.DOE 1A 4 I 0324.0631




b. Procedures shall be established and maintained to define the methodsand requirements for control of liquid and gaseous radioactivedischarges within the limits of DCPP Technical SpecificationSections 6.8.4.g. These procedures shall address the following:

1. Limitations on the operability of radioactive liquid and gaseousmonitoring instrumentation including surveillance requirementsand setpoint determination in accordance with methodology in theOff-site Dose Calculation Procedure, (ODCP).

2. Limitations on the concentrations of radioactive material releasedin liquid effluents to UNRESTRICTED AREAS conforming to10 CFR Part 20, Appendix B, Table 2, Column 2,

Monitoring, sampling, and analysis of radioactive liquid andgaseous effluents in accordance with 10 CFR 20.1302 and withthe methodology and parameters in the ODCP,

i

4.

5.

6.

Limitations on the annual and quarterly doses or dose commitmentto a MEMBER OF THE PUBLIC from radioactive materials inliquid effluents released from each unit to UNRESTRICTEDAREAS conforming to Appendix I to'10 CFR Part 50,

Determination of cumulative and projected dose contributionsfrom radioactive effluents for the current calendar quarter andcurrent calendar year in accordance with the methodology andparameters in the ODCP at least every 31 days,

Limitations on the operability and use of the liquid and gaseouseffluent treatment systems to ensure that the appropriate portionsof these systems are used to reduce releases of radioactivity whenthe projected doses in a 31-day period would exceed 2 percent ofthe guidelines for the annual dose or dose commitmentconforming to Appendix 1 to 10 CFR Part 50,

7. Limitations on the dose rate resulting from radioactive materialreleased in gaseous effluents from the site to areas at or beyondthe SITE BOUNDARYshall be limited to the following:

a)

b)

For noble gases: Less than or equal to a dose rate of500 mrem/yr to the whole body and less than or equal to a

dose rate of 3000 mrem/yr to the skin, and

For Iodine-131, for Iodine-133, for tritium, and for allradionuclides in particulate form with half-lives greater than8 days: Less than or equal to a dose rate of 1500 mrem/yrto any organ.

0) 317002.DOE iA 5 0324.0631




C.

d.

8. Limitations on the annual and quarterly air doses resulting fromnoble gases released in gaseous effluents from each unit to areas

beyond the SITE BOUNDARYconforming to Appendix I to10 CFR Part 50,

9. Limitations on the annual and quarterly doses to MEMBERS OFTHE PUBLIC from Iodine-131, Iodine-133, tritium, and allradionuclides,in particulate form with half-lives greater than 8

days in gaseous effluents released from each unit to areas beyondthe SITE BOUNDARYconforming to Appendix I to 10 CFR Part50, and

10. Limitations on the annual dose or dose commitment to anyMEMBER OF THE PUBLIC due to releases of radioactivity andto radiation from uranium fuel cycle sources conforming to 40CFR Part 190.

Sampling and analysis methods associated with effluent monitoringactivities shall be controlled in accordance with CY1.DC4, " Control ofMaterial and Equipment Used For Analysis for Chemistry andRadiochemistry Program."

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

An onsite meteorological program shall be established and maintainedin accordance with the requirements of Regulatory Guide 1.23,February 1972, to provide sufficient data for the performance of doseassessments.

The collection and processing of technical data required to support theAnnual Radioactive Effluent Release Report and non-routine reports tothe NRC to comply with DCPP Technical SpecificationsSections 6.9.1.6 shall be defined as part of this program. Theprocessing of these reports shall be performed in accordance withXI1, "Regulatory Interface."

4.2.4 Offsite Dose Calculation Procedures

a. Offsite Dose Calculation Procedures (ODCP) shall be established andmaintained to define and control the methods for determining offsitedoses. 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 shalladdress the following:

1. Methods for determining monitoring instrumentation alarmsetpoints are addressed in accordance with a Department-LevelAdministrative Procedure (DLAP) under CY2.

013i 7002.DOE lA 6 0324.0631




4.2.5

4.2.6

4.2.7

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 CF4, "Modification Control."

Environmental Radiological Monitoring Procedure

a. An Environmental Radiological Monitoring Procedure (ERMP) shall be

established and shall contain a description of sample locations, types ofsample locations, methods and frequency of analysis, and reportingrequirements.

Radwaste Treatment Systems

a. Radwaste Treatment Systems shall be provided to control the processingand release of radioactive materials in gaseous and liquid effluent incompliance with Technical Specification requirements. The design ofthese systems shall be controlled in accordance with CF3, "DesignControl," and the requirements of Regulatory Guide 1. 143, October1979.

b. Approval of changes to the Radwaste Treatment Systems shall beprocessed in accordance with the requirements of CF4, "ModificationControl."

Quality Assurance Requirements

In addition to requirements specified in earlier sections and thoserequirements utilizing procedures in the section 6.2 of this PD, the ControlProgram shall be subject to the Quality Assurance requirements specified inCY1, "Chemistry/Radiochemistry."

RESPONSIBILITIES

5.1 'he Senior Vice President and General Mana er NPG - is responsible for establishingthe policy and general requirements for the Radiological Monitoring and ControlsProgram, for providing management support and guidance for the program'simplementation, and ensuring compliance with all regulatory requirements is

maintained.

5.2 The Vice President and Plant Mana'r P&PM - is responsible for the overalldevelopment, implementation, and maintenance of the Radiological Monitoring and

Controls Program in accordance with the requirements of this PD.

01317002.DOE IA 7 I 0324.0631

PACIFIC GAS ANDELECTRIC COMPANYNUCLEAR POWER GENERATION



5.3

5.4

5.5

5.6

5.7

The Vice President Nuclear Technical Services NPG - is responsible for ensuring thatsupport from reporting departments is provided for the Radiological Monitoring andControls Program.

The Mana er - DCPP erations Services - is responsible for the direct implementationof the Radiological Monitoring and Controls Program with the exception of the designof radwaste treatment and effluent monitoring systems.

The Mana er - En ineerin Services and the Mana er Re lato and Desi n ServicesQ<&D~S - are responsible for maintaining the design bases for installed plant radwastetreatment and effluent monitoring systems, structures, and components and providingtechnical support to the plant for the operation and maintenance of these systems.

The Mana er - Technical and Ecolo ical Services TES - is responsible for performingradiological laboratory analysis for the Radiological Environmental MonitoringProgram and preparing and reviewing the Annual Radiological EnvironmentalOperating Report.

Other PG&E De artments - called upon to support NPG activities associated with theRadioactive Effluent Control program are responsible for performing their activities inaccordance with the requirements of this PD.

5.8

5.9

The Mana er - Nuclear ali Services N S - is responsible for auditing theRadiological Monitoring and Controls Program at least once every 12 months.

The Mana er - Maintenance Services - is responsible for maintaining the radiationmonitoring systems and the hardware and software for the Rad Effluent program.

KEY IMPLEMENTINGDOCUMENTS

6.1

6.2

Inter-Departmental Administrativ'e Procedures (IDAPs)

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

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

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

Department-Level Administrative Procedures (DLAPs)

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

Ol317002.DOE 1A 8 0324.0631




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"

7.1.2

"Records" provides for the retention of Radiological Monitoring and ControlsProgram records.

CF3, "Design Control"

"Design Control" addresses the implementation of design activities forinstalled radwaste treatment and effluent monitoring systems in accordancewith the requirements of NRC Regulatory Guide 1.143.

7.1.3 CF4, "ModificationControl"

"Modification Control" addresses the implementation of modificationactivities for installed effluent monitoring systems.

7.1.4 CY1, "Chemistry/Radiochemistry"

7.1.5

7.1.6

"Chemistry/Radiochemistry" addresses the methods forchemistry/radiochemistry sampling and analysis of liquid and gaseousradioactive effluents in support of this PD.

OM7, "Problem Resolution"

"Problem Resolution" addresses deficiencies identified during theimplementation of the. radioactive effluent control program. OM7 alsoaddresses evaluating nonconformances for reportability in accordance withTechnical

Specifications.'Q1,

"Personnel Training and Qualification"

"Personnel Training and Qualification" identifies training and qualificationrequirements for personnel.

7.1.7 XI1, "Regulatory Interface"

"Regulatory Interface" addresses the process for required reporting andcommunication with outside agencies.

7.1.8 CY2.ID1, "Radioactive Effluent Controls Program"

l

8. RECORDS

None

"Radioactive Effluent Controls Program" contains the general programrequirements to ensure the requirements of 10 CFR Part 20 and10 CFR Part 50, Appendix I, are met.

Oi317002.DOE 1A 9 0324.0631

PACIFIC GAS ANDELECTRIC COMPANYNUCLEARPOWER GENERATION


NUIHBER CY2REVISION 2APAGE 10 OF 12

9. APPENDICES

9.1 Graded Quality Assurance Requirements for Radiological Monitoring and ControlsProgram

10. ATTACHMENTS

None

11. REFERENCES

11.1 Title 10, Code of Federal Regulations,

11.1.1 Part 20, ."Standards for Protection Against Radiation"

11.2

11.3

11.4

11.5

11.6

11.7

11.1.2 Part 50, Appendix I,

11.1.3 Part 50, Appendix A, GDC 60, 64,

Title 40, Code of Federal Regulations, "Environmental Radiation Protection Standardsfor Nuclear Power Operations."

Regulatory Guide 1.109, Revision 1, October 1977, "Calculation of Annual Doses to, Man From Routine Releases of Reactor Effluents for the Purpose of EvaluatingCompliance with 10 CFR Part 50 Appendix I."

Regulatory Guide 1.143, October 1979, "Design Guidance for Radioactive WasteManagement Systems, Structures, and Components Installed in Light-Water-CooledNuclear Power Plants."

Regulatory Guide'1.21, Revision 1, June 1974, "Measuring, Evaluating, and ReportingRadioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid andGaseous Effluent from Light-Water-Cooled Nuclear Power Plants."

Regulatory„Guide 1.23, February 1972, "Onsite Meteorological Programs."

Regulatory Guide 4.15, Revision 1, February 1979, "Quality Assurance ForRadiological Monitoring Programs (Normal Operations) - Effluent Streams and theEnvironment."

11.8 Regulatory Guide 4.1, Revision 1, April 1975, "Programs for Monitoring Radioactivityin the Environs of Nuclear Power Plants."

11.9 Diablo Canyon Nuclear Power Plant Facility Operating Licenses (Unit 1, Unit 2)

12. SPONSOR

David Chen

01317002.DOE 1A 10 1 0324.0631


TITLE; RADIOLOGICALMONITORINGAND CONTROLSPROGRAM


APPENDIX 9.1

GRADED QA REQUIREMENTSFOR RADIOLOGICALMONITORINGAND CONTROLS PROGRAM

The basis for these Graded QA requirements is to comply w'ith the regulations of 10 CFR 20, 10 CFR 50,

40 CFR 190, the Technical Specifications and Regulatory Guides 1.21; and 4:15.

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.

GRADED ACTIVITIES

3.

2.3

2.1 Installed radiation monitors required per the Technical Specifications 3/4.3.3.6, and

6.8.4.b for monitoring radioactive effluents during plant operations or accidents shall be

calibrated at prescribed intervals.

2.2 Sampling and analysis of liquid and gaseous effluents shall be performed in accordance

with CY1, "Chemistry/Radiochemistry."

Calculations, computer programs, and procedures for evaluating the dose associated

with radioactive effluents shall be performed in accordance with approved quality

related procedures.

GRADED RE UIREMENTS

3.1

3.2

3.3

3.4

3.5

Effluent releases shall be maintained ALARAand shall be performed in accordance

with the requirements of this Program Directive (CY2) to limit the concentrations,

doses and doserates as specified in DCPP Technical Specifications 6.8.4,NRC regulations 10 CFR 20,.10 CFR 50 Appendix I, and EPA regulation 40 CFR 190.

The Annual Radiological Environmental Operating Report shall be developed in

accordance with Technical Specification 6.9.1.5.

The Annual Radioactive Effluent Release Report shall be developed in accordance withTechnical Specification 6.9.1.6.

Records that support and document radioactive effluent releases and dose evaluations

shall be controlled in accordance with AD10, "Records."

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."

Ol317002.DOE iA ii i 0324.063l




3.6

3.7

3.8

3.9

APPENDIX 9.1 Continued

NotiTications and reports 'to and correspondence with regulatory agencies shall be donein accordance with XI1, "Regulatory Interface."

Written plans, procedures and instructions for implementing Radiological Monitoringand Controls Program shall be prepared, processed, and controlled in accordance withAD1, "Administrative Control Program."

Procurement of quality-related equipment or services shall be in accordance withwritten procedures. Applicable regulatory requirements, design bases, and any otherrequirements necessary to assure adequate quality shall be included in or invoked byreference in documents for procurement of items or services. Test or acceptancerequirements and documentation to be submitted by the supplier shall be identified inthe procurement documents. Receipt inspection requirements, ifrequired, shall beidentified in the procurement documents.

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

3.10 Periodic assessments shall be performed at least annually to review the content andimplementation of the Radiological Monitoring and Controls Program.

013 l7002.DOE iA 12 i 0324.063 I

fC

0

0

Attachment 3PG&E Letter DCL 99-055

Radioactive Effluent Control Program (RECP)

(CY2.ID1 Revision 3A)

PACIFIC GAS AND ELECTRIC COMPANYNUCLEARPOWER GENERATIONINTER-DEPARTMENTALADMINISTRATIVEPROCEDURE

I'UMBERCY2.ID1REVISION 3A

PAGE 1 OF 37

TITLE: RADIOACTIVEEFFLUENT CONTROLS PROGRAM

03/24/98EFFECTIVE DATE

SPONSORING ORGANIZATION: CHEMISTRYPROCEDURE CLASSIFICATION: QUALITYRELATED

REVIEW LEVEL: "A"

SECTION

SCOPE .......................DISCUSSION.RESPONSIBILITIESINSTRUCTIONSRECORDSAPPENDICES .

REFERENCESSPONSOR.

1. SCOPE

TABLEOF CONTENTS

PAGE

~ ~ ~ ~ ~ ~ 0 ~ 23

....... 6

....... 6

....... 6.6

This procedure contains the general program requirements of the Radioactive EffluentControls Program. This program ensures that the requirements of 10 CFR Part 20 and10 CFR Part 50 Appendix I are met.

DISCUSSION

2.1

2.2

This procedure provides the general requirements for Radioactive Effluent ControlsProgram in accordance with the Technical Specifications and the implementationGeneric Letter 89-01, "Implementation of Programmatic Controls for RadiologicalEffluent Technical Specifications in the Administrative Controls Section of theTechnical Specifications and the Relocation of Procedural Details of RETS to theOff-Site Dose Calculation Manual or to the Process Control Program."

The following Technical Specification definitions are applicable.

2.2.1 The OFF-SITE DOSE CALCULATIONPROCEDURE (ODCP) shallcontain the methodology and parameters used in the calculation of off-sitedoses resulting from radioactive gaseous and liquid effluents, in thecalculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints.

01192903. DOC 1A 1

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' UNCONTROLLEDPROCEDURE-DO NOTUSE TO'PERFORN NORKorISSUEFOR USE-,'ACIFICGAS ANDELECTRIC COMPANY

NUCLEAR POWER GENERATION


NUMBER CY2.IDlREVISION 3APAGE 2 OF 37

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 gaseousand 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 inthis procedure by reference. Therefore, the requirements for revisions to thisprocedure also apply to CAP A-8.

2.3.1 The procedural requirements that ensure compliance with TechnicalSpecification 6.8.4.g are contained in CAP A-5, CAP A-6, and CAP A-8,"Off-Site Dose Calculation Procedure (ODCP)."

RESPONSIBILITIES

3.1

3.2

3.3

The Director, Chemistry is responsible for:

3.1.1 Implementation'of the Off-Site Dose Calculation Procedure in a manner thatmeets regulatory requirements and preparing the Annual RadiologicalEffluent Release Report.

3.1.2 Providing direction to the Operations Staff in the processing of radioactivewaste streams.

3.1.3 Ensuring preparation, review and approval of the Nonroutine RadiologicalEnvironmental Operating Report when required by Appendix 6.1 ~ 11.1. TESnotifies DCPP in accordance with 3.5.

3.1.4 Ensuring that a comparison of the Annual Radioactive Effluent ReleaseReport and the Annual Radiological Environmental Operating Report isperformed.

3.1.5 Ensuring that dose commitment increases due to the Land Use Census inaccordance with Appendix 6.1. 12.1 are determined and communicatedpromptly to RP.

The Director, Radiation Protection is responsible for:

3.2.1 Ensuring the performance of the annual land use census and that the resultsare provided to Chemistry so that Chemistry can established the doserequirements of Appendix 6.1.12.1.

3.2.2 Ensuring that the results of the annual Land Use Census are provided to TESfor inclusion in the Annual Radiological Environmental Operating Report.

3.2.3 Ensuring that changes to the Environmental Radiological MonitoringProcedure are provided to Chemistry for inclusion in the Annual RadiologicalEffluent Release Report.

The Director, Chemical and Environmental, TES is responsible for ensuring that REMPsample results exceeding the criteria of Appendix 6.1.11.1 are communicated promptlyto the Director, Chemistry and the Director, RP at DCPP.

01192903.DOC 1A 2

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...'.. UNCONTROElEDPROCEDURE=-DO NOT'VSE TOPERFORM O'ORKorISSVEFOR VSE ":,



NUMBER CY2.ID1REVISION 3APAGE 3 OF 37

INSTRUCTIONS

4.1 Administrative Requirements

4.1.1 Appendix 6.1 of this procedure contains the operational requirements of theRadioactive Effluent Controls Program.

4.1.2 The operational requirements are implemented by equipment controlguidelines (reference OP1.DC16), CAP A-8, and XI1.ID2, "RegulatoryReporting Requirements and Reporting Process."

a. The Equipment Control Guidelines implement those requirements thatare related to equipment and have specific allowed outage times oroperator actions.

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

c. XI1.ID2 implements the reporting requirements.

4.2 Reporting Requirements

4.2.1 Annual Radioactive Effluent Release Report

a. Report Schedule

1. Annual Radioactive Effluent Release Reports covering theoperation of the unit during the previous calendar year shall besubmitted before May 1 of each year.

b. The Annual Radioactive Effluent Release Reports shall include:

1. A summary of the quantities of radioactive liquid and gaseouseffluents and solid waste released from the unit as outlined inRegulatory Guide 1.21, "Measuring, Evaluating, and ReportingRadioactivity in Solid Wastes and Releases of RadioactiveMaterials in Liquid and Gaseous Effluents fromLight-Water-Cooled Nuclear Power Plants," Revision 1, June1974, with data summarized on a quarterly basis following theformat of Appendix B thereof. For solid wastes, the format forTable 3 in Appendix B shall be supplemented with three additionalcategories; class of solid wastes (as defined by 10 CFR Part 61),type of container (e.g., LSA, Type A, Type B, Large Quantity)and SOLIDIFICATIONagent or absorbent (e.g., cement, ureaformaldehyde);

01192903. DOC lA 3

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TITLE: RADIOACTIVEEFFLUENT CONTROLS PROGRAMI


2. A list and description of unplanned releases from the site toUNRESTRICTED AREAS of radioactive materials in gaseousand liquid effluents made during the reporting period;

3. Any changes made during the reporting period to the PCP,RMCP, ERMP, and ODCP, pursuant to Specifications 6.13 and6. 14, respectively.

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

A listing of new locations for dose calculations and/orenvironmental monitoring identified by the Land Use Censuspursuant to Appendix 6.1.

An explanation as to why the inoperability of liquid or'gaseouseffluent monitoring instrumentation was not corrected within thetime specified in Appendix 6.1; and

Description of the events leading to liquid holdup tanks or gasstorage tanks exceeding the limits of Technical Specifications3.11.1.4 or 3.11.2.6.

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

A summary of the evaluation that led to the determination that thechange could be made in accordance with 10 CFR 50.59;

Sufficient detailed information to totally support the reason for thechange without benefit of additional or supplemental information;

A detailed description of the equipment, components andprocesses involved and the interfaces with other plant systems;

An evaluation of the change which shows the predicted releases ofradioactive materials in liquid and gaseous effluents and/orquantity of solid waste that differ from those previously predictedin the License application and amendments thereto;

An evaluation of the change which shows the expected maximumexposures to a MEMBER OF THE PUBLIC in theUNRESTRICTED AREA and to the general population that differfrom those previously estimated in the License application andamendments thereto;

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

01192903.DOC IA 4

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PACIFIC GAS AND ELECTRIC COMPANYNUCLEARPOVVER GENERATION



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

result of the change; and

8. Documentation of the fact that the change was reviewed and foundacceptable by the PSRC.

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

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

1. An annual summary of hourly meteorological data collected overthe previous year. This annual summary may be either in theform of an hour-by-hour listing on magnetic tape/hard disk orother media of wind speed, wind direction, atmospheric stability, =

and precipitation (ifmeasured), or in the form ofjoint frequencydistributions of wind speed, wind direction, and atmosphericstability or the licensee has the option of retaining this summary ofrequired meteorological data on site in a file that shall be providedto the NRC upon request;

2. An assessment of the radiation doses due to the radioactive liquidand gaseous effluents released from the unit or station during theprevious calendar year;

3. An assessment of the radiation doses from radioactive liquid andgaseous effluents to MEMBERS OF THE PUBLIC due to theiractivities inside the SITE BOUNDARY (see TechnicalSpecification Figure 5.1-3) during the report period;

4. - Allassumptions used in making these assessments, i.e., specificactivity, exposure time and location. The meteorologicalconditions concurrent with the time of release of radioactivematerials in gaseous effluents, as determined by samplingfrequency and measurement, shall be used for determining thegaseous pathway doses. The assessment of radiation doses shallbe performed in accordance with the methodology and parametersin the OFF-SITE DOSE CALCULATIONPROCEDURE(ODCP); and

5..An assessment of radiation doses to the likely most exposedMEMBER OF THE PUBLIC from reactor releases and othernearby uranium fuel cycle sources, including doses from primaryeffluent pathways and direct radiation, for the previous calendaryear to show conformance with 40 CFR Part 190, "EnvironmentalRadiation Protection Standards for Nuclear Power Operation."Acceptable methods for calculating the dose contribution fromliquid and gaseous effluents are given in Regulatory Guide 1.109,Rev. 1, October 1977.

01192903.DOC 1A 5

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4.3

4.4

e. A single submittal may be made for a multiple unit plant. The submittalshould combine those sections that are common to all units at the plant;however, for units with separate radwaste system, the submittal shallspecify the releases of radioactive material from each unit.

Revisions to the RECP

4.3.1 The requirements for revision to the RECP also apply to CAP A-8.

4.3.2 The requirements are provided in Technical Specification 6.14.2.

Major changes to Liquid, Gaseous, and Solid Radwaste Treatment Systems

4.4.1 Major changes to the liquid, gaseous, and solid radwaste treatment systemsshall become effective upon review and acceptance by the PSRC provided thechange could be made in accordance with 10 CFR 50.59.

5. RECORDS

5.1 Data Sheets and records willbe maintained in the Records Management System (RMS)in accordance with CY1.DC1, "Analytical Data Processing Responsibilities."

APPENDICES

6.1 Operational Requirements of the Radioactive Effluent Controls Program

REFERENCES

7.1 CAP A-8, "Off-site Dose Calculation Procedure (ODCP)."

7.2 CAP A-5, "Liquid Radwaste Discharge Management."

7.3 CAP A-6, "Gaseous Radwaste Discharge Management."

7.4 RP1.ID11, "Environmental Radiological Monitoring Procedure."

7.5 OP1.DC16, "Control of Plant Equipment Not Required by the TechnicalSpecifications."

7.6 XI1.ID2, "Regulatory Reporting Requirements and Reporting Process."

7.7 Regulatory Guide 1.21, Revision 1, June 1974.

7.8 Regulatory Guide 1.109, Revision 1, October 1977.

7.9 License Amendment Request 93-04.

SPONSOR

David Chen

0l 192903.DOC lA 6

' UilfCONTROI.IEIJ:PROCEOURE-OO NOMSE TOPERFORM'kYORKorISSUEFOR 'USE,,"PACIFIC GAS ANDELECTRIC COMPANYNUCLEARPOWER GENERATION

TITLE: RADIOACTIVEEFFLUENT CONTROLS PROGRAMl

NUMBER CY2.IDIREVISION 3APAGE 7 OF 37

APPENDIX 6.1OPERATIONALREQUIREMENTS OF THE RADIOACTIVEEFFLUENT

CONTROLSPROGRAM

6.1.1 Radioactive Li uid Effluent Monitorin Instrumentation (Also covered by ECG 39.3)

Commitment for 0 eration

6.1.1.1 The radioactive liquid effluent monitoring instrumentation channels shown inTable 6.1.1-1 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure thatthe limits of Commitment 6.1.3.1 are not exceeded. The Alarm/Trip Setpoints of thesechannels shall be determined in accordance with the methodology and parameters in theOFF-SITE DOSE CALCULATIONPROCEDURE (ODCP).

~Aticebitit: At eii times.

Action:

With a radioactive liquid effluent monitoring instrumentation channelAlarm/Trip Setpoint less conservative than required by the aboveCommitment, immediately suspend the release of radioactive liquid effluentsmonitored by the affected channel or declare the channel inoperable.

With less than the minimum number of radioactive liquid effluent monitoringinstrumentation channels OPERABLE, take the ACTION shown inTable 6.1.1-1. Restore the inoperable instrumentation to OPERABLE statuswithin the time specified in the ACTION, or explain in the next AnnualRadioactive Effluent Release Report why this inoperability was not correctedwithin the time specified.

Surveillance Re uirements

6.1.1.2 Each radioactive liquid effluent monitoring instrumentation channel shall bedemonstrated OPERABLE by performance of the CHANNELCHECK, SOURCECHECK, CHANNELCALIBRATIONand CHANNELFUNCTIONALTEST at thefrequencies shown in Table 6.1.1-2.

6.1.1.3 At least one saltwater pump shall be determined operating and providing dilution to thedischarge structure at least once per 4 hours whenever dilution is required to meet thelimits of Commitment 6.1.3.1.

0l 192903.DOC IA 7

"'NCONITROllEDPROCEDVRE-DONOTUSE TO PERFORM OORKor ISSVEFOR USE "PACIFIC GAS AND ELECTRIC COMPANYNUCLEAR POWER GENERATION




TABLE6.1.1-1RADIOACTIVELI UID EFFLUENT MONITORINGINSTRUMENTATION

INSTRUMENT

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

b. Steam Generator Blowdown Tank (RM-23)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line (FR-20)¹

b. Steam Generator Blowdown Effluent

Lines (FR-53)

Oily Water Separator EffluentLine (FR-251)¹

adioactivity Monitor Not Providing AutomaticTermination of Release

c,Oily Water Separator Effluent Line (RM-3)¹

1. Radioactivity Monitors Providing Alarm and AutomaticTermination of Release

MINIMUMCHANNELS'OPERABLE ACTION

¹ This Radioactive Liquid Effluent Monitoring Instrumentation is common to both units.

01192903.DOC 1A 8

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ACTION 1

ACTION 2

ACTION 3

ACTION4


TABLE6.1.1-1 ContinuedACTION STATEMENTS

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 14

days provided that prior to initiating a release:

At least two independent samples are analyzed in accordance withCommitment 6.1.3.2.

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

Otherwise, suspend release of radioactive effluents via this pathway.

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 30days provided grab samples are analyzed for radioactivity (beta or gamma) at a lowerlimit of detection of no more than 107 microcuries/ml:

At least once per 12 hours when the specific activity of the secondary coolantis greater than 0.01 microcuries/gram DOSE EQUIVALENTI-131, or

b. At least once per 24 hours when the specific activity of the secondary coolantis less than or equal to 0.01 microcuries/gram DOSE EQUIVALENTI-131.

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 30days provided that, at least once per 12 hours, grab samples are collected and analyzedfor radioactivity (beta or gamma) at a lower limitof detection of no more than

10'icrocuries/mlor transfer the oily water separator effluent to the Liquid RadwasteTreatment System.

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 30days provided the flow rate is estimated at least once per 4 hours during actual releases.Pump performance curves may be used to estimate flow.

01192903.DOC 1A 9

VNCONTROI.I.EDPROCEDURE- DONPACIFI AND ELECTRIC COMPANYNUCLEARPOWER GENERATION


E TO PERFORM O'ORKorISSUE FOR USE':.=':NUMBER CYREVISION 3APAGE 10 OF 37

APPENDIX 6.1. Continued

TABLE6.1.1-2RADIOACTIVELI UID EFFLUENT MONITORINGINSTRUMENTATIONSURVEILLANCERE UIREMENTS

INSTRUMENT1. Radioactivity Monitors Providing Alarm and Automatic Termination

of Release

a. Liquid Radwaste Effluent Line (RM-18)

b. Steam Generator Blowdown Tank (RM-23)

2. Flow Rate Measurement Devices

a. Liquid Radwaste Effluent Line (FR-20)

b. Steam Generator Blowdown Effluent Line (FR-53)

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

3. Radioactivity Monitor Not Providing Automatic Termination ofRelease

Oily Water Separator Effluent Line (RM-3)

CHANNELCHECK

D

D

D(4)

D(4)

Daily(4)

D

SOURCECHECK

N.A.

N.A.

N.A.

CHANNELCALIBRATION

R(3)

R(3)

R(3)

CHANNELFUNCTIONAL

TEST

Q(1)

Q(1)

Q

Q

Q

Q(2)

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(2)

(y (3)

(4)

(5)


TABLE6.1.1-2TABLENOTATION

The CHANNELFUNCTIONALTEST shall also demonstrate that automatic isolation of thispathway and Control Room alarm annunciation occurs ifany 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), orc. Instrument indicates a downscale failure (alarm only), ord. Instrument controls not set in operate mode (alarm only).

The CHANNELFUNCTIONALTEST shall also demonstrate that Control Room alarmannunciation occurs ifany of the following conditions exist:

a. Instrument indicates measured levels above the Alarm Setpoint, orb. Circuit failure, orc. Instrument indicates a downscale failure, ord. Instrument controls not set in operate mode.

The initial CHANNELCALIBRATIONshall be performed using one or more of the referencestandards certified by the National Bureau of Standards (NBS) or using standards that have beenobtained from suppliers that participate in measurement assurance activities with NBS. Thesestandards shall permit calibrating the system over its intended range of energy and measurementrange. For subsequent CHANNELCALIBRATION,sources that have been related to the initialcalibration shall be used.

CHANNELCHECK shall consist of verifying indication of flow during periods of release.CHANNELCHECK for FR-251 shall be made once per calendar day*, and for FR-20 and FR-53shall be made at least once per 24 hours on days on which continuous, periodic, or batch releasesare made.

Fre uenc Notation

NotationDDailyMQRpN.A.

FrequencyAt least once per 24 hoursAt lease once per calendar day*At least once per 31 daysAt least once per 92 daysAt least once per 18 monthsCompleted prior to each releaseNot Applicable

* The frequency "once per calendar day" could result in two successive channel checks nearly 48 hoursapart over a two day period. This frequency is different from and should not be confused with thefrequency notation "D" (at least once per 24 hours) defined in Technical Specifications.

01192903.DOC lA 11

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6.1.2 Radioactive Gaseous Effluent Monitorin Instrumentation (Also covered by ECG 39.4)


6.1.2.1 The radioactive gaseous effluent monitoring instrumentation channels shown inTable 6.1.2-1 shall be OPERABLE with their.Alarm/Trip Setpoints set to ensure thatthe limits of Commitment 6.1.6.1 is not exceeded. The Alarm/Trip Setpoints of thesechannels meeting Commitment 6.1.6.1 shall be determined and adjusted in accordancewith the methodology and parameters in the ODCP.

~Aplicability: As shown in Table 6.1.2-1.

Action:

a. With a radioactive gaseous effluent monitoring instrumentation channelAlarm/Trip Setpoint less conservative than required by the aboveCommitment, immediately suspend the release of radioactive gaseouseffluents monitored by the affected channel, or declare the channelinoperable.

b. With the number of OPERABLE radioactive gaseous effluent monitoringinstrumentation channels less than the Minimum Channels OPERABLE, takethe ACTION shown in Table 6.1.2-1. Restore the inoperable instrumentationto OPERABLE status within the time specified in the ACTIONor explain inthe next Annual Radioactive Effluent Release Report why this inoperabilitywas not corrected within the time specified.


6.1.2.2 Each radioactive gaseous effluent monitoring instrumentation channel shall bedemonstrated OPERABLE by performance of the CHANNELCHECK, SOURCECHECK, CHANNELCALIBRATIONand CHANNELFUNCTIONALTEST at thefrequencies shown in Table 6.1.2-2.

01192903.DOC 1A 12

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TABLE6.1.2-1RADIOACTIVEGASEOUS EFFLUENT MONITORINGINSTRUMENTATION

INSTRUMENT1. Gaseous Radwaste System

Noble Gas ActivityMonitor - Providing

Alarm and Automatic Termination of Release (RM-22)

2. Plant Vent System

a. Noble Gas ActivityMonitor Providing Alarm (RM-14 orRM-14R)

b. Iodine Sampler

c. Particulate Sampler

d. Flow Rate Monitor (FR-12)

e. Iodine Sampler Flow Rate Monitor

3. Containment Purge System

Noble Gas ActivityMonitor - Providing Alarm and AutomaticTermination of Release (RMMAor 44B)

MINIMUMCHANNELOPERABLE

2n)

APPLICABILITY ACTION

2 channels required in Modes 1, 2, 3 and 4. Only 1 channel required in Mode 6 during Core Alterations or movement of irradiated fuel withincontainment.

01192903.DOC 1A 13

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TABLE 6.1.2-1 ContinuedTABLENOTATIONS

At all times.MODES 1-4; also MODE 6 during CORE ALTERATIONSor movement of irradiated fuel withincontainment.

ACTION 5

ACTION 6

ACTION7

ACTION 8

ACTION 9

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, the contents of the tank(s) may be released to theenvironment 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, andb. 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.

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 30days provided the flow rate is estimated at least once per 4 hours.

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via this pathway may continue for up to 30days provided grab samples are taken at least once per 12 hours and these samples areanalyzed for radioactivity within 24 hours.

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, immediately suspend containment PURGING of radioactiveeffluents via this pathway.

With the number of channels OPERABLE less than required by the Minimum ChannelsOPERABLE requirement, effluent releases via the affected pathway may continue forup to 30 days provided samples are continuously collected with auxiliary samplingequipment as required in Commitment Table 6.1.6-1.

NOTE FOR ACTION 9: To respond to the low flow alarm, determine that a simplefixcannot be made and that an auxiliary sampler is needed. Move the sampler in, hookup 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 thecommitment for operation (PSRC Interpretation 85-07).

01192903.DOC IA 14

J

I~ 4

",«; UllfCOJITROIIEDPROCEDURE-DOllf E TOPERFORN lYORKorJSSUEFOR USE""PACIFI ANDELECTRIC COMPANY NUMBER CYNUCLEARPOWER GENERATION REVISION 3A

PAGE 15 OF 37TITLE: RADIOACTIVEEFFLUENT CONTROLS PROGRAM


TABLE6.1.2-2RADIOACTIVEGASEOUS EFFLUENT MONITORINGINSTRUMENTATIONSURVEILLANCERE UIREMENTS

INSTRUMENT

1. Gaseous Radwaste System

Noble Gas ActivityMonitor - Providing

Alarm and Automatic Termination of Release (RM-22)

2. Plant Vent System

a. Noble Gas ActivityMonitor Providing Alarm (RM-14 or RM 14R)

b. Iodine Sampler

c. Particulate Sampler

d. Flow Rate Monitor (FR.12)

e. Iodine Sampler Flow Rate Monitor

3. Containment Purge System

Noble Gas ActivityMonitor - Providing Alarm and Automatic Termination of Release

(RM44A or 44B)

CHANNEL

CHECK

W(4)

W(4)

SOURCE

CHECK

M

N.A.

N.A.

N.A.

N.A.

CHANNEL

CALIBRATION

R(3)

R(3)

N.A.

R(3)

CHANNEL

FUNCTIONAL

TEST

Q(1)

Q(2)

Q(1)

MODES FOR

WHICH

SURVEILLANCE

IS REQUIREO

01 192903.DOC IA 15

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TABLE6.1.2-2 ContinuedTABLENOTATIONS

At all times.** MODES 1-4; also MODE 6 during CORE ALTERATIONSor movement of irradiated fuel within

containment.

(1) The CHANNELFUNCTIONALTEST shall also demonstrate that automatic isolation of thispathway and control room alarm annunciation occurs ifany of the following conditions exists:

(2)

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

b. Instrument indicates a downscale failure (alarm only), orc. Instrument controls not set in operate mode (alarm only).

The CHANNELFUNCTIONALTEST shall also demonstrate that control room alarmannunciation occurs ifany of the following conditions exist:

a. Instrument indicates measured levels above the Alarm Setpoint, orb. Circuit failure, orC. Instrument indicates a downscale failure, ord. Instrument controls not set in operate mode.

(3) The initial CHANNELCALIBRATIONshall be performed using one or more of the referencestandards certified by the National Bureau of Standards (NBS) or using standards that have beenobtained from suppliers that participate in measurement assurance activities with NBS. Thesestandards shall permit calibrating the system over its intended range of energy and measurementrange. For subsequent CHANNELCALIBRATION,sources that have been related to the initialcalibration shall be used.

(4) The CHANNELCHECK shall consist of verifying that the iodine cartridge and particulate filterare installed in the sample holders.

Fre uenc Notation

NotationDWMQRPN.A.

FrequencyAt least once per 24 hoursAt least once per 7 daysAt least once per 31 daysAt least once per 92 daysAt least once per 18 monthsCompleted prior to each releaseNot Applicable

01192903.DOC 1A 16

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6.1.3 Li uid Effluents - CONCENTRATION


6.1.3.1 The concentration of radioactive material released in liquid effluents toUNRESTRICTED AREAS (see TS Figure 5.1-.3) shall be. limited to theconcentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2for radionuclides other than dissolved or entrained noble gases. Fordissolved or entrained noble gases, the concentration shall be limited to2 x 10 microcurie/ml total activity.

~Alicabilit: At all times.

Action:

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


6.1.3.2 Radioactive liquid wastes shall be sampled and analyzed according to thesampling and analysis program of Table 6.1.3-1.

6.1.3.3 The results of the radioactivity analyses shall be used in accordance with themethodology and parameters in the ODCP to assure that the concentrations atthe point of release are maintained within the limits of Commitment 6.1.3.1.

01192903.DOC IA 17

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TABLE6.1.3-1RADIOACTIVELI UID WASTE SAMPLING AND ANALYSISPROGRAM

LIQUIDRELEASETYPE

1. Batch WasteReleaseTanks

'4'.

ContinuousReleases'"

Steam GeneratorBlowdown Tank

3. ContinuousReleases'"

Oily WaterSeparator Effluent

SAMPLINGFREQUENCY

pEach Batch

pOne Batch/M

Each Batch

Each Batch

DGrab Sample

Grab Sample

D

Grab Sample

D

Grab Sample

D

Grab Sample

MINIMUMANALYSIS

FREQUENCYp

Each Batch

M

Composite"'omposite"'

Composite"'omposite"i

Composite"i

W

Composite"'YPE

OFACTIVITYANALYSIS

Principal GammaEmitters t@

I-131

Dissolved andEntrained Gases

(Gamma emitters)

H-3

Gross Alpha

Sr-89, Sr-90

Fe-55

Principal GammaEmitters

"'-131

Dissolved andEntrained Gases

(Gamma emitters)

H-3

Gross Alpha

Sr-89, Sr-90

Fe-55

Principal GammaEmitters'@

LOWER LIMITOFDETECTION

(LLD)(pCi/ml)"

Sx10'x104

lx105

lx10'x10'x10

lx10~

Sx10'xl0~

lxl0~

ix10~

lx10'x10-s

lx10+

Sx10'

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(1) The LLD is defined, for the purposes of these specifications, as the smallest concentration ofradioactive material in a sample that willyield a net count, above system background, that willbedetected with 95% probability with only 5% probability of falsely concluding that a blankobservation represents a "real" signal.

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

LLD— 4.66 sb

E + V + 2.22 x 10 + Y + exp(-Mt)

Where:

LLDSb

EV

2.22 x 10'

ht

the "a priori" lower limitof detection (microcuries per unit mass or volume),the standard deviation of the background counting rate or of the counting rateof a blank sample as appropriate (counts per minute),the counting efficiency (counts per disintegration),the sample size (units of mass or volume),the number of disintegrations per minute per microcurie,the fractional radiochemical yield, when applicable,the radioactive decay constant for the particular radionuclide (sec'), andthe elapsed time between the midpoint of sample collection and the time ofcounting (sec).

(2)

(3)

(4)

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

It should be recognized that the LLD is defined as an ~ariori (before the fact) limit representingthe capability of a measurement system and not as an ~aosterior (atter the fact) limit for a

particular measurement.

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

To be representative of the quantities and concentrations of radioactive materials in liquideffluents, 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 thecomposite sample to be representative of the effluent release.

A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling foranalyses, each batch shall be isolated, and then thoroughly mixed, by a method described in theODCP, to assure representative sampling.

01192903.DOC 1A 19

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(5)

(6)



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.

The principal gamma emitters for which the LLD specification applies include the followingradionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141.Ce-144 shall also be measured but with an LLDof Sx10~. This list does not mean that only thesenuclides are to be considered. Other gamma peaks that are identifiable, together with those of theabove nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent ReleaseReport.

(7) Frequency Notation:

Notation FrequencyD At least once per 24 hours.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.

6.1.4 Li uid Effluents - Dose


6.1.4.1 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactivematerials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (seeTS Figure 5.1-3) shall be limited:

During any calendar quarter to less than or equal to 1.5 mrem to the wholebody and to less than or equal to 5 mrem to any organ, and

During any calendar year to less than or equal to 3 mrem to the whole bodyand to less than or equal to 10 mrem to any organ.

~Atieebitit: At eii times.

Action:

With the calculated dose from the release of radioactive materials in liquideffluents exceeding any of the above limits, prepare and submit to theCommission within 30 days, pursuant to Technical Specification 6.9.2, a

Special Report that identifies the cause(s) for exceeding the limit(s) anddefines the corrective actions that have been taken to reduce the releases andthe proposed corrective actions to be taken to assure that subsequent releaseswillbe in compliance with the above limits.

01192903.DOC 1A 20

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6.1.4.2 Cumulative dose contributions from liquid effluents for the current calendar quarter andthe current calendar year shall be determined in accordance with the methodology andparameters in the ODCP at least once per 31 days.

6.1.5 Li uid Radwaste Treatment S stem


6.1.5.1 The Liquid Radwaste Treatment System* shall be OPERABLE and appropriate portionsof the system shall be used to reduce the radioactive materials in liquid wastes prior totheir discharge when the projected doses due to the liquid effluent, from each unit, toUNRESTRICTED AREAS (see TS Figure 5.1-3) would exceed 0.06 mrem to thewhole body or 0.2 mrem to any organ in a 31-day period.

~Alicabilit: At aii times.

Action:

With any portion of the Liquid Radwaste Treatment System not in operationand with radioactive liquid waste being discharged without treatment and inexcess of the above limits, prepare and submit to the Commission within30 days, pursuant to Technical Specification 6.9.2, a Special Report whichincludes the following information:

Explanation of why liquid radwaste was being discharged withouttreatment, identification of any inoperable equipment or subsystems,and the reason for the inoperability,

Action(s) taken to restore the inoperable equipment to OPERABLEstatus, and

3. Summary description of action(s) taken to prevent a recurrence.


6.1.5.2 Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall beprojected at least once per 31 days, in accordance with the methodology and parametersin the ODCP when Liquid Radwaste Treatment Systems are not being fullyutilized.

6.1.5.3 The installed Liquid Radwaste Treatment System shall be considered OPERABLE bymeeting Commitments 6.1.3.1 and 6.1.4.1.

* The Liquid Radwaste Treatment System is common to both units.

01192903.DOC 1A 21

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6.1.6 Gaseous Effluents - Dose Rate



6.1.6.1 The dose rate due to radioactive materials released in gaseous effluents from the site toareas at or beyond the SITE BOUNDARY (see TS Figure 5.1.-3).shall be limited to thefollowing:

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

b. For Iodine-131, for Iodine-133, for tritium, and for all radionuclides inparticulate form with half-lives greater than 8 days: Less than or equal to1500 mrem/yr to any organ.

~Attcabtttt: At all times.

Action:

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


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

6.1.6.3 The dose rate due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulateform with half-lives greater than 8 days in gaseous effluents shall be determined to bewithin the above limits in accordance with the methodology and procedures of theODCP by obtaining representative samples and performing analyses in accordance withthe sampling and analysis program specified in Table 6.1.6-1.

01192903.DOC IA 22

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TABLE6.1.6-1RADIOACTIVEGASEOUS WASTE SAMPLING AND ANALYSISPROGRAM

GASEOUS RELEASE TYPE

1. Waste Gas Decay Tank

2. Containment Purge

3. Plant Vent

SAMPLINGFREQUENCY

PEach TankGrab Sample

PEach

Purge"'rab

Sample

M<'>

Grab Sample

MP) (8

Grab Sample

MINIMUMANALYSIS

FREQUENCY

Each Tank

PEach

Purge"'r>

TYPE OFACTIVITYANALYSIS

Principal Gamma Emitters+(noble gases)


I-131, I-133

Principal Gamma Emitters(particulates)

H-3


H-3

LOWER LIMITOFDETECTION (LLD)

(pCi/ml) <'>

Ix10~

Ixl04

1x10 ~

1x10

Ix10~

1x10

1x10

011%%)3.DOC 1A 23

',"'IIJOOIVTROIIEOPROOEOURE -OOR E TOPERFO8N EFORKorISSUEFORUSE QPACIFI AS AND ELECTRIC COMPANY NUMBER CY2. INUCLEARPOWER GENERATION REVISION 3A



TABLE6.1.6-1 Continued

4. AllRelease Types as listed in 1., 2.,3., above, at the plant vent

Continuous'

CharcoalSample

W'1 I-131

I-133

1x10'x10

5. Steam Generator Blowdown Tank Vent

Continuous'@

Continuous'@

Continuous'@

M1'0

w<'>

ParticulateSample

MCompositeParticulateSample

CompositeParticulateSample

M(8)

Principal Gamma Emitters+

Gross Alpha

Sr-89, Sr-90


1x10"

1x10"

1x10"

Ix10~

01192903.DOC 1A 24

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TABLE 6.1.6-1 ContinuedTABLENOTATIONS

(1) The LLD is defined, for purposes of these specifications, as the smallest concentration ofradioactive material in a sample that willyield a net count, above system background, that willbedetected with 95% probability with only 5% probability of falsely concluding that a blankobservation represents a "real" signal.

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

I.LD = '

E V 2.22x10 Y exp(-Xb,t)

Where:

LLD

EV

2.22 x 10'

the "a priori" lower limitof detection (microcuries per unit mass or volume),the standard deviation of the background counting rate or of the counting rateof a blank sample as appropriate (counts per minute),the counting efficiency (counts per disintegration),the sample size (units of mass or volume),the number of disintegrations per minute per microcurie,the fractional radiochemical yield, when applicable,the radioactive decay constant for the particular radionuclide (sec'), andthe elapsed time between the midpoint of sample collection and the time ofcounting (sec).

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

It should be recognized that the LLD is defined as an ~ariort (before the fact) limit representingthe capability of a measurement system and not as an ~aosterior (after the fact) limit for a

particular measurement.

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

(3) Tritium grab samples shall be taken a least once per 24 hours when the refueling canal is flooded.

01192903. DOC lA 25

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TABLE6.1.6-1 ContinuedTABLENOTATIONS Continued

(4) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hoursafter changing or after removal from sampler. Sampling shall also be performed at least once per24 hours for at least 7 days following each shutdown, startup or THERMALPOWER 'change

exceeding 15% of RATED THERMALPOWER within a 1-hour period and analyses shall becompleted within 48 hours of changing. When samples collected for 24 hours are analyzed, thecorresponding LLD's may be increased by a factor of 10. This requirement does not apply if: (1)analysis shows that the DOSE EQUIVALENTI-131 concentration in the reactor coolant has notincreased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity hasnot increased more than a factor of 3.

(5)

(6)

(7)

(8)

(9)

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

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the timeperiod covered by each dose or dose rate calculation made in accordance with Commitments6.1.6.1, 6.1.7.1, and 6.1.8.1.

The principal gamma emitters for which the LLD specification applies include the followingradionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 in noble gas releases andMn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, Ce-141, and Ce-144 inIodine and particulate releases. This list does not mean that only these nuclides are to beconsidered. 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.

Grab samples shall be taken and analyzed at least once per 31 days whenever there is fiowthrough the steam generator blowdown tank. Releases of radioiodines shall be estimated based onsecondary coolant concentration and partitioning factors during releases or shall be measured.

Fre uenc Notation

NotationWMQP

FrequencyAt least once per 7 daysAt least once per 31 daysAt least once per 92 daysCompleted prior to each release

0l 192903.DOC lA 26

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6.1.7 Dose - Noble Gases



6.1.7.1 The air dose due to noble gases released in gaseous effluents, from each unit, to areasat or beyond the SITE BOUNDARY(see TS Figure 5.1-3).shall be limited to thefollowing:

During any calendar quarter: Less than or equal to 5 mrad for gammaradiation 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 gammaradiation and less than or equal to 20 mrad for beta radiation.

~Aticabiltt: At all times

Action:

With the calculated air dose from radioactive noble gases in gaseous effluentsexceeding any of the above limits, prepare and submit to the Commissionwithin 30 days, pursuant to Technical Specification 6.9.2, a Special Reportthat identifies the cause(s) for exceeding the limit(s), defines the correctiveactions that have been taken to reduce the releases and the proposedcorrective actions to be taken to assure that subsequent releases willbe incompliance with the above limits.


6.1.7.2 Cumulative dose contributions for the current calendar quarter and current calendaryear for noble gases shall be determined in accordance with the methodology andparameters in the ODCP at least once per 31 days.

6.1.8 Dose - Iodine-131 Iodine-133 Tritium and Radioactive Material in Particulate Form


6.1.8.1 The dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium andall radionuclides in particulate form with half-lives greater than 8 days in gaseouseffluents released, from each unit, to areas at and beyond the SITE BOUNDARY (seeTS Figure 5.1-3) shall be limited to the following:

During any calendar quarter: Less than or equal to 7.5 mrem to any organand,

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

~Aticabilit: At all times.

i

01192903.DOC 1A 27

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Action:


With the calculated dose from the release of Iodine-131, Iodine-133, tritium,and radionuclides in particulate form with half-lives greater than 8 days, ingaseous effluents exceeding any of the above limits, prepare and submit tothe Commission within 30 days, pursuant to Technical Specification 6.9.2, a

Special Report that identifies the cause(s) for exceeding the limit(s), definesthe corrective actions that have been taken to reduce the releases and theproposed actions to be taken to assure that subsequent releases willbe incompliance with the above limits.


6.1.8.2 Cumulative dose contributions for the current calendar quarter and current calendaryear for Iodine-131, Iodine-133, tritium, and radionuclides in particulate form withhalf-lives greater than 8 days shall be determined in accordance with the methodologyand parameters in the ODCP at least once per 31 days.

6.1.9 Gaseous Radwaste Treatment S stem


The GASEOUS RADWASTE SYSTEM and the VENTILATIONEXHAUSTTREATMENTSYSTEM shall be OPERABLE. The appropriate portions of thesesystems shall be used to reduce releases of radioactivity when the projected doses in 31days due to gaseous effluent releases, from each unit, to areas at and beyond the SITEBOUNDARY (see TS Figure 5.1-3), would exceed 0.2 mrad to air from gammaradiation or Oe4 mrad to air from beta radiation or 0.3 mrem to any organ of a

MEMBER OF THE PUBLIC.

~Alicabiltt: At all times.

Action:

With radioactive gaseous waste being discharged without treatment and inexcess of the above limits, prepare and submit to the Commission within 30days, pursuant to Technical Specification 6.9.2, a Special Report thatincludes the following information:

1. Identification of the inoperable equipment or subsystems and the reasonfor inoperability,

2. Action(s) taken to restore the inoperable equipment to OPERABLEstatus, and

3. Summary description of action(s) taken to prevent a recurrence.

01192903.DOC iA 28

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6.1.9.2 Doses due to gaseous releases from each unit to areas at and beyond the SITEBOUNDARYshall be projected at least once per 31 days, in accordance with themethodology and parameters in the ODCP when Gaseous Radwaste Treatment Systemsare not being fullyutilized.

6.1.9.3 The installed VENTILATIONEXHAUST TREATMENTSYSTEM and GASEOUSRADWASTE SYSTEM shall be considered OPERABLE by meeting Commitments6.1.6.1 and 6.1.7.1 or 6.1.8.1.

6.1.10 Total Dose


6.1. 10.1 The annual (calendar year) dose or dose commitment to any MEMBER OF THEPUBLIC due to releases of radioactivity and to radiation from uranium fuel cyclesources shall be limited to less than or equal to 25 mrems to the whole body or anyorgan, except the thyroid, which shall be limited to less than or equal to 75 mrems.

~Aplicabilitlr, At all times.

Action:

With the calculated doses from the release of radioactive materials in liquidor gaseous effluents exceeding twice the limits of Commitments 6.1.4.1.a,6.1.4.1.b, 6.1.7.1.a, 6.1.7.1.b, 6.1.8.1.a, or 6.1.8.1.b, calculations shall bemade including direct radiation contributions from the units and from outsidestorage tanks to determine whether the above limits of RegulatoryCommitment 6.1.10.1 have been exceeded. Ifsuch is the case, prepare andsubmit to the Commission within 30 days, pursuant to Technical Specification6.9.2, a Special Report that defines the corrective action to be taken toreduce subsequent releases to prevent recurrence of exceeding the abovelimits and includes the schedule for achieving conformance with the abovelimits. This Special Report, as defined in 10 CFR 20.2203(a), shall includean analysis that estimates the radiation exposure (dose) to a MEMBER OFTHE PUBLIC from uranium fuel cycle sources, including all effluentpathway and direct radiation, for the calendar year that includes the release(s)covered by this report. It shall also describe levels of radiation andconcentrations of radioactive material involved, and the cause of the exposurelevels or concentrations. Ifthe estimated dose(s) exceeds the above limits,and ifthe release condition resulting in violation of 40 CFR Part 190 has notalready been corrected, the Special Report shall include a request for a

variance in accordance with the provisions of 40 CFR Part 190. Submittal ofthe report is considered a timely request, and a variance is granted until staffaction on the request is complete.

01192903.DOC IA 29

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6.1.10.2 Cumulative dose contributions from liquid and gaseous effluents shall be determined inaccordance with Commitments 6.1.4.2, 6.1.7.2, and 6.1.8.2, and in accordance withthe methodology and parameters in the ODCP.

6.1.10.3 Cumulative dose contributions from direct radiation from the units and from outsidestorage tanks shall be determined in accordance with the methodology and parametersin the ODCP. This requirement is applicable only under conditions set forth inACTION a. of Commitment 6.1.10.1.

6.1.11 Radiolo ical Environmental Monitorin

6.1.11.1 Commitment for 0 eration - The Radiological Environmental Monitoring Programshall be conducted as specified in RP1.ID11, "Environmental Radiological MonitoringProcedure."

~Aticabitit: At all times.

Action:

With the confirmed level of radioactivity as the result of plant effluents in anenvironmental sampling medium at a specified location exceeding the"Reporting Levels for Nonroutine Operating Reports" in RP1.ID11 whenaveraged over any calendar quarter, prepare and submit to the Commissionwithin 30 days from the end of the quarter, pursuant to Specification 6.9.2, aNonroutine Radiological Environmental Operating Report that identifies thecause(s) for exceeding the limit(s) and defines the corrective actions to betaken to reduce radioactive effluents so that the potential annual dose to a

MEMBER OF THE PUBLIC is less than the calendar year limits ofCommitment 6.1.4.1, 6.1.7.1, or 6.1.8.1. A confirmatory reanalysis of theoriginal, a duplicate, or a new sample may be desirable, as appropriate. Theresults of the confirmatory analysis shall be completed at the earliest timeconsistent with the analysis, but in any case within 30 days. When more thanone of the radionuclides from "Reporting Levels for Nonroutine OperatingReports" in RP1.ID11 are detected in the sampling medium, this report shallbe submitted if:

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

01 l92903.DOC 1A 30

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t


When radionuclides other than those in the "Reporting Levels for NonroutineOperating Reports" in RP1.ID11 are detected and are the result of planteffluents, a Nonroutine Radiological Environmental Operating Report shallbe submitted ifthe potential annual dose to a MEMBER OF THE PUBLICfrom all radionuclides is equal to or greater than. the calendar year limits ofCommitment 6.1.4.1, 6.1.7.1, or 6.1.8.1. This report shall include anevaluation of any release conditions, environmental factors, or other aspectsnecessary to explain the anomalous result.

6.1.12 LAND.USE CENSUS

6.1.12.1 A Land Use Census shall be conducted as specified in RP1.ID11, "EnvironmentalRadiological Monitoring Procedure."

~Aticabitit: At aii times.

Action:

With a Land Use Census identifying a location(s) that yields a calculated doseor dose commitment greater than the values currently being calculated inCommitment 6.1.8.2, identify the new location(s) in the next AnnualRadioactive Effluent Release Report.

With a Land Use Census identifying a location(s) that yields a calculated doseor dose commitment (via the same exposure pathway) 20% greater than at a

location from which samples are currently being obtained in accordance withCommitment 6.1.11.1, add the new location(s) within 30 days to theRadiological Environmental Monitoring Program given in the ERMP. Thesampling location(s), excluding the control station location, having the lowestcalculated dose or dose commitment(s), via the same exposure pathway, maybe deleted from this monitoring program after October 31 of the year inwhich this Land Use Census was conducted. Submit in the next AnnualRadioactive Effluent Release Report documentation for a change in theERMP including a revised figure(s) and table(s) for the ERMP reflecting thenew location(s) with information supporting the change in sampling locations.

Bases

6.1.1 Radioactive Li uid Effluent Monitorin 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 liquideffluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted inaccordance with the methodology and parameters in the ODCP to ensure that the alarm/trip willoccur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITYand use of thisinstrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 ofAppendix A to 10 CFR Part 50.

01192903.DOC 1A 31

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-',.':;"'.".'l/ItfCOIYTROllED'PROCEDURE-DO'iVOTVSE,TO PERFORM O'ORKor'ISSUEFOR VSE ',PACIFIC GAS AND ELECTRIC COMPANYNUCLEAR POWER GENERATION

TITLE: RADIOACTIVEEFFLUENT CONTROLS PROGRAMt


6.1.2

6.1.3


Radioactive Gaseous Effluent Monitorin 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 ofgaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjustedin accordance with the methodology and parameters in the ODCP to ensure that the alarm/trip willoccur prior to exceeding the limits of NUREG 0133. The OPERABILITYand use of thisinstrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 ofAppendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitors used to showcompliance with the gaseous effluent release requirements of Regulatory Commitment 6.1.7.1shall be such that concentrations as low as 1 x 10~ pCi/ml are measurable.

Li uid Effluents - Concentration

This Regulatory Commitment is provided to ensure that the concentration of radioactive materialsreleased in liquid waste effluents to UNRESTRICTED AREAS willbe less than the concentrationlevels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2. This limitation providesadditional assurance that the levels of radioactive materials in bodies of water inUNRESTRICTED AREAS willresult in exposures within: (1) the Section II.Adesign objectivesof Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of10 CFR 20.1301(e) to the population. The concentration limit for dissolved or entrained noblegases is based upon the assumption that Xe-135 is the controlling radioisotope and its EffluentConcentration Limit (ECL) in air (submersion) was converted to an equivalent concentration inwater 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 effluentsfrom all units at the site.

The required detection capabilities for radioactive materials in liquid waste samples are tabulatedin terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and otherdetection limits can be found in Currie, L.A., "Lower Limitof Detection: Definition andElaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300 (revisedannually).

01192903.DOC lA 32

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PACIFIC GAS AND ELECTRIC COMPANYNNUCLEARPOWER GENERATION



6.1.4 Li uid Effluents - Dose


6.1.5

This Regulatory Commitment is provided to implement the requirements of Sections II.A, III.Aand IV.Aof Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements theguides set forth in Section II.Aof Appendix I. The ACTION statements provide the requiredoperating flexibilityand at the same time implement the guides set forth in Section IV.AofAppendix I to assure that the releases of radioactive material in liquid effluents toUNRESTRICTED AREAS willbe kept "as low as is reasonably achievable." The dosecalculation methodology and parameters in the ODCP implement the requirements in Section HI.Aof Appendix I that conformance with the guides of Appendix I be shown by calculationalprocedures based on models and data, such that the actual exposure of a MEMBER OF THEPUBLIC through appropriate pathways is unlikely to be substantially underestimated. Theequations specified in the ODCP for calculating the doses due to the actual release rates ofradioactive materials in liquid effluents are consistent with the methodology provided inRegulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of ReactorEffluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents fromAccidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April1977.

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

Li uid Radwaste Treatment S stem

The OPERABILITYof the Liquid Radwaste Treatment System ensures that this system willbeavailable 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 providesassurance that the releases of radioactive materials in liquid effluents willbe kept "as low as isreasonably 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 inSection II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use ofappropriate portions of the Liquid Radwaste Treatment System were specified as a suitablefraction of the dose design objectives set forth in Section II.Aof Appendix I, 10 CFR Part 50, forliquid effluents.

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

01192903. DOC 1A 33

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~ TITLE: RADIOACTIVEEFFLUENT CONTROLS PROGRAM


6.1.6 Gaseous Effluents - Dose Rate


This Regulatory Commitment is provided to ensure that the dose at any time at and beyond theSITE BOUNDARYfrom gaseous effluents from all units on the site willbe within the annual dose

limits of NUREG 0133 to UNRESTRICTED AREAS. The annual dose limits are the doses to beassociated with the concentrations of 10 CFR Part 20, Appendix'B, Table 2, Column 1. Theselimits provide reasonable assurance that radioactive material discharged in gaseous effluents willnot 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 tobe specified in Appendix B, Table 2 of 10 CFR Part 20 (10 CFR Part 20.1302(c)). ForMEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, theoccupancy of that MEMBER OF THE PUBLIC willusually be sufficiently low to compensate forany increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examplesof 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 THEPUBLIC at or beyond the SITE BOUNDARYto less than or equal to 500 mrems/year to thewhole body or to less than or equal to 3000 mrems/year to the skin. These release rate limits alsorestrict, at all times, the corresponding thyroid dose rate above background to a child via theinhalation pathway to less than or equal to 1500 mrems/year.

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

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

01192903.DOC 1A 34

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6.1.7 Dose - Noble Gases


This Regulatory Commitment is provided to implement the requirements of Sections II.B, III.Aand IV.Aof Appendix I, 10 CFR Part 50. The Limiting Condition for Operation implements theguides set forth in Section II.B of Appendix I. The ACTION statements provide the requiredoperation flexibilityand at the same time implement the guides set'forth in Section IV.AofAppendix I to assure that the releases of radioactive material in gaseous effluents toUNRESTRICTED AREAS willbe kept "as low as is reasonably achievable." The SurveillanceRequirements implement the requirements in Section III.Aof Appendix I that conformance withthe guides of Appendix I be shown by calculational procedures based on models and data suchthat the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways isunlikely to be substantially underestimated. The dose calculation methodology and parametersestablished in the ODCP for calculating the doses due to the actual release rates of radioactivenoble gases in gaseous effluents are consistent with the methodology provided in RegulatoryGuide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluentsfor 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 andDispersion 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 andbeyond the SITE BOUNDARYare based upon the historical average atmospheric conditions.

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

01192903.DOC lA 35

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6.1.8


Dose - Iodine-131 Iodine-133 Tritium and Radioactive Material in Particulate Form

This Regulatory Commitment is provided to implement the requirements of Sections II.C, III.A,and IV.Aof Appendix I, 10 CFR Part 50. The Limiting Conditions for Operation are the guidesset forth in Section II.C of Appendix I. The ACTION statements provide the required operatingflexibilityand at the same time implement the guides set forth in Sections.A of Appendix I toassure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREASwillbe kept "as low as is reasonably achievable." The ODCP calculational methods specified inthe Surveillance Requirements implement the requirements in Section III.Aof Appendix I thatconformance with the guides of Appendix I be shown by calculational procedures based onmodels and data, such that the actual exposure of a MEMBER OF THE PUBLIC throughappropriate pathways is unlikely to be substantially underestimated. The ODCP calculationalmethodology and parameters for calculating the doses due to the actual release rates of the subjectmaterials are consistent with the methodology provided in Regulatory Guide 1.109, "Calculationof Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose ofEvaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 andRegulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion ofGaseous Effluents in Routine Releases from Light-Water-Cooled Reactors,: Revision 1, July1977. These equations also provide for determining the actual doses based upon the historicalaverage 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 dependentupon the existing radionuclide pathways to man in the areas at and beyond the SITEBOUNDARY. The pathways that were examined in the development of the calculations were:(1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafyvegetation with subsequent consumption by man, (3) deposition onto grassy areas where milkanimals 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 effluentsfrom each unit at the site. For units with shared Radwaste Treatment Systems, the gaseouseffluents from the shared system are proportioned among the units sharing that system.

01192903.DOC IA 36

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6.1.9 Gaseous Radwaste Treatment S stem

The OPERABILITYof the GASEOUS RADWASTE SYSTEM and the VENTILATIONEXHAUST TREATMENTSYSTEM ensures that the systems willbe available for use whenevergaseous effluents require treatment prior to release to the environment. The requirement that theappropriate portions of these systems be used, when specified, provides reasonable assurance thatthe releases of radioactive materials in gaseous effluents willbe kept "as low as is reasonablyachievable." This specification implements the requirements of 10 CFR 50.36a, General DesignCriterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D ofAppendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions ofthe systems were specified as a suitable fraction of the dose design objectives set forth in SectionsII.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.

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

6.1 ~ 10 Total Dose

This Regulatory Commitment is provided to meet the dose limitations of 40 CFR Part 190 thathave been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires thepreparation and submittal of a Special Report whenever the calculated doses due to releases ofradioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole bodyor any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. Forsites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OFTHE PUBLIC willexceed the dose limits of 40 CFR Part 190 ifthe individual reactors remainwithin twice the dose design objectives of Appendix I, and ifdirect radiation doses from the unitsand from outside storage tanks are kept small. The Special Report willdescribe a course of actionthat should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to withinthe 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that thedose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources isnegligible, with the exception that dose contribution from other nuclear fuel cycle facilities at thesame site or within a radius of 8 km must be considered. Ifthe dose to any MEMBER OF THEPUBLIC 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 190have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and10 CFR 20.2203(a), is considered to be a timely request and fulfillsthe requirements of40 CFR Part 190 until NRC staff action is completed. The variance only relates to the limits of40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of10 CFR Part 20, as addressed in Regulatory Commitment 6.1.3.1 and 6.1.6.1. An individual isnot considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged incarrying out any operation that is part of the nuclear fuel cycle.

01192903.DOC lA 37


Off-Site Dose Calculation Procedure (ODCP)

(CAP A-8 Revision 22)

***ISSUED FOR USE BY:PACIFIC GAS AND ELECTRIC COMPANYNUCLEAR POWER GENERATIONDIABLOCANYONPOWER PLANTCHEMICALANALYSISPROCEDURE

DATE: EXPIRES:NUMBER CAP A-8

REVISION 22

PAGE 1 OF 64

UNITS

TITLE: OFF-SITE DOSE CALCULATIONS

EFFECTIVE DATE

PROCEDURE CLASSIFICATION: QUALITYRELATED

SECTION

SCOPE ...DISCUSSIONRESPONSIBILITIES ...PRERE UISITES

TABLEOF CONTENTS

PRECAUTIONSINSTRUCTIONS

Liquid Effluents.Liquid Effluents - Dose Calculation.10 CFR 20, Appendix B, Table 2, Column 2, Effluent (liquid) Concentration Limit (ECL)Calculation.Liquid Effluent Radiation Monitor Set Point MethodologyDose Projection (for Liquid Effluents) .

Liquid LimitingFlow Rate Methodology - ECL Based.Liquid Limiting Flow Rates - LLD Based.

Gaseous Effluents.Plant Vent Noble Gas Monitor - RE-14 HASPContainment Purge - RE-44 HASP .

Dose To Critical Receptor Due To Radioiodines, Tritium and Particulates Released in GaseousEffluents.

40 CFR 190 Dose CalculationsACCEPTANCE CRITERIA .

REFERENCESRECORDSAPPENDICESATTACHMENTSSPONSOR .

PAGE

23

3

3

3

444

67

13

15

16

18

2128

3442474747484848

Table 6.1 - Typical Liquid Effluent Discharge Pathway Allocation Factors.Table 6.2 - Typical Gaseous Effluent Discharge Pathway Allocation Factors ..............Table 6.3 - Expected On-Site Distance and Visitation Times for Members of the Public.

Figure 6.1 - Liquid Discharges (LRW) Monitored for RadioactivityFigure 6.2 - Gaseous Discharges (GRW) Monitored for Radioactivity

8

20.. 46

4950

CAPA8.DOC 08 1

PACIFIC GAS ANDELECTRIC COMPANYDIABLOCANYONPOWER PLANT

TITLE: OFFQITE DOSE CALCULATIONS

NUMBER CAP A-'8

REVISION 22PAGE 2 OF 64

UNITS I AND 2

SCOPE

This procedure describes the methodology for the following:

EffluentType

Liquids

Gases

RECP orTech Spec

6.1.1.1 (RECP)6.1.3.1 (RECP)

6.1.2.1 (RECP)6.1.6.1 (RECP)

RECP or Tech Spec Surveillance Requirement

Determination of alarm/trip setpoints for RE-18;RE-23, and RE-3

Determination of alarm/trip setpoints for RE-22,RE-14, and RE-14R

Implements

10 CFR 20 App. BTable 2, Col. 2

NUREG 0133

Liquids

Liquids

Liquids

Gases

6.1.3.2 (RECP) Prerelease analyses of effluents6.1.3.3 (RECP)

Postrelease analysis of effluents

6.1.4.2 (RECP) Dose calculations

6.1.5.2 (RECP) Dose projections

6.1.6.2 Dose Rate calculations, Noble Gases, Total Body andSkin

10 CFR20 App. BTable 2, Col. 2

10 CFR 50 App. I

10 CFR 50 App. I

NUREG 0133

Gases 6.1.6.3 Dose Rate calculations, Iodines, Particulates and, NUREG 0133Radionuclides other than Noble Gases, per organ, perage group

Gases

Gases

Gases

6.1.7.2 (RECP) Noble Gas AirDose Calculations

6.1.8.2 (RECP)

6.1.9.2 (RECP)

Iodines, Particulates, and Radionuclides other thanNoble Gases Organ Dose Calculations per age group

Noble Gases, Iodines, Particulates, and Radionuclidesother than Noble Gases, Dose Projection

10 CFR 50 App. I

10 CFR 50 App. I

10 CFR 50 App. I

LiquidsandGases

DirectRadiation

6.1.10.2(RECP)4.4.2.b.l(RECP)

6.1.10.3(RECP)

Cumulative Dose from: Liquids, Noble Gases,Iodines, Particulates, and Radionuclides other thanNoble Gases per age group, per organ

Direct Radiation Dose Rate and Dose Calculations tounrestricted areas due to plant and high radwastestorage sky-shine

40 CFR 190

40 CFR 190

The calculational methodology for doses are based on models and data that make it unlikely tosubstantially underestimate the actual exposure of an individual through any of the appropriatepathways. Tables containing the values for the various parameters used in these expressions arealso included.

CAPA8.DOC 08 2



NUMBER CAP A-8REVISION 22PAGE 3 OF 64

UNITS 1 AND 2

2. DISCUSSION

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 routineradioactive 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 througheither 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.).

Further, the conduct of the Environmental Radiological'Monitoring Procedure (ERMP) is foundin RP1.ID11.

3. RESPONSIBILITIES

3.1 The Director, Chemistry is the overseeing authority of responsibility for ensuring thatthe off-site dose calculational procedure (ODCP) meets all RECP and Tech Specrequirements with regards to calculated doses delivered by the plant to the unrestrictedarea surrounding the site.

3.2

3.3

3.4

3.5

3.6

The Senior Radiochemistry Engineer assumes the overall responsibility for ensuringthat this procedure's program is followed and implemented where appropriate,especially in regards to RECP or Tech Spec requirements.

The Radiochemistry Effluents Engineer has the responsibility of correct and timelyimplementation of all the procedure's calculational methodology, where appropriate, foreach radioactive effluent released. Furthermore this engineer is responsible for:reviewing the results; cross (spot) checking the calculations; and maintaining an updatedarchive of post release calculated doses for annual report purposes.

The Senior Engineer Tech Maintenance Computer Group assures that any supportingcomputer software is maintained current and compatible with the procedure'scalculational methodology and that the computer hardware is maintained operable at alltimes.

The Radiochemistry staff engineer provides an oversight of the effluents program'sODCP to: confirm compliance with RECP or Tech Specs; provide technical support;recommend or design improvements to the dose calculational methodology and theeffluent program control; and investigate long-term planning toward effluent related

'ctivitiesand their associated dose calculations.

Responsibilities as described in CY1, " Chemistry and Radiochemistry," andCY1.DC1, "Analytical Data Processing Responsibilities," apply.

Not Applicable

PRECAUTIONS

Not Applicable

CAPA8.DOC 08 3

PACIFIC GAS AND ELECTRIC COMPANYDIABLOCANYONPOWER PLANT



UNITS 1 AND 2

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 tounrestricted areas using the following expression:

D, = F, 6t gA,,C,. eI

Where:

Do The dose commitment to organ, o, in mrem.

Fc Near field average dilution factor during the period of therelease. It is defined as:

Where:

Waste FlowDilutionFlow x Z

(2)

C;

Z is the site specific factor for the mixing effect of thedischarge structure. Specifically, it is the credit taken fordilution which occurs between the discharge structure andthe body of water which contaminates fish or invertebratesin the liquid ingestion pathway. For DCPP Z = 5.

The time period for the release in hours.

The site specific ingestion dose commitment factor toorgan, o, due to radionuclide, i, in mrem/hr per pCi/ml as

defined by Equation 3.

Concentration of radionuclide, i, in the undiluted liquideffluent, in pCi/ml.

Decay constant of radionuclide, i.

Time interval between end of sampling and midpoint ofrelease.

CAPA8.DOC 08 4




UNITS 1 AND 2

The site specific ingestion dose commitment factor, A;„ is defined as:

A. =1c U„BF. +U,BI. DF.

Where:

Units conversion factor of 1. 14x10- in units ofpCi/pCi x ml/1 x yr/hr.

(3)

UF —— Saltwater fish consumption rate in kilograms of fish peryear. DCPP value for this parameter is 21 kg/yr and istaken from NUREG 0133, Section 4.3.1.

BFi

BI;

DF;

Saltwater bioaccumulation factor for nuclide, i, in fishflesh in units of pCi/Kg per pCi/1. Values for BF; aretaken from Table A-1 of Reg. Guide 1.109.

Saltwater invertebrate consumption rate in kilograms peryears. DCPP value for this parameter is 5 kg/yr and istaken from NUREG 0133, Section 4.3.1.

Saltwater bioaccumulation factor for nuclide, i, ininvertebrate flesh in units of pCi/Kg per pCi/1. Values forBI; are taken from Table A-1 of Reg. Guide 1.109.

Adult ingestion dose conversion factor for nuclide, i, inmrem per pCi ingested, from Table E-11 of RegulatoryGuide 1.109, with the following exceptions: H-3, Br-82,Sb-124 and Sb-125 ingestion dose conversion factors aretaken from NUREG/CR-4013. As-76, Sn-113, Sn-117mand Sb-122 ingestion dose conversion factors werecalculated by ORNL using ICRP-2 methodology.

The site specific values for A;, are listed in Table 10.1.

Units 1 and 2 share a common liquid radwaste (LRW) treatment system. Theeffluent doses due to releases discharged via the common LRW areapportioned between the units with 50% credited to Unit 1 and 50% creditedto Unit 2.

CAPA8.DOC 08 5


TITLE: OFFNITE DOSE CALCULATIONS


UNITS I AND 2

6.1.2 10 CFR 20, Appendix B, Table 2, Column 2, Effluent (liquid) ConcentrationLimit (ECL) Calculation

a. The ECL for the identified mixture of radionuclides in the "j " batch ofliquids is calculated as follows:

Where:

pc.„.ECL =

ECL

(4)

ECL1

C0

ECL0

The unrestricted area, total undiluted ECL for the"j " particular mixture of identified radionuclides,in pCi/ml.

The concentration of radionuclide "i," in pCi/mlfor the "j "mixture.

The ECL in unrestricted area water forradionuclide "i," in general, in pCi/ml (from10 CFR 20, Appendix B, Table 2, Column 2).

CAPAS. DOC 08 6




UNITS 1 AND 2

b. The overall ECL for simultaneous discharges is given by Equation 5.

Where:

ECLovcrall

C;

ECLJ

ge,.c,.j~l

overall n

,, ECL;

The unrestricted area ECL for the current radionuclidemixture for concurrent "j"discharges (in pCi/ml).

The total activity concentration for the "j " individualstream in pCi/ml.

The total ECL for the "j " individual mixture (orstream) determined as defined in Equation 4, inpCi/ml.

The ratio of an individual discharge "j " pathwayflowrate to the sum total of all individual undilutedpathway flowrates as defined by:

Where:

fjJ

J

fj = Undiluted effluent flowrate for pathway, "j".

6.1.3 Liquid Effluent Radiation Monitor Set Point Methodology

Introduction

The DCPP radiological effluent controls program requires that theliquid effluent monitors be operable with their alarm/trip set points set

to ensure that the effluent concentration limits of 10 CFR 20 are notexceeded.

The alarm/trip set point for the liquid effluent radiation monitors is

derived from the concentration limitset 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 betweenthe point of discharge and the unrestricted area boundary.

CAPA8.DOC 08 7




UNITS 1 AND 2

b. Allocation and Safety Factors

The limits of RCMP 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 unrestrictedarea. The DCPP High Alarm Set Point (HASP) methodology makesuse of an Allocation Factor (AF) to limit'he effluent concentrationsfrom simultaneous liquid discharges. The Allocation Factors can be

adjusted based upon, operational requirements with the restriction thatthe sum of the Allocation Factors must be less than or equal to 1.

Typical Allocation Factors are shown Table 6. 1.

Table 6. 1

Typical Liquid Effluent Discharge Pathway Allocation Factors

Discharge PathwayOily Water Separator RE-3 0.01

Rad Monitor Allocation Factor (AF)

Liquid Radwaste SystemSteam Generator Blow Down (Unit 1)

Steam Generator Blow Down (Unit 2)Miscellaneous

RE-18RE-23 (U1)RE-23 (U2)

none

0.900.040.040.01

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 the10 CFR 20 concentration limits.

Tritium Correction Factor „

As result of an aggressive liquid radwaste treatment program, the liquideffluents 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 highertritium concentration in liquid effluents when compared with the lowgamma emitter concentrations. As a result, standard HASPmethodology results in very low set points. In some cases thecalculated set points are barely above the monitor background.

The liquid HASP methodology used by DCPP uses a TritiumCorrection Factor (TCF) which assumes a constant, but conservativetritium concentration in the liquid effluent. This results in an

operationally reasonable set point while ensuring that the liquid effluentconcentrations released to the unrestricted areas do not exceed the limitsof 10 CFR 20.

CAPA8.DOC 08 8




UNITS 1 AND2

The Tritium Correction Factor is defined as shown in Equation 7.

H3I C H3

@if (7)

Where:

ECLii3 = effluent concentration limitof tritium

CH3 = concentration of tritium in the release mix,pre-dilution (pCi/ml)

conservative dilution flow rate (gpm)

conservative undiluted effluent flow rate (gpm)

The concentration of tritium, Cio, is conservatively estimated.

d. Liquid Effluent Radiation Monitor Set Point Calculations

The High Alarm Set Point (HASP) are calculated to ensure that theliquid effluent concentration limits of 10 CFR 20 are not exceeded.The set points represent the maximum operational set point. The actualset point used by operations willbe equal to or less than the actualvalue as determined by the HASP methodology described in thissection.

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 thereis primary to secondary leakage does this become a potentialliquid effluent discharge point. In order to insure that nounplanned or unmonitored releases take place by way of the OilyWater Separator, RE-3 serves to monitor the discharge even whenno activity has been identified in the effluent. When no significantprimary to secondary leakage is taking place or when no activityhas been identified in the Oily Water Separator, the High AlarmSet Point for RE-3 is calculated as shown in Equation 8.

HASP~ 3= 3 X BEGD~ 3 (8)

CAPA8.DOC 08 9




UNITS 1 AND 2

In the event that primary to secondary leakage results in activitybeing detected in the Oily Water Separator, Equation 9 willbeused to calculate a High Alarm Set Point value. The greaterHASP value as determined by Equation 8 or Equation 9 willbeused.

HASP~, = BKGD„~, + (AF)(SF) x gk„C gc, zcl.,IwH3

x TCF (9)

Where:

HASPRE.3

BKGDRa.3

(AF)

high alarm setpoint for RE-3 (cpm)

background reading for RE-3 (cpm)

allocation factor for the oily water separatoreffluent system from Table 6. 1 (0.01)

(SF) safety factor for RE-3 (0.9)

monitor response factor (cpm/pCi/ml)

concentration of gamma emitting isotopes in therelease mix, pre-dilution (pCi/ml)

dilution flow rate (gpm)

undiluted effluent flow rate (gpm)

C; concentration of isotope "i," in the release mix,pre-dilution (pCi/ml)

ECL;

TCF

effluent concentration limitof isotope "i"

tritium correction factor as defined by Equation 7.

CAPA8.DOC 08 10




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 Systemliquid effluent radiation monitor is calculated as shown inEquation 10.

HASP'8 BICGDI ia + (AF)(SF) x g k> C|,gC, ECI,,islt3

x TCF (IP)

Where:

HASPRE.is high alarm setpoint for RE-18 (cpm)

BKGDRE.is —— background reading for RE-18 (cpm)

(AF)

(SF)

C;

ECL;

TCF

allocation factor for the liquid radwasteeffluent system from Table 6. 1 (0.9)

safety factor for RE-18 (0.9)

monitor response factor (cpm/NCi/ml)

concentration of gamma emitting isotopes inthe release mix, pre-dilution (pCi/ml)



concentration of isotope "i," in the release

mix, pre-dilution (pCi/ml)

effluent concentration limitof isotope "I"

tritium correction factor as defined byEquation 7.

CAPA8.DOC 08 11

. PACIFIC GAS AND ELECTRIC COMPANYDIABLOCANYONPOWER PLANT

TITLE: OFFGITE DOSE CALCULATIONS


UNITS 1 AND 2

3. Set Point Methodology for RE-23 HASP: Steam GeneratorBlowdown Tank

The High Alarm Set Point for the RE-23, Steam GeneratorBlowdown Tank liquid effluent radiation monitor, is calculated as

shown in Equation 11.

HASP~ „= BKGD~ „+(AF)(SF) x gk, C„gC, ECL,

iwH3

x TCF (11)

Where:

HASPaE.23 high alarm setpoint for RE-23 (cpm)

BKGDaE.23= background reading for RE-23 (cpm)

(AF)

(SF)

allocation factor for the steam generatorblowdown effluent system for each unit fromTable 6. 1 (0.04)

safety factor for RE-23 (0.9)

monitor response factor (cpm/pCi/ml)

concentration of gamma emitting isotopes inthe release mix, pre-dilution (pCi/ml)



C;

ECL;

TCF

concentration of isotope "i," in the release

mix, pre-dilution (pCi/ml)

effluent concentration limitof isotope "i"

tritium correction factor as defined byEquation 7.

CAPA8.DOC 08 12




UNITS 1 AND 2

6.1.4 Dose Projection (for Liquid Effluents)

The projected dose contributions from each reactor unit due to liquideffluents for the current calendar month, quarter and current calendar yearmust be determined in accordance with the methodology and parameters inthe ODCP at least every 31 days.

The purpose of this is to determine ifappropriate treatment of liquidradioactive materials in relation to maintaining releases "as low as reasonablyachievable," is necessary.

Projections willbe made, at least by the end of each month with attention tothe frequency requirement contained in the radiological effluent controlsprogram.

The projected dose from each reactor unit is given by:

Where:

P P,U 2 P,Coo> (12)

D„= Projected Dose.

DP U= Projected dose attributed to reactor unit, U.

DP c,„, = Projected dose common to both reactor units.

The 31-day projected dose is calculated by Equation 13.

Phf Chf CBDh' 31

D„+d„+ dP

(T+ t)(13)

Where:

Dphil= Monthly projected Dose

D~" = Previous Month's Actual Dose

d~' Current Month Actual Dose to date

dp —— Projected Dose from Current Batch Release

T = Number of days in the previous month

t = Number of days into the present month

CAPA8.DOC 08 13




UNITS 1 AND 2

Projected quarterly doses are determined by Equation 14.

D" d dz)CQ dCQ (9 )

A A P+ 2-t( )

Where:

(14)

g)CQP Projected dose for the current calendar

quarter.

Current quarter to date actual dose.

g) QPrevious quarter's actual dose.

dCBP 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.

a" =d" +(366-i) "(15)

Where:

ICYP Projected dose for the current calendar

year.

dCY Current year to date actual dose.

g) PY Previous year's actual dose.

dCBP Projected dose as a result of the current

batch release.

T = Number of days in the previous year.

t = Number of days into the presentyear.

CAPAS.DOC 08 14

PACIFIC GAS AND ELECTRIC COMPANYDIABLOCANYON POWER PLANT



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-23as well as for releases from the Condensate Demineralizer Regenerate wastetank or miscellaneous release points is established in order to provide furthercontrol over the effluent releases. The release rate limit is determined by the

effluent concentration and the 10 CFR 20 Effluent Concentration Limits(ELCs) as shown in Equation 16.

Where:

F (AF) (SF) (TCF)CiZ —,'

«-3 J-'Cli

(16)

Maximum operational undiluted liquid radwaste effluentdischarge flow rate (gpm)

AF

Expected dilution flow rate (gpm)

allocation factor for the liquid radwaste effluent sourcefrom Table 6. 1.

TCF

safety factor (0.9).

tritium correction factor as defined by Equation 7.

C; concentration of isotopes "i" in the release mix,pre-dilution (pCi/ml)

ECL; effluent concentration limit.of isotope "i" (pCi/ml)

When the term g —='0 then the Limiting Flow Rate is calculated by:CI

„„3ECI-I

f= F(AF)(SF)(TCF)

Where the terms are as previously defined.

(17)

CAPA8.DOC 08 15

PACIFIC GAS ANDELECTRIC COMPANYDIABLOCANYON POWER PLANT



UNITS 1 AND 2

6.1.6 Liquid Limiting Flow Rates - LLDBased

When there is no primary to secondary leakage, the Oily Water Separatorand various miscellaneous release points are assumed to be uncontaminated.Furthermore, in order to establish practical operational flow rate limits forany sources when they are considered uncontaminated, Equation 18 is used.While no activity may be present, Equation 18 assumes a concentration equalto the Lower Limitof Detection for the nuclides listed in CY2.ID1,Appendix 6.1, Table 6.1.3-1.

Where:

F (~F) (Sr)4.3

(18)

Maximum operational undiluted liquid radwaste effluentdischarge flow rate (gpm)-

Expected dilution flow rate (gpm)

AF allocation factor for the liquid radwaste effluent sourcefrom Table 6. 1.

SF

4.3

safety factor (0.9).

Total ECL fraction as given by:

Z—'LD(

I ECL,

Where:

LLD; = Lower limitof detection for isotope "i" from CY2.ID1,Appendix 6.1, Table 6.1.3-1. (pCi/ml)

ECL; =. effluent concentration limitof isotope "i" (pCi/ml)

CAPAS.DOC 08 16



NUMBER CAP A-SREVISION 22PAGE 17 OF 64

UNITS 1 AND 2

6.1.7 Unplanned Liquid Releases

An unplanned release is an unexpected and potentially unmonitored release tothe environment due to operational error or equipment malfunctions.

a. Unmonitored unplanned releases shall have a report written by theRadiochemistry Effluents Engineer describing the event with a

calculation, ifpossible, of the percent of Tech Spec release rate limit.This will then be forwarded to PSRC for review and then forwarded toNSOC for review.

b. Monitored unplanned releases which exceed 1% of the RMCP release

rate limitwillalso have a report written describing the event and mustbe forwarded to the PSRC for review, and then forwarded to NSOC forreview.

c. Describe these unplanned releases in the Annual Radioactive EffluentRelease Report. This report willalso be forwarded to PSRC andNSOC for review.

CAPA8.DOC 08 t7




UNITS 1 AND 2

6.2 Gaseous Effluents

The only significant path for gaseous radioactive releases to the environment duringnormal 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 radioiodinesand 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 fumehood, the main condenser Nash vacuum pump discharge, hot machine shop vent, etc.,'are considered miscellaneous release sources. These miscellaneous release sources arenot continuously monitored but can have dose rates and dose calculated for their path tothe unrestricted area.

6.2.1 Meteorological Methodology

The equations for determining gaseous effluent concentration limits, highalarm setpoints, dose rates, and critical receptor doses make use of thehistorical average atmospheric conditions in accordance with methodologiesof Regulatory Guides 1.109 and 1.111 and NUREGs 0133 and 0472. Thehistorical average dispersion (X/Q) and deposition (D/Q) values are derivedfrom the methodology of Regulatory Guide 1.111 as implemented byNUREG 2019 (computer code XOQDOQ). The DCPP dispersion anddeposition values are based on the latest five years of meteorological data andare updated when the value of X/Q or D/Q changes by more than ten percent.The present values are listed in Table 10.2.

Long-term releases are characterized as those that are generally continuousand stable in release rate, such as normal ventilation systems effluents.Doses due to long-term releases are modeled using historical annual averagedispersion and deposition values in accordance with the guidance ofRegulatory Guide 1.109, Regulatory Guide 1.111, NUREG 0133 andNUREG 0472.

Short-term releases are defined as those which occur for a total of 500 hoursor less in a calendar year but not more than 150 hours in any quarter. Inaccordance with NUREG 0133 and based upon an operational history thathas demonstrated short term gaseous releases can be characterized as randomin both time of day and duration, historical average atmospheric dispersionand deposition values are used to model doses due to short-term releases.

CAPA8.DOC '08 18




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 dueto radioactive materials released in gaseous effluents from the site toareas at or beyond the site boundary for noble gases to less than orequal 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 inparticulate 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 instantaneousconcentrations of radionuclides in gaseous effluents at the siteboundary.

l. Allocation and Safety Factors

The limits set forth by RCMP 6.1.6.1 are site limits which requirethat the set point methodology must ensure simultaneous releasesdo not exceed the off-site dose rate limits set forth by RCMP6.1.6.1(a) and 6.1.6.1(b). The DCPP High Alarm Set Pointmethodology makes use of an Allocation Factor (AF) to limit thenoble gas effluent dose rate from simultaneous atmosphericreleases.

The Allocation Factors can be adjusted based upon operationalrequirements with the following restrictions:

The sum of the Allocation Factors for RE-14, the SGBDtank vents, and miscellaneous release points from both unitsmust 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 IodineMonitor) and RE-28 (Plant Vent Particulate Monitor) are set

equal to the Allocation Factor for RE-14.

CAPA8.DOC 08 l9



NUMBER CAP'A-8REVISION 22PAGE 20 OF 64

UNITS I AND 2

~ Typical Allocation Factors are shown in Table 6. 2.

Table 6. 2

Typical Gaseous Effiuent Discharge Pathway Allocation Factors

Discharge PathwayPlant Vent - NG MonitorPlant Vent Iodine MonitorPlant Vent Part MonitorWaste Gas Decay TankContainment PurgeSGBD Tank VentMiscellaneous

Rad MonitorRE-14RE-24RE-28RE-22RE-44

Allocation Factor (AF)0.480.480.480.100.380.010.01

An additional level of conservatism in the HASP methodology is

implemented by the use of a Safety Factor (SF). The SafetyFactor is defined as 0.9 and provides for a High Alarm Set Pointat 90% of the dose rate limits.

CAPA8.DOC 08 20




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 monitorRE-14. RE-14 provides alarm function only.The High Alarm Set Point methodology for RE-14 is given byEquation 19.

" ' (NG) = AF SF472x F „x (Xig) x 294

Where:

(19)

4M'"C (NG) the maximum allowable total concentration of noble gas inthe plant vent, in pCi/cc

AF the allocation factor for the plant vent for one unit fromTable 6. 2 (0.48)

SF = a safety factor to ensure that dose rate limits of theradiological effluent controls are not exceeded (0.9)

500 = (mrem/yr) the site Total Body dose rate limit forinstantaneous releases

472 = the conversion constant to cc/sec from cfm

Fpp = total flow rate in the plant vent, in cfm (maximum plant ventflow rate is 263,000 cfm)

(xigi = the maximum historical site boundary dispersion factor,Sam

based on 5 year averages derived from the meteorologicaldata base, from Table 10.2.

294 = the whole body dose factor (mrem/yr/pCi/m ) for Xe-133 as

presented in Table 10.3, (for the plant vent HASP, therelease is assumed to be all Xe-133)

CAPA8.DOC 08 21




UNITS 1 AND 2

2. PLANT VENT NOBLE GAS MONITOR - RE-14 SCALING

In order to correlate the readings of RE-14 to noble gasconcentration during periods between samplings, the concentrationis scaled according to Equation 20.

CPMT

CPM(2o)

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).

Scaled concentration of noble gas (pCi/cc).

CAPAS. DOC 08 22




UNITS 1 AND 2

3. PLANT VENT IODINEMONITOR - RE-24

The Plant Vent Iodine concentration is monitored by rad monitorRE-24. RE-24 provides alarm function only. The alarm setpointmethodology is based upon the assumption that RE-24 responds

only to I-131. The methodology also presumes a release mixturebased upon the RCS source term.

The High Alarm Set Point methodology of RE-24 is given byEquation 21.

"""C (Iodine) =SFx AFx f 1500Y '''472xF x X/Q .gP"f (21)

PY Max . i iI

Where:

'"CT(iodine) the maximum allowable concentration ofI-131 in the plant vent

AF The allocation factor for the plant vent forone unit from Table 6. 2 (0.48)

SF

ft-131

A safety factor to insure that the dose ratelimits of the radiological effluent controlsare not exceeded (0.9).

fraction of the total non-noble gasconcentration that is due to 1-131. Definedas:

f CI }31I-1 31

I

(22)

CAPA8.DOC 08 23




UNITS I AND 2

1500 = (mrem/yr) the site organ dose rate limit forradioiodines and for all radioactivematerials in particulate form andradionuclides other than noble gases withhalf-lives greater than 8 days.


FP„ total flowrate in the plant vent, in cfm(maximum plant vent flowrate is

263,000 cfm)

x/g) = maximum historical site boundaryIhfar

dispersion factor, based on 5 year averagesderived from the meteorological database,from Table 10.2.

i

PW Inhalation dose factor for nuclide "i"(mrem/yr/pCi/m ) for child age group forworst case organ, from Table 10.4. Dosefactors are based upon NUREG 0133methodology. Inhalation dose conversionfactors are taken from Reg. Guide 1.109,Rev 1, Table E-9, with the followingexceptions: H-3, Sb-124 and Sb-125inhalation dose conversion factors takenfrom NUREG/CR-4013.

fraction of total non-noble gas concentration(excluding tritium) that is due to nuclide, i,and defined as:

C,f=-pc, (23)

CAPA8.DOC 08 24


TITLE: OFFAITE DOSE CALCULATIONS


UNITS I AND 2

4. PLANT VENT PARTICULATEMONITOR - RE-28

The Plant Vent Particulate concentration is monitored by radmonitor RE-28. The alarm setpoint methodology is based uponthe assumption of a 5% cross talk from the iodine channel. Thisis due to the retention of a small portion of iodine on the

particulate filter. A release mixture based upon the RCS sourceterm is also assumed.

The High Alarm Set Point methodology for RE-28 is given byEquation 24.

C (Particulates) = AFxSFx 0.05x f . + flodines Particulates

1500x

472xF x y/Q gP fI

(24)

Where:

'"Cr(Particulate) Maximum allowable particulate concentration inthe plant vent

AF The allocation factor for the plant vent for oneunit from Table 6. 2 (0.48)

SF A safety factor to insure that the dose rate limitsof the radiological effluent controls are notexceeded (0.9).

0.05

fiodines

Fraction of total iodine activity retained onparticulate filter.

Fraction of the total non-noble gas concentrationthat is due to iodines.

fParticuhtes Fraction of the total non-noble gas concentrationthat is due to particulates.

1500

472

FP„

(mrem/yr) the site organ dose rate limit forradioiodines and for all radioactive materials inparticulate form and radionuclides other thannoble gases with half-lives greater than 8 days.

Conversion constant to cc/sec from cfm

Total flowrate in the plant vent, in cfm (maximumplant vent flowrate is 263,000 cfm)

CAPAS. DOC 08 25




'NITS 1 AND 2

(x/g) = Maximum historical site boundary dispersionIhfax

factor, based on 5 year averages derived from themeteorological database, from Table 10.2

PW Inhalation dose factor for nuclide "i"(mrem/yr/NCi/m3) for child age group for worstcase organ, from Table 10.4. Dose factors arebased upon NUREG 0133 methodology.Inhalation dose conversion factors are taken fromReg. Guide 1.109, Rev 1, Table E-9, with thefollowing exceptions: H-3, Sb-124 and Sb-125inhalation dose conversion factors taken fromNUREG/CR-4013.

Fraction of total non-noble gas concentration(excluding tritium) that is due to nuclide, i, as

defined by Equation 23.

5. WASTE GAS DECAYTANKMONITOR - RE-22 HASP

Effluent releases from the Waste Gas Decay Tank are monitoredby rad monitor RE-22. RE-22 provides alarm and automaticrelease termination functions.

The High Alarm Set Point methodology for RE-22 is given byEquation 25.

'"'r(NG)= AFxSFx472 x F~ x (XiQ) x 1.34x10'25)

Where:

22M-C,(NG) the maximum allowable total concentration ofnoble gas in the gas decay tank exhaust line, inpCi/cc

472

F6d1

the conversion constant to cc/sec from cfm

maximum flow rate for the gas decay tank system(31 cfm)

AF = the allocation factor for the waste gas decay tankfor one unit from Table 6. 2 (0.10)

SF = safety factor to ensure that dose rate limits of theradiological effluent controls are not exceeded

(0.9)

3000 = (mrem/yr) the site Skin dose rate limit forinstantaneous releases

CAPA8.DOC 08 26




UNITS 1 AND 2

(y,/gl = the maximum historical site boundary dispersionIhoax

factor, based on 5 year averages derived from themeteorological data base from Table 10.2.

1.34x10 = -. the skin dose factor for Kr-85 (mrem/yr/ pCi/m,from Table 10.3.

CAPAS.DOC 08 27




UNITS I AND 2

6. CONTAINMENTPURGE - RE-44 HASP

The Containment Purge is monitored by rad monitor R&44.RE-44 provides alarm and automatic release terminationfunctions.

The HASP for RE-44 must limit the noble gas dose rate for skinand total body exposure. In order to address this, two set pointsare calculated. One set point is calculated based upon limiting thetotal body dose rate and the other limits the skin dose rate. Themore limiting set point is used. The High Alarm Set Pointmethodology for RE-44 is given by Equations 26 and 27.

a) Limiting Concentration Based on Total Body Dose

44AIEMmg (gG) gF gF472x Fg x(X/Qj x QKifi(NG)

(26)

Where: .

'"CT(NG) = the maximum allowable total co'ncentrationof noble gas in the containment purgeexhaust line, in pCi/cc

AF the allocation factor for the containment purgefor one unit from Table 6. 2 (0.38)

SF safety factor to ensure that dose rate limits ofradiological effluent controls are not exceeded

(0 9)

500 (mrem/yr) the site Total Body dose rate limit forinstantaneous releases


F„= maximum flow rate in the containment purgesystem (maximum containment purge flowrate is

55,000 cfm)

the maximum historical site boundary dispersion

factor, based on 5 year averages derived from themeteorological data base averages, fromTable 10.2.

CAPA8.DOC 08 28




UNITS 1 AND 2

K; = whole body dose factor (mrem/yr/pCi/m ) forisotope "i." Dose factors are presented inTable 10.3.

f; = the fraction of the concentration of the individualnoble gas radionuclide, "i," in the total mix ofnoble gas effluents in the containment purge line.

b) Limiting Concentration Based on Skin Dose

Skin dose should be calculated whenever an RCS samplecontains a significant fraction of Kr-85.

The concentration limitcalculated by this method shouldthen be compared to the concentration limitcalculated fortotal body dose. The smaller concentration should beselected as the limiting concentration for the HASP.

44~////,'~C (NG) = AF SF3000

(2'7)472 x F„x (X/Q) xg(L +11M') fj(7VG)

/

Where:

44A/BMaxc (NG) the maximum allowable total concentration ofnoble gas in the plant vent in pCi/cc

AF the allocation factor for the containment purgefor one unit from Table 6. 2 (0.38)

SF safety factor to ensure that dose rate limits of theradiological effluent controls are not exceeded

(0 9)

3000 = (mrem/yr) the site Total Body dose rate limit forinstantaneous releases.


F« = total flow rate in the containment purge system,in cfm (maximum containment purge flow rate is

55,000 cfm)

the maximum historical site boundary dispersion

factor, based on 5 year averages derived from themeteorological data base averages, fromTable 10.2.

CAPA8.DOC 08 29




UNITS 1 AND 2

L; = the skin dose factor (mrem/yr per pCi/m ) forisotope "i." Dose factors are presented inTable 10.3.

1.1 = Conversion factor mrem/mrad. Converts airdose to skin dose.

M; = the beta dose factor (mrad/yr per pCi/m ) forisotope 'i." Dose factors are presented inTable 10.3.

f; = the fraction of the concentration of the individualnoble gas radionuclide, "i," in the total mix ofnoble gas effluents in the containment purge line.

Mode 6 Particulate activity.

The HASP calculation specified in this section based upon Noble Gas

effluent limitations conservatively bounds the Tech. Spec. requirementfor particulate activity in Mode 6. The FSAR expected case accidentfor Mode 6 is a containment fuel handling accident which does notinclude a particulate release. Therefore, the HASP for RM-44 in thissection conservatively satisfies the Tech. Spec. (Ref.: 8.15)

CAPA8.DOC 08 30




UNITS 1 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 noblegases is given by:

(y/Q), gK,.Q,.(500 mrem/yearI

Where:

(28)

Q/ The release rate of radionuclide i in units ofpCi/sec.

Allother 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:

(y/Q) g(L.,+1.1M.,)Q.,<3000mrem/yearI

Where the terms are as previously defined.

(29)

CAPA8.DOC 08 31




UNITS 1 AND 2

c. Radioiodine, Tritium and Particulate Dose Rate Methodology

The dose rate to organ, o, due to radioiodines, tritium and particulatesreleased in gaseous effluents is given by:

(yiQ) gP,. Q, <1500 mrem/yearI

Where:

P;, = K'BR) DFA;,

and

(30)

K' 10 pCi/pCi

BR = The breathing rate of the a child age group inm /yr. The default value of 3700 m /yr is takenfrom Table E-5 of Reg. Guide 1.109.

DFA; The inhalation dose factor for organ o, for thechild age group for radionuclide, i, fromTable E-9 of Reg. Guide 1.109 in mrem/pCi, withthe following exceptions: H-3, Sb-124 and Sb-

125 inhalation dose conversion factors taken fromNUREG/CR4013.


Values for P;, are listed in Table 10.6.

CAPAS.DOC 08 32




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:

D = 3.17 x10 (X/Q), gM,Q,I

Where:

D,„= Gamma air dose in mrad.

3. 17x10 = Conversion constant yr/sec.

(31)

M; = Gamma air dose factor for nuclide i, in mrad/yrper pCi/m . Values are listed in Table 10.3.

Q, = Total release of noble gas radionuclide, i, in pCi.


b. Beta AirDose

The beta air dose due to immersion in a cloud of noble gases is givenby:

D~

——3.17x10 '(X/Q)„, gN,Q,I

Where:

(32)

D,ti = Beta air dose in mrad.

N; = Beta air dose factor for nuclide i, in mrad/yr perNCi/m . Values are listed in Table 10.3.


CAPAS.DOC 08 33




UNITS 1 AND 2

6.2.5 Dose To Critical Receptor Due To Radioiodines, Tritium and ParticulatesReleased in Gaseous Effluents

a. Calculation Methodology

The dose to an individual (critical receptor) due to radioiodines, tritiumand particulates released in gaseous effluents with half-lives greater than8 days is determined based upon the methodology described inNUREG 0133. This methodology makes use of the maximumindividual concept described in Regulatory Guide 1.109. Themaximum individual is characterized as maximum with regard to foodconsumption, occupancy, and other usage parameters. This concepttherefore models those individuals within the local population withhabits representing reasonable deviations from the average. In allphysiological and metabolic respects, the maximum individual isassumed 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 ofthe maximum individual. The critical receptor is defined as thatindividual that receives the largest dose based upon the combination ofdose pathways that have been shown to actually exist. The criticalreceptor concept is applied at that location where the combination ofdispersion'(X/Q), deposition (D/Q), existing pathways, occupancy time,receptor age group, and effluent source term indicates the maximumpotential exposure. The inhalation and ground plane exposurepathways are considered to exist at all locations. The grass-cow-milk,grass-cow-meat, and vegetation pathways are considered based on theiractual existence in the vicinityof the plant.

The dose pathways that have been shown to actually exist at DCPP arethe ground plane, inhalation and the vegetation pathways. These dose

pathways are reviewed yearly and updated based upon the annual landuse census survey in order to insure that actual exposure to anindividual willnot be substantially underestimated.

The locations of the pathways and descriptions are listed in Table 10.2.

CAPA8.DOC 08 34

PACIFIC GAS AND ELECTRIC COMPANYDIABLOCANYONPOVVER PLANT



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 maximumexposed individual (Critical Receptor) due to radioactive gaseouseffluent releases is calculated for all radionuclides identified in gaseous

effluents released to unrestricted areas using the followin'g expression:

D,~ = 3.17 x 10 'c„QR,. Q,.I

(33)

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 ) from Table 10.2.

Critical receptor D/g for ground plane and allingestion pathways (1/m ) from Table 10.2.

Raipo Site specific dose factor for age group a, radionuclidei, pathway p, and organ j (mrem/yr per pCi/m forinhalation and tritium pathways - mrem/yr perpCi/(sec m ) for ground plane and ingestionpathways). These dose factors are listed inTable 10.6.

The site specific dose factors are calculated based

upon NUREG 0133 methodology. Alldoseconversion 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 pCi.

CAPA8.DOC 08 35




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.

(XIQ)„ZI:iQiPRRL — ' 100%

(048) (500 mreinl yr)(34)

Where:

X/Ql = The maximum site boundary dispersion factor based on/hfdf

5 year averages from Table 10.2.

Whole body dose factor (mrem/yr per lhCi/m ) forisotope "i." Dose factors are presented in Table 10.3.

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.

CAPA8.DOC 08 3G




UNITS 1 AND 2

The Expected Reading (ER) is the anticipated monitor response based uponthe known plant vent concentration and the monitor response factors. TheExpected Readings for RE-22, RE-44, RE-14 are given by Equations 35, 36,aild 37.

BR(~ „) = BKG(~ „)+gk())F „)) C())) „),I

(35)

Where:

ER(RE-22)

BKG(RE.22) =

k(RE-22) i

Expected reading on monitor RE-22 in pCi/cc.

Monitor background.

Noble gas monitor response factor for nuclide "i" formonitor RE-22.

C(RE 22»= Concentration of nuclide "i" seen by RE-22.

ER(~ )4)——BKG(~ „) +CCSP(~ 44)gk(~ 4)), C(~ ~)) (36>

I

Where:

ER(REg4)

BKG(RE~4) =


Monitor background.

CCSP(RE~4) = Conversion constant setpoint for monitor RE-44.

k(RE44)i Noble gas monitor response factor for nuclide "i" formonitor RE-44.

C(REM)i Concentration of nuclide "i" seen by monitor RE-44.

ER())) )4>—BKG(() )))+ CCSP())E ))>gk())) )))) C())) ))))

I(37)

Where:

ER(RE-14)

BKG(RE.i4) =

CCSP(RE-14)


Monitor background.

Conversion constant setpoint for monitor RE-14.

CAPA8.DOC 08 37

PACIFIC GAS AND ELECTRIC COMPANYDIABLOCANYONPONER PLANT



UNITS 1 AND 2

ktRE-14) i Noble gas monitor response factor for nuclide "i" formonitor RE-14.

C<RE i4i;—— Concentration of nuclide "i" seen by monitor RE-14.

Generally ifthe Expected Reading (ER) is greater than the existing HASPsetting (an "administrative limit" as set by CY2.DC1) then no release shouldbe 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 andparticulates for each unit is calculated based upon the 1500 mrem/yr organdose rate limit. The dose rate is calculated for the inhalation pathway to thechild age group using the highest (worst case) organ dose factor for nuclide.The Percent Release Rate Limitbased on the worst case organ is given byEquation 38.

(x/Q)„,„Z11" QII'RRL, = x 100%

(0.48) (1500 >nrem/ yr)(38)

Where:

(x/Q)„=PW'he

maximum site boundary dispersion factor based on 5

year averages from Table 10.2.

Inhalation dose factor for nuclide "i" (mrem/yr/pCi/m3) forchild age group for worst case organ, from Table 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 followingexceptions: H-3, Sb-124 and Sb-125 inhalation doseconversion factors taken from NUREG/CR-4013.

Q/ 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.

CAPA8.DOC 08 38




UNITS 1 AND 2

6.2.8 Alternate Dose Methodologies

For purposes of routine gaseous effluent dose assessment, the methodology ofNUREG 0133 (described in Section 6.2.5) willbe used. However, DCPPmay elect to utilize the dose methodologies of Regulatory Guide 1.109 or theGASPAR computer code for special purposes such as evaluation of potentialnew 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 yearmust be determined in accordance with the methodology and parameters inthe ODCP at least every 31 days.

The purpose of this is to determine ifappropriate treatment of gaseousradioactive materials in relation to maintaining releases "as low as reasonablyachievable," is necessary.

Projections willbe made, at least by the end of each month with attention tothe frequency requirement contained in radiological effluent controlsprogram.

The projected dose from each reactor unit is given by:

Where:

>p = >p,v +is>p,c... (39)

Dp = Projected Dose.

Dp U Projected dose attributed to reactor unit, U.

Dp c Projected dose common to both reactor units.

CAPAS. DOC 08 39



NUMBER CAP A-8'REVISION 22PAGE 40 OF 64

UNITS 1 AND 2

The 31 day projected dose is calculated by Equation 40.

Dhi 31DA"+ dA '+dP

(T+ i)(40)

Where:

Projected Dose for the next 31 day period.

@Phd Previous Month's Actual Dose

Current Month Actual Dose to date

dcB Projected Dose from Current Batch Release

Number of days in the previous month

Number of days into the present month

Projected quarterly doses are determined by Equation 41.

a"Q+d" +d"gCQ dCQ (92 )

A A P(41)

Where:

L)CQProjected dose for the current calendar quarter.

d Current quarter to date actual dose.

g)PQPrevious quarter's actual dose.

dcBProjected dose as a result of the current batch release.

T = Number of days in the previous quarter.

t = Numberofdays,into thepresentquarter.

CAPA8.DOC 08 40




UNITS 1 AND 2

Projected yearly doses are determined by Equation 42.

D""+1 Y+1D Y =1 -Y+(366-t)

P A (y+()(42)

Where:

g)CY Projected dose for the current calendar year.

1CY

DPY

Current year to date actual dose.

Previous year's actual dose.

1cBProjected 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

An unplanned release is an unexpected and potentially unmonitoredrelease to the environment due to operational error or equipmentmalfunctions.

1. Unmonitored unplanned releases shall have a report written by theRadiochemistry Effluents Engineer describing the event with a

calculation, ifpossible, of the percent of RMCP release rate limit.This will then be forwarded to PSRC for review and thenforwarded to NSOC for review.

2. Monitored unplanned releases which exceed 1% of the RMCPrelease rate limitwillalso have a report written describing theevent and must be forwarded to the PSRC for review, and thenforwarded to NSOC for review. For purposes of classificationonly, unplanned release puffs through the plant vent may use onehour integrated resolution times.

3. Describe these unplanned releases in the Annual radioactiveeffluent release report. This report willalso be forwarded toPSRC and NSOC for review.

CAPA8.DOC 08 41


TITLE: OFF-SITE DOSE CALCULATIONSI

NUMBER CAP A-8REVISION 22PAGE " 42 OF 64

UNITS 1 AND 2

6.3 40 CFR 190 Dose Calculations

6.3.1 Pathways

Calculation of total uranium fuel cycle dose for purposes of demonstratingcompliance with 40 CFR 190 requires the contributions from liquid and

gaseous effluent as well as direct radiation from the units and outside storagetanks to be considered. The total uranium fuel cycle dose to any member ofthe public willbe calculated by summing the following doses:

~ Direct Radiation Dose

~ Liquid Effluent Dose

~ Noble Gas Dose

~ Radioiodine, Tritium and Particulate Gaseous Effluent Dose

6.3.2 Methodology

Direct Radiation Dose

Determination of direct radiation dose from the reactor units and fromoutside storage tanks may be made by direct survey measurements,derived from environmental TLD data, or calculated by shielding code.

The direct radiation dose willalso take into account residence timesnear the site based upon land use census information.

The direct radiation determination using environmental TLD is given byequation 43.

2

b + X 9/7 X e-6.38 (43)

where:

D', b= the dose rate at the site boundary

D'~ = the dose rate from the dosimetry reading

r„, = the distance from the point source to the dosimetry

CAPA8.DOC 08 42




UNITS 1 AND2

b. Noble Gas Dose

The noble gas skin dose and total body dose contributions to the totaluranium fuel cycle dose to a member of the public willbe determinedas shown in Equations 44 and 45.

Noble Gas Total Body Dose = 3.17 x 10'(X/L7) g K,g, (44)I

Noble Gas Skin Dose = 3.17x)0'(X/Q) g(L, +1)M ) g, (aS)R

I

Where:

3.17x10 Conversion constant yr/sec.

(x/())„

K;

Maximum historical dispersion factor forreceptor of interest, based on 5 year averagesfrom Table 10.2.

Gamma air dose factor for nuclide i, in mrem/yrper pCi/m . Values are listed in Table 6.3.

Gamma air dose factor for nuclide i, in mrem/yrper pCi/m . Values are listed in Table 6.3.

Conversion factor mrem/mrad. Converts airdose to skin dose.

M; = Gamma air dose factor for nuclide i, in mrad/yrper pCi/m . Values are listed in Table 10.3:

QI Total release of noble gas radionuclide, i, inNCi/sec.

CAPA8.DOC 08 43




UNITS 1 AND 2

Liquid and Gaseous Effluent Dose

The doses from liquid effluents and radioiodines, tritium and

particulates in gaseous effluents willbe determined by Equations 1 and

33, respectively.

For purposes of calculating the dose required by the radiologicaleffluent controls, more realistic assumptions concerning the liquid and

gaseous effluent dose pathways willbe used, based upon the mostrecent land use census data as well as the latest environmentalmonitoring information.

These assumptions may include, but not be limited to: more realisticliquid dilution factors; location and age of actual individuals, sitespecific food pathway parameters, and documentation of true foodconsumption. Other assumptions may be used provided they can besubstantiated by census or direct measurement.

CAPAS.DOC 08 44


TITLE: OFF-SITE DOSE CALCULATIONSi

j


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, butonly in the owner controlled area up to the protected area boundary. The mostcommon public access activities are: tours to the simulator (training building) or BioLab, policemen using the shooting range (most frequent activity), cattle drives throughto adjacent properties, and visits of American Indians to on-site burial grounds (closestto the plant).

Exposure to members of the public due to liquid releases while on-site is highly unlikelyand therefore not addressed. Exposure due to gaseous releases and direct radiation arecredible and therefore are considered.

The dose to members of the public during on-site activities willbe 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 R s

in Equation 33 consider only the inhalation and ground plane pathway and exclude theinfant age group.

The X/Q and D/Q values are modified using logarithmic extrapolation from the siteboundary to the on-site location of interest as shown in Equations 46 and 47.

log [X/Q ]

log [X/Q ]> >—log [X/Q ]ttoe [log (dist on - site.) —iog (dist S tt)]..

log (dist.S.B.) —log (dist. loc.)

+ log [X/Q ]

log [D/Q]

log [D//Q]» —log [D/Q],"'log (dist on - site) —iog (dt'st g tt)]..log (dist.S./3.) —log (dist. loc.)

+ log [D/Q]

(47)

Based upon Regulatory Guide 1.111, these equations can be expected to providereasonable dispersion and deposition estimates for distances as close as 200 meters.

Determination of direct radiation dose from the reactor units and from outside storagetanks may be made by direct survey measurements, derived from environmental TLDdata, 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 willbe calculated.

CAPAS.DOC 08 45




UNITS 1 AND 2

The activities of the members-of-the-public while on-site (described above), are at orbeyond 200 meters. Table 6. 3 details the types of on-site activities thatmembers-of-the-public might be expected to participate in at DCPP. The sectors andclosest distances in which they may visit as well as expected visitation duration are alsoshown (based on Security Section information).

Table 6. 3

Expected On-Site Distances and Visitation Times for Members of the Public

ONSITE MEMBEROF THE PUBLIC

Police atshooting range

Tour Participants

Simulator Bldg

Bio Lab

Overlook

American Indiansat burial grounds

Ranch hands drivingcattle around site

SECTOROF

VISITATION

SE

S (SE)

SSE (SE)

E

NWNNW

NWNNWNNNENE

CLOSEST POINTOF APPROACH

TO PLANT

700m

310m

460m

210m

200m200m

250m350m320m450m630m

AVERAGE EXPECTEDVISITATION

TIME PER YEAR

208 hours

4 hours

6 hours

1 hours

96 hours96 hours

1 hour1 hour1 hour1 hour1 hour

i

CAPAS.DOC 08 46




UNITS 1 AND 2

7. ACCEPTANCE CRITERIA

7.1 There is no quantitative acceptance for this procedure. Ifthe task or analysis has been

accomplished within the bounds of this procedure, it is considered acceptable.

8. REFERENCES

8.1

8.2

8.3

8.4

8.5

8.6

8.7

Draft Radiological Tech Specs for PWRs, NUREG No. 0472; May 1978;

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.

Calculation of Annual Doses to Man From Routine Releases of Reactor Effluents forthe Purpose of Evaluating Compliance with 10 CFR 50, Appendix I, Regulato'ryGuide 1.109, Rev. 1, October 1977.

Preparation of Radiological Effluent Tech Specs for Nuclear Power Plants,NUREG No. 0133, October 1978.

LADTAPII - Technical Reference and User Guide, NUREG/CR-4013.

Methods for Demonstrating LWR Compliance with the EPA Uranium Fuel CycleStandard 40 CFR 190, NUREG No. 0543, January 1980.

Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents inRoutine 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-S, "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.DC1, "Radiation Monitoring System High Alarm Setpoint Control Procedure."

8.14 CY2.ID1, "Radiological Monitoring and Controls Program" (RMCP)

8.15 "Setpoint Calculation for Containment Ventilation Exhaust Monitor,"Calc ¹ NSP-18'-39-44, 10/92 and 11/92 and AR A0430610.

RECORDS

9.1 Data Sheets and records willbe maintained in the Records Management System (RMS)in accordance with CY1.DC1, "Analytical Data Processing Responsibilities."

CAPAS.DOC 08 47




UNITS I AND 2

10. APPENDICES

10.1 Tables

11. ATTACHMENTS

None

12. SPONSOR

Jeff Gardner

CAPA8.DOC 08 48

PACIFI AND ELECTRIC COMPANYDIABLOCANYONPOWER PLANT


NUMBER CREVISION 22PAGE 49 OF 64UNITS I AND 2

FIGURE 6.1LIQUIDDISCHARGES (LRW) MONITORED FOR RADIOACTIVITY

LIQUIDRADWASTE SYSTEM CONDENSATE DEMINERALIZER STEAM GENERATORREGENERATE TANKS BLOWDOWN

BA EVAPFEED DEMONS

HIGH LOWCONDVCllVITY CONDVCTMlY

MOBAERADWASTE TANK TANKPROCESSING SYSTEM(MRPSj

SGRDTANK

STEAM CENERATO

PRIMARYSYS.PRIMARY

WATER STORATANK

WASTE BORIC ACID

RETURN TOCONDENSOR

CLOSED DRAINS

MEOT

SCBT

EDR

SURFACECAPACITY

FOR EDR/FBA

LAUNDRYDISTAIATE

TANK

OILYWATERSEPARATOR

DEMWREGENRECVR

CARMEIXAFATER

MIXEDBED

OEM/NB

PROCESSWASTE

RECEIVER

OILYWATER

SEPARATOR

RES

EOIAPMENTDRAW

RECEIVER

P A

LAUNDRYdHOT

I~ F F

FCV4IT RCV-Id

RE Id

ONCE THROVGH COOUNG WATER

DIABLOCOVE

00692302

CAPA8.DOC 08 49

PACIFIC AND ELECTRIC COMPANYDIABLOCANYONPOWER PLANT


NUMBER CAPREVISION 22PAGE 50 OF 64UNITS 1 AND 2

FIGURE 6.2GASEOUS RELEASES (GRW) MONITORED FOR RADIOACTIVITY

II PZR

REVEFTANK

II r-II II I REACTII

DRAWTANKII

I II II II IIIIlIIIII II II II II II II II II II II II IIIII

I VOL

I

R

SURGE TANKI

~Qm

.0

I uaUID

I I I

I I I

WASTE1.'I

COMPRESSORMCNSTSEP.

WASTE01

COMPRESSORMOISTSEP.

HOLDUP

N1 I

@ --Q.-gVGISD

TANK I

--g-'GVO

IHOLDUP

TAIJK I01

II

ISSCEGVPDRASITANK

N2

okCOMPONENT

I I WATERHEAD 'fARK

XD P.

Qmaav +DECAY

HEPAFILTER

GASDECAYTANK

ppppppSIG 14~

I~SIG14

I DVMPTO ATMOSPHERE DVMP 'fO ATMOSPHERE DVMPTO ATMOSPHERE

~:„pxog P P P,",". xoxoxoxo pxoxoxoxo"„".".'=,

I II

@n. 22.TS.S T4I I I

ISJG ATMOS.1.1

v UNIT2

III VNIT2II

II

GASDECAYTANK

L 24 I

15R IS

PIANTVENTTO

ATMOSPHERE

14/14R24J24R24J2SRST20

44A

ATMOS.44B

FUEL HANDVNGBLDG. VENT

III

SJG BD TANKI GIANDSTEAM

ATMOS.

00692319

CAPA8.DOC 08 50



TITLE: OFF-SITE DOSE CALCULATIONS UNITS 1AND2

Nuclide Tot Body

TABLE 10.1

LRW COMPOSITE DOSE FACTORS A'<)

FOR ADULTS AT A SALTWATERSITE(mrem/hour per pCi/ml)

organ "o"

Thyroid Kidney Lung GI-LLI Bone LiverH-3Na-24Cr-51Mn-54Mn-56Fe-55Fe-59Co-57Co-58Co-60Ni-65CU-64Zn-65Zn-69As-76Br-82Br-84Rb-86Rb-88Rb-89Sr-89Sr-90Sr-91Sr-92Y-90Y-91mY-92Zr-95Zr-97Nb-95Mo-99Tc-101Ru-103Ru-105Ru-106Ag-110mSn-113Sn-117mSb-122Sb-124

1.61E-01 1.61E-01 1.61E-01 1.61E-01 1.61E-01 0.00E+00 1.61E-014.57E-01 4.57E-01 4.57E-01 4.57E-01 4.57E-01 4.57E-01 4.57E-01-5.58E+00 3.34E+00 1.23E+00 7.40E+00 1.40E+03 O.OOE+00 0.00E+001.35E+03 O.OOE+00 2.10E+03 O.OOE+00 2.16E+04 0.00E+00 7.06E+033.15E+01 0.00E+00 2.26E+02 0.00E+00 5.67E+03 0.00E+00 1.78E+028.23E+03 0.00E+00 O.OOE+00 1.97E+04 2.03E+04 5.11E+04 3.53E+047.27E+04 0.00E+00 0.00E+00 5.30E+04 6.32E+05 8.06E+04 1.90E+052.36E+02 0.00E+00 O.OOE+00 O.OOE+00 3.59E+03 0.00E+00 1.42E+021.35E+03 O.OOE+00 O.OOE+00 0.00E+00 1.22E+04 O.OOE+00 6.03E+023.82E+03 O.OOE+00 0.00E+00 0.00E+00 3.25E+04 O.OOE+00 1.73E+031.20E+01 0.00E+00 0.00E+00 0.00E+00 6.65E+02 2.02E+02 2.62E+011.01E+02 0.00E+00 5.40E+02 0.00E+00 1.83E+04 0.00E+00 2.14E+022.32E+05 O.OOE+00 3.43E+05 O.OOE+00 3.23E+05 1.61E+05 5.13E+054.56E+01 O.OOE+00 4.26E+02 O.OOE+00 9.85E+Ol 3.43E+02 6.56E+024.42E+01 O.OOE+00 8.72E+01 0.00E+00 O.OOE+00 0.00E+00 4.62E+014.07E+00 0.00E+00 O.OOE+00 0.00E+00 4.67E+00 0.00E+00 O.OOE+009.39E-02 0.00E+00 0.00E+00 0.00E+00 7.37E-07 0.00E+00 0.00E+00

2.91E+02 0.00E+00 O.OOE+00 0.00E+00 1.23E+02 O.OOE+00 6.24E+029.49E-01 0.00E+00 0.00E+00 0.00E+00 2.47E-11 0.00E+00 1.79E+008.34E-01 0.00E+00 0.00E+00 0.00E+00 6.89E-14 0.00E+00 1.19E+001.43E+02 0.00E+00 O.OOE+00 0.00E+00 8.00E+02 4.99E+03 0.00E+002.83E+03 0.00E+00 0.00E+00 0.00E+00 3.55E+03 1.41E+05 O.OOE+003.71E+00 0.00E+00 0.00E+00 0.00E+00 4.37E+02 9.18E+01 O.OOE+001.51E+00 0.00E+00 0.00E+00 0.00E+00 6.90E+02 3.48E+01 0.00E+001.63E-01 0.00E+00 O.OOE+00 0.00E+00 6.42E+04 6.06E+00 0.00E+002.22E-03 0.00E+00 O.OOE+00 O.OOE+00 1.68E-01 5.73E-02 0.00E+001.56E-02 0.00E+00 0.00E+00 0.00E+00 9.32E+03 5.32E-01 0.00E+00

3.46E+00 0.00E+00 8.02E+00 0.00E+00 1.62E+04 1.59E+01 5.11E+008.13E-02 0.00E+00 2.68E-01 0.00E+00 5.51E+04 8.81E-01 1.78E-011.34E+02 0.00E+00 2.46E+02 0.00E+00 1.51E+06 4.47E+02 2.49E+022.43E+01 0.00E+00 2.89E+02 O.OOE+00 2.96E+02 0.00E+00 1.28E+021.88E-01 0.00E+00 3.46E-01 9.81E-03 5.77E-14 1.33E-02 1.92E-02

4.60E+01 O.OOE+00 4.07E+02 0.00E+00 1.25E+04 1.07E+02 O.OOE+003.51E+00 0.00E+00 1.15E+02 0.00E+00 5.44E+03 8.89E+00 0.00E+002.01E+02 0.00E+00 3.06E+03 O.OOE+00 1.03E+05 1.59E+03 0.00E+008.60E+02 O.OOE+00 2.85E+03 0.00E+00 5.91E+05 1.56E+03 1.45E+033.53E+03 9.85E+02 O.OOE+00 0.00E+00 O.OOE+00 6.06E+04 1.66E+038.76E+02 2.52E+02 O.OOE+00 0.00E+00 O.OOE+00 3.02E+03 3.41E+026.65E+00 3.09E-01 0.00E+00 1.18E+01 O.OOE+00 2.19E+01 4.47E-011.09E+02 6.70E-01 0.00E+00 2.15E+02 7.84E+03 2.76E+02 5.22E+00

Sb-125 4.20E+01h

1.79E-01 O.OOE+00 1.36E+02 1.94E+03 1.77E+02 1.97E+00

CAPA8.DOC 08 51




UNITS 1 AND 2

Nuclide

LRW COMPOSITE DOSE FACTORS, A;„FOR ADULTS AT A SALTWATERSITE

(mrem/hour per pCi/ml)organ "o"

Tot Body Thyroid Kidney Lung GI-LLI Bone . LiverTe-129mTe-129Te-131mTe-131Te-132I-131I-132I-133I-134I-135Cs-134Cs-136Cs-137Cs-138Ba-139Ba-140Ba-141La-140La-142Ce-141Ce-143Ce-144Pr-144Nd-147

'-187N -239

1.47E+02 3.20E+02 3.89E+03 0.00E+00 4.69E+03 9.31E+02 3.47E+026.19E-01 1.95E+00 1.07E+01 0.00E+00 1.92E+00 2.54E+00 9.55E-01

5.71E+01 1.08E+02 6.94E+02 O.OOE+00 6.80E+03 1.40E+02 '6.85E+015.03E-01 1.31E+00 6.99E+00 0.00E+00 2.26E-01 1.59E+00 6.66E-011.24E+02 1.46E+02 1.27E+03 0.00E+00 6.24E+03 2.04E+02 1.32E+021.79E+02 1.02E+05 5.35E+02 0.00E+00 8.23E+01 2.18E+02 3.12E+029.96E+00 9.96E+02 4.54E+01 0.00E+00 5.35E+00 1.06E+01 2.85E+013.95E+01 1.90E+04 2.26E+02 0.00E+00 1.16E+02 7.45E+01 1.30E+025.40E+00 2.62E+02 2.40E+01 0.00E+00 1.32E-02 5.56E+00 1.51E+012.24E+01 4.01E+03 9.75E+01 0.00E+00 6.87E+01 2.32E+01 6.08E+011.33E+04 0.00E+00 5.27E+03 1.75E+03 2.85E+02 6.84E+03 1.63E+042.04E+03 0.00E+00 1.57E+03 2.16E+02 3.21E+02 7.16E+02 2.83E+037.85E+03 O.OOE+00 4.07E+03 1.35E+03 2.32E+02 8.77E+03 1.20E+045.94E+00 0.00E+00 8.81E+00 8.70E-01 5.12E-05 6.07E+00 1.20E+012.30E-01 O.OOE+00 5.23E-03 3.17E-03 1.39E+01 7.85E+00 5.59E-031.08E+02 O.OOE+00 7.02E-01 1.18E+00 3.38E+03 1.64E+03 2.06E+001.29E-01 O.OOE+00 2.68E-03 1.63E-03 1.80E-09 3.81E+00 2.88E-032.10E-01 0.00E+00 O.OOE+00 0.00E+00 5.83E+04 1.57E+00 7.94E-019.13E-03 0.00E+00 0.00E+00 0.00E+00 2.68E+02 8.06E-02 3.67E-022.63E-01 O.OOE+00 1.08E+00 0.00E+00 8.86E+03 3.43E+00 2.32E+004.94E-02 0.00E+00 1.97E-01 0.00E+00 1.67E+04 6.04E-01 4.46E+029.59E+00 O.OOE+00 4.43E+01 0.00E+00.6.04E+04 1.79E+02 7.47E+019.64E-04 O.OOE+00 4.44E-03 0.00E+00 2.73E-09 1.90E-02 7.87E-032.74E-01 O.OOE+00 2.68E+00 0.00E+00 2.20E+04 3.96E+00 4.58E+00

2.68E+00 O.OOE+00 0.00E+00 0.00E+00 2.51E+03 9.16E+00 7.66E+001.91E-03 0.00E+00 1.08E-02 0.00E+00 7.11E+02 3.53E-02 3.47E-03

Dose factors are based upon NUREG 0133 methodology.

CAPA8.DOC 08 52




UNITS '1 AND 2

TABLE10.2SUMMARYOF LANDUSE CENSUS EVALUATION

Sector

N

NNE

NE

ENE

SE

SSE

SSW

SW

WSW

WNW

NNW

ReceptorDescription

none

residence

residence

residence

residence

residence

residence

residence

residence +garden

oat hay andsugar peas

none

none

none

none

none

none

none

none

highest siteboundary

dispersion value

residence

residence

residence

Distance(miles)

443.3

4.9

4.4

5.0

4.0

3.7

4.5

3.3

0.5

1.2

3.6

1.5

5.56E-08

8.55E-08

4.39E-08

4.94E-08

3.58E-08

4.86E-08

5.92E-08

4.10E-08

1.88E-07

5.24E-06

1.57E-06

2.19E-07

5.89E-07

D/Q

1.00E-10

1.67E-10

8.10E-11

1.02E-10

6.80E-11

1.30E-10

1.65E-10

1.05E-10

1.20E-09

2.00E-08

6.31E-09

7.10E-10

1.85E-09

Comments

no receptors within 5 miles

unknown use

cabin - limited use

full'time occupancy

full time occupancy

full time occupancy

cabin - unknown use

part time occupancy

full time occupancy

field workers present only during the

day - critical receptor ground plane,inhalation, and vegetation ingestiondose assessed at this location

no receptors within 5 miles

over water

over water

over water

over water

over water

over water

over water

Gas effluent dose rates. PRRLs andHASPs evaluated at this location.

cabin - very limited use

full time occupancy

full time occupancy (trailer) - criticalreceptor ground plane and inhalationdose assessed at this location

CAPA8.DOC 08 53

PAC S AND ELECTRIC COMPANYDIABLOCANYONPOWER PLANT


NUMBER CAREVISION 22PAGE 54 OF 64UNITS 1 AND 2

TABLE 10.3

GRW DOSE FACTORS FOR NOBLE GASES

Radionuclide

Kr-83mKr-85mKr-85Kr-87Kr-88Kr-89Kr-90Xe-131mXe-133mXe-133Xe-135mXe-135Xe-137Xe-138Areal

Whole BodyDose Factor

K;(mrem/yr per pCi/m )

7.56E-021.17E+031.61E+015.92E+031.47E+041.66E+041.56E+049.15E+012.51E+022.94E+023.12E+031.81E+031.42E+038.83E+038.84E+03

Skin Dose FactorL;

(mrem/yr per pCi/m )

1.46E+031.34E+039.73E+032.37E+031.01E+047.29E+034.76E+029.94E+023.06E+027.11E+021.86E+031.22E+044.13E+032.69E+03

Gamma AirDose Factor

M;(mrad/yr per pCi/m )

1.93E+011.23E+031.72E+016.17E+03

- 1.52E+041.73E+041.63E+041.56E+023.27E+023.53E+023.36E+031.92E+031.51E+039.21E+039.30E+03

Beta AirDose Factor

N;(mrad/yr per pCi/m )

2.88E+021.97E+031.95E+031.03E+042.93E+031.06E+047.83E+031.11E+031.48E+031.05E+037.39E+022.46E+031.27E+044.75E+033.28E+03

From Table B-1 of Regulatory Guide 1.109 (Rev. 1, Oct. 1977)

CAPA8.DOC 08 54




UNITS 1 AND 2

TABLE 10.4

CHILD INHALATIONPATHWAYDOSE FACTORS FOR WORST CASE ORGAN

RadionuclideH-3CR-51MN-54FE-59CO-58CO-60ZN-65RB-86SR-89Y-90SR-90ZR-95NB-95RU-103RU-106AG-110MSB-124SB-125TE-129MI-131I-133CS-134CS-136CS-137BA-140CE-141CE-144ND-147

p 1Y

i

6.40E+021.70E+041.58E+061.27E+061.11E+067.07E+069.95E+051.98E+052.16E+062.68E+053.85E+072.23E+066. 14E+056.62E+051.43E+075.48E+063.24E+062.32E+061.76E+061.62E+073.85E+061.01E+061.71E+059.07E+051.74E+065.44E+051.20E+073.28E+05

CAPA8.DOC 08 55



NUMBERREVISIONPAGE

UNITS

CAP A-82256 OF 64

1AND2

Nuclide

TABLE 10.5

GROUND PLANE DOSE FACTORS

GRW DOSE PARAMETERS Ri.ot FOR RADIOIODINES, RADIOACTIVEPARTICULATES, AND ANYRADIONUCLIDEOTHER THAN

NOBLE GAS (IPT), GASEOUS EFFLUENTS(GRW),'NY

AGE GROUP, GROUND PLANE PATHWAY(mrem/yr per pCi/(sec m'))

Bone Liver T Body Thyroid Kidney Lung GI-LLIH-3 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00CR-51 4.65E+06 4.65E+06 4.65E+06 4.65E+06 4.65E+06MN-54 1.38E+09 1.38E+09 1.38E+09 1.38E+09 1.38E+09FE-59 2.73E+08 2.73E+08 2.73E+08 2.73E+08 2.73E+08CO-58 3.80E+08 3.80E+08 3.80E+08 3.80E+08 3.80E+08

0.00E+004.65E+061.38E+092.73E+083.80E+08

0.00E+004.65E+061.38E+092.73E+083.80E+08

CO-60ZN-65RB-86SR-89Y-90SR-90 0.00E+00 O.OOE+00ZR-95 2.45E+08 2.45E+08NB-95 1.37E+08 1.37E+08RU-103 1.08E+08 1.08E+08RU-106 4.20E+08 4.20E+08

0.00E+002.45E+081.37E+081.08E+084.20E+08

2.15E+10 2.15E+10 2.15E+107.46E+08 7.46E+08 7.46E+088.98E+06 8.98E+06 8.98E+062.16E+04 2.16E+04 2.16E+044.50E+03 4.50E+03 4.50E+03

2.15E+107.46E+088.98E+062.16E+044.50E+03

2.15E+107.46E+088.98E+062.16E+044.50E+03

0.00E+00 0.00E+002.45E+08 2.45E+081.37E+08 1.37E+081.08E+08- 1.08E+084.20E+08 4.20E+08

2.15E+107.46E+088.98E+062.16E+044.50E+030.00E+002.45E+081.37E+081.08E+084.20E+08

2.15E+107.46E+088.98E+062.16E+044.50E+030.00E+002.45E+081.37E+081.08E+084.20E+08

AG-110M 3.45E+09 3.45E+09 3.45E+09 3.45E+09 3.45E+09SB-124 5.99E+08 5.99E+08 5.99E+08 5.99E+08 5.99E+08SB-125 2.34E+09 2.34E+09 2.34E+09 2.34E+09 2.34E+09TE-129M 1.98E+07 1.98E+07 1.98E+07 1.98E+07 1.98E+07I-131 1.72E+07 1.72E+07 1.72E+07 1.72E+07 1.72E+07I-133 2.45E+06 2.45E+06 2.45E+06 2.45E+06 2.45E+06CS-134 6.90E+09 6.90E+09 6.90E+09 6.90E+09 6.90E+09CS-136 1.51E+08 1.51E+08 1.51E+08 1.51E+08 1.51E+08CS-137 1.03E+10 1.03E+10 1.03E+10 1.03E+10 1.03E+10BA-140 2.05E+07 2.05E+07 2.05E+07 2.05E+07 2.05E+07CE-141 1.37E+07 1.37E+07 1.37E+07 1.37E+07 1.37E+07CE-144 6.96E+07 6.96E+07 6.96E+07 6.96E+07 6.96E+07ND-147 8.39E+06 8.39E+06 8.39E+06 8.39E+06 8.39E+06

3.45E+095.99E+082.34E+091.98E+071.72E+072.45E+066.90E+091.51E+081.03E+102.05E+071.37E+076.96E+078.39E+06

3.45E+095.99E+082.'34E+091.98E+071.72E+072.45E+066.90E+091.51E+081.03E+102.05E+071.37E+07„6.96E+078.'39E+06


CAPAS.DOC 08 56


TITLE OFF-SITE DOSE CALCULATIONS


UNITS 1 AND 2

TABLE 10.6

GRW DOSE PARAMETERS FOR RADIOIODINES, RADIOACTIVEPARTICULATES, ANDANYRADIONUCLIDEOTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW)

CAPA8.DOC 08 57




UNITS 1 AND 2

TABLE 10.6.1

GRW DOSE PARAMETERS FOR RADIOIODINES, RADIOACTIVEPARTICULATES, AND ANYRADIONUCLIDEOTHER THANNOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), INFANT

AGE GROUP, INHALATIONPATHWAYORGAN "0" (mrem/yr per pCi/m ) Rg ( y t

Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLIH-3 O.OOE+00 3.68E+02 3.68E+02CR-51 0.00E+00 0.00E+00 8.95E+01MN-54 O.OOE+00 2.53E+04 4.98E+03FE-59 1.36E+04 2.35E+04 9.48E+03CO-58 O.OOE+00 1.22E+03 1.82E+03

3.68E+025.75E+010.00E+000.00E+000.00E+00

3.68E+02 3.68E+02 3.68E+021.32E+01 1.28E+04 3.57E+024.98E+03 1.00E+06 7.06E+030.00E+00 1.02E+06 2.48E+040.00E+00 7.77E+05 1. 11E+04

CO-60ZN-65RB-86SR-89Y-90

0.00E+00 8.02E+03 1.18E+041.93E+04 6.26E+04 3.11E+04O.OOE+00 1.90E+05 8.82E+043.98E+05 O.OOE+00 1. 14E+043.29E+03 O.OOE+00 8.82E+01

0.00E+000.00E+000.00E+000.00E+000.00E+00


SR-90ZR-95NB-95RU-103RU-106

1.55E+071.15E+051.57E+042.02E+038.68E+04

0.00E+00 3.12E+052.79E+04 2.03E+046.43E+03 3.78E+03O.OOE+00 6.79E+020.00E+00 1.09E+04

O.OOE+000.00E+000.00E+00O.OOE+00O.OOE+00

O.OOE+00 1.12E+07 1.31E+053.11E+04 1.75E+06 2.17E+044.72E+03 4.79E+05 1.27E+044.24E+03 5.52E+05 1.61E+041.07E+05 1.16E+07 1.64E+05

AG-110M 9.98E+03 7.22E+03SB-124 3.79E+04 5.56E+02SB-125 5.17E+04 4.77E+02TE-129M 1.41E+04 6.09E+03I-131 3.79E+04 4.44E+04

5.00E+031.20E+041.09E+042.23E+031.96E+04

0.00E+001.01E+026.23E+015.47E+031.48E+07

1.09E+04 3.67E+060.00E+00 2.65E+060.00E+00 1.64E+063.18E+04 1.68E+065.18E+04 O.OOE+00

3.30E+045.91E+041.47E+046.90E+041.06E+03

I-133CS-134CS-136CS-137BA-140

1.32E+043.96E+054.83E+045.49E+055.60E+04

1.92E+04 5.60E+037.03E+05 7.45E+041.35E+05 5.29E+046.12E+05 4.55E+045.60E+01 2.90E+03

3.56E+060.00E+000.00E+00O.OOE+000.00E+00


CE-141 2.77E+04 1.67E+04 1.99E+03CE-144 3.19E+06 1.21E+06 1.76E+05ND-147 7.94E+03 8.13E+03 5.00E+02

0.00E+000.00E+000.00E+00

5.25E+03 5.17E+05 2.16E+04 .

5.38E+05 9.84E+06 1.48E+053.15E+03 3.22E+05 3.12E+04


cApAs.Doc 08 58



TITLE:t OFF-SITE DOSE CALCULATIONS UNITS 1 AND 2

TABLE 10.6.2

GRW DOSE PARAMETERS FOR RADIOIODINES, RADIOACTIVEPARTICULATES, AND ANYRADIONUCLIDEOTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), CHILD AGE

GROUP, INHALATIONPATHWAYORGAN "0" (mrem/yr per pCi/m ) R»„„.,

NuclideH-3CR-51MN-54FE-59CO-58

Bone Liver T Body Thyroid0.00E+00 6.40E+02 6.40E+02 6.40E+020.00E+00 0.00E+00 1.54E+02 8.55E+010.00E+00 4.29E+04 9.51E+03 0.00E+002.07E+04 3.34E+04 1.67E+04 0.00E+000.00E+00 1.77E+03 3.16E+03 0.00E+00

Kidney Lung GI-LLI6.40E+02 6.40E+02 6.40E+022.43E+01 1.70E+04 1.08E+031.00E+04 1.58E+06 2.29E+040.00E+00 1.27E+06 7.07E+040.00E+00 1.11E+06 3.44E+04

CO-60ZN-65RB-86SR-89Y-90

0.00E+004.26E+040.00E+005.99E+054.11E+03

1.31E+04 2.26E+041.13E+05 7.03E+041.98E+05 1.14E+050.00E+00 1.72E+040.00E+00 1.11E+02

0.00E+000.00E+000.00E+000.00E+000.00E+00


SR-90ZR-95NB-95RU-103RU-106AG-110MSB-124SB-125TE-129MI-131

3.85E+07 0.00E+00 7.66E+051.90E+05 4.18E+04 3.70E+042.35E+04 9.18E+03 6.55E+032.79E+03 0.00E+00 1.07E+031.36E+05 0.00E+00 1.69E+041.69E+04 1.14E+04 9.14E+035.74E+04 7.40E+02 2.00E+049.84E+04 7.59E+02 2.07E+041.92E+04 6.85E+03 3.04E+034.81E+04 4.81E+04 2.73E+04

0.00E+000.00E+00O.OOE+00O.OOE+000.00E+000.00E+001.26E+029.10E+016.33E+031.62E+07


I-133CS-134CS-136CS-137BA-140

1.66E+046.51E+056.51E+049.07,E+057.40E+04

2.03E+041.01E+061.71E+058.25E+056.48E+01

7.70E+032.25E+051.16E+051.28E+054.33E+03

3.85E+060.00E+000.00E+000.00E+000.00E+00

3.38E+04 0.00E+003.30E+05 1.21E+059.55E+04 1.45E+042.82E+05 1.04E+052.11E+01 1.74E+06

5.48E+033.85E+034.18E+033.62E+031.02E+05

CE-141CE-144ND-147

3.92E+04 1.95E+04 2.90E+03 0.00E+006.77E+06 2.12E+06 3.61E+05 0.00E+001.08E+04 8.73E+03 6.81E+02 0.00E+00

8.55E+03 5.44E+05 5.66E+041.17E+06 1.20E+07 3.89E+054.81E+03 3.28E+05 8.21E+04


cApAs.Doc 08 59



TITLE: OFF-SITE DOSE CALCULATIONS UNITS 1 AND2

TABLE 10.6.3

GRW DOSE PARAMETERS FOR RADIOIODINES, RADIOACTIVEPARTICULATES, AND ANYRADIONUCLIDEOTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), TEEN AGE

GROUP, INHALATIONPATHWAYORGAN "0" (mrem/yr per pCi/m ) R, t p,

Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLIH-3CR-51MN-54FE-59CO-58

0.00E+00 7.25E+02 7.25E+02 7.25E+02 7.25E+02 7.25E+02 7.25E+020.00E+00 0.00E+00 1.35E+02 7.50E+01 3.07E+01 2.10E+04 3.00E+030.00E+00 5.11E+04 ~ 8.40E+03 0.00E+00 1.27E+04 1.98E+06 6.68E+041.59E+04 3.70E+04 1.43E+04 0.00E+00 0.00E+00 1.53E+06 '.78E+050.00E+00 2.07E+03 2.78E+03 0.00E+00 0.00E+00 1.34E+06 9.52E+04

CO-60ZN-65RB-86SR-89Y-90

0.00E+003.86E+040.00E+004.34E+052.98E+03

1.51E+04 1.98E+041.34E+05 6.24E+041.90E+05 8.40E+040.00E+00 1.25E+040.00E+00 8.00E+01

0.00E+000.00E+000.00E+000.00E+000.00E+00

0.00E+00 8.72E+068.64E+04 1.24E+060.00E+00 0.00E+000.00E+00 2.42E+060.00E+00 2.93E+05

2.59E+054.66E+041.77E+043.71E+055.59E+05

SR-90ZR-95NB-95RU-103RU-106AG-110MSB-124SB-125TE-129MI-131

3.31E+07 0.00E+001.46E+05,4.58E+041.86E+04 1.03E+042.10E+03 0.00E+009.84E+04 0.00E+001.38E+04 1.31E+044.30E+04 7.94E+027.38E+04 8.08E+021.39E+04 6.58E+033.54E+04 4.91E+04

6.66E+053.15E+045.66E+038.96E+021.24E+047.99E+031.68E+041.72E+042.25E+032.64E+04

O.OOE+000.00E+000.00E+000.00E+000.00E+000.00E+009.76E+017.04E+Ol4.58E+031.46E+07

0.00E+00 1.65E+076.74E+04 2.69E+061.00E+04 7.51E+057.43E+03 7.83E+051.90E+05 1.61E+072.50E+04 6.75E+060.00E+00 3.85E+060.00E+00 2.74E+065.19E+04 1.98E+068.40E+04 O.OOE+00

7.65E+051.49E+059.68E+041.09E+059.60E+052.73E+053.98E+059.92E+044.05E+056.49E+03

I-133CS-134CS-136CS-137,BA-140CE-141CE-144ND-147

1.22E+04 2.05E+04 6.22E+03 2.92E+065.02E+05 1.13E+06 5.49E+05 0.00E+005.15E+04 1.94E+05 1.37E+05 0.00E+006.70E+05 8.48E+05 3.11E+05 '.00E+005.47E+04 6.70E+01 3.52E+03 0.00E+002.84E+04 1.90E+04 2.17E+03 0.00E+004.89E+06 2.02E+06 2.62E+05 0.00E+007.86E+03 8.56E+03 5.13E+02 0.00E+00

3.59E+04 0.00E+003.75E+05 1.46E+051.10E+05 1.78E+043.04E+05 1.21E+052.28E+01 2.03E+068.88E+03 6.14E+051.21E+06 1.34E+075.02E+03 3.72E+05

1.03E+049.76E+031.09E+048.48E+032.29E+051.26E+058.64E+051.82E+05


CAPA8.DOC 08 60



NUMBER, CAP A-8REVISION 22PAGE 61 OF 64

UNITS 1 AND 2

Nuclide Bone Liver T Body GI-LLI

TABLE 10.6.4

GRW DOSE PARAMETERS FOR RADIOIODINES, RADIOACTIVEPARTICULATES, AND ANYRADIONUCLIDEOTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), ADULT

AGE GROUP, INHALATIONPATHWAYORGAN "0" (mrem/yr per pCi/m ) Ri jh It

Thyroid Kidney LunH-3 O.OOE+00 7.18E+02 7.18E+02CR-51 0.00E+00 0.00E+00 1.00E+02MN-54 0.00E+00 3.96E+04 6.30E+03FE-59 1.18E+04 2.78E+04 1.06E+04CO-58 0.00E+00 1.58E+03 2.07E+03

7.18E+025.95E+01O.OOE+00O.OOE+00O.OOE+00

7.18E+02 7.18E+02 7.18E+022.28E+01 1.44E+04 3.32E+039.84E+03 1.40E+06 7.74E+040.00E+00, 1.02E+06 1.88E+050.00E+00 9.28E+05 1.06E+05

CO-60ZN-65RB-86SR-89Y-90SR-90ZR-95NB-95RU-103RU-106

0.00E+00 1.15E+04 1.48E+043.24E+04 1.03E+05 "4.66E+04O.OOE+00 1.35E+05 5.90E+043.04E+05 0.00E+00 8.72E+032.09E+03 0.00E+00 5.61E+012.87E+07 O.OOE+00 5.77E+051.07E+05 3.44E+04 2.33E+041.41E+04 7.82E+03 4.21E+031.53E+03, 0.00E+00 6.58E+026.91E+04 0.00E+00 8.72E+03

O.OOE+000.00E+00O.OOE+000.00E+000.00E+000.00E+000.00E+00O.OOE+000.00E+000.00E+00

0.00E+00 5.97E+06 2.85E+056.90E+04 8.64E+05 5.34E+040.00E+00'.OOE+00 1.66E+04O.OOE+00 1.40E+06 3.50E+05O.OOE+00 1.70E+05 5.06E+050.00E+00 9.60E+06 7.22E+055.42E+04 1.77E+06 1.50E+057.74E+03 5.05E+05 1.04E+055.83E+03 5.05E+05 1.10E+051.34E+05 9.36E+06 9.12E+05

AG-110MSB-124SB-125TE-129MI-131

1.08E+043.12E+045.34E+049.76E+032.52E+04

1.00E+04 5.94E+035.89E+02 1.24E+045.95E+02 1.26E+044.67E+03 1.58E+033.58E+04 2.05E+04

O.OOE+007.55E+015.40E+013.44E+031.19E+07

1.97E+040.00E+000.00E+003.66E+046.13E+04

4.63E+06 3.02E+052.48E+06 4.06E+051.74E+06 1.01E+051.16E+06 3.83E+05O.OOE+00 6.28E+03

I-133CS-134CS-136CS-137BA-140

8.64E+03 1.48E+04 4.52E+033.73E+05 8.48E+05 7.28E+053.90E+04 1.46E+05 1.10E+054.78E+05 6.21E+05 4.28E+053.90E+04 4.90E+01 '.57E+03

2.15E+060.00E+000.00E+000.00E+000.00E+00


CE-141 1.99E+04 1.35E+04 1.53E+03CE-144 " 3.43E+06 1.43E+06 1.84E+05ND-147 5.27E+03 6.10E+03 3.65E+02

0.00E+000.00E+00O.OOE+00

6.26E+03 3.62E+05 1.20E+058.48E+05 7.78E+06 8.16E+053.56E+03 2.21E+05 1.73E+05


CAPA8.DOC 08 61




UNITS 1 AND 2

TABLE 10.6.5

GRW DOSE PARAMETERS'OR RADIOIODINES, RADIOACTIVEPARTICULATES, AND ANYRADIONUCLIDEOTHER THANNOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), CHILD AGE

GROUP, VEGETATIONPATHWAYORGAN "0" (mrem/yr per pCi/(sec m')) R, p's(

Nuclide Bone Liver T Body Thyroid Kidney Lun GI-LLI'-3

CR-51MN-54FE-59CO-58CO-60ZN-65RB-86SR-89Y-90

0.00E+00 2.29E+03 2.29E+03 2.29E+030.00E+00 0.00E+00 1.17E+05 6.49E+040.00E+00 6.65E+08 1.77E+08 0.00E+003.97E+08 6.42E+08 3.20E+08 0.00E+000.00E+00 6.45E+07 1.97E+08 0.00E+000.00E+00 3.78E+08 1.12E+09 0.00E+008.12E+08 2.16E+09 1.35E+09 0.00E+000.00E+00 4.54E+08 2.79E+08 0.00E+003.59E+10 0.00E+00 1.03E+09 0.00E+002.31E+04 0.00E+00 6.18E+02 0.00E+00


SR-90ZR-95NB-95RU-103RU-106

1.87E+123.86E+064.12E+051.53E+077.45E+08

0.00E+00 3.77E+ 10 0.00E+008.50E+05 7.56E+05 0.00E+001.61E+05 1.15E+05 0.00E+000.00E+00 5.89E+06 0.00E+000.00E+00 9.30E+07 0.00E+00


AG-110M 3.21E+07SB-124 3.52E+08SB-125, 4.99E+08TE-129M 8.40E+08I-131 1.43E+08

2.17E+074.57E+063.85E+062.35E+081.44E+08

1.74E+071.23E+081.05E+081.30E+088.17E+07

0.00E+007.78E+054.62E+052.71E+084.75E+10


I-133CS-134CS-136CS-137BA-140

3.52E+061.60E+ 108.18E+072.39E+102.77E+08

4.36E+062.63E+102.25E+082.29E+102.43E+05

1.65E+06 8.09E+085.55E+09 0.00E+001.46E+08 0.00E+003.38E+09 0.00E+001.62E+07 0.00E+00

7.26E+068.16E+091.20E+087.46E+097.90E+04

0.00E+00 1.76E+062.93E+09 1.42E+081.79E+07 7.90E+062.68E+09 1.43E+081.45E+05 1.40E+08

CE-141 6.55E+05 3.27E+05 4.85E+04CE-144 1.27E+08 3.98E+07 6.78E+06ND-147 . 7.27E+04 5.89E+04 4.56E+03

0.00E+000.00E+00O.OOE+00

1.43E+05 0.00E+00 4.08E+082.21E+07 0.00E+00 1.04E+103.23E+04 O.OOE+00 9.33E+07


For Tritium the units of the dose parameters are mrem/yr per pCi/m for all pathways, and they must'e

multiplied by X/Q.

cApAs.Doc 08 62




UNITS 1 AND 2

TABLE 10.6.6

GRW DOSE PARAMETERS FOR RADIOIODINES, RADIOACTIVEPARTICULATES, AND ANYRADIONUCLIDEOTHER THAN NOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), TEEN AGE

GROUP, VEGETATIONPATHWAYORGAN "0" (mrem/yr per pCi/(sec m')) R,<,,

Nuclide Bone Liver T Body Thyroid Kidney Lun GI-LLIH-3 0.00E+00 1.47E+03 1.47E+03 1.47E+03, 1.47E+03 1.47E+03 1.47E+03CR-51 0.00E+00 0.00E+00 6.16E+04 3.42E+04 1.35E+04 8.79E+04 1.03E+07MN-54 0.00E+00 4.54E+08 9.01E+07 0.00E+00 1.36E+08 0.00E+00 9.32E+08FE-59 1.79E+08 4.18E+08 1.61E+08 0.00E+00 0.00E+00 1.32E+08 9.89E+08CO-58 0.00E+00 4.37E+07 1.01E+08 0.00E+00 0.00E+00 0.00E+00 6.02E+08CO-60ZN-65RB-86SR-89Y-90

0.00E+004.24E+08O.OOE+001.51E+101.24E+04

2.49E+08 5.60E+081.47E+09 6.86E+082.75E+08 1.29E+080.00E+00 4.33E+080.00E+00 3.35E+02

0.00E+000.00E+000.00E+000.00E+000.00E+00


SR-90 9.22E+11 0.00E+00 1.84E+10 0.00E+00ZR-95 1.72E+06 5.44E+05, 3.74E+05 0.00E+00NB-95 1.93E+05 1.07E+05 5.90E+04 0.00E+00RU-103 6.82E+06 0.00E+00 2.91E+06 0.00E+00RU-106 3.09E+08 0.00E+00 3.90E+07 0.00E+00AG-110M 1.52E+07 1.44E+07 8.73E+06 0.00E+00SB-124 1.55E+08 2.85E+06 6.03E+07 3.51E+05SB-125 2.14E+08 2.34E+06 5.01E+07 2.05E+05TE-129M 3.61E+08 1.34E+08 5.72E+07 1.17E+08I-131 7.68E+07 1.08E+08 5.78E+07 3.14E+10

0.00E+00 0.00E+00 2.11E+107.99E+05 0.00E+00 .1.26E+091.04E+05 O.OOE+00 4.58E+082.40E+07 0.00E+00 5.69E+085.97E+08 0.00E+00 1.48E+102.74E+07 0.00E+00 4.03E+090.00E+00 1.35E+08 3.11E+090.00E+00 1.88E+08 1.67E+091.51E+09 0.00E+00 1.36E+091.85E+08 0.00E+00 2.13E+07

I-133CS-134CS-136CS-137BA-140

1.93E+067.10E+094.35E+071.01E+ 101.38E+08

3.28E+061.67E+101.71E+081.35E+101.69E+05

1.00E+067.75E+091.15E+084.69E+098.90E+06

4.58E+080.00E+000.00E+000.00E+000.00E+00

5.75E+06 O.OOE+005.31E+09 2.03E+099.31E+07 1.47E+074.59E+09 '.78E+095.74E+04 1.14E+05

2.48E+062.08E+081.38E+071.92E+082.13E+08

CE-141 2.83E+05 1.89E+05 2.17E+04 0.00E+00CE-144 5.27E+07 2.18E+07 2.83E+06 0.00E+00ND-147 3.67E+04 .4.00E+04 2.39E+03 0.00E+00

8.89E+04 0.00E+00 5.40E+081.30E+07 0.00E+00 1.33E+102.35E+04 0.00E+00 1.44E+08

Dose factors are based upon NUREG 0133 methodology.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.

CAPA8.DOC 08 63




UNITS j. AND 2

TABLE 10.6.7

GRW DOSE PARAMETERS FOR RADIOIODINES, RADIOACTIVEPARTICULATES, AND ANYRADIONUCLIDEOTHER THANNOBLE GAS (IPT), GASEOUS EFFLUENTS (GRW), ADULT

AGE GROUP VEGETATIONPATHWAYORGAN 0 (mrem/yr pcI'Ci/(sec m )) Rl yy

Nuclide Bone Liver T Body Thyroid Kidney Lun GI-LLIH-3CR-51MN-54FE-59CO-58

0.00E+000.00E+000.00E+001.26E+080.00E+00

1.29E+03 1.29E+03 1.29E+03 1.29E+03 1.29E+03 1.29E+030.00E+00 4.64E+04 2.77E+04 1.02E+04 6.15E+04 1.17E+073.13E+08 5.97E+07 O.OOE+00 9.31E+07 0.00E+00 9.5SE+082.96E+08 1.13E+08 0.00E+00 0.00E+00 8.27E+07 9.87E+083.08E+07 6.90E+07 0.00E+00 0.00E+00 0.00E+00 6.24E+08

CO-60ZN-65 *

RB-86SR-89Y-90

0.00E+00 1.67E+08 3.69E+083.17E+08 1.01E+09 4.56E+080.00E+00 2.20E+08 1.03E+OS9.95E+09 0.00E+00 2.86E+081.33E+04 0.00E+00 3.57E+02

0.00E+00 O.OOE+00 0.00E+00 3. 14E+090.00E+00 6.75E+OS O.OOE+00 6.36E+080.00E+00 O.OOE+00 0.00E+00 4.34E+070.00E+00 0.00E+00 O.OOE+00 1.60E+090.00E+00 O.OOE+00 0.00E+00 1.41E+08

SR-90 6.95E+11 0.00E+00 1.40E+10 0.00E+00 O.OOE+00ZR-95 1.18E+06 3.77E+05, 2.55E+05 O.OOE+00 5.92E+05NB-95 1.43E+05 7.95E+04 4.27E+04 O.OOE+00 7.86E+04RU-103 '.77E+06 0.00E+00 2.05E+06 0.00E+00 1.82E+07RU-106 1.93E+08 0.00E+00 2.44E+07 0.00E+00 3.72E+08

0.00E+00 1.75E+100.00E+00 1.20E+090.00E+00 4.83E+080.00E+00 5.57E+08O.OOE+00 1.25E+10

AG-110M 1.05E+07SB-124 1.04E+08SB-125 1.37E+08TE-129M 2:51E+08I-131 8.07E+07


1.92E+07, 0.00E+00 3.98E+090.00E+00 8.08E+07 2.95E+090.00E+00 1.05E+08 1.50E+091.05E+09 0.00E+00 1.26E+091.98E+08 O.OOE+00 3.05E+07

I-133CS-134CS-136CS-137BA-140

2.08E+064.67E+094.25E+076.36E+091.29E+08

3.62E+06 1.10E+06 5.32E+OS 6.31E+06 0.00E+00 3.25E+061.11E+10 9.08E+09 0.00E+00 3.59E+09 1.19E+09 1.94E+081.68E+08 - 1.21E+08 0.00E+00 9.33E+07 1.28E+07 1.90E+078.70E+09 5.70E+09 O.OOE+00 2.95E+09 9.81E+08 1.68E'+081.61E+05 8.42E+06 0.00E+00 5.49E+04 9.24E+04 2.65E+08

CE-141CE-144ND-147

1.97E+05 1.33E+05 ~ 1.51E+04 0.00E+00 6.19E+04 0.00E+00 5.09E+083.29E+07 1.38E+07 1.77E+06 0.00E+00 8.16E+06 0.00E+00 1.11E+103.37E+04 3.90E+04 2.33E+03 O.OOE+00 2.28E+04 0.00E+00 1.87E+08


For Tritium the units of the dose parameters are mrem/yr per pCi/m for all pathways, and they mustbe multiplied by X/Q.

cApAs.Doc 08 64

Attachment 5PG&E Letter DCL 99-055

Process Control Program

(RP2. DC2 Revision 4)

*~*ISSUED FOR USE BY:PACIFIC GAS ANDELECTRIC COMPANYNUCLEAR POWER GENERATIONDIABLOCANYONPOWER PLANTADMINISTRATIVEPROCEDURE

DATE: EXPIRES: 4>k8

NUMBER RP2.DC2

REVISION 4

PAGE 1 OF 6

'ITLE: PROCESS CONTROL PROGRAM

EFFECTIVE DATESPONSORING ORGANIZATION:RADIATIONPROTECTION

PROCEDURE CLASSIFICATION: QUALITYRELATEDREVIEW LEVEL: "A"

SCOPE

DISCUSSION

2.1 Solidification is the conversion of wet radioactive wastes into a form that meets

shipping and burial ground requirements.

This procedure implements the requirements of 10 CFR 50.36a and General DesignCriterion 60 of Appendix A to 10 CFR Part 50. The process parameters included inestablishing the PROCESS CONTROL PROGRAM may include, but are not limited to,waste type, waste pH, waste/liquid/SOLIDIFICATION agent/catalyst ratios, waste oilcontent, waste principal chemical constituents, and mixing and curing times.

This procedure contains the individual procedures necessary to perform PCP samplesolidifications.

2.2

2.3

Cement Solidification willnot be utilized to stabilize resin or floor drain sludges. OnlyNRC approved binders (e.g., bitumen, VinylEster Styrene, Aztec) may be used tosolidify resin or floor drain sludges to meet waste form stability.

RESPONSIBILITIES

2.4

1.1 ~ The purpose of the Process Control Program (PCP) is to define the necessary programguidance used at DCPP to ensure that SOLID RADIOACTIVEWASTEMANAGEMENTactivities, in solidifying wet radioactive waste for disposal, conformto the Code of Federal and State Regulations and the Waste Burial Site License Criteria.

3.1

3.2

3.3

3.4

3.5

Plant Manager has the overall responsibility for the Solid Radioactive Waste activities atDCPP.

Director, Radiation Protection is responsible for the implementation of the requirementsof this procedure.

Radwaste Engineer is responsible for the development and review of proceduresrelating to the requirements of this procedure.

.The Radwaste Foreman is responsible for the implementation of procedures relating tothe requirements of this procedure.

The Quality Control Section is responsible for verification of compliance with the

Quality requirements.

RP2DC2.DOC


TITLE: PROCESS CONTROL PROGRAM

NUMBER RP2.DC2REVISION 4PAGE- 2OF6

4.2

5.1 GENERAL

It is the policy of Pacific Gas and Electric Company to conscientiously apply emphasisand attention to those activities associated with generation, processing, packaging,storage and disposal of radioactive waste generated at the Diablo Canyon Power Plantand to maintain a high level of assurance that radioactive waste products meet or exceed

the applicable Federal and State regulations and the Radioactive Waste Burial SiteLicense Criteria.

4.1 This procedure with the attachments and any changes thereto requires review by thePlant Staff Review Committee and submission to the US NRC in the Annual EffluentReport for the period in which the changes were made.

Any major change to the Solid Radwaste Treatment System shall be'reported to the USNRC in the Annual Effluent Report for the period in which the evaluation was reviewed

by the PSRC. The discussion of each change shall contain the items listed inAttachment 7.2. This information may be submitted as part of the annual FSAR updatein lieu of the Annual Effluent Report.

INSTRUCTIONS

5.2 WET WASTE

5.2.1 LIQUID/WETWASTE

5.2.2

Liquid/Wet wastes at DCPP are processed to a condition meeting shipping-and disposal criteria. These criteria include requirements for immobilization,stability and limits on Free Standing Water (FSW). Specific instructions onprocessing and required FSW limits are contained in plant procedures and/orqualified vendor procedures.

CONTAINERS, SHIPPING CASKS AND PACKAGING

Solid radioactive waste is processed, packaged and shipped in accordancewith DCPP procedures and/or qualified vendor procedures. Theseprocedures provide specific instructions which ensure the container, shippingcasks, and packaging methods comply with the applicable Code of FederalRegulations, State Regulations and the Radioactive Waste Burial Site LicenseCriteria.

5.2.3 SHIPPING AND DISPOSAL

Solid radioactive waste is prepared, loaded and shipped to a Federal and/orState Licensed Radioactive Waste Disposal Facility (Burial Ground) inaccordance with DCPP procedures and/or qualified vendor procedures.These procedures provide specific instructions which ensure the shipmentsmeet the intended Burial Site License Requirements as well as'pplicableFederal and State Regulations.

RP2DC2.DOC 1B



NUMBER RP2.DC2REVISION 4PAGE 3 OF 6

5.2.4

5.2.5

5.2.6

LABORATORYMIXINGOF SAMPLES

Qualified vendor procedures reviewed by the PSRC and approved by thePlant Manager provide written instructions on sampling, processing and

handling waste for the determination of process parameters prior to the actualfull scale solidification. These procedures contain the description of thelaboratory mixing methods used for these samples.

SOLIDIFICATIONPROCESS

Qualified vendors used by DCPP for radioactive waste solidification are

required to provide the Process Control Program and written procedures.These procedures and changes thereto must be reviewed by the PSRC and

approved by the Plant Manager prior to use. Further, the vendors arerequired to have a topical report, as referenced, on the waste forms whichwillbe solidified at DCPP. This topical report should demonstratecompliance with the NRC requirements for waste form. These documentsshould include:

a. Description of the solidification process.

b. Type of solidification used.

c. Process control parameters.

d. Parameter boundary conditions.

e. Proper waste form properties.

f. Specific instructions to ensure the systems are operated withinestablished process parameters.

SAMPLING PROGRAM FOR SOLIDIFICATION

Vendors, utilized by DCPP for radioactive waste solidification, are requiredto include in their approved procedures, requirements to sample at least everytenth batch of the same waste type to ensure solidification and to provideactions to be taken ifa sample fails to verify solidification. After a test

specimen failure, initial test specimens from three consecutive batches of thatwaste type must demonstrate solidification before testing requirements ofevery tenth batch can be resumed. Verification of such sampling is to beaccomplished by completing Form 69-10350, "Processing Control Program(PCP) Verification." (See Attachment 7.1.) These forms willbe maintainedin the Radiation Protection Section and the Records Management System(RMS). These procedures and changes thereto must be reviewed by thePSRC and approved by the Plant Manager prior to use.

RP2DC2.DOC




5.3

5.4

5.2.7 WASTE FORM VERIFICATION

Vendors utilized by DCPP to process wet wastes are required to include intheir procedures provisions to verify that the solidification and/or FSWCriteria in the Federal and State regulations and the Burial Site License aremet for the specific type of waste being processed.

5.2.8 CORRECTIVE ACTIONS FOR FREE STANDINGWATER

Vendors utilized by DCPP to process wet wastes are required to include intheir approved procedures provisions for correcting processed waste in whichfree standing water in excess of the FSW Criteria is detected.

5.2.9 EXOTHERMIC PROCESSES

Vendors utilized by DCPP for radioactive waste solidification that utilize an

exothermic solidification method are required to include in their approvedprocedures:

a. Waste/binder temperature monitoring to mitigate the consequence ofadverse exothermic reactions which may occur in the full scale

solidification but might not be noticeable in the specimen tests.

b. Specific process control parameters that shall be met before capping thecontainer.

OILYWASTE

Oily wastes at DCPP are processed in accordance with approved vendor procedures.These procedures specify the proper methods to treat oily wastes to comply withFederal and State regulations and applicable Burial Site License Criteria.

SPECIAL CASES

Based upon previous industry experience, DCPP foresees the potential for situationsarising that may be beyond existing plant capabilities. Anticipating this possibility,provisions are made herein to accommodate such situations in a timely manner by usingspecial techniques or processes. These special cases would be controlled as follows:

5.4.1 Implementing procedures would be developed comparable to those used fornormal plant solid waste activities based on the guidance of this PCP andincorporating the applicable provisions for process control and testing.

5.4.2 The implementing procedure would receive PSRC review and Plant Managerapproval prior to use.

5.4.3 Use of this provision and supporting information would be included in thenext Annual Effluent Report to the NRC.

RP2DC2. DOC 1B


TITLE: PROCESS CONTROL PROGRAMti

A


5.5 REMEDIALACTIONS

5.5.1

5.5.2

For waste forms which do not meet Federal, State and burial site regulationsand requirements, suspension of shipment of the inadequately processed

waste and correction of the PCP, procedures or processing equipment shallbe performed as necessary to prevent recurrence.

For waste forms not prepared in accordance with the PCP, testing of thewaste to verify shipping and burial site requirements shall be performed and

appropriate administrative action taken to prevent recurrence.

5.6 VENDOR PROCEDURES AND REPORTS

5.6.1 The following are maintained in Document Control Master File, Catalog No.TK 9400/SEG-1.

a. Scientific Ecology Group, Inc., Procedure SS-042, Process ControlProgram for Radwaste Solidification Service at Diablo Canyon PowerPlant, Rev. F

b. Topical Report TR-002, 10 CFR 61 Qualified Radioactive WasteForms, Rev. 1

Scientific Ecology Group, Inc., Procedure GS-007, "TemperatureIndicating Device Comparison Test," Rev. E

Scientific Ecology Group, Inc., Procedure GS-006, "Calibration for a

Triple Beam Balance," Rev. D

e. Scientific Ecology Group, Inc., Procedure SS-049, "PenetrometerCheck Procedure," Rev. D

5.6.2 The following are maintained in Document Control Master File, Catalog No.TK 94001 DTI-1.

Diversified Technologies, Procedure DT-DCPP-10, PCP for VERISolidification, Rev 14

Topical Report DTI-VERI-100-NP-A. VERI (VinylEster Resin InSitu) Solidification Process for Low-Level Radioactive Waste, Rev 1

c. Diversified Technologies, Procedure QIP 20-17, pH Meter Calibration,Rev. 3

d. Diversified Technologies, Procedure QIP 20-18,,Conductivity MeterCalibration, Rev. 2

e. Diversified Technologies, Procedure QIP-20-16, ThermocoupleCalibration, Rev 1

f. Diversified Technologies, Procedure DT-DCPP-VES-14, PCP for VESSolidification, Rev B.

RP2DC2.DOC




6. RECORDS

g. Topical Report DNS-RSS-200-NP, The Dow Waste SolidificationProcess for Low-Level Radioactive Waste (Docket Number WM-82).

6.1 Records of PCP specimen results and Form 69-10350 shall be submitted to the RecordsManagement System on a shipment basis by container per RCP RW-4.

7. ATTACHMENTS

7.1 Form 69-10350, "Process Control Program (PCP) Verification," 06/03/93

7.2 "Major Change to the Solid Radwaste Treatment System Evaluation," 06/03/93

REFERENCES

8.1 Title 10 Code of Federal Regulations.

8.2 NUREG 0472 and 0473.

8.3 NUREG-0800, 11.4 US NRC Standard Review Plan Solid Waste ManagementSystems.

8.4

8.5

8.6

8.7

8.8

RCP.DC3, "Dewatering Control Program."

RP2.DC4, "Mobile Service Operating Procedure for Low-Level Radioactive WasteProcessing."

NRC Information Notice 88-08, Chemical Reactions with Radioactive WasteSolidification Agents.

Scientific Ecology Group, Inc., Quality Management System Manual, Rev. 5,Document Control Master File, Catalog No. TK-9400/SEG-3.

Technical Position on Waste Form, Revision 1, US NRC, January 1991.

SPONSOR

C. Clint Miller

RP2DC2.DOC 1B

69-10350 06/03/93DIABLOCANYONPOWER PLANT

RP2.DC2

Page 1 of 1

TITLE: PROCESS CONTROL PROGRAM (PCP) VERIFICATION

WASTE BATCH PCPDATE

Number Type Passed FailedNOTES OPERATOR VERIFIER

RP2DC2. DOC

06/03/93DIABLOCANYONPOWER PLANT

RP2.DC2ATTACHMENT7.2

Page 1 of 1

TITLE: MAJOR CHANGE TO THE SOLID RADWASTE TREATMENTSYSTEMEVALUATION

1. A summary of the evaluation that led to the determination that the change could be made inaccordance with 10 CFR 50.59;

2. Sufficient detailed information to totally support the reason for the change without benefit ofadditional or supplemental information;

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

4. An evaluation of the change which shows the predicted releases of radioactive materials in liquid andgaseous effluents and/or quantity of solid waste that differ from those previously predicted in theLicense application and amendments thereto;

5. An evaluation of the change which shows the expected maximum exposures to a MEMBER OF THEPUBLIC 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 effluent and insolid waste, to the actual releases for the period prior to when the changes are to be made;

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

8. Documentation of the fact that the change was reviewed and found acceptable by the PSRC.

RP2DC2.DOC

Attachment 6PG8E Letter DCL 99-055

Process Control Program


~~~ ISSUED FOR USE BY:PACIFIC GAS ANDELECTRIC COMPANYNUCLEARPOWER GENERATIONDIABLOCANYONPOWER PLANTADMINISTRATIVEPROCEDURE

DATE: EXPIRES:NUMBER RP2.DC2

REVISION 5

PAGE 1 OF 6


(o-EFFECTIVE DATE

SPONSORING ORGANIZATION:RADIATIONPROTECTIONPROCEDURE CLASSIFICATION:QUALITYRELATED

REVIEW LEVEL: "A"

l. SCOPE

The purpose of the Process Control Program (PCP) is to define the necessary programguidance used at DCPP to ensure that SOLID RADIOACTIVEWASTEMANAGEMENTactivities, in solidifying wet radioactive waste for disposal, conformto the Code of Federal and State Regulations and the Waste Burial Site License Criteria.

DISCUSSION

2.1

2.2

2.3

SolidiTication is the conversion of wet radioactive wastes into a form that meetsshipping and burial ground requirements.

This procedure implements the requirements of 10 CFR 50.36a and General Design, Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in

establishing the PROCESS CONTROL PROGRAM may include, but are not limited to,waste type, waste pH, waste/liquid/SOLIDIFICATION agent/catalyst ratios, waste oilcontent, waste principal chemical constituents, and mixing and curing times.

This procedure contains the individual procedures necessary to perform PCP samplesolidifications.

2.4 Cement Solidification willnot be utilized to stabilize resin or floor drain sludges. OnlyNRC approved binders (e.g., bitumen, Vinyl Ester Styrene, Aztec) may be used tosolidify resin or floor drain sludges to meet waste form stability.

RESPONSIBILITIES

3.1

3.2

3.3

3.4

3.5

Plant Manager has the overall responsibility for the Solid Radioactive Waste activities atDCPP.

Director, Radiation Protection is responsible for the implementation of the requirementsof this procedure.

Radwaste Engineer is responsible for the development and review of proceduresrelating to the requirements of this procedure.

The Radwaste Foreman is responsible for the, implementation of procedures relating tothe requirements of this procedure.

NQS is responsible for verification of compliance with the Quality requirements.

01180405.DOA 1B I I 0806.08l3




4,

5.

PRERE UISITES

4.2

4.1 This procedure with'the attachments and any changes thereto requires review by thePlant Staff Review Committee and submission to the US NRC in the Annual EffluentReport for the period in which the changes were made.

Any major change to the Solid Radwaste Treatment System shall be reported to the USNRC in the Annual Effluent Report for the period in which the evaluation was reviewedby the PSRC. The discussion of each change shall contain the items listed inAttachment 7.2. This information may be submitted as part of the annual FSAR updatein lieu of the Annual Effluent Report.

INSTRUCTIONS

5.1

5.2

GENERAL

It is the policy of Pacific Gas and Electric Company to conscientiously apply emphasis ~

and attention to those activities associated with generation, processing, packaging,storage and disposal of radioactive waste generated at the Diablo Canyon Power Plantand to maintain a high level of assurance that radioactive waste products meet or exceedthe applicable Federal and State regulations and the Radioactive Waste Burial SiteLicense Criteria.

WET WASTE

5.2.1 LIQUID/WETWASTE

5.2.2

5.2.3

Liquid/Wet wastes at DCPP are processed to a condition meeting shippingand disposal criteria. These criteria include requirements for immobilization,stability and limits on Free Standing Water (FSW). Specific instructions onprocessing and required FSW limits are contained in plant procedures and/orqualified vendor procedures.

CONTAINERS, SHIPPING CASKS AND PACKAGING

Solid radioactive waste is processed, packaged and shipped in accordancewith DCPP procedures and/or qualified vendor procedures. Theseprocedures provide specific instructions which ensure the container, shippingcasks, and packaging methods comply with the applicable Code of FederalRegulations, State Regulations and the Radioactive Waste Burial Site LicenseCriteria.

SHIPPING AND DISPOSAL

Solid radioactive waste is prepared, loaded and shipped to a Federal and/orState Licensed Radioactive Waste Disposal Facility (Burial Ground) inaccordance with DCPP procedures and/or qualified vendor procedures.These procedures provide specific instructions which ensure the shipmentsmeet the intended Burial Site License Requirements as well as applicableFederal and State Regulations.

01180405.DOA IB 2 I 0806.08i3




5.2.4

5.2.5

5.2.6

LABORATORYMIXINGOF SAMPLES

Qualified vendor procedures reviewed by the PSRC and approved by thePlant Manager provide written instructions on sampling, processing and

handling waste for the determination of process parameters prior to the actualfull scale solidification. These procedures contain the description of thelaboratory mixing methods used for these samples.

SOLIDIFICATIONPROCESS

Qualified vendors used by DCPP for radioactive waste solidification are

required to provide the Process Control Program and written procedures.These procedures and changes thereto must be reviewed by the PSRC and

approved by the Plant Manager prior to use. Further, the vendors arerequired to have a topical report, as referenced, on the waste forms whichwillbe solidified at DCPP. This topical report should demonstratecompliance with the NRC requirements for waste form. These documentsshould include:

a. Description of the solidification process.

b. Type of solidification used.

c. Process control parameters.

d. Parameter boundary conditions.

e. Proper waste form properties.

Specific instructions to ensure the systems are operated withinestablished process parameters.

SAMPLING PROGRAM FOR SOLIDIFICATION

Vendors, utilized by DCPP for radioactive waste solidification, are requiredto include in their approved procedures, requirements to sample at least everytenth batch of the same waste type to ensure solidification and to provideactions to be taken ifa sample fails to verify solidification. After a testspecimen failure, initial test specimens from three consecutive batches of thatwaste type must demonstrate solidification before testing requirements ofevery tenth batch can be resumed. Verification of such sampling is to beaccomplished by completing Form 69-10350, "Processing Control Program(PCP) Verification." (See Attachment 7.1.) These forms willbe maintainedin the Radiation Protection Section and the Records Management System(RMS). These procedures and changes thereto must be reviewed by thePSRC and approved by the Plant Manager prior to use.

01180405.DOA tB 3 I 0806.0813


TITLEi PROCESS CONTROL PROGRAM


5.2.7

5.2.8

5.2.9

WASTE FORM VERIFICATION

Vendors utilized by DCPP to process wet wastes are required to include intheir procedures provisions to verify that the solidification and/or FSWCriteria in the Federal and State regulations and the Burial Site License aremet for the specific type of waste being processed.

CORRECTIVE ACTIONS FOR FREE STANDING WATER

Vendors utilized by DCPP to process wet wastes are required to include intheir approved procedures provisions for correcting processed waste in whichfree standing water in excess of the FSW Criteria is detected.

EXOTHERMIC PROCESSES

Vendors utilized by DCPP for radioactive waste solidification that utilize anexothermic solidification method are required to include in their approvedprocedures:

a. Waste/binder temperature monitoring to mitigate the consequence ofadverse exothermic reactions which may occur in the full scalesolidification but might not be noticeable in the specimen tests.

b. Specific process control parameters that shall be met before capping thecontainer.

5.3 OILYWASTE

Oily wastes at DCPP are processed in accordance with approved vendor procedures.These procedures specify the proper methods to treat oily wastes to comply withFederal and State regulations and applicable Burial Site License Criteria.

5.4 SPECIAL CASES

Based upon previous industry experience, DCPP foresees the potential for situationsarising that may be beyond existing plant capabilities. Anticipating this possibility,provisions are made herein to accommodate such situations in a timely manner by usingspecial techniques or processes. These special cases would be controlled as follows:

5.4.1 Implementing procedures would be developed comparable to those used fornormal plant solid waste activities based on the guidance of this PCP andincorporating the applicable provisions for process control and testing.

5.4.2 The implementing procedure would receive PSRC review and Plant Managerapproval prior to use.

5.4.3 Use of this provision and supporting infor'mation would be included in thenext Annual Effluent Report to the NRC.

01 i80405.DOA 1B 4 I 0806.08l3



TH'LE: PROCESS CONTROL PROGRAM

5.5

5.6

REMEDIALACTIONS

5.5.1 For waste forms which do not meet Federal, State and burial site regulations

and requirements, suspension of shipment of the inadequately processed

waste and correction of the PCP, procedures or processing equipment shall

be performed as necessary to prevent recurrence.

5.5.2 For waste forms not prepared in accordance with the PCP, testing of the

waste to verify shipping and burial site requirements shall be performed and

appropriate administrative action taken to prevent recurrence.

VENDOR PROCEDURES AND REPORTS

5.6.1 The following are maintained in Document Control Master File, Catalog No.

TK 9400/MMT-1.

ao

b.

Molten Metal Technology of Tennessee, Inc., ProcedurePG01-PCP-05-001, Process Control Program for MMT-TNRadwaste

Solidification at Diablo Canyon Power Plant, Rev. 0

Topical Report TR402, 10 CFR 61 Qualified Radioactive Waste

Forms, Rev. 1

c. Molten Metal Technology of Tennessee, Inc., ProcedureMMT-P-01-035, "Temperature Indicating Device Comparison Test,"Rev. 0

d. Molten Metal Technology of Tennessee, Inc., ProcedureMMT-P-01-020, "Calibration for a Triple Beam Balance," Rev. 0

e. Molten Metal Technology of Tennessee, Inc., ProcedureMMT-P-01-024, "Penetrometer Check Procedure," Rev. 0

5.6.2 The following are maintained in Document Control Master File, Catalog No.TK 94001 DTI-1.

a.

b.

Diversified Technologies, Procedure DT-DCPP-10, PCP for VERISolidification, Rev 14

Topical Report DTI-VERI-100-NP-A. VERI (VinylEster Resin InSitu) Solidification Process for Low-Level Radioactive Waste, Rev 1

c. Diversified Technologies, Procedure QIP 20-17, pH Meter Calibration,Rev. 3

d. Diversified Technologies, Procedure QIP 20-18, Conductivity MeterCalibration, Rev. 2

01180405.DOA IB 5 I 0806.0813




6. RECORDS

e. Diversified Technologies, Procedure QIP-20-16, ThermocoupleCalibration, Rev 1

f. Diversified Technologies, Procedure DT-DCPP-VES-14, PCP for VESSolidification, Rev B.

g. Topical Report DNS-RSS-200-NP, The Dow Waste SolidificationProcess for Low-Level Radioactive Waste (Docket Number WM-82).

7.

8.

6.1 Records of PCP specimen results and Form 69-10350 shall be submitted to the Records

Management System on a shipment basis by container per RCP RW4.

ATTACHMENTS

7.1 Form 69-10350, "Process Control Program (PCP) Verification," 06/03/93

7.2 "Major Change to the Solid Radwaste Treatment System Evaluation," 06/03/93

REFERENCES

8.1 Title 10 Code of Federal Regulations.

8.2 NUREG 0472 and 0473.

8.3 NUREG-0800, 11.4 US NRC Standard Review Plan Solid Waste ManagementSystems.

8.4

8.5

8.6

8.7

8.8

RCP.DC3, "Dewatering Control Program."

RP2.DC4, "Mobile Service Operating Procedure for Low-Level Radioactive WasteProcessing."

NRC Information Notice 88-08, Chemical Reactions with Radioactive WasteSolidification Agents.

Molten Metal Technology of Tennessee, Inc., Quality Assurance Program DescriptionNo. 71-0870, Rev. 0.

Document Control Master File, Catalog No. TK-9400/MMT-3.

Technical Position on Waste Form, Revision 1, US NRC, January 1991.

01180405.DOA 1B' I 0806.0813

69-10350 06/03/93DIABLOCANYONPOWER PLANT.

RP2.DC2

Page 1 of 1

TITLE: PROCESS CONTROL PROGRAM (PCP) VERIFICATION~1

WASTE BATCH PCPDATE

Number T)g)e Passed F I dNOTES OPERATOR

01180405.DOA 1B 7 I 0806.0813

06/03/93DIABLOCANYONPOWER PLANT

RP2.DC2ATTACHMENT7.2

Page 1 of 1

TITLE: MAJOR CHANGE TO THE SOLID RADWASTE TREATMENTSYSTEMEVALUATION

1. A summary of the evaluation that led to the determination that the change could be made inaccordance with 10 CFR 50.59;

2. Sufficient detailed information to totally support the reason for the change without benefit ofadditional or supplemental information;

3. A detailed description of the equipment, components and processes involved and the interfaces withother 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 THEPUBLIC 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 effluent and insolid waste, to the actual releases for the period prior to when the changes are to be made;

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

8. Documentation of the fact that the change was reviewed and found acceptable by the PSRC.

0118040S.DOA 1B 8 I 0806.0813

Attachment 7PG8E Letter DCL 99-055

Mobile Service Operating Proceduresfor Low-Level Radioactive Waste Processing


***ISSUED FOR USE BY:PACIFIC GAS AND ELECTRIC COMPANYNUCLEAR POWER GENERATIONDIABLO.CANYONPOWER PLANTADMINISTRATIVEPROCEDURE

DATE: EXPIRES:NUMBER RP2.DC4

REVISION 9

PAGE 1 OF 4

TITLE: MOBILESERVICE OPERATING PROCEDURES FORLOW-LEVELRADIOACTIVEWASTE PROCESSING

EFFECTIVE DATESPONSORING ORGANIZATION: RADIATIONPROTECTION

PROCEDURE CLASSIFICATION: QUALITYRELATEDREVIEW LEVEL: "A"

2.

SCOPE

1.1 This procedure discusses the use of Mobile Service equipment for the proper, safe andeffective processing and/or preparation of radioactive waste for disposal.

DISCUSSION

2.2

RESPONSIBILITIES

2.1 This procedure contains the individual Operating Procedures necessary to operate themobile processing and solidification equipment.

The processing conditions necessary to assure that packaged radioactive wastes meet thevarious governmental and Burial Site Facility Waste Form requirements are addressedin RP2.DC2, "Process Control Program."

3.1

3.2

The Mobile Service Contractor is responsible for:

3.1.1 Setup of the processing equipment,

3.1.2 Operation of the processing equipment,

3.1.3 Dismantling of the processing equipment.

Radiation Protection is responsible for:

3.2.1

3.2.2

Performing surveys on the equipment prior to receipt,

Assuring that the DOT requirements have been met prior to the release of theequipment.

3.3 The Radwaste Engineer is responsible for:

3.3.1 Assuring the most current revisions of the Operating Procedures are on file,

3.3.2 Assuring that the requirements (ifany) for effluent monitoring have beendiscussed with the Chemistry Engineer for radioactive effluents and theimplementation of such requirements is conducted.

3.4

3.5

The Chemistry Engineer for radioactive effluents is responsible for evaluating andreporting effluents from mobile low-level radwaste processing equipment in the AnnualEffluent Report as necessary.

NQS is responsible for the verification that operations have been conducted inaccordance with the appropriate Operating Procedures.

RP2DC4.DOC 18 1



NUMBER RP2.DC4REVISION 9PAGE — 2 OF 4

Changes to the dewatering and solidification procedures contained herein due to changes inequipment which would affect any waste form parameters require submission to the US NRC inthe Annual Effluent Report for the period in which the changes were made.

INSTRUCTIONS

Follow the appropriate steps in the individual Operating Procedures as listed below:

5.1 Liquid Waste Processing

NOTE: Copies of the following procedures are maintained in Document ControlMaster File, Catalog No. TK-9400/DTI-3.

5.1.1 Diversified Technologies, Procedure DT-DCPP-01, WPS WasteProcessing, Rev. 0.

5.1.2 Diversified Technologies, Procedure DT-DCPP-02, WPS Adding Media,Rev. 0.

5.1.3 Diversified Technologies, Procedure DT-DCPP-03, WPS Sluicing Media,Rev. 0.

5.1.4 Diversified Technologies, Procedure DT-DCPP-04, WPS Leak Check,Rev. 0.

5.1.5 Diversified Technologies, Procedure DT-DCPP-08, RO-NF Operations,Rev. 0.

NOTE: A copy of the following procedure is maintained in Document Control MasterFile, Catalog No. TK-9400/SEG-2.

5.1.6 Molten Metal Technology of Tennessee, Inc., Procedure PG01-P-10-001,Procedure For Processing Sluice Water In A NUHIC-80B or Resin ExpressContainer at Diablo Canyon, Rev. 0.

NOTE: A copy of the following procedure is maintained in Document Control MasterFile, Catalog No. TK-9400/CNS-3.

5.1.7 Chem Nuclear Systems, Inc., WI-CNSI-96-020, Work Instruction forLoading a CNS-21-300 Thermex Liner at PG&E Diablo Canyon PowerPlant, Rev. 0.

5.2 Dewatering

NOTE: Copies of the following procedures are maintained in Document ControlMaster File, Catalog No. TK-9400/SEG-2.

5.2.1

5.2.2

Scientific Ecology Group, Inc., Procedure DW-008, Dewatering Procedurefor 182 Ft3 Liners at Diablo Canyon Power Plant, Rev. D.

Molten Metal Technology of Tennessee, Inc., Procedure PG01-P-06-001,"Transfer and Dewatering Activated Carbon and/or Bead Resin at DiabloCanyon for Offsite Processing," Rev 0.

RP2DC4.DOC 18 2




5.2.3

5.2.4

5.2.5

5.2.6

5.2.7

5.2.8

Molten Metal Technology of Tennessee, Inc., Procedure PG01-P-03-001,Procedure for Dewatering Bead Resin, Filter Sludge, Zeolites, or CarbonMedia with the MMT-TNDewatering System at Diablo Canyon, Rev. 0.

Scientific Ecology Group, Inc., Procedure HC-001, Procedure for Using theSEG Barrier-Plus High Integrity Container, Rev. 4.

Scientific Ecolog Group, Inc., Procedure HC-002, Procedure for Sealingthe Barrier Plus High Integrity Container for DCPP, Rev. 0.

Scientific Ecology Group, Inc., Procedure DC-P-03-002, Procedure ForFilter Packaging Using the HIC Overpack Assembly at Diablo Canyon,Rev. 0.

TFC Nuclear Associates, Inc., Procedure For Handling and Storage of HighIntegrity Containers, Rev. 4.

TFC Nuclear Associates, Inc., Procedure For Sealing High IntegrityContainers, Rev. 4.

NOTE: The following manual is maintained in Document Control Master File,Catalog No. TK-9400/SEG-2A.

5.2.9 Scientific Ecology Group, Inc., Rad Services Manual for RADLOKHighIntegrity Containers, RSM-014, Rev 0.

5.3 Solidification

NOTE: Copies of the following procedures are maintained in Document ControlMaster File, Catalog No. TK-9400/SEG-2.

5.3.1 Scientific Ecology Group, Inc., Procedure SS-039, Setup/Removal ProcedureFor SEG Radwaste Solidification System No. 8960 for use at Diablo CanyonPower Plant, Rev. D.

5.3.2 Scientific Ecology Group, Inc., Procedure SS-038, Operating Procedure forSEG Radwaste Solidification System No. 8960 for use at Diablo CanyonPower Plant, Rev. E.

5.3.3

5.3.4

Scientific Ecology Group, Inc., Procedure SS-054, Procedure for Using theL-80GE Filter Encapsulation Liner at Diablo Canyon, Rev. 0.

Scientific Ecology Group, Inc., Procedure SS-040, Contamination BibInstallation/Removal, Rev. C.

5.3.5 Scientific Ecology Group, Inc., Procedure SS-041, Procedure for RemoteLid Closure, Rev. C.

NOTE: Copies of the following procedures are maintained in Document ControlMaster File, Catalog No. TK-94001 DTI-2.

5.3.6 Diversified Technologies, Procedure DT-DCPP-09, Waste Transfer, Rev. 9.

5.3.7 Diversified Technologies, Procedure DT-DCPP-11, Filter Loading, Rev. 9.

RP2DC4.DOC 18 3




5.3.8 Diversified Technologies, Procedure DT-DCPP-12, Media Treatment,Rev. 11.

5.3.9 Diversified Technologies, Procedure DT-DCPP-13, VERI Solidification,Rev. 15.

5.3.10 Diversified Technologies, Procedure DT-DCPP-15, VES Solidification,Rev. B.

5.4 Dry Active Waste Processing

NOTE: The following procedure is maintained in Document Control Master File,Catalog No. TK-9400/SEG-2.

Scientific Ecology Group, Inc., Procedure DC-P-28-001, Procedure for using the SEGBarrier Plus ™Container with Foam Lining at Diablo Canyon, Rev. A.

'ECORDS

Records of dewatering and solidification operations shall be submitted to the Records ManagementSystem on a shipment basis by container per RCP RW-4.

7. REFERENCES

7.1 Scientific Ecology Group, Inc., Quality Management System Manual, Rev. 5,Document Control Master File, Catalog No. TK-9400/SEG-3.

7.2 RP2.DC2, "Process Control Program."

7.3 RP2.DC3, "Dewatering Control Program."

7.4 Diversified Technologies, Procedure QIP 20-09, Operator Training Procedure, Rev. 1.

SPONSOR

C. Clint Miller

RP2DC4.DOC lB 4


1998 Land Use Census

DCPP Radiation Protection personnel conducted a land use census in the vicinity of

DCPP for 1998. The land use census is based on Nuclear Regulatory Commission

(NRC), Regulatory Guide 4.8, "Environmental Technical Specifications for Nuclear

Power Plants", and required by DCPP Program Directive CY2, "Radiological

Monitoring and Controls Program." The census is conducted at least once per year

during the growing season for the Diablo Canyon environs.

The land use census identifies the nearest milk animal and garden greater

than 50 square meters (500 square feet), producing broadleaf vegetation, in each ofthe landward meteorological sectors within a distance of 8 kilometers (5 miles) of the

plant. DCPP IDAP RP1.ID11 "Environmental Radiological Monitoring Program",

requires that the nearest residence be identified in each of the landward sectors

within a distance of 5miles.'he

land use census was performed by directly contacting individual landowners or

tenants, or by aerial surveys. The landowners or tenants were contacted between

July 7 and August 31, 1998. The aerial survey was done on August 5, 1998.

The census identified one household garden in the east sector at 4.5 miles, greaterthan 50 square meters (500 square feet) that produces broadleaf vegetation. No

milk animals were identified within the first 5 miles in any sector. Much of the area

surrounding the plant site is used for cattle grazing. Goats were used for weed

abatement for approximately 7 months within the area surrounding the plant site. Afarm is located on the coastal plateau in the east southeast (ESE) sector, along the

site access road. The farm starts at approximately 3.3 miles and extends to 4.5

miles from the plant, but produces only legumes and cereal grass (grains).

A total of 13 residences were identified within the 5-mile radius of the plant that were

confirmed to be, or appeared to be, occupied during 1998. The nearest residence is

1.2 miles north-west (NW) of the plant. Table 1 summarizes the results of the land

use census and Figure 1 shows the locations of the farm and residences in the

vicinity of DCPP.


Table 1

Land Use Census

Distance in Miles from the Unit 1 Center Line to theNearest MilkAnimal, Residence, Vegetable Garden

2p/'egree(a) Nearest Nearest

Radial Sector MilkAnimal Residencekm (mi)

ResidenceAzimuthDegree

NearestVegetable

Gardenkm (mi)

NW None 1.93 (1.2) 319.5 None

NNW None 2.41 (1.5) 331 None

None None None

NNE

NE

ENE

None

None

None

None

5.3 (3.3)

7.89 (4.9)

7.08 (4.4)

5.95 (3.7)

018.5

036

063.5

097.5

None

None

None

None

ESE

SE

None

None

None

7.24 (4.5)

None

None

098 7.24 (4.5)

5.28(3.3)"'one

Table Notation:

Sectors not shown contain no land beyond the site boundary, other thanislets not used for the purposes indicated in this table.

'" The vegetable garden indicated is the farm along the site access road; however,it does not grow broadleaf vegetation.

To Monterey To Paso Robics

CAYUCOS ~ ATASCADERO

MORRO BAY

~ SANTAMARGARITA

NW

NNW

A

Osp so NE

eV+o ENE

C n.Rd.pre<+

SAN LUIS OBISBO

A

PacificOcean

UNITS 1 AND 2DIABLOCANYON SITE

0 Gardens or Farm

A Residences

5 Mile

ESE

0+

e~+

.g+ SE

@0co

0 1 2 3 4 5

SCALE IN MILES

8~~~0,

GROVER ~reCITY <e

Figure L Units 1 and 2 Diablo Canyon Power Plant-Land Use Census.

970338/landuso/pc19

~rk

)

~ i

tgl

· enclosure pg8 e letter dcl 99-055 diablocanyon power plant annual radioactive effluent release...

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